Abstracts Book Migration 2013 - Loughborough University

Migration 2013
The Brighton Centre, UK
September 8 – 13, 2013
Book of Abstracts

Conference Committees
International Steering Committee:
Th. Fanghänel (Chairman) (Germany)
D.L. Clark (USA)
S.B. Clark (USA)
H. Geckeis (Germany)
H. Nitsche (USA)
P. Toulhoat (France)
International Scientific Committee:
U. Berner (Switzerland) K. Morris (UK)
D. Bosbach (Germany) T. Ohnuki (Japan)
V. Brendler (Germany) T. Payne (Australia)
J. Bruno (Spain) Ch. Poinssot (France)
N.D. Bryan (UK)
D. Reed (USA)
P. de Canniere (Belgium) K. Spahiu (Sweden)
N. Evans (UK) S.L.S. Stipp (Denmark)
A. Felmy (USA) B.S. Tomar (India)
B. Grambow (France) V. Vallet (France)
Y. Ikeda (Japan) L. van Loon (Switzerland)
W.H. Kim (Korea)
Wang Xiangke (China)
Liu Chunli (China)
J.-I. Yun (Korea)
Scientific Secretary:
Th. Rabung (Germany)
N. Evans (UK)
Local Committee:
N. Evans (Loughborough University) N.D. Bryan (Manchester University)
K. Morris (Manchester University) N. Smart (NDA-RWMD)
S. Williams (NDA-RWMD) D. Read (Loughborough University)
A. Bath (GeolSoc) R. Shaw (BGS)
T. Milodowski (BGS) G. Shaw (Nottingham University)
I. Burke (Leeds University) M. Townshend (Golder Associates)
A. Ware (RSC) P. Thompson (AWE)
G. Law (Manchester University) J. Renshaw (Birmingham University)
L. Vivian (AWE)
Conference address:
Dr Nick Evans
Senior Lecturer in Radiochemistry
Department of Chemistry
Loughborough University
Loughborough
LE 11 3TU
2

Supported by:
Migration 2013 is being organised by Loughborough University and the Royal Society of
Chemistry and is supported by:
ANALYTICAL DIVISION
CINCH-II: Cooperation in education and
training In Nuclear Chemistry
3

Migration is also supported by the:
European Commission through Talisman
Karlsruhe Institute for Technology (KIT)
4

Background
The MIGRATION conferences provide an international forum for the timely exchange of scientific
information on chemical processes controlling the migration behaviour of actinides and fission
products in natural aquifer systems. Experimental investigations and predictive modelling of these
processes are the main topics of the conferences. The information generated from the MIGRATION
conferences is the basis for the mechanistic understanding of the migration behaviour of long-lived
radionuclides in the geosphere, which is essential for the long-term performance assessment of
nuclear waste disposal.
The first MIGRATION conference was held in 1987 in Munich, Germany. It was followed by
MIGRATION ‘89 in Monterey, California, USA; MIGRATION ‘91 in Jerez de la Frontera, Spain;
MIGRATION ‘93 in Charleston, South Carolina, USA; MIGRATION ‘95 in Saint-Malo, France;
MIGRATION ‘97 in Sendai, Japan, MIGRATION ‘99 at Lake Tahoe, Nevada, USA, MIGRATION
‘01 in Bregenz, Austria, MIGRATION ‘03 in Gyeongju, Korea, MIGRATION ‘05 in Avignon,
France, MIGRATION ‘07 in Munich, MIGRATION ’09 in Kennewick, Washington, USA and
Migration 2011 in Beijing, China
Conference papers that were peer-reviewed and accepted for publication have appeared in special
issues of Radiochimica Acta 44/45 (1988), 52/53 (1991), 58/59 (1992), 66/67 (1994), 74 (1996), 82
(1998), 88 (no. 9/10) (2000), 90 (2002), 92 (2004), 94 (2006), 96 (2008) in the Journal of
Contaminant Hydrology 13 (no. 1/4) (1993), 21 (no. 1/4) (1996), 26 (no. 1/4) (1997), 35 (no. 1/3)
(1998), 47 (no. 2/4) (2001), 61 (no. 1/4) (2003) and in Physics and Chemistry of the Earth 31
(2006), 33 (2008). Proceedings of the last conference are printed in the journals Radiochimica Acta
98 (2010) and Physics and Chemistry of the Earth 35 (2010).

Scope
The MIGRATION conferences focus on recent developments in the fundamental chemistry of
actinides, fission and activation products in natural aquifer systems, their interactions and migration
in the geosphere, and the processes involved in modelling their geochemical behaviour.
The sessions in MIGRATION’11 cover the following areas:
A Aquatic chemistry of actinides and fission products
1) Solubility and dissolution
2) Solid solution and secondary phase formation
3) Complexation with inorganic and organic ligands
4) Redox reactions and radiolysis effects
5) Solid-water interface reactions
6) Colloid formation
7) Experimental methods
8) Computational chemistry
B Migration behaviour of radionuclides
1) Sorption/desorption phenomena in dynamic systems
2) Diffusion and other migration processes
3) Colloid migration
4) Effects of biological and organic materials
5) Field and large scale experiments
6) Natural analogues
C Geochemical and transport modelling
1) Data selection and evaluation
2) Coupling chemistry and transport
3) Development and application of models
4) Model validation
5) Safety assessment and repository concepts
UK Special Session
Environmental Behaviour of Radionuclides after the Fukushima Accident
International Programmes

Programme
SUNDAY (8. September)
12:00 noon REGISTRATION
5:00 pm OPENING SESSION
Chair: Thomas Fanghänel (EU), Nick Evans (UK)
Thomas Fanghänel, Director of the European Commission, Joint Research Centre, Institute for
Transuranium Elements,
Chairman of the Migration International Steering Committee
WELCOME ADDRESS
Ute Blohm-Hieber, Directorate General for Energy, Directorate D – Nuclear Safety & Fuel Cycle,
Head of Unit D2 – Nuclear Energy Technology, Nuclear Waste and Decommissioning
THE EC POLICY ON NUCLEAR WASTE DISPOSAL – THE WASTE DIRECTIVE
Jon Martin, Head of Research, NDA-RWMD
AN INTEGRATED RESEARCH PROGRAMME IN SUPPORT OF THE SAFE DISPOSAL OF
NUCLEAR WASTE IN THE UK

MONDAY (9. SEPTEMBER)
SESSION 1
A1: SOLUBILITY AND DISSOLUTION
Chair:
K. Spahiu (Sweden) and D. Bosbach (Germany)
8:30 am SOLUBILITY, SPECIATION AND THERMO-DYNAMICS OF ACTINIDES
AND FISSION PRODUCTS
M. Altmaier (INVITED) (Germany)
9:15 am EVALUATION OF DISCREPANCIES IN TETRAVALENT OXIDE
SOLUBILITY VALUES BY ISOTOPIC EXCHANGE AND ITS IMPACT ON
THE SAFETY ASSESSMENT
T. Suzuki-Muresan , J. Vandenborre, B. Grambow, A. Valls, L. Duro (France,
Spain)
9:40 am REDOX CHEMISTRY, SOLUBILITY AND HYDROLYSIS OF
TECHNETIUM IN DILUTE TO CONCENTRATED NaCl and MgCl 2
SOLUTIONS
E. Yalcintas, X. Gaona, M. Altmaier, A.C. Scheinost, T. Kobayashi, H. Geckeis
(Germany, Japan)
10:05 am LONG-TERM AQUEOUS ALTERATION KINETICS OF A 99 Tc -DOPED
SON68 BOROSILICATE GLASS
S. Rolland, M. Tribet, M. Magnin, V. Broudic, S. Peuget, A. Janssen, T. Wiss, C.
Jégou, P. Toulhoat (France, EU)
A1-1
A1-2
A1-3
A1-4
10:30 am BREAK
SESSION 2 A5: SOLID-WATER INTERFACE REACTIONS
Chair: T. Payne (Australia) and M. Bradbury (Switzerland)
10:50 am WET CHEMISTRY AND SITE-SELECTIVE LUMINESCENCE
SPECTROSCOPY STUDIES ON THE UPTAKE OF HEXAVALENT
ACTINIDES BY CEMENTITIOUS MATERIALS
J. Tits, T. Stumpf, C. Walther, E. Wieland (Switzerland, Germany
11:15 am THE BIGRAD CONSORTIUM - THE FATE OF TECHNETIUM AND
URANIUM DURING MAGNETITE CRYSTALLISATION AT
HYPERALKALINE pH
T. A. Marshall, K. Morris, G.T.W. Law, J.F.W. Mosselmans, S. Shaw (UK)
11:40 am SPECIATION OF NEPTUNIUM UPTAKE BY OPALINUS CLAY
T. Reich, S. Amayri , J. Drebert , D.R. Fröhlich, D. Grolimund , U. Kaplan
(Germany, Switzerland)
12:05 pm UNDERSTANDING THE MECHANISM OF Eu(III), Np(V), AND U(VI)
SORPTION TO HEMATITE USING VARIABLE TEMPERATURE BATCH
REACTIONS
S.L. Estes, B.A. Powell (USA)
A5-1
A5-2
A5-3
A5-4
12:30 pm BREAK

SESSION 3
A8: COMPUTATIONAL CHEMISTRY STUDIES
Chair:
V. Vallet (France) and D.L. Clark (USA)
2:00 pm COMPUTATIONAL CHEMISTRY FOR HEAVY ELEMENTS:
PRINCIPLES AND APPLICATIONS OF DENSITY FUNCTIONAL
THEORY
N. Kaltsoyannis (INVITED) (UK)
2:45 pm DENSITY FUNCTIONAL MODELING OF URANYL ADSORPTION ON
SOLVATED 2:1 CLAY MINERALS
A. Kremleva, S. Krüger, N. Rösch (Germany)
3:10 pm ION SUBSTITUTION OF STRONTIUM AND NICKEL INTO CALCITE
STUDIED BY DENSITY FUNCTIONAL THEORY
M.P. Andersson, H. Sakuma, S.L.S. Stipp (Denmark, Japan)
3:35 pm RADIUM SOLUBILITY IN THE PRESENCE OF BARITE: SORPTION
EXPERIMENTS AND ATOMISTIC MODELLING
F. Brandt, M. Klinkenberg, V. Vinograd, K. Rozov,
D. Bosbach (Germany)
A8-1
A8-2
A8-3
A8-4
4:00 pm BREAK
SESSION 4
B2: DIFFUSION AND OTHER MIGRATION
PROCESSES
Chair:
P. de Canniere (Belgium) and J. Samper (Spain)
4:20 pm ASSESSING THE SORPTION PROPERTIES OF CANADIAN
SEDIMENTARY ROCKS UNDER SALINE CONDITIONS
P. Vilks, N.H. Miller, T. Yang (Canada)
4:45 pm GAS INDUCED RADIONUCLIDE TRANSPORT IN DISTURBED AND
UNDISTURBED BOOM CLAY
E. Jacops, T. Maes, N. Maes, E. Weetjens, G. Volckaert (Belgium)
5:10 pm DIFFUSION OF 85 Sr 2+ IN COMPACTED MONTMORILLONITE
SEEMINGLY AGAINST ITS OWN CONCENTRATION GRADIENT
M. A. Glaus, J. Eikenberg , S. Frick , M. Rüthi , L.R. Van Loon (Switzerland)
B2-1
B2-2
B2-3

SESSION 5
POSTER SESSION I (7:00 – 10:00 PM)
SPONSORED BY AWE
Chair:
PA1
PA1-1
PA1-2
PA1-3
PA1-4
PA1-5
PA1-6
PA1-7
PA1-8
PA1-9
PA1-10
PA1-11
T. Rabung (Germany) and C. Landesman (France)
SOLUBILITY AND DISSOLUTION
GASEOUS RELEASE OF CARBON-14 FROM IRRADIATED MATERIALS
G.M.N. Baston, N.A. Hodge, S.J. Williams (UK)
ELICITATION OF DISSOLUTION RATE DATA FOR POTENTIAL WASTEFORM
TYPES FOR PLUTONIUM UNDER REPOSITORY CONDITIONS
G. Deissmann, S. Neumeier, F. Brandt, G. Modolo, D. Bosbach (Germany)
SOLUBILITY AND TRLFS STUDY OF Nd(III) AND Cm(III) IN DILUTE TO
CONCENTRATED ALKALINE NaCl-NaNO 3 AND MgCl 2 -Mg(NO 3 ) 2 SOLUTIONS
M. Herm, X. Gaona, Th. Rabung, C. Crepin, V. Metz, M. Altmaier, H. Geckeis (Germany,
France)
SOLUBILITY AND HYDROLYSIS OF U(VI) AT 80°C UNDER ACIDIC TO
HYPERALKALINE PH CONDITIONS
X. Gaona, M. Marques, B. Baeyens, M. Altmaier (Germany, Switzerland)
EFFECTS OF SURFACE MORPHOLOGY ON DISSOLUTION OF ThO 2
E. Myllykylä, T. Lavonen, K. Ollila (Finland)
THE EFFECT OF THE CONCENTRATION OF SILICATES ON THE SOLUBILITY OF
Pu(IV) IN SODIUM BICARBONATE/CARBONATE SOLUTIONS
T. Yamaguchi, K. Henmi, Y. Iida, H. Okamoto, T. Tanaka (Japan)
LEACHING TEST OF CeO 2 UNDER GROUNDWATER CONDITIONS
N. Rodriguez, J. Cobos, E. Iglesias, C. Palomo, J. Nieto, J. M. Cobo, L. Serrano, S. Durán, J.
Quiñones (Spain)
INFLUENCE OF PARTICLE SIZE, CARBONATES AND ORGANIC MATTER ON THE
SOLUBILITY OF ThO 2 (cr)
S. Salah, D. Liu, L. Wang (Belgium, China)
ADOPT PELLET LEACHING PROPERTIES, A COMPARISON WITH UO 2 PELLET
K. Nilsson, O. Roth, M. Jonsson (Sweden)
Np(V) SOLUBILITY IN DILUTE TO CONCENTRATED MgCl 2 SOLUTIONS
V.G. Petrov , X. Gaona , D. Fellhauer , J. Rothe, K. Dardenne, S.N. Kalmykov, M. Altmaier
(Russia, Germany)
THE DISPOSAL OF SPENT NUCLEAR FUEL: THE EFFECT OF SURFACE DEFECTS
ON DISSOLUTION RATE
C.L. Corkhill, J.W. Bridge, P. Hillel, L.J. Gardner, M.C. Stennett, R. Tapper, N. C. Hyatt
(UK, USA)

PA1-12
PA1-13
PA3
PA3-1
PA3-2
PA3-3
PA3-4
PA3-5
PA3-6
PA3-7
PA3-8
PA3-9
PA3-10
THE EFFECT OF ADVA CAST 551 SUPERPLASTICISER ON RADIONUCLIDE
SOLUBILITY
R. Beard, A.P. Clacher , M.M. Cowper (UK)
ELICITATION OF URANIUM SOLUBILITY TO SUPPORT THE DISPOSAL OF U 3 O 8
T. Beattie, M. Couch, C.P. Jackson (Switzerland, UK)
COMPLEXATION WITH INORGANIC AND ORGANIC
LIGANDS
CHAINS AND LAYERS OF URANYL FORMATES TEMPLATED BY PROTONATED
DIAMINES
Qianqian Zhu, Luhua Wang, Chunli Liu, Zheming Wang (China)
THERMODYNAMIC STUDIES OF COMPLEX FORMATION OF TRIVALENT Nd, Am
BY SMALL ORGANIC LIGANDS WITH MICRO TITRATION CALORIMETRY
M. Acker, M. Müller, S. Taut, A. Barkleit, J. Schott, G. Bernhard (Germany)
TWO SYSTEMS OF [DabcoH 2 ] 2+ /[PipH 2 ] 2+ -URANYL-OXALATE SHOWING
REVERSIBLE CRYSTAL TO CRYSTAL TRANSFORMATIONS CONTROLLED BY
THE DIAMMONIUM/URANYL/OXALATE RATIOS IN AQUEOUS SOLUTIONS
([DabcoH 2 ] = 1,4-Diazabicyclo-[2.2.2]-OctaneH 2 and [PipH 2 ] = PiperazineH 2 )
L. H. Wang, R. Shang, Z. Zheng, C. L. Liu, Z. M. Wang (China)
THE THERMODYNAMICS OF THE COMPLEXATION OF Cm(III) WITH SMALL
ORGANIC LIGANDS UNDER SALINE CONDITIONS AND INCREASED
TEMPERATURES
A. Skerencak-Frech, D.R. Fröhlich,, P.J. Panak (Germany)
EFFECT OF ISOSACCHARINIC ACID ON THE SORPTION OF EUROPIUM(III) AND
PLUTONIUM(IV) ON CEMENT CSH PHASES
M. García-Gutiérrez , T. Missana , H. Rojo , H. Galán, M. Mingarro, U. Alonso (Spain,
Switzerland)
SOLUBILITY AND NMR STUDIES OF Ca-GLUCONATE AND C-Np(IV)-GLUCONATE
SYSTEMS IN DILUTE TO CONCENTRATED ALKALINE CaCl 2 SOLUTION
X. Gaona, C. Adam, H. Rojo, M. Böttle, P. Kaden, M. Altmaier (Germany)
COMPLEXATION OF Nd(III)/Cm(III) WITH GLUCONATE IN ALKALINE NaCl, AND
CaCl 2 SOLUTIONS: SOLUBILITY AND TRLFS STUDIES
H. Rojo, X. Gaona, Th. Rabung, M. Garcia, T. Missana, M. Altmaier (Germany, Spain)
A STUDY OF TERNARY Ca-UO 2 -CO 3 COMPLEXATION UNDER NEUTRAL TO
WEAKLY ALKALINE CONDITIONS
J.-Y. Lee, J.-I. Yun (Korea)
USE OF NMR TO DETERMINE STRUCTURE OF PLUTONIUM SIDEROPHORE
COMPLEXES: Pu(IV)(H 2 DFOB) and DFOBPu(IV)di-μ-OH-DFOBPu(IV)
M.A. Boggs, M. Zavarin, A.B. Kersting (USA)
THE BIGRAD CONSORTIUM - MONITORING REDOX SPECIATION OF URANIUM
USING OPTICAL FINGERPRINTING
L.S. Natrajan, A.N. Swinburne, S.D. Woodall , S. Randall, K. Smith, N.D. Bryan, K. Morris
(UK)

PA4
PA4-1
PA4-2
PA4-3
PA4-4
PA4-5
PA4-6
PA4-7
PA4-8
PA4-9
PA4-10
PA4-11
PA6
PA6-1
PA6-2
REDOX REACTIONS AND RADIOLYSIS EFFECTS
EXAMINATION OF THE EFFECT OF ALPHA RADIOLYSIS ON PLUTONIUM(V)
SORPTION TO QUARTZ USING MULTIPLE PLUTONIUM ISOTOPES
A.E. Hixon, Y. Arai, B.A. Powell (USA)
INVESTIGATION OF THE REDOX AND COMPLEXATION BEHAVIOUR OF
URANIUM BY ORGANIC ACIDS USING CYCLIC VOLTAMMETRY
M.Q. Chew, N.D.M. Evans, R.J. Mortimer, C. Boxall (UK)
CUPROUS HYDROXIDE AS AN INTERMEDIATE STEP IN OXIDATION OF COPPER
IN AQUEOUS SOLUTIONS
I.L. Soroka, P.A. Korzhavyi, N.V. Tarakina, M. Jonsson (Sweden, Germany)
THIOSULPHATE, KEY INTERMEDIATE DETERMINING ABIOTIC SELENIUM
REDOX TRANSFORMATIONS IN PRESENCE OF PYRITE
E. Breynaert, W. Wangermez, T.N. Parac-Vogt, C.E.A. Kirschhock, A. Maes (Belgium)
THE BIGRAD CONSORTIUM - NEPTUNIUM BIOGEOCHEMICAL INTERACTIONS
WITH THE MANGANESE CYCLE
G.T.W. Law, C.L. Thorpe, P. Bots, S. Shaw, K. Law, T. Marshall, F.R. Livens, J.R. Lloyd,
M.A. Denecke, J. Rothe, K. Dardenne, K. Morris (UK, Germany)
ROLE OF Fe(II) ON ACTINIDE REDOX PROCESSES AT MINERAL SURFACES
Y. Wen, D. Renock, L. Shuller-Nickles (USA)
STRUCTURAL INVESTIGATION OF SOLID SOLUTIONS IN THE SYSTEM USiO 4 –
ThSiO 4
S. Labs, S. Weiss, C. Hennig, H. Curtius, D. Bosbach (Germany)
ON FISSION PRODUCT ALLOY PARTICLES AND THEIR CATALYTIC PROPERTIES
D. Cui, S. Hovmöller, W. Wan, Y. Yun, M. Granfors, L. Jeanett, K. Spahiu (Sweden)
SOLUBILITY OF TcO 2· xH 2 O( S ) IN DILUTE TO CONCENTRATED NaNO 3 SOLUTIONS
T. Kobayashi, T. Sasaki, A. Kitamura (Japan)
SURFACE SCIENCE STUDY OF SPENT FUEL CORROSION PROCESSES USING
THIN FILM MODEL SYSTEMS
T. Gouder (EC)
EFFECT OF NITROUS ACID ON REDUCTION OF Np(VI) IN IRRADIATED
SOLUTIONS OF NITRIC ACID
A. Paulenova, M. Precek, B. Mincher, S. Mezyk (USA, Czech Republic)
COLLOID FORMATION
INTERACTION OF RARE EARTH ELEMENTS AND SUSPENDED MATTERS
CONTAINED IN HORONOBE DEEP GROUNDWATER
A. Kirishima, A. Kuno, H. Amamiya, H. Murakami, Y. Amano, T. Iwatsuki, T. Mizuno, T.
Kubota, T. Sasaki, N. Sato (Japan)
SIZE AND ELEMENTAL COMPOSITION ANALYSES OF GRANITIC
GROUNDWATER BY FLOW-FIELD FLOW FRACTIONATION
T. Hamamoto, T. Saito, S. Tanaka (Japan)

PA6-3
PA6-4
PA6-5
PA6-6
PA6-7
PA6-8
PB2
PB2-1
PB2-2
PB2-3
PB2-4
PB2-5
PB2-6
PB2-7
PB2-8
SYNTHESIS AND CHARACTERIZATION STUDIES OF SONOLYTIC COLLOIDAL
SPECIES OF PLUTONIUM(IV)
V. Morosini, T. Chave, P. Moisy, C. Den Auwer, S. Nikitenko (France)
SONOCHEMICAL REDUCTION OF Pu(IV) IN AQUEOUS NITRIC SOLUTIONS
M. Virot, L. Venault, P. Moisy, V. Morosini, S. I. Nikitenko (France)
MONTMORILLONITE COLLOID SIZE HETEROGENEITY - IMPACT ON
RADIONUCLIDE SORPTION CAPACITIES
K.K. Norrfors, M. Bouby, S. Heck, N. Finck, R. Marsac, T. Schäfer, H. Geckeis, S. Wold
(Germany, Sweden)
MONTMORILLONITE COLLOID SIZE HETEROGENEITY - IMPACT ON STABILITY
IN SUSPENSION
K.K. Norrfors, M. Bouby, S. Heck, N. Finck, R. Marsac, T. Schäfer, H. Geckeis, S. Wold
(Germany, Sweden)
ACTINIDE SOLUBILITY AND SPECIATION IN THE WASTE ISOLATION PILOT
PLANT (WIPP) REPOSITORY
D.T. Reed, M. Borkowski, J.S. Swanson, M.K. Richmann, J.F. Lucchini, K. Simmons, D.
Cleveland (USA)
ACTINIDE COLLOIDS AND NANOPARTICLES: RELEVANCE TO LEGACY WASTE,
CLEAN-UP AND GEOLOGICAL DISPOSAL
J. Rochford, S. Heath (UK)
DIFFUSION AND OTHER MIGRATION PROCESSES
DIFFUSION PROPERTIES IN LOW PERMEABILITY MEDIA
Y. Xiang, D. Loomer, T. Yang, S. Hirschorn, M. Jensen, T. Al (Canada)
IODIDE DIFFUSION THROUGH COMPACTED BENTONITE B75: CONSISTENT
EVALUATION TAKING INTO ACCOUNT DIFFUSE LAYER INHOMOGENEITY
E. Hofmanová, P. Večerník, D. Vopálka (Czech Republic)
POTENTIAL FOR BUOYANT NON-AQUEOUS PHASE LIQUID (NAPL) TO MIGRATE
IN THE FREE PHASE FROM A GDF
S. Watson, S. Benbow, N. Chittenden, A. Lansdell, M.O. Rivett, G. Towler, A. Herbert, G.
Carpenter, S. Norris, S. Williams (UK)
THE PUZZLING RARE EVENTS OF HIGH 238 Pu CONTENT IN THE GROUND LEVEL
ATMOSPHERE
H. Wershofen, R. Kierepko, J.W. Mietelski, R. Anczkiewicz, Z. Holgye, K. Isajenko, J. Kapała,
A. Komosa (Germany, Poland, Czech Republic)
ROLE OF THE CLAY AGGREGATES IN THE CONTAINMENT PROPERTIES OF THE
CALLOVO-OXFORDIAN CLAYSTONES: A CASE STUDY WITH COMPACTED
CRUSHED SAMPLES AND MACRO-FRACTURED SAMPLES
S. Savoye, C. Imbert, A. Fayette, B. Grenut, J.-C. Robinet (France)
COMBINED STUDIES OF RADIONUCLIDE ADSORPTION AND DIFFUSION
R.A. Wogelius, A. van Veelen, B. Zou, G. Law, K. Morris, M.P. Ryan (UK)
DETERMINING STRONTIUM AND CAESIUM MASS TRANSFER PARAMETERS
WITHIN INTACT CRYSTALLINE ROCKS
B. Zou, T. Ohe, G. Grime, R. Wogelius (UK, Japan)
CHARACTERISING GAS MIGRATION THROUGH VARIABLY SATURATED MEDIA:
A NUMERICAL MODEL
D. Huxtable, D. Read, G. Shaw (UK)

PB2-9
PB2-10
PB2-11
PB2-12
PB2-13
PB2-14
PB2-15
PB2-16
PB2-17
PB2-18
PB2-19
PB2-20
PB4
PB4-1
BENCHMARK EXPERIMENTS FOR THE INVESTIGATION OF THE DIFFUSIVE
BEHAVIOUR OF 85 Sr 2+ IN COMPACTED Na-ILLITE
M.A. Glaus, L.R. Van Loon, L. Van Laer, M. Aertsens, C. Bruggeman, J. Govaerts, N. Maes
(Switzerland, Belgium)
MATRIX DIFFUSION MEASUREMENTS ON A DRILL CORE SAMPLE FROM
ONKALO, OLKILUOTO
J. Kuva, M. Voutilainen, P. Kekäläinen, J. Timonen, P. Hölttä, M. Siitari-Kauppi, K.
Hänninen, K. Helariutta, L. Koskinen (Finland)
3D ANALYSIS OF FLUID FLOW IN FISSURED SALT ROCK -WITHDRAWN
M. Wolf, F. Enzmann, J. Kulenkampff, J. Lippmann-Pipke (Germany)
MATRIX DIFFUSION AND SORPTION OF Cs + , Na + , I - AND HTO IN GRANODIORITE:
LABORATORY RESULTS AND THEIR EXTRAPOLATION TO THE IN-SITU
CONDITION
Y. Tachi, T. Ebina, H. Takahashi, K. Nemoto, T. Suyama, A. Martin (Japan, Switzerland)
DIFFUSION AND SORPTION OF Cs + , I - AND HTO IN COMPACTED SODIUM
MONTMORILLONITE AS A FUNCTION OF DRY DENSITY
Y. Tachi, K. Yotsuji (Japan)
BEHAVIOUR OF SELENIUM IN GRANITIC ROCK
J. Ikonen, P. Sardini, M. Voutilainen, K. Hänninen, L. Jokelainen, R. Pehrman, A. Martin, M.
Siitari-Kauppi (Finland, France, Switzerland)
SORPTION AND DIFFUSION OF Zn ONTO Na-ILLITE UNDER A WIDE VARIETY OF
CONDITIONS
L. van Laer, T. Kupcik, C. Bruggeman, N. Maes, T. Schäfer (Belgium, Germany)
INFLUENCE OF A SALINE PLUME (NaNO 3 ) ON RADIONUCLIDE MOBILITY IN THE
CALLOVO-OXFORDIAN CLAY ROCK
V. Blin, P. Arnoux, D. Hainos, J. Radwan (France)
INFLUENCE OF CLAY CONTENT ON HTO AND 36 Cl TRANSPORT PROPERTIES IN
CALLOVO-OXFORDIAN CLAY ROCK : PERCOLATION EXPERIMENTS AND
MODELLING
C. Landesman, S. Ribet , C. Bailly, J.C. Robinet, B. Grambow (France)
POROSITY, DIFFUSIVITY AND HYDRAULIC CONDUCTIVITY IN GRANITIC ROCK
MATRIX: LABORATORY MEASUREMENTS AND NUMERICAL MODELLING
V. Havlová, J. Najser, L. Gvoždík, K. Sosna, P. Večerník, J. Záruba, P. Dobeš (Czech
Republic)
MASS TRANSPORT IN SHALE MATRIX UNDER SALINE CONDITIONS
P. Vilks, N.H. Miller, J.G. Miller, T. Yang (Canada)
THE BIGRAD CONSORTIUM - MIGRATION OF KEY RADIONUCLIDES THROUGH
HOLLINGTON SANDSTONE
O. Preedy, M. Felipe Sotelo, N.D.M. Evans (UK)
EFFECTS OF BIOLOGICAL AND ORGANIC
MATERIALS
ASSESSING THE SUITABILITY OF HYDROXYAPATITE BIOMINERALS FOR THE
REMEDIATION AND IMMOBILISATION OF AQUEOUS RADIONUCLIDES
S. Handley-Sidhu, R.A.D Pattrick, J.M. Charnock, J.R Lead, B. Stolpe, L.E. Macaskie (UK,
USA)

PB4-2
PB4-3
PB4-4
PB4-5
PB4-6
PB4-7
PB4-8
PB4-9
PB4-10
PB4-11
PB4-12
PB6
PB6-1
PB6-2
BACTERIAL DIVERSITY IN MONT TERRI OPALINUS CLAY AND THE INFLUENCE
OF THE BACTERIAL SPOROMUSA SP. ISOLATE ON PLUTONIUM SPECIATION
H. Moll, L. Lütke, V. Bachvarova, A. Geissler, S. Selenska-Pobell, G. Bernhard (Germany)
BIODEGRADATION OF CELLULOSE DEGRADATION PRODUCTS
S. Rout, P. Shaw, A. McCarthy, D. Rooks, P. Loughnane, C. Doulgeris, P. Humphreys, A.
Laws (UK)
BIOSORPTION AND MECHANISM OF URANIUM (VI) ON BACILLUS SP ISOLATED
FROM AERATED ZONE SOIL
Xiaolong Li, Congcong Ding, Jiali Liao, Yuanyou Yang, Dong Zhang, Jijun Yang, Jun Tang,
Ning Liu (China)
THE IMPACT OF ALKALIPHILIC AND ALKALITOLERANT MICROORGANISMS
FROM A HYPER-ALKALINE SPRING ON THE TRANSPORT CHARACTERISTICS OF
SANDSTONE
S. Smith, J. Lloyd, J. West (UK)
THE BIGRAD CONSORTIUM - GEOMICROBIOLOGY OF CEMENTITIOUS NUCLEAR
WASTE
A.J. Williamson, K. Morris, G.T.W. Law, S. Shaw, C. Boothman, J.R. Lloyd (UK)
COMPARISON OF MICROBIOLOGICAL INFLUENCES ON THE TRANSPORT AND
CHEMICAL PROPERTIES OF INTACT SANDSTONE AND ITS RELEVANCE TO
GEODISPOSAL
J. Wragg, K. Bateman, A.E. Milodowski, J.M. West (UK)
MICROBIAL DEGRADATION OF ISA UNDER HIGH PH CONDITIONS
REPRESENTATIVE OF INTERMEDIATE LEVEL WASTE
N.M. Bassil, N. Bryan, J.R. Lloyd (UK)
REDUCTION OF SELENITE BY BIOFILMS OF AN IRON-REDUCING BACTERIUM
Y. Suzuki,* H. Saiki, A. Kitamura, H. Yoshikawa (Japan)
MOLECULAR CHARACTERIZATION OF U(VI) ASSOCIATION WITH U RESISTANT
CLAY BACTERIAL ISOLATE UNDER NEUTRAL CONDITIONS
M. Lopez Fernandez, I. Sanchez Castro , M. Romero Gonzalez, A. Guenther, P.L. Solari, M.L.
Merroun (Spain, UK, Germany)
BIOREMEDIATION OPTIONS FOR REMOVING URANIUM FROM
GROUNDWATER
L. Newsome, K. Morris, D. Trivedi, J. Lloyd (UK)
THE BIGRAD CONSORTIUM - THE HYDROGEN DRIVEN
GEOMICROBIOLOGY OF CEMENTITIOUS NUCLEAR WASTE
M.J.C. Crouch, K. Morris, D. Engelberg, J. Small, J.R. Lloyd (UK)
NATURAL ANALOGUES
RADIUM NATURAL ANALOGUE REQUIREMENTS FOR ROBUST SAFETY CASE
DEVELOPMENT
W.R. Alexander, I.G. McKinley, E.M. Scourse, S.M.L. Hardie, E. Klein (Switzerland, UK)
NATURAL ANALOGUES FOR RADIONUCLIDES
J.S. Zakharova, A.D. Wheatley, O. Voitsekhovich (UK, Ukraine)

PC1
PC1-1
PC1-2
PC1-3
PC1-4
PC3
PC3-1
PC3-2
PC3-3
PC3-4
PC3-5
PC3-6
PC3-7
PC4
PC4-1
PC4-2
DATA SELECTION AND EVALUATION
SORPTION VALUES IN SORPTION DATA BASES FOR ARGILLACEOUS ROCKS
AND BENTONITE: A COMPARISON BETWEEN DERIVED AND MEASURED
VALUES
B. Baeyens, M. Marques Fernandes, M.H. Bradbury (Switzerland)
RADIONUCLIDE SOURCE TERM ESTIMATIONS FOR THE PRELIMINARY SAFETY
ASSESSMENT GORLEBEN (VSG)
M. Altmaier, B. Kienzler, Ch. Bube, V. Metz, H. Geckeis (Germany)
AN ATEMPT TO SELECT THERMODYNAMIC DATA AND TO EVALUATE THE
SOLUBILITY OF RADIOELEMENT WITH UNCERTAINTY UNDER HLW DISPOSAL
CONDITIONS
T. Yamaguchi, S. Takeda, Y. Nishimura, Y. Iida, T. Tanaka (Japan)
THEREDA REVISITED - WHAT HAPPENED SO FAR
V. Brendler, F. Bok, M. Altmaier (Germany)
DEVELOPMENT AND APPLICATION OF MODELS
DISCUSSION OF SOME UNCERTAINTIES IN EVALUATION OF DIFFUSION
EXPERIMENTS WITH COMPACTED BENTONITE
D. Vopálka, A. Vetešník, E. Hofmanová (Czech Republic)
NUCLIDE TRANSPORT OF N-MEMBER DECAY CHAIN IN FRACTURED ROCKS
WITH STAGNANT WATER ZONE- MODEL DEVELOPMENT AND SIMULATIONS
P. Shahkarami, L. Liu, L. Moreno, I. Neretnieks (Sweden)
DISCRETE FRAGMENT MODEL FOR APPARENT FORMATION CONSTANTS OF
METAL COMPLEXES WITH HUMIC SUBSTANCES
T. Sasaki, H. Yoshida, S. Aoyama, T. Kobayashi, I. Takagi, H. Moriyama (Japan)
ANALYSIS OF SENSITIVITY OF MODEL OF TRANSPORTING SUBSTANCES FROM
RADIOACTICE WASTE REPOSITORY
J. Chudoba (Czech Republic)
REACTIVE TRANSPORT MODELING OF 129 I MIGRATION IN GRANITIC ROCKS
E. Orucoglu, B. van den Akker, J. Ahn (USA)
PHREEQC TRANSPORT MODELLING OF CS-137 TO ASSESS THE IMPACT OF A
NEW LEAK TO GROUND INTERACTING WITH EXISTING CONTAMINATION
L. Abrahamsen, J. Graham (UK)
CHARACTERISATION OF GROUNDWATER PATHWAYS AND MODELLING
SOLUTE TRANSPORT THROUGH THE BEDROCK AT OLKILUOTO, FINLAND
L. Hartley, J. Hoek , S. Baxter , L. Koskinen UK, Finland)
MODEL VALIDATION
DIFFUSION AND FLOW IN THIN SLITS WITH VARYING APERTURE:
EXPERIMENTAL STUDIES ON VIZUALIZATION AND EVALUATION OF THE Q-
EQUIVALENT TRANSPORT MODEL
H. Winberg , I. Neretnieks, L. Moreno, L. Liu (Sweden)
BENCHMARKING EXERCISE OF REACTIVE TRANSPORT CODES USING THE
THERMOCHIMIE DATABASE
M. Grivé, A. Idiart, D. García, E. Colàs (Spain)

PIP
PIP-1
PIP-2
PIP-3
PIP-4
PIP-5
PIP-6
PIP-7
INTERNATIONAL PROGRAMMES
EMSL RADIOCHEMISTRY ANNEX: A NEW INTERNATIONAL USER-FACILITY FOR
THE STUDY OF RADIOLOGICAL SAMPLES
N.J. Hess, A.A. Campbell (USA)
FORGE - FATE OF REPOSITORY GASES
R. Shaw (UK)
CROCK: CRYSTALLINE ROCK RETENTION PROCESSES - A 7TH FRAMEWORK
PROGRAMME COLLABORATIVE PROJECT (2011-2013)
T. Rabung, D. Garcia, J. Molinero (Germany, Spain)
FP7 COLLABORATIVE PROJECT FIRST-NUCLIDES
B. Kienzler, V. Metz, L. Duro, A. Valls (Germany, Spain)
THE EUROPEAN NUCLEAR ENERGY FORUM (ENEF) GUIDELINES FOR
ESTABLISHING AND NOTIFYING NATIONAL PROGRAMMES UNDER THE
EURATOM WASTE DIRECTIVE
G. Buckau (EC)
TALISMAN - A LARGE INTERNATIONAL EC FP7 EURATOM FRAMEWORK
PROJECT
Bourg, S., Altmaier, M., Bryan, N., Collings, P., Dacheux, N 5 , Duplantier, B 6 , Ekberg, Ch.,
Grolimund, D., Natrajan, L., Poinssot, Ch., Raison, Ph., Schaefer, Th 2 , Scheinost, A.,
Schimmelpfennig, B.
CINCH–II PROJECT – NEXT STEP IN THE COORDINATION OF EDUCATION IN
NUCLEAR- AND RADIOCHEMISTRY IN EUROPE
Jan John, Václav Čuba, Mojmír Němec, Teodora Retegan, Christian Ekberg,
Gunnar Skarnemark, Jukka Lehto, Teija Koivula, Paul J. Scully, Clemens Walther,
Jan-Willem Vahlbruch, Nick Evans, David Read, Eric Ansoborlo, Bruce Hanson,
Lindis Skipperud, Brit Salbu, Jon Petter Omtvedt

TUESDAY (10. SEPTEMBER)
SESSION 6
C2: COUPLING CHEMISTRY AND TRANSPORT
Chair:
L. van Loon (Switzerland) and O. Kolditz (Germany)
8:30 am MODELING REACTIVE TRANSPORT PROCESSES IN POROUS MEDIA
WITH OPENGEOSYS
H. Shao, O. Kolditz (INVITED) (Germany)
9:15 am LONG-TERM REACTIVE TRANSPORT SIMULATION OF THE
INTERACTIONS OF CORROSION PRODUCTS AND COMPACTED
BENTONITE IN A HLW REPOSITORY AND SENSITIVITY ANALYSES
TO KEY PARAMETERS
J. Samper, A. Naves, L. Montenegro, A. Mon, B. Pisani (Spain)
9:40 am NUMERICAL ANALYSIS OF MULTI-MINERALS TRANSFER
BETWEEN THE HOST ROCK AND GROUNDWATER AGAINST
EXPERIMENT FOR NUCLEAR WASTE DISPOSAL
XiaoHui Chen, S. Thornton, J. Small, E. Moyce, S. Shaw, T. Milodowski, C.
Rochelle (UK)
10:05 am HIGH-PERFORMANCE REACTIVE TRANSPORT MODELLING OF
RADIONUCLIDE MIGRATION. THE COMSOL-PHREEQC APPROACH
J. Molinero, A. Nardi, D. García, C. Domènech, M. Grivé, B. Cochepin
(Spain, France)
C2-1
C2-2
C2-3
C2-4
10:30 am BREAK
SESSION 7
Chair:
B5: FIELD AND LARGE SCALE EXPERIMENTS
G. Bracke (Germany) and K. Morris (UK)
10:50 am TEASING OUT BIOGEOCHEMICAL CONTROLS ON THE TRANSPORT
OF PU: EXAMPLES FROM FIELD AND LABORATORY STUDIES
A.B Kersting, M. Zavarin, B.A. Powell , P. Zhao, J. Begg, Z. Dai, M. Boggs
(USA)
11:15 am LONG TERM DIFFUSION EXPERIMENT LTD PHASE I: EVALUATION
OF RESULTS AND MODELLING
V. Havlová, A. Martin, J. Landa, F. Sus, M. Siitari-Kauppi, J. Eikenberg, P.
Soler, J. Miksova (Czech Republic, Switzerland, Finland, Spain))
11:40 am TRACERS BEHAVIOR INTO THE GROUNDWATER OF A FRENCH
NUCLEAR WASTE DISPOSAL
O. Péron, S. Razafindratsima, A. Piscitelli, C. Gégout, V. Schneider, F.
Barbecot, E. Giffaut, J-C. Robinet, G. Montavon (France, Canada)
12:05 pm DENITRIFICATION PROCESSES IN THE MONT TERRI IN SITU
BITUMEN NITRATE CLAY INTERACTION EXPERIMENT AND
MODELLING EFFECTS ON Eh AND RADIONUCLIDE MIGRATION
J. Small, L. Abrahamsen, A. Albrecht, N. Bleyen, E. Valcke (UK, France,
Belgium)
B5-1
B5-2
B5-3
B5-4
12:30 pm BREAK

SESSION 8
C5: SAFETY ASSESSMENT AND REPOSITORY
CONCEPTS
Chair:
C. Liu (China) and Steve Williams (UK)
2:00 pm PRELIMINARY SAFETY ASSESSMENT OF THE GORLEBEN SITE
G. Bracke K. Fischer-Appelt (INVITED) (Germany)
C5-1
2:35 pm COMMON CHALLENGES FOR THE UK AND JAPANESE GEOLOGICAL
DISPOSAL PROGRAMMES
E.M. Scourse, T. Beattie, S.M.L. Hardie, I.G. McKinley (Switzerand, UK)
3:00 pm CHALLENGING THE FUNCTION OF ENGINEERED BARRIERS IN THE
REPOSITORY FOR STORAGE OF SPENT NUCLEAR FUEL
W. Forsling (Sweden)
C5-2
C5-3
3:25 pm BREAK
SESSION 9
A3: COMPLEXATION WITH INORGANIC
AND ORGANIC LIGANDS
Chair:
P. Panak (Germany) and D. Reed (USA)
3:45 pm INVESTIGATION OF THE SYSTEM Ln(III)/An(III)-B(OH) 3 -ORGANICS
J. Schott, M. Acker, J. Kretzschmar, A. Barkleit, S. Taut, V. Brendler, G.
Bernhard (Germany)
4:10 pm INFLUENCE OF THIOL COMPOUNDS ON STABILIZING BULK U(IV)
SPECIES AND SPECTROSCOPIC DETERMINATION OF U(VI) AND U(IV)
UNDER THIOL-RICH CONDITIONS
W. Cha, E. C. Jung, H-R. Cho, S. Y. Lee , M. H. Baik (Korea)
4:35 pm COMPLEXION OF Nd(III)/Cm(III) WITH BORATE IN DILUTE TO
CONCENTRATED ALKALINE NaCl, MgCl 2 AND CaCl 2 SOLUTIONS:
SOLUBILITY AND TRLFS STUDIES
K. Hinz, M. Altmaier, X. Gaona, T. Rabung, D. Schild, C. Adam, H. Geckeis
(Germany)
A3-1
A3-2
A3-3
5:00 pm BREAK

SESSION 10
POSTER SESSION II (7:00 – 10:00 PM)
SPONSORED BY AWE
Chair: H. Geckeis (Germany) and M. Glaus (Switzerland)
PA2
SOLID SOLUTION AND SECONDARY PHASE
FORMATION
PA2-1
PA2-2
PA2-3
PA2-4
PA2-5
RECRYSTALLIZATION OF BARITE IN THE PRESENCE OF RADIUM - A
MICROSCOPIC AND SPECTROSCOPIC STUDY
M. Klinkenberg, F. Brandt, U. Breuer, D. Bosbach (Germany)
ISOTOPIC EXCHANGE OF 45 Ca AND 14 C ON CALCITE
J. Lempinen, S. Kallio, M. Hakanen, J. Lehto (Finland)
ACTINIDE BORATES FORMATION AT NORMAL AND EXTREME
CONDITIONS
E.V. Alekseev, S. Wu, S. Wang, W. Depmeier, T.E. Albrecht-Schmitt (Germany, China,
USA)
MECHANISM OF ENHANCED INCORPORATION OF STRONTIUM IN CALCITE
VIA AMORPHOUS CALCIUM CARBONATE PRECIPITATION
J. Littlewood, I.T. Burke, S. Shaw, C.L. Peacock, D. Trivedi (UK)
THERMODYNAMIC AND STRUCTURAL DATA FOR THE RADIUM AND
BARIUM SULFATE SYSTEM
N. Torapava, H. Hedström, C. Ekberg (Sweden)
PA5
PA5-1
PA5-2
PA5-3
PA5-4
PA5-5
PA5-6
SOLID-WATER INTERFACE REACTIONS
60 CO SORPTION ON LAYERED DOUBLE HYDROXIDES FROM AQUEOUS
SOLUTIONS
N. A. Konovalova, S. A. Kulyukhin, E. P. Krasavina, I. A. Rumer (Russia)
APPROACHES TO SURFACE COMPLEXATION MODELING OF Ni(II) ON
CALLOVO-OXFORDIAN CLAY STONE
Z. Chen, Z. Guo, X. Wang, S. Razafindratsima, J.C. Robinet, G. Montavon, C. Landesman
(France, China)
EFFECT OF B-SITE VACANCY ON CESIUM ADSORPTION TO LAYERED
PEROVSKITE KCa 2 Nb 3 O 10
Zhu-Ling Jiang, TsingHai Wang, Chu-Fang Wang (Taiwan)
ENTHALPY MEASUREMENT OF THE PROTONATION OF GAMMA-ALUMINA BY
CALORIMETRIC TITRATION
Y. Saito, A. Kirishima, N. Sato (Japan)
RADIOTRACER EXCHANGE STUDIES ON THE REVERSIBILITY OF INTERACTION
PROCESSES RELATED TO HUMIC-BOUND METAL TRANSPORT
H. Lippold, J. Lippmann-Pipke (Germany)
REDISTRIBUTION OF FORMS OF Cs 137 , Sr 90 AND Co 60 IN PROCESS OF LONG-
TIME INTERACTION WITH BACKFILL MATERIALS
E.E. Ostashkina, O.A. Burlaka, Z.I. Golubeva, G.A. Varlakova (Russia)

PA5-7
PA5-8
PA5-9
PA5-10
PA5-11
PA5-12
PA5-13
PA5-14
PA5-15
PA5-16
PA5-17
PA5-18
PA5-19
PA5-20
PA5-21
REVERSIBILITY IN RADIONUCLIDE/BENTONITE TERNARY SYSTEMS
N. Sherriff, R. Issa, P. Ivanov, T. Griffiths, N. Grieg, G. Bozhikov, S. Dmitriev, K. Morris,
N.D. Bryan (UK, Russia)
SORPTION BEHAVIOR OF HYDROGEN SELENIDE (HSe-) ONTO IRON-
CONTAINING MINERALS
Y. Iida, T. Yamaguchi, T. Tanaka (Japan)
SORPTION OF 99 TC TO CEMENTITIOUS MATERIALS UNDER REDUCING AND
OXIDIZING CONDITIONS
S.L. Estes, D.I. Kaplan, B.A. Powell (USA)
SORPTION OF CESIUM ON PEAT
M. Lusa, S. Virtanen, J. Lempinen, A.T.K. Ikonen, A.-M. Lahdenperä, J. Lehto (Finland)
SORPTION OF THORIUM FROM AQUEOUS SOLUTIONS USING GRAPHENE OXIDE
ANG IRRADIATED GRAPHENE OXIDE
Li Yan, Zhong Zheng, Chunli Wang, Chunli Liu, Wangsuo Wu (China)
SORPTION OF U(VI) ON LAYERED DOUBLE HYDROXIDES OF Mg, Al, AND Nd
FROM COMPLEX CHEMICAL COMPOSITION SOLUTIONS
N. A. Konovalova, S. A. Kulyukhin, E. P. Krasavina (Russia)
SYNTHESIS AND CHARACTERIZATION OF Ba 2 T iSi 2 O 8 FOR SORPTION STUDIES
A.S. Kar, M. Sahu, M. Keskar, B.S. Tomar (India)
URANYL IONS SORPTION TO TiO 2 AND INTERACTIONS WITH FA SORPTION:
EXPERIMENTS AND MODELING
Y. Ye, X. Wang, N. Guo, Z. Guo, W. Wu (China)
MODELING THE ADSORPTION OF Eu(III) AND Am(III) ON GRANITE USING A
GENERALIZED COMPOSITE APPROACH
Z. Guo, Z. Chen, Q. Jin, W. Wu (China)
THE BIGRAD CONSORTIUM - THE INFLUENCE OF IRON OXIDE
CRYSTALLIZATION ON THE MOBILITY OF NEPTUNIUM
P. Bots, S. Shaw, G.T.W. Law, T. Marshall, J.F.W. Mosselmans, F.R. Livens, M.A. Denecke,
J. Rothe, K. Dardenne, K. Morris (UK, Germany)
MECHANISM OF STRONTIUM SORPTION AND INCORPORATION DURING THE
ALKALINE ALTERATION OF SEDIMENTS BY KOH DOMINATED CEMENT
LEACHATE
S.H. Wallace, S. Shaw, K. Morris, J.S. Small, I.T. Burke (UK)
RETENTION OF SELENIUM ON GRANITE: ANALYSIS OF GRANITIC SURFACE
FOR THE PURPOSE OF DETERMINATION OF SELENIUM SPECIES
K. Videnská, V. Havlová, M. Vašinová Galiová, V. Kanický, P. Sajdl (Czech Republic)
SELENIUM FORMS ON ÄSPÖ DIORITE: ANALYSIS OF ROCK SURFACE USING
LA-ICP MS AND XPS
K. Videnská, V. Havlová, M. Vašinová Galiová, V. Kanický, P. Sajdl (Czech Republic)
A NANO TO MACRO SCALE INVESTIGATION OF MULTI-SITE CAESIUM
SORPTION TO ILLITE AND SEDIMENT
A.J. Fuller, S. Shaw, C.L. Peacock, D. Trivedi. I.T. Burke (UK)
Tc(VII) IMMOBILIZATION ON GRANITOID ROCKS FROM ÄSPÖ (SWEDEN)
Y. Totskiy, F. Huber, T. Schäfer, H. Geckeis (Germany)

PA5-22
PA5-23
PA5-24
PA5-25
PA5-26
PA5-27
PA5-28
PA5-29
PA5-30
PA5-31
PA5-32
PA5-33
PA5-34
PA5-35
PA5-36
INVESTIGATION OF ACTINIDE AND LANTHANIDE SORPTION ON CLAY
MINERALS UNDER SALINE CONDITIONS
A. Schnurr, R. Marsac, T. Rabung, J. Lützenkirchen, H. Geckeis (Germany)
SORPTION OF Np(V) ONTO Na -BENTONITE AND GRANITE: EFFECT OF
EQUILIBRIUM TIME, pH, IONIC STRENGTH AND TEMPERATURE
P. Li, Z. Liu, Z. Guo, W. Wu (China)
COMPARISON OF URANYL ADSORPTION ON IRON(III) OXYHYDROXIDES
K. Niida, T. Saito, S. Tanaka (Japan)
URANYL COORDINATION CHEMISTRY ON Mg-RICH MINERALS: POLARISATION
DEPENDENT EXAFS
A. van Veelen, R. Copping, G. La 3 , A.J. Smith, J.R. Bargar, D.K. Shuh, R.A. Wogelius
EFFECT OF AGING ON THE REVERSIBILITY OF Pu(IV) SORPTION TO GOETHITE
J.C. Wong, M. Zavarin, J.D.C. Begg, A.B. Kersting, B.A. Powell (USA)
INVESTIGATIONS OF THE SORPTION OF U(VI) ONTO SiO 2 IN THE PRESENCE OF
PHOSPHATE: IN SEARCH OF A TERNARY SURFACE COMPLEX
M.J. Comarmond, H. Foerstendorf, R. Steudtner, E. Chong, K. Heim, K. Müller, K. Gückel,
V. Brendler, T.E. Payne (Australia, Germany)
U(VI) SURFACE DISTRIBUTION ON ÄSPÖ DIORITE UNDER ANOXIC CONDITIONS
U. Alonso, T. Missana, A. Patelli, D. Ceccato, M. García-Gutiérrez, V. Rigato (Spain, Italy)
IMMOBILISATION OF TECHNETIUM-99 ON BACKFILL CEMENT: SORPTION
UNDER STATIC AND SATURATED FLOW CONDITIONS
C.L. Corkhill, J.W. Bridge, P. Hillel, L.J. Gardner, M.C. Stennett, R. Tappero, N.C. Hyatt
(UK)
SPECIATION OF PLUTONIUM DURING DIFFUSION IN OPALINUS CLAY
S. Amayri, U. Kaplan, J. Drebert, J. Rosemann, D. Grolimund, T. Reich (Germany,
Switzerland)
INFLUENCE OF SOIL PROPERTIES AND PH CHANGES IN AMERICIUM SORPTION-
DESORPTION ON SOILS
O. Ramírez-Guinart, M. Vidal, A. Rigol (Spain)
THE INFLUENCE OF DIFFERENT MINERAL SURFACE PROPERTIES AND THE
PRESENCE OF NICKEL(II) ON EUROPIUM(III) RETENTION AT VARIOUS OXIDE
MINERALS
S. Virtanen, J. Knuutinen, N. Huittinen, T. Rabung, H. Geckeis, J. Lehto (Finland, Germany)
STUDY OF THE Cs SORPTION IN KAOLINITE AND APPLICATION TO Cs
SORPTION MODELLING IN MIXED CLAY SYSTEMS
A. Benedicto, T. Missana, M. Garcia-Gutierrez (Spain)
CHARACTERIZING URANIUM AND THORIUM IN SOILS: COMPLEMENTARY
INSIGHTS FROM ISOTOPIC EXCHANGE AND SINGLE EXTRACTIONS
H. Ahmed, S. Young, G. Shaw (UK)
PREDICTIONS OF NpO 2 + IONIC EXCHANGE ON MONTMORILLONITE IN
NATURAL WATERS
A. Benedicto, J. D. Begg, P. Zhao, A. B. Kersting, M. Zavarin (USA, Spain)
EXAFS INVESTIGATION ON Eu(III)-SILICA-HUMIC ACID SORPTION SYSTEM:
EFFECT OF ADDITION ORDER
S. Kumar, S. Kasar, A.S. Kar, B.S. Tomar (India)

PA5-37
PA5-38
PA5-39
PA5-40
PA5-41
PA5-42
PA5-43
PA5-44
PA5-45
PA5-46
PA5-47
PA5-48
PA5-49
STUDIES ON URANIUM(VI) SORPTION ONTO MONTMORILLONITE IN HIGHLY
CONCENTRATED BACK-GROUND ELECTROLYTES
K Fritsch, K Schmeide, G Bernhard (Germany)
STUDY OF EUROPIUM AND NICKEL INTERACTION WITH CALCITE - BATCH
EXPERIMENTS AND SPECTROSCOPIC CHARACTERIZATION
A. Sabau , N. Jordan , C. Lomenech , N. Marmier, V. Brendler, A. Barkleit , S. Surblé , N.
Toulhoat , Y. Pipon , N. Moncoffr , E. Giffaut (France, Germany)
SELECTIVE REMOVAL OF METALS FROM AQUEOUS SOLUTIONS USING SILICA
ATTACHED LIGANDS
J. Holt, S. Christie, N. Evans, S. Edmondson (UK)
EFFECTS OF CEMENT SUPERPLASTICIZERS ON Eu SORPTION ONTO KIVETTY
GRANITE
S. Holgersson (Sweden)
URANIUM (VI) SORPTION ONTO ROCK SAMPLES FROM AREAS OF THE
PROPOSED HLW AND SNF REPOSITORY IN RUSSIA (NIZHNEKANSKY MASSIVE)
N.V. Kuzmenkova, V.G. Petrov, I.E. Vlasova, V.A. Petrov, V.V. Poluektov, S.N. Kalmykov
(Russia)
SORPTION OF SELENIUM OXYANIONS ONTO HEMATITE
N. Jordan, S. Domaschke, H. Foerstendorf, A.C. Scheinost, S. Weiß, K. Heim (Germany)
EFFECT OF NATURAL ORGANIC LIGANDS ON PLUTONIUM SORPTION TO
MONTMORILLONITE: OBSERVATIONS ON DIFFERENCES DUE TO ORDER OF
ADDITION
M.A. Boggs, M. Zavarin, B.A. Powell, A.B. Kersting (USA)
SORPTION AND SPECIATION OF MOLYBDENUM ON BOREAL FOREST SOIL
SAMPLES
M. Söderlund, J. Lehto (Finland)
Sr-85 AND Eu-152 SORPTION ON MX-80 BENTONITE COLLOIDS
P. Hölttä, O. Elo, S. Jortikka, S. Niemiaho, M. Lahtinen, J. Lehto (Finland)
Eu AND Cm SORPTION ONTO UNPURIFIED ILLITE: BATCH-TYPE EXPERIMENTS
AND TIME RESOLVED LASER FLUORESCENCE SPECTROSCOPY (TRLFS)
T. Kupcik, R. Marsac, M. Hedde, T. Rabung, T. Schäfer, M. Marques Fernandes, B. Baeyens,
H. Geckeis (Germany, Switzerland)
INTERACTION OF Ni(II) WITH CLAY MINERALS STUDIED BY MACROSCOPIC
AND MICROSCOPIC APPROACH
Shitong Yang, Guodong Sheng, Xiaoli Tan, Jun Hu, Xiangke Wang (China)
SURFACE MODIFIED MINERALS FOR RADIONUCLIDE SEQUESTRATION
H. Gillings, S.E. Dann, D. Read (UK)
RADIONUCLIDE INTERACTIONS WITH THE Fe 1-x O SURFACE UNDER GDF
CONDITIONS
O. Preedy, A van Veelen, M.P. Ryan, R.A. Wogelius, K. Morris, G.T.W. Law, N.A. Burton, F.
Mosselmans, N.D.M. Evans (UK)

PA7
PA7-1
PA7-2
PA7-3
PA7-4
PA7-5
PA7-6
PA7-7
PA7-8
PA7-9
PA7-10
EXPERIMENTAL METHODS
DEVELOPMENT OF DIFFUSIVE GRADIENTS IN THIN-FILMS TECHNIQUE FOR
THE DETERMINATION OF TECHNETIUM-99 IN FRESHWATERS
J.J. Surman, J.M. Pates, H. Zhang (UK)
THE ENHANCEMENT OF ANALYTICAL METHOD FOR IODINE-129
DETERMINATION IN LOW-LEVEL RADIOACTIVE WASTE
Yi-Kong Hsieh, TsingHai Wang, Li-Wei Jian, Wei-Han Chen, Tsuey-Lin Tsai, Chu-Fang
Wang (Taiwan)
SELECTIVE CHEMILUMINESCENCE SPECROSCOPY OF ACTINIDES AND
LANTHANIDES IN SOLUTIONS
I.N. Izosimov, N.G. Firsin, N.G. Gorshkov, V.A. Mikhalev, S.N. Nekhoroshkov (Russia)
SPECTROPHOTOMETRIC DETERMINATION OF MICROAMOUNTS OF THORIUM
WITH THORIN IN THE PRESENCE OF CETYLPRIDINIUM CHLORIDE AS
SURFACTANT IN PERCHLORIC ACID
Muhammad Haleem Khan, Syed Manzoor Hussain Bukhari, Akbar Ali (Pakistan)
APPLICATION OF ACCELERATOR MASS SPECTROMETRY TO MIGRATION
STUDIES OF ACTINIDES AT A LEGACY WASTE DISPOSAL SITE
T.E. Payne, J.J. Harrison, D.P. Child, K.L. Wilsher, S. Thiruvoth, M.A.C. Hotchkis, A.
Ikeda Ohno, M. P. Johansen (Australia)
ICP-MS MEASUREMENT OF SAMPLES WITH HIGH SALINITY - SAMPLE CLEAN
UP OR TRANSIENT MEASUREMENT
C. Hein, J.M. Sander, R. Kautenburger, H.P. Beck, G. Kickelbick (Germany)
BATCH EXPERIMENTS MADE EASY - AUTOMATION OF ICP-MS BATCH
EXPERIMENTS
J.M. Sander , C. Hein , R. Kautenburger , H.P. Beck , G. Kickelbick (Germany)
ELECTROCHEMICAL ASSESSMENT OF REDOX PROPERTIES OF
ARGILLACEOUS SEDIMENTS USING ORGANIC ELECTRON TRANSFER
MEDIATORS
A.L. Hoving, T. Behrends, M. Sander, N. Maes, C. Bruggeman (Netherlands, Switzerland,
Belgium)
THE USE OF DIAMOND LIGHT SOURCE TO STUDY RADIONUCLIDES BY X-RAY
ABSORPTION SPECTROSCOPY
F. Mosselmans, G. Law, K. Morris, S. Shaw, S. Parry (UK)
SPECIATION OF 79 SE AT ULTRATRACE LEVELS USING AN AUTOMATED
CHROMATOGRAPHIC SYSTEM COUPLED TO HIGH RESOLUTION ICPMS.
APPLICATION TO NUCLEAR SPENT FUEL CORROSION STUDIES
L. Aldave de las Heras, M. Sandow, D. Serrano-Purroy, R. Sureda Pastor, S. Van Winckel,
J.P. Glatz (EU, Spain)

PA7-11
PA7-12
PA7-13
THE COMMISSIONING AND VALIDATION OF LOW-FLOW SAMPLING
TECHNIQUE
K. Farrow, R. Stoate, L. Vivian, L. Penrose, R. Hackett, J. Rice (UK)
DEVELOPING A ROBUST ANALYTICAL METHOD FOR RADIOANALYTICAL
SEPARATION OF AMERICIUM-241
K. Farrow, G. Murphy, T. Cartledge (UK)
ANALYSIS OF URANIUM AND PLUTONIUM USING ACTINIDE RESIN
A.R. King, D. Knight, A. Fairhead (UK)
PA8
PA8-1
COMPUTATIONAL CHEMISTRY
QUANTUM CHEMICAL INVESTIGATION OF THE SORPTION OF SELENITE ON
THE CALCITE (104) SURFACE AND INCORPORATION INTO THE BULK PHASE
R. Polly, F. Heberling, B. Schimmelpfennig, H. Geckeis (Germany)
PA8-2
PB1
PB1-1
PB1-2
PB1-3
PB1-4
PB1-5
QUANTUM CHEMICAL INVESTIGATION OF THE SORPTION OF SELENITE ON
THE CALCITE (104) SURFACE AND INCORPORATION INTO THE BULK PHASE
Xiaoyu Liu, Zhong Zheng, Na Zhang, Chunli Liu (China)
SORPTION/DESORPTION PHENOMENA IN
DYNAMIC SYSTEMS
AGING EFFECT OF SORPTION OF 32-YEAR-OLD PLUTONIUM COMPLEXES ON
SAVANNAH RIVER SITE SEDIMENTS
H.P. Emerson, B. A. Powell (USA)
MIGRATION OF URANIUM THROUGH SANDSTONE IN THE ALKALINE
DISTURBED PLUME FROM A CEMENTITIOUS REPOSITORY
M. Felipe-Sotelo, A.E Milodowski , N. Bryan, N.D.M. Evans (UK)
ANALYSING METAL SPECIATION AND MOBILITY IN CLAY - FROM ICP-MS
BATCH EXPERIMENTS TO A NEW APPROACH OF MINIATURISED CLAY
COLUMN EXPERIMENTS (MCCE) USING LC-ICP-MS
R. Kautenburger, C. Hein, J.M. Sander, H.P. Beck, G. Kickelbick (Germany)
STUDY OF SORPTION AND DESORPTION BEHAVIOUR OF RADIONUCLIDES IN
COLUMN FILLED WITH CRUSHED GRANITE
K. Videnská, Š. Palágyi, K. Štamberg, V. Havlová, H. Vodičková (Czech Republic)
STUDY ON THE INTERACTION BETWEEN HUMIC ACIDS AND GRANITE FROM
BEISHAN AREA, CHINA
Chunli Wang, Chun Li, Chunli Liu (China)

PB1-6
DETERMINATION OF MIGRATION PARAMETERS OF CRYSTALLINE ROCKS:
APPLICATION OF ELECTRO-MIGRATION METHOD ON SAMPLES WITH
DIFFERENT LENGTHS
P. Vecerník, V. Havlová (Czech Republic)
PB1-7
THE INFLUENCE OF HUMIC ACID ON THE MIGRATION OF CAESIUM,
NICKEL AND EUROPIUM CATIONS THROUGH QUARTZ SAND
S. L. Jain, N. Evans, N. Bryan (UK)
PB3
PB3-1
PB3-2
PB3-3
PB3-4
PB3-5
PB5
PB5-1
PB5-2
PB5-3
PB5-4
COLLOID MIGRATION
HYDRO-GEOCHEMICAL EFFECTS ON COLLOID MIGRATION IN THE
ARTIFICIAL, SINGLE-FRACTURED GRANITIC ROCK
S. Lee, J.-W. Kim, M.-H. Baik, J. Jeong (Korea)
COLLOIDAL TRANSPORT IN A PODZOLIC SOIL CONTAMINED BY DEPLETED
URANIUM: SITE INVESTIGATION, LABORATORY STUDIES AND MODELING
S. Harguindeguy, P. Crançon, G. Lespes, L. De Windt, F. Pointurier, M. Potin Gautier
(France)
THRESHOLD OF BENTONITE COLLOID EROSION UNDER STATIC AND FLOW
CONDITIONS: RELEVANCE OF SCENARIO EVOLUTION
U. Alonso, T. Missana, M. García-Gutiérrez, N. Albarran, M. Mingarro (Spain)
THE EFFECT OF COLLOIDS ON RADIONUCLIDE TRANSPORT IN COLUMN
EXPERIMENTS
S. Niemiaho, P. Hölttä, M. Voutilainen, J. Lehto (Finland)
THE EFFECT OF COLLOIDS ON RADIONUCLIDE TRANSPORT IN COLUMN
EXPERIMENTS
S. Niemiaho, P. Hölttä, M. Voutilainen, J. Lehto (Finland)
FIELD AND LARGE-SCALE EXPERIMENTS
AN OVERVIEW OF THE LONG-TERM DIFFUSION TEST, GRIMSE TEST SITE,
SWIZTERLAND
A. Martin, M. Siitari-Kauppi, V. Havlová, Y. Tachi, J. Miksova (Switzerland, Finland, Czech
Republic, Japan)
IN-SITU MIGRATION EXPERIMENTS AT THE ROCK FRACTURES OF KURT
(KAERI UNDERGROUND RESEARCH TUNNEL)
J.-W. Kim, J.-K. Lee, M.-H. Baik, J. Jeong (Korea)
THE LATEST RESULTS ON COLLOID ASSOCIATED RADIONUCLIDE MOBILITY
FROM THE CFM PROJECT, GRIMSEL (SWITZERLAND)
T. Schäfer, I. Blechschmidt, M. Bouby, S. Büchner, J. Brendlé, G. Darbha, H. Geckeis, T.
Kupcik, R. Götz, W. Hauser, S. Heck, F. Huber, M. Lagos, A. Martin (Germany, Switzerland,
France)
QUANTIFYING 14 CH 4 MIGRATION AND FATE FOLLOWING SUB-SURFACE
RELEASE TO AGRICULTURAL SOIL
G. Shaw, B. Atkinson, W. Meredith, C. Snape, A. Hoch, D. Lever (UK)

PB5-5
PB5-6
PB5-7
PC2
PC2-1
PC2-2
PC2-3
PC2-4
PC2-5
PC2-6
REPRO - THE IN SITU RADIONUCLIDE MIGRATION EXPERIMENT AT ONKALO
UNDERGROUND FACILITY
A. Poteri, K. Helariutta, J. Ikonen, M. Voutilainen , M. Siitari-Kauppi, P. Andersson, J.
Byegård, K Nilsson, M. Skålberg, J. Kuva, P. Kekäläinen, J. Timonen, P. Pitkänen, K.
Kemppainen, J. Liimatainen, I. Aaltonen, L. Koskinen (Finland, Sweden)
CONTRASTING ACTIVITIES OF FALLOUT RADIONUCLIDES BETWEEN TWO
TYPES OF ARCTIC SOILS
E. Łokas, P. Wachniew , M. Gąsiorek (Poland)
AN IN-SITU EXPERIMENT TO STUDY SORPTION AND DIFFUSION OF SEVERAL
RADIONUCLIDES IN UNDISTURBED GRANITIC ROCK AT NATURAL
CONDITIONS IN THE ÄSPÖ HARD ROCK LABORATORY, SWEDEN. LTDE-SD
(LONG TERM SORPTION DIFFUSION EXPERIMENT)
J. Byegård, E. Gustafsson, S. Höglund, M. Skålberg, H. Widestrand (Sweden)
COUPLING CHEMISTRY AND TRANSPORT
RADIONUCLIDES TRANSPORT OF PYRO-PROCESSED WASTE IN KOREAN
REFERENCE DISPOSAL SYSTEM
C.H. Kang, J.T. Jeong (Korea)
NUMERICAL MODELING OF IRON-CORROSION AND INTERACTION WITH
BENTONITE IN CLAY FORMATIONS
C. Hansmeier, G. Bracke, B. Reichert (Germany)
IMPACT OF POROSITY CLOGGING ON DIFFUSION RATE: EXPERIMENTS VERSUS
MODELING
I. Fatnassi, S. Savoye, P. Arnoux, P. Gouze, O. Bildstein, V. Detilleux, C. Wittebroodt
(France, Belgium)
COUPLED THCM MODELS OF HEATING AND HYDRATION EXPERIMENTS
PERFORMED ON SAMPLES OF COMPACTED FEBEX BENTONITE IN CONTACT
WITH CONCRETE AND CARBON STEEL
J. Samper, A. Mon, L. Montenegro, J. Cuevas, R. Fernández, M.J. Turrero, E. Torres, A.
Naves (Spain)
QUANTIFYING PLUTONIUM SORPTION AND DESORPTION RATES FROM
MINERAL SURFACES: A NUMERICAL APPROACH TO MODELING BATCH AND
FLOW CELL EXPERIMENTAL DATA
M. Zavarin, B.A. Powell, J. Begg, A.B. Kersting (USA)
ORCHESTRA: A FRAMEWORK FOR INTEGRATING DETAILED REACTIVE
TRANSPORT PROCESSES OF RADIONUCLIDES INTO PERFORMANCE
ASSESSMENT MODELS
J.C.L. Meeussen, E. Rosca-Bocancea J. Grupa, T.J. Schröder (Netherlands)

PC5
PC5-1
PC5-2
PC5-3
PC5-4
PC5-5
PFS
PFS-1
PFS-2
PFS-3
SAFETY ASSESSMENT AND REPOSITORY
CONCEPTS
DECOMPOSITION OF U(VI)-ARSENAZO III COMPLEX BY FENTON REACTION
INDUCED BY GAMMA IRRADIATION
Z. Zheng, M.L. Kang, C.L. Liu, T. Chen, B. Grambow, L. Duro, T. Suzuki-Muresan (China,
France, Spain)
THE BIGRAD CONSORTIUM - SPECTROSCOPIC STUDIES OF THE
TRANSFORMATION OF MINERALS IN SANDSTONE UNDER HYPERALKALINE
CONDITIONS
A.F. Seliman, S.A. Banwart, M.E. Romero-Gonzales (UK)
RADIATION INDUCED CORROSION OF COPPER IN DEEP GEOLOGICAL
REPOSITORIES FOR SPENT NUCLEAR FUEL
Å. Björkbacka, S. Hosseinpour, M.C. Johnson, C. Leygraf, M. Jonsson (Sweden)
GEOCHEMICAL MODEL OF CRUSHED GRANITE DISSOLUTION AT 70°C
I. Brusky, J. Sembera (Czech Republic)
INVENTORY OF A DEEP GEOLOGICAL RADIOACTIVE WASTE REPOSITORY IN A
SALT FORMATION
H. Seher, H. Fischer, J. Larue (Germany)
ENVIRONMENTAL BEHAVIOUR OF RADIO-
NUCLIDES AFTER THE FUKUSHIMA ACCIDENT
FIXATION AND ITS DYNAMCIS OF RADIOACTIVE CESIUM IN FUKUSHIMA
SOILS
T. Saito , H. Makino , S. Tanaka (Japan)
QUANTIFICATION OF THE NATURAL REDISTRIBUTION OF RADIOCAESIUM IN
THE AFTERMATH OF THE FUKUSHIMA DAI-ICHI NUCLEAR ACCIDENT
H. Sato, K. Iijima, T. Niizato, S. Nakayama, S.M.L. Hardie, I.G. McKinley (Japan,
Switzerland)
DEVELOPMENT OF A BOX-MODELLING APPROACH TO INTEGRATE
CONTAMINATED SITE UNDERSTANDING IN FUKUSHIMA
S.M.L. Hardie, I.G. McKinley, T.M. Beattie, E.M. Scourse, L. Klein (Switzerland, UK)
PUK
PUK-1
PUK-2
SPECIAL UK SESSION
SYNTHESIS OF SUPERPLASTICISERS TAILORED FOR APPLICATIONS IN
NUCLEAR DECOMMISSIONING AND STORAGE
M. Isaacs, S. Edmondson, S. Christie, D. Read (UK)
CARBONATION OF REPOSITORY CEMENT: IMPACT OF CO 2 ON CEMENT
MINERALOGY, WATER CHEMISTRY AND PERMEABILITY
G. Purser, C. Rochelle, A. Milodowski, D. Noy, J. Harrington D. Wagner, A. Butcher (UK)

PUK-3
PUK-4
PUK-5
PUK-6
PUK-7
PUK-8
PUK-9
PUK-10
THE ROLE OF MONITORED NATURAL ATTENUATION IN MANAGEMENT OF
RADIOACTIVELY CONTAMINATED GROUND FROM THE SELLAFIELD MAGNOX
SWARF STORAGE SILO
J. Graham (UK)
A STATISTICAL APPROACH TO INVESTIGATING ENHANCEMENT OF
POLONIUM-210 IN THE EASTERN IRISH SEA ARISING FROM DISCHARGES
FROM A FORMER PHOSPHATE PROCESSING PLANT
A. Dewar, W. Camplin, J. Barry, P. Kennedy (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM
K. Morris, N.D. Bryan, N.D.M Evans, F.R. Livens, J.R. Lloyd,
J.F.W. Mosselmans, A. Milodowski, S. Shaw, J.S. Small, S. Thornton (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM: MODELLING AND SYNTHESIS
J.S. Small, L. Abrahamsen, X. Chen, H Steele, S. F. Thornton (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM: WORK PACKAGE 2 – RADIONUCLIDE FORM,
REACTION, AND TRANSPORT
N.D. Bryan, K. Morris, G.T.W. Law, N.D.M Evans, F.R. Livens,
J.R. Lloyd, J.F.W. Mosselmans, S. Shaw, K. Smith, A. Stockdale, P. Bots, M. Felipe-Sotelo
(UK)
MICROBIOLOGICAL CROSS-CUTTING RESEARCH; THE NERC
BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT BIGRAD
CONSORTIUM
J.R. Lloyd, A. Williamson, C. Boothman, A. Rizoulis, N. Bassil, S. Smith, K. Morris, S. Shaw,
G.T.W. Law, A. Milodowski, J.M. West, J.S. Small, J.F.W. Mosselmans (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM: WORK PACKAGE 1 – GEOSPERE EVOLUTION
S. Shaw, A.E. Milodowski, E.B.A. Moyce C. Rochelle, G.T.W. Law, J.F.W. Mosselmans, P.
Bots, T.A.M. Marshall, M.E. Romero-Gonzalez, A. Seliman, K. Morris (UK)
THE BIGRAD CONSORTIUM – RADIONUCLIDE SORPTION AT HIGH PH IN
CALCITE SYSTEMS
Kurt Smith, Nick Bryan, Katherine Morris

WEDNESDAY (11. SEPTEMBER)
SESSION 11
B4: EFFECTS OF BIOLOGICAL AND
ORGANIC MATERIALS
Chair:
H. Nitsche (USA) and T. Ohnuki (Japan)
8:30 am MICROBIAL METABOLISM; CRITICAL CONTROLS ON RADIONUCLIDE
MIGRATION IN NATURAL AND ENGINEERED ENVIRONMENTS
J.R. Lloyd, A. Rizoulis, N. Bassil, L. Newsome, C.L. Thorpe, V.E. Evans, A.R.
Brown, R. Kimber, C. Boothman, G.T.W. Law, F.R. Livens, K. Morris
(INVITED) (UK)
9:15 am AQUEOUS STABLE 127 I AND RADIOACTIVE 129 I SPECIATION AND
UPTAKE BY SUBSURFACE SEDIMENTS IN AN ARID ENVIRONMENT
D. I. Kaplan, C. Xu, S. Zhang, H.-S. Chang, H.-P. Li, Y.-F. Ho, K.A. Schwehr,
C.M. Yeager, D. Wellman, P.H. Santschi (USA)
9:40 am EXPERIMENTS TO ASSESS THE MOBILITY OF NICKEL ISOTOPES
THROUGH A CEMENTITIOUS BACKFILL IN THE PRESENCE OF
CELLULOSE DEGRADATION PRODUCTS, USING 63 NI AS A TRACER
J. Hinchliff, M Felipe-Sotelo, D. Read, N.D.M. Evans, S. J. Williams, D. Drury,
A.E. Milodowski (UK))
10:05 am BIOSORORPTION OF ACTINIDES TOWARDS HALOPHILIC
MICROORGANISMS
D.T. Reed, J.S. Swanson, K. Simmons, J.F. Lucchini, D. Cleveland, M.K. Richman
(USA)
B4-1
B4-2
B4-3
B4-4
10:30 am BREAK
SESSION 12
A3: COMPLEXATION WITH INORGANIC AND
ORGANIC LIGANDS
Chair:
N. Kaltsoyannis (UK) and M. Altmaier (Germany)
10:50 am OXYGEN EXCHANGE BETWEEN URANYL(VI) AND WATER:
BINUCLEAR SCENARIOS IN ACID AND IN BASE
S. Tsushima, A. Rossberg, H. Moll (Germany)
11:15 am THE BIGRAD CONSORTIUM - RADIONUCLIDE INTERACTIONS WITH
NATURAL ORGANIC MATTER AT HIGH pH
A. Stockdale, N.D. Bryan, S. Lofts, E. Tipping (UK)
11:40 pm A COMBINED EXAFS AND TRLFS SPECTROSCOPIC STUDY TO
DETERMINE THE THERMODYNAMIC AND STRUCTURAL PROPERTIES
OF TRIVALENT ACTINIDE COMPLEXES WITH ORGANIC AND
INORGANIC LIGANDS AT ELEVATED TEMPERATURES
D.R. Fröhlich, A. Skerencak-Frech, P.J. Panak (Germany)
A3-4
A3-5
A3-6
12:05 pm FREE

THURSDAY (12. SEPTEMBER)
SESSION 13
A4: REDOX REACTIONS AND RADIOLYSIS
Chair:
B. Grambow (France) and W.H. Kim (Korea)
8:30 am IMPACT OF WATER RADIOLYSIS ON URANIUM DIOXIDE CORROSION
A. Traboulsi, J. Vandenborre, G. Blain, J. Barbet, M. Fattahi (France)
8:55 am REDOX REACTION OF Se(IV)/Se(VI) WITH NATURAL PYRRHOTITE
AND IRON SELENIDES
M. L Kang, B. Ma, F. Bardelli, F. R Chen, C. L Liu, L. Charlet, J.L Xie (China,
France)
9:20 am INTERFACIAL RADIATION CHEMISTRY IN GEOLOGICAL
REPOSITORIES FOR SPENT NUCLEAR FUEL
M. Jonsson (Sweden)
9:45 am IMPACT OF Fe MINERALS ON THE SAFE DISPOSAL OF NUCLEAR
WASTE
C. Pearce, K. Rosso, J. Liu, S. Luksic, M. Schweiger, J. McCloy, A. Felmy (USA)
A4-1
A4-2
A4-3
A4-4
10:10 am BREAK
SESSION 14
C3: DEVELOPMENT AND APPLICATION OF
MODELS
Chair:
J. Bruno (Spain) and N.D. Bryan (UK)
10:30 am MODELLING OF Tc (IV) INTERACTION WITH DISSOLVED ORGANIC
MATTER FROM BOOM CLAY BY A KINETIC APPROACH
C. Bruggeman, A. Maes, N. Maes, E. Martens, J. Vancluysen, M. Van Gompel
(Belgium)
10:55 am RADIONUCLIDE SOLUBILITY CALCULATIONS IN CRYSTALLINE AND
SEDIMENTARY GROUNDWATERS FROM THE CANADIAN SHIELD
D. García, L. Duro, E. Colàs, V. Montoya (Spain)
11:20 am DEVELOPING QUALITY ASSURED SORPTION DATABASES FOR
PERFORMANCE ASSESSMENT
E. Klein, S.M.L. Hardie, E.M. Scourse and I.G. McKinley (Switzerland, UK)
11:45 am DIFFUSION OF CESIUM IN GRIMSEL GRANODIORITE: SIMULATIONS
IN TIME DOMAIN WITH HETEROGENEOUS SORPTION PROPERTIES
M. Voutilainen, P. Sardini , M. Siitari-Kauppi , A. Martin, J. Timonen (Finland,
France, Switzerland)
C3-1
C3-2
C3-3
C3-4
12:10 pm BREAK

SESSION 15
A5: SOLID-LIQUID INTERFACE REACTIONS
Chair:
M.A. Denecke (UK) and J. Lehto (Finland)
2:00 pm THE BIGRAD CONSORTIUM - ALTERATION OF SANDSTONE BY
ALKALINE CEMENT LEACHATES AND ITS EFFECT ON URANIUM
SPECIATION
E.B.A. Moyce, K. Morris, T. A. Marshal, S. Shaw (UK)
2.25 pm 99-TECHNETIUM BEHAVIOR UPON INTERACTIONS WITH HEMATITE
N.A. Wall, L.C. Gribat (USA)
A5-5
A5-6
2:50 pm
3:15 pm
3:40 pm
DEVELOPMENT AND TESTING OF COMPONENT ADDITIVITY SURFACE
COMPLEXATION MODELS OF NEPTUNIUM AND PLUTONIUM
SORPTION TO SOILS
B.A. Powell, S. Herr, N. Conroy, J. Wong, Yu Xie, M. Zavarin, A.B. Kersting (USA)
SORPTION OF Cs(I) AND Sr(II) ON SMECTITE RICH CLAY: BATCH
SORPTION AND MODELLING
S. Kasar, A.S. Kar, S. Kumar, A.S. Pente, C.P. Kaushik, R.K. Bajpai, B.S. Tomar
(India)
BREAK
A5-7
A5-8
SESSION 16
A2: SOLID SOLUION AND SECONDARY
PHASE FORMATION
Chair:
J.-I. Yun (Korea) and D Read (UK)
4:00 pm URANIUM OXIDATION AND MIGRATION THROUGH CRYSTALLINE
ROCK - TRANSPORT PATHWAYS AND THE ROLE OF SECONDARY
PHASES IN RETARDATION
D. Read, S. Black, D. Thornley, A. Milodowski, M. Siitari-Kauppi, W. E. Falck
(UK, Finland, France)
4:25 pm FROM URANOTHORITES TO COFFINITE: A SOLID SOLUTION ROUTE
TO THE THERMODYNAMIC PROPERTIES OF USIO4
S. Szenknect, D.T. Costin, N. Clavier, A. Mesbah, C. Poinssot, P. Vitorge, N.
Dacheux (France)
4:50 pm TETRAVALENT CATION COPRECIPITATION WITH CLAY MINERALS
N. Finck, M. Bouby, K. Dardenne, H. Geckeis (Germany)
A2-1
A2-2
A2-3
5:15 pm INVESTIGATIONS INTO THE FORMATION OF NEPTUNIUM(IV)-SILICA
COLLOIDS
R. Husar, S. Weiß, C. Hennig, H. Zänker, G. Bernhard (Germany)
5:40 pm LIMITED REACTION OF BENTONITE AND SWELLING CLAYS IN LOW
ALKALI CEMENT LEACHATES: FINAL RESULTS FROM THE CYPRUS
NATURAL ANALOGUE PROJECT (CNAP)
W.R. Alexander, A.E. Milodowski, S.J. Kemp, J.C. Rushton, P. Korkeakoski, S.
Norris and P. Sellin (Switzerland, UK, Finland, Sweden)
A2-4
A2-5

FRIDAY (13. SEPTEMBER)
Chair:
V. Brendler (Germany) and J. Lloyd (UK)
8:30 am THE UNUSUAL CHEMISTRY OF PLUTONIUM
D.L. Clark (INVITED) (USA)
SESSION 17
B2: DIFFUSION AND OTHER MIGRATION
PROCESSES
9:15 am REACTIVE TRANSPORT OF U(VI) THROUGH POROUS MEDIA AMENDED
WITH PHOSPHATE TO INDUCE IN SITU URANIUM IMMOBILIZATION
V.S. Mehta, F. Maillot, Z. Wang, J.G. Catalano, D.E. Giammar (USA)
9:40 am URANIUM MIGRATION AND RETENTION MECHANISM IN THE PROCESS
WASTE OF A CONVERSION FACILITY
T. Fernandes, L. Duro, Th. Schäfer, P. Masqué, A. Delos, J.S. Flinois, G. Videau
(Spain, Germany, France)
10:05 am DISSOLVED ORGANIC MATTER AND COLLOID MEDIATED TRANSPORT IN
BOOM CLAY: SIZE EFFECTS
D. Durce, C. Bruggeman, N. Maes (Belgium)
B2-4
B2-5
B2-6
10:30 am BREAK
SESSION 18
ENVIRONMENTAL BEHAVIOUR OF RADIONUCLIDES
AFTER THE FUKUSHIMA ACCIDENT
Chair:
P. Toulhoat (France) and U. Berner (Switzerland)
10:50 am RADIOCESIUM IN THE ENVIRONMENT IN THE AFTERMATH OF THE
FUKUSHIMA DAIICHI ACCIDENT: STATUS OF RADIOACTIVITY,
DECONTAMINATION WORK AND MIGRATION RESEARCH PROJECT
M. Yui (INVITED) (Japan)
11:35 am ENVIRONMENTAL BEHAVIOR OF RADIOCESIUM AFTER FUKUSHIMA DAIICHI
NUCLEAR POWER PLANT ACCIDENT
T. Ohnuki, F. Sakamoto, N. Kozai, S. Yamasaki (Japan)
12:00 am EFFECT OF SOIL PARAMETERS ON SORPTION PROPERTIES OF
ACTINIDES AND FISSION PRODUCTS : DEPTH PROFILE DISTRIBUTION
OF FALLOUT RADIONUCLIDES IN SOILS AFFECTED BY FUKUSHIMA
NUCLEAR POWER PLANT ACCIDENT
S. Mishra, A. Sorimachi, M. Hosoda, S. Tokonami, T. Ishikawa, S.K. Sahoo (Japan)
FS-1
FS-2
FS-3
12:30 pm WRAP-UP
T. Fanghänel H. Geckeis, (EU, Germany)
13:00 pm END

SESSION 1
A1: SOLUBILITY AND DISSOLUTION
SOLUBILITY, SPECIATION AND THERMO-DYNAMICS OF ACTINIDES AND
FISSION PRODUCTS
M. Altmaier (INVITED) (Germany)
EVALUATION OF DISCREPANCIES IN TETRAVALENT OXIDE SOLUBILITY
VALUES BY ISOTOPIC EXCHANGE AND ITS IMPACT ON THE SAFETY
ASSESSMENT
T. Suzuki-Muresan , J. Vandenborre, B. Grambow, A. Valls, L. Duro (France, Spain)
REDOX CHEMISTRY, SOLUBILITY AND HYDROLYSIS OF TECHNETIUM IN
DILUTE TO CONCENTRATED NaCl and MgCl 2 SOLUTIONS
E. Yalcintas, X. Gaona, M. Altmaier, A.C. Scheinost, T. Kobayashi, H. Geckeis (Germany,
Japan)
LONG-TERM AQUEOUS ALTERATION KINETICS OF A 99 Tc -DOPED SON68
BOROSILICATE GLASS
S. Rolland, M. Tribet, M. Magnin, V. Broudic, S. Peuget, A. Janssen, T. Wiss, C. Jégou, P.
Toulhoat (France, EU)
A1-1
A1-2
A1-3
A1-4
A1-1
SOLUBILITY, SPECIATION AND THERMODYNAMICS OF ACTINIDES
AND FISSION PRODUCTS
M. Altmaier
Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Germany.
Long-term disposal of nuclear waste in deep underground repositories is the safest option to separate highly
hazardous radionuclides from the environment. On an international level three different host rock formations
(crystalline, clay, rock salt) are currently considered. With respect to predicting the long-term safety of a
repository, specific (geo)chemical scenarios for each host rock must be analyzed which require a detailed
and reliable understanding of solubility, speciation and thermodynamics of actinides and long-lived fission
products. Radionuclide solubility and speciation strongly depend on chemical conditions (pH, E h , matrix
electrolyte system and ionic strength) with additional factors like the presence of inorganic and organic
ligands or temperature further impacting solution chemistry. The broad set of potential geochemical
boundary conditions defines both the need to investigate fundamental aspects of aqueous actinide chemistry
and the need to supply reliable thermodynamic data for applied model calculations. As this input is an
obvious key requirement needed to derive reliable solubility limits for source term estimations and PA, the
precise description of aqueous and solid radionuclide speciation is similarly important for neighboring fields.
For example, the mechanistic understanding and quantitative description of solid liquid interphase reactions,
waste dissolution processes, secondary phase and solid solution formation, must be based upon qualitatively
and quantitatively correct radionuclide speciation models.
The main driving force for progress in aquatic actinide and fission product chemistry and thermodynamics
over the last decade is related to the frequent use of advanced spectroscopic tools for investigating chemical
speciation at the molecular level. The technical progress in modern analytics has been directly translated into
a significantly improved understanding of actinide and fission product chemistry. A review on recent
advances in aqueous actinide chemistry and thermodynamics has currently been made available in the
literature [1]. Based upon new chemical and structural information, a much more detailed picture of
radionuclide speciation is becoming available, generating advanced process understanding and input for
more realistic chemical models. Computational chemistry also has developed into a valuable tool for actinide
science as it offers information from a quantum chemical perspective to complement conventional chemical
approaches. Thermodynamic data and databases have been significantly improved by the Thermodynamic
Database Project of the OECD-NEA (http://www.oecd-nea.org/dbtdb). The NEA-TDB project has published

a series of expertly reviewed and evaluated compilations of consistent thermodynamic data being widely
accepted as reference values for many key elements and ligand systems.
In the present contribution, an overview on recent studies in aquatic actinide and fission product chemistry
will be given. The need to continuously dedicate research efforts to aqueous actinide and fission product
chemistry and thermodynamics will be critically reflected against the progress made in this research field
over the last decades. It is concluded that it remains essential to improve the quality of chemical models and
reduce present systematic and data uncertainties as a key contribution to repository safety. It will be
especially important to fill relevant data gaps and improve data for more complex geochemical systems, e.g.
high ionic strength systems, kinetically controlled processes, redox transformations or aquatic chemistry at
elevated temperatures. The need for international networking and scientific exchange is likewise emphasized
and pointed out as a powerful instrument to continuously and efficiently optimize the scientific description of
solubility, speciation and the thermodynamics of actinides and fission products.
[1] Altmaier, M., Gaona, X., Fanghänel, Th., Recent Advances in Aqueous Actinide Chemistry and Thermodynamics,
Chem. Rev. (2013), 113, 901-943.
A1-2
EVALUATION OF DISCREPANCIES IN TETRAVALENT OXIDE SOLUBILITY VALUES BY
ISOTOPIC EXCHANGE AND ITS IMPACT ON THE SAFETY ASSESSMENT
Suzuki-Muresan T. *(1) , Vandenborre J. (1) , Grambow B. (1) , A. Valls (2) , L. Duro (2)
(1) SUBATECH / CNRS-IN2P3 / Université de Nantes, 4 rue Alfred Kastler, BP 20722, 44307 Nantes cedex
03, France
(2) Amphos 21, P. García Faria 49-51, 1-1, E-08019-Barcelona, Spain
7* Corresponding author: suzuki@subatech.in2p3.fr
Most of the primary solids expected to form under repository conditions after the eventual release of the
radionuclides present in the different waste forms are oxides. These oxides form in a first stage a rather
amorphous oxide characterized by a hydrous surfaces. These solids are sparingly soluble, so that apparent
equilibrium states are reached in relatively short time periods. Nevertheless, the amorphous nature of these
surfaces enhances the dynamism of the systems, so that they are not under real equilibrium, but in a
continuous evolution. The understanding of the kinetics of alteration of this type of solid phase is the focus
of this work in the framework of the European project SKIN (FP7-269688, 2011-2013). Specific
investigations are focused on the study of the kinetics of dissolution of tetravalent oxides under conditions
close to equilibrium, in particular on the study of the solid-liquid interface, including monitoring the
composition of the surface through isotopic exchange techniques.
Thorium oxide is a suitable candidate to study the mechanism controlling the equilibrium at the interface
solid and solution, avoiding the oxidation states and focusing on the solid-liquid interface. Previous studies
on dissolution of ThO 2 (c) indicate that the dissolution is very dependent on the material at the grain
boundary [1,2]. 80 % of the XPS accessible near surface region of sintered thorium oxide is represented by
the less reactive ThO 2 (cr) grains. The remaining 20 % corresponds to ThO x (OH) y (H 2 O) z which is largely
associated with grain boundaries. The empirical solubility data does not correspond to thermodynamic bulk
phase/solution equilibrium, as measured solution concentrations are controlled by specific site exchange
mechanisms at the solid/solution interface. Therefore for sparingly soluble solids, one needs to quantify the
specific surface site involved in the attachment and detachment rates if one wants to assess solubility
constraints.
Different crystalline states of thorium oxide surfaces without the grain boundaries were synthesized and
studied in this work and compared with crushed and washed HTR sphere of ThO 2 . Four sets of were studied:
(i) Crushed ThO 2 spheres, (ii) Initial ThO 2 spheres, (iii) ThO 2 powder synthesized at 1300°C and (iv) ThO 2
powder synthesized at 700°C. The characterization of the crystalline state of the solids was realized by X-
Rays Diffraction by measuring the Full Width Half Middle onto the XRD peaks as a function of temperature
of synthesis (700° and 1300°C). After about 50 days, a pseudo steady-state equilibrium is reached for all

samples. The concentration of Th for the powder synthesized at 1300°C and the crushed sphere are close
(2.98×10 -8 mol/L and 2.95×10 -8 mol/L, respectively). It seems that the global and the kinetic dissolution
behaviors are similar for these two samples. The crystalline state of the samples is involved in the dissolution
mechanism during the initial kinetic dissolution and the reaching of the pseudo steady-state equilibrium. The
ratio between grain and grain boundaries and the global and kinetic dissolution behavior can be described as
following: ThO 2 powder synthesized at 700°C > Initial ThO 2 spheres > Crushed ThO 2 spheres = ThO 2
powder synthesized at 1300°C. In the aim of understanding the surface behavior at the interface
solid/solution under conditions close to equilibrium, a spike of radioactive tracer, here 229 Th, is added in
solution. Depending on the crystalline surface state, one can observe isotopic exchange between 229 Th spike
and 232 Th from the solid or adsorption phenomena due to the existence of the grain boundaries. An example
of isotopic exchange 229 Th/ 232 Th result is shown in Figure 1 for a powder of ThO 2 synthesized at 700°C.
In addition, a literature review of thorium solubility has been done in the aim to assess the impact of the
uncertainties of the solubility values on the safety of disposal. The solubility product values range between
ThO 2 (cr) (log K°sp=-56.9 [3] and ThO x (OH) y (H 2 O) z (s) (log K°sp=-45.5 [4,5] while an average value was
suggested for the amorphous phase [6] (log K°sp=-47). However, most oxides, like for example SiO 2 , form
OH groups at the surface but this does not lead to discrepancy between solid state thermodynamic data and
thermodynamic data obtained from solubility values. Hence, despite the careful recent review of
thermodynamic data (Rand et al. 2008) it is still not clear how to predict thorium oxide solubility. Therefore,
the large uncertainty in solid phase properties and solubility product values requires a new approach linking
bulk solid, solid surface and solubility properties. We use solid analysis, leaching experiment and isotopic
exchange to contribute to the understanding of the discrepancy in the solubility data. A model of safety case
has been chosen in the framework of the SKIN project which takes into account the transfer of radionuclides
of interest from the source term to the sink compartment. The concentration of Th has been assessed from the
first and last compartments of bentonite and granite. In the model, thorium solubility in the canister and bentonite is
7.9·10 -7 M and in granite is 8.9·10 -9 M. Those values are in agreement with literature data.
The work will be presented in four parts: (i) the kinetics of dissolution of thorium oxides powders and
spheres; (ii) the advance results on the isotopic exchange 229 Th/ 232 Th; (iii) the impact of the crystalline
surface state on the thorium oxide solubility; (iv) the evolution of the concentration of Th in all compartments
of the system (safety case model) from the first one of bentonite to the last of granite.
[ 232 Th] (mol/L)
5.0x10 -6 [232Th]
[229Th]
4.0x10 -6
3.0x10 -6
2.0x10 -6
1.0x10 -6
8.0x10 -9
6.0x10 -9
4.0x10 -9
2.0x10 -9
[ 229 Th] (mol/L)
0.0
0 100 200 300 400
Time duration (days)
0.0
Figure 1. [ 232 Th] and [ 229 Th] as a function of contact time at pH = 3.2, monitored by Q-ICP-MS
[1] Vandenborre, J., Abdelouas, A., Grambow, B. (2008). “Discrepancies in Thorium Oxide Solubility Values: a New
Experimental Approach to Improve Understanding of Oxide Surface at Solid/Solution Interface.” Radiochimica Acta
96, 515-520.
[2] Vandenborre, J., Grambow, B., Abdelouas, A. (2010). “Discrepancies in Thorium Oxide Solubility Values: Study of
Attachment/Detachment Processes at the Solid/Solution Interface.” Inorganic Chemistry 49, 8736-8748.
[3] Rai, D., Moorw, D.A., Oakes, C.S., Yui, M. (2000). “Thermodynamic model for the solubility of thorium dioxide in
the Na + -Cl - -OH - -H 2 O system at 23°C and 90°C.” Radiochim. Acta, 88, 297-306

[4] Felmy, A.R., Rai, D., Mason, M.J. (1991). “The solubility of hydrous thorium(IV) oxide in chloride media :
development of an aqueous ion-interaction model.” Radiochim. Acta, 55, 177-185
[5] Rai, D., Felmy, A.R., Sterner, S.M., Moore, D.A., Mason, M.J., Novak, C.F. (1997) “The solubility of Th(IV) and
U(IV) hydrous oxides in concentrated NaCl and MgCl 2 solutions.” Radiochim. Acta, 79, 239-247
[6] Bitea, C., Müller, R., Neck, V., Walther, C., Kim, J.I. (2003) Study of the generation and stability of thorium(IV)
colloids by LIBD combined with ultrafiltration. Colloids Surf., A, 217, 63-70
A1-3
REDOX CHEMISTRY, SOLUBILITY AND HYDROLYSIS OF TECHNETIUM IN DILUTE TO
CONCENTRATED NaCl and MgCl 2 SOLUTIONS
E. Yalcintas 1)* , X. Gaona 1) , M. Altmaier 1) , A.C. Scheinost 2) , T. Kobayashi 3) , H. Geckeis 1)
1) Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Germany
2) Institute of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf, Germany
3) Department of Nuclear Engineering, Kyoto University, Japan
Technetium–99 is a β–emitting fission product highly relevant for the safety assessment of nuclear waste
repositories due to its significant inventory in radioactive waste, long half-life (t ½ ~211.000 a) and redox
sensitivity. Tc(VII) and Tc(IV) are the most stable redox states of Tc in the absence of any complexing
ligands under sub-oxic/oxidizing or reducing aqueous environments. Tc(VII) exists as highly soluble and
mobile TcO 4 − pertechnetate anion, whereas Tc(IV) forms sparingly soluble hydrous oxides (TcO 2 ∙xH 2 O(s)).
Due to the very different mobility of both redox states, the redox chemistry of Tc as well as Tc(IV) solubility
and hydrolysis are of special relevance in the context of radioactive waste disposal. An appropriate
understanding of the chemistry of Tc in concentrated NaCl and MgCl 2 solutions of extremely high ionic
strength is required in the performance assessment of underground repositories in rock-salt formations.
12
8
4
"Redox neutral"
(pe+pH)=13.8
0.8
0.6
0.4
0.2
pe
0
-4
-8
-12
1 bar H 2
(g)
(pe+pH)=0
Na2S2O4
Sn(II)
HQ
Fe(II)/Fe(III)
Fe powder
Magnetite
Mackinawite
Siderite
-0.8
0 2 4 6 8 10 12 14
pH c
Fig. 1: Redox behavior of Tc(VII)/Tc(IV) in 5.0 M NaCl. Open symbols indicating reduction to Tc(IV); solid symbols
corresponding to samples with no reduction of Tc(VII). Dashed line indicates the Tc(VII)/Tc(IV) borderline calculated
according with the NEA-TDB.
0.0
-0.2
-0.4
-0.6
E h
[V]

The redox behaviour of Tc(VII)/Tc(IV) was investigated in 0.5–5.0 M NaCl and 0.25–4.5 M MgCl 2
solutions with [Tc(VII)] 0 = 10 –5 M in the presence of homogenous (Na 2 S 2 O 4 , Sn(II), Fe(II)/Fe(III), Fe
powder) and heterogeneous (magnetite, mackinawite, siderite) reducing systems. Samples were prepared,
stored and handled in Ar-glove boxes at 22±2°C. Technetium concentration, pH c and E h were monitored at
regular time intervals, and represented in Pourbaix diagrams calculated according to the NEA–TDB data
selection for Tc [1]. A decrease of [Tc] in the aqueous phase was interpreted as reduction of Tc(VII) to
Tc(IV) and subsequent precipitation of TcO 2 ∙xH 2 O(s). The redox state of Tc in the aqueous phase was further
analysed by solvent extraction techniques for selected samples. XANES and EXAFS were used to
characterize the redox state and molecular environment of Tc in the heterogeneous reducing systems
evaluated.
The solubility of Tc(IV) was studied in 0.5–5.0 M NaCl and 0.25–4.5 M MgCl 2 solutions under Aratmosphere
at 22±2°C. A Tc(VII) stock solution was electrochemically reduced to Tc(IV) and precipitated as
TcO 2 ⋅xH 2 O(s) in a Na 2 S 2 O 4 solution at pH c ∼12. About 5 mg of the resulting solid phase were added to each
independent batch experiment with the corresponding background electrolyte and reducing system (Na 2 S 2 O 4 ,
Sn(II) or Fe powder). Technetium concentration, pH c and E h were monitored with time. After reaching
equilibrium conditions, solid phases of selected batch experiments were characterized by XRD and
quantitative chemical analysis.
The redox distribution of Tc determined experimentally is in good agreement with the thermodynamically
calculated borderline for the chemical reaction TcO 4 − + 4H + +3e − ⇔ TcO 2 ∙xH 2 O(s) + (2−x) H 2 O (Fig.1), thus
indicating that E h and pH are robust and reliable parameters for the prediction of Tc redox behaviour in
homogeneous systems. Higher E h values are observed for samples in highly saline solutions (4.5 M MgCl 2 )
regardless of the reducing system; this was found to influence Tc redox chemistry consistently with the
thermodynamically calculated Tc(VII)/Tc(IV) borderline. In heterogeneous systems, EXAFS indicates the
predominance of Tc(IV) surface complexes and the absence of Tc solid phases.
Tc(IV) solubility data obtained in NaCl and MgCl 2 solutions reveal an increased [Tc] at 4 < pH c < 10. This
can be related to the predominance of TcOOH + and TcO(OH) 3 – aqueous species, respectively, in good
agreement with the current hydrolysis scheme selected in the NEA–TDB [1]. The pH–independent solubility
reaction TcO 2 ∙xH 2 O(am) ⇔ TcO(OH) 2 (aq) + (1–x) H 2 O is proposed to control solubility at 4 < pH c < 10.
Thermodynamic and activity models (SIT, Pitzer) for the system Tc 4+ –H + –Na + –Mg 2+ –OH – –Cl – are derived
based on the newly generated experimental solubility data.
[1] R. Guillaumont, T. Fanghänel, V. Neck, J. Fuger, D. A. Palmer, I. Grenthe and M. H. Rand, eds.
"Chemical Thermodynamics 5. Update on the Chemical Thermodynamics of Uranium, Neptunium,
Plutonium, Americium and Technetium". ed. E. NEA OECD(2003).
A1-4
LONG-TERM AQUEOUS ALTERATION KINETICS OF A 99 Tc-DOPED SON68 BOROSILICATE
GLASS
S. Rolland (1) , M. Tribet (1)* , M. Magnin (1) , V. Broudic (1) , S. Peuget (1) , A. Janssen (2) , T. Wiss (2) , C.
Jégou (1) and P. Toulhoat (3,4)
(1) CEA/DEN/DTCD/SECM/LMPA, Site de Marcoule, Bâtiment 166, B.P.17171, F-30207 Bagnols-sur-Cèze
(2) European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany
(3) CNRS/ISA, Institut des Sciences Analytiques, Université de Lyon, 5 rue de la Doua, F- 69100 Villeurbanne
(4) INERIS, Parc Technologique Alata, BP 2, F-60550 Verneuil en Halatte
*corresponding author
Fission products and minor actinides are currently immobilized by vitrification in a borosilicate glass known
as “R7T7”. One of the major functions of the glass matrix is to retain the radioactive elements in the event of
water intrusion in the geological disposal, expected after several thousand years. The alteration of SON68
glass (nonradioactive surrogate of R7T7 glass) has been extensively studied during the last thirty years and
the key mechanisms and kinetics of borosilicate glass alteration have been determined [1]: the first step
corresponds to a congruent dissolution characterized by the initial dissolution rate. Then in-situ

econdensation of dissolved species results in the formation of a gel that decreases the alteration rate by
several orders of magnitude to a residual alteration rate. The present work focuses on the study of the effects
of electronic dose rate, which would induce electronic excitations and ionisations, on this residual alteration
rate by studying a beta-doped glass containing 99 Tc. Moreover, the behaviour of a long-life isotope, 99 Tc
(T 1/2 = 2.1×10 5 years), which contributes to the long-term R7T7 glass activity, is considered.
A 0.16%wt 99 Tc-doped SON68 borosilicate glass is leached here under static conditions, in argon
atmosphere at 90°C and at a high surface-area-to-volume ratio (initial S/V = 25 cm -1 ), in order to reach
quickly the residual alteration rate. In addition, a reference experiment is performed by leaching a nonradioactive
SON68 borosilicate glass in the same conditions (initial S/V = 26 cm -1 , 90°C), but outside the
shielded cell.
The alteration rate is monitored by the releases of glass alteration tracer elements (B, Na and Li), which are
measured by ICP-AES. Most of the leachate samples are analyzed directly. The last sampling is filtered
(0.45 µm) and ultrafiltered (10 000 Dalton) to check for colloids. The releases in solution are expressed in
terms of normalized mass losses (NL in g.m -2 ) [2]. Radiation effects on the glass leached and its gel network
are characterised on raw grains by Scanning Electron Microscopy (SEM) analyses, and on ground altered
glass powder by Transmission Electron Microscopy (TEM) analyses. Technetium releases are also measured
by radiometry and its chemical oxidation state is assessed by measuring both pH and redox potential of the
leachate.
Results do not highlight any significant effect of beta irradiation on the residual alteration stage of this doped
glass: similar results have been observed on the reference experiment. These observations are consistent with
SEM and TEM characterizations, which show that a protective layer can be formed under beta irradiation.
Concerning the behaviour of technetium, as described in figure 1, congruence between boron and technetium
releases is observed, traducing the predominance of Tc(VII) in solution under these conditions (Eh = 380
mV/SHE, pH = 8 - 8,5).
Figure 1. Normalized mass losses of technetium versus time. Comparisons between direct, filtered and
ultrafiltered fractions. Comparison with boron normalized mass losses.
[1] P. Frugier, S. Gin, Y. Minet, T. Chave, B. Bonin, N. Godon, J.E. Lartigue, P. Jollivet, A. Ayral, L. De Windt, G. Santarini, J.
Nucl. Mater. 380, 8 (2008).
[2] S. Rolland, M. Tribet, P. Jollivet, C. Jégou, V. Broudic, C. Marques, H. Ooms and P. Toulhoat, J. Nucl. Mater. 433 (2013).

SESSION 2
A5: SOLID-WATER INTERFACE REACTIONS
WET CHEMISTRY AND SITE-SELECTIVE LUMINESCENCE SPECTROSCOPY STUDIES
ON THE UPTAKE OF HEXAVALENT ACTINIDES BY CEMENTITIOUS MATERIALS
J. Tits, T. Stumpf, C. Walther, E. Wieland (Switzerland, Germany)
THE BIGRAD CONSORTIUM - THE FATE OF TECHNETIUM AND URANIUM DURING
MAGNETITE CRYSTALLISATION AT HYPERALKALINE pH
T. A. Marshall, K. Morris, G.T.W. Law, J.F.W. Mosselmans, S. Shaw (UK)
SPECIATION OF NEPTUNIUM UPTAKE BY OPALINUS CLAY
T. Reich, S. Amayri , J. Drebert , D.R. Fröhlich , D. Grolimund , U. Kaplan (Germany,
Switzerland)
UNDERSTANDING THE MECHANISM OF Eu(III), Np(V), AND U(VI) SORPTION TO
HEMATITE USING VARIABLE TEMPERATURE BATCH REACTIONS
S.L. Estes, B.A. Powell (USA)
A5-1
A5-2
A5-3
A5-4
A5-1
WET CHEMISTRY AND SITE-SELECTIVE LUMINESCENCE SPECTROSCOPY STUDIES ON
THE UPTAKE OF HEXAVALENT ACTINIDES BY CEMENTITIOUS MATERIALS
J. Tits (1) , T. Stumpf (2) , C. Walther (3) , E. Wieland (1)
1) Paul Scherrer Institute, Laboratory for Waste Management, CH-5232 Villigen-PSI, Switzerland
2) Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal, P.O. Box 3640, D-76021
Karlsruhe, Germany
3) Institut für Radioökologie und Strahlenschutz, Leibniz Universität Hannover, D-30419 Hannover,
Germany
Cementitious materials are an important component in the multi-barrier concepts being developed in many
countries for the safe disposal of low and intermediate level radioactive waste in deep geological
repositories. In the past, studies on the retention of radionuclides by cementitious materials have focused
predominantly on adsorption as the most relevant uptake process (e.g., [1]). However, other potentially
important immobilization processes, such as incorporation in the solid matrix, may take place.. Calcium
silicate hydrates (C-S-H) are major constituents of cementitious materials. They are characterized by high
recrystallization rates and in recent years several studies have shown that incorporation is an important
retardation process in C-S-H phases for trivalent and tetravalent actinides (e.g., [2,3]).
In the present study, wet chemistry and luminescence spectroscopy experiments have been performed with
the aim of determining whether surface adsorption or incorporation is controlling the uptake of hexavalent
actinides (An(VI)), such as U(VI) and Np(VI), by C-S-H phases and hardened cement paste (HCP). Batch
sorption experiments were carried out on HCP, C-S-H phases, and titanium dioxide. (TiO 2 is a solid phase
which is stable under high pH conditions and is often used as a model material in surface complexation
studies.) Comparison of the sorption behaviour of An(VI) on TiO 2 and on cementitious materials allows the
influence of incorporation processes and surface complexation on immobilization to be determined. The
local coordination environment of U(VI) on TiO 2 , C-S-H phases and HCP was probed with luminescence
spectroscopy.
The pH dependence of the An(VI) sorption behaviour on TiO 2 and on C-S-H phases appeared to be nearly
2-
identical; R d values were found to decrease with the increasing predominance of AnO 2 (OH) 4 aqueous
species, showing that these actinyl anions are not taken up by these sorbents.
Luminescence spectroscopy investigations were conducted using both non-selective and site-selective laser
excitation at cryogenic temperatures (~15 K). Non-selective indirect laser excitation typically results in
poorly resolved U(VI) emission spectra with three to five broad bands corresponding to the coupling of the

O=U=O symmetric stretch vibration with the pure electronic charge transfer transition. These broad band
structures result from the superposition of many narrower bands associated with slightly different noninteracting
U(VI) luminescence centres in the nearly amorphous cementitious material and on the TiO 2
surface. Each of these U(VI) luminescence centres has a slightly different coordination environment since
the exact position of neighbouring atoms is not well defined. Under site-selective direct excitation, using a
tuneable laser as a narrow-band light source, the spectra of U(VI) sorbed on C–S–H phases exhibit detailed
fine structure (Fig. 1b). The spectrum of U(VI) sorbed on TiO 2 , however, remains poorly resolved (Fig. 1a)
due to the considerable energy transfer to neighbouring uranyl luminescence centres on the TiO 2 surface with
a slightly lower excitation energy. The smaller energy transfer effects in the site-selective luminescence
spectrum of U(VI) sorbed on C-S-H phases suggests a homogeneous distribution of the sorbed U(VI) over
the entire C-S-H matrix (i.e., incorporation), thus resulting in longer distances between neighbouring U(VI)
luminescence centres.
Evaluation of the highly resolved luminescence spectra of U(VI) sorbed on C–S–H phases recorded at
different excitation wavelengths allowed the identification and characterisation of two different U(VI) sorbed
species in different geometries (Spectrum b in Fig. 1). By comparison with the peak positions of the
luminescence spectra of U(VI) sorbed on TiO 2 , the first species was determined to be sorbed on the C-S-H
surface. The strong red-shift of the peak positions of the second species indicates that a considerably
different ligand field determines the fluorescence emission. This suggests a significantly different
coordination environment. Both the absence of energy transfer effects in the luminescence spectra of U(VI)
sorbed on C-S-H phases, and the presence of two sorbed species, one of which exhibits a strong red-shifted
luminescence spectrum, are strong indications for the incorporation of U(VI) in the C-S-H matrix.
The site selective luminescence spectrum of U(VI) sorbed on HCP (spectrum c in Fig. 1) exhibit peak
positions similar to those of U(VI) sorbed on C-S-H phases, confirming that both sorbed U(VI) species have
similar coordination environments in both sorbents.
This study thus shows that the uptake of hexavalent actinides by cementitious materials is governed by two
processes: surface adsorption and incorporation in the C-S-H matrix.
[1] Evans, N.D.M. (2008). “Binding mechanisms of radionuclides to cement.” Cem. Concr. Res. 38: 543-553.
[2] Tits, J., Stumpf, T., Rabung, T., Wieland, E. Fanghänel, T. (2003). „Uptake of Cm(III) and Eu(III) by calcium
silicate hydrates: A solution chemistry and time-resolved laser fluorescence spectroscopy study.” Environ. Sci.
Technol. 37: 3568-3573.
[3] Gaona, X., Dähn, R., Tits, J., Scheinost, A., Wieland, E. (2011). „Uptake of Np(IV) by C-S-H phases and cement
paste: an EXAFS study.” Environ. Sci. Technol. 45: 8765-8771.
Normalized intensity (A.U.)
Excitation
(a)
(b)
(c)
ν Ε
(2) ν S
(2)
ν E
(1) ν S
(1) ν S
(1)
Emission
TiO 2
C-S-H
HCP
22000 21000 20000 19000 18000 17000
Energy (cm -1 )
Figure 1: Excitation spectra and site-selective luminescence spectra (λ ex = 500 nm) of U(VI) sorbed on (a)
TiO 2 , (b) C-S-H phases, and (c) HCP at pH = 13.3. The U(VI) loading = 10 -3 mol kg -1 . ν E (1),
ν E (2) are charge transfer transitions of species 1 and 2. ν S (1) and ν S (2) are vibronic transitions
originating from the O=U=O symmetric stretch vibration.

A5-2
THE BIGRAD CONSORTIUM - THE FATE OF TECHNETIUM AND URANIUM DURING
MAGNETITE CRYSTALLISATION AT HYPERALKALINE pH
T. A. Marshall 1* , K. Morris 1 , G.T.W. Law 2 , J.F.W. Mosselmans 3 and S. Shaw 1
1) Research Centre for Radwaste and Decommissioning and Williamson Research Centre, School of Earth,
Atmospheric and Environmental Sciences, The University of Manchester, M13 9PL. UK.
2) Centre for Radiochemistry Research and Research Centre for Radwaste and Decommissioning, School of
Chemistry, The University of Manchester, M13 9PL. UK.
3) Diamond Light Source Ltd., Diamond House, Harwell Science and Innovation Campus, Didcot,
Oxfordshire, OX11 0DE. UK.
Geological disposal of legacy radioactive wastes stored at Earth's surface is now the predominant
management pathway for these materials and is a task of importance for nuclear power generating countries.
In many radioactive waste disposal scenarios, intermediate level wastes (ILW) are grouted and emplaced in a
geological disposal facility (GDF) which will have cement present as a ubiquitous engineering material.
Furthermore, cementitious materials have been considered as backfill for some GDF concepts. Therefore,
post-closure leaching of cementitious materials in a GDF is expected to create hyperalkaline conditions in
and around the repository, resulting in mineral alteration and crystallisation in the repository components and
host rock. Iron within the host rock derived from the alkaline breakdown of Fe-bearing silicate minerals (e.g.
biotite, chlorite), corrosion products formed within the repository, or iron flocs in the waste will form iron
(oxyhydr)oxide minerals. The formation and re-crystallisation of these reactive minerals may sequester
radionuclides through reduction to less soluble forms and/or incorporation into stable secondary iron oxide
phases, therefore they may prove key to the fate of radionuclides in such environments [1, 2].
To evaluate the significance of these processes, ferrihydrite was crystallised to magnetite (via addition of
Fe 2+ (aq)) under CO 2 -free, anoxic conditions in three synthetic cement leachates spiked with either Tc(VII) or
U(VI). The leachates represent early (pH 13.1), middle (pH 12.5) and late (pH 10.5) stage cement evolution
for a cementitious ILW GDF. After ageing, the headspace was flushed with CO 2 -free air to reoxidise for 21
days. Parallel experiments were set up compare the radionuclide sorption behaviour onto pre-formed
magnetite. These samples were allowed an equilibration period of 1 day prior to oxidation for 21 days.
Solution samples were taken throughout to determine the partitioning of Tc/U between the solid and solution
and X-ray Absorption Spectroscopy was used to characterise radionuclide associations with the solid phases
after ageing and oxidation.
All Tc(VII) was removed from solution rapidly (minutes) upon addition of the Fe 2+ (aq) with some oxidative
remobilisation after 21 days and with differences in the remobilisation extent with different chemical
conditions. Very little variation was observed between the EXAFS at the different pH conditions, and only
minor differences were observed between the reduced and oxidised samples even though significant
remobilisation of Tc to solution had occurred. Initial analysis of the EXAFS spectra show that some fraction
of the Tc is likely to be incorporated within the reduced mineral and that this phase may be resistant to
reoxidation. Similarly, all U(VI) was removed from solution during magnetite ageing, and analysis of the
EXAFS spectra suggested U(IV) incorporation into the reduced mineral phase. Upon reoxidation, some U
was remobilised to solution and again variable levels of remobilisation were seen in the different chemical
treatments.
These data suggest that Tc and U can become reduced and immobilised within the lattice of stable iron oxide
phases and that some fraction of the radionuclide is then recalcitrant to oxidative remobilisation. These data
will be discussed in the context of geochemical and spectroscopic data for these samples and in terms of the
wider significance of these processes in geological disposal and contaminated land.
[1] S. Kerisit, A.R. Felmy, and E.S. Ilton (2011). Atomistic simulations of uranium incorporation into iron (hydr)oxides.
Environmental Science and Technology 45; 2770-2776.
[2] W. Um, H-S Chang, J.P Icenhower, W.W. Lukens, R.J, Serne, N.P. Qafoku, J.H. Westsik, Jr., E.C. Buck, and S.C. Smith (2011).
Immobilisation of 99-technetium (VII) by Fe(II)-goethite and limited reoxidation. Environ. Sci. Technol. 45; 4904-4913.

A5-3
SPECIATION OF NEPTUNIUM UPTAKE BY OPALINUS CLAY
T. Reich (1) , S. Amayri (1) , J. Drebert (1) , D. R. Fröhlich (1) , D. Grolimund (2) , U. Kaplan (1)
(1) Institute of Nuclear Chemistry, Johannes Gutenberg-Universität Mainz, 55099 Mainz, Germany
(2)
Swiss Light Source, Paul Scherrer Institut, 5232 Villigen PSI, Switzerland
Argillaceous rock formations are under consideration as a potential host rock for the construction of highlevel
nuclear waste repositories. Under environmental conditions the most stable oxidation states of 237 Np
(t 1/2 =2.1 × 10 6 a) are Np(IV) and Np(V). We have investigated the sorption and diffusion of the more mobile
Np(V) in Opalinus Clay (OPA, Mont Terri, Switzerland) [1-3]. OPA, which is present in Switzerland and
southern Germany, possesses a micro-scale heterogeneity and is composed of several types of clay minerals,
but also of calcite, quartz and iron(II)-bearing minerals. In our previous diffusion [1] and anaerobic sorption
experiments [2], we observed higher distribution coefficients, K d , than expected from batch experiments
performed in air, indicating that a partial reduction of Np(V) to Np(IV) had occurred. To test this hypothesis,
different sorption and diffusion samples with Np(V) were prepared for spatially resolved molecular-level
investigations at the microXAS beamline at the Swiss Light Source (PSI, Villigen, Switzerland) [4]. By
combining synchrotron-based micro-X-ray fluorescence (µ-XRF) mapping, micro-X-ray absorption nearedge
structure spectroscopy (µ-XANES) and micro-X-ray diffraction (µ-XRD), the spatial distribution of
Np, its oxidation state and reactive mineral phases of OPA were determined.
Two kinds of samples, i.e. OPA thin sections and pieces from OPA bore cores, were prepared for microbeam
measurements. After establishing the minimum loading of Np that is necessary for detecting evaluable
signals by deposition and evaporation of an 85 µM Np(V) solution on OPA thin sections under ambient-air
conditions, thin sections were inserted into sorption cells and brought into contact with 8 µM Np(V) at pH 7
under anaerobic conditions. The solution was removed from the cell after a contact time of five days. Finally,
an intact OPA bore core was placed in a diffusion cell and preconditioned with synthetic OPA pore water
(pH 7.6, I=0.4 M) for more than six weeks under aerobic conditions. Then, the clay was contacted from one
side with 8 µM Np(V) in synthetic OPA pore water. After one month of Np diffusion, the clay core was
removed from the cell, dried and cut into small pieces for measurements at the SLS.
Elemental distributions of Ca, Fe and Np have been determined by µ-XRF mapping. As an example, Fig. 1
shows the elemental distributions of these elements on the aerobic sample prepared by diffusion of Np(V) in
an OPA bore core. Regions of high Np concentration were subsequently investigated by Np L III -edge µ-
XANES. In most samples Np spots with considerable amounts of tetravalent Np could be found, even when
the experiments were performed under ambient-air conditions. In some cases, almost pure Np(IV) L III -edge
XANES spectra were recorded (see Fig. 2). In case of the anaerobic sorption sample, a clear correlation
between Np and Fe was observed by µ-XRF, indicating that iron(II)-bearing minerals could be responsible
for the reduction of Np(V). µ-XRD measurements of this sample showed that pyrite is at least one of the
redox-active phases determining the speciation of Np in OPA. In this case, Np was accumulated on pyrite,
indicating that the reduction of Np occurred near the surface. By comparison of these spatially resolved
micro-beam investigations with Np L III -edge EXAFS measurements on OPA powder samples, we could
conclude that Np(IV) tends to agglomerate and that Np(V) is distributed more homogeneously on the clay
surface since the “hot spots” contain mostly Np(IV) [4]. In conclusion, our investigations showed that the
highly mobile Np(V) will be immobilized in Opalinus Clay by reduction to Np(IV), further consolidating the
suitability of argillaceous rocks with regard to the long-term storage of Np.
This work was financed by the Federal Ministry of Economics and Technology (BMWi, contract no.
02E10166) and ACTINET-I3 (contract no. 232631). D. R. Fröhlich was supported by a fellowship of DFG-
GRK 826.

Figure 1. Elemental distributions of Np, Fe and Ca measured by µ-XRF mapping on an aerobic sample
prepared by diffusion of Np(V) in an OPA bore core (2.5 × 2 mm; step size 20 µm; E=17.65 keV) [4].
Figure 2. Np L III -edge XANES spectrum collected on an anaerobic sorption sample prepared by contacting
an OPA thin section with 8 µM Np(V) solution at pH 7.6 together with the reproduction using reference
spectra of Np(IV) and Np(V) [4].
[1] T. Wu, S. Amayri, J. Drebert, L. R. Van Loon and T. Reich (2009). "Neptunium(V) sorption and diffusion in
Opalinus Clay." Environ. Sci. Technol. 43: 6567-6571
[2] D. R. Fröhlich. S. Amayri, J. Drebert and T. Reich (2011). "Sorption of neptunium(V) on Opalinus Clay under
aerobic/anaerobic conditions." Radiochim. Acta 99: 71-77
[3] D. R. Fröhlich, S. Amayri, J. Drebert and T. Reich (2012). „Influence of temperature and background electrolyte on
the sorption of neptunium(V) on Opalinus Clay.“ Appl. Clay Sci. 69: 43-49
[4] D. R. Fröhlich, S. Amayri, J. Drebert, D. Grolimund, J. Huth, U. Kaplan, J. Krause and T. Reich (2012). “Speciation
of Np(V) uptake by Opalinus Clay using synchrotron microbeam techniques.” Anal. Bioanal. Chem. 404: 2151-2162
A5-4
Understanding the Mechanism of Eu(III), Np(V), and U(VI) Sorption to Hematite using Variable
Temperature Batch Reactions
Shanna L. Estes, Brian A. Powell
Environmental Engineering and Earth Sciences, Clemson University, 342 Computer Court, Anderson, South
Carolina 29625, United States
Sorption to minerals and soils greatly influences the transport of actinides in the subsurface. However,
actinide sorption to mineral surfaces has not been greatly studied outside of 25 °C, despite the range of
temperatures expected in natural environments and the high temperature expected in future nuclear waste

epositories. Additionally, the mechanisms controlling actinide sorption to mineral surfaces are not well
understood, and examining sorption reactions at variable temperatures can provide valuable insight into these
mechanisms and their associated thermodynamics. In this work, we hypothesized that actinide sorption
reactions are driven by the entropically favorable displacement of solvating water molecules during the
formation of inner-sphere surface complexes. Our goal for this work was to combine variable-temperature
batch, spectroscopic, and modeling techniques in order to gain a better understanding of actinide sorption
mechanisms and thermodynamics.
We collected sorption edge data for Eu(III) at 15, 25, 35, and 50 °C and demonstrated that Eu(III) sorption to
hematite increased by approximately 24% over this temperature range. 1 We also determined from EXAFS
data and computational modeling that Eu(III) sorption to hematite was endothermic and resulted in the
formation of mononuclear bidentate surface complexes ((≡FeO) 2 Eu + ), where Eu(III) had a reduced
coordination number (5) compared to the free aqua ion (8). 1 These results suggested the loss of hydrating
waters from the Eu(III) coordination sphere during sorption.
We also studied Np(V) sorption to hematite over the same temperature range, and U(VI) sorption to hematite
at 15, 25, 35, 50, and 80 °C. Based on the effective charge on these ions, we would expect the sorption
affinity as a function of pH to increase in the order of Np(V), Eu(III), and U(VI), and indeed this trend was
observed. However, unlike Eu(III) sorption to hematite, U(VI) sorption increased by only approximately
16%, and the increase in Np(V) sorption with temperature was minimal. Additionally, and dissimilar from
Eu(III) sorption to hematite, surface complexation modeling of U(VI) sorption data required the use of two
surface species ((≡FeO) 2 UO 2 and (≡FeOH) 2 UO 2 2+ ) with varying relative concentrations depending on the
system temperature. Furthermore, EXAFS analyses available in the literature generally suggest that U(VI)
forms bidentate inner-sphere surface complexes to hematite 2 and ferrihydrite 3 with an overall coordination
number of 7 – 8 (including axial oxygens), representing a loss of two coordinating waters compared to
aqueous U(VI) species. Additionally, Arai and co-workers measured both bidentate inner-sphere surface
complexes and outer-sphere complexes for Np(V) sorption to hematite. 4 Therefore, these data suggest the
entropy change associated with sorption of U(VI) and Np(V) to hematite may be less favorable due to
displacement of fewer hydrating water molecules from these ions relative to Eu(III). Isothermal titration
calorimetry experiments are ongoing to verify our measured enthalpy and entropy values.
(1) Estes, Shanna L.; Arai, Yuji; Becker, Udo; Fernando, Sandra; Yuan, Ke; Ewing, Rodney C.; Zhang, Jiaming;
Shibata, Tomohiro; Powell, Brian A. A self-consistent model describing the thermodynamics of Eu(III) sorption to
hematite, submitted to J. Am. Chem. Soc.
(2) Bargar, John R.; Reitmeyer, Rebecca; Lenhart, John J.; Davis, James A. Characterization of U(VI)-carbonato
ternary complexes on hematite: EXAFS and electrophoretic mobility measurements. Geochim. Cosmochim. Acta 2000,
64, 2737 – 2749.
(3) Rossberg, André; Ulrich, Kai-Uwe; Weiss, Stephan; Tsushima, Satoru; Hiemstra, Tjisse; Scheinost, Andreas C.
Identification of uranyl surface complexes on ferrihydrite: Advanced EXAFS data analysis and CD-MUSIC modeling.
Environ. Sci. Technol. 2009, 43, 1400 – 1406.
(4) Arai, Yuji; Moran, P. B.; Honeyman, B. D.; Davis, J. A. In situ spectroscopic evidence for neptunium(V)-carbonate
inner-sphere and outer-sphere ternary surface complexes on hematite surfaces. Environ. Sci, Technol. 2007, 41, 3940 –
3944.

SESSION 3
A8: COMPUTATIONAL CHEMISTRY STUDIES
COMPUTATIONAL CHEMISTRY FOR HEAVY ELEMENTS: PRINCIPLES AND
APPLICATIONS OF DENSITY FUNCTIONAL THEORY
N. Kaltsoyannis (INVITED) (UK)
DENSITY FUNCTIONAL MODELING OF URANYL ADSORPTION ON SOLVATED 2:1
CLAY MINERALS
A. Kremleva, S. Krüger, N. Rösch (Germany)
ION SUBSTITUTION OF STRONTIUM AND NICKEL INTO CALCITE STUDIED BY
DENSITY FUNCTIONAL THEORY
M.P. Andersson, H. Sakuma, S.L.S. Stipp (Denmark, Japan)
RADIUM SOLUBILITY IN THE PRESENCE OF BARITE: SORPTION EXPERIMENTS
AND ATOMISTIC MODELLING
F. Brandt, M. Klinkenberg, V. Vinograd, K. Rozov, D. Bosbach (Germany)
A8-1
A8-2
A8-3
A8-4
A8-1
COMPUTATIONAL CHEMISTRY FOR HEAVY ELEMENTS: PRINCIPLES AND
APPLICATIONS OF DENSITY FUNCTIONAL THEORY
Nikolas Kaltsoyannis
Department of Chemistry, University College London, 20 Gordon Street, London WC1H 0AJ, UK
The computational chemistry of molecules and solids is nowadays dominated by studies employing density
functional theory (DFT). Although not perfect, this approach offers a computationally feasible and generally
reliable route to the calculation of many properties of interest, such as geometric and electronic structures,
reaction pathways and energetics, and a host of spectroscopic properties including vibrational, electronic and
spin resonance data. It is a particularly valuable tool for the investigation of actinide systems, where the
radioactivity of the transuranic elements precludes their experimental study at all but specialist nuclear
laboratories.
In this presentation I shall begin by describing the fundamentals of DFT and its most common
implementation via the Kohn-Sham approach. I shall then move on to illustrate its application to problems in
actinide and fission product chemistry. This will include a discussion of our recent work on the complexation
of the uranyl ion UO 2 2+ with the cellulose degradation products, α-isosaccharinate and D-gluconate [1, 2],
and also our studies of the molecular-level speciation of Sr 2+ with water [3] and our use of the periodic
electrostatic embedded cluster method to model the interactions of Sr 2+ /water complexes with the surface of
mineral brucite (Mg(OH) 2 ). The approach we are developing to study Sr/brucite is in principle applicable to
the interactions of radiotoxic species with mineral surfaces throughout the geosphere.
[1] K. H. Birjkumar, N. D. Bryan and N. Kaltsoyannis “Computational investigation of the speciation of uranyl
gluconate complexes in aqueous solution” Dalton Transactions 40 (2011) 11248.
[2] K. H. Birjkumar, N. D. Bryan and N. Kaltsoyannis “Is gluconate a good model for isosaccharinate in uranyl(VI)
chemistry? A DFT study” Dalton Transactions 41 (2012) 5542.
[3] A. Kerridge and N. Kaltsoyannis “Quantum chemical studies of the hydration of Sr 2+ in vacuum and aqueous
solution” Chemistry, A European Journal 17 (2011) 5060.

A8-2
DENSITY FUNCTIONAL MODELING OF URANYL ADSORPTION ON SOLVATED 2:1 CLAY
MINERALS
Alena Kremleva, Sven Krüger, Notker Rösch
Department Chemie & Catalysis Research Center, Technische Universität München, 85748 Garching,
Germany
Clays are considered as host rock formations for highly radioactive waste as they are regarded as natural and
technical barriers against actinide distribution in the environment. Adsorption of actinides on clay minerals,
which represent the dominant fraction of the mineral content of clay rocks, is an efficient retardation
mechanism besides precipitation or absorption. Thus, an understanding of actinide adsorption on clay
mineral surfaces at the atomic level is an important prerequisite for a thorough modeling of actinide
distribution in the environment. In this context, systematic studies of the interaction of solvated actinide ions
with clay mineral surfaces by quantum chemistry methods is a useful approach to speciation and energetics.
We examined uranyl adsorption on clay minerals at the density functional level with the plane-wave based
projector augmented wave (PAW) approach as implemented in the program VASP. Mineral surfaces were
modeled with the periodic supercell approach; surface solvation was approximated by about 2-3 layers of
adsorbed water molecules 1,2 . To start, we characterized the reactive edge surfaces of the neutral 2:1 clay
mineral pyrophyllite and created the corresponding surfaces of charged model minerals by cation
substitution. The permanently charged clay minerals beidellite and montmorillonite were represented in this
way by model structures with substitutions in the tetrahedral and octahedral layers, respectively. A
permanent charge of 0.5 e per unit cell was chosen, balanced by solvated Na + counter ions in the interlayer
space. We explored the solvation of various edge surfaces, their relative stability, and the effect of the
charged defects. We calculated lower surface energies for surfaces with substitutions below the surface than
for those with substitutions directly on the surface.
For all model minerals we compared bidentate uranyl(VI) adsorption on various edge surfaces. Two main
types of adsorption sites were studied: (i) aluminol sites involving only AlO(H) surface groups and (ii)
mixed sites composed of aluminol and silanol groups, jointly coordinating uranyl. As long as they are not on
the surface, substitutions do not affect the adsorption sites of uranyl. Due to their various surface groups, the
types of adsorption complexes determined depend on the orientation of the edge surfaces. Structural
parameters of the adsorbed species are very similar for adsorption on neutral pyrophyllite and charged clay
minerals when their substitutions lie below the surface. Noticeable variations appear only when the
substitution modifies the adsorption site, e.g., MgOOH instead of AlOOH on montmorillonite or AlO-AlO
instead of SiO-AlO on beidellite. This finding agrees with our previous result that U-O bond lengths in the
equatorial plane to the surface correlate with the formal charge of the corresponding surface O center. In
agreement with this paradigm, significant structural effects are lacking in case of sub-surface substitutions
which do not affect the formal charges of surface O centers. In contrast, substitutions on the surface lower
the formal charges of surface O centers and therefore, in most cases, lead to shorter U-O bonds to the
surface.
Energetically we characterized uranyl adsorption by surface complex formation energies, formally calculated
as energies of the exchange of two surface protons by the uranyl ion. Direct comparison of formation
energies for uranyl adsorbed on pyrophyllite, beidellite, and montmorillonite do not show a clear preference
of specific adsorption sites. To determine uranyl adsorption energies, we estimated relative deprotonation
energies of surface groups and used experimental values of the pH of zero point charge as reference.
Preliminary results on (010) and (100) surfaces suggest that uranyl adsorption on pyrophyllite and beidellite
takes place only at rather high pH, presumably on mixed aluminol-silanol adsorption sites. Adsorption of
uranyl on montmorillonite occurs already at low pH on aluminol sites. With increasing pH also other types
of sites become available for adsorption.
1
B. Martorell, A. Kremleva, S. Krüger, N. Rösch. J. Phys. Chem. C 2010, 114, 13287
2
A. Kremleva, S. Krüger, N. Rösch. Geochim. Cosmochim. Acta 2011, 75, 706

A8-3
ION SUBSTITUTION OF STRONTIUM AND NICKEL INTO CALCITE STUDIED BY DENSITY
FUNCTIONAL THEORY
M.P. Andersson 1)* , H. Sakuma 1,2) , S.L.S. Stipp 1)
1)
Nano-Science Center, Department of Chemistry, University of Copenhagen, DK-2100 Copenhagen,
Denmark
2) Department of Earth and Planetary Sciences, Graduate School of Science and Engineering, Tokyo
Institute of Technology, 2-12-1 Ookayama, Meguro-ku, Tokyo 152-8551, Japan.
Calcite, the rhombohedral polymorph of CaCO 3 , is a common mineral in the Earth’s crust, present in soil,
sediments and rocks and it readily forms as a secondary phase in the high pH waters associated with the
breakdown of concrete, such as would happen in an aging radioactive waste repository. Divalent cations,
with atomic radius equivalent to or smaller than calcium, often substitute into calcite with little or no
disruption of the atomic structure and its rhombohedral lattice can tolerate a certain amount of larger divalent
cations before the orthogonal polymorph, aragonite, becomes the stable phase. Thus, calcite is a possible
sequestering agent for nickel and strontium, either through uptake during precipitation or as a result of
recrystallisation of existing material in dynamic equilibrium. Experimental thermodynamic and kinetic data
are essential for risk assessment modelling and to predict the transport of radioactive Ni 2+ and Sr 2+ but for
some conditions, experiments are difficult, if not impossible. Computational approaches are invaluable for
filling in missing data and for providing insight into the molecular scale processes responsible for uptake and
release. We have used plane-wave density functional theory of a slab of calcite with the {10.4} cleavage
termination to study the thermodynamics of ion substitution into the surface and bulk, for Sr 2+ and Ni 2+ . The
reaction is investigated using models for hydrated ions and hydrated calcite surfaces that are effective for ion
exchange of a calcite Ca 2+ ion for another cation, M 2+ , according to the reaction:
M(H 2 O) 6 2+ + calcite-Ca (hydrated) -> calcite-M (hydrated) + Ca(H 2 O) 6
2+
calcite-Ca (hydrated) is a slab model of a calcite {10.4} surface with a monolayer of water (Fig. 1) and
calcite-M is the same slab with one Ca replaced by M. Our model for hydrated divalent cations includes the
first hydration shell, assumed to consist of six water molecules. We obtain a linear relationship with a slope
very close to 1, between the binding enthalpy of six water molecules to divalent ions and the free energy of
hydration of that ion, which gives confidence that our calculated energy differences between hydrated
divalent cations reflect free energy differences in aqueous solution.
Our calculations show that that Sr 2+ uptake is endothermic both in surface and bulk sites of calcite. Ni 2+
uptake on the other hand is exothermic both in surface sites and bulk sites, with surface sites being more
stable than bulk sites. This is consistent with experimental data[1,2] showing that calcite acts as a sink for
Ni 2+ and although Sr 2+ does substitute for Ca 2+ , the quantity is small.
Figure 1. Surface slab model used for the hydrated calcite surface
[1] J.M. Zachara, C.E. Cowan, C.T. Resch, Geochimica Et Cosmochimica Acta, 55 (1991) 1549-1562.
[2] U. Hoffmann, S.L.S. Stipp, Geochimica Et Cosmochimica Acta, 65 (2001) 4131-4139.

A8-4
RADIUM SOLUBILITY IN THE PRESENCE OF BARITE: SORPTION EXPERIMENTS AND
ATOMISTIC MODELLING
F. Brandt 1) , M. Klinkenberg 1) , V. Vinograd 2) , K. Rozov 1) , D. Bosbach 1)
1) Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research – Nuclear Waste
Management and Reactor Safety (IEK-6), 52425 Juelich, Germany
2) Institute of Geosciences, Goethe University, Frankfurt (Germany)
Phase relations in the BaSO 4 -RaSO 4 -H 2 O system are important for the understanding of the role of baritetype
minerals in controlling the solubility of radium in natural waters. A number of recent studies were
concerned with the assessment of radium mobility in ground waters, which could come into contact with
radioactive waste [1, 2]. It has been argued that the radium concentration in the aqueous phase will be
controlled by the solubility of Ra-bearing sulfates, which will form due to chemical reactions between the
waste and the sulfate in the ground water. Pure RaSO 4 is not a likely phase to be formed in this process. It is
expected that radium will form a solid solution with barite due to its chemical similarity with barium. Barite
occurs as a trace constituent in clay formations e.g. the Opalinus Clay formation, which is currently
investigated in Switzerland as a host rock for the disposal of spent nuclear fuels. The composition of pore
waters extracted from the Opalinus Clay is close to barite saturation [3]. Thus, a realistic scenario for the
corrosion of a spent fuel disposed in a clay formation should consider an aqueous phase, which is saturated
or nearly saturated with respect to Ba-sulfate. However, due to a lack of reliable data, the solid solution
system RaSO 4 -BaSO 4 -H 2 O is currently not considered in long term safety assessments for nuclear waste
repositories. For instance, the solubility of the pure RaSO 4 endmember is poorly constrained between
logK RaSO4 = -10.26 to -10.41 by only very few experimental data [4, 5]. Furthermore, available literature data
for the interaction parameter W BaRa , which describes the non-ideality of the solid solution, vary by about one
order of magnitude [6, 7]. The final radium concentration in this system is extremely sensitive to the amount
of barite, the difference in the solubility products of the end-member phases, and the degree of non-ideality
of the solid solution phase.
We have combined a macroscopic experimental approach with atomistic calculations and thermodynamic
modeling to study in detail how a radium containing solution will equilibrate with solid BaSO 4 under
repository relevant conditions. Batch recrystallization experiments were carried out with an initial Ra/Ba
ratio of 0.3 (5 . 10 -6 mol/L Ra) at near-neutral pH. At room temperature, a significant decrease of the radium
concentration to 3.5 . 10 -9 to 7.0 . 10 -9 mol/L (solid/liquid ratio: 5 g/L) occurred within the first 70 days of the
experiment. At 90 °C and the same solid/liquid ratio, a faster decrease of the radium concentration in
solution is observed compared to room temperature. Here, an apparent equilibrium is reached at 3 . 10 -8 mol/L
of radium after 30 days. The decrease of the radium concentration is apparently not related to the specific
surface area of the barite crystals. First principle calculations indicate that a regular solid solution model can
be applied. The thermodynamic parameters for the solid solution have been derived from the change in total
energy of a 2×2×2 supercell of barite due to the insertion of a single substitutional defect of radium into the
barite structure. The computed value of W BaRa = 2.50 ± 1.00 kJ/mol implies a non-ideal solid solution. The
results of thermodynamic modeling indicate a good agreement of the apparent final Ra (aq) equilibrium
concentration from experimental data at RT and 90°C with the lower limit of the computed W BaRa (1.50
kJ/mol) and a solubility of the RaSO 4 end member of logK RaSO4 = -10.41. These thermodynamic parameters
for the solid solution can now be applied in the safety assessment for direct disposal of spent fuel elements.
[1] Berner, U. and Curti, E., 2002. Internal report TM-44-02-04. Paul Scherrer Institute, Villigen, Switzerland.
[2] Grandia, F., Merino, J., and Bruno, J., 2008. Technical Report TR-08-07. Svensk Kärnbränslehantering AB.
Swedish Nuclear Fuel and Waste Management Company. Sweden.
[3] Pearson, F. J. , Arcos, D. , Bath, A., Boisson, J.-Y., Fernández, A. M., Gäbler, H.-E., Gaucher, E., Gautschi, A.,
Griffault, L., Hernán, P., and Waber, H. N. Mont Terri Project, 2003. Reports of the FOWG, Geol, Series No. 5, Bern.
[4] Lind, S. C., Underwood, J. E., and Whittemore, C. F., 1918. J Am Chem Soc 40, 465-472.
[5] Paige, C. R., Kornicker, W. A., Hileman, O. E., and Snodgrass, W. J., 1998. Geochim. Cosmochim. Acta 62, 15-23.
[6] Zhu, C., 2004. Geochim. Cosmochim. Acta 68, 3327-3337.
[7] Curti, E., Fujiwara, K., Iijima, K., Tits, J., Cuesta, C., Kitamura, A., Glaus, M. A., and Muller, W., 2010. Geochim.
Cosmochim. Acta 74, 3553-3570.

SESSION 4
B2: DIFFUSION AND OTHER MIGRATION
PROCESSES
ASSESSING THE SORPTION PROPERTIES OF CANADIAN SEDIMENTARY ROCKS
UNDER SALINE CONDITIONS
P. Vilks, N.H. Miller, T. Yang (Canada)
GAS INDUCED RADIONUCLIDE TRANSPORT IN DISTURBED AND UNDISTURBED
BOOM CLAY
E. Jacops, T. Maes, N. Maes, E. Weetjens, G. Volckaert (Belgium)
DIFFUSION OF 85 Sr 2+ IN COMPACTED MONTMORILLONITE SEEMINGLY AGAINST ITS
OWN CONCENTRATION GRADIENT
M. A. Glaus, J. Eikenberg , S. Frick , M. Rüthi , L.R. Van Loon (Switzerland)
B2-1
B2-2
B2-3
B2-1
ASSESSING THE SORPTION PROPERTIES OF CANADIAN SEDIMENTARY ROCKS UNDER
SALINE CONDITIONS
Peter Vilks 1) , Neil H. Miller 1) , Tammy Yang 2)
1) Atomic Energy of Canada Limited, Whiteshell Laboratories, Pinawa, Manitoba, Canada, R0E 1L0
2)
Nuclear Waste Management Organization, 22 St. Clair Avenue East, 6 th Floor, Toronto, Ontario, Canada,
M4T 2S3
The Nuclear Waste Management Organization’s (NWMO) Adaptive Phased Management (APM) Technical
Program is intent on advancing the understanding of solute migration in deep crystalline and sedimentary
groundwater systems. In these systems, groundwaters may contain brines with total dissolved solid (TDS)
concentration up to 200-375 g/L. As part of NWMO’s Technical Program, a database of sorption coefficients
(K d ) for Canadian sedimentary rocks in saline solutions has been developed. The initial database was
compiled from the literature and augmented with sorption values from sorption tests [1, 2]. This work
describes both batch and mass transport sorption tests conducted in a reference Na-Ca-Cl brine solution with
TDS of 275 g/L for sedimentary rocks.
Batch sorption experiments to measure specific sorption coefficients for elements Ni, Cu, Pb, Zr and U in the
reference brine solution were conducted using shale, limestone and bentonite to i) augment the sorption
database; and ii) improve understanding of sorption mechanisms in highly saline solutions. The sorption tests
were conducted using single elements U(VI) and Zr(IV), as well as multiple elements Ni(II), Cu(II), Pb(II)
and U(VI). Batch sorption tests were also performed with a dilute reference solution (TDS = 0.2 g/L), to
provide a reference case where sorption would not be suppressed by the presence of concentrated salt
solutions. The experimental results demonstrated that the sorption coefficients of Ni(II), Cu(II) and Pb(II)
were lower in the reference brine by factors of 20 to 300 in comparison to those measured in the dilute
solution, while K d values for U(VI) was lower by factors of 1.5 to 8. Assuming that montmorillonite and
illite could be used to approximate bentonite and shale, sorption modelling was performed for these minerals
using a 2-site protolysis non-electrostatic surface complexation and cation exchange model. In the brine,
where surface complexation is the dominant sorption mechanism, simulated K d values for Ni and Pb were
identical to measured values, while simulated K d values for Cu and Zr were within measured values by
factors of 1.5 and 1.9, respectively. In dilute reference solution, where cation exchange is the dominant
sorption mechanism, simulated K d values for Ni, Cu and Pb were within a factor of 3 to 4 of the measured
values.
Mass transport tests for a low permeability (3x10 -21 m 2 ) shale were performed with both advective and
diffusive mass transport methods using the reference brine solution to investigate how the measured sorption
coefficients compare to values measured using batch sorption tests. The sorbing tracers included Ni, Cu, Pb

and U, with Li being used as a conservative tracer. A through-diffusion test was performed to study sorption
during mass transport by diffusion only. Tracer diffusion was characterized by monitoring tracer
concentrations that diffuse through the shale sample and by determining tracer diffusion profiles within the
rock at the end of the diffusion test. An advective transport test with induced hydraulic flow was performed
using the High Pressure Radioisotope Migration (HPRM) apparatus [1] in an attempt to obtain transport
information on a shorter time scale than possible with a diffusion test. The results of both advective and
diffusive mass transport tests were interpreted with compartment models set up within the AMBER
modelling environment, and designed to simulate the geometry of both tests. The K d values derived from
both the advective transport and the through-diffusion tests were consistent with K d values determined by
batch tests.
[1] Vilks, P. 2011. Sorption of Selected Radionuclides on Sedimentary Rocks in Saline Conditions – Literature Review,
Nuclear Waste Management Organization technical report, NWMO TR-2011-12, Toronto, Canada.
[2] Vilks, P., N. H, Miller and K. Felushko. 2011. Sorption Experiments in Brine Solutions with Sedimentary Rock and
Bentonite, Nuclear Waste Management Organization technical report, NWMO TR-2011-11, Toronto, Canada.
B2-2
GAS INDUCED RADIONUCLIDE TRANSPORT IN DISTURBED AND UNDISTURBED BOOM
CLAY
E. Jacops, T. Maes, N. Maes, E. Weetjens, G. Volckaert
Belgian Nuclear Research Centre, SCK•CEN, Institute for Environment, Health and Safety, Boeretang 200,
B-2400 Mol, Belgium
The Belgian agency for radioactive waste and enriched fissile materials ONDRAF/NIRAS presently
considers Boom Clay as a potential host formation for the disposal of high-level and long-lived radioactive
waste. In argillaceous formations like Boom Clay, radionuclide transport is dominated by solute diffusion.
However, other phenomena like gas-induced transport may also occur, but are poorly described and
quantified [2]. The present study focusses on the potential of gas-induced radionuclide transport after gas
breakthrough in a clayey host rock (disturbed and undisturbed conditions) and at interfaces with engineered
barriers. It was performed within the framework of the FP7 project FORGE (Fate of Repository Gasses).
In geological repositories for radioactive waste, gas can be generated by different mechanisms like anaerobic
corrosion of metals, radiolysis of water and organic materials and microbial degradation of various organic
wastes. The gas generated in the waste and engineered barriers will dissolve in the pore water and will be
transported away from the repository by diffusion as dissolved species. However if the gas generation rate is
larger than the capacity for diffusive transport of dissolved gas, the pore water will get oversaturated and a
free gas phase will be formed, leading to a potential gas pressure build-up [1]. In case the gas pressure would
exceed a local threshold value (e.g. tensile strength of concrete barriers or lithostatic stress in the Boom Clay
host formation) this could lead to formation of discrete pathways. During a local gas breakthrough event in
the clay, some water could be expelled by the gas phase. Depending on the timing of gas breakthrough,
dissolved radionuclides and contaminants could be driven out of the clay faster than the normally expected
diffusive transport.
To test the potential for gas-driven radionuclide transport, a column experiment was designed in which a
water saturated clay core (h~4cm, Ø= 3.8 cm) was put directly on top of a thin BC core (h~1cm) which had
been previously saturated with a tracer solution. To mimic the presence of radionuclides in the pore water, a
non-radioactive NaI solution of 0.01 mol/l was used as tracer solution. A He gas pressure (P~0.5-5 MPa) was
applied at the bottom end of the iodide saturated plug while at the top end of the column, a known volume of
natural pore water was put in contact with the plug. The gas pressure was stepwise increased until gas
breakthrough occurred. Upon gas breakthrough, the water on top of the column was expelled and analysed
for its iodide content. The measured concentration of I was linked to the amount of NaI saturated pore water
that was displaced, taking into account the natural background concentration of I and the transport of I due to
diffusion [2]. Different types of experiments have been performed: experiments with undisturbed and
disturbed (artificially fissured and let to self-seal for 1 night or 1 week) Boom Clay samples with different

orientation wrt. bedding plane (parallel or perpendicular to the bedding plane). To investigate the role of
interfaces, experiments on combined Boom Clay – bentonite samples were also conducted.
Figure 1: Basic concept of the tests to study gas driven tracer transport in the interface Boom Clay –
Bentonite. Blue = NaI saturated Boom Clay core. Yellow = half Boom Clay core. Pink = half bentonite core.
Figure 2: Average % of desaturation for different sample types
Based on the obtained results, we can state that the transport of radionuclides and contaminants due to a gas
breakthrough is indeed possible but remains very limited. Figure 2 shows that the amount of I that was
transported and consequently the degree of desaturation due to gas breakthrough was very low (< 0.5%). The
effect of different parameters (orientation of the sample with respect to bedding plane, time of sealing of the
fractures and porous medium type) on the degree of desaturation was investigated, but only orientation wrt.
bedding plane played a significant role. This could be explained by the plate-structure of the clay: parallel
oriented clay plates can more easily re-join and sealing will be more efficient. Another significant parameter
is the location of gas breakthrough: in 53% of the experiments, gas breakthrough pathways were observed at
the interface clay - cell. In only 16% of the experiments gas breakthrough pathways were going through the
clay core. So gas did find the weakest path which was in most cases the interface clay – cell. Transport
through combined Boom Clay – bentonite samples is comparable to transport through disturbed Boom Clay.
Finally, it is important to note that during the experiments, observations of gas flow rather indicated another
flow mechanism than classical visco-capillary 2-phase flow.
[1] L. Yu., E. Weetjens (2009). "Summary of gas generation and migration – Current State-of-the-Art". SCK•CEN
report (ER-108).
[2] E. Jacops, N. Maes, G. Volckaert, J. Govaerts, T. Maes (2012). "Results and interpretation of gas-driven
radionuclide transport in disturbed and undisturbed Boom Clay and Boom Clay – bentonite interfaces". SCK•CEN
report (ER-222).
B2-3

DIFFUSION OF 85 Sr 2+ IN COMPACTED MONTMORILLONITE SEEMINGLY AGAINST ITS
OWN CONCENTRATION GRADIENT
M. A. Glaus (1) , J. Eikenberg (2) , S. Frick (1) , M. Rüthi (2) , L.R. Van Loon (1)
(1) Laboratory for Waste Management, Paul Scherrer Institut, 5232 Villigen PSI, Switzerland.
(2) Radioanalytics, Paul Scherrer Institut, 5232 Villigen PSI, Switzerland
Enhanced diffusion rates of many cations compared to uncharged tracers in compacted smectite clays is well
documented [1, 2], albeit still controversially discussed phenomenon. Surface diffusion combined with
aqueous phase diffusion or diffusion in different pore types of the clay have often been invoked as
explanations [3]. However the magnitude of the driving force and the assignment of parameter values remain
arbitrary to a large degree because the individual fluxes cannot be observed independently.
Here we report the seeming “uphill diffusion” of a 85 Sr 2+ tracer in compacted sodium montmorillonite, i.e.
transport in the direction from a low to a high tracer concentration reservoir. The classical set-up for tracer
through-diffusion experiments has two sides of a compacted clay sample in contact with identical electrolyte
solutions, whereby one of these reservoir solutions contains a radioactive isotope. In contrast the setup for
the present experiments uses a clay sample equilibrated with two different electrolyte solutions, but
containing the same amounts of the radioisotope on both sides. These initial conditions lead to a subsequent
decrease of the tracer concentration in both reservoirs because of the uptake of the cationic radiotracer by the
clay. However, after a given time, the tracer concentration starts to increase in the reservoir with the higher
background electrolyte concentration, while it continuously decreases in the other reservoir. Consequently
the tracer diffuses in a seeming “uphill manner” from lower to higher tracer concentration.
The effect can be explained as follows. Owing to a residual mobility of tracer cations bound to the basal
exchange sites present in the interlayer porosity, not only the tracer concentration differences in the external
aqueous phase, but also those with respect to the tracer concentration in the interlayer porosity are
determining the magnitude and direction of diffusion. The conditions of the experiments were chosen in such
a way that the gradients in the aqueous phase and in the interlayer porosity were opposite (cf. Figure 1). In
view of the observed seeming “uphill fluxes” of the tracer i.e. against the gradient in the aqueous phase, it
can be concluded that the concentration gradient of 85 Sr 2+ in the interlayer porosity is the dominant driving
force for diffusion under the experimental conditions.
We observed such phenomena in similar earlier experiments with a 22 Na + tracer [4] and clay samples
compacted to a bulk dry density of 1900 kg m -3 . The present experiments with 85 Sr 2+ were carried out at a
bulk dry density of 1300 kg m -3 . The use of a bi-valent cationic tracer results in steeper concentration
gradients in the interlayer porosity, and the lower bulk dry density leads to generally higher diffusivities.
These circumstances lead to an increased dynamical behaviour of the system compared to the experiments
with the 22 Na + tracer. Thus the “experimentally accessible window of observation” is increased. A
measurable change in the background electrolyte concentration was observed accordingly, which in turn had
a clear effect on the diffusive behaviour of the tracer. A model in which the transport of the tracer is coupled
to the diffusive transport of the electrolyte enabled the underlying processes to be quantitatively described.
The experiments clearly demonstrate that the physical transport behaviour of tracer cations in smectite clays
cannot be described independently of the chemical behaviour of these species.

Figure 1. Through-diffusion experiment under salt gradient conditions. The clay sample (1) is in contact
with a low-salinity reservoir (2) and a high-salinity reservoir (3) via a circulation system. The blue colours
represent the concentrations of the background electrolyte; the red bars in the upper part of the figure
indicate that the experiment started with equal initial concentrations of the 85 Sr 2+ tracer in the two reservoirs.
The solution concentrations of the background electrolyte and the tracer species in a later steady-state flux
situation with a maintained constant concentration difference of the background salt are shown in the lower
scheme. Owing to the differences in competition between Na + and Sr 2+ on both sides of the cell, an internal
non-zero gradient of the tracer cations is induced. Note that this gradient is opposite to that of the
background ions.
1. Melkior, T.; Yahiaoui, S.; Thoby, D.; Motellier, S.; Bartès, V., Diffusion coefficients of alkaline cations in Bure
mudrock. Phys. Chem. Earth 2007, 32, 453–462.
2. Glaus, M. A.; Frick, S.; Rossé, R.; Van Loon, L. R., Comparative study of tracer diffusion of HTO, 22 Na + and 36 Cl –
in compacted kaolinite, illite and montmorillonite. Geochim. Cosmochim. Acta 2010, 74, (7), 1999–2010.
3. Bourg, I. C.; Sposito, G.; Bourg, A. C. M., Modeling the diffusion of Na+ in compacted water-saturated Nabentonite
as a function of pore water ionic strength. Appl. Geochem. 2008, 23, 3635–3641.
4. Glaus et al., in preparation

SESSION 5
POSTER SESSION I
PA1
PA1-1
PA1-2
PA1-3
PA1-4
PA1-5
PA1-6
PA1-7
PA1-8
PA1-9
PA1-10
PA1-11
PA1-12
PA1-13
SOLUBILITY AND DISSOLUTION
GASEOUS RELEASE OF CARBON-14 FROM IRRADIATED MATERIALS
G.M.N. Baston, N.A. Hodge, S.J. Williams (UK)
ELICITATION OF DISSOLUTION RATE DATA FOR POTENTIAL WASTEFORM
TYPES FOR PLUTONIUM UNDER REPOSITORY CONDITIONS
G. Deissmann, S. Neumeier, F. Brandt, G. Modolo, D. Bosbach (Germany)
SOLUBILITY AND TRLFS STUDY OF Nd(III) AND Cm(III) IN DILUTE TO
CONCENTRATED ALKALINE NaCl-NaNO 3 AND MgCl 2 -Mg(NO 3 ) 2 SOLUTIONS
M. Herm, X. Gaona, Th. Rabung, C. Crepin, V. Metz, M. Altmaier, H. Geckeis (Germany,
France)
SOLUBILITY AND HYDROLYSIS OF U(VI) AT 80°C UNDER ACIDIC TO
HYPERALKALINE PH CONDITIONS
X. Gaona, M. Marques, B. Baeyens, M. Altmaier (Germany, Switzerland)
EFFECTS OF SURFACE MORPHOLOGY ON DISSOLUTION OF ThO 2
E. Myllykylä, T. Lavonen, K. Ollila (Finland)
THE EFFECT OF THE CONCENTRATION OF SILICATES ON THE SOLUBILITY OF
Pu(IV) IN SODIUM BICARBONATE/CARBONATE SOLUTIONS
T. Yamaguchi, K. Henmi, Y. Iida, H. Okamoto, T. Tanaka (Japan)
LEACHING TEST OF CeO 2 UNDER GROUNDWATER CONDITIONS
N. Rodriguez, J. Cobos, E. Iglesias, C. Palomo, J. Nieto, J. M. Cobo, L. Serrano, S. Durán, J.
Quiñones (Spain)
INFLUENCE OF PARTICLE SIZE, CARBONATES AND ORGANIC MATTER ON THE
SOLUBILITY OF ThO 2 (cr)
S. Salah, D. Liu, L. Wang (Belgium, China)
ADOPT PELLET LEACHING PROPERTIES, A COMPARISON WITH UO 2 PELLET
K. Nilsson, O. Roth, M. Jonsson (Sweden)
Np(V) SOLUBILITY IN DILUTE TO CONCENTRATED MgCl 2 SOLUTIONS
V.G. Petrov , X. Gaona , D. Fellhauer , J. Rothe, K. Dardenne, S.N. Kalmykov, M. Altmaier
(Russia, Germany)
THE DISPOSAL OF SPENT NUCLEAR FUEL: THE EFFECT OF SURFACE DEFECTS
ON DISSOLUTION RATE
C.L. Corkhill, J.W. Bridge, P. Hillel, L.J. Gardner, M.C. Stennett, R. Tapper, N. C. Hyatt
(UK, USA)
THE EFFECT OF ADVA CAST 551 SUPERPLASTICISER ON RADIONUCLIDE
SOLUBILITY
R. Beard, A.P. Clacher , M.M. Cowper (UK)
ELICITATION OF URANIUM SOLUBILITY TO SUPPORT THE DISPOSAL OF U 3 O 8
T. Beattie, M. Couch, C.P. Jackson (Switzerland, UK)

PA1-1
GASEOUS RELEASE OF CARBON-14 FROM IRRADIATED MATERIALS
G.M.N. Baston 1 , N.A. Hodge 2 and S.J. Williams 3 *
1 AMEC, Building 150, Harwell Oxford, Didcot, OX11 0QB, UK
2 National Nuclear Laboratory, Preston Laboratory A709, Springfields, Preston, PR4 0XJ, UK
3 NDA Harwell Office, Building 587, Curie Avenue, Harwell Oxford, Didcot, OX11 0RH, UK
* corresponding author Steve.Williams@nda.gov.uk
Carbon-14 is a key radionuclide in the assessment of the safety of a geological disposal facility for
radioactive waste because of the calculated radiological consequences of gaseous carbon-14 bearing species.
It may be that such calculations are based on overly conservative assumptions and that better understanding
could lead to considerably reduced radiological consequences from these wastes. For a waste stream
containing carbon-14 to be an issue:
• there must be a significant inventory of carbon-14 in the waste stream; and
• that waste stream has to generate carbon-14 bearing gas; and
• a bulk gas phase has to entrain the carbon-14 bearing gas: and
• these gases must migrate through the engineered barriers in significant quantities; and
• these gases must migrate through the overlying geological environment (either as a distinct gas phase or
as dissolved gas); and
• these gases must interact with materials in the biosphere (i.e. plants) in a manner that leads to significant
doses and risks to exposed groups or potentially exposed groups.
The Radioactive Waste Management Directorate of the UK’s Nuclear Decommissioning Authority is
undertaking a programme of research to investigate the release of carbon-14 from irradiated materials
following closure of a UK geological disposal facility (GDF) to address the second bullet point above. Some
of these materials contain a significant proportion of the UK carbon-14 inventory in intermediate–level waste
(ILW) (e.g. graphite) whereas others may be in a chemical form that could result in rapid generation of
gaseous carbon-14 after resaturation of waste packages sometime after GDF closure (e.g. irradiated Magnox
or metallic uranium). This paper will describe the results of recent studies to measure the gaseous release of
carbon-14 from the leaching of irradiated graphite in alkaline water and from the corrosion of irradiated
Magnox alloy under similar conditions.
It has been shown previously that small amounts of gaseous carbon-14 are released during leaching of
irradiated UK graphites from the Windscale Advanced gas-cooled reactor (WAGR) and from the air-cooled
British Experimental Pile‘0’ (BEPO) [1].
These experiments were carried out under alkaline oxic conditions and apparatus was developed to
discriminate organic/hydrocarbon- 14 C and inorganic ( 14 CO/ 14 CO 2 ) species in the gas phase by means of
selective oxidation and capture. Results for BEPO graphite showed that a small fraction of the graphite
carbon-14 inventory was released to the gas phase (~0.004% as CO/CO 2 and~0.001% associated with
organic/hydrocarbon compounds) with a larger fraction, about 0.1%, released to solution over the fourteen
month duration of the experiment (Figure 1). The understanding of the effect of aqueous conditions on the
rate of release of carbon-14 from irradiated graphite and its chemical form is being developed in a further
study using irradiated graphite from a Magnox reactor at Oldbury in the UK. This is investigating the effect
of reducing conditions, pH and graphite surface area. Results so far show that the proportion of the carbon-
14 inventory released to the gas phase appears to be less than from the irradiated BEPO graphite. In addition,
the partitioning between inorganic and organic carbon-14 may be affected by the redox potential of the
solution. The new results will be discussed and compared to the previous studies [1].

100
0.0045%
90
Cumulative C-14 released (Bq)
80
70
60
50
40
30
20
10
Experimental C-14 as CO
Experimental C-14 as organic
Model C-14 as CO
Model C-14 as organic
0.0040%
0.0035%
0.0030%
0.0025%
0.0020%
0.0015%
0.0010%
0.0005%
Cumulative C-14 released (fraction of inventory)
0
0.0000%
0 2000 4000 6000 8000 10000 12000
Time (Hours)
Figure 1 The 14 CO and 14 C-organic release from irradiated BEPO graphite
Initial results from a similar study to measure the release of gaseous carbon-14 from the corrosion of
Magnox alloy under alkaline conditions will also be presented.
This work was funded by NDA-RWMD
[1] G.M.N. Baston, T.A. Marshall, R.L. Otlet, A.J. Walker , I.D. Mather and S.J. Williams, Rate and speciation of
volatile carbon-14 and tritium releases from irradiated graphite, Mineralogical Magazine 78, 3293-3302, (2012).
PA1-2
ELICITATION OF DISSOLUTION RATE DATA FOR POTENTIAL WASTEFORM TYPES FOR
PLUTONIUM UNDER REPOSITORY CONDITIONS
G. Deissmann 1, 2) , S. Neumeier 1) , F. Brandt 1) , G. Modolo 1) , D. Bosbach 1)
1) Institut für Energie- und Klimaforschung – Nukleare Entsorgung und Reaktorsicherheit (IEK-6),
Forschungszentrum Jülich GmbH, D-52425 Jülich, Germany
2) Brenk Systemplanung GmbH, D-52080 Aachen, Germany
Various plutonium isotopes are generated during the operation of nuclear reactors from the uranium present
in the nuclear fuels through capture of neutrons. The plutonium contained within spent nuclear fuels can be
recovered during reprocessing. At present, separated stocks of UK civil plutonium (about 90 t HM as PuO 2 ) are
held in storage as zero value asset, since there is no final decision on plutonium disposition in the UK.
Recently, the UK government stated that its preliminary preferred policy on the long-term management of
plutonium is reuse as MOX fuel, but consideration of disposal options will continue [1]. However,
irrespective of future UK government strategies for plutonium disposition, at least a portion of the UK
plutonium inventory (i.e. some tonnes) is likely to be designated for geological disposal.
This paper will describe outcome and conclusions of an elicitation exercise regarding the dissolution rates of,
and the plutonium release from, potential wasteforms for plutonium, performed on behalf of the NDA
RWMD. The elicited distributions provide a 'non-conservative lower bound', a 'conservative upper bound'
and where possible a 'best current estimate' for the dissolution rates and the plutonium release rates. These

data can be used in Post-Closure Safety Assessment Models to calculate performance measures such as
annual individual risk related to plutonium disposal with time. Generic candidate wasteform types included
in this study were borosilicate glasses and phosphate glasses, ceramic wasteforms, and low-specification
"storage" MOX. Due to the character of the current UK disposal programme (‘generic stage’, in which no
preferred disposal concept has yet been selected), a range of possible environmental conditions in the
repository near-field, including effects of an alkaline plume potentially arising from a co-located
cementitious LILW-repository module, and various disposal scenarios were considered.
Experimental data on the durability of plutonium wasteforms under repository-relevant conditions are
generally limited or in many cases absent (e.g. [2, 3]). Due to this limited knowledge base, the initial part of
the work comprised the development of conceptual approaches to evaluate the plutonium release rates from
various wasteforms. Following the compilation and analysis of relevant data and information, the respective
distributions of the wasteform dissolution rates and plutonium release rates for the conditions expected in a
geological repository in the UK were derived. The information basis regarding wasteform durability and
leaching resistance is rather diverse for the different matrices. Information on the performance of plutoniumbearing
glasses is rather limited to date. However, considerable knowledge exists from laboratory and in-situ
studies about borosilicate-based nuclear waste glasses for disposal of reprocessing wastes and their long-term
performance. In contrast, the database regarding the long-term leaching behaviour of phosphate-based
glasses was found to be comparatively small. Data with respect to the performance of ceramic wasteforms in
the repository environment are rather scarce, and a systematic approach to an understanding of the aqueous
durability of the various ceramic matrices (regarding, e.g., the dissolution behaviour as function of the
crystallographic structure, chemical composition, lattice substitutions, radiation damage, etc.) is still lacking.
Furthermore, even for nominally similar matrices, the data can show a considerable spread and are often
difficult to compare, due to different processing and fabrication routes employed, different experimental
conditions, as well as the usage of plutonium surrogates in some experiments. Experimental investigations
on the dissolution and long-term performance of storage MOX and/or calcined PuO 2 are also rather limited
to date. Thus the assessments and the elicitation of the dissolution rates of this wasteform was based on
experiments and modelling studies related to the matrix dissolution of spent nuclear fuels (i.e. UOX and
MOX), and on the understanding of relevant processes affecting their long-term behaviour in a repository,
which has been significantly expanded throughout the last decades. The dissolution rates derived for the
different (generic) plutonium wasteforms under conditions relevant for a UK geological disposal facility are
summarised in table 1.
Table 1. Elicited dissolution rates (in g m -2 d -1 ) for potential plutonium wasteforms under conditions relevant
for a UK geological disposal facility
Wasteform Lower bound Best estimate Upper bound
Borosilicate glass

PA1-3
SOLUBILITY AND TRLFS STUDY OF Nd(III) AND Cm(III) IN DILUTE TO CONCENTRATED
ALKALINE NaCl–NaNO 3 AND MgCl 2 –Mg(NO 3 ) 2 SOLUTIONS
M. Herm 1 , X. Gaona 1 , Th. Rabung 1 , C. Crepin 2 , V. Metz 1 , M. Altmaier 1 , H. Geckeis 1
1 Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Karlsruhe, Germany
2 Ecole National Supérieure de Chimie de Montpellier, Montpellier, France
In long–term performance assessment analyses for deep geological nuclear waste repositories a reliable
prediction of the chemical behavior of actinides in aqueous solutions is necessary. Although geological or
geotechnical barriers may prevent formation water from contacting the waste, intrusion of aqueous solutions
into a repository has to be taken into account. Porewater in certain sedimentary bedrocks as well as water
potentially intruding salt rock repositories will be characterized by high ionic strength and high Na + , Mg 2+
and Cl − concentrations. In repositories with waste from nuclear fuel reprocessing, high nitrate concentrations
(≥ 1.0 M) and slow nitrate reduction kinetics have to be taken into account which may impact the
mobilization of actinides. An(III) and An(IV) are the most relevant actinide redox states under the reducing
conditions which develop after the closure of deep underground repositories for nuclear waste.
Nitrate complexes of actinides are reported to be weak but clearly more stable than the corresponding
chloride counterparts. Accounting for the An(III)–NO 3 system, the last NEA–TDB update [1] selected only
thermodynamic data for the complex AmNO 3 2+ . Recent Cm(III)–TRLFS studies in the presence of elevated
nitrate concentrations proposed the formation of both CmNO 3 2+ [2–3] and Cm(NO 3 ) 2 + [2] complexes. So far,
all these studies focus on acidic conditions, leaving aside the assessment of nitrate effects under neutral to
alkaline repository–relevant pH conditions.
Solubility experiments were conducted with Nd(OH) 3 (am) in 0.1–5.0 M NaCl–NaNO 3 and 0.25–4.5 M
MgCl 2 –Mg(NO 3 ) 2 mixtures at 7.5 ≤ pH c ≤ 13 and 0 ≤ [NO − 3 ] ≤ 7.0 M at 22 ± 2°C under inert gas (Ar)
atmosphere. Samples were equilibrated for t ≤ 500 days, and pH c and [Nd] (ICP–MS) monitored at regular
time intervals. After ensuring equilibrium conditions, solid phases were characterized by XRD and SEM–
EDX. Aqueous speciation was further investigated by TRLFS with ~10 −7 M Cm(III) in 5.0 M NaCl–NaNO 3
and 0.25/3.5 M MgCl 2 –Mg(NO 3 ) 2 mixtures as background electrolyte. The original pH c (pH max ∼9) was
titrated to pH c = 1 with HCl of corresponding ionic strength and Cm(III) fluorescence spectra collected at
each individual pH c . The comparison of Nd(III) solubility in MgCl 2 –Mg(NO 3 ) 2 systems with nitrate–free
MgCl 2 solutions [4] shows a clear increase of [Nd(III)] for [Mg 2+ ] ≥ 2.5 M, [NO − 3 ] ≥ 1.0 M and pH c 8–9. In
contrast to Mg–bearing systems, no effect of nitrate on Nd(III) solubility is observed in dilute to concentrated
NaCl–NaNO 3 solutions even up to 5.0 M NaNO 3 . XRD and SEM–EDX results confirm that Nd(OH) 3 (am) is
the solubility controlling solid phase in all systems with [Cl − ] ≤ 5.0 M. TRLFS studies in 3.5 M MgCl 2 –
Mg(NO 3 ) 2 with 0 ≤ [NO − 3 ] ≤ 7.0 M solutions indicate the formation of CmNO 2+ 3 and Cm(NO 3 ) + 2 complexes
in acidic to near neutral pH conditions as previously described in [2]. Peak deconvolution of the complete set
of fluorescence spectra further reveals the formation of ternary Cm–OH–NO 3 species under weakly alkaline
pH c conditions, likely CmOHNO +
3 and Cm(OH) 2 NO 3 (aq). These spectroscopic results confirm that the
observed increase in solubility for Nd(OH) 3 (am) is not a simple matrix effect but is related to a genuine
complexation reaction with nitrate.
The combination of slope analysis, solid phase characterization and TRLFS indicates the equilibrium
reaction Nd(OH) 3 (am) + 2H + + NO 3 − ⇔ NdOHNO 3 + + 2H 2 O in concentrated nitrate–bearing systems and
permits to further extend the chemical and thermodynamic models described in [4] for Ln(III) and An(III) to
Ln 3+ /An 3+ –H + –Na + –Mg 2+ –Ca 2+ –OH − –Cl − –NO 3 − systems.
[1] R. Guillaumont, Th. Fanghänel, V. Neck, J. Fuger, D.A. Palmer, I. Grenthe, M.H. Rand, Elsevier, Amsterdam
(2003).
[2] A. Skerencak, P.J. Panak, W. Hauser, V. Neck, R. Klenze, P. Lindqvist-Reis, Th. Fanghänel, Radiochim. Acta 97,
385–393 (2009).
[3] L. Rao, G. Tian, Dalton Trans. 40, 914–918 (2011).
[4] V. Neck, M. Altmaier, Th. Rabung, J. Lützenkirchen, Th. Fanghänel, Pure Appl. Chem. 81, 1555–1568 (2009).

PA1-4
SOLUBILITY AND HYDROLYSIS OF U(VI) AT 80°C UNDER ACIDIC TO HYPERALKALINE
pH CONDITIONS
X. Gaona 1 , M. Marques 2 , B. Baeyens 2 , M. Altmaier 1
1 Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Karlsruhe, Germany
2 Laboratory for Waste Management, Paul Scherrer Institut, Villigen PSI, Switzerland
Temperature is one of the parameters that will vary during the different phases of operation of a high level
radioactive waste (HLW) repository. Elevated temperature conditions (up to 200°C depending on host rock
system and repository concept, e.g. [1]) will affect actinide chemistry in the near-field of a HLW repository.
The hydrolysis of U(VI) has been thoroughly studied at 25°C. The thermodynamic data selection resulting
from the NEA–TDB reviews [2] is thus very complete (as log K°, Δ r G° m or Δ f G° m ) and includes most of the
hydrolysis species expected to form from very acidic to hyperalkaline pH conditions. On the contrary, a very
limited number of studies is dedicated to assess the effect of temperature on the hydrolysis of U(VI), which
results in enthalpy and entropy data selection in [2] being restricted to UO 2 OH + . More recent calorimetric
and spectroscopic studies have extended this list to (UO 2 ) 2 (OH) + 2 , (UO 2 ) 3 (OH) 2+ 4 , (UO 2 ) 3 (OH) +
5 and
(UO 2 ) 4 (OH) + 7 [3, 4]. However, no Δ f H° m and S° m data are available for neutral and anionic U(VI) hydrolysis
species prevailing in near-neutral to alkaline, repository-relevant, pH conditions. Regarding solid phases,
both enthalpy, entropy and C° p,m data are selected in the NEA–TDB for UO 3 2H 2 O(cr) based on
calorimetric studies. No data (even at 25°C) are selected for ternary Na–U(VI)–OH solid phases reported to
control the solubility of U(VI) under alkaline conditions [5].
Undersaturation solubility experiments were conducted with UO 3 2H 2 O(cr) and Na 2 U 2 O 7 ⋅H 2 O(cr) solid
phases synthesized and characterized in a previous solubility study at 25°C [5]. All samples were prepared in
an inert gas (N 2 ) glovebox under exclusion of O 2 and CO 2 . U(VI) solid phases were distributed in several
independent batch experiments (4–5 mg solid per sample), arranged in two series of 0.5 M NaCl–NaOH
solutions with 4 ≤ pH m ≤ 7 (UO 3 2H 2 O(cr)) and 8 ≤ pH m ≤ 12 (Na 2 U 2 O 7 ⋅H 2 O(cr)) and stored in an oven at
80°C (in a N 2 glovebox). Samples were equilibrated for ∼1 year and monitored at regular time intervals for
dissolved uranyl, [U], and pH m . Triplicate aliquots (supernatant, 0.1 µm filtration and 10 kD ultrafiltration)
were taken for the quantification of [U] by ICP–MS. After reaching equilibrium conditions, solid phases of
selected samples were characterized by XRD, SEM–EDS and chemical analysis.
The results indicate that the solubility of UO 3 2H 2 O(cr) slightly decreases (∼0.5 log–units) at 80°C
compared to similar solubility experiments at 25°C. The species UO 2 2+ is expected to prevail within 4 ≤ pH m
≤ 6 in 0.5 M NaCl, and thus this observation can be related to the effect of temperature on
log *K s,0 (UO 3 2H 2 O(cr)). On the contrary, a significant increase in solubility (1–2 log units) is observed at
80°C for Na 2 U 2 O 7 ⋅H 2 O(cr). Similar to 25°C, the solubility of U(VI) as a function of pH shows a well-defined
slope of +1 in the pH m region 9–12, likely corresponding to the equilibrium 0.5 Na 2 U 2 O 7 ⋅H 2 O(cr) + 2 H 2 O
⇔ Na + + UO 2 (OH) 4 2– + H + . This observation is consistent with the expected increase of hydrolysis of metal
ions with increasing temperature as a result of the increase of K w of H 2 O. The combination of slope analysis
with accurate solid phase characterization (XRD, SEM–EDS, chemical analysis) allows the development of a
thermodynamic model for the system UO 2 2+ –H + –Na + –OH – –Cl – at 80°C valid within 4 ≤ pH m ≤ 12.
[1] Nagra, 2002. Project Opalinus Clay – safety report: demonstration of disposal feasibility for spent fuel, vitrified
high-level waste and long-lived intermediate-level waste (Entsorgungsnachweis). Nagra Technical Report NTB 02-05
[2] Guillaumont, R., Fanghänel, J., Neck, V., Fuger, J., Palmer, D.A., Grenthe, I., Rand, M.H. (2003) Chemical
Thermodynamics 5. Update on the Chemical Thermodynamics of Uranium, Neptunium, Plutonium, Americium and
Technetium. NEA OECD, Elsevier.
[3] Zanonato, P. L., Di Bernardo, P., Bismondo, A., Liu, G., Chen, X., Rao, L. (2004). Hydrolysis of uranium(VI) at
variable temperatures (10-85 °C). Journal of the American Chemical Society, 126, 5515-5522.
[4] Crea, F., De Stefano, C., Pettignano, A., Sammartano, S. (2004). Hydrolysis of dioxouranium(VI): a calorimetric
study in NaCl and NaClO 4 at 25°C. Thermochimica Acta, 414, 185–189
[5] Altmaier, M., Neck, V., Metz, V., Müller, R., Schlieker, M., Fanghänel, Th. (2003). Solubility of U(VI) in NaCl and
MgCl 2 solution. Migration Conference 2003, Gyeongju (Korea).

PA1-5
EFFECTS OF SURFACE MORPHOLOGY ON DISSOLUTION OF ThO 2
E. Myllykylä (1) T. Lavonen (1) and K. Ollila (1)
(1) VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo –Finland
The objective of this study was to investigate the effects of solid surface evolution on the dissolution rate of
ThO 2 during dissolution process. In the dissolution experiments both the solution phase and the dissolving
surfaces are analysed. With fluorite structure ThO 2 can be considered as an analogue for spent fuel matrix
UO 2 . However, unlike U(IV), Th(IV) is not redox active. ThO 2 appears also as solid solution in spent fuel
matrix. The next generation applications of nuclear energy have shown interest towards thorium. As a fuel
thorium has many beneficial properties, such as high fusion temperature, good sintering capability, resistance
against radiation damages, greater abundance in the Earth’s crust compared to U, and the possibility for
transmutation.
In literature, the solubility values of ThO 2 , as well as the hydrolysis constants of thorium show great
discrepancies [1]. The main reasons for the discrepancies are the tendency of Th to undergo polynucleation
and colloid formation, its strong absorption to surfaces, and the low solubility of Th 4+ hydroxide and hydrous
oxide.
The solubility product values have been observed to vary between ThO 2 (cr) (log K°sp = - 56.9) and
ThO x (OH) y (H 2 O) z (s) (log K°sp = -45.5) depending on the crystallinity of the Th phase. In addition to
crystallinity, the surface phenomena have been mentioned as a possible factor affecting the solubility
properties. Vanderborre et al. 2010 [1] used solid analyses, a leaching experiment, and isotopic exchange in
their study to contribute to the understanding of the discrepancy in the solubility values and to describe the
reversibility in the exchange mechanism. They observed the dissolution mainly to occur on grain boundaries
and variation between different sites “local solubility”. The
229 Th spiking revealed dynamic
dissolution/precipitation reactions on the solid/solution interface. One aim of this study is to investigate
whether the evolution or the history of the surfaces have an effect on the dissolution rate of ThO 2 .
Prior to the experiments, the sintered ThO 2 pellets were prepared in a way to obtain an ideal composition and
microstructure similar to UO 2 fuel [2, 3]. The methodology of preparation has been described in detail
elsewhere [4].
The optimized sintering temperature for ThO2 powder was 1750 °C, in which the density of the prepared
ThO 2 pellets was 93 % of the theoretical density. The surface topography and crystal orientation of the ThO 2
pellet’s top and cross section surfaces were analysed with confocal profilometry and EBSD (Electron
Backscatter Diffraction). The surface analyses showed the microstructure similar to UO 2 fuel pellets,
randomly oriented crystals with grain size from 10 to 30 μm.
For the experiments, ThO 2 pellets were crushed to pieces in a mortar. The suitable size of particles from 2 to
4 mm, were selected and washed with isopropanol and ethanol in order to remove fines. Smaller particles of
ThO 2 were produced by high voltage pulse power fragmentation method, SelFrag. The produced material
was washed, magnetically separated from the impurities generated during the fragmentation, and sieved with
sieving canvas to selected particle size fraction of 80 to 160 µm.
The pre-tests of solubility and dissolution rate studies were conducted with 2 to 4 mm particles in 0.1 M
NaCl and 0.01 M NaCl (with 2 mM NaHCO 3 ) solutions under atmospheric conditions. The reaction time of
pre-tests was 115 days and the tests were run in triplicate. The actual experiments were conducted with two
particle sizes 60 to 180 μm and 2 to 4 mm in 0.1 M NaCl in Ar glove box. The ThO 2 samples had a different
history; they were either intact particles, particles from pre-tests or particles leached already with HNO 3 . In
some of the tests isotopic exchange with 229 Th was used to discover the reversibility of the surface reactions.
In addition to the dissolution rate experiments, the solubility of ThO 2 was studied as a function of pH. The
experiments were started with tests in 1 M HNO 3 (~ pH 1) at 80 °C under Ar atmosphere. After heating, test

vessels were removed to 25 °C and let settle to equilibrium. When the concentration of Th had reached
constant value, the solution was diluted to next pH value. Some of the particles were kept even longer in 1 M
HNO 3 at 80 °C under Ar atmosphere to see some evolution of the dissolving surfaces. So far the altered
ThO 2 surfaces have been studied with SEM, but some EBSD and profilometer analyses are in plans, as some
changes were observed in the SEM results. The concentration of Th in unfiltered and ultrafiltered samples
was analysed with HR-ICP-MS, as well.
In pre-tests, the solubility and the dissolution rate increased in the solution containing carbonate due to the
formation of carbonate/hydroxide complexes, and the dissolution rates were evaluated to be 1·10 -13 mol dm -3
d -1 in carbonate containing solution and 4·10 -14 mol dm -3 d -1 in 0.1 M NaCl solution. The SEM images of the
HNO 3 leached ThO 2 particles revealed that the nature of the dissolution/precipitation phenomenon varied
among the particles present in same solution. Even in a same particle, there were some surfaces and grains
which seemed to have dissolved at grain boundaries and surfaces which showed different
dissolution/precipitation mechanism (Figure 1).
This study has been a part of REDUPP project. The research leading to these results has received funding
from the European Atomic Energy Community's Seventh Framework Programme (FP7) under grant
agreement No. 269903.
Figure 1. SEM image of ThO 2 particle after 4 weeks of leaching with 1 M HNO 3 at 80°C. The upper part
shows disappearance of clear grain boundaries and variation in orientations of the grain surfaces, while in the
middle area some grains resemble intact grains before leaching. The lower part shows dissolution mainly
having occurred from the grain boundaries.
[1] Vandenborre J., Grambow B. and Abdelouas A. (2010). Discrepancies in Thorium Oxide Solubility Values: Study
of Attachment/Detachment processes at the Solid/Solution Interface, Inorganic Chemistry, 29, 8736-8748
[2] Forsyth R., 1987. Fuel rod D07/B15 from Ringhals 2 PWR: Source material for corrosion / leach tests in
groundwater. Fuel rod / pellet characterisation program part 1. SKB Technical Report TR 87-02. Svensk
Kärnbränslehantering AB.
[3] Forsyth R., 1995. Spent nuclear fuel. A review of properties of possible relevance to corrosion processes. SKB
Technical Report TR 95-23. Svensk Kärnbränslehantering AB.
[4] REDUPP Deliverable 1.1 (found on the REDUPP web page, www.skb.se/REDUPP)
PA1-6

THE EFFECT OF THE CONCENTRATION OF SILICATES ON THE SOLUBILITY OF Pu(IV) IN
SODIUM BICARBONATE/CARBONATE SOLUTIONS
T. Yamaguchi 1) , K. Henmi 1) , Y. Iida 1) , H. Okamoto 2) , T. Tanaka 1)
1) Waste Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency
2-4, Shirakata, Tokai, Ibaraki 319-1195, Japan
2) Operation Planning Section, Department of Fukushima Technology Development, Japan Atomic Energy
Agency, 2-4, Shirakata, Tokai, Ibaraki 319-1195, Japan
Although silicates are ubiquitous in natural groundwater and potentially affect the solubility of actinides, no
consensuses have been achieved on how to deal with the effects in the safety assessment of geological
disposal system because of lacking quantitative knowledge. In this study, the solubility of 239 Pu(IV) was
determined at the pH range 7.0 - 12.4 in Na + -HCO 3 - -CO 3 2- solutions whose total carbonate concentration
ranged 3x10 -4 - 0.1 mol L -1 . Then the Pu solutions were added with Na 2 SiO 3 to get the dissolves silicate
concentration of 1.25x10 -5 – 3.6x10 -4 mol L -1 to observe the effect of the silicate concentration on the
solubility of Pu(IV). The solubility of Pu(IV) were negatively correlated with the pH and positively
correlated with the total carbonate concentration. No obvious correlation was observed between the
solubility of Pu and the silicate concentration. Neither formation of aqueous silicate complexes nor silicate
compounds was significant under the experimental conditions.
OECD NEA TDB [1] selected thermodynamic data for USiO 4 (coffinite) as a silicate compound of U(IV). If
the coffinite limits the concentration of uranium in aqueous solution, the solubility of uranium must be lower
than the case that the UO 2 (am) does by around 6 orders of magnitude. Such a data have not been reported.
Rai et al. [2] reported that the solubility of Th(IV) was lower in the presence of silicates than the ordinary
solubility of ThO 2 (am) by 2.7 orders of magnitude. The reason for the low solubility, however, was not
identified as the formation of thorium silicates. The silicates may also enhance the solubility of actinides via
complexation with silicates. Although the silicate complexation was observed for An(III) at the neutral to
weak alkaline conditions [3-5], for An(IV) only around pH 1 [6]. In this study, the effect of the dissolved
silicates on the solubility of Pu(IV) was experimentally examined in order to judge whether the silicate
compounds can be considered as solubility limiting solids for An(IV).
A 239 Pu(IV) stock solution was prepared as a 2.1x10 -3 mol L -1 solution in 1 mol L -1 HNO 3 . A 50 µL aliquot of
the Pu(IV) stock solution was added to 40 mL of mixtures of NaCl, NaHCO 3 , Na 2 CO 3 , NaOH solutions. The
ionic strength of the solution was 0.1, 1.0 or 2.0 mol kg -1 , and the total carbonate concentration (C t = [HCO 3 - ]
+ [CO 3 2- ]) 3×10 -4 - 10 -1 mol L -1 , respectively. The pH was adjusted to values between 9 and 12. The solutions
were stored for 8 weeks under Ar at room temperature (20 - 28°C). The Pu concentration was determined by
alpha spectrometry after separating any solids with 5,000 NMWL (nominal molecular weight limit)
polysulfon filters. The storage and the concentration determination were repeated after adding 450 µL of 10 -2
mol L -1 Na 2 SiO 3 solution and after further adding 1000 µL of the Na 2 SiO 3 solution.
The solubility of Pu(IV) showed a negative correlation to the pH and a positive correlation to the total
carbonate concentration as shown in Fig. 1. No obvious change in the solubility of Pu(IV) with increasing
silicate concentration up to 3.6×10 -4 mol L -1 as shown in Fig. 2. Neither formation of aqueous silicate
complexes nor silicate compounds was significant under the experimental conditions.

Fig. 1 Solubility of Pu(IV) in the absence
of silicates
Fig. 2 Solubility of Pu(IV) in the absence of silicates
(56 th d), at the silicate concentration of 1.1×10 -4
mol L -1 (112 th d) and at 3.6×10 -4 mol L -1 (168 th d).
[1] I. Grenthe et al., “Chemical Thermodynamics of Uranium” North-Holland, Amsterdam (1992).
[2] D. Rai et al., Radiochim. Acta 93, 443 (2005).
[3] P. J. Panak et al., Radiochim. Acta 93, 133 (2005).
[4] Z. Wang et al., Radiochim. Acta 93, 741 (2005).
[5] P. N. Pathak, G. R. Choppin, Radiochim. Acta 94, 81 (2006).
[6] A. B. Yusov et al., Radiochim. Acta 92, 869 (2004).
PA1-7
LEACHING TEST OF CEO 2 UNDER GROUNDWATER CONDITIONS.
N. Rodriguez, J. Cobos, E. Iglesias, C. Palomo, J. Nieto, J. M. Cobo, L. Serrano, S. Durán, J. Quiñones
CIEMAT Avenida Complutense, 40- Edif. 12
28040 – Madrid – Spain
Problems related with waste management in the current uranium oxide nuclear fuel cycle mainly due to its
radiotoxicity have been well characterized. The formation of fission products (PF), plutonium and minor
actinides, generated during fuel irradiation, is responsible of such radiotoxicity. One of the proposals
assessed in the non-proliferation treaty of nuclear weapons (SNE-TP) to minimize Pu inventories, is to reuse
the Pu (advanced closed cycles). International projects are proposing different alternatives into this subject.
Different matrices behaviour either metal or ceramic is being under investigation.
Due to its abundance, irradiation stability and temperature stability range, ThO 2 can be considered as a
candidate material as Pu matrix. Up to 75% of initial Pu can be incinerated per path and 232 Th produces 1 ,
during irradiation, less inventories of fissile elements.

Cerium oxide is a good surrogate for plutonium dioxide for the studies involving ceramic phases., both have
the same fluorite type crystallographic structure and comparable thermodynamic properties 2,3 . The aim of
this work is to study the effect of the dissolution of Ce and CeO 2 (cr) under spent nuclear fuel storage
conditions. To assess the dissolution behaviour of this material as a pellet form and as powder at room
temperature, leaching tests were carried out. Batch experiments were performed under demineralised and
carbonated synthetic water with an atmospheric control. The leaching tests were carried out into a glove box
with argon atmosphere. Preleaching characterization of all the samples was performed by XRD, BET Laser
Diffraction system and Scanning Electron Microscopy.
Concerning the calculation of Ce solubility, thermodynamic PHREEQC code was applied, using a Ciemat
database. Solubility limiting solid phase selected to this work were CeO 2 and Ce(OH) 4 . The results were
compared with the experimental data.
The trend of cerium concentration as a function of time in all the experiments increased until a steady-state
level was reached. This work confirms the stability of the CeO2, also under high carbonated media.
1
E. Shwageraus, P. Hejzlar, and M. S. Kazimi, "Use of Thorium for Transmutation of Plutonium and Minor Actinides
in PWRs," Nuclear Technology, vol. 147, pp. 53-68, 2004
2 H. S. Kim, C. Y. Joung, B. H. Lee, J. Y. Oh, Y. H. Koo, and P. Heimgartner, "Applicability of CeO2 as a surrogate for
PuO2 in a MOX fuel development," Journal of Nuclear Materials, vol. 378, pp. 98-104, 200
3 M. Stan, T. J. Armstrong, D. P. Butt, T. C. Wallace, Y. S. Park, C. L. Haertling, T. Hartmann, and R. J. Hanrahan Jr,
"Stability of the perovskite compounds in the Ce-Ga-O and Pu-Ga-O systems," Journal of the American Ceramic
Society, vol. 85, pp. 2811-2816, 2002
PA1-8
INFLUENCE OF PARTICLE SIZE, CARBONATES AND ORGANIC MATTER ON
THE SOLUBILITY OF THO 2 (cr)
S. Salah 1) , D. Liu 2) , L. Wang 1)
1)
Belgian Nuclear Research Centre, SCK•CEN, Expert Group Waste & Disposal, Boeretang 200, B-2400
Mol, Belgium
2)
State Nuclear Power Research Institute, 100029, Beijing, China
In order to assess the safety of geological radioactive waste disposal in Boom Clay (BC), which is
investigated as one of the potential host formations in Belgium, it is important to know which effects may
influence the solubility of waste relevant radionuclides under BC conditions. Therefore, first the solubility of
crystalline ThO 2 (cr) was studied at I = 0.01 M NaClO 4 (S/L ratio: 10 g/L) in the pH range 2-10 and in
carbon-free atmosphere (N 2 glovebox, 25°C) enabling comparison of the results with literature data.
Afterwards, in order to study the effect of carbonates and organic matter (both present in BC), additional
experiments were done under in-situ conditions (Ar-0.4% CO 2 ) using Synthetic Boom Clay Water (SBCW:
0.015 M NaHCO 3 ) and Real Boom Clay Water (RBCW: 0.015 M NaHCO 3 with ~50-200 mg/L dissolved
organic carbon), respectively.
The Th-concentrations were determined after 90 days using ultracentrifugation at 694,000 x g (0.5 h and 2 h)
and high-speed centrifugation at 108,000 x g (1h).
Results indicated that with increasing pH log 10 [Th] decreased, but from pH > 5.5 the Th-concentrations
approached quasi-constant and pH-independent values ranging between log 10 [Th] -9.8 and -10.4.
Solubility calculations performed with GWB [1] and using thermodynamic data from [2], revealed that the
experimentally determined thorium solubilities at the different pH values could not be reproduced by using
the solubility constant for ThO 2 (cr).
ThO 2 (cr) + 2 H 2 O = Th 4+ + 4 OH - log 10 K° s,0 = -54.2 ± 1.1

Only after adaptation of the latter, consistent values between the experimentally determined Th solubilities
and the modeled ones could be obtained. In Tab. 1, the adapted solubility constants are summarized.
Table 1: Adapted ThO 2 (cr) solubility constants
pH 2 3 4 5.5 8 10
logK° sp -53.9 -50.6 -51.5 -48.9 -48.6 -49.0
It is well known that variations in solubility can be due to surface hydration and/or particle/crystallite size
variation [2]. It can be assumed that the ThO 2 powder that was used in the experiments was not
homogeneous and had a certain size distribution. The relation between logK° sp of small ThO 2 particles
(particle size d) and large crystals of ThO 2 (cr) was derived by [3] and is given by:
logK° sp (ThO 2 , particle size d) = -(54.2 ± 1.1) + 22.6/d (nm)
Applying Schindler’s equation to the solubility products fitted to the experimental data (Tab. 1), leads to
particle sizes ranging between ~4 and 66 nm.
In Fig. 1a, the effect of particle size on the solubility constant of ThO 2 (cr) predicted by the equation of
Schindler is illustrated. For comparison, also the solubility constants for microcrystalline ThO 2 , fresh
ThO 2 (am,hyd) and dried Th(OH) 4 (am) are plotted on the solubility curve. Marked in red are the data derived
from the experimental data of this study. It can be seen that all points are lying on the solubility curve,
confirming the predicted values. Furthermore, it can be seen that only the solubility constant derived for the
lowest pH value (pH = 2) actually refers to crystalline ThO 2 (cr) and that the solubility is mainly determined
by crystallites > 30 nm. However, with increasing pH, the solubility is determined by fractions of smaller
size and the solubility products are ranging between the one for microcrystalline ThO 2 (-53.0 ± 0.5) and
amorphous, hydrated ThO 2 (-47.8 ± 0.6).
By assuming different initial size fraction distributions and S/L ratios, [2] derived ThO 2 (cr) solubility
limiting curves for different particle sizes as function of pH (Fig. 1b). When plotting the experimental data
into this graph, it can be deduced that the solubility at pH = 2 is mainly determined by large particles, i.e. >
30 nm, while at pH 3-4, particles < 30 nm represent the solubility limiting fraction, and at pH > 5.5 particles
< 10 nm determine the solubility.
pH=5.5-10
RBCW
pH=2
694,000 g/0.5h
694,000 g/2h
SBCW
Figure 1 a) Effect of particle size on the solubility constant of ThO 2 (cr) predicted with the equation
of Schindler, b) Solubility curves for different size fractions
(adapted figures from [2])
The log 10 [Th] concentrations in SBCW after ultracentrifugation (0.5 and 2h) were measured to be -10.4 and -
9.7. These values are thus lying within the range of concentrations determined for the experiments using 0.01
M NaClO 4 (pH range 5.5-10). This reveals that complexation of thorium with bicarbonate (HCO 3 - ) does not
seem to have a solubility increasing effect.

In contrast, the log 10 [Th] concentrations measured in RBCW after ultracentrifugation (0.5 and 2h) are around
2 log units higher (i.e. log 10 [Th] = -8.1 and -7.8) than the values determined in NaClO 4 and SBCW,
respectively. This clearly shows that organic matter (here: humic and fulvic acids) have a solubility
increasing effect and that organic complexation cannot be neglected when trying to model the experimentally
determined solubilities.
[1] Bethke C. M. (2010). The Geochemist's Workbench Release 8.0 (four volumes), Hydrogeology Program. University
of Illinois, Urbana, Illinois.
[2] Rand M. et al. (2009). Chemical Thermodynamics of Thorium, OECD NEA, Issy-les-Moulineaux, France.
[3] Schindler P. W. (1967). Heterogeneous equilibria involving oxides, hydroxides, carbonates, and hydroxide
carbonates, Equilibrium concepts in natural water systems, Vol. 67, Adv. Chem. Ser., p. 196-221.
PA1-9
ADOPT PELLET LEACHING PROPERTIES, A COMPARISON WITH UO 2 PELLET
Kristina Nilsson a , Olivia Roth b , and Mats Jonsson a
a Applied Physical Chemistry, KTH Royal Institute of Technology, School of Chemical Science and
Engineering, SE-100 44 Stockholm, Sweden.
b Studsvik Nuclear AB, SE-611 82 Nyköping, Sweden
According to the KBS-3 method, the spent nuclear fuel will be stored in a deep repository, 500 meters below
ground level in the bedrock; kept in copper canisters that are surrounded by bentonite clay. In the event of a
copper canister failure and groundwater diffusing through the bentonite clay, the spent nuclear fuel will
come in contact with the groundwater. The UO 2 matrix will in that case govern the release of radionuclides
into the biosphere. Understanding the oxidative dissolution behavior of UO 2 in water solution is of
importance when assessing the safety of a deep repository 1 .
ADOPT (Advanced Doped Pellet Technology) is a new type of fuel developed by Westinghouse with
properties such as enlarged grain size and reduced Fission Gas Release (FGR). To achieve this UO 2 is doped
with chromium and aluminium oxides 2 . These dopants could influence the leaching properties of the fuel
compared to fuel based on pure UO 2 .
In this work we have performed a kinetic and mechanistic study of ADOPT pellet leaching by evaluating the
reactivity of H 2 O 2 towards ADOPT pellets in aqueous solutions containing HCO 3 - . The dissolution of
uranium has been studied in parallel with H 2 O 2 consumption. The catalytic decomposition of H 2 O 2 on
ADOPT fuel has been quantified and compared to pure UO 2 . In addition, studies of γ-irradiation induced
dissolution of ADOPT pellets as well as the influence of H 2 on the oxidative dissolution process have been
carried out. The Arrhenius parameters for the reaction between H 2 O 2 and ADOPT fuel have also been
determined.
Furthermore, leaching experiments of irradiated ADOPT fuel have been performed. The results of these
leaching studies are compared to numerical simulations of radiation induced dissolution of spent nuclear fuel
taking the results from the model studies above into account.
1 Jonsson, M., Radiation Effects on Materials Used in Geological Repositories for Spent Nuclear Fuel. ISRN
Materials Science, 2012. 2012: p. 13
2 Arborelius, J., et al., Advanced doped UO2 pellets in LWR applications. Journal of Nuclear Science and
Technology, 2006. 43(9): p. 967-976

PA1-10
Np(V) SOLUBILITY IN DILUTE TO CONCENTRATED MgCl 2 SOLUTIONS
V.G. Petrov (1) , X. Gaona (2) , D. Fellhauer (2) , J. Rothe (2) , K. Dardenne (2) , S.N. Kalmykov (1) , M. Altmaier (2)
(1)
Lomonosov Moscow State University, Department of Chemistry, Moscow, Russia
(2)
Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Karlsruhe, Germany
The understanding of actinide solubility and speciation in concentrated chloride media is of particular
relevance for the disposal of radioactive waste in underground salt-rock formations. In the unlikely event of
water intrusion into such repositories, NaCl and MgCl 2 will be the main brine components potentially
contacting the radioactive waste. 237 Np is considered as a relevant dose contributor in performance
assessment (PA) in the long term due to its long half-life (t 1/2 = 2.14·10 6 a) and high radiotoxicity. The
importance of Np is also related to its redox sensitivity and the high mobility of Np(V), particularly due to its
high solubility and weak sorption. Although Np(V) solubility was previously investigated in NaCl, NaClO 4
and CaCl 2 solutions, there are no systematic solubility studies dedicated to the MgCl 2 system. In this system,
the maximum pH is fixed by the precipitation of Mg(OH) 2 (in [MgCl 2 ] ≤ 2 m) or Mg 2 (OH) 3 Cl⋅4H 2 O(cr) (in
[MgCl 2 ] > 2 m) at pH max ∼9 [1], which prevents hydrolysis of Np(V).
In the present study the solubility of Np(V) was investigated from undersaturation in dilute to concentrated
MgCl 2 solutions. All batch samples were prepared and stored at 22±2°C in inert gas (Ar) gloveboxes under
exclusion of oxygen and CO 2 . Np(V) was precipitated as NpO 2 OH(am,fresh) in a diluted alkaline solution
(pH ∼11.5), separated from the supernatant and distributed in five experimental series with 0.10, 0.25, 2.11,
3.87 and 5.15 m MgCl 2 and 7.5 ≤ pH m ≤ 9 (∼15 mg per independent batch experiment). Samples were
equilibrated for ∼6 months and monitored at regular time intervals for [Np] and pH m . Additional experiments
from oversaturation were carried out in 0.25 and 2.11 m MgCl 2 solutions to assess the possible formation of
Mg–Np(V)–OH ternary solid phases as previously described for Np(V) in CaCl 2 solutions under analogous
pH m and ionic strength conditions [2]. Solid phases of selected solubility samples were characterized by
XAFS, XRD, SEM–EDS and quantitative chemical analysis. Likewise, Np(V) speciation in the aqueous
phase was investigated by UV–vis/NIR and XAFS.
In contrast to previous experimental evidence in NaCl or CaCl 2 solutions, no visible solid phase
transformation occurs in Np(V) solubility experiments in dilute to concentrated MgCl 2 solutions within the
timeframe of the experiment. The solubility curve follows a well-defined slope of –1 in all MgCl 2 solutions
(0.1–5.15 m) within the pH range considered (7.5 ≤ pH m ≤ 9) (Figure 1). This observation can be properly
explained by the predominance of the solubility reaction NpO 2 OH(am,fresh) + H + ⇔ NpO 2 + + H 2 O. A
significant increase in solubility is observed with increasing MgCl 2 concentration (∼1.5 log-units between
0.25 m and 5.15 m MgCl 2 ). A similar effect observed in CaCl 2 solutions (∼1.8 log-units between 0.25 m and
5.26 m CaCl 2 ) was explained by considering the predominance of the aqueous ternary species Ca[NpO 2 Cl] 2+
in [CaCl 2 ] ≥ 3.91 m [2].
The combination of slope analysis, aqueous speciation (UV–vis/NIR and XAFS) and solid phase
characterization (XRD, XAFS, SEM–EDS, chemical analysis) allows the development of thermodynamic
and activity (SIT, Pitzer) models for the system NpO 2 + –H + –Mg 2+ –OH – –Cl – at 25°C.

-1
pH m
max
-2
log m Np
-3
-4
-5
0.10 m
0.25 m
2.11 m
3.87 m
5.15 m
7 8 9
pH m
Figure 1. Solubility of Np(V) in MgCl 2 solutions as function of pH m and ionic strength. Solid line is the
calculated Np(V) concentration for 0.1 M MgCl 2 according to NEA–TDB [3]. Aqueous MgCl 2 systems
limited to pH m < ~9 due to precipitation of solid Mg–OH–Cl phases.
This work was supported by ACTINET–i3 Integrated Infrastructure Initiative, the Ministry of Education and
Science of Russian Federation (Contract 11.519.11.5011) and Russian Foundation for Basic Research (grant
12-03-31548).
[1] Altmaier M., Metz V., Neck V., Müller R., Fanghänel T. Solid-liquid equilibria of Mg(OH) 2 (cr) and
Mg 2 (OH) 3 Cl⋅4H 2 O(cr) in the system Mg–Na–H–OH–Cl–H 2 O at 25°C. Geochimica et Cosmochimica Acta, 3595–3601
(2003).
[2] Fellhauer, D. Untersuchungen zur Redoxchemie und Löslichkeit von Neptunium und Plutonium, PhD thesis,
Universität Heidelberg (2013).
[3] Guillaumont, R., Fanghänel, J., Neck, V., Fuger, J., Palmer, D.A., Grenthe, I., Rand, M.H. (2003) Chemical
Thermodynamics 5. Update on the Chemical Thermodynamics of Uranium, Neptunium, Plutonium, Americium and
Technetium. NEA OECD, Elsevier.
PA1-11
IMMOBILISATION OF TECHNETIUM-99 ON BACKFILL CEMENT: SORPTION UNDER
STATIC AND SATURATED FLOW CONDITIONS
C. L. Corkhill 1)* , J. W. Bridge 2) , P. Hillel 3) , L. J. Gardner 1) , M. C. Stennett 1) , R. Tappero 4)
and N. C. Hyatt 1) .
1 The Immobilisation Science Laboratory, Department of Materials Science and Engineering,
The University of Sheffield, UK
2
Centre for Engineering Sustainability, School of Engineering, University of Liverpool, UK
3 Department of Nuclear Medicine, Hallamshire Hospital, Sheffield, UK
4 National Synchrotron Light Source, Brookhaven National Laboratory, Upton, New York, USA
Technetium-99, a β-emitting radioactive fission product of 235 U, formed in nuclear reactors, presents a major
challenge to nuclear waste disposal strategies. Its long half-life (2.1 x 10 5 years) and high solubility under

oxic conditions as the pertechnetate anion [Tc(VII)O 4 ] is particularly problematic for long-term disposal of
radioactive waste in geological repositories.
In this study, we investigate the effectiveness of the backfill cement, Nirex Reference Vault Backfill
(NRVB) and other cements commonly used in nuclear waste disposal scenarios (crushed Ordinary Portland
Cement (OPC) and OPC combined with Blast Furnace Slag (BFS) or Pulverised Fly Ash (PFA)) for their
effectiveness towards immobilisation of Tc(VII). Sorption in batch experiments was shown to be dependent
on cement type; NRVB, OPC and OPC/PFA weakly sorbed pertechnetate, while the BFS-containing OPC
cement sorbed ~50% of the injected Tc(VII). Oxidation state µ-XRF mapping, combined with µ-XANES
performed on a BFS-containing cement reacted with Tc(VII) showed that immobilisation in this cement was
due to a rapid reductive-precipitation mechanism, with Tc(IV) precipitates localised on the surface of BFS
particles. The cements that displayed poor sorption of technetium were found to contain only Tc(VII). Timelapse,
non-invasive, quantitative radiographic imaging of a 99m Tc radiotracer through the different cement
compositions was performed to investigate Tc(VII) immobilisation under dynamic conditions. A standard
medical gamma camera was used to monitor pulse-inputs of ~15MBq 99m Tc under saturated conditions and at
a constant flow of 0.33ml/min. Dynamic gamma imaging was conducted every 30s for 2 hours. Spatial
moments analysis of the resulting 99m Tc plume provided information about the relative changes in mass
distribution of the radionuclide in the various cement materials. 99m Tc advected through NRVB demonstrated
typical conservative transport behaviour, while OPC and OPC/PFA produced a slight reduction in 99m Tc
centre of mass transport velocity over time. BFS-containing cement was shown to be most effective at
immobilising 99m Tc under dynamic, rapid-flow conditions, with up to 50% of the injected activity retained
irreversibly by the cement, indicating that the determined sorption mechanism has a significant effect on the
transport of technetium.
PA1-12
The effect of ADVA Cast 551 superplasticiser on radionuclide solubility
R. Beard 1 , A. P. Clacher 2 , and M. M. Cowper 3
1 NDA RWMD, NDA Harwell Office, Building 587, Curie Avenue, Harwell, Didcot, OX11 0RH
2 AMEC Analytical Services, 601 Faraday Street, Birchwood Park, Warrington, WA3 6GN
3 AMEC Nuclear Limited, Building 150, Harwell, Didcot, OX11 0QB
The addition of superplasticisers to cementitious materials is often employed to improve the workability of
the wet material and potentially allow a reduction in the quantity of water required in the mixture.
Cementitious materials are used as waste-conditioning grouts for intermediate-level wastes in the UK and
would be part of the engineered barrier system for the resulting waste packages in a geological disposal
facility (GDF). If cement additives were to be included in the formulations of such materials, it is important
to understand the effect they might have on the behaviour of radionuclides after closure of a GDF.
Potentially radionuclides could form complexes with the additives (or their degradation products), resulting
in increases in solubility or decreases in sorption. Previous work in the UK with a variety of available
superplasticisers in the 1980/90s showed large increases in solubility for some radionuclides under alkaline
conditions.
The newer generation comb polymer superplasticisers now available offer the potential for better
characterisation and understanding of their impact on radionuclide behaviour and are being investigated in
the UK. AMEC is performing a programme of work for the UK Nuclear Decommissioning Authority (NDA)
Radioactive Waste Management Directorate (RWMD) involving one of these, ADVA Cast 551, as an
example of a typical superplasticiser. Initial studies were designed to see whether the solubility increases
observed with the older superplasticisers in the 1980/90s were also observed with ADVA Cast 551.
The first study 1 investigated the solubilities of Pu(IV) and U(VI) in irradiated and unirradiated Ca(OH) 2
solution with 0.01%, 0.1%, 0.3% and 1% w/v concentrations of ADVA Cast 551 (‘free solution tests’).
These data were compared to solubilities measured in Ca(OH) 2 in the absence of superplasticiser as a
baseline. For U(VI) solubility increased by up to three orders of magnitude (not solubility-limited) in the
unirradiated 1% w/v solution but there was no increase in the 0.01% w/v solution after 30,000 nominal

molecular weight cut-off (NMWCO) filtration compared to the baseline. Pu(IV) solubility increased by three
or four orders of magnitude at both 1% and 0.01% w/v concentrations of ADVA Cast 551. These were not
solubility-limited, Irradiation to 1 MGy of 0.01 and 1% w/v ADVA Cast 551 solutions resulted in decreases
in total organic carbon content and reduced increases in solubility for both Pu(IV) and U(VI) at both
loadings.
A subsequent study 2 extended the programme to Ni(II), Am(III), Th(IV) and Nb(V). For Ni(II), Am(III) and
Th(IV), significant solubility increases of two to four orders of magnitude (in some cases these were
inventory limited) were found over the range of ADVA Cast 551 concentrations following 10,000 NMWCO
filtration. For Nb(V), smaller solubility increases were observed, which were found to be equilibration time
dependent; this was attributed to the rates of formation of an solubility-limiting calcium niobate phase.
Subsequent to those studies a more detailed examination of the effects of ADVA Cast 551 under conditions
that are may be more representative of release from waste encapsulation grouts has been undertaken. OPCbased
cement formulations (CEM I 42.5 N Ordinary Portland Cement and Ground Granulated Blast-furnace
Slag at mass ratios of 1:3 and 1:9 containing 0.5% by mass ADVA Cast 551). After a period of curing, the
cements were leached as either crushed samples or monolith in de-ionised water for up to 6 months. In some
cases, porewaters were also squeezed from cement monoliths. The effects of these leachates and porewater
solutions on the solubilities of U(VI), Am(III), Th(IV) and Ni(II) have been investigated.. These results will
be discussed and compared with the observations made in the ‘free solution tests’ in order to understand the
impact of retention or degradation of the superplasticiser components in cement matrices on radionuclide
behaviour.
The authors wish to thank the UK Nuclear Decommissioning Authority for funding this work.
1
Effect of ADVA Cast 551 on the Solubility of Plutonium (IV) and Uranium (VI). A.P. Clacher and M.M. Cowper,
Serco/TAS/003145/001 Issue 2, 2011
2
Solubility Studies: Effect of Adva Cast 551 in Low Concentration, A.P. Clacher, T. Marshall and S.W. Swanton,
Serco, NPO004236, December 2010
PA1-13
ELICITATION OF URANIUM SOLUBILITY TO SUPPORT THE DISPOSAL OF U 3 O 8
T. Beattie 1)* M. Couch 2) , C. P. Jackson 2)
1) MCM-International, Baden, Switzerland
2) Amec, Building 150, RSRL Harwell, Didcot, Oxfordshire, UK
Uranium, with a range of U-235 enrichments, is stored at a number of nuclear sites in the UK. A major
component of this material includes depleted uranium (DU), with a U-235 content less than ~0.7% formed as
a by-product of the uranium enrichment process used in the manufacture of nuclear fuels.
This material is not currently declared as waste, but is considered in research and performance assessments
undertaken by the UK Nuclear Decommissioning Authority (NDA) because it may be declared as waste in
the future and is therefore included in the UK Radioactive Waste Baseline Inventory. Current planning
assumptions consider that uranium (currently stored as UF 6 ) would be converted to oxide (mainly U 3 O 8 , with
small amounts of UO 2 and UO 2 F 2 ), compacted and disposed in the cementitious ILW disposal area of a
geological disposal facility [1]. These packaging assumptions are consistent with the MRWS White Paper [2],
but it is recognised that the actual packaging may be different in a disposal facility tailored to a specific site.
As part of the NDA generic research programme, elicitation of the solubility of U 3 O 8 in the near field was
undertaken for two pH ranges (12.3 to 13.5 and 9.0 to 12.3) [3]. As in previous work [4], a formal structured
approach was used to elicit probability distributions with a group of experts using a pre-prepared uranium
data pack of internationally available uranium solubility data of relevance to the conditions under
consideration [5].
This paper describes the main discussion held during the stage of encoding the uncertainties for the
quantities of interest, particularly regarding the consideration of solubility limiting secondary phases (such as

calcium uranate and calcium uranium silicates) and how such phases would evolve and be altered by the
redox potential conditions during the late post-closure period. The discussion culminated in expert agreement
that previously elicited distributions for uranium would be appropriate to use for U 3 O 8 to support the NDA
2010 generic Post-closure performance assessments [5]. However, in preparation for future site-specific
safety assessments, work should be undertaken to prepare a quantified description of redox potential, rather
than considering the solubility of uranium(IV) and uranium(VI) separately. This recommendation is being
taken forward as part of a wider programme of work initiated by the NDA (Uranium Integrated Project
Team).
This work highlighted the importance of the continued production and peer reviewed publication of high
quality solubility data at measured redox potentials for uranium secondary phases under in cementitious
disposal conditions. It also highlighted the requirements for greater understanding and representation of the
complexity in the treatment of uranium secondary phase formation in relation to calcium availability in the
disposal system.
[1] Nuclear Decommissioning Authority, Geological Disposal:Generic Disposal SystemTechnical Specification, NDA
Report NDA/RWMD/044, 2010.
[2] Defra, BERR, Welsh Assembly Government, Department of the Environment Northern Ireland, Managing
Radioactive Waste Safely: A framework for implementing geological disposal, Cm7386, June 2008. ISBN 0101738625.
[3] M. Couch, C. P. Jackson, Elicitation of the solubility of U 3 O 8 in the near field, Serco report SERCO/E.004133/001,
In Publication.
[4] D. Swan, C. P. Jackson, Formal Structured Data Elicitation for Uranium Solubility in the Near Field, Serco report
SERCO/ENV/0290, 2007.
[5] Nuclear Decommissioning Authority, Geological Disposal: Post-closure Safety Assessment, NDA Report
NDA/RWMD/030, 2010.

PA3
PA3-1
PA3-2
PA3-3
PA3-4
PA3-5
PA3-6
PA3-7
PA3-8
PA3-9
PA3-10
COMPLEXATION WITH INORGANIC AND ORGANIC
LIGANDS
CHAINS AND LAYERS OF URANYL FORMATES TEMPLATED BY PROTONATED
DIAMINES
Qianqian Zhu, Luhua Wang, Chunli Liu, Zheming Wang (China)
THERMODYNAMIC STUDIES OF COMPLEX FORMATION OF TRIVALENT Nd, Am
BY SMALL ORGANIC LIGANDS WITH MICRO TITRATION CALORIMETRY
M. Acker, M. Müller, S. Taut, A. Barkleit, J. Schott, G. Bernhard (Germany)
TWO SYSTEMS OF [DabcoH 2 ] 2+ /[PipH 2 ] 2+ -URANYL-OXALATE SHOWING
REVERSIBLE CRYSTAL TO CRYSTAL TRANSFORMATIONS CONTROLLED BY
THE DIAMMONIUM/URANYL/OXALATE RATIOS IN AQUEOUS SOLUTIONS
([DabcoH 2 ] = 1,4-Diazabicyclo-[2.2.2]-OctaneH 2 and [PipH 2 ] = PiperazineH 2 )
L. H. Wang, R. Shang, Z. Zheng, C. L. Liu, Z. M. Wang (China)
THE THERMODYNAMICS OF THE COMPLEXATION OF Cm(III) WITH SMALL
ORGANIC LIGANDS UNDER SALINE CONDITIONS AND INCREASED
TEMPERATURES
A. Skerencak-Frech, D.R. Fröhlich,, P.J. Panak (Germany)
EFFECT OF ISOSACCHARINIC ACID ON THE SORPTION OF EUROPIUM(III) AND
PLUTONIUM(IV) ON CEMENT CSH PHASES
M. García-Gutiérrez , T. Missana , H. Rojo , H. Galán, M. Mingarro, U. Alonso (Spain,
Switzerland)
SOLUBILITY AND NMR STUDIES OF Ca-GLUCONATE AND C-Np(IV)-GLUCONATE
SYSTEMS IN DILUTE TO CONCENTRATED ALKALINE CaCl 2 SOLUTION
X. Gaona, C. Adam, H. Rojo, M. Böttle, P. Kaden, M. Altmaier (Germany)
COMPLEXATION OF Nd(III)/Cm(III) WITH GLUCONATE IN ALKALINE NaCl, AND
CaCl 2 SOLUTIONS: SOLUBILITY AND TRLFS STUDIES
H. Rojo, X. Gaona, Th. Rabung, M. Garcia, T. Missana, M. Altmaier (Germany, Spain)
A STUDY OF TERNARY Ca-UO 2 -CO 3 COMPLEXATION UNDER NEUTRAL TO
WEAKLY ALKALINE CONDITIONS
J.-Y. Lee, J.-I. Yun (Korea)
USE OF NMR TO DETERMINE STRUCTURE OF PLUTONIUM SIDEROPHORE
COMPLEXES: Pu(IV)(H 2 DFOB) and
DFOBPu(IV)di-μ-OH-DFOBPu(IV)
M.A. Boggs, M. Zavarin, A.B. Kersting (USA)
THE BIGRAD CONSORTIUM - MONITORING REDOX SPECIATION OF URANIUM
USING OPTICAL FINGERPRINTING
L.S. Natrajan, A.N. Swinburne, S.D. Woodall , S. Randall, K. Smith,
N.D. Bryan, K. Morris (UK)

PA3-1
CHAINS AND LAYERS OF URANYL FORMATES TEMPLATED BY PROTONATED DIAMINES
Qianqian Zhu, Luhua Wang, Chunli Liu*, Zheming Wang*
Beijing National Laboratory of Molecular Sciences, Radiochemistry and Radiation Chemistry Key
Laboratory for Fundamental Sciences, State Key Laboratory of Rare Earth Materials Chemistry and
Applications, College of Chemistry and Molecular Engineering,
Peking University, Beijing 100871, China.
Fax: 86-10-62751708, E-mail: liucl@pku.edu.cn, zmw@pku.edu.cn
The development of the coordination chemistry of uranyl in recent years has led to a richness in structures of
such compounds [1]. While many kinds of ligands, such as phosphates and carboxylates, have been widely
investigated, the smallest and simplest carboxylate group, formate, has rarely been explored [2], and the
reason was unknown. The formate ligand possesses all the functionalities of carboxylate, showing different
bridging coordination modes, anti-anti, syn-anti, etc., and it has a small steric effect. These are beneficial for
the formation of extended structures [3]. As a natural extension from transition metal and lanthanide formate
systems to uranium ones [3, 4], we have started to investigate the ammonium-uranyl-formate systems. Here
we report four uranyl formate compounds, in which three are formulated as [AH 2 ][U 2 O 4 (HCOO) 6 ] for 1,
AH 2 = H 2 NCH 2 CH 2 NH 3 , 2, AH 2 = CH 3 NH(CH 2 CH 2 ) 2 NHCH 3 , and 3, AH 2 = (CH 3 ) 2 NHCH 2 CH 2 NH(CH 3 ) 2 ,
and 4 is [(CH 3 ) 2 NHCH 2 CH 2 CH 2 NH 3 ][U 4 O 8 (HCOO) 10 ]. 1 and 2 possess anionic uranyl formate chains, with
an arc section in 1 but flat one in 2, though the uranyl-formate-uranyl linkages are the same (Figure 1a, 1b). 3
and 4 possess anionic uranyl formate layers showing different uranyl-formate-uranyl linkages or network
topologies (Figure 1c, 1d). In all of the four compounds, the uranyl ion is coordinated by five oxygen atoms,
each from one formate ligand in the equatorial plane, resulting in the typical pentagonal bipyramid.
The anionic chains of 1 and 2 could be described as two uranyl arrays, in which the uranyl ions are
connected by anti-anti formate bridges along the chain direction, and cross-linked by syn-anti formate
bridges in 1, resulting in an arc section of the chain, but anti-anti ones in 2, showing flat section. The
structures could also be considered as [(UO 2 ) 3 (HCOO) 3 ] triangles sharing their two edges, with each uranyl
having an extra terminal formate ligand. The uranyl-formate layer of 3 possesses dimeric units of
[(UO 2 ) 2 (HCOO) 2 ] in which the two uranyl ions are bridged by two syn-anti formate ligands, and each
dimeric unit further connects four neighborhood ones through formate bridges, resulting a (4, 4) network if
the dimers are considered as the nodes. However, if each uranyl is as a node, the network is a honeycomb or
(6, 3) one. In 3, each uranyl has an extra terminal formate ligand. The layer of 4 consists of
[(UO 2 ) 3 (HCOO) 3 ] triangles and [(UO 2 ) 4 (HCOO) 4 ] squares. By sharing their formate edges, anti-anti or synanti,
the triangles and squares form a 3 3 4 3 5 4 network in which each uranyl is a 5-connected node. The
protonated diamine cations locate in the inter-chain or inter-layer region, and their NH donors form hydrogen
bonds with the formate ligands in the anionic uranyl-formate components. These results clearly demonstrate
that the anionic uranyl formate structures could be modulated or templated by the protonated diamine
cations, as what have been well documented for the transition metal and lanthanide systems [2, 3], and these
imply the promising potentials and opportunity in the research of such systems.

Figure 1. The extended anionic uranyl-formate structures templated by different protonated diamines: chains
in 1 (a) and 2 (b) and layers in 3 (c) and 4 (d).
[1] (a) Andrews, M. B.; Cahill, C. L., Chem. Rev., 2013, 113 (2), 1121-1136; (b) Jones, M. B.; Gaunt, A. J., Chem. Rev.,
2013, 113 (2), 1137-1198; (c) Qiu, J.; Burns, P. C., Chem. Rev., 2013, 113 (2), 1097-1120.
[2] Wang, Z.; Hu, K.; Gao, S.; Kobayashi, H., Adv. Mater., 2010, 22 (13), 1526-1533.
[3] Thuery, P., Inorg. Chem. Commun., 2008, 11 (2008), 616–620.
[4] (a) Li, M.; Liu, B.; Wang, B.; Wang, Z.; Gao, S.; Kurmoo, M., Dalton Trans., 2011, 40 (22), 6038-6046; (b) Liu, B.;
Zheng, H.-B.; Wang, Z.-M.; Gao, S., Crystengcomm 2011, 13 (17), 5285-5288.
PA3-2
THERMODYNAMIC STUDIES OF COMPLEX FORMATION OF TRIVALENT Nd, Am BY SMALL
ORGANIC LIGANDS WITH MICRO TITRATION CALORIMETRY
M. Acker (1) , M. Müller (1,3) , S. Taut (1) A. Barkleit (2,3) , J. Schott (2) , G. Bernhard (2,3)
1 Technische Universität Dresden, Central Radionuclide Laboratory, Dresden, Germany
2 Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology, Dresden, Germany
3 Technische Universität Dresden, Department of Chemistry and Food Chemistry, Radiochemistry, Dresden,
Germany
It is well known, that natural organic matter (NOM) like macromolecules, e.g. humic and fulvic substances,
or bioligands, like siderophores as well as water soluble small molecules, like acetate, citrate, tartrate or
lactate can significantly influence the radionuclide migration due to their complexation properties. Thus, a
full thermodynamic description of the interactions between radionuclides and NOM is of fundamental
interest for modelling and prediction of radionuclide transport in the near and far field of a nuclear waste
repository as well as in the environment.
Up to now, most of the data in thermodynamic databases consist of the equilibrium constants of actinide
interactions with some organic ligands. And most of them are available only at standard temperature at 25 °C

and low ionic strength (< 1m). There is a significant lack of the fundamental data regarding the complexation
of actinides with NOM at elevated temperatures and at higher ionic strengths.
In principle, missing data can be obtained by extrapolation. The well-known van’t Hoff approach is often
used to extrapolate complexation constants to higher temperature given that the complexation enthalpies at
standard temperature of 25 °C are known. In few cases where complexation constants are available for
different temperatures the complexation enthalpy can be derived from the van’t Hoff plot, i.e. a plot of ln K
vs. 1/T. In both cases, a constant complexation enthalpy over the considered temperature range is assumed.
However, such extrapolations may result in large uncertainties because (i) the complexation enthalpies can
be strongly temperature-dependent, (ii) the temperature range of extrapolation is far too wide, or (iii) the
used thermodynamic data (equilibrium constants) are subject to major uncertainties [1].
Titration calorimetry is the method of choice to determine precise complexation enthalpies at elevated
temperatures and ionic strengths. However, its application to actinide complexation had been limited due to
the inappropriate technical design of most commercial titration calorimeters which need large sample
volumes (> 1,5 ml) making them not applicable for handling of actinide solutions. A further limitation is the
analysis of the titration curves, which can be very complex and require a lot of experience.
In the present work we describe the capabilities and limitations of a commercial micro titration calorimeter.
Due to the very small reaction volume of only 200 µl, it is suitable for the handling of actinides and to study
their complexation behavior with small organic ligands at elevated temperatures and ionic strength. The
reaction volume and the calorimetric detection limit (~ 1 µJ) together with the magnitude of ∆H and
equilibrium constant determine the upper and lower limit of equilibrium constant that can be determined. In
our case the lower limit of equilibrium constant is round log K~2.
We investigate the complex formation of Nd(III) as inactive chemical analogue for trivalent actinides with
tartrate and malonate at different ionic strength (0.25-3mol dm -3 NaCl) and different temperature (25-60°C)
and compared them with data already available for lanthanides in the literature [e.g. 2,3,4,5,6].
Multiple calorimetric titrations with varying concentrations of metal ion and ligand at different temperature
and ionic strength were carried out to investigate the influence of the species distribution on the accuracy and
precisions of the determined enthalpy values. An example of such calorimetric titration curve of Nd(III)
malonate complexation for given condition is shown in figure 1.
From these experiments a set of basic titration experiments was extracted, which are absolutely necessary to
determine the Am(III) complexation enthalpies with both ligands at acceptable uncertainties and at the
lowest possible Am(III) concentration.
The calorimetric titration curves were analysed with the program HYP∆H [7] to obtain the enthalpies for
given target reactions. Additionally, for both metal ions the species distribution and complexation constants
were determined and verified by UV/Vis-absorption spectrometric titrations using a long path capillary cell
(0.5 and 1 m path length) for low metal concentration under the same conditions as the calorimetric titration
experiments [8].
[1] L. Rao, Chem. Soc. Rev. 36, 881-892 (2007)
[2] H. Kitano et. al. Radiochim. Acta 94, 541-547 (2006)
[3] V. K. Rao et. al. Inorg. Chim. Acta,128, 131-134 (1987)
[4] P. Thakur et. al. J Solution Chem, 38, 265-287 (2009)
[5] S. Ramamoorthy, J. Inorg. Nucl. Chem. 34, 1977-1987 (1972)
[6] A. E. Martell, NIST Critically Selected Constants of Metal Complexes, Ver 7, Texas A&M University, Texas 2003
[7] P. Gans et. al. J. Solution Chem. 37, 467-476 (2008)
[8] M. Müller et. al. J. of Radioanal. and Nucl. Chem. 286, 175-180 (2010)

83,72
Time (min)
0 10 20 30 40 50 60 70 80 90 100 110 120
62,79
µWatts/sec
41,86
20,93
0,00
kJ mol -1 of injectant
10,00
7,50
5,00
2,50
0,00
0,0 0,5 1,0 1,5 2,0 2,5 3,0 3,5
Molar Ratio (malonate/Nd)
Figure 1. Calorimetric titration of the Nd malonate system (T = 25°C, I = 2.0 m NaCl, pcH= 6, titrant:
100 mM malonic acid/ Na malonate, sample: 6 mM Nd)
PA3-3
TWO SYSTEMS OF [DABCOH 2 ] 2+ /[PIPH 2 ] 2+ -URANYL-OXALATE SHOWING REVERSIBLE
CRYSTAL TO CRYSTAL TRANSFORMATIONS CONTROLLED BY THE
DIAMMONIUM/URANYL/OXALATE RATIOS IN AQUEOUS SOLUTIONS ([DABCOH 2 ] = 1,4-
DIAZABICYCLO-[2.2.2]-OCTANEH 2 AND [PIPH 2 ] = PIPERAZINEH 2 )
L. H. Wang, R. Shang, Z. Zheng, C. L. Liu*, Z. M. Wang*
Beijing National Laboratory of Molecular Sciences, Radiochemistry and Radiation Chemistry Key
Laboratory for Fundamental Sciences, State Key Laboratory of Rare Earth Materials Chemistry and
Applications, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871, P. R.
China
Crystal to crystal transformation has attracted much attention in recent years [1-4]. The structural
alternations such as mononuclear to polynuclear species and change in structural dimensionality [2],
framework breathing and layer sliding [3], and the associated changes in physical properties [2, 3] resulting
from such alterations are of significant potential in applications, as well as being of fundamental scientific
interest. In many such cases, the transformations are the formation of the new phases through the reassembly
of the tectons [4]. Here we present two systems of [dabcoH 2 ] 2+ -uranyl-oxalate and [pipH 2 ] 2+ -uranyl-oxalate
in which [dabcoH 2 ] 2+ and [pipH 2 ] 2+ are cations of doubly protonated 1,4-diazabicyclo-[2.2.2]-octane (dabco)
and piperazine (pip) respectively. Each system yielded two different crystals and showed the reversible
crystal to crystal transformations between them in aqueous solutions, to be controlled by the ratio of
reactants or tectons (Figure 1). The four compounds in pairs are [dabcoH 2 ][UO 2 (C 2 O 4 ) 2 (H 2 O)]·2H 2 O (dabco1)
and [dabcoH 2 ][(UO 2 ) 2 (C 2 O 4 ) 3 (H 2 O) 2 ]·2H 2 O (dabco2), and [pipH 2 ][UO 2 (C 2 O 4 ) 2 (H 2 O)]·4H 2 O (pip1) and

[pipH 2 ][(UO 2 ) 2 (C 2 O 4 ) 3 (H 2 O) 2 ]·2H 2 O (pip2), and their structures were determined by single crystal X-ray
diffraction.
Besides the cations and lattice water, dabco1 and pip1 contain mononuclear anions of
[UO 2 (C 2 O 4 ) 2 (H 2 O)] 2− , while dabco2 and pip2 possess dinuclear anions of [(UO 2 ) 2 (C 2 O 4 ) 3 (H 2 O) 2 ] 2− , and in all
structures uranium ion shows pentagonal bipyramid environment made of equatorial oxalate, water and
apical oxygen. The needle-like crystals of dabco1 belong to the chiral space group P 6 5 22. In the structure
the [UO 2 (C 2 O 4 ) 2 (H 2 O)] 2− anions form anionic helixes along 6 5 axis and they are further linked by N−H···O
H-bonds between the inter-helix [dabcoH 2 ] 2+ cation and oxalates of the anion. Disorder of lattice water and
[dabcoH 2 ] 2+ cation is observed in dabco1. The thin plate crystals of pip1 in space group P⎺1 possesses a
lamellar structure, with dense layers of [pipH 2 ] 2+·2[UO 2 (C 2 O 4 ) 2 (H 2 O)] 2− pillared by other
crystallographically unique [pipH 2 ] 2+ cations, and the structure contains lattice water forming branched
zigzag water chains. Block crystals of dabco2 and pip2, in monoclinic space group C 2/c and P 2 1 /c,
respectively, are of similar layer-like structures, possessing the dinuclear [(UO 2 ) 2 (C 2 O 4 ) 3 (H 2 O) 2 ] 2− anion with
one tetradentate oxalate bridge. The anions and diammonium cations form chains by N−H···O H-bonds
between them then further form stacked layers via O−H···O H-bonds among the coordinated and lattice
water and the oxalate ligands. The ratio of diammonium/uranyl/oxalate in the reaction systems controlled the
final outcomes and the occurrence of transformation, no transformation, and reverse transformation, as
monitored by time-dependent snapshots and powder X-ray diffraction patterns. The thermal stability, UV-
Vis, IR and Raman spectra, and luminescence were also investigated.
Figure 1. The morphologies of crystals of the four compounds, dabco1 (top left), dabco2 (bottom left), pip1
(top right) and pip2 (bottom right) and the transformations controlled by the diammonium/uranyl/oxalate
ratios.
[1] Desiraju, G. R. Angew. Chem., Int. Ed. 2007, 46, 8342.
[2] (a) Duan, Z. M.; Zhang, Y.; Zhang, B.; Zhu, D. B. J. Am. Chem. Soc. 2009, 131, 6934. (b) Sun, J.; Dai, F. N.; Yuan,
W. B.; Bi, W. H.; Zhao, X. L.; Sun, W. M.; Sun, D. F. Angew. Chem., Int. Ed. 2011, 50, 7061.
[3] (a) Serre, C.; Mellot-Draznieks, C.; Surblé, S.; Audebrand, N.; Filinchuk, Y.; Férey, G. Science 2007, 315, 1828. (b)
Aijaz, A.; Sañudo, E. C.; Bharadwaj, P. K. Cryst. Growth Des. 2011, 11, 1122.
[4] (a) Grohol, D.; Clearfield, A. J. Am. Chem. Soc. 1997, 119, 4662. (b) Andrews, M. B.; Cahill, C. L. Angew. Chem.,
Int. Edit. 2012, 51, 6631.

PA3-4
THE THERMODYNAMICS OF THE COMPLEXATION OF Cm(III) WITH SMALL ORGANIC
LIGANDS UNDER SALINE CONDITIONS AND INCREASED TEMPERATURES
A. Skerencak-Frech a,b , D.R. Fröhlich a,b , P.J. Panak a,b
a Institut für Nukleare Entsorgung, Karlsruher Institut für Technologie, 76021 Karlsruhe, Germany
b
Physikalisch-Chemisches Institut, Universität Heidelberg, 69120 Heidelberg, Germany
The long-term safety assessment of a nuclear waste repository in deep geological formations requires a wellfounded
thermodynamic description of the processes relevant for the migration and retention of the actinides.
In the past, a broad variety of complexation reactions of actinides in their most important oxidation states
were studied and the according stability constants (log β 0 n) as well as the thermodynamic data (Δ r H 0 m, Δ r S 0 m)
are listed in thermodynamic databases.[1,2] However, the majority of this data is valid only for 25°C. Due to
the radioactive decay, temperatures above 25°C will prevail in the near field of a nuclear waste repository.
This will have a distinct impact on the geochemistry of the actinides. Thus, thermodynamic data for the
actinides at T ≥ 25°C is of high importance for a comprehensive long-term safety assessment. Besides rock
salt and granite, clay formations are considered as potential host rocks for a nuclear waste repository. Small
organic ligands (e.g. acetate (≤ 1865 μM), propionate (≤ 127 μM), etc.) are present in the pore water of clay
rocks.[3,4] Additionally, the pore waters of different origins vary considerably in their ionic strength and
electrolyte composition (NaCl, CaCl 2 , MgCl 2 , etc.). Only little information is available on the complexation
of trivalent actinides with carboxylic acids under high salinity and at increased temperature.
In the present work the formation of [Cm(L) n ] 3-n complexes (L = propionate, acetate; n = 1, 2) is studied in
different ionic media (NaClO 4 , NaCl, CaCl 2 ) and T = 20 to 90°C. Using time resolved laser fluorescence
spectroscopy (TRLFS), the molar fractions of the individual Cm(III) species are determined as a function of
the temperature. The results of the Cm(III) propionate system at T = 20 and 90°C are displayed in figure 1.
Fig. 1 Molar fractions of the [Cm(Prop) n ] 3-n species (n = 0,1,2) in aqueous NaClO 4 , NaCl and CaCl 2
at 20 and 90°C.
At 25°C the speciation is governed mainly by Cm 3+ and the [Cm(Prop)] 2+ complex. The [Cm(Prop) 2 ] +
complex is formed only at higher ligand concentrations and remains a minor species. As the temperature
increases the chemical equilibrium is shifted towards the higher complexed species. Thus, the Cm 3+ aquo-ion
decreases rapidly and the [Cm(Prop) 2 ] + is the dominating species at higher [Prop - ]. The small differences in

the speciation of the NaClO 4 and NaCl media are attributed solely to the different activity coefficients of the
reactive species. However, in the CaCl 2 medium considerably higher ligand concentrations are required for
the formation of the complexed species. This effect is attributed to the additional formation of a [Ca(Prop)] +
species, which reduces the amount of the free ligand in solution. Using the experimental speciation, the
conditional stability constants (log K’ n (T)) are calculated and extrapolated to zero ionic strength with the
specific ion interaction theory (SIT), yielding the thermodynamic log K 0 n(T) values (figure 2). Furthermore,
the Δε n (T) values are determined.
Fig. 2 Thermodynamic stability constants of [Cm(L) n ] 3-n complexes (L = propionate, acetate, n = 1,
2) in aqueous NaClO 4 , NaCl and CaCl 2 at T = 20 - 90°C.
The log K 0 1(T) and log K 0 2(T) values of the Cm(III) propionate and acetate complexes are in very good
agreement. The log K 0 1(T) values increase slightly by 0.3 orders of magnitude with the temperature, whereas
the log K 0 2(T) values show a distinctively stronger temperature dependency and increase by almost one order
of magnitude. At 25°C, the log K 0 1(T) and log K 0 2(T) values of the Cm(III) propionate and acetate complexes
are by about 0.4 higher than the respective log K 0 n([Eu(L) n ] 3-n ) values.[5] The temperature dependency of the
stability constants is fitted by the integrated van’t Hoff equation, yielding the thermodynamic constants of
the respective complexation reactions (Δ r H 0 m, Δ r S 0 m). The results show positive values for Δ r H 0 m and Δ r S 0 m.
Thus, the formation of the complexation of Cm(III) with carboxylic acids is entropy driven.
The present work provides a detailed thermodynamic study of the complexation of Cm(III) with small
organic ligands under saline conditions and at increased temperatures. The results give new insights into the
geochemistry of trivalent actinides under near field conditions and are important for the modeling of their
migration behavior in clay rock formations. Thus, the thermodynamic data determined here are of high value
for a comprehensive long-term safety assessment of a nuclear waste repository.
[1] Guillaumont, R, et. al., OECD, NEA-TDB. Chemical Thermodynamics Vol. 5. Update on the Chemical
Thermodynamics of Uranium, Neptunium, Plutonium, Americium and Technetium. Elsevier , Amsterdam, (2003).
[2] Hummel, W., et. al., Nagra/PSI Chemical Thermodynamic Data Base, 01/01, (2001).
[3] Courdouan, A., et. al., Appl. Geochem., 22, 1537-1548, (2007).
[4] Courdouan, A., et. al., Appl. Geochem., 22, 2926-2939, (2007).
[5] Wood, S.A., Eng. Geo., 34, 229-259, (1993).

PA3-5
EFFECT OF ISOSACCHARINIC ACID ON THE SORPTION OF EUROPIUM(III) AND
PLUTONIUM(IV) ON CEMENT CSH PHASES.
M. García-Gutiérrez 1) , T. Missana 1) , H. Rojo 2) , H. Galán 1) , M. Mingarro 1) , U. Alonso 1)
(1) CIEMAT, Dpt. of Environment, Av. Complutense 40, 28040 Madrid (SPAIN)
(2) PSI, Laboratory for Waste Management, Paul Scherrer Institut 5232 Villigen (Switzerland)
Cement materials are widely used for stabilizing and conditioning of radioactive waste in low- and
intermediate-level repositories. Cement determines the chemical and physical properties of the repository
near-field for a long period of time, and provides alkaline conditions, which favour radionuclide
immobilization by sorption and low radionuclide solubility. Organic compounds as tissues, cotton or paper
(i.e. cellulosic material) present with the radioactive waste are degraded in these alkaline conditions. The
main alkaline degradation product of the cellulose products is the isosaccharinic acid (ISA) that has a high
capacity to form organic complexes with the radionuclide. These complexes are generally very stable and
can enhance the radionuclide mobility [1, 2, 3]. Therefore, it is important to analyse its behaviour in the
frame of radioactive waste repositories.
Ca(ISA) 2 was synthesized by lactose alkaline degradation in saturated aqueous Ca(OH) 2 as described by
Whistler & BeMiller [4]. The NaISA form was obtained by treating the Ca(ISA) 2 with an ion exchange resin
in the Na + form.
The Eu(III) and Pu(IV) radioactive elements were chosen as representative of trivalent lanthanides an
tetravalent actinides, respectively. Under reducing conditions, Pu(IV) as Pu(OH) 4 (aq) is the species
dominating the plutonium aqueous speciation. The 152 Eu activity was measured by γ-counting (CobraII,
Packard) and 238 Pu activity by liquid scintillation counting (2700TR TriCarb, Packard).
Calcium silicate hydrates (CSH) phases are the most important constituents, in volume, of hydrated cement
paste and present a high specific surface. Synthetic CSH phases with different CaO:SiO 2 mol ratios (1.5, 1.0
and 0.8), simulating the cement paste evolution (from non-degraded to degraded cement) were prepared in a
glove box under N 2 atmosphere to avoid CO 2 effects. The pH of the waters in equilibrium with the three
CSH phases used in the experiments was 12.7, 11.8 and 10.5 respectively.
Batch-type experiments, in glove box under N 2 atmosphere, have been performed to investigate the effect of
ISA concentration on the sorption of 152 Eu(III) and 238 Pu(IV) on CSH phases. First, sorption kinetic studies
were carried out with the three CSH phases to obtain the equilibrium time. A very short time was needed for
the equilibrium with 238 Pu. After, sorption isotherms with different ISA concentration (in a range from 10 -6 to
10 -2 M, and without ISA), were performed.
High ISA concentrations in solution, around 10 -4 M, are able to significantly decrease the sorption capacity
of Pu(IV) and Eu(III), two or three order of magnitude in the distribution coefficient, on CSH phases. This
means that the complex Pu(IV)-ISA and Eu(III)-ISA formed, are very stable, and can enhance the mobility
of both radionuclides.
Acknowledgements: This research is funded by the Spanish Government (Project CeluCem CTQ2011-
28338)
[1] Vercammen, K., Glaus, M.A., Van Loon, L.R. (2001) Complexation of Th(IV) and Eu(III) by α-isosaccharinic acid
under alkaline conditions”, Radiochim. Acta 89, 393-401.
[2] Tits, J., Wieland, E., Bradbury, M.H. (2005) The effect of isosaccharinic acid and gluconic acid on the retention of
Eu(III), Am(III) and Th(IV) by calcite. Applied Geochemistry 20, 2082-2096.
[3] Hakanen, M. and Ervanne, H. (2006) The influence of Organic Cement additives on Radionuclide Mobility. A
literature Survey. POSIVA Working Report 2006-06.
[4] Whistler, R.L. and BeMiller, J.N. (1963) In: Methods in Carbohydrate Chemistry, Vol. 2: Reactions of
Carbohydrates. M.L. Wolfrom and J.N. BeMiller Eds. Academic Press, New York, p. 477-479.

PA3-6
SOLUBILITY AND NMR STUDIES OF Ca–GLUCONATE AND Ca–Np(IV)–GLUCONATE
SYSTEMS IN DILUTE TO CONCENTRATED ALKALINE CaCl 2 SOLUTIONS
X. Gaona 1 , C. Adam 1 , H. Rojo 1,2 , M. Böttle 1 , P. Kaden 1 , M. Altmaier 1
1 Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Karlsruhe, Germany
2 Laboratory for Waste Management, Paul Scherrer Institut, Villigen PSI, Switzerland
Radionuclide solubility and sorption in cementitious and saline systems can be affected by the presence of
organic ligands. Gluconic acid (GLU, C 6 H 12 O 7 ) is a poly–hydroxocarboxylic acid expected in repositories
for low and intermediate–level radioactive waste (LILW) as component of cementitious materials. The
formation of very stable An(IV)–GLU complexes has been reported in the literature. The stability of these
complexes can be further increased in the presence of Ca 2+ due to the formation of ternary Ca–An(IV)–GLU
complexes [1, 2]. All previous studies available in the literature have focussed on low ionic strengths and
low Ca 2+ concentrations. In CaCl 2 –rich brines, which can be generated by the corrosion of cementitious
waste forms in certain saline systems, the stabilization of binary Ca–GLU complexes may decrease the
availability of free GLU in solution and thus outcompete the formation of complexes with An(IV). The
development of complete and correct chemical, thermodynamic and activity models for the binary and
ternary systems Ca–GLU and Ca–An(IV)–GLU for dilute to concentrated CaCl 2 solutions is thus required to
properly assess the impact of this ligand on the mobilization of An(IV) under repository-relevant conditions.
The solubility of Ca(GLU) 2 (s) was investigated at 8 ≤ pH m ≤ 12 by analyzing [Ca] and [GLU] tot (as TOC).
The studies were complemented with 13 C–NMR measurements of the supernatant solution. The aqueous
complexation of calcium with GLU was further studied by 13 C–NMR in the absence of solid phase with
[GLU] = 0.02 M, pH m = 10 and varying [Ca] tot . In both approaches, dilute to concentrated NaCl (0.1–5.0 M)
and CaCl 2 (0.25–4.5 M) solutions were used as background electrolyte systems. NMR measurements were
performed on a Bruker Avance III 400 spectrometer operating at 100.63 MHz for 13 C and equipped with a
broadband observe probe head (BBFO).
Np(IV) was obtained by chemical reduction of a Np(V) stock solution with Na 2 S 2 O 4 0.05 M at pH = 8, and
quantitatively precipitated at pH m = 12 as NpO 2 (am,hyd). The resulting solid phase was divided in different
series (∼2 mg 237 Np per batch experiment) with 0.1–3.5 M CaCl 2 as background electrolyte, with 10 –6 ≤
[GLU] tot ≤ 10 –2 M and 9 ≤ pH m ≤ 12. All samples were prepared and stored at 22±2°C in Ar gloveboxes
under exclusion of oxygen and CO 2 .
13 C–NMR revealed clear shifts of C1–C6 resonance signal positions of gluconate with increasing [Ca]. The
observed shifts are consistent with the single formation of CaGLU + both in NaCl and CaCl 2 solutions with
[Ca] tot ≤ 0.5 M. Above this Ca concentration, the strong shift in C4 hints towards the formation of a new
species, likely a cyclic Ca–GLU structure with certain similarities to the lactone-ring reported to form under
acidic conditions. Ca(GLU) 2 (s) sets very high GLU solubility limits in both NaCl and CaCl 2 solutions. At
high CaCl 2 concentrations (
3.5 M), a
0.5 M.
The presence of GLU induces a significant increase on NpO 2 (am,hyd) solubility in CaCl 2 solutions (Figure
1). This increase is stronger at higher concentrations of Ca 2+ , thus indicating the formation of ternary
complexes analogous to those described in the literature for Th(IV) (CaAn IV (OH) 4 GLU 2 (aq)) [1, 2].
Independent of [CaCl 2 ] the system reaches a saturation level at [Np] 10
–5.5 M. This observation can be
explained by the formation of a new Ca–Np(IV)–GLU solid phase, although this hypothesis remains to be
further confirmed by appropriate solid phase characterization (XRD, SEM–EDS) currently under way.

log m Np
-3
-4
-5
-6
-7
-8
CaCl 2
0.1 M
0.25 M
1.0 M
2.0 M
4.5 M
slope 2
-9
NpO 2
(am,hyd) + 2H 2
O ⇔ Np(OH) 4
(aq) [NEA-TDB]
-10
-6 -5 -4 -3 -2 -1
log [GLU] tot
Figure 1. Solubility of NpO 2 (am,hyd) at pH m = pH max ∼12 under increasing [GLU] tot in 0.1–4.5 M CaCl 2
solutions.
[1] Tits, J., Wieland, E., Bradbury, M. (2005). The effect of isosaccharinic acid and gluconic acid on the retention of
Eu(III), Am(III) and Th(IV) by calcite. Applied Geochemistry, 20, 2082–2096
[2] Gaona X., Montoya, V., Colàs, E., Grivé, M., Duro, L. (2008). Review of the complexation of tetravalent actinides
by ISA and gluconate under alkaline to hyperalkaline conditions. Journal of Contaminant Hydrology, 102, 217-227.
PA3-7
COMPLEXATION OF Nd(III)/Cm(III) WITH GLUCONATE IN ALKALINE NaCl AND CaCl 2
SOLUTIONS: SOLUBILITY AND TRLFS STUDIES
H. Rojo 1,2 , X. Gaona 1 , Th. Rabung 1 , M. Garcia 3 , T. Missana 3 , M. Altmaier 1
1 Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Karlsruhe, Germany
2 Laboratory for Waste Management, Paul Scherrer Institut, Villigen PSI, Switzerland
3 CIEMAT, Research Centre for Energy, Environment and Technology, Madrid, Spain
Gluconic acid (GLU) is a polyhydroxy carboxylic acid expected in repositories for low and intermediate–
level radioactive waste as component of cementitious materials. GLU is also considered a close analogue of
iso-saccharinic acid (ISA), a polyhydroxy carboxylic ligand resulting from the degradation of cellulose under
alkaline conditions. The presence of these ligands can affect the solubility and sorption of radionuclides in
cementitious systems, which is of particular interest for assessing the safety of nuclear waste disposal.
Formation of very stable An(III)–GLU/ISA complexes has been reported in the literature, although in
contrast to An(IV), no ternary species Ca–An(III)–GLU/ISA have been described so far. The latter may play
a relevant role in cementitious and saline environments, where high Ca 2+ concentrations are expected in
certain scenarios.
Undersaturation solubility studies with Nd(OH) 3 (am) were conducted in inert gas (Ar) gloveboxes at 22 ±
2°C. Samples were prepared with dilute NaCl (0.1 M) and CaCl 2 (0.1 and 0.25 M) solutions as background
electrolyte. Parallel experimental series were prepared at constant pH c ca. 12 and 10 –6 ≤ [GLU] ≤ 10 –2 M,
and with [GLU] = constant = 10 –3 M and 9 ≤ pH c ≤ 13. The concentration of Nd in the aqueous solution was
quantified by ICP–MS after 10 kD ultrafiltration, whereas [GLU] tot in solution was measured as total organic
carbon (TOC). Solid phases before and after solubility experiments were characterized by XRD. TRLFS
measurements were performed with 1×10 –7 M Cm(III) per sample in NaCl (0.1 M) and CaCl 2 (0.1 and 0.25
M) solutions. In the NaCl systems, three different concentration levels of Ca 2+ were fixed at 0, 1×10 –3 M and

1×10 –2 M. The initial concentration of GLU in all samples (1×10 –6 M) was increased to 3×10 –3 M by step–
wise additions of NaGLU–NaCl or Ca(GLU) 2 –CaCl 2 solutions.
The solubility of Nd(OH) 3 (am) remains unaffected by GLU in 0.1 M NaCl solutions. XRD and TOC
analyses confirm that no secondary (crystalline) solid phases are forming and all GLU remains in solution
under these conditions. On the other hand, the solubility of Nd(OH) 3 (am) in 0.1 and 0.25 M CaCl 2 solutions
is clearly increased by GLU under hyperalkaline conditions. The species forming are pH–dependent and in
all cases involve the coordination of Ca 2+ . Slope analysis of Nd(OH) 3 (am) solubility under increasing
gluconate concentration (log [Nd] vs. log [GLU] tot ) likely indicates the formation of a Nd–GLU complex
with stoichiometry 1:2. At [GLU] tot ≥ 10 –3 M, no further increase of the Nd(III) concentration is observed,
resulting in an upper solubility limit at [Nd] ca. 10 –6.5 M which suggests the formation of a new Ca–Nd–GLU
solubility limiting solid phase.
TRLFS indicates the formation of one main Cm(III)–GLU species in NaCl systems and 10 –3 M ≤ [Ca] tot ≤
10 –2 M. Consistent with Nd(III) solubility data, TRLFS confirms the key role of Ca 2+ in the complexation
process and likely formation of a Ca–Cm(III)–GLU complex with Ca:GLU ratio 1:1. Two further Ca–Cm–
GLU species were identified in 0.1 M and 0.25 M CaCl 2 solutions by TRLFS, pointing to the complex
chemistry and speciation in the Ca–GLU system as observed by 13 C–NMR under analogous conditions [1].
Only weak Cm–GLU complexes were observed by TRLFS under the absence of Ca.
These results indicate that gluconate has a limited impact on the solubility of An(III) under repositoryrelevant
conditions.
[1] Gaona, X., Adam, C., Rojo, H., Böttle, M., Kaden, P., Altmaier, M. (2013). Solubility and NMR studies of Ca–
gluconate and Ca–Np(IV)–gluconate systems in dilute to concentrated alkaline CaCl 2 solutions. This conference.
PA3-8
A STUDY OF TERNARY Ca-UO 2 -CO 3 COMPLEXATION UNDER NEUTRAL TO WEAKLY
ALKALINE CONDITIONS
J.-Y. Lee and J.-I. Yun
Department of Nuclear and Quantum Engineering, KAIST
219 Daehak-ro, Yuseong-gu, Daejeon 305-701, Republic of Korea
The formation of Ca-UO 2 -CO 3 species has been considered primarily due to the predominant calcium and
carbonate ions in the natural groundwater system [1]. Even though several studies have addressed the
reaction behavior of Ca-UO 2 -CO 3 species, accurate spectroscopic properties and formation constants are still
uncertain, especially for CaUO 2 (CO 3 ) 3 2- species [1-3]. The objective of the present work was to investigate
the chemical behavior of Ca-UO 2 -CO 3 species under natural analogue system by combining time-resolved
laser fluorescence spectroscopy (TRLFS) with slope analysis.
Samples were prepared by dissolution of UO 2 (NO 3 ) 2·6H 2 O in 2 M HNO 3 . Total concentration of
uranium(VI) was kept at 10 -4 M. The initial CO 3 2- concentration was set to 5×10 -2 M with NaHCO 3 and
samples were equilibrated with atmospheric CO 2 (g). The pH was adjusted in the range of 7.4 to 9.0 with
TRIS, MES, and HClO 4 . The Ca 2+ concentration in aqueous samples was controlled by dissolving Ca(ClO 4 ) 2 .
The slope analysis by adding EDTA for the titration of Ca 2+ concentration was performed to identify the
formed uranyl complexes. Ionic strength was maintained at 0.1 M with addition of Na/HClO 4 . For the
excitation of Ca-UO 2 -CO 3 samples, a ns-pulsed fourth harmonic (266 nm) of Nd:YAG laser was utilized.
The fluorescence emission was measured by the Czerny-Turner type spectrometer coupled with an ICCD
camera. The overall fluorescence efficiency of the whole detection system including spectrometer and ICCD
was corrected with a fluorescence standard reference material (NIST, SRM 936a). All of the chemicals used
in the present work were analytical grade.
As shown in Fig. 1(a), measured fluorescence spectra revealed Ca-UO 2 -CO 3 species based on the
fluorescence peak wavelengths at 466, 485, 505, 527, and 551 nm, which are consistent with literature data

[1,3,4]. The increment in fluorescence intensity supports further complexation of ternary Ca-UO 2 -CO 3
species with raising Ca 2+ concentration. Fig. 1(b) represents the fluorescence lifetimes of Ca-UO 2 -CO 3
species as a function of Ca 2+ concentration. The lifetime behavior was represented by an S-shaped curve
ranging from a minimum value to a maximum value at low and high Ca 2+ concentration, indicating the
transition of uranium(VI) species from CaUO 2 (CO 3 ) 3 2- to Ca 2 UO 2 (CO 3 ) 3 (aq) species. From the result, the
fluorescence lifetimes of ternary Ca-UO 2 -CO 3 species were found to be 12.7±0.2 ns and 29.2±0.4 ns for
CaUO 2 (CO 3 ) 3 2- and Ca 2 UO 2 (CO 3 ) 3 (aq), respectively, even though indistinguishable fluorescence peak shapes
were monitored for both uranium(VI) species. In addition, the formation constants of ternary Ca-UO 2 -CO 3
species were calculated using slope analysis in terms of relative fluorescence intensity versus Ca 2+
concentration. Predominant formation of Ca 2 UO 2 (CO 3 ) 3 (aq) was identified at neutral pH while the transition
of ternary Ca-UO 2 -CO 3 species from Ca 2 UO 2 (CO 3 ) 3 (aq) to CaUO 2 (CO 3 ) 3 2- was confirmed at weakly basic
conditions due to limited soluble Ca 2+ ion concentration. The formation constants were determined to be log
β 0 113 = 27.27±0.14 for CaUO 2 (CO 3 ) 3 2- and log β 0 213 = 29.81±0.19 for Ca 2 UO 2 (CO 3 ) 3 (aq) using specific ion
interaction theory (SIT) [5]. The result reveals that ternary Ca-UO 2 -CO 3 species were major aqueous
uranium(VI) species at neutral to weakly alkaline conditions in the presence of Ca 2+ and CO 3 2- ions.
Fig. 1: (a) Fluorescence spectra and (b) fluorescence lifetimes of ternary Ca-UO 2 -CO 3 species as a function
of Ca 2+ concentration at pH 7.4.
[1] G. Bernhard, G. Geipel, T. Reich, V. Brendler, S. Amayri, H. Nitsche, Radiochim. Acta. 89, 511 (2001).
[2] W. Dong, S. C. Brooks, Environ. Sci. Technol. 40, 4689 (2006).
[3] O. Prat, T. Vercouter, E. Ansoborlo, P. Fichet, P. Perret, P. Kurttio and L. Salonen, Environ. Sci. Technol. 43, 3941
(2009)
[4] S. N. Kalmykov, G. R. Choppin, Radiochim. Acta. 88, 603 (2000).
[5] L. Ciavatta, Ann. Chim. Rome. 70, 551 (1980).
PA3-9
USE OF NMR TO DETERMINE STRUCTURE OF PLUTONIUM SIDEROPHORE COMPLEXES:
Pu(IV)(H 2 DFOB) and
DFOBPu(IV)di-μ-OH-DFOBPu(IV)
Boggs, M. A. 1 Zavarin, M. 1 Kersting A.B. 1
1. Glenn T. Seaborg Institute, Lawrence Livermore National Laboratory, 7000 East Avenue,
Livermore, California 94551 USA
Deferrioxamine B (DFOB) is a microbial siderophore that forms very strong metal ligand complexes. DFOB
is a linear molecule consisting of three hydroxamic acids linked by succinic acid and diaminopentane chains.
An important trait of DFOB is that while it exists primarily as a linear chain in solution it forms a cyclical
complex around most metals upon complexation. Structural changes of this nature are ideal for NMR
spectroscopy and can lead to insights into metal ligand structures. NMR also offers the unique ability to take
advantage of shifts in spectra due to plutonium paramagnetism to obtain solution state structures.

Plutonium (Pu) complexes with DFOB are of interest as they have been shown to decrease sorption of Pu to
iron bearing minerals (goethite) 1 while increasing sorption to clays (montmorillonite) 2 . Such differences in
behavior may be attributed to Pu(IV)-DFOB complexes being able to reside in the interlayer of clays through
an ion exchange process 3 . Studies have also observed an increase in cellular uptake of Pu 4 in the presence of
DFOB that was attributed to similarities between the plutonium and iron DFOB complexes. Its high affinity
to clays and microbial uptake makes the nature and structure of Pu(IV)-DFOB complexes particularly
important for understanding Pu environmental behavior.
Pu(IV)-DFOB complexes were prepared in D 2 O by adding equimolar (1 mM) amounts of Pu(IV) and DFOB.
The pH of these complexes was then adjusted with saturated NaOH in D 2 O to circa 1.0 or 7.0. Two distinct
complexes were observed and identified by UV-Vis 5 . Rapid conversion between the two complexes is
observed as the pH is shifted below or above ~ pH 4. Experiments were carried out over the course of
roughly 5 months with no changes in spectra that would signify a deterioration of the Pu(IV)-DFOB
complexes, indicating these complexes are stable.
1 H spectra of Pu(IV)-DFOB complexes (Figure 1) show a large amount of peak broadening and chemical
shifts, compared to uncomplexed DFOB and the Al-DFOB complex, which can be associated with changes
in conformation and the presence of a paramagnetic metal center. The change from a linear to a cyclical form
causes protons in previously equal environments to no longer be degenerate. 13 C DEPT NMR spectra showed
that unlike the Al-DFOB complex, Pu(IV)-DFOB complexes did not have rapid isomeric conversions.
Peak assignments were made through the 2D techniques of Correlation Spectroscopy (COSY) and
Heteronuclear Single Quantum Correlation (HSQC). While the spectra for the complexes studied were very
complex, we were able to identify the majority of the peaks present, and obtain basic structural information.
In order to expand on this information, variable temperature 1 H spectra were acquired and we were able to
observe pseudo contact shifts (PCS) and show that there are protons in different molecular orientations with
respect to the magnetic field of plutonium. Furthermore, by using the variable temperature method we were
able to determine, through PCS, the distance of the shifted protons from the metal center.
Diffusion Ordered Spectroscopy (DOSY) was used to determine the rates of diffusion and size of the two
Pu(IV)-DFOB complexes. These results indicate that the complex at pH 7.0 is on the order of twice the size
of the lower pH complex. The high pH complex was previously been identified 5 as Pu(IV)DFOB(OH) 2 ,
while our results suggest that it is more likely a dimeric DFOBPu(IV)di-μ-OH-DFOBPu(IV) species. The
existence of a dimer is corroborated by the rapid interconversion between the Pu monomer and dimer to the
monomeric Pu(IV)(H 2 DFOB) as well as by the different 1D and 2D spectra acquired.
The work presented here demonstrates how solution state NMR can be applied to answer complicated
questions concerning actinide complexation chemistry. Using relatively simple NMR techniques we were
able to deconstruct complex spectra and characterize two plutonium complexes.
1 Personal communication
2 Boggs, M.A; Abstract Submitted for Migrations 2013
3 Haack, E. A.; Johnston, C. T.; Maurice, P. A., Geochimica Et Cosmochimica Acta 2008, 72, 3381-3397
4 John, S. G.; Ruggiero, C. E.; Hersman, L. E.; Tung, C. S.; Neu, M. P., Environmental Science & Technology 2001, 35,
2942-2948
5 Boukhalfa, H.; Reilly, S.; Neu, M., Inorganic Chemistry 2007, 46, 1018-1026

PA3-10
THE BIGRAD CONSORTIUM - MONITORING REDOX SPECIATION OF URANIUM USING
OPTICAL FINGERPRINTING
L.S. Natrajan (1) , A.N. Swinburne (1) , S.D. Woodall (1) , S. Randall (1) , K. Smith (1,2) , N.D. Bryan (1) , K. Morris (2)
(1) The Centre for Radiochemistry Research, School of Chemistry, The University of Manchester, Oxford
Road, Manchester, M13 9PL, UK
(2)
The Research Centre for Radwaste and Decommissioning, School of Earth, Atmospheric and
Environmental Sciences, The University of Manchester, Oxford Road, Manchester, M13 9PL, UK
The UK currently holds a substantial nuclear legacy arising from fission activities, with a large proportion of
high activity wastes that pose a radiological threat to natural and engineered environments. The decision to
dispose of these high level wastes in a suitable geological disposal facility (GDF) has provided some of the
most demanding technical, and environmental challenges facing the UK in the coming century. Currently,
the plan is to place waste containers in cementitious vaults that will be sealed in order to restrict the mobility
of the radioactivity released under these conditions. However, since the cementitious material is highly
alkaline, reaction of the pore fluid with the surrounding rock will ultimately alter the chemical and physical
characteristics of the rock, potentially acting to transport radioactive material and affecting the retention or
release of radioactive materials.
In order to address these issues, we have begun a programme of work to establish a comprehensive
understanding of the electronic properties and physical and chemical properties of the radioactive actinide
metals using state of the art emission spectroscopic techniques. Our aim is to aid in the elucidation of
migration pathways in immobilized wastes as well as actinyl uptake and accumulation in the natural
environment. Despite the importance of monitoring speciation, techniques for measuring actinide
concentration and movement currently tend to be based on radiometric assay, which are destructive and
provide limited information regarding oxidation state or chemical form. By contrast, time resolved emission
spectroscopy, is a sensitive alternative technique to study the electronic properties of f-element compounds
and can be achieved remotely using fibre optic technology. Since actinide ions, in particular, uranium,
plutonium and neptunium can co-exist in several oxidation states depending on the surrounding physical
conditions, chemical behaviour is inherently linked to oxidation state. However, the speciation and
environmental fate following biological and chemical induced redox processes remain to be fully understood.

Figure 1. Ligands chosen as models for carboxylate and phosphate
uranium species
Our approach to this is to firstly use coordination chemistry to synthesise uranium compounds and ligands
that model environmentally complexed species and use optical spectroscopy to understand and map both the
chemical and physical behaviour of these species (Figure 1). We have recently established that U(IV)
complexes are emissive and will demonstrate that uranium in the +IV and +VI oxidation states can be
detected simultaneously at relatively low concentrations. Time gating techniques enable the long lived
uranyl(VI) species to be separated from the shorter lived uranium(IV) species. Furthermore, the form of the
emission spectra of uranyl(VI) compounds are extremely sensitive to the nature of the ligand bound in the
equatorial plane and the complex nuclearity (extent of aggregation), potentially giving a sensitive method of
assessing the solution forms of uranium in environmental conditions. We will next discuss how the optical
properties of these model compounds can be applied to disproportionation reactions and redox events in
solution and finally present our results of adsorbed and incorporated uranium speciation in calcite materials.
[1] L.S. Natrajan (2012), “Recent Developments in the Photophysics and Photochemistry of Actinide Coordination
Compounds”, Coord. Chem. Rev., 256: 1583.
[2] M.P. Redmond, D. Whittaker, S.D. Woodall, S.M. Cornet, D. Collison, M. Heliwell, L.S. Natrajan (2011), “Probing
the local coordination environment and solution nuclearity of uranyl complexes by emission spectroscopy, Dalton
Trans., 15: 3761.
[3] L.S. Natrajan (2012), “The First Structural and Spectroscopic Study of a Paramagnetic 5f DO3A Complex”, Dalton
Trans., 41: 13167.
[4] E. Hashem, A.N. Swinburne, C. Schulzke, R.C. Evans, J.A. Platts, A. Kerridge, L.S. Natrajan, R.J. Baker (2013),
“Emission Spectroscopy of U(IV) Compounds; A Combined Synthetic, Spectroscopic and Computational Study“ RSC
Advances, 3: 4350.
PA4
PA4-1
PA4-2
PA4-3
PA4-4
PA4-5
PA4-6
REDOX REACTIONS AND RADIOLYSIS EFFECTS
EXAMINATION OF THE EFFECT OF ALPHA RADIOLYSIS ON PLUTONIUM(V)
SORPTION TO QUARTZ USING MULTIPLE PLUTONIUM ISOTOPES
A.E. Hixon, Y. Arai, B.A. Powell (USA)
INVESTIGATION OF THE REDOX AND COMPLEXATION BEHAVIOUR OF
URANIUM BY ORGANIC ACIDS USING CYCLIC VOLTAMMETRY
M.Q. Chew, N.D.M. Evans, R.J. Mortimer, C. Boxall (UK)
CUPROUS HYDROXIDE AS AN INTERMEDIATE STEP IN OXIDATION OF COPPER
IN AQUEOUS SOLUTIONS
I.L. Soroka, P.A. Korzhavyi, N.V. Tarakina, M. Jonsson (Sweden, Germany)
THIOSULPHATE, KEY INTERMEDIATE DETERMINING ABIOTIC SELENIUM
REDOX TRANSFORMATIONS IN PRESENCE OF PYRITE
E. Breynaert, W. Wangermez, T.N. Parac-Vogt, C.E.A. Kirschhock, A. Maes (Belgium)
THE BIGRAD CONSORTIUM - NEPTUNIUM BIOGEOCHEMICAL INTERACTIONS
WITH THE MANGANESE CYCLE
G.T.W. Law, C.L. Thorpe, P. Bots, S. Shaw, K. Law, T. Marshall, F.R. Livens, J.R. Lloyd,
M.A. Denecke, J. Rothe, K. Dardenne, K. Morris (UK, Germany)
ROLE OF Fe(II) ON ACTINIDE REDOX PROCESSES AT MINERAL SURFACES
Y. Wen, D. Renock, L. Shuller-Nickles (USA)

PA4-7
PA4-8
PA4-9
PA4-10
PA4-11
STRUCTURAL INVESTIGATION OF SOLID SOLUTIONS IN THE SYSTEM USiO 4 –
ThSiO 4
S. Labs, S. Weiss, C. Hennig, H. Curtius, D. Bosbach (Germany)
ON FISSION PRODUCT ALLOY PARTICLES AND THEIR CATALYTIC PROPERTIES
D. Cui, S. Hovmöller, W. Wan, Y. Yun, M. Granfors, L. Jeanett, K. Spahiu (Sweden)
SOLUBILITY OF TcO 2· xH 2 O( S ) IN DILUTE TO CONCENTRATED NaNO 3 SOLUTIONS
T. Kobayashi, T. Sasaki, A. Kitamura (Japan)
SURFACE SCIENCE STUDY OF SPENT FUEL CORROSION PROCESSES USING
THIN FILM MODEL SYSTEMS
T. Gouder (EC)
EFFECT OF NITROUS ACID ON REDUCTION OF Np(VI) IN IRRADIATED
SOLUTIONS OF NITRIC ACID
A. Paulenova, M. Precek, B. Mincher, S. Mezyk (USA, Czech Republic)
PA4-1
EXAMINATION OF THE EFFECT OF ALPHA RADIOLYSIS ON PLUTONIUM(V) SORPTION
TO QUARTZ USING MULTIPLE PLUTONIUM ISOTOPES
Amy E. Hixon (1) , Yuji Arai (2) , and Brian A. Powell (1)
(1) Department of Environmental Engineering & Earth Sciences,
Clemson University, Clemson, SC, USA
(2) School of Agriculture, Forestry, and Environmental Science,
Clemson University, Clemson, SC, USA
Plutonium contamination is a worldwide problem due to nuclear weapons production and testing,
commercial nuclear power, and nuclear waste repositories. The presence of plutonium in the geosphere poses
a long-term environmental concern due to its toxicity and the long half-lives of several isotopes. Therefore,
understanding the mechanisms responsible for enhancing or retarding the mobility of plutonium in the
environment is important for risk management. Surface-mediated reduction of relatively mobile Pu(V) to
relatively immobile Pu(IV) has been observed on a variety of surfaces. Although numerous studies have
observed the capacity of pure minerals to reduce or oxidize plutonium, the exact mechanism(s) responsible
for surface-mediated redox reactions remain unclear. Hypotheses include: (i) radiolysis at the mineral surface
[1-2], (ii) electron transfer via ferrous iron or manganese in the mineral structure [1, 3-6], (iii)
disproportionation of Pu(V) [1-2, 4, 7], (iv) electron shuttling due to the semiconductor properties of the
mineral [5, 7], and (v) reduction based on a Nernstian favorability of Pu(IV) surface complexes or
nanocolloidal precipitates [5, 8].
Batch sorption experiments were used to examine Pu(V) sorption to quartz as a function of time and total
alpha radioactivity. Systems were prepared using different ratios of 238 Pu and 242 Pu to maintain a constant
plutonium total concentration of 4 x 10 -8 M while varying the total alpha radioactivity from
1500 Bq/L to 36000 Bq/L. This experimental design was used to examine the influence of radiolysis on
plutonium sorption and reduction. The conceptual model is that radiolysis products formed from the
radioactive decay of plutonium can reduce Pu(V) to Pu(IV) either on the mineral surface or in solution. Thus,
the samples containing greater alpha activity should exhibit greater sorption and reduction. The oxidation
state of plutonium associated with quartz was examined using solvent extraction, X-ray absorption
spectroscopy (XAS), and electron microscopy. The high-purity quartz used in this study is not a
semiconductor and trace metals such as iron or manganese, which can facilitate electron shuttling and
electron transfer mechanisms, are below detection limits. Therefore, based on the hypotheses discussed
above, reduction of plutonium in these systems is due to radiolysis, Pu(V) disproportionation, or a
thermodynamic favorability of sorbed Pu(IV) hydrolysis products.

Plutonium remained soluble at pH 3. At pH 5, the fraction of sorbed plutonium increased with increasing
time. Increasing the pH resulted in an increase in fraction sorbed at all equilibration times and alpha
activities. After approximately one year, the fraction of sorbed plutonium added as Pu(V) converged to the
values for plutonium added as Pu(IV), which provides indirect evidence of Pu(V) reduction to Pu(IV). This
was confirmed by solvent extraction results and XAS fits, which show a significant fraction of Pu(IV)
associated with the quartz in both batch sorption and XAS samples. Over the two-year time period of these
experiments, Pu(V) remained stable in the aqueous phase. Therefore, either reduction of Pu(V) to Pu(IV) is
surface mediated or sorption of Pu(IV) occurs instantaneously following aqueous reduction of Pu(V) to
Pu(IV). Plutonium nanocolloids were observed using electron microscopy, indicating that the final reduced
phase may be a solid Pu(IV) species at high plutonium concentrations. Regardless of whether Pu(IV) is
present as a monomeric sorbed species or nanocolloidal precipitate, Pu(V) reduction to Pu(IV) occurred in
these systems.
An investigation of the kinetics of Pu(V) sorption to quartz showed that there was no significant dependence
of the reaction rate constant on the total alpha activity of the solution. Therefore, radiolysis at the mineral
surface does not appear to be the main mechanism responsible for surface-mediated reduction of plutonium
in these systems. A comparison of the reaction rate constants to the rate of disproportionation showed that
disproportionation of Pu(V) also does not appear to be responsible for surface-mediated reduction of
plutonium in these systems. Thermodynamic calculations indicate that the reduction of Pu(V), facilitated by
oxidation of water, would become energetically favorable if the potential of the Pu(V)/Pu(IV) couple was
increased by 0.28 V. Therefore, it is reasonable to suggest that a thermodynamic gradient to the formation of
a surface bound, hydrolyzed Pu(IV) complex or a precipitated Pu(IV) nanocolloid drives Pu(V) reduction to
Pu(IV) in the presence of quartz.
[1] Sanchez, A.L.; Murray, J.W.; Sibley, T.H. Geochim. Cosmochim. Acta 49 (1985) 2297.
[2] Romanchuk, A. Yu.; Kalmykov, S.N.; Aliev, R.A. Radiochim. Acta 99 (2011) 137.
[3] Shaughnessy, D.; Nitsche, H.; Booth, C.H.; Shuh, D.K.; Waychunas, G.A.; Wilson, R.E.; Gill, H.; Cantrell, K.J.;
Serne, R.J. Environ. Sci. Technol. 37 (2003) 3367.
[4] Powell, B.A.; Fjeld, R.A.; Kaplan, D.I.; Coates, J.T.; Serkiz, S.M. Environ. Sci. Technol. 39 (2005) 2107.
[5] Powell, B.A.; Fjeld, R.A.; Kaplan, D.I.; Coates, J.T.; Serkiz, S.M. Environ. Sci. Technol. 39 (2005) 2107.
[6] Powell, B.A.; Duff, M.C.; Kaplan, D.I.; Fjeld, R.A.; Newville, M.; Hunter, D.B.; Bertsch, P.M.; Coates, J.T.; Eng,
P.; Rivers, M.L.; Sutton, S.R.; Triay, I.R.; Vaniman, D.T. Environ. Sci. Technol. 40 (2006) 3508.
[7] Keeney-Kennicutt, W.L. and Morse, J.W. Geochim. Cosmochim. Acta 49 (1985) 2577.
[8] Powell, B.A.; Dai, Z.; Zavarin, M.; Zhao, P.; Kersting, A.B. Environ. Sci. Technol. 45 (2011) 2698.
PA4-2
INVESTIGATION OF THE REDOX AND COMPLEXATION BEHAVIOUR OF URANIUM BY
ORGANIC ACIDS USING CYCLIC VOLTAMMETRY
(1)
M. Q. Chew (1) , N. D. M. Evans (1) , R. J. Mortimer (1) , C. Boxall (2)
Chemistry Department, Loughborough University, Loughborough, Leics, LE11 3TU, UK.
(2)
Nuclear Engineering Department, Lancaster University, Bailrigg, Lancaster, LA1 4YW, UK.
The current option for the management of Intermediate-Level Waste (ILW) and High-Level Waste (HLW) in
the UK is to store it in a deep underground Geological Disposal Facility (GDF). This may subsequently be
backfilled with a cementitious material generating hyperalkaline conditions and high concentrations of
calcium from the cement. 1 The preferred buffer/backfill material is a cement-based grout known as the Nirex
Reference Vault Backfill (NRVB). 2,3 The GDF will develop highly alkaline porewater with an initial pH of
about 13-13.5. However, this pH will decrease as ingressing groundwater dissolves the NaOH and KOH
present. Nevertheless, slightly soluble cement mineral phases should buffer the porewater pH at 12-12.5 for
about 10 5 years. 4,5 Under these conditions, most actinides and fission products will be sparingly soluble. The
corrosion of the stainless steel canisters containing the waste used for disposal will lead to reducing
conditions thereby promoting a low Eh environment. 2

Uranium is a major component of the UK waste inventory, mainly from spent fuel. The redox behaviour of
uranium in the nuclear wastes must therefore be understood in the context of prevailing GDF chemistry,
where the reduced state of uranium, i.e. U(IV) is preferred as it is generally less soluble and therefore less
mobile in groundwater. The mobility and solubility of uranium depends significantly on its oxidation state.
The GDF will not be homogeneous at any one time, so there will be areas of oxidising and reducing
potentials. This could mean that both U(VI) and U(IV) are present together at any one time.
Electrochemical experiments are needed to determine which uranium species is/are present at a particular pH
and to model the redox behaviour of uranium in a potential GDF. The main aim of this project is to deduce
peak potentials for the various uranium redox couples in aqueous solutions across the pH range and in
particular the hyperalkaline range using cyclic voltammetry. The ionic strength of aqueous solutions used for
the electrochemical measurements is much higher than in deep groundwaters, where the temperature is also
higher than 25 o C. 6 Data in the literature have been obtained only under acidic conditions where they were
subsequently extrapolated to obtain data for alkaline conditions in some reports. Is this valid? Experiments
are therefore needed to obtain fundamental data under alkaline conditions to fill in gaps in the literature.
Besides radionuclides, anthropogenic organic materials will be present in UK’s nuclear wastes, including
decontamination agents such as EDTA, natural polyhydroxylated carboxylic acids like α-isosaccharinic acid
(α-ISA), which is the main degradation product formed by the anaerobic, alkaline degradation of cellulose
found in soil and waste materials and gluconic acid. 7-12 α-ISA is of particular interest because it may form in
ILW stores when cellulose is degraded by the calcium hydroxide in the cement backfill materials. In
addition, α-ISA and its derivatives have potential applications in decontamination and remediation processes
for materials and sites contaminated with actinides. 13 These ligands could have an important effect on the
migration of radionuclides due to their ability to form complexes, thereby enhancing their solubility and
mobility in the cement pore water, sorbing onto cement surface sites and competing with other anions. 5
The peak potential of each uranium redox reaction was measured and the potentials were subsequently
compared in ligand and non-ligand systems (see Fig. 1). The redox behaviours of uranium in the presence of
gluconic acid and α-ISA across the pH range were compared to those of uranium in the presence of EDTA at
similar pH. Their behaviours were compared to obtain their similarities and differences in terms of peak
shape, peak potential, reversibility/re-oxidisability, current response and etc. Analysis and comparison of the
similarities and differences were needed be able to better understand the complexation effects of these
ligands with uranium at different pH. The peak potentials of uranium across the pH range in various
experimental conditions were tabulated to confirm if uranium observed the ’59 mV movement per pH unit’
theory under the different experimental conditions described in this work. In addition, the disproportionation
of U(V) to U(IV) and U(VI) was investigated using a scan rate study.
Graph of Current (A) versus Potential (V) of 3 mM Uranium Solution in 0.2 M
Sodium Carbonate and 0.8 M Sodium Chloride Background Electrolyte
Mixture obtained using Glassy Carbon Working Electrode at 0.1 V/s
4.00E-05
2.00E-05
0.00E+00
-1.6 -1.4 -1.2 -1 -0.8 -0.6 -0.4 -0.2
-2.00E-05
0 0.2 0.4 0.6
Current (A)
-4.00E-05
-6.00E-05
-8.00E-05
-1.00E-04
-1.20E-04
pH 2.14; 0.1 M
EDTA absent
pH 2.21; 0.1 M
EDTA present
Potential (V) vs. Ag/AgCl (3 mol dm -3 NaCl)
-1.40E-04

Fig.1. Uranium cyclic voltammogram in the absence and presence of EDTA in excess sodium carbonate at
ca. pH 2.
1. D. P. Rai, L. F. Rao and Y. X. Xia, Solubility of Crystalline Calcium Isosaccharinate, J. Sol. Chem., 1998, 27,
1109.
2. A. J. Francis, R. Cather and I. G. Crossland, Development of the Nirex Reference Vault Backfill: Report on
Current Status in 1994, United Kingdom Nirex Ltd., Harwell, United Kingdom, 1997.
3. I. G. Crossland and S. P. Vines, Why a Cementitious Repository?, United Kingdom Nirex Ltd., Harwell, United
Kingdom, 2001.
4. U. Berner, A Thermodynamic Description of the Evolution of Pore Water Chemistry and Uranium Speciation
During the Degradation of Cement, Paul Scherrer Institute, Cillingen, Switzerland, 1990.
5. I. Pointeau, D. Hainos, N. Coreau and P. Reiller, Effects of Organics on Selenite Uptake by Cementitious
Materials, Waste Manage., 2006, 26, 733.
6. H. Capdevila and P. Vitorge, Temperature and Ionic Strength Influence on U(VI/V) and U(IV/III) Redox Potentials
in Aqueous Acidic and Carbonate Solutions, J. Radioanal. Nucl. Chem., 1990, 143, 403.
7. B. F. Greenfield, M. H. Hurdus, N. J. Pilkington, M. W. Spindler and S. J. Williams, The Degradation of Cellulose
in the Near-Field of a Radioactive Waste Repository, Mat. Res. Soc. Symp. Proc., Boston, United States of America,
1994, 333, 705.
8. B. F. Greenfield, A. D. Moreton, M. W. Spindler, S. J. Williams and D. R. Woodwark, The Effects of the
Degradation of Organic Materials in the Near-Field of a Radioactive Waste Repository, Mat. Res. Soc. Symp. Proc.,
Strasbourg, France, 1992, 257, 299.
9. B. F. Greenfield, G. F. Holtom, M. H. Hurdus, N. O'Kelly, N. J. Pilkington, A. Roseaver, M. W. Spindler and S. J.
Williams, The Identification and Degradation of Isosaccharinic Acid, A Cellulose Degradation Product, Mat. Res. Soc.
Symp. Proc., 1995, 353, 1151.
10. B. F. Greenfield, W. N. Harrison, G. P. Robertson, P. J. Somers and M. W. Spindler, Mechanistic Studies of the
Alkaline Degradation of Cellulose in Cement, Harwell, United Kingdom, 1993.
11. R. L. Whistler and J. N. BeMiller, Alkaline Degradation of Polysaccharides, Adv. Carbohydr. Chem., 1958, 13,
289.
12. R. L. Whistler and J. N. BeMiller, α-D-Isosaccharino-1,4-Lactone, Carbohydr. Chem., 1963, 2, 477.
13. R. P. Bontchev, R. Moore, M. Tucker and K. Holt, Structure and Aqueous Solubility of Sodium Isosaccharinate in
Environmental and Waste Management: Advancements through the Environmental Management Science Program,
227 th American Chemical Society Meeting, Anaheim, United States of America, 2004.
PA4-3
CUPROUS HYDROXIDE AS AN INTERMEDIATE STEP IN OXIDATION OF COPPER IN
AQUEOUS SOLUTIONS
I.L. Soroka 1 , P. A. Korzhavyi 2 , N. V. Tarakina 3 and M. Jonsson 1
1 Applied Physical Chemistry, School of Chemical Science and Engineering, KTH Royal Institute of
Technology, SE-100 44 Stockholm, Sweden
2 Department of Material Science and Engineering, KTH Royal Institute of Technology, SE-100 44 Stockholm
3 Experimentelle Physik III, Physikalisches Institut and Wilhelm Conrad Röntgen - Research Centre for
Complex Material Systems, Universität Würzburg, Am Hubland, D-97074 Würzburg, Germany
The potential degradation of the outer copper shell is a key issue in the long term safety assessment of the
deep repository for spent nuclear fuel, since a copper canister completely isolates the spent nuclear fuel from
the surrounding groundwater. Numerous studies on copper corrosion in different environments have been
presented; however, the mechanism of copper oxidation under anoxic conditions is still not completely
understood. Understanding the mechanism of copper corrosion in aqueous solutions and under the irradiation
opens up the possibility of predicting and possibly controlling the copper canister degradation.
Corrosion of copper in groundwater environments can be enhanced by groundwater components, such as
sulphides, oxygen and chlorides. Also, there are studies which claim that copper can corrode in anoxic
conditions [1] with the formation of CuOH. The corrosion of copper in oxygen-free water is not energetically
favourable and it is not expected to occur at any significant rate [2]. However, it was shown that the adsorbed
OH monolayers on copper surface can be more stable than bulk Cu 2 O [3]. In connection to this, there are
several questions to be addressed: if the chemisorbed Cu-OH monolayer can grow to form a bulk CuOH

phase; how stable is this phase; under which conditions solid CuOH can form and is this compound of
importance in the copper corrosion process?
Our studies show that solid CuOH exists and that it is metastable. We investigate the mechanism of copper
oxidation in aqueous environment through the possible formation of cuprous hydroxide as an intermediate
step.
The experimental investigation of the chemical properties of cuprous hydroxide, which had been
theoretically proved to be metastable in solid form [4], is performed. It was shown that cuprous hydroxide
may be synthesized by at least three different methods, such as: the reduction of Cu 2+ to Cu + with free
radicals in aqueous solutions; the reduction reaction of Cu 2+ with EDTA and the formation of CuOH layers
on the surface of the Cu-CuH powder stored in water [5]. The aim of these studies is to verify experimentally
that copper (I) hydroxide exists in a solid form and to find the way to distinguish between CuOH and Cu 2 O,
as well as to demonstrate that these are indeed two different compounds. Our study reveals that cuprous
hydroxide does exist in solid form and most likely has a hydrated form CuOH×H 2 O [5]. Moreover, it may
form on the surfaces of metallic copper and copper (I) compounds [6] when exposed to aqueous solutions
and to ionizing radiation. The formation of cuprous hydroxide may be an intermediate step in oxidation of
copper in aqueous solutions. At the same time, cuprous hydroxide may form layers which protect Cu surface
from further oxidation [6]. The investigation of the process of cuprous hydroxide formation and the study of
CuOH physical and chemical properties are of key importance for understanding the copper corrosion
mechanism.
[1] G. Hultquist, M.J. Graham, O. Kodra, S. Moisa, R. Liu, U. Bexell and J.L. Smialek “Corrosion of copper in distilled
water without molecular oxygen and the detection of produced hydrogen”, Swedish Radiation Safety Authorities, SSM,
report 2013:07
[2] I Puigdomenech, C Taxén Technical Report TR-00-13, SKB, Swedish Nuclear Fuel and Waste Management Co. ,
Stockholm, 2000.
[3] E. Protopopoff, P. Markus, Electrochim. Acta 51, 408 (2005)
[4] P. A. Korzhavyi, I. L. Soroka, E. I. Isaev, C. Lilja and Börje Johansson, PNAS 2012, 109, pp.686-689.
[5] I. L. Soroka, A. Shchukarev, M. Jonsson, N. V. Tarakina and P. A. Korzhavyi Cuprous hydroxide in solid form:
Does it exist?” in manuscript.
[6] I. L. Soroka, N. V. Tarakina, P. A. Korzhavyi, L. Belova, V. Stepanenko d and M. Jonsson ”Effect of synthesis
temperature on the morphology and solid-state stability of copper hydride desert-rose nanoparticles” in manuscript.
PA4-4
THIOSULPHATE, KEY INTERMEDIATE DETERMINING ABIOTIC SELENIUM REDOX
TRANSFORMATIONS IN PRESENCE OF PYRITE
E. Breynaert 1 , W. Wangermez 1 , T.N. Parac-Vogt 2 , C.E.A. Kirschhock 1 , A. Maes 1
1) KULeuven – Center for Surface Chemistry and Catalysis, Department M 2 S, kasteelpark arenberg 23 – box
2461, B-3001 Heverlee, Belgium. eric.breynaert@biw.kuleuven.be
2) KULeuven – Department of Chemistry, Celestijnenlaan 200F, B-3001, Leuven, Belgium.
The geochemical behaviour and bio-availability of selenium have an unexpectedly intricate impact on
modern society. While selenium is an essential micronutrient for many living organisms, the window
between deficiency and toxicity is very narrow (0.04 ppm ; essential; 0.04 – 0.1 ppm beneficial; 3 ppm
toxic). Due to its similarity to sulphur, it is commonly encountered in subsurface deposits such as coal and
uranium, phosphate and sulphidic transitionmetal ores. The release of selenium to the environment is closely
associated with the economic exploitation of such deposits. Because of its significant contribution to longterm
radiation exposure, 79 Se is considered as one of the important isotopes in the inventory of the long-lived
radioactive waste produced by nuclear industry.
Considering exclusively abiotic processes, selenium interaction with abundant iron sulphide minerals
significantly contributes to the mechanisms limiting its bio-availability. Using X-ray Absorption
Spectroscopy (XAS), Breynaert et al. demonstrated the interactions of Se(IV) with Fe(II) monosulphide (e.g.
troilite, mackinawite) and Fe(II) disulphide (e.g. pyrite) in circum-neutral anoxic aqueous conditions differ

completely.[1] In line with other studies, [2, 3] FeSe x was observed as solid phase reaction product upon
equilibrating Se(IV) with troilite. Se(IV) interaction with pyrite produced Se 0 , an outcome which could not
readily be explained. [1] Since pyrite (FeS 2 ) is one of the main minerals governing the redox conditions of
the Boom Clay (potential storage site in Belgium), it was expected that selenite also would be reduced to Se 0
in Boom Clay conditions. The combination of XAS with size fractionation and chemical analysis not only
confirmed formation of Se0, but also demonstrated a strong correlation between the formation of Se 0 and the
pyrite content of different grain size fractions encountered in the Boom Clay. These results confirmed the
adsorption/reduction based reaction scheme previously derived from experiments evaluating the geochemical
behaviour of selenite in the Boom Clay system [4]. In all experiments, the total Se concentration in solution
decreased as function of time after addition of selenite to Boom Clay batch systems, hence leading to an
increasing association of Se with the solid phase [4-6].
Although these studies pinpointed pyrite as the main factor determining the geochemical behaviour of
selenite in Boom Clay, the data did not allow formulating a molecular level mechanism for the reduction of
Se(IV) by pyrite or Boom Clay. In addition, speciation analysis (SEC + IC) on artificial Boom Clay
porewater solutions equilibrated with selenite had indicated the presence of significant concentrations of
dissolved non-selenite Se species (up to ~10 -7 M in the systems studied).[7]
This paper demonstrates how the action of intermediary thiosulphate (S 2 O 3 2- ) species in the reaction
mechanism of Se(IV) reduction with pyrite explains the different reaction products in presence of FeS and
FeS 2 , respectively. This mechanism explains the presence of low concentrations of dissolved non-selenite
species in combination with Se 0 solid phase reaction products in pyrite-containing/-derived systems. The
proposed model is in line with state-of-the art pyrite oxidation chemistry and was demonstrated by
combination of XAS and NMR spectroscopy in Boom Clay batch systems, and in synthetic samples
designed to isolate specific reaction intermediates occurring in chemical systems containing pyrite and
selenite.
[1] E. Breynaert, et al., ES&T, 42 (2008) 3595-3601.
[2] A.C. Scheinost, L. Charlet, ES&T, 42 (2008) 1984-1989.
[3] A.C. Scheinost, et al., J. Contam. Hydrol., 102 (2008) 228-245.
[4] C. Bruggeman, et al., Env. Poll., 137 (2005) 209-221.
[5] E. Breynaert, et al., ES&T, 44 (2010) 6649-6655.
[6] C. Bruggeman, et al., Radiochim. Acta, 90 (2002) 629-635.
[7] C. Bruggeman, et al., Appl. Geochem., 22 (2007) 1371-1379.
PA4-5
THE BIGRAD CONSORTIUM – NEPTUNIUM BIOGEOCHEMICAL INTERACTIONS WITH
THE MANGANESE CYCLE
Gareth T.W. Law 1 , Clare L. Thorpe 2 , Pieter Bots 2 , Sam Shaw 2 , Kathleen Law 1 , Tim Marshall 2 , Francis R.
Livens 1 , Jon R. Lloyd 2 , Melissa A. Denecke 3 , Jörg Rothe 3 , Kathy Dardenne 3 , Katherine Morris 1
1) Centre for Radiochemistry Research, School of Chemistry, The University of Manchester, Manchester, M13
9PL, UK
2) Research Centre for Radwaste and Decommissioning, School of Earth, Atmospheric and Environmental
Sciences, The University of Manchester, Manchester, M13 9PL, UK
3) Institut für Nukleare Entsorgung (INE), Forschungszentrum Karlsruhe, D-76021 Karlsruhe, Germany
Neptunium is a key risk-driving radionuclide in geological disposal and it is becoming apparent that
microbially induced redox cycling is pertinent to geological disposal environments [1]. Furthermore, Np is
predicted to be the most mobile transuranic element in the sub-surface at nuclear contaminated sites;
however, there is a paucity of information concerning neptunium’s environmental behavior and this is
particularly true when one considers the biogeochemical interactions of Np with the Mn cycle (Mn is a
ubiquitous element implicated in actinide biogeochemical cycling [e.g. 2, 3]). To further explore the
relationships between Np and Mn, we have characterised Np biogeochemical behaviour in a range of Mn
dominated experimental systems. These included: (i) a low-level (0.3 μM 237 Np) bioreducing δMnO 2 -rich
sediment system; (ii) a high level (0.3 mM 237 Np) bioreducing δMnO 2 -rich sediment system; and (iii) a range

of synthetic Mn mineral systems (pure-phase δMnO 2 , tri-clinic birnessite, acid birnessite, hausmannite, and
rhodochrosite). In all experiments, NpO 2 + was added to groundwater and sediment / mineral mixtures at
concentrations below saturation. During a timecourse of sampling, aqueous Np was tracked alongside the
evolving stable element geochemistry. After reaction, the sediment / mineral phases were then sampled and
collected under anaerobic conditions for XAS (EXAFS and XANES) analysis of the Np-oxidation state and
coordination environment. Neptunium L III -edge spectra (17.610 keV) were collected at ambient temperature
in fluorescence mode by a 5 pixel solid-state detector (LEGe Canberra) using Ge(422) monochromator
crystals.
In the low level (0.3 μM 237 Np) δMnO 2 -rich sediment system, a small proportion of the added NpO 2
+
sorbed to the oxic sediments. Acetate addition then stimulated microbially-mediated bioreduction and whilst
the rate of Np sorption to sediment remained low during denitrification, it increased considerably during both
Mn-reduction and Fe(III) reduction. Indeed, by the onset of sulfate reduction, the vast majority (> 99 %) of
the added NpO 2 + had been removed from solution. Select, parallel XAS experiments (Np(V) ~ 0.3 mM)
underwent a similar biogeochemical evolution and samples were collected for XAS analysis under defined
terminal electron accepting process (namely, aerobic respiration; denitrification; Mn reduction (in the
absence of measureable Fe(III) reduction); Fe(III) reduction; and SO 4 2- reduction). Subsequent XAS analysis
revealed that Np(V) was wholly reduced to Np(IV) concomitant with Mn-reducing conditions although it
was not clear whether reduction was directly (enzymatically) facilitated by the indigenous bacteria or
whether an indirect, redox reaction dominated. In the pure-mineral systems, Np reactivity varied across the
range of minerals with some evidence for abiotic reduction in Mn(II/III)-bearing phases. These data suggest
a direct link between the Mn and Np biogeochemical cycles, and suggest that Np(IV) will be the dominant
species under mildly reducing biogeochemical conditions.
[1] T. Rizoulis, et al., Min. Mag. 26, 3621 (2012).
[2] G. T. W. Law, et al., Env. Sci. Technol. 44, 8924 (2010).
[3] B. A. Powell, et al., Env. Sci. Technol. 40, 3508 (2006).
PA4-6
ROLE OF FE(II) ON ACTINIDE REDOX PROCESSES AT MINERAL SURFACES
Y. Wen (1) , D. Renock (2) , L. Shuller-Nickles (1)
(1) Environmental Engineering and Earth Science, Clemson University, 342 Computer Court, Anderson, SC
29625, USA
(2) Department of Earth Sciences, Dartmouth College, 6105 Fairchild Hall, Hanover, NH 03755, USA
Actinide mobility in aqueous systems is affected by many factors, particularly oxidation state. In natural
systems, certain minerals and microbes can facilitate reduction of these elements, producing less mobile
tetravalent U, Np, and Pu species. Fe-bearing minerals are of great interest since they are ubiquitous in
geologic environments, have high reactivity, and have high sorption capacities. Studying absorption and
redox reactions between Np, Pu, U species and Fe-bearing minerals will provide more knowledge for risk
assessment associated with actinide transportation in the environment.
To study the redox behavior of actinides at mineral surfaces, it is crucial to gain information about redox
potentials, activation energies, and reaction rates. Cyclic voltammetry (CV) is used to acquire redox
potentials of actinides at mineral surfaces by forcing the actinides to oxidize and reduce on a mineral powder
microelectrode (PME, or cavity microelectrode). Compared with traditional electrodes, the PME is easy to
prepare, free of adhesive, and can efficiently limit mass transfer effects. Also, PME can allow wider scan
range and higher scan rates 1 . Through varying preconditioning times, reaction rates can be determined by
examining the amount of analyte reduced, which is reflected in CV peak heights.
Cyclic voltammetry of actinyl complexes using natural magnetite, pyrite, and hematite PMEs indicates the
role of the mineral surface in actinyl redox reactions. Reduction of U(VI) to U(IV) is known to occur with
the presence of structural Fe(II) in minerals 2 . For example, with the magnetite PME, as illustrated in Figure
1, U(IV) to U(VI) oxidation (Peak I at +0.12 V) and the U(V)/U(VI) redox couple (Peak II at -0.05 V for

oxidation, Peak III at -0.23 V for reduction) were readily identified. Fe(II)/Fe(III) couple was observed at
+0.45 V for oxidation (Peak IV) and +0.38 V for reduction (Peak V). Through preconditioning at -0.25 V for
1-20 minutes, U(IV) can build up in the system. Longer preconditioning time results in higher U(IV) to
U(VI) oxidation peak (Peak I), which indicates observable reduction of U(VI) to U(IV) occurs in minutes
and even longer to fully reduce U(VI) to U(IV). Since U(VI) to U(IV) reduction is a slower reaction than the
experimental conditions, which are measured in seconds, U(IV) is quickly depleted in the first cycle, and
Peak I is replaced by U(V) to U(VI) oxidation Peak II. Additional Fe(II) in the solution was observed to
increase Peak I intensity, indicating enhanced U(IV) reduction and subsequent IV to VI oxidation.
Figure 2. Uranium CV on magnetite with and without Fe(II). Preconditioned at -0.25 V for 15 min,
scan rate 50 mV/s. Scan range is -0.55 V to +0.7 V. After acquiring uranium CV without Fe(II), the
cell solution was spiked with 3.85 mM FeCl 2 solution to acquire the uranium CV with Fe(II). Peak I
(+0.12 V) is the U(IV) to U(VI) oxidation peak. Peak II (-0.05 V) is the U(V) to U(VI) oxidation
peak. Peak III (-0.23 V) is the U(VI) to U(V) reduction peak. Peak IV (+0.45 V) shows Fe(II) to
Fe(III) oxidation; while Peak V (+0.38 V) shows Fe(III) to Fe(II) reduction. Changes in Peaks I and
III intensities reflect the amount of uranium involved in the redox reactions
[1] Vivier V, Cachet-Vivier C, Wu BL, Cha CS, Nedelec JY, Yu LT (1999). "Cavity microelectrode for studying
powder materials at a high potential scan rate." Electrochemical and Solid State Letters. 1999; 2(8):385-7.
[2] Skomurski FN, Ilton ES, Engelhard MH, Arey BW, Rosso KM (2011). "Heterogeneous reduction of U6+ by
structural Fe2+ from theory and experiment." Geochim Cosmochim Acta. 2011; 75(22):7277-90.
PA4-7
STRUCTURAL INVESTIGATION OF SOLID SOLUTIONS IN THE SYSTEM
USiO 4 – ThSiO 4
S. Labs 1) , S. Weiss 2) , C. Hennig 2) , H. Curtius 1) , D. Bosbach 1)
1) Forschungszentrum Jülich, Institute of Energy and Climate Research – IEK-6, Nuclear Waste
Management, 52425 Jülich, Germany.
2) Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology, Bautzener Landstraße 400,
01314 Dresden, Germany.
UO 2 is considered to be the most stable uranium phase in a nuclear waste repository for spent nuclear fuel in
a deep geological formation. It is therefore assumed to be the uranium solubility controlling phase in long-

term safety assessments. However, there are indications that in nature UO 2 can react in a silica rich
environment (e.g. in contact with ground water or pore water with silica concentration
> 10 −4 mol/L) to coffinite (Langmuir’s criterion) [1]:
UO 2 + H 4 SiO 4 USiO 4 + 2 H 2 O
As SNF consists to > 90% of UO 2 , the consideration of coffinite as a potential secondary phase would
greatly benefit a responsible safety analysis. Coffinite has been studied extensively before; however, most
investigations rely on the use of natural samples. Unfortunately precise thermodynamic data regarding the
stability of coffinite are missing due to the fact that coffinite samples from nature contain typically
significant amounts of impurities. Up to date the synthesis of pure coffinite remains a challenge. However,
the available estimates regarding coffinite stability suggest, that uranium solubility would be significant
lower compared to UO 2 as the solubility controlling phase [2].
Coffinite, USiO 4 and U x Th (1-x) SiO 4 uranothorite solid solutions with x = 0.9 – 0.15 were prepared via a
hydrothermal route. The obtained samples were characterized by X-ray diffraction (XRD) concerning phase
purity and crystallinity. Scanning electron microscopy (SEM) and High resolution transmission electron
microscopy (HRTEM) measurements were performed to further exclude the presence of impurities and
especially minor amounts of the dioxide as far as possible. The short range order was then studied with
EXAFS.
The X-ray diffraction measurements show the expected zircon-type pattern for pure USiO 4 . No other
crystalline uranium phase is observed but amorphous silica is present in the sample. Lattice parameters could
be determined from XRD as c = 6.2610(4) Å and a = 6.9862(2) Å by Rietveld refinement. These values are
slightly smaller than those given by Pointeau et al. [3] but in very good agreement with the ones originally
published by Hoekstra & Fuchs [4] and Fuchs & Gebert [5]. The interatomic distances obtained from
EXAFS are in good accordance with those calculated from X-ray data.
In the XRD measurements U x Th (1-x) SiO 4 uranothorite also exhibits the pattern for spacegroup I 4 1 /amd.
While the presence of the dioxide can be excluded, amorphous silica, and in some samples α-SiO 2 , are
observed as additional phases. According to X-ray data U x Th (1-x) SiO 4 forms a complete solid solution under
hydrothermal conditions. Refined lattice parameters of the uranothorite phase as well as cell volume follow
Vegard’s law and regularly decrease with increasing uranium content. This is in good agreement with the
slightly smaller ionic radius of U 4+ (1.00 Å) compared to Th 4+ (1.05 Å) in the eightfold coordination. [6]
First results from EXAFS measurements of Th L III - and U L III -edge show that the short range order of U and
Th is the same. The interatomic distances derived from the measurements feature a correlation between
increasing uranium content and decreasing metal-oxygen and metal-metal bond-lengths. This strongly
indicates that no local clusters or superstructure exist and hence confirms the presence of a complete solid
solution in the system USiO 4 – ThSiO 4 .
We gratefully acknowledge that beam time at the ROBL beamline (ESRF) was granted through the
ACTINET-i3 (call 4), proposal No. AC4-JRP04.
[1] D. Langmuir, Geochim. Cosmochim. Acta 42, 547-569, (1978).
[2] Ferriss et al.; Amer. Min. 95, 229 – 241, (2010).
[3] V. Pointeau et al., J. Nuc. Mat. 393, 449-458, (2009).
[4] H.R. Hoekstra, L.H. Fuchs, Science, 123, 105-105, (1956).
[5] L.H. Fuchs, E. Gebert, Amer. Min., 43, 243-245, (1958).
[6] Shannon, R. D. Acta Crystallogr., Sect. A, 32, 751–767 (1976).

PA4-8
ON FISSION PRODUCT ALLOY PARTICLES AND THEIR CATALYTIC PROPERTIES
D. Cui (1,2) , S. Hovmöller (2) , W. Wan (2) , Y. Yun (2) , M. Granfors (1) , L. Jeanett (1) , K. Spahiu (3)
(1) Studsvik Nuclear AB, Sweden, (2)Stockholm University, Sweden, (3) SKB, Sweden
Spent nuclear fuel (SNF) contains >90% UO 2 and 3% fission products. Some fission products in SNF exist
as nm to µm sized fission product alloy particles (FPAPs) (26.5 Mo-32.7Ru-7.6Tc-23.1Pd-6.3Rh-4.9Te) [1,
2, 3]. Our recent work proved that FPAPs are both source terms of the radionuclide 99 Tc and catalysts that
can activate molecular H 2 t generated from iron waste canister interaction with groundwater [4]. The
activated hydrogen H . from FPAPs can reduce and immobilize dissolved radionuclide species Np(V), U(VI),
Tc(VII), Pu(VI), Se(VI)/(IV)[4] and slow down or block the dissolution of SNF and FPAPs [5]. To enrich
our knowledge about SNF, A H 3 PO 4 digested residue from SNF was further characterized by leaching test,
XRD, TEM-EDS and 3D electron diffraction. Through above characterizations and rinsing and leaching
experiments, it is proved that in SNF, 99 Tc, 129 I and 79 Se exist in a form that is insoluble in H 3 PO 4 . The
structures of nm sized FPAPs with hexagonal symmetry and lattice parameters (a = 2.840Å and c= 4.572),
were demonstrated for the first time by a recently developed rotating 3D electron diffraction method.
The catalytic capacity of FPAPS was investigated through an isotope experiment between D in D 2 gas phase
and H in H 2 O phase. With presences of a hydrogen catalyst, D 2 can be sorbed on the catalyst, activated to D .
and three isotopic exchange processes should occur, as listed below as Fig. 1.
Fig. 1. Gas compositions in the autoclave with 17ml 10bar D 2 , 33.6ml H 2 O, and 4mg FPAPs
D 2 + H 2 O ↔ DH + DHO
DH + H 2 O ↔ H 2 + DHO
D 2 + DHO ↔ DH + D 2 O
D/H/DH ratios in gas samples in three autoclaves containing water and one of following potential catalysts:
1) FPAPs, 2)UO 2 powder and 3) spent fuel simulator (simfuel) at different times were analyzed by gas mass
spectrometry. H 2 O/DHO/D 2 O ratios in water phase were also measured. In the experiment with FPAPs, as
shown in Fig. 1, all D 2 in gas phase was exchanged by H 2 during 175 days. DH in gas phase increased to
max. value (20%) at 60 days and then gradually decreased. FPAPs are proved to be excellent catalyst, but
other two solids don’t display any significant catalytic effects.
[1] J. Nucl. Mater. 131 (1985) 463
[2] Radiochimica Acta 92(551-555) (2004)
[3] Journal of Nuclear Materials 420, 1-3 (2012)328-33.
[4] Applied Catalysis B: Environmental 94,173–178 (2010)
[5] Energy & Environmental Science 2011, 4, 2537-2545, (2010)

PA4-9
SOLUBILITY OF TcO 2·xH 2 O(s) IN DILUTE TO CONCENTRATED NaNO 3 SOLUTIONS
T. Kobayashi 1) , T. Sasaki 1) , A. Kitamura 2)
1) Department of Nuclear Engineering, Kyoto University, Kyoto daigaku-Katsura, Nishikyo,
Kyoto 615-8540, Japan
2) Radionuclide Migration Research Group, Geological Isolation Research and Development
Directorate, Japan Atomic Energy Agency, Muramatsu, Tokai, Ibaraki, Japan
For the safety assessment of radioactive waste disposal, it is necessary to predict radionuclide solubility
limits under relevant disposal conditions. Technetium ( 99 Tc) as a fission product of 235 U and 239 Pu has a long
half-life (2.1•10 5 y) and is redox sensitive (IV/VII) in a groundwater system. The solubility limit is
controlled by TcO 2·xH 2 O(s) under reducing conditions, while Tc exists as anion TcO − 4 which is highly
soluble and mobile under oxidizing conditions [1]. Some of low-level wastes generated from nuclear
reprocessing process contain considerable amount of nitrate salt, which may affect the redox condition of
groundwater, hence the distribution of Tc species. In the present study, we focus on the solubility of
TcO 2·xH 2 O(s) in the existence of the dilute to concentrated NaNO 3 in order to investigate the potential effect
of nitrate on the Tc species.
Sample solutions were prepared by adding pre-prepared TcO 2·xH 2 O solid phase to 10 ml NaNO 3 , and
NaClO 4 and NaCl solutions as a comparison in an Ar glove box at 25±1ºC. The initial solid phase of
TcO 2·xH 2 O precipitated by reducing NH 4 TcO 4 solution with Na 2 S 2 O 4 and washed with Milli-Q water. The
pH and ionic strength (I) of the sample solutions were set to be pH = 8, 10, and 12.5, and I = 0.01, 0.1, 1 and
6, respectively. No reducing chemical was used in the sample solution. Total Tc concentration was about
5·10 -5 mol/l (M) if the initial solid phase was completely dissolved. After given periods, pH and Eh of the
sample solution were measured and Tc concentration determined after 10kDa (3 nm) ultrafiltration using
Liquid Scintillation Counting (LSC) with a detection limit of about 10 −8 M. In order to investigate a possible
formation of colloidal species, the sample solutions were also filtered with 0.45µm filter to determine Tc
concentration.
At I = 1.0 and 6.0 (NaNO 3 and NaClO 4 ), Tc concentration increased up to near total Tc concentration level
(5·10 -5 M) at neutral pH and slightly increased at pH 12.5. The observed Eh values at pH 8 and 10 were
higher than the calculated Tc(IV)/Tc(VII) equilibrium line [1], and slightly higher than the line at pH 12.5,
indicating that the initial TcO 2·xH 2 O(s) can be oxidized to TcO 4 − . At low ionic strength in NaNO 3 and
NaClO 4 solutions, however, the Tc concentration remains low after 11 weeks in spite of the higher Eh values
than the Tc(IV)/Tc(VII) equilibrium line, suggesting the oxidation and dissolution of TcO 2·xH 2 O(s) at low
ionic strength is extremely slow. In NaCl solutions, the Eh values were similar to those in NaNO 3 and
NaClO 4 solutions, and initial TcO 2·xH 2 O(s) was rapidly dissolved within 2 weeks. In the presence of Cl ion
under oxidizing condition, an accelerated oxidation and dissolution may occur due to a Cl ion-mediated
effect on the surface of TcO 2·xH 2 O(s) [3].
[1] J. Rard, M. Rand, G. Anderegg, H. Wanner : Chemical Thermodynamics of Technetium, Elsevier, North-Holland,
Amsterdam (1999).
[2] T. Kobayashi, A.C. Scheinost, D. Fellhauer, X. Gaona, M. Altmaier, Radiochim. Acta, (2013) in press.
[3] K. Lieser, Ch. Bauscher, T. Nakashima, Dissolution of TcO 2 in aqueous solutions under various conditions,
Radiochim. Acta, 42, 191 (1987).

PA4-10
Surface Science Study of Spent Fuel Corrosion Processes using Thin Film Model Systems
Thomas Gouder
European Commission, Joint Research Centre – Institute for Transuranium Elements (JRC-ITU),
Herrmann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany
Corrosion of spent fuel is a surface process which proceeds via alteration of structure and composition of the
topmost surface layers. This process depends as well on the material properties (grain boundaries, defects,
composition, etc.) as on the environmental properties (pH, oxidants, complexing agents, etc.). Understanding
the mechanisms of these reactions beyond simple rationalization is a difficult task, far from being completed.
It necessitates a multitechnique approach to determine surface and environmental parameters, and the use of
simplified model systems, representative of the fuel, yet avoiding its high complexity ("single parameter
studies").
An overview of present and planned surface science facilities at ITU will be given. Special emphasis is given
to the combination of spectroscopy techniques properties (XPS, UPS, TPD, HREELS, Kelvin probe) probing
the average surface, and mapping techniques probing local reactivity aspects. The projected use of AFM for
studying local reactivity aspects under controlled atmosphere (combined with electrochemistry) will be
presented. Another important aspect is the gap between ultra-high vacuum conditions, imposed for most
surface spectroscopies, and the ambient reaction conditions (aqueous environment) of spent fuel during
storage. To assess it, diagnostic techniques running under both conditions are planned (e.g. HREELS and
Raman, probing vibrational frequencies under UHV and ambient pressure). Fuel model systems and their
preparation by thin film growth methods (sputter co-deposition) will be discussed. Using these techniques,
samples with widely varying composition and oxidation state can be prepared in-situ. The methods are being
used to study the influence of fission products (e.g. the 4d transition metals forming the α-particles) on the
redox potential, and to perform systematic studies of surface reactions of mixed oxides.
All these techniques are now being combined in one single surface science lab station. To allow for
flexibility, a modular setup has been chosen, where the various techniques are mounted in individual
chambers which can be connected to the station on demand.
A recent example of a surprising surface reaction will be presented. We studied the surface interaction of
PuO 2 , UO 2+x , and U-Th and U-Pu mixed oxides (TOX and MOX) with ice. Ice was chosen to fix water in
high concentrations to the surface under ultrahigh vacuum conditions. The study was motivated by the
possible role of water in promoting further oxidation of PuO 2 . To our surprise we found that in presence of
UV light and ice, PuO 2 undergoes a fast surface reduction to Pu 2 O 3, while UO 2+x (x = 0 to 1) is also reduced,
but to a lesser extent. In MOX, UO 2+x reduction is more pronounced, and this is directly related to the
presence of Pu. A similar observation is made for TOX, where U reduction is also faster than in pure UO 2+x .
Surface reduction is shown to be a true interfacial process, where the solid absorbs light and undergoes a
photochemical reaction. It is not a simple surface reaction of the oxides with ice photolysis products.

PA4-11
EFFECT OF NITROUS ACID ON REDUCTION OF Np(VI) IN IRRADIATED SOLUTIONS OF
NITRIC ACID
A. Paulenova 1)* , M. Precek 1,2) , B. Mincher 3) , S. Mezyk 4)
1) Department of Nuclear Engineering, Oregon State University, Corvallis, OR 97331, USA
2) Institute of Physics of the Academy of Sciences of the Czech Republic Na Slovance 1999/2, 182 21
Praha 8, Czech Republic
3)
Division of Radiochemistry and Aqueous Separations, Idaho National Laboratory, Idaho Falls, ID
83415, USA
4)
Department of Chemistry, California State University at Long Beach, Long Beach, CA 90820
In aqueous solutions of nitric acid, typical for radiochemical separations, neptunium may exist as Np(V),
Np(IV) and/or Np(VI). The small cationic charge of the (Np V O 2 ) + species causes its lowest separation yields,
while the higher cationic charges of both (Np VI O 2 ) 2+ and Np 4+ are separated with satisfactory results. The
success of a selected separation method relies on the ability to maintain Npin an appropriate oxidation state.
Regardless of the initial valence state of Np, a series of redox reactions leading to the generation of mixed
valence solution occurs. The valence state of Np in solutions of nitric acid is a function of the concentration
of nitrous acid, [1] as the nitrous acid reduces Np(VI), favoring production of Np(V):
2NpO 2 + + 3H + + NO 3 - ⇋ 2NpO 2 2+ + HNO 2 + H 2 O (1)
On the other hand, HNO 2 appears to behave as a catalyst, promoting the rate of oxidation of higher relative
abundances of Np(V) by nitric acid [1]. Explanation of this behavior [2] lies in the catalytic effect of nitrous
acid, which is facilitated by the • NO 2 radical from N 2 O 4 , which is formed by interaction of nitric and nitrous
acids:
• NO 2 + • NO 2 + H 2 O ⇋ N 2 O 4 + H 2 O ⇋ HNO 2 + H + -
+ NO 3 (1)
The • NO 2 radical then acts as the actual oxidizing agent:
NpO 2 + + • NO 2 → NpO 2 2+ + NO 2
- (2)
and the oxidation of Np(VI) in the presence of low concentrations of HNO 2 may actually be due to reaction
of Np(V) with the nitrogen dioxide radical ( • NO 2 ), rather than with HNO 2 itself. [3] This study was
conducted to analyze the effect of gamma radiation on Np redox speciation in aqueous nitric acid solutions,
primarily to identify the importance of the radiolytic production of nitrous acid. Dependence of radiolytic
yield of HNO 2 on HNO 3 concentration in aqueous and organic phases was measured. It was found that
generation of HNO 2 levels off down due to decreased availability of H 2 O molecules for the reaction with the
• NO 2 radical, which is an important source of HNO 2 through a disproportionation reaction [3]:
2 • NO 2 + H 2 O → HNO 2 + HNO 3 (3)
The • NO 2 radicals then instead increasingly contribute to the production of nitrated organic compounds by
reacting with organic radicals resulting from the radiolysis of 30%TBP/n-dodecane, such as are the
hydrocarbon radicals and radicals resulting from TBP radiolysis [4]:
• NO 2 + • R → NO 2 -R (5)
• NO 2 + (C 4 H 9 O) 2 ( • C 4 H 8 O)PO → (C 4 H 9 O) 2 (O 2 N-C 4 H 8 O)PO (4)
[1] Siddal, T H; Dukes, E K., J. Am. Chem. Soc., 1959, 81(4), 790–794.
[2] Tochiyama, et al. J. Nucl. Sci. Techn., 1995, 32(2), 118–124.
[3] Vladimirova, M.V. Radiochemistry, 1995, 37(5), 410–416.
[4] Mincher, B. J.; Modolo, G.; Mezyk, S. P. Solvent Extr. Ion Exch, 2009, 27(1), 1–25

PA6
PA6-1
PA6-2
PA6-3
PA6-4
PA6-5
PA6-6
PA6-7
PA6-8
COLLOID FORMATION
INTERACTION OF RARE EARTH ELEMENTS AND SUSPENDED MATTERS
CONTAINED IN HORONOBE DEEP GROUNDWATER
A. Kirishima, A. Kuno, H. Amamiya, H. Murakami, Y. Amano, T. Iwatsuki, T. Mizuno, T.
Kubota, T. Sasaki, N. Sato (Japan)
SIZE AND ELEMENTAL COMPOSITION ANALYSES OF GRANITIC
GROUNDWATER BY FLOW-FIELD FLOW FRACTIONATION
T. Hamamoto, T. Saito, S. Tanaka (Japan)
SYNTHESIS AND CHARACTERIZATION STUDIES OF SONOLYTIC COLLOIDAL
SPECIES OF PLUTONIUM(IV)
V. Morosini, T. Chave, P. Moisy, C. Den Auwer, S. Nikitenko (France)
SONOCHEMICAL REDUCTION OF Pu(IV) IN AQUEOUS NITRIC SOLUTIONS
M. Virot, L. Venault, P. Moisy, V. Morosini, S. I. Nikitenko (France)
MONTMORILLONITE COLLOID SIZE HETEROGENEITY - IMPACT ON
RADIONUCLIDE SORPTION CAPACITIES
K.K. Norrfors, M. Bouby, S. Heck, N. Finck, R. Marsac, T. Schäfer, H. Geckeis, S. Wold
(Germany, Sweden)
MONTMORILLONITE COLLOID SIZE HETEROGENEITY - IMPACT ON
STABILITY IN SUSPENSION
K.K. Norrfors, M. Bouby, S. Heck, N. Finck, R. Marsac, T. Schäfer, H. Geckeis, S. Wold
(Germany, Sweden)
ACTINIDE SOLUBILITY AND SPECIATION IN THE WASTE ISOLATION PILOT
PLANT (WIPP) REPOSITORY
D.T. Reed, M. Borkowski, J.S. Swanson, M.K. Richmann, J.F. Lucchini, K. Simmons, D.
Cleveland (USA)
ACTINIDE COLLOIDS AND NANOPARTICLES: RELEVANCE TO LEGACY
WASTE, CLEAN-UP AND GEOLOGICAL DISPOSAL
J. Rochford, S. Heath (UK)

PA6-1
INTERACTION OF RARE EARTH ELEMENTS AND SUSPENDED MATTERS CONTAINED IN
HORONOBE DEEP GROUNDWATER
A. Kirishima 1) , A. Kuno 1) , H. Amamiya 2) , H. Murakami 2) , Y. Amano 2) , T. Iwatsuki 3) , T. Mizuno 2) ,
T. Kubota 4) , T. Sasaki 5) , N. Sato 1)
1) Institute of Multidisciplinary Research for Advanced Materials (IMRAM), Tohoku University, Katahira 2-
1-1, Aoba-ku, Sendai 980-8577, Japan
2) Horonobe Underground Research Unit, Geological Isolation Research and Development Directorate,
Japan Atomic Energy Agency (JAEA), Hokushin 432-2, Horonobe-cho, Hokkaido 098-3224, Japan
3) Tono Geoscientific Research Unit, Geological Isolation Research and Development Directorate, Japan
Atomic Energy Agency (JAEA), Yamanouchi 1-64,Akeyo-cho,Mizunami city, Gifu 509-6132, Japan
4) Kyoto University Research Reactor Institute, 2, Asashiro-Nishi, Kumatori-cho, Sennan-gun, Osaka 590-
0494, Japan
5) Department of Nuclear Engineering, Kyoto University, Kyoto daigaku-Katsura, Nishikyo, Kyoto 615-8530,
Japan
For better understanding of the migration behavior of minor actinides (MA) in groundwater, the interaction
of rare earth elements (REEs) and suspended matters contained in Horonobe deep groundwater was studied.
The groundwater sample was collected under inert gas in a stainless bottle from a packed section in a
borehole drilled to 140 m depth in drift at Horonobe underground research laboratory (URL), Hokkaido,
Japan. The Horonobe URL was constructed for the geoscientific research of sedimentary rock strata and it is
known that the groundwater from this site contains a relatively high concentration of fossil sea water. 10 ppb
of the rare earth elements, i.e., Y, La, Ce,Pr,Nd,Sm,Eu,Gd,Tb,Dy,Er,Tm and Yb were
spiked to the sample groundwater as chemical analogues of trivalent actinides. Then, that groundwater was
sequentially filtrated using 0.2 µm, 10 kDa, 3 kDa and 1 kDa filters under inert gas conditions. After that, the
filtrate solutions were analyzed by ICP-MS to determine the concentrations of the solution fraction of the
REEs at each filtration step. On the other hand, the surface of spent filters were analyzed by the neutron
activation analysis (NAA) conducted at Kyoto university research reactor (KUR), SEM-EDX and TOF-
SIMS element mapping to determine the amount and chemical state of collected fraction of the REEs on
each pore size filter.
The result of filtrate solution analysis by ICP-MS indicated that less than 30 % of La, Ce, Pr, Nd, Sm, Eu
passed through the 0.2 µm filter, while the collection amount of these elements were decreased in the order
of filter pore sizes 10 kDa, 3 kDa and 1 kDa. An ionic radius dependency was observed in the ratio of the
solution fraction of REEs for each filtration steps, thereby, those ratios of heavier REEs (Gd, Tb, Dy, Ho, Er,
Tm, Yb) and Y were always higher than those of lighter REEs having larger ionic radius. From the result of
NAA, it was shown that almost same amount of REEs were collected on the 0.2 µm and 10 kDa filters while
much less was on the 3 kD and 1 kDa filters when it was standardized by the volume of filtrated
groundwater. The equilibrium speciation calculation indicates that 95 - 100 % of the spiked REEs precipitate
as phosphate salts under the chemical composition and condition of the Horonobe groundwater. The element
mapping of the spent filter surface by SEM-EDX and TOF-SIMS expressed that REEs were collected as
phosphate particle on the 0.2 µm filter, and were collected as a pseudo colloids with hydro-carbon on the 10
kDa filter.
These results indicate that the major fraction of the spiked REEs precipitated at the bottom of the stainless
bottle as the rare earth phosphate and dispersed particles of that phosphate were collected by the 0.2 µm
filter, furthermore, the rest of REEs formed pseudo colloids with suspended organic matters in the
groundwater and passed through the 0.2 µm, 10 kDa and 3 kDa filters depending on their particle sizes.
From this chemical analogue study, it is suggested that the migration behavior of MA in the Horonobe
groundwater system seems to be regulated by the formation of phosphate precipitation and small percentage
of MA would be carried in the groundwater as pseudo colloids like MA-humic substance complex.

PA6-2
SIZE AND ELEMENTAL COMPOSITION ANALYSES OF GRANITIC GROUNDWATER BY
FLOW-FIELD FLOW FRACTIONATION
T. Hamamoto 1 , T. Saito 2 , S. Tanaka 1
1 Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-
1 Hongo, Bunkyo-ku, Tokyo 113-8656, Japan
2 Nuclear Professional School, School of Engineering, The University of Tokyo, 2-22 Shirakata Shirane,
Tokai-mura, Ibaraki, 319-1188, Japan
In geological disposal of radioactive wastes, colloids may have large impacts on the migration of
radionuclides in geosphere. However, qualitative and quantitative understanding of groundwater nanocolloids
is still premature for the safety assessment of geological disposal, partly because of limited
availability of deep groundwater samples and also because of low concentration and complexity of colloids
there. Flow-field flow fractionation (FlFFF) enables continuous size fractionation of colloids without
significantly disturbing their physicochemical conditions. By combining with ICP-MS, one can
simultaneously analyze the size distribution and elemental composition of colloids and trace elements
trapped by colloids. In this study, size fractionation of nano-colloids in granitic groundwater was performed
by FlFFF with gradually decreasing cross flow in order to extend the measured size range; major elements
constituting colloids in the groundwater and guest trace elements associated with them were assessed from
the ICP-MS measurements of the fractionated colloids.
Asymmetrical FlFFF (Postnova, AF2000 FOCUS) with a 300 Da ultrafiltration membrane was used.
The effluent was 5 mM NaCl, which passed through a UV/Vis detector (SPD-20A, Shimadzu) and a
fluorescence detector (RF-10AXL, Shimadzu) for the determination of organic colloids, and ICP-MS
(7500cx, Agilent) for the elemental compositions. In this study, we used programmed cross flow to extend
the range of size fractionation; the cross flow rate was decreased from 0.9 to 0 mL/min over 30-min of the
elution step, which corresponded to the fractionation up to 100 nm (in diameter). The cross flow was kept 0
mL/min for additional 30 min as the washing step, where colloids in 100-450 nm were recovered. Linear
calibration curves between the retention times and hydrodynamic diameters were obtained, using five
globular proteins. The methods for groundwater sampling and pre-treatment were similar to those in the
previous study [1], except that preliminary evacuated Teflon-coated glass vials were used for the sampling in
this study. Granitic groundwater was sampled at the 09MI21 borehole at the access/research gallery at a
depth of 300 m in the Mizunami Underground Research Laboratory, operated by Japan Atomic Energy
Agency.
The size distributions of organic colloids detected by UV/Vis and fluorescence and 18 elements by ICP-
MS were obtained, among which Si, Al, and Ca were relatively abundant and, thus, considered to be major
constituents of inorganic colloids. Some examples of the obtained size distributions were shown in Figure 1.
The left sides of each graph are the elution step and the right sides are the washing step. Organic colloids
existed in a relatively small size range (< 10 nm), where all elements except Al and Si were found as well.
Inorganic colloids consisting of Al and Si as major constituents were found at > 80 nm together with Mg and
Fe. Natural actinide elements (U and Th) were also associated with the inorganic colloids, although the
concentrations of these elements were relatively small. The observed trends were similar among the
groundwater samples obtained from the different sections of the borehole, suggesting the presence of small
organic colloids and large inorganic colloids as potential carriers of trace elements in this groundwater.

Figure 1. Size distribution of organic colloids and Al, Fe, Eu in the groundwater sampled in the 09MI21 borehole. The
right sides are the washing step where colloids in 100-450 nm were recovered.
[1] Saito, T.; Suzuki, Y.; Mizuno, T., Size and elemental analyses of nano colloids in deep granitic groundwater:
Implications for transport of trace elements. Colloids Surf. A, in press.
PA6-3
SYNTHESIS AND CHARACTERIZATION STUDIES OF SONOLYTIC COLLOIDAL SPECIES
OF PLUTONIUM (IV)
V. Morosini 1) , T. Chave 1) , P. Moisy 2) , C. Den Auwer 3) , S. Nikitenko 1)
1) Institut de Chimie Séparative de Marcoule, ICSM-UMR5257 CEA/CNRS/UM2/ENCSM, Site de
Marcoule, Bât. 426, BP 17171, 30207 Bagnols sur Cèze, France
2) Département RadioChimie et Procédés, Direction de l’Energie Nucléaire, CEA/DEN/DRCP,
Site de Marcoule 30207 Bagnols sur Cèze, France
4) Institut de Chimie de Nice, ICN-UMR7272 CNRS/UNS, 06108 Nice, cedex 2, France
Since their discovery, production and first use, radioactive contaminant particles, also known as “hot
particles”, are of a great concern, because of their potential hazards in the environment. Their release is
mostly due to nuclear accidents, i.e. at Chernobyl or Fukushima and military tests before 1973. Formation,
structure and behavior of nano/micrometer-sized actinide particles containing colloids are of a particular
interest, considering their potential to be transported in the hydrosphere and geosphere, and consequently
being ultimately transferred in the biosphere.
The main objective of this work is to improve our knowledge about the mechanisms of formation, the
composition and the chemical reactivity of plutonium (IV) colloids. To this end, discrete and structurecontrolled
plutonium particles prepared in well-defined conditions are needed.
In this work is highlighted for the first time the sonochemical synthesis of plutonium colloids from pure
water dispersed plutonium oxide PuO 2 under a 20 kHz ultrasounds field (P US =0.34 W.mL -1 ) under argon or
argon/carbon monoxide atmosphere.
Sonochemistry is based on ultrasound effects arise from acoustic cavitation. In fact, when submitted
into a liquid, power ultrasound induces the nucleation, growth, and violent collapse of cavitation bubbles.
This is responsible for in situ radical formation. In pure water, homolytic split of H 2 O molecules leads to the
formation of H 2 within the bubble and H 2 O 2 at the bubble–liquid interface due to the mutual recombination
of H atoms and OH . radicals, respectively.[1]

H 2 O ))))) H + HO .
2H H 2
2HO . H 2 O 2
The violent collapse of the cavitation bubble creates a “hot spot” where the energy stored is released,
enabling localized extreme pressure and temperature.[2-4] Additionally, the collapsing bubble emits a shock
wave [5] and in case of heterogeneous cavitation, the implosion of the bubble next to a solid surface leads to
microdamages on sonicated surfaces, which along with other effects such as microstreamings and
microturbulences can lead to phenomena such as erosion, fusion, fragmentation, inclusion, dissolution.[6-10]
According to that, the chemical and physical aspects of sonochemistry lead to the formation of stable
plutonium (IV) colloids.
The stability over time of the as-prepared colloids, along with variations of salinity level and their redox
behavior are studied. The speciation of these plutonium colloids are carried out by XAS and UV/Vis
spectroscopy.
[1] Fischer, C.H., Hart, E.J., Henglein, A., J. Phys. Chem., 1986. 90(9): p. 1954-1956.
[2] Bernstein, L.S., et al., J. Phys. Chem., 1996. 100(16): p. 6612-6619.
[3] Didenko, Y.T., Pugach, S.P. J. Phys. Chem., 1994. 98(39): p. 9742-9749.
[4] Engel, V., et al., J. Phys. Chem., 1987. 87(8): p. 4310-4314.
[5] Pecha, R., Gompf, B., Phys. Rev. Lett., 2000. 84(6): p. 1328-1330.
[6] Kuppa, R., Moholkar, V.S., Ultrason. Sonochem., 2010. 17(1): p. 123-131.
[7] Mason, T.J.L., John P., J. Chem. Tech. & Biotech. Vol. 79. 2004
[8] Prozorov, T., Prozorov, R., Suslick, K.S., JACS, 2004. 126(43): p. 13890-13891.
[9] Suslick, K.S. and G.J. Price, Annual Review of Materials Science, 1999. 29: p. 295-326.
[10] Virot, M., et al., J. Phys. Chem. C, 2010. 114(30): p. 13083-13091.
PA6-4
SONOCHEMICAL REDUCTION OF Pu(IV) IN AQUEOUS NITRIC SOLUTIONS
M. Virot 1) , L. Venault 2) , P. Moisy 2) , V. Morosini 1) , S. I. Nikitenko 1)
1) ICSM-UMR5257 CNRS/CEA/UM2/ENCSM, Site de Marcoule,
BP 17171, 30207 Bagnols sur Cèze, France
2) CEA/DEN/DRCP, Radiochemistry and Process Department,
BP 17171, 30207 Bagnols sur Cèze, France
The propagation of ultrasound (16 kHz < f < 1 MHz) in a liquid media may lead to acoustic cavitation which
is the nucleation, growth, and rapid implosive collapse of vapour filled micro-bubbles. In homogeneous
systems, the implosion of these bubbles goes with the in-situ generation of transient extreme conditions in
terms of temperature and pressure which promote the formation of excited species and free radicals in
solution [1,2]. In heterogeneous systems, acoustic cavitation involves the production of micro-jets and shock
waves at the vicinity of the treated solid boundary which can be related to several applications such as
erosion, extraction, cleaning, dissolution, etc. [1-3]. Because of these particular occurrences, power
ultrasound has been showed to enhance the yields and accelerate the reaction rates of numerous
homogeneous and heterogeneous chemical systems.
In this study, the behaviour of Pu(IV) in aqueous nitric solutions is considered under 20 kHz ultrasound
irradiation. The experiments are performed at ~25°C under Ar flow in a 50 mL home-made reactor and the
reactions are followed using a UV-Vis spectrometer. In the presence of HNO 2 scavengers (ex: sulfamic acid
or hydrazinium nitrate), ultrasound irradiation of Pu(IV) leads to the accumulation of Pu(III) in solution.
Pu(IV) reduction rate is found to be function of the acoustic intensity delivered to the solution. The reduction
follows a first-order reaction mechanism and is not complete. In addition, a steady-state is observed for
prolonged irradiation. The behaviour of Pu(IV) and Pu(III) in this system is attributed to reactions with
sonochemically generated hydrogen peroxide, H 2 O 2 , in agreement with the following reactions that can also
be observed without ultrasound application [4].

2 Pu 4+ + H 2 O 2 → 2 Pu 3+ + O 2 + 2 H +
2 Pu 3+ + H 2 O 2 + 2 H + → 2 Pu 4+ + 2 H 2 O
The kinetics observed for Pu(IV) sonochemical reduction are however found to be faster than the kinetics
related to H 2 O 2 generation under ultrasound at similar conditions. The various investigations show that this
difference can be attributed to the mechanical erosion of the ultrasonic tip resulting from the acoustic
cavitation generated at the vicinity of the solid surface. The erosion of the tip produces titanium particles and
their subsequent solubilization in the sonicated solution (checked with SEM and ICP-AES) may be involved
for reactions with Pu(IV) in agreement with the following equation.
Ti (s) + 3 e - → Ti 3+
Ti 3+ + H 2 O → TiO 2+ + 2 H + + e -
Pu 4+ + Ti 3+ + H 2 O → Pu 3+ + TiO 2+ + 2 H +
[1] K. S. Suslick, D. J. Flannigan, Ann. Rev. Phys. Chem. 59, 659 (2008).
[2] T. G. Leighton, The Acoustic Bubble, Academic Press: London (1994).
[3] M. Virot, T. Chave, S. I. Nikitenko, D. G. Shchukin, T. Zemb, H. Mohwald, J. Phys. Chem. C 114,
13083 (2010).
[4] J. M. Cleveland, The Chemistry of Plutonium, American Nuclear Society, La Grange Park (1979).
PA6-5
MONTMORILLONITE COLLOID SIZE HETEROGENEITY
- IMPACT ON RADIONUCLIDE SORPTION CAPACITIES
K. K. Norrfors 1),2) , M. Bouby 1) , S. Heck 1) , N. Finck 1) , R. Marsac 1) , T. Schäfer 1) , H. Geckeis 1) , S. Wold 2)
1): Institute for Nuclear Waste Disposal (INE), Karlsruhe Institute of Technology (KIT), P.O. Box 3640, D-
760 21 Karlsruhe, Germany
2): Department of Chemistry/Applied Physical Chemistry, KTH Royal Institute of Technology, Teknikringen
30, SE-100 44 Stockholm, Sweden
Bentonite buffers may release colloidal particles (montmorillonite) under certain conditions, i.e. after a
glacial melt water intrusion, which thus may act as radionuclide (RN) carriers. During transport in bedrock
fractures, the colloids may be separated according to their size. Smaller colloids, presenting a larger exposed
surface area suggesting then a higher RNs sorption potential 1 and a facilitated transport, may enhance the
RNs mobility, if transported longer then the larger colloids. Therefore, it is of high interest to carefully study
the effect of the montmorillonite colloid size on their radionuclide sorption capacities, namely K D -values.
Conceptual models used for RNs and colloidal transport 2 do not take into account either eventual size
heterogeneity effects or sorption kinetics. Therefore, this kind of study can contribute to improvements of
these transport models.
MX-80 bentonite was equilibrated with a carbonated synthetic groundwater (SGW) of low ionic strength
(simulating ice age melt water). After a first sedimentation step, seven colloidal clay suspensions, S i , were
obtained by sequential (ultra-)centrifugation at different speeds/times in presence or not of organic matter
(fulvic acids, FA). The stability of the colloids was examined in different electrolytes and pH in a separate
study 3 . The size distribution of each colloid suspension was characterized by Photon Correlation
Spectroscopy (PCS) and Asymmetric Flow Field-Flow Fractionation (AsFlFFF)/UV-Vis/LLS/ICP-MS.
Broad size distributions were obtained with peak maxima positions located at 55 nm, 80 nm, 100 nm, 130
nm, 160 nm, 170 nm and 210 nm. The ion and colloid concentrations in all suspensions have been measured
by IC and ICP-OES. Those analyses indicate the presence of calcite on the colloids. They suggest the
dissolution of NaCl, gypsum and celestite accompanied by ionic exchange processes. The mineral phase
composition, analyzed by XRD, does not change for the different colloidal suspensions. Two types of
sorption experiment were then performed.

Batch sorption tests in SGW were first performed by using 233 U, 242 Pu, 232 Th, 99 Tc and 237 Np 4 on each of the
seven isolated montmorillonite colloidal suspensions and followed over time. At a first glance, it appears that
the 232 Th and the 242 Pu are associated with the clay colloids independently of their size, while the 99 Tc, 237 Np
and 233 U (due to the presence of carbonates) are not clay-colloid bound, whatever the colloidal size, and
remain in suspensions. These three elements do not form eigen-colloids under the present conditions. One
has to note the release of naturally 238 U present in the colloidal suspensions, attributed to the presence of
carbonates in the synthetic water used, which seems to be enhanced for the smallest size fractions.
A second sorption test was carried out in SGW, starting with the first clay colloidal suspension S 0 , obtained
after the initial sedimentation step only. Three elements, 152 Eu, 232 Th and 238 U, were first added to this
colloidal suspension. After equilibration during 2 weeks, the S 0 suspension was analyzed by AsFlFFF/UV-
Vis./LLS/ICP-MS and then sequentially centrifuged and re-analyzed after each centrifugation step. The aim
was to detect a preferential sorption onto specific size class and to compare these additional results to those
obtained in the first sorption test. Data treatment is in progress.
Complementary sorption reversibility/desorption tests are on-going. They consist in: 1) varying, after
sorption, the chemical composition of the clay colloid suspensions by (i) decreasing the pH, (ii) increasing
the ionic strength or (iii) adding FA or crushed granodiorite as competing ligand and mineral surface,
respectively, 2) with time follow the elemental release obtained from RNs loaded solid clay residues,
presenting the different size fractions, isolated and placed in contact with fresh synthetic carbonated water.
A summary of the results achieved so far and the conclusions drawn will be discussed.
[1] A.S. Madden, M.F. Hochella, T.P. Luxton, Insights for size-dependent reactivity of hematite nanomineral surfaces
through Cu 2+ sorption, Geochim. Cosmochim. Acta, 70 (2006) 4095-4104.
[2] F. Vahlund, H. Hermansson, A direct numerical approach to solving the transport equations for radionuclide
transport in fractured rock, SKB internal report R-04-50, (2006).
[3] K.K. Norrfors, M. Bouby, S. Heck, N. Finck, R. Marsac, T. Schäfer, H. Geckeis, S. Wold, Montmorillonite colloid
size heterogeneity - Impact on stability in suspension (contribution submitted for a poster presentation), in: Migration
2013, Brighton (UK), 2013.
[4] S. Wold, Sorption of prioritized elements on montmorillonite colloids and their potential to transport radionuclides,
SKB internal report TR-10-20, (2010).
PA6-6
MONTMORILLONITE COLLOID SIZE HETEROGENEITY - IMPACT ON STABILITY IN SUSPENSION
K. K. Norrfors 1),2) , M. Bouby 1) , S. Heck 1) , N. Finck 1) , R. Marsac 1) , T. Schäfer 1) , H. Geckeis 1) , S. Wold 2)
1): Institute for Nuclear Waste Disposal (INE), Karlsruhe Institute of Technology (KIT), P.O. Box 3640, D-76021
Karlsruhe, Germany
2): Department of Chemistry/Applied Physical Chemistry, KTH Royal Institute of Technology, Teknikringen 30, SE-100
44 Stockholm, Sweden
Highly compacted bentonites are envisaged as an engineered barrier in most designs of high level radioactive
waste repositories. The main qualities of bentonite are a low hydraulic permeability that minimizes the water
intrusion to the canister, a high sorption capacity and a stable environment where radionuclides (RNs) can be
transported only by diffusion. Bentonites are composed mainly of clay minerals from the smectite group.
Smectites are usually of very small size (< 2 µm) and can easily be dispersed in water. If the chemical
conditions ensure a colloidal stability (low salinity as in ice age fresh melt water), bentonite colloidal
particles can be released and act as carriers by transporting RNs over long distances. Given the physical
properties of the media (fracture geometry, presence of fracture filling material, etc.), the transport of
nanoparticles of small sizes may be favoured over that of larger colloids or vice-versa. Therefore, the
stability of clay colloids according to their size heterogeneity is one of the key questions for predicting their
influence on the RNs migration.
In this study, unpurified MX-80 bentonite was first equilibrated and sedimented in low ionic strength
carbonated synthetic ground water. Sequential and direct centrifugations have been used to separate and
isolate smectite (montmorillonite) colloidal fractions. The separation was performed after a first equilibration

and sedimentation step in carbonated synthetic groundwater with low ionic strength. Seven colloidal clay
suspensions of various size distributions were obtained and characterized by PCS and AsFlFFF/UV-
Vis/LLS/ICP-MS. IC and ICP-OES have been used to determine the ion and colloid concentrations in all
suspensions, and the mineral phase composition was analyzed by XRD. Stability studies were performed at
pH 7 and 8 in different ionic strengths (ISs) (0.01M - 3M) in NaCl, CaCl 2 , MgCl 2 electrolytes according to
the experimental protocol described in [1]. Stability ratios (W) were calculated under these experimental
conditions after monitoring the initial agglomeration rate by PCS measurements. Additional stability studies
were performed with well-characterized carboxylated polystyrene nanoparticle standards (LC) of different
sizes (24nm, 40nm, 60nm, 81nm, 97nm, 217nm, 420nm and 497nm) in order to determine more precisely
the influence on the stability of the exposed surface area, the total mass, the number of particles and the total
charge present initially in the samples.
The results show that addition of 0.1M NaCl or 0.03M IS CaCl 2 or MgCl 2 is enough to destabilize all
montmorillonite fractions, even though it is hard to determine the initial agglomeration rate for the largest
montmorillonite colloidal fraction since these data are widely scattered. The experimental data further show
that the stability ratios depend on the number of particles present in the samples, as expected, for both the
fractionated montmorillonite and the LC. This is deduced from the decrease in W with the particle size in the
samples having the same initial mass concentration. Additionally, it was found that increasing the particle
size results in a faster initial agglomeration rate for the LC, when comparing the samples with the same
initial number of particles, which is in line with the literature [2]. Furthermore, MgCl 2 affects more the
colloidal montmorillonite suspensions than CaCl 2 , as explained by the Schulze-Hardy rule.
[1] R. Kretzschmar, H. Holthoff, H. Sticher, Influence of pH and Humic Acid on Coagulation Kinetics of Kaolinite: A
Dynamic Light Scattering Study, Journal of Colloid and Interface Science, 202 (1998) 95-103.
[2] S.H. Behrens, D.I. Christl, R. Emmerzael, P. Schurtenberger, M. Borkovec, Charging and Aggregation Properties of
Carboxyl Latex Particles: Experiments versus DLVO Theory, Langmuir, 16 (2000) 2566-2575.
PA6-7
ACTINIDE SOLUBILITY AND SPECIATION IN THE WASTE ISOLATION PILOT PLANT
(WIPP) REPOSITORY
D. T. Reed, M. Borkowski, J. S. Swanson, M. K. Richmann, J. F. Lucchini,
K. Simmons, and D. Cleveland
Earth and Environmental Sciences Division,
Los Alamos National Laboratory, Carlsbad NM, 88220, USA
The Waste Isolation Pilot Plant (WIPP) transuranic repository remains a cornerstone of the U.S. Department
of Energy's (DOE) nuclear waste management effort. Waste disposal operations began at the WIPP on
March 26, 1999 but a requirement of the repository license is that the WIPP needs to be recertified every five
years for its disposal operations. The WIPP is now pursuing it third recertification (to be submitted in March
2014) and there are many ongoing discussion about possible expanded missions and additional nuclear
repository concepts in a Salt geology.
The overall ranking of actinides, from the perspective of potential contribution to release from the WIPP, is:
Pu ~ Am > U >> Th and Np and remains unchanged from past recertifications. The oxidation state
distribution of key multivalent actinides also remains unchanged: U – 50% U(IV) and 50% U(VI); Pu – 50%
Pu(III) and 50% Pu(IV); with Am/Cm as the III oxidation state and thorium as the IV oxidation state. In this
recertification cycle, thorium solubility studies in brine were completed and the PA approach to define the
continuation of colloids to the actinide source term was re-examined and will be updated. These data [1-4]
continue to extend our understanding of high ionic-strength actinide chemistry.
Solubility of thorium in WIPP brine
After equilibration for two years in carbonate-free brine, the measured solubility of thorium was 6-7×10 -7 M
and was essentially independent of pH, brine composition and carbonate concentration from pC H+ = 6.5 to
11.5 (see Figure 1). In this timeframe, sequential filtration to ~ 10 nm pore size and ultracentrifugation up to

1,000,000 g resulted in a significantly lower ( 9. These longer-term data are in relatively good agreement with the model-predicted thorium
solubility.
Colloidal fraction of the actinides in the WIPP
The mineral, intrinsic and microbial contribution to the WIPP mobile colloidal actinide source term model
was re-examined in light of recent literature results and new WIPP-specific data. Colloidal species can
potentially contribute to the WIPP actinide source term and are currently accounted for in WIPP PA. In the
current model, four types of colloids are identified: intrinsic, mineral, microbial and humic.
Our WIPP-specific experiments indicate that nano-sized intrinsic colloids (typically < 10 nm) are almost
always present in brine, even after multiyear equilibration times. The biosorption of actinides towards WIPPindigenous
microorganisms was also determined and used to update the bio-colloid model. These new data,
although not complete, provide significant improvement in our understanding of the potential contribution of
colloidal species to the WIPP actinide source term. Additionally, some inconsistencies between the known
solution chemistry and literature observation are addressed. A summary of the updated parameters and
supporting data will be provided.
Figure 1. The concentration of
thorium measured in WIPP
simulated brines both ERDA-6
and GWB) as a function of
time, filtration, and the
presence of carbonate (circles:
oversaturation, squares: undersaturation).
Although high, but
meta-stable, concentrations
were initially present, in time
the measured concentration
decreased to values that are
conservatively below the
WIPP model-predicted values.
Th(IV) Concentration, M
1E-4 - GWB 2 years CO2-free
- ERDA 2 years CO2-free
- GWB ultracentrifuged
1E-5
- GWB 4 years CO2-free
- GWB 2 years with carbonate
- model calculated
1E-6
1E-7
1E-8
1E-9
1E-10
8.5 9.0 9.5 10.0 10.5
pC H+
[1] J. S. Swanson, D. M. Norden , H. M. Khaing, and D. T. Reed (2013). “Degradation of Organic Complexing Agents
by Halophilic Microorganisms in Brines,” Geomicrobiology Journal, vol. 30:3, 189-198.
[2] D. A. Ams, J. S. Swanson, J.E. S. Szymanowski, J. B. Fein, M. Richmann, and D. T. Reed (2013). “The Effect of
High Ionic Strength on Neptunium (V) Adsorption to a Halophilic Bacterium,” Geochimica et Cosmochimica
Acta, vol. 110: 45-57.
[3] M. Borkowski, M.K. Richmann, J. F. Lucchini, and D. T. Reed (2012). “Solubility of An(IV) in WIPP Brine:
Thorium Analogy Studies in WIPP Simulated Brine,” LA‐UR 12‐24417, Los Alamos National Laboratory report,
Carlsbad Operations.
[4] D. T. Reed, J. S. Swanson, J. F. Lucchini, and M.K. Richmann (2013). “Intrinsic, Mineral and Microbial Colloid
Enhancement Parameters for the WIPP Actinide Source Term,” LA‐UR 13‐20858, Los Alamos National
Laboratory report, Carlsbad Operations.
[5] M. Altmaier, V. Neck, and Th. Fanghänel (2004). “Solubility and Colloid formation of Th(IV) in concentrated
NaCl and MgCl 2 Solutions.” Radiochimica Acta, vol. 92: 537–43.

PA6-8
ACTINIDE COLLOIDS AND NANOPARTICLES: RELEVANCE TO LEGACY WASTE, CLEAN-
UP AND GEOLOGICAL DISPOSAL
J. Rochford and S. Heath.
Centre for Radiochemistry Research, School of Chemistry, The University of Manchester, Brunswick Street,
Manchester, M13 9PL, UK.
The hydrolysis of actinides needs to be understood at a molecular level to ensure thorough decontamination
of waste water streams in effluent treatment facilities and if final disposal is to take place. The hydrolysis of
actinides can be more fully understood by attempting to control complexation and hydrolysis reactions by
restricting the number of free binding sites that are available for hydrolysis. The speciation of early actinides
with simple organic ligands, primarily those which can occur in storage ponds and effluent streams, have
been studied in aqueous systems over a wide pH range. The effects of eight ligands (thme, TEA, bicine,
Heidi, NTA, ADA, citric acid and 5Me-HXTA) on the speciation and hydrolysis of Th 4+ and UO 2+ 2 have
been studied using various methods including
1 H/ 13 C NMR spectroscopy, UV/Vis spectroscopy,
potentiometry and elemental analysis. The results have been interpreted using speciation modelling
(PHREEQC).
The alcohol functionalities show no binding between 2 < pH < 13. The pK a of an alcohol group is typically
around 14 and hydrolysis has been seen at pH 5 for both Th 4+ and UO 2 2+ ; so occurs before deprotonation and
binding can take place. Reflecting this, at the high pH found in storage ponds, alcohol groups will not alter
the chemistry of the actinides. Ligands containing one carboxylic acid functional group have been observed
to bind between 2 < pH < 5. Through the presence of broad signals in the NMR, it has been determined that
the ligands are fluxional and that as pH increases, hydrolysis occurs with the precipitation of hydroxide
species. As further carboxylate groups are incorporated in the ligand, they become multidentate, and as a
consequence, they become less labile preventing hydrolysis until the solution is at a higher pH. In some
instances (citric acid and NTA) the actinide species is soluble up to pH 12.
Ligands containing both carboxylic acid and amide groups have also been investigated. At pH < 7 both the
amide and carboxylate groups bind to the actinide, stabilising the species with respect to hydrolysis.
However, as pH is increased further, only the carboxylate groups bind to the metal allowing partial
hydrolysis to occur on the free site and by pH 9 a precipitate forms.
The functionalities of organic ligands have been shown to have a profound effect on the hydrolysis and
partial hydrolysis of the actinides. This work is crucial to the nuclear industry and its environment impact.

PB2
PB2-1
PB2-2
PB2-3
PB2-4
PB2-5
PB2-6
PB2-7
PB2-8
PB2-9
PB2-10
PB2-11
PB2-12
PB2-13
PB2-14
DIFFUSION AND OTHER MIGRATION PROCESSES
DIFFUSION PROPERTIES IN LOW PERMEABILITY MEDIA
Y. Xiang, D. Loomer, T. Yang, S. Hirschorn, M. Jensen, T. Al (Canada)
IODIDE DIFFUSION THROUGH COMPACTED BENTONITE B75: CONSISTENT
EVALUATION TAKING INTO ACCOUNT DIFFUSE LAYER INHOMOGENEITY
E. Hofmanová, P. Večerník, D. Vopálka (Czech Republic)
POTENTIAL FOR BUOYANT NON-AQUEOUS PHASE LIQUID (NAPL) TO MIGRATE
IN THE FREE PHASE FROM A GDF
S. Watson, S. Benbow, N. Chittenden, A. Lansdell, M.O. Rivett, G. Towler, A. Herbert, G.
Carpenter, S. Norris, S. Williams (UK)
THE PUZZLING RARE EVENTS OF HIGH 238 Pu CONTENT IN THE GROUND LEVEL
ATMOSPHERE
H. Wershofen, R. Kierepko, J.W. Mietelski, R. Anczkiewicz, Z. Holgye, K. Isajenko, J. Kapała,
A. Komosa (Germany, Poland, Czech Republic)
ROLE OF THE CLAY AGGREGATES IN THE CONTAINMENT PROPERTIES OF THE
CALLOVO-OXFORDIAN CLAYSTONES: A CASE STUDY WITH COMPACTED
CRUSHED SAMPLES AND MACRO-FRACTURED SAMPLES
S. Savoye , C. Imbert , A. Fayette , B. Grenut , J.-C. Robinet (France)
COMBINED STUDIES OF RADIONUCLIDE ADSORPTION AND DIFFUSION
R.A. Wogelius, A. van Veelen, B. Zou, G. Law, K. Morris, M.P. Ryan (UK)
DETERMINING SR AND CS MASS TRANSFER PARAMETERS WITHIN INTACT
CRYSTALLINE ROCKS
B. Zou, T. Ohe, G. Grime, R. Wogelius (UK, Japan)
CHARACTERISING GAS MIGRATION THROUGH VARIABLY SATURATED MEDIA:
A NUMERICAL MODEL
D. Huxtable, D. Read, G. Shaw (UK)
BENCHMARK EXPERIMENTS FOR THE INVESTIGATION OF THE DIFFUSIVE
BEHAVIOUR OF 85 S r 2+ IN COMPACTED Na-ILLITE
M.A. Glaus, L.R. Van Loon, L. Van Laer, M. Aertsens, C. Bruggeman, J. Govaerts, N. Maes
(Switzerland, Belgium)
MATRIX DIFFUSION MEASUREMENTS ON A DRILL CORE SAMPLE FROM
ONKALO, OLKILUOTO
J. Kuva, M. Voutilainen, P. Kekäläinen, J. Timonen, P. Hölttä, M. Siitari-Kauppi, K.
Hänninen, K. Helariutta, L. Koskinen (Finland)
WITHDRAWN
MATRIX DIFFUSION AND SORPTION OF Cs + , Na + , I - AND HTO IN GRANODIORITE:
LABORATORY RESULTS AND THEIR EXTRAPOLATION TO THE IN-SITU
CONDITION
Y. Tachi, T. Ebina, H. Takahashi, K. Nemoto, T. Suyama, A. Martin (Japan, Switzerland)
DIFFUSION AND SORPTION OF Cs + , I - AND HTO IN COMPACTED SODIUM
MONTMORILLONITE AS A FUNCTION OF DRY DENSITY
Y. Tachi, K. Yotsuji (Japan)
BEHAVIOUR OF SELENIUM IN GRANITIC ROCK
J. Ikonen, P Sardini, M. Voutilainen, K. Hänninen, L. Jokelainen, R. Pehrman, A. Martin, M.
Siitari-Kauppi (Finland, France, Switzerland)

PB2-15
PB2-16
PB2-17
PB2-18
PB2-19
PB2-20
SORPTION AND DIFFUSION OF Zn ONTO Na-ILLITE UNDER A WIDE VARIETY OF
CONDITIONS
L. van Laer, T. Kupcik, C. Bruggeman, N. Maes, T. Schäfer (Belgium, Germany)
INFLUENCE OF A SALINE PLUME (NaNO 3 ) ON RADIONUCLIDE MOBILITY IN THE
CALLOVO-OXFORDIAN CLAY ROCK
V. Blin, P. Arnoux, D. Hainos, J. Radwan (France)
INFLUENCE OF CLAY CONTENT ON HTO AND 36 Cl TRANSPORT PROPERTIES IN
CALLOVO-OXFORDIAN CLAY ROCK : PERCOLATION EXPERIMENTS AND
MODELLING
C. Landesman, S. Ribet , C. Bailly, J.C. Robinet, B. Grambow (France)
POROSITY, DIFFUSIVITY AND HYDRAULIC CONDUCTIVITY IN GRANITIC ROCK
MATRIX:
LABORATORY MEASUREMENTS AND NUMERICAL MODELLING
V. Havlová, J. Najser, L. Gvoždík, K. Sosna, P. Večerník, J. Záruba, P. Dobeš (Czech
Republic)
MASS TRANSPORT IN SHALE MATRIX UNDER SALINE CONDITIONS
P. Vilks, N.H. Miller, J.G. Miller, T. Yang (Canada)
THE BIGRAD CONSORTIUM - MIGRATION OF KEY RADIONUCLIDES THROUGH
HOLLINGTON SANDSTONE
O. Preedy, M. Felipe Sotelo, N.D.M. Evans (UK)
PB2-1
DIFFUSION PROPERTIES IN LOW PERMEABILITY MEDIA
Y. Xiang 1 , D. Loomer 1 , T. Yang 2 , S. Hirschorn 2 , M. Jensen 2 , T. Al 1
1 University of New Brunswick, Department of Earth Sciences, Fredericton, NB E3B 5A3, Canada
2 Nuclear Waste Management Organization, 22 St. Clair Avenue East, Sixth Floor, Toronto, ON M4T 2S3,
Canada
The Nuclear Waste Management Organization’s (NWMO) Adaptive Phased Management (APM) Technical
Program is intent on advancing the understanding of pore fluid evolution and solute migration in deep seated
sedimentary and crystalline groundwater systems. As part of this program, innovative laboratory methods are
being developed and tested to improve the basis for characterizing the in-situ effective diffusive properties of
sedimentary and crystalline rock. These methods include through-diffusion and radiography techniques in
which the diffusive migration of conservative, anionic and weakly sorbed solutes in saturated and partially
saturated media are being explored. This presentation provides an overview of advances in application of
such techniques for the characterization of low hydraulic conductivity (10 -14 m/s), low porosity (0.01-0.08)
Ordovician age carbonates and shale within the Michigan Basin.
An X-ray radiography technique has been developed to reliably observe the transient evolution of 3-
dimensional tracer porewater concentrations in centimeter scale rock core samples from which effective
diffusion coefficients are estimated by inverse means [1-3]. The X-ray radiography technique offers shorter
measurement time in comparison to the conventional through-diffusion method because diffusion
coefficients can be determined during the transient period prior to steady-state. A comparison of effective
diffusion coefficients estimated by X-ray radiography and the through-diffusion techniques for the nonsorbed
anionic tracer iodide yields consistent values for the shale and carbonate rock samples [2, 3].
The X-ray radiography technique was also used to investigate reactive transport of a non-conservative solute
(cesium) in shale from the Michigan Basin in Canada and in the Opalinus Clay in Switzerland [3]. Reactivetransport
modelling, which couples diffusion and ion-exchange, correlated well with experimental data and

was successfully used to quantify cesium effective diffusion coefficients and the cation exchange capacity.
The resulting diffusion coefficients and cation exchange capacities are in reasonable agreement with
published values measured in both laboratory and in-situ diffusion experiments.
In addition to the X-ray source, a technique employing an Am-241 gamma source is being developed to
measure diffusion coefficients for partially-saturated sedimentary rock samples and for low permeability,
low porosity (0.001) crystalline rocks.
To assess the influence of rock core sample stress relief on estimated effective diffusion coefficients,
through-diffusion measurements have been conducted with samples under replicated in-situ stress
conditions. The in-situ stress through-diffusion measurements were performed on shale and limestone core
samples obtained from about 500-700 m below ground surface. The experiments involved HTO and iodide
tracers in synthetic porewaters with ionic strength of 5 or 8 mol/kg, under confining pressures of 12-17 MPa.
The effective diffusion coefficients determined under representative in-situ confining pressures are 20 to
40% lower than the diffusion coefficients determined at ambient laboratory pressure.
[1] Cavé, L.C., T. Al, Y. Xiang and P. Vilks. 2009. A technique for estimating one-dimensional diffusion coefficients in
low-permeability sedimentary rock using X-ray radiography: Comparison with through-diffusion measurements. J.
Cont. Hydrol. 103, 1.
[2] Cavé, L.C., T.A. Al and Y. Xiang. 2009. X-ray radiography techniques for measuring diffusive properties of
sedimentary rocks. Nuclear Waste Management Organization Report NWMO TR-2009-03. Toronto, Canada.
[3] Cavé, L.C., T. Al, Y. Xiang and D. Loomer. 2010. Investigations of diffusive transport processes in sedimentary
rock. Nuclear Waste Management Organization Report NWMO TR-2010-04. Toronto, Canada.
PB2-2
IODIDE DIFFUSION THROUGH COMPACTED BENTONITE B75: CONSISTENT
EVALUATION TAKING INTO ACCOUNT DIFFUSE LAYER INHOMOGENEITY
E. Hofmanová (1,2) , P. Večerník (2) , D. Vopálka (1) ,
(1) Department of Nuclear Chemistry, Czech Technical University, CZ-11519 Prague, Czech Republic
(2) ÚJV Řež, a.s., CZ-25068, Husinec - Řež, Czech Republic
Compacted bentonite B75 represents proposed buffer/backfill material in the Czech concept of high level
radioactive waste repository (HLWR). Iodide, as the most stable iodine species under in situ conditions of
the deep repository, plays a crucial role in the performance assessment due to its negligible interaction with
materials of engineered or natural barriers. Nevertheless, iodide transport in these barriers can be reduced by
anion exclusion effect. Therefore it is essential to understand iodide diffusive behaviour with the aim to
obtain precise diffusion coefficients necessary for prediction of radioactive iodine migration.
This study presents a series of iodide through-diffusion experiments in compacted bentonite B75 to dry
densities from 1300 kg/m 3 to 1900 kg/m 3 (see Fig. 1A). Several authors performed out-diffusion after
through-diffusion experiments in order to verify obtained diffusion coefficient [e.g. 1,2]. However, diffusion
coefficients obtained from the out-diffusion step were higher than that from the initial through-diffusion step.
This difference is likely caused by increased porosities at the clay boundaries [3]. In our work, we propose
that tracer concentration profiles could be more beneficial for the verification because they bring not only
concentration profile in the layer but also porosity profiles revealing increased porosity at the clay boundary
(see Fig. 1B).
Apparent and effective diffusion coefficients were evaluated from both data sets (break-through curves and
concentration profiles), using own computer module EVALDIFF. The module can respect real concentration
changes in both reservoirs during the experiment, radioactive decay, and influence of separating filters [4]. It
also enables to model diffusion in the layer with heterogeneous porosity. The experimental data were fitted
with regard to the increased water content at the bentonite boundary zones (multiple porosity model). Results
were compared with values obtained by homogenous (single) porosity modelling. E.g. obtained apparent
diffusion coefficient D a (single) = (1,42 ± 0,11)·10 -10 m 2 /s for dry density of 1600 kg/m 3 is in a good

agreement with value of D a = (1,85 ± 0,06) ·10 -10 m 2 /s determined on similar type of bentonite for the same
compaction [1].
The difference between single and multiple porosity approaches was not significant at the confidence level of
95 %. However, neglecting the effect of porosity heterogeneity, results in a systematic error in the
determination of the diffusion coefficients, these being very important parameters for predicting long term
safety of HLWR.
Figure1. Iodide break-through curves (A) and total porosity profiles (B) for different bentonite dry densities
(■ 1300 kg/m 3 , ○ 1450 kg/m 3 , ▲ 1600 kg/m 3 , □ 1780 kg/m 3 , ● 1900 kg/m 3 ).
[1] T. Wu, J. Li, W. Dai, G. Xiao, F. Shu, J. Yao, Y. Su, L. Shi, Sci. China. Chem. 55, 1760 (2012).
[2] L. R. Van Loon, M. A. Glaus, W. Müller, Appl. Geochem. 22, 2536 (2007).
[3] M. A Glaus, S. Frick, R. Rossé, L. R. Van Loon, J. Contam. Hydrol. 123, 1 (2011).
[4] D. Vopálka, H. Filipská, A. Vokál, Mat. Res. Soc.–Symp. Proc. 932, 983 (2006).
PB2-3
POTENTIAL FOR BUOYANT NON-AQUEOUS PHASE LIQUID (NAPL) TO MIGRATE IN THE
FREE PHASE FROM A GDF
S. Watson 1 , S. Benbow 1 , N. Chittenden 1 , A. Lansdell 2 , M O Rivett 3 , G Towler 1 , A. Herbert 3 , G. Carpenter 2 ,
S. Norris 4* , S. Williams 4
1 Quintessa Limited, Henley-on-Thames, Oxfordshire, UK.
2 AMEC Ltd, Gloucester, UK
3
School of Geography, Earth & Environmental Sciences, University of Birmingham, Birmingham, UK
4 NDA Harwell Office, Building 587, Curie Avenue, Harwell Oxford, Didcot, UK
The inventory of intermediate-level wastes (ILW) destined for disposal in a geological disposal facility
(GDF) will contain some non-aqueous phase liquids (NAPLs) that will have been immobilised prior to
disposal. These are organic liquids, such as oils and solvents, that have limited miscibility with water and
will thus form a separate phase if present in sufficient amounts. NAPLs may also be released from some of
the organic waste materials (for example plastics) as they degrade. NAPLs may have an affinity for some
radionuclides, which could then partition preferentially into the NAPL phase with potential implications for
safety. Furthermore, in sufficient quantities light NAPLs (LNAPLs), with densities less than that water,
could potentially agglomerate and form a discrete phase within a GDF.
The possible impact of NAPLs has, in the past, been identified as a potential viability challenging issue for a
GDF. It is therefore important to demonstrate understanding of the influence these NAPLs might have on the
overall evolution of the performance of a disposal system, and in particular how they might influence
radionuclide transport.
For NAPLs to be an issue and for there to be a potential NAPL pathway from a GDF to the surface:
• There must be a significant inventory of NAPLs and/or precursor in the waste; and

• NAPLs must be able to escape from the waste package in sufficient amount; and
• NAPLs must be able to migrate through backfill in sufficient amount; and
• NAPLs must be able to migrate through the geosphere in sufficient amount.
To progress understanding of the significance of NAPLs in the context of a safety case for a GDF, a study
has been undertaken and has recently reported [1]. This drew together information from a wide range of
sources, to develop safety arguments concerning the potential for a discrete NAPL phase to be present in
each of the barriers that comprise the multi-barrier system.
The developed safety arguments support the case that it is unlikely that significant quantities of NAPL would
be able to escape from the waste packages, and even less likely that NAPL would be sufficiently persistent in
the disposal environment that it could accumulate at the highest point of a vault and be transported into the
geosphere. These conclusions are valid for all of RWMD’s generic environments, and while some significant
uncertainties remain the approach taken has provided multiple lines of evidence to support the proposition
that the release of a discrete NAPL phase originating from the ILW will not present a significant challenge to
the safety case.
The presence of NAPLs is unlikely to have a significant impact on the expected flows of groundwater and
gas in a disposal system. In terms of the approach to developing safety arguments, the safety case is built
around the evidence that supports the statements ‘NAPLs not significant’ (i.e. the expected inventory is
below a threshold value defined to ensure that the impact on disposal system performance is negligible) and
‘Impact confined to the waste package’.
The key arguments revolve around the likely inventory of NAPL within a waste package and whether or not
sufficient NAPL will accumulate in the source zone (or at other interfaces between materials if NAPL
escapes from the waste package) to form a mobile mass. While previous pessimistic estimates of the amount
of NAPL that might be generated within a waste package would likely result in NAPL saturations that permit
migration, more realistic estimates based on experiments conducted on typical NAPL precursor materials
suggest that the amount of NAPL that would persist in the source region for long enough to potentially
migrate out of the waste package is too small to exceed the NAPL residual saturation. In this case, NAPL
would not be released as a distinct phase.
Even for the pessimistic inventory, the available evidence and understanding suggest that the vast majority of
the NAPL generated in the waste packages will dissolve in the groundwater and then potentially degrade to
soluble degradation products. Any migration in the free phase would be slow owing to the properties of the
NAPLs expected to be generated from ILW, which are viscous and not especially buoyant.
Although there are significant uncertainties, these conclusions appear to be robust to a range of potentially
realistic scenarios.
This work was funded by NDA RWMD.
[1] Potential for Buoyant Non-Aqueous Phase Liquid to Migrate in the Free Phase from a GDF, S. Watson, S. Benbow,
N. Chittenden, A. Lansdell, M O Rivett and G Towler, Quintessa / AMEC / University of Birmingham report to NDA
RWMD, 2012.

PB2-4
THE PUZZLING RARE EVENTS OF HIGH 238 Pu CONTENT IN THE GROUND LEVEL
ATMOSPHERE
Herbert Wershofen 1 , Renata Kierepko 2 , Jerzy W. Mietelski 2 , Robert Anczkiewicz 3 , Zoltan Holgye 4 ,
Krzysztof Isajenko 5 , Jacek Kapała 6 , Andrzej Komosa 7
1 Physikalisch-Technische Bundesanstalt, AG 6.12, Bundesallee 100, 38116 Braunschweig, Germany
2 The Henryk Niewodniczański Institute of Nuclear Physics Polish Academy of Sciences, Radzikowskiego 152,
Kraków, Poland,
3 Institute of Geological Sciences, Polish Academy of Sciences, Senacka 1, Kraków, Poland,
4 National Radiation Protection Institute, Bartoskova 28, Prague, Czech Republic,
5 Central Laboratory for Radiation Protection, Konwaliowa 7, Warsaw, Poland,
6 Białystok Medical University, Mickiewicza 2A, Białystok, Poland,
7 Department of Chemistry, The Maria Curie-Skłodowska University, Plac Marii Curie-Skłodowskiej 2,
Lublin, Poland.
Long term monitoring of radionuclides in the ground level air started during the nuclear bomb test era. In
some countries the national monitoring was just using gross alpha/beta counting in some other more
sophisticated spectrometric methods were applied. In some countries collected air filters were preserved and
the further development of the measurements techniques allowed retrospective analyses of plutonium content
in ground-level air. The available record of Pu in the air starts in some cases (like PTB, Germany, for
instance) from sixties XX century, in other cases (like in IFJ PAN, Poland) it started from 1990.
Following decreasing of Pu content in air the aerosols were collected from bigger and bigger volume using
high volume aerosols samplers. With modern sampler, efficient, selective radiochemical procedure and low
background alpha spectrometers during one week measurements the MDC can go down to even fraction of
nano-Bq/m 3 . Our laboratories stated collaboration in this field in late nineties of XX C. The question which
arise from very beginning was surprising enhanced medium value for 238 Pu / 239+240 Pu activity ratio noticed in
PTB record first time.
The main goal of our cooperation was to study the correlations between observed variation of isotopic ratio
in the long base, continent scale. Pu in air is considered to be present mainly due to the re-suspension of Pu
attached to tiny soil particles. At least in Europe, the majority of plutonium present in soils is just that of
global fallout origin (i.e. 238 Pu/ 239+240 Pu, activity ratio, R a = 0.03, 240 Pu/ 239 Pu, mass ratio, R m = 0.18). There are
some relatively small areas in Europe with soils of lower ratios (remains of the Palomares accident) or higher
ratios (for example the Chernobyl area). Some Pu in ground-level air can also origin from sea-spray
containing particles with both the Pu ratios higher than the global fallout ratios mostly due to Sellafield
discharges). The study on the isotopic composition made by alpha spectrometry or mass spectrometry reveal
some unusual activity ratios, which cannot be explained just by re-suspension. The most striking are rare
events of very unusual, high ratios of 238 Pu/ 239+240 Pu which happened sometime. The identification of
possible origin is very difficult, since to achieve low detection limit the air sampling last for a month at least
and during this time the averaging the direction of incoming air masses occur.
The paper will summarize measurement results, including data on such rare events observed in of long-term
records of plutonium monitoring in ground-level air in Middle Europe.

PB2-5
ROLE OF THE CLAY AGGREGATES IN THE CONTAINMENT PROPERTIES OF THE
CALLOVO-OXFORDIAN CLAYSTONES: A CASE STUDY WITH COMPACTED CRUSHED
SAMPLES AND MACRO-FRACTURED SAMPLES
S. Savoye (1) , C. Imbert (2) , A. Fayette (1) , B. Grenut (1) , J.-C. Robinet (3)
(1) CEA, DEN, DPC, Laboratory of Radionuclides Migration Measurements and Modeling, F-91191 Gifsur-Yvette,
France. (sebastien.savoye@cea.fr)
(2) CEA, DEN, DPC, Laboratory of Concrete and Clay Behavior Studies, F-91191 Gif-sur-Yvette,
France. (christophe.imbert@cea.fr)
(3) Andra, 1 rue Jean Monnet, F-92298 Châtenay-Malabry, France. (jean-charles.robinet@andra.fr)
The safety of a radioactive waste disposal seated in an argillaceous formation, such as the Callovo-Oxfordian
formation (France), is based, among others, on the specific properties of the host-rock, especially due to the
presence of clay mineral phases. In addition to their ability for retaining the radionuclides onto their surfaces,
clay phases allow argillaceous rocks to swell. This permits their use as potential backfill material in the
repository but this would also be responsible for the self-sealing of the fractures created during the
excavation, and that could be initially conductive [1]. Whereas permeability measurements in clayrock
samples either under crushed and weakly-compacted form or strongly cracked have been described in the
literature [1, 2, 3], no study dealing with the diffusive transport in such materials is, as far as we know,
available.
In the present work, in addition to hydraulic characterization, the diffusion of two water tracers (HTO or
HDO), and when possible, of an anion ( 36 Cl - ) and a cation ( 22 Na + ) was studied through clayey samples
originating from the Callovo-Oxfordian formation. For the backfill-type samples, two grain size distributions
(0-250 µm and 0-1000 µm) and two values of dry density (1.4 and 1.7 g.cm -3 ) were investigated. For the
fractured samples, submillimeter planar fractures were obtained by breaking the rock disk in two halves that
were subsequently tied together by their edges with glue.
During the hydration of the fractured samples, the plexiglass®-made cells allowed us to visualize the fast
closure of the fractures (< 24 hours), associated to a sharp drop of the hydraulic conductivity values (from
10 -6 m/s to 10 -10 -10 -11 m/s in less than 100 hours). Moreover, the backfill-type samples showed an evolution
of their hydraulic conductivity (K H ) in agreement with their expected pore space geometry: from a K H equal
to 3 × 10 -10 m/s for the highest dry density and the finer grain size distribution, to a K H equal to 80 × 10 -10 m/s
for the lowest dry density and the coarser grain size distribution.
The diffusion experiments lead to values of effective diffusion coefficient (D e ) and porosity for the backfilltype
samples that follow the empirical Archie’s law with a cementation factor characteristic of granular
materials, while those obtained on the macro-fractured samples are slightly higher than those determined on
mechanically-undamaged samples (D e
fractured
= ~ 2 x D e undamaged ). Moreover, anionic exclusion and enhanced
diffusion for cation, observed in sound samples [4], still occur in the compacted crushed samples. Theses
samples were finally 3D-imaged under water-saturated state within their diffusion cells, using X-ray
computed microtomography. The main structural features (macropores, clay aggregates…) were
characterized from the micro-CT images. All these results demonstrate the major role played by the clay
aggregates in the diffusion processes.
[1] Bock, H. et al., 2010. Self-sealing of fractures in argillaceous formations in the context of geological disposal of
radioactive waste: Review and Synthesis. NEA report n° 6184.
[2] Tang, C.S. et al. 2011. A study of the hydro-mechanical behaviour of compacted crushed argillite. Eng. Geol. 118,
93-103.
[3] Zhang, C.L. 2011. Experimental evidence for self-sealing of fractures in claystone. Phys. Chem. Earth 36, 1972-
1980.
[4] Savoye, S. et al., 2011. Effect of temperature on the containment properties of argillaceous rocks: the case study of
Callovo–Oxfordian claystones. J. Contam. Hydrol. 125, 102–112.

PB2-6
COMBINED STUDIES OF RADIONUCLIDE ADSORPTION AND DIFFUSION
R.A. Wogelius 1 *, A. van Veelen 1 , B. Zou 1 , G. Law 2, K. Morris 3 , M.P. Ryan 4
1 University of Manchester, Williamson Centre for Molecular Environmental Sciences, School of Earth,
Atmospheric and Environmental Sciences, Oxford Road, Manchester, M13 9PL, United Kingdom
(*correspondence: roy.wogelius@manchester.ac.uk)
2 University of Manchester, Centre for Radiochemistry Research, School of Chemistry, Oxford Road,
Manchester, M13 9PL, UK
3 University of Manchester, Research Centre for Radwaste and Decommissioning, School of Earth,
Atmospheric and Environmental Sciences, Oxford Road, Manchester, M13 9PL, UK
4 Imperial College London, Department of Materials, London SW7 2AZ, UK
Because of the slow rates of diffusion in low permeability rock matrices and the generally low
concentrations of radionuclides involved in disposal scenarios, experiments to resolve the physical and
chemical parameters needed for use in risk models is challenging. Typically, radionuclide mobility in a
geological disposal facility depends on both the physical properties of the host rock and on reactions at the
mineral-fluid interface. Thus, in order to model systems over time periods pertinent to GDF lifetime
performance (10 5 to 10 6 years), quantitative measurements of surface uptake and diffusive flux are needed.
Work presented here will give an overview of spectroscopic measurements designed to provide chemical
details of the process of surface attachment of key radionuclides (U, Tc, Sr, Cs) to a number of mineral
surfaces (magnesite, brucite, portlandite, magnetite, phyllosilicate, K-feldspar, plagioclase feldspar) reacted
under a range of repository conditions (pH 6-13, reducing and oxidizing conditions, low and high PCO 2 ).
Unambiguous information regarding the surface chemistry of adsorbates is provided via x-ray absorption
spectroscopy, determined in both standard and glancing incidence modes.
Furthermore, x-ray surface scattering and surface diffraction have been used to follow the evolution of
surface morphology and also constrain thin film thickness, changes in surface electron density, and to
determine the crystal structure of ultra-fine surface precipitates. Standard adsorption experiments have also
been completed in order to measure contaminant uptake as a function of solution concentration. The detailed
spectroscopy and partition coefficient information has been used in designing and executing related diffusive
flux experiments. Rock matrices with reactivities controlled by minerals used in the adsorption studies have
been used in flow-through (micro-reactor) and in situ (Multiple internal reflection-FTIR) experiments in
order to make rapid measurements of diffusion coefficients in systems where surface adsorption can be
robustly constrained. For comparison to values deduced from breakthrough curves, surface measurements of
contaminant uptake and diffusion length have been completed using proton induced x-ray emission (PIXE)
coupled with Rutherford Back-Scattering (RBS).
PB2-7
Determining Sr and Cs Mass Transfer Parameters within Intact Crystalline Rocks
B. Zou 1) , T. Ohe 2) , G. Grime 3) , R. Wogelius 1)
1) Williamson Research Centre for Molecular Environmental Science and School of Earth, Atmospheric
and Environmental Sciences, University of Manchester, Oxford Road, Manchester M13 9PL, UK
2) Energy, Engineering and Science Department, Tokai University, 1117 Kita Kaname, Hiratsuka-shi,
Kanagawa, 259-1292, Japan
3) University of Surrey Ion Beam Centre, Guildford GU2 7XH, UK
The engineered and surrounding geological barriers are intended to preclude the development of meso-scale
fracture pathways in a geological disposal facility, therefore diffusive flux should be considered as an
alternative risk for the loss of containment, especially for the extended period of time over which
sequestration is required. “Crystalline rock” (low permeability) with quartz and feldspar as prominent
minerals is a typical choice for a repository host and is a favoured medium in Sweden, Britain, and Japan.

Many factors related to the matrix porosity influence diffusion such as connected pore volumes, tortuosity,
and constrictivity, all of which make efforts to study diffusion difficult. Therefore, a novel method has been
developed, referred to as the Micro-Reactor Simulated-Channel (MRSC) which is able to determine effective
diffusion coefficients on non-crushed rock samples rapidly with minimum contamination (Okuyama et al,
2008; Ohe et al., 2012). The concept is similar to the micro-chemical reactor, with a very thin fluid channel
in the middle of the solid system increasing the rate of surface reactions due to a high surface area to liquid
volume ratio. This enables fast measurements using only small reactor volumes. The MRSC consists of two
injection syringe pumps, a reaction unit, an auto sampler, and a drainage tank: the experimental set up is
shown in Figure 1 (Okuyama et al 2007). Unlike the conventional column method, an intact hard rock
sample can be used for the determining the effective diffusion coefficient and sorption coefficients
simultaneously and also creates product surfaces that are amenable for direct chemical analysis.
Figure 1. Set up of micro-reactor simulated channel method (MRSC) (Okuyama et al 2008)
In the MRSC experiments, the aqueous fluids containing dissolved Sr and Cs have been pumped through a
simulated rock fracture at a constant rate. 90 Sr and 137 Cs isotope are both radioisotopes of great concern with
half-lives of approximately 30 years. For convenience in these experiments we have used non-radioactive Sr
and Cs. The outlet solute concentration was monitored as a function of time; and the effluent was collected
for determining the breakthrough curves. Breakthrough curves were constructed by time resolved analysis of
the effluent solution until steady state was observed. Because the outlet concentration is not equal to inlet
concentration, Sr and Cs are lost through two dominant processes; 1) diffusion into grain boundaries,
micropores, and microfractures, and 2) sorption from the simulated fracture onto mineral surfaces.
Concentrations remaining behind are much higher than would be predicted from simple surface adsorption
onto the exposed plates, and hence it is concluded that appreciable Sr and Cs has diffused into the grain
boundaries.
In order to further constrain the tracer inventory between fine particulates and mineral surfaces, an analytical
regimen was developed using ion beam techniques. This allowed us to fully characterise the reacted surfaces
and carry out depth-profiling. In particular, we designed these measurements to take advantage of the
extremely low background radiation levels at the Sr and Cs Kα energies in particle induced X-ray emission
(PIXE) analysis in order to accurately quantify Sr and Cs taken up by mineral surfaces in the MRSC
experiments (Wogelius et al., 1992). In addition, the depth profiling capability of Rutherford backscattering
spectrometry (RBS) was used to non-destructively constrain diffusion profiles below the surface with

concentration levels down to a few parts per million from specimen areas of 1 µm in diameter. The mass
balance calculations as the result of the ion beam analyses are able to fully quantify the amount of Sr and Cs
present at the rock surface which provides a critical constraint in terms of the mass transfer parameters
produced from MRSC solution analysis.
Ohe T., Zou B., Noshita K., Gomez-Morilla I., Jeynes C., Morris P.M., and Wogelius R.A. (2012) Adsorption and
Diffusion of Sr in Simulated Rock Fractures Quantified via Ion Beam Analysis. Mineralogical Magazine 76, 3203-
3215.
Okuyama, K., Sasahira, A, Noshita, K., and Ohe, T (2008). A method for determining both diffusion
and sorption coefficients of rock medium within a few days by adopting a micro-reactor
technique. Applied Geochemistry 23, 2130-2136.
Wogelius, R.A., Fraser, D.G., Feltham, D. and Whiteman, M. (1992) Trace elements in dolomite:
proton microprobe data and constraints on fluid compositions. Geochimica et Cosmochimica Acta,
56, 319-334.
PB2-8
Characterising gas migration through variably saturated media: a numerical model
Darren Huxtable *1 , David Read 2 , George Shaw 1
1 School of Bioscience, University of Nottingham, UK 2 The Department of Chemistry, University of
Loughborough, UK.
Molecular diffusion is the principal mechanism that determines gas migration in most soil profiles.
Numerous models have been proposed to ‘predict’ effective gas diffusion coefficients in soils but the
accuracy of these models needs to be tested and improved through experimental investigation of the
variables that influence gas diffusion at various scales. The fate and transport of labeled methane and carbon
dioxide are of particular interest for the energy sector – shale gas and methane are forecast to account for
63% of North American production by 2030. Tracking the movement of these gases is made complex by
biotic and abiotic reactions in the soil. Large errors may be introduced by taking surface flux measurements
of these gases directly. There are also problems associated with using 14 C directly as an in-situ tracer.
This study aims to bring new insight into gas transport modeling by building on studies that demonstrate how
depletion of radon in soil gas may be used as an indicator to map subsurface plumes of non-aqueous phase
liquid (NAPL) contamination. The ubiquitous generation of radon ( 222 Rn) in soil from its parent nuclide,
226 Ra, and the fact that radon is markedly more soluble in organic solvents than in water, a) allows its use as
an inert tracer of gas movement through the soil profile and b) provides a means by which the distribution of
NAPL in the subsurface can be determined.
A two-dimensional model has been developed to support and optimise experimental design and
interpretation. The model employs an implicit finite difference scheme to solve Fick’s second law and is
being paramaterised and validated using data collected from parallel laboratory soil/glass bead column
studies as well as in-situ field studies on a 15 x 15 m site at Sutton Bonington (undisturbed for over 10
years). The fundamental variables that determine the production and fate of radon at the site are being
examined through a programme of experiments. These include detailed analyses of radon mass exhalation of
the soils as a function of depth; solubility experiments, to determine the partitioning of radon in a NAPL (in
this case vegetable oil) as a function of temperature and a 2D microcosm experiment to examine the
behaviour of radon in the presence of a NAPL in an homogenous porous medium (glass beads). Detailed
baseline measurements are being conducted at the site on a monthly basis and soil variables are being
monitored as a time series using an array of 30 soil gas probes to assess the influence of known volumes of
vegetable oil buried at depth (≥0.7 m).
The inputs to the model are variables taken from experimental data: soil porosity, organic matter content,
moisture, generation of 222 Rn from 226 Ra and the solubility of 222 Rn in both vegetable oil and water.
Boundary conditions applied to the model are in line with those applied experimentally: atmospheric
concentration of 222 Rn as the upper boundary and concentration at relaxation depth as the lower. The model
will be developed further to integrate the effects of advection, or ‘atmospheric pumping’, an effect observed

in the field that may result in gas transport orders of magnitude greater than molecular diffusion. A resulting
four-phase gas migration model will then be developed and verified using data obtained in the laboratory and
in the field, with comparison made to similar studies investigating effective diffusion coefficients of soils.
Radon calibrated flux measurements of biologically active gases will enable more accurate estimations of
their rates of movement through soils. By interpreting the behaviour of radon and its interaction with an
organic solvent under variably saturated conditions, it should be possible to determine qualitatively and
quantitatively, the influence of a non-aqueous phase liquid in the subsurface.
PB2-9
BENCHMARK EXPERIMENTS FOR THE INVESTIGATION OF THE DIFFUSIVE BEHAVIOUR
OF 85 Sr 2+ IN COMPACTED SODIUM-ILLITE
M.A. Glaus, L.R. Van Loon
Laboratory for Waste Management, Paul Scherrer Institut, 5232 Villigen PSI, Switzerland.
L. Van Laer, M. Aertsens, C. Bruggeman, J. Govaerts, N. Maes
Belgian Nuclear Research Centre (SCK•CEN), Expert Group Waste & Disposal, 2400 Mol, Belgium.
The question concerning the driving force for diffusion in compacted clay minerals and clay rocks is crucial
for a process-based understanding of the transport behaviour of radionuclides in these media. In the case of
montmorillonite clear evidence has been found that the diffusion of tracer cations is driven by their gradients
in the total concentration in the clay (e.g. ref [1]). This phenomenon is mostly called 'surface diffusion' and
has been attributed to the mobility of these cations in the interlayer pore space providing a separate pathway
for diffusion. The situation in illite is less clear because of the presence of immobilised and merely nonexchangeable
potassium cations in the interlayer pore space. However tracer diffusion experiments with
22 Na + in a Na-exchanged illite indicated that surface diffusion is also effective under the conditions of high
degrees of compaction [2]. In the frame of the Catclay EU project, the question of the importance of surface
diffusion in compacted illite and clay rocks is being investigated in detail for different types of cations.
In the course of this project the diffusive behaviour of Sr 2+ tracer ions in compacted Na-exchanged illite is
being investigated. If surface diffusion is active, large tracer fluxes are to be expected representing a
challenge not only in experimental respect but also from the point of modelling. The possibility of drawing
conclusions biased by systematic errors from such experiments should not be underestimated, although the
procedures involved look rather straightforward. Issues with confining filters and inappropriate boundary
conditions are just two examples that can be mentioned. In view of the expected difficulties benchmark
experiments for measuring the diffusion of 85 Sr 2+ in compacted illite under a variety of solution compositions
were carried out at SCK•CEN and at the Laboratory for Waste Management at PSI. The goal of this
benchmark study was to compare different experimental techniques and model approaches. A representative
example is shown in Figure 1 in which the results of through diffusion experiments in compacted illite with
85 Sr 2+ at 0.1 M external background electrolyte (NaClO 4 ) concentration are compared. The setup for these
experiments varied in a couple of respects, among which the most important was probably the functional
property of the confining filters. While the filter solutions represented a diffusive resistance in the SCK•CEN
setup (left hand figure) owing to the use of stagnant pore solutions, these were advectively flushed in the PSI
setup (right hand figure) which leads to a homogenisation of the solutions in the filter pores. The flat tracer
profile measured after through diffusion in the left hand figure clearly indicates that diffusion is dominated
by the diffusive resistance of the filters. Further differences between the data shown in Figure 1 can be
explained by the different geometric bounding properties of the reservoir solutions and the clay plugs leading
to different capacities of these domains for the 85 Sr 2+ tracer. The evaluation of best-fit parameters from these
experiments is still on-going. The present contribution will give an overview of the most important results
and conclusions drawn during the course of this exercise.

Figure1. Results of through-diffusion experiments in illite with 85 Sr 2+ at 0.1 M background electrolyte
concentration. Shown are the tracer concentrations in the source reservoir, the tracer flux (J) at the
downstream boundary and the tracer profiles (C total ) measured after through-diffusion (insert plots). The left
hand plot gives results obtained with diffusion cells using stagnant filters (SCK-CEN) and the right hand plot
results from a diffusion cell with advectively flushed filters (PSI).
1. M. A. Glaus et al., Diffusion of 22 Na and 85 Sr in montmorillonite: Evidence of interlayer diffusion being the
dominant pathway at high compaction. Environ. Sci. Technol. 41, 478 (2007).
2. M. A. Glaus, S. Frick, R. Rossé, L. R. Van Loon, Comparative study of tracer diffusion of HTO, 22 Na + and 36 Cl – in
compacted kaolinite, illite and montmorillonite. Geochim. Cosmochim. Acta 74, 1999 (2010).
Acknowledgment: We thank Nagra, PSI and NIRAS for the partial financial support. The research leading to
these results has further received funding from the European Atomic Energy Community's Seventh
Framework Programme (FP7/2007-2011) under grant agreement n° 249624.
PB2-10
MATRIX DIFFUSION MEASUREMENTS ON A DRILL CORE SAMPLE FROM ONKALO,
OLKILUOTO
J. Kuva (1) , M. Voutilainen (2) , P. Kekäläinen (1) , J. Timonen (1) , P. Hölttä (2) , M. Siitari-Kauppi (2) , K. Hänninen (2) ,
K. Helariutta (2) and L. Koskinen (3)
(1) University of Jyväskylä, Department of Physics, P.O. Box 35, 40014 University of Jyväskylä, Finland
(2) University of Helsinki, Department of Chemistry, P.O. Box 55, 00014 University of Helsinki, Finland
(3) Posiva Oy, Olkiluoto, 27160 Eurajoki, Finland
The nuclear waste produced by the nuclear power plants of TVO (Teollisuuden Voima Oy) and Fortum will
be deposited in an underground repository. When analyzing the safety of such repository, diffusion of solutes
into rock matrix is an important factor. It is vital to know how far and fast nuclear waste would spread if a
container broke in the repository. In order to make such predictions, information about the diffusive and
other transport properties of the surrounding bedrock is needed. An important process retarding the migration
of nuclear elements is believed to be matrix diffusion. Advective transport dominates the migration of
radionuclides in water conducting fractures, but molecular diffusion allows them to migrate from the flow
into the surrounding rock matrix. For some elements sorption is also important here, and serves to retard the
radionuclides even more. The tunnel leading to the underground repository facilities is now in use as a test
facility called ONKALO, where these properties are being measured in situ. In order to determine parameters
for the in-situ measurements, and to validate transport models under well controlled conditions, parallel
laboratory experiments were conducted.
For these laboratory experiments we constructed two similar experimental setups, one for gas phase and one
for water phase. In both experiments a steady carrier flow was induced in a thin flow channel around a

cylindrical sample, and a short pulse of tracer was injected into the flow. The breakthrough curve of the
tracer was then measured and fitted by a mathematical model for an infinitely deep matrix, which takes into
account longitudinal diffusion and Taylor dispersion in the flow channel [1]. This setup which was similar to
that of the in-situ experiments, was planned to depict a scenario where radionuclides are being transported in
a flow path in the bedrock.
The gas phase method was developed so as to study diffusion without the long measuring times typical of
water phase measurements [2]. In the gas phase, nitrogen was used as the carrier and helium as the tracer.
The breakthrough curve was measured with a helium mass spectrometer. A schematic representation of the
setup is shown in Fig. 1.
Figure 1. Schematic representation of the measurement setup.
With this measurement we were able to demonstrate the retardation effect caused by matrix diffusion and to
determine a diffusion coefficient of 3.0-3.4 ⋅10 -8 m 2 /s for a sample from the REPRO-niche at ONKALO.
Data for one measurement along with a mathematical fit is shown in Fig. 2. We also measured the
longitudinal diffusion coefficient in the flow channel to be 6.4-7.7 ∙10 -5 m 2 /s, which is consistent with the
diffusion coefficient of helium in nitrogen, 6.98 ∙ 10 -5 m 2 /s.
Figure 2. The breakthrough curve of a sharp pulse of helium in the experimental setup shown in Fig. 1. Influence
of advective transport dominates the early part of the curve, while the effect of matrix diffusion is seen in its late
time behaviour as a strong tailing of the curve.
Water phase measurements were first tested with iodine as the tracer, and then conducted with sodium,
chloride and tritium as tracers. The flow speed used was 20 µl/min in a 1.25 mm wide flow channel. The
retarding effect of matrix diffusion was observed in the breakthrough curves. The diffusion coefficients are
yet to be determined. First results will be presented.
[1] Kekäläinen, P., M. Voutilainen, A. Poteri, P. Hölttä, A. Hautojärvi and J. Timonen (2011) “Solutions to and
validation of matrix-diffusion models.” Transport Porous Med. 87: 125-149
[2] Väätäinen, K., J. Timonen and A. Hautojärvi (1993) “Development of a gas method for migration studies in
fractured and porous media.” Mat. Res. Soc. Symp. Proc. 294: 851-856

PB2-11
WITHDRAWN
PB2-12
MATRIX DIFFUSION AND SORPTION OF Cs + , Na + , I – and HTO IN GRANODIORITE :
LABORATORY RESULTS AND THEIR EXTRAPOLATION TO THE IN-SITU CONDITION
Y. Tachi (1) , T. Ebina (2) , H. Takahashi (1) , K. Nemoto (1) , T. Suyama (1) , A. Martin (3)
(1) Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki, 319-1194, Japan
(2)
NESI Inc., Tokai, Ibaraki, 319-1112, Japan
(3)
Nagra, Hardstrasse 73, 5430 Wettingen, Switzerland
Matrix diffusion and sorption of radionuclides in crystalline host rocks are key processes controlling the safe
geological disposal of radioactive waste. Diffusion and sorption parameters have been derived from
laboratory experiments using drilled or crushed rock samples. These samples have different properties in
porosity, pore-connectivity and reactive surface area in comparison with real in-situ conditions. The
laboratory measurements can therefore lead to overestimation of matrix diffusion and sorption properties. To
set reliable parameters of matrix diffusion and sorption relevant to performance assessment, it is necessary to
understand in detail the processes of diffusion and sorption both in laboratory and in-situ experiments, and to
develop a method for extrapolating from the laboratory to in-situ conditions. In this work, these issues were
investigated in connection with the Long Term Diffusion (LTD) project [1] at the Grimsel Test Site (GTS),
Switzerland. Diffusion and sorption parameters for cationic Cs + and Na + , anionic I – and neutral HTO tracers
were determined by laboratory experiments using GTS rock samples and the LTD test results [2] were
interpreted based on the laboratory results and their extrapolation to the in-situ condition.
Effective diffusivities (D e ), rock capacity factors () and distribution coefficients (K d ) of 137 Cs + , 22 Na + , 125 I –
and HTO in the granodiorite samples (20mm x t5mm) from the GTS, which were sampled from drillcore
for the in-situ LTD test, were measured under synthetic groundwater condition by through-diffusion tests.
From the breakthrough curves, the derived D e values of Cs + > Na + > HTO > I – which is as expected. The
and K d values for these tracers show the same trends that would be obtained from ion exchange sorption
mechanisms. These results imply that cation excess and anion exclusion effects may be an important
mechanism in these samples, although these effects have been observed for compacted bentonites and
argillaceous rocks [3]. Depth profiles of Cs + and Na + obtained by abrasing samples after a diffusion period
showed typical dual profiles which are similar with existing observations in granitic rocks [e.g., 4, 5]. The
modelling results by separating near and far profiles indicated that near surface profiles could be reasonably
well interpreted with higher and lower D e values. The K d values were compared with batch sorption results
using crushed samples with different particle size, indicating that crushing of samples had a strong influence
on cationic sorption.
From the in-situ LTD experiment where radionuclides were allowed to diffuse into the rock matrix
continuously for 2.5 years, the tracer depletion curves in the circulating solution through a packed-off
borehole, and the tracer profiles in the rock by overcoring and analyzing were obtained for HTO, 22 Na + ,
134 Cs + [2]. These in-situ diffusion behavior were calculated by using GoldSim code and transport parameters
based on laboratory results and their extrapolations to in-situ conditions. The diffusion and sorption
parameters are extrapolated from laboratory to in-situ conditions by considering difference in porosity [6]
and assuming their linear relation to porosity. The borehole damaged zone (BDZ) with thickness of 1mm
was assumed to have high sorption capacity based on near surface profiles observed in laboratory tests.
These model calculations could the in-situ LTD results of Cs + and Na + fit reasonably well, although it was
difficult to interpret the HTO results.
These comparative experimental and modelling studies between laboratory experiments and in-situ LTD
tests gave key findings: (1) Cation excess diffusion may be a key mechanism in granitic rock, like in
compacted clays and argillaceous rock. (2) BDZ and near surface diffusion and sorption is critically
important to evaluate transport parameters both in lab and in-situ tests. (3) Difference of porosity between

lab samples and in-situ are a key factor to scale laboratory data to in-situ conditions. Further studies are
needed to understand both laboratory and in-situ diffusion behaviors in more detail.
*This study was partly funded by the Ministry of Economy, Trade and Industry of Japan. The authors greatly
appreciate the contributions from Tokyo Nuclear Service Inc. for experimental work.
[1] Martin A., Siitari-Kauppi M., Havlová V., Tachi Y. and Miksova J. (2013). “An overview of the Long-term
Diffusion Test, Grimsel Test Site, Switzerland.” MIGRATION 2013.
[2] Soler J.M., Landa J., Havlová V., Tachi Y., Ebina T., Sardini P., Siitari-Kauppi M. and Martin A. (2013). “LTD
Experiment - Post-mortem modelling of monopole 1.” Nagra Technical Report 12-53.
[3] Tachi Y., Yotsuji K., Seida Y., Yui M. (2011). “Diffusion and sorption of Cs + , I - and HTO in samples of the
argillaceous Wakkanai Formation from the Horonobe URL, Japan: Clay-based modeling approach.” Geochim.
Cosmochim. Acta 75, 6742-6759.
[4] Idemitsu K., Furuya H., Hara T., Inagaki Y. (1992) “Migration of cesium, strontium and cobalt in water-saturated
Inada granite.” Journal of Nuclear Science and Technology 29, 454-460.
[5] Byegård J., Johansson H., Skålberg M., Tullborg E.-L. (1998). “The interaction of sorbing and non-sorbing tracers
with different Äspö rock types. Sorption and diffusion experiments in the laboratory scale.” SKB Technical Report TR
98-18.
[6] Kelokaski M., Siitari-Kauppi M., Kauppi I., Hellmuth K.-H., Möri A., Biggin C., Kickmaier W., Inderbitzin L.,
Martin A. (2010). “GTS Phase VI – Pore Space Geometry Project. Characterisation of Pore Space Geometry by 14 C-
MMA Impregnation.” Nagra Technical Report 05-03.
PB2-13
DIFFUSION AND SORPTION OF Cs + , I – and HTO IN COMPACTED SODIUM
MONTMORILLONITE AS A FUNCTION OF DRY DENSITY
Y. Tachi (1) , K. Yotsuji (1)
(1) Japan Atomic Energy Agency (JAEA), Tokai, Ibaraki, 319-1194, Japan
Diffusion and sorption of radionuclides in compacted bentonite are the key processes in the safe geological
disposal of radioactive waste. To set reliable parameters under various conditions such as geochemical
parameters and buffer material designs relevant to performance assessment, it is necessary to understand
detailed processes of diffusion and sorption, and to develop mechanistic models, which describe them. One
of key challenges for these issues is the applicability of such models to complex/compacted systems by
considering the narrower porespace and electrostatic influence. For this purpose, the integrated sorption and
diffusion (ISD) model in compacted bentonite systems has been developed, and applied for various
radionuclides such as cations (Cs + , Na + , Sr 2+ ), anion (I – ), and complex Np(V) species as a function of
porewater salinity [1, 2, 3]. In this study, the diffusion and sorption behaviors of monovalent cationic Cs + ,
anionic I – and neutral HTO in compacted montmorillonite as a function of dry density (0.8–1.8 Mg m –3 ) were
investigated by experimental and modeling approaches.
Effective diffusion coefficients (D e ) and distribution coefficients (K d ) of Cs + , I – and HTO in compacted
sodium montmorillonite (Kunipia F ® ) saturated with 0.1 M NaCl solution were measured under aerobic
conditions at room temperature. Compacted samples with different dry densities, 0.8, 1.4, 1.8 Mg/m 3 were
used to investigate the effect of dry density on diffusion and sorption in the compacted montmorillonite. The
diffusion tests were conducted by the through-diffusion method with variable concentrations in the inlet and
outlet reservoirs coupled with depth profile measurement in compacted samples [1]. Then, D e and K d values
were derived by multiple curve analysis (including tracer depletion, breakthrough and depth concentration
curves), which could be fitted with a conventional diffusion model using only one set of parameters. The D e
values obtained were in the sequence of Cs + , (10 –10 m 2 /s) > HTO (10 –11 m 2 /s) > I – (10 –12 –10 –13 m 2 /s).
Although the D e values of HTO and I – decreased as dry density increased, those of Cs + were mostly
independent of dry density, as shown in Fig.1(a). The K d values obtained for Cs + were shown in Fig.1(b) as a
function of dry density and were compared with K d values obtained by batch sorption [1]. Although several
trends are reported on the effect of compaction on Cs + sorption [4, 5, 6 ; see Fig.1(b)], K d values of Cs +
obtained in this work were independent on dry density and were consistent with batch-derived values.

Sorption and diffusion behaviors were interpreted based on the latest integrated sorption and diffusion (ISD)
models [3]. The diffusion model based on a simplified pore structure and the electrical double layer (EDL)
theory, describes the change of ionic concentration and viscoelectric effects caused by the electrostatic
interaction with negatively charged clay surfaces. This model could quantitatively express the cation excess
and anion exclusion behaviors, where D e (Cs + ) > D e (HTO) > D e (I – ), and their dependence on salinity [1]. As
shown in Fig.1(a), their dependence on dry density can be also quantitatively accounted by change in
electrostatic effect as a function of density (pore aperture). The D e trend of Cs + were interpreted as coupled
effects of increase in cation excess effect and decrease in geometric factor as density increased. On the other
hand, the D e trend of I – were more pronounced by anion exclusion effect, as compared with neutral HTO.
The sorption model considering ion exchange reactions provided reasonable account of Cs + sorption trends
as functions of dry density as shown in Fig.1(b).
The ISD model could express diffusion and sorption behaviors of monovalent cation, anion and neutral
species as a function of porewater salinity and dry density. Further study are needed to test and improve the
model in order to reach better performance for various radionuclides and a wide range of conditions.
De (m 2 s -1 )
10 -8
10 -9
10 -10
10 -11
(a)
Cs +
HTO
I -
10 -12
EDL diffusion model
10 -13 0.6 0.8 1 1.2 1.4 1.6 1.8 2
Dry density (Mg m -3 )
Kd (m 3 kg -1 )
10 1 (b)
10 0
10 -1
This study
Oscarson et al. (1994)
Molera and Eriksen (2002)
Van Loon and Glaus (2008)
Iex Model
10 -2
0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2
Dry density (Mg m -3 )
Fig.1. Comparison between measured and modeled (a) D e values of Cs + , I – and HTO, (b) K d values of Cs + in
compacted montmorillonite as a function of dry density.
*This study was partly funded by the Ministry of Economy, Trade and Industry of Japan. The authors greatly
appreciate the contributions from Tokyo Nuclear Service Inc. for experimental work.
[1] Tachi Y., Yotsuji K., Seida Y. and Yui M. (2009). “Diffusion of cesium and iodine in compacted sodium
montmorillonite under different saline conditions.” In: Scientific Basis for Nuclear Waste Management XXXIII, Mater.
Res. Soc. Symp. Proc., 1193, 545–552.
[2] Tachi Y., Nakazawa T., Ochs M., Yotsuji K., Suyama T., Seida Y., Yamada N. and Yui M. (2010). “Diffusion and
sorption of neptunium(V) in compacted montmorillonite: effects of carbonate and salinity.” Radiochim. Acta 98, 711–
718.
[3] Tachi Y. and Yotsuji K. (2012). “Diffusion and sorption of Sr 2+ in compacted sodium montmorillonite as a function
of porewater salinity.” Clays in Natural and Engineered Barriers for Radioactive Waste Confinement. 5 th International
meeting. Abstract P/MT/DP/20.
[4] Oscarson D.W., Hume H.B. and King F. (1994). “Sorption and diffusion of cesium on compacted bentonite.” Clays
Clay Miner., 42, 731–736.
[5] Molera M. and Eriksen T. (2002). “Diffusion of 22 Na + , 85 Sr 2+ , 134 Cs + and 57 Co 2+ in bentonite clay compacted to
different densities: experiments and modeling.” Radiochim. Acta, 90, 753–760.
[6] Van Loon L.R. and Glaus M.A. (2008). “Mechanical compaction of smectite clays increases ion exchange
selectivity for cesium.” Environ. Sci. Technol., 42, 1600–1604.

PB2-14
BEHAVIOUR OF SELENIUM IN GRANITIC ROCK
J. Ikonen (1) , P Sardini (2) M. Voutilainen (1) , K. Hänninen (1) , L. Jokelainen (1) , R. Pehrman (1) , A. Martin (3) and M.
Siitari-Kauppi (1)
(1)
Laboratory of Radiochemistry, University of Helsinki, A.I. Virtasen aukio 1, 00014 Helsingin yliopisto,
Finland
(2) IC2MP, University of Poitiers, UMR 7285 CNRS, France
(3)
Nagra (National Cooperative for the Disposal of Radioactive Waste), Hardstrasse73, 5430 Wettingen,
Switzerland
In many countries, high-level radioactive waste is planned to be disposed of in deep-lying crystalline rock.
There has been a tendency to play down the potential role of the geosphere as a safety barrier in repository
performance assessment. Among other reasons, current uncertainties in transport pathway definition and
pore space characterisation of crystalline rock as well as the in situ diffusion properties of radioactive
elements can be mentioned. Repository safety evaluation today requires going from the laboratory and
surface-based field work to the underground repository level. Little is known about the changes to rock
transport properties during sampling and decompression.
Diffusion and sorption of radionuclides have been studied extensively in the laboratory scale. However, only
few long-term in situ experiments have been carried out to evaluate matrix diffusion and sorption properties
of radionuclides. These experiments are time consuming and cost intensive, and it is commonly accepted that
laboratory scale tests are well-established approaches for characterizing properties of geological materials.
There is variation in distribution coefficients and diffusion coefficients as well as in porosity values
depending whether the determinations are based on laboratory experiments or in situ experiments. In order to
assess the relevance of laboratory experiments, the Swiss National Cooperative for Disposal of Radioactive
Waste (Nagra) have been conducting extensive in situ experiments at the Grimsel test site (GTS) in the field
of radionuclide transport and retention. The second Long Term Diffusion (LTD) experiment has been started
in autumn 2012. In this experiment, radionuclides H-3, Na-22, Cs-134, Cl-36 and Ba-133 as well as
nonradioactive element Se, are circulated in a packed-off interval.
In the Laboratory of Radiochemistry, University of Helsinki, there have been running two laboratory scale
selenium diffusion experiments to support the above mentioned long term in situ experiment. The
experiments are conducted in two rock blocks; Kuru grey granite and Grimsel granodiorite. In the Kuru grey
granite, there is an inlet hole for tracer and the observation boreholes at different distances from it. In the
Grimsel granodiorite there are two inlet holes and observation holes are placed around them; some are in the
direction of the rock foliation and some against it. Changes of the selenium concentration in the inlet and
observation holes are followed using ICP-MS technology. Experiments are conducted in oxic conditions.
Parallel to the diffusion experiments, selenium sorption onto Grimsel granodiorite and Kuru grey granite was
studied with batch experiments and geochemical modelling. Sorption was studied as a function of grain size,
pH, Se concentration and pe. These experiments were done with Se-75 and stable selenium in groundwater
simulants.
Diffusion experiment has been running in the Kuru grey granite for three years and significant breakthrough
of selenium is observed at a distance of 1 cm. Breakthrough has also been observed to 1 cm distance in
Grimsel granodiorite. Diffusion of selenium has been found to be faster in the direction of the foliation than
against it, as can be seen in Figure 1. Diffusion and distribution coefficients of selenium in Kuru grey granite
and Grimsel granodiorite will be determined.

10000
1000
Se / ppb
100
10
A1
B1
1
0 100 200 300 400 500 600
Time / d
Figure 1. Breakthrough curve of selenium in the Grimsel granodiorite at distance of 1 cm. Observation hole
B1 is in the direction of the rock foliation and A1 against it.
PB2-15
BEHAVIOUR OF SELENIUM IN GRANITIC ROCK
J. Ikonen (1) , P Sardini (2) M. Voutilainen (1) , K. Hänninen (1) , L. Jokelainen (1) , R. Pehrman (1) , A. Martin (3) and M.
Siitari-Kauppi (1)
(1)
Laboratory of Radiochemistry, University of Helsinki, A.I. Virtasen aukio 1, 00014 Helsingin yliopisto,
Finland
(2) IC2MP, University of Poitiers, UMR 7285 CNRS, France
(3)
Nagra (National Cooperative for the Disposal of Radioactive Waste), Hardstrasse73, 5430 Wettingen,
Switzerland
In many countries, high-level radioactive waste is planned to be disposed of in deep-lying crystalline rock.
There has been a tendency to play down the potential role of the geosphere as a safety barrier in repository
performance assessment. Among other reasons, current uncertainties in transport pathway definition and
pore space characterisation of crystalline rock as well as the in situ diffusion properties of radioactive
elements can be mentioned. Repository safety evaluation today requires going from the laboratory and
surface-based field work to the underground repository level. Little is known about the changes to rock
transport properties during sampling and decompression.
Diffusion and sorption of radionuclides have been studied extensively in the laboratory scale. However, only
few long-term in situ experiments have been carried out to evaluate matrix diffusion and sorption properties
of radionuclides. These experiments are time consuming and cost intensive, and it is commonly accepted that
laboratory scale tests are well-established approaches for characterizing properties of geological materials.
There is variation in distribution coefficients and diffusion coefficients as well as in porosity values
depending whether the determinations are based on laboratory experiments or in situ experiments. In order to
assess the relevance of laboratory experiments, the Swiss National Cooperative for Disposal of Radioactive
Waste (Nagra) have been conducting extensive in situ experiments at the Grimsel test site (GTS) in the field
of radionuclide transport and retention. The second Long Term Diffusion (LTD) experiment has been started
in autumn 2012. In this experiment, radionuclides H-3, Na-22, Cs-134, Cl-36 and Ba-133 as well as
nonradioactive element Se, are circulated in a packed-off interval.
In the Laboratory of Radiochemistry, University of Helsinki, there have been running two laboratory scale
selenium diffusion experiments to support the above mentioned long term in situ experiment. The
experiments are conducted in two rock blocks; Kuru grey granite and Grimsel granodiorite. In the Kuru grey

granite, there is an inlet hole for tracer and the observation boreholes at different distances from it. In the
Grimsel granodiorite there are two inlet holes and observation holes are placed around them; some are in the
direction of the rock foliation and some against it. Changes of the selenium concentration in the inlet and
observation holes are followed using ICP-MS technology. Experiments are conducted in oxic conditions.
Parallel to the diffusion experiments, selenium sorption onto Grimsel granodiorite and Kuru grey granite was
studied with batch experiments and geochemical modelling. Sorption was studied as a function of grain size,
pH, Se concentration and pe. These experiments were done with Se-75 and stable selenium in groundwater
simulants.
Diffusion experiment has been running in the Kuru grey granite for three years and significant breakthrough
of selenium is observed at a distance of 1 cm. Breakthrough has also been observed to 1 cm distance in
Grimsel granodiorite. Diffusion of selenium has been found to be faster in the direction of the foliation than
against it, as can be seen in Figure 1. Diffusion and distribution coefficients of selenium in Kuru grey granite
and Grimsel granodiorite will be determined.
10000
1000
Se / ppb
100
10
A1
B1
1
0 100 200 300 400 500 600
Time / d
Figure 1. Breakthrough curve of selenium in the Grimsel granodiorite at distance of 1 cm. Observation hole
B1 is in the direction of the rock foliation and A1 against it.
PB2-16
INFLUENCE OF A SALINE PLUME (NANO 3 ) ON RADIONUCLIDE MOBILITY IN THE
CALLOVO-OXFORDIAN CLAY ROCK
V. Blin, P. Arnoux, D. Hainos, J. Radwan
(1) CEA, DEN, DPC, Laboratory of Radionuclides Migration Measurements and Modeling, F-91191 Gifsur-Yvette,
France. (virginie.blin@cea.fr)
The Callovo-Oxfordian clay rock formation (COx) has been proposed in France as a potential host rock for
deep radioactive waste disposal. In a general way, in order to evaluate the impact of a deep geological
disposal, it is necessary to determine the transfer properties of the constituent barriers of the disposal:
engineered barriers as well as the geological barrier (clay rock).
The clay rock has a very low permeability and diffusion is likely to be the predominant mechanism for solute
transfer. There is already a good understanding of the Callovo-Oxfordian clay rock chemistry, ions transfer
in its porosity and a large dataset on their transfer parameters. The negative charged surface of clay minerals
induces specific behaviour of the ions: while a diffusion front of anionic species is delayed by the physical
phenomenon called anionic exclusion [1], cationic species are delayed by chemical sorption on clay mineral
surfaces and their diffusion is accelerated [2].

Among the different kinds of intermediate level and long-lived waste packages, some of them, such as
reprocessing sludge, are going to release large amounts of chemical species in the system, i.e. soluble salts.
There is still a lack of knowledge concerning the behaviour and properties of these “co-contaminants” and
their potential effect on the clay rock confinement properties.
This study is thus dedicated to a better comprehension of the processes happening in the near-field of a
disposal after waste packages degradation, focusing on the effect of a particular soluble salt, sodium nitrate.
In order to quantify its effect on radionuclides transfer, several through-diffusion experiments have been
performed using:
(i) various concentrations of NaNO 3 , i.e. 0, 0.5, 1.5 and 3M,
(ii) different tracers: tritiated water as a reference tracer, 36 Cl - as an anionic tracer, 22 Na + as a weak s
orbing cationic tracer and 137 Cs + as a high sorbing cationic tracer.
The rock samples have been equilibrated with the different concentrations of NaNO 3 before spiking the
diffusion cells with the radioactive tracers.
Analyses of the results based on a Fick’s second law for one-dimensional reactive transport have shown the
persistence of an anionic exclusion phenomenon, even at high NaNO 3 concentration (see Figure), as
previously noticed on compacted bentonite by Ishidera et al. [3].
The influence of sodium nitrate increasing concentration on 22 Na + and 137 Cs + transfer is clearly seen on
experimental normalized fluxes through the samples. Its quantification is still on progress.
ANDRA is gratefully acknowledged for financial support.
[1] Descostes, M., Blin, V., Bazer-Bachi, F., Meier, P., Grenut, B., Radwan, J., Schlegel, M., Buschaert, S., Coelho, D.,
Tevissen, E. Diffusion of anionic species in Callovo-Oxfordian argillites and Oxfordian limestones (Meuse/Haute-
Marne, France). Appl. Geochem. 2008, 23, 655-677.
[2] Melkior T., Yahiaoui S., Thoby D., Motellier S., Barthes, V. Diffusion coefficients of alkaline cations in Bure
mudrock. Physics and Chemistry of the Earth, Parts A/B/C 2007, 32, 453-462.
[3] Ishidera T., Miyamoto S., Sato H. Effect of sodium nitrate on the diffusion of Cl(-) and I(-) in compacted bentonite.
J. of Nuclear Science and Technology 2008, 45, issue 7, 610-616.
Figure: Ratios of diffusion coefficients in free water (D 0 ), effective diffusion coefficients in COx clay rock
(D e ), and diffusivities in COx clay rock (D e /D 0 ) of 36 Cl - and tritiated water (HTO) vs. sodium nitrate
concentration.
PB2-17

INFLUENCE OF CLAY CONTENT ON HTO AND 36 Cl TRANSPORT PROPERTIES IN
CALLOVO-OXFORDIAN CLAY ROCK : PERCOLATION EXPERIMENTS AND MODELLING
C. Landesman (1) , S. Ribet (1) , C. Bailly (1) , J.C. Robinet (2) , B. Grambow (1)
(1) SUBATECH, 4 rue A. Kastler, 44307 Nantes Cedex, France
(2) ANDRA, Research and Development Division, 1/7 rue Jean Monnet – 92298 Châtenay-Malabry, France
In many international projects, clay-rock formations are intensively studied as potential hosts for radioactive
waste disposal facilities. In France, the Callovo-Oxfordian (COx) clay-rock has been selected as a reference
host rock due to its multiple favourable properties mainly its low permeability and high sorption capacity. At
the sedimentary formation scale, the COx layer (∼130m) is characterized by mineral content variations
mainly clay amount, notably in its upper part. This study is focused on the potential effects of the clay
content variations and its associated pore space geometry on RN transport properties (hydraulic properties,
accessible porosities, diffusion coefficients…).
In order to investigate the interrelationship between mineral composition, RN retention and macroscopic
transport properties, a specific strategy has been developed in Subatech for 7 years which allows probing and
characterization of natural rock pore water compositions with respects to dissolved concentrations of major
and trace elements as well as dissolved organic matter [1,2], identification of mineral/water interaction
processes, studies of water migration (hydrodynamics) in intact clay-rock, transport properties of major ions
and of RN in the interconnected saturated water pore network (considering anion exclusion and cation
retention).
Our strategy is based on two complementary approaches : i) an experimental approach based on a specific
transport experiment set-up (flow-through reactor) using cm-sized clay-rock core which allows to work with
sound samples under strictly controlled experimental conditions simulating in situ conditions (representative
synthetic pore water, control of P CO2 , anoxic conditions, …), ii) a coupled chemistry/transport modeling
taking into account the mineral composition notably the different clay phases, the pore water chemistry and
the interactions between clay surfaces and ions within the pore water (surface complexation and cation
exchange processes, anion porosity…).
Clay-rocks samples originated from the Meuse/Haute Marne URL (borehole EST 423; depth = 501m, C2c ).
Cylindrical cores are precisely machined (L=36 mm; D = 25 mm), embedded in epoxy resin and inserted in
stainless steel flow through reactors. Using a HPLC pump (40 bars), synthetic pore water is injected at
reactor inlet under controlled atmosphere (N 2 /CO 2 1%). Darcy velocities, calculated from measured outlet
flow rates, range from 0.2 to 1.5 10 -8 m s -1 . Water tracer (HTO) and anion tracer ( 36 Cl) are injected, in the
synthetic pore water through a 10 µL chromatography injection loop. HTO and 36 Cl breakthrough curves are
modeled with the 1D Advection-Dispersion-Reaction equation using PHREEQC geochemical code, coupling
chemistry and transport. Hydrodynamic parameters fitted are the accessible HTO and anion species
porosities and the hydrodynamic dispersion coefficient for each tracer. Typical HTO and 36 Cl experimental
and modeled breakthrough curves are shown below. For clay-rich samples (>15%wt.), typical values range
from 7 to 11% vol. and from 14 to 17% for anion and HTO accessible porosities respectively however, for
clay-poor sample (3%wt.) anion accessible porosity is about 4% vol. and HTO accessible porosity decreases
to less than 6%. HTO effective diffusion coefficient shows the same trend i.e. decreasing from 1.5–3.5 10 -11
m 2 s -1 for clay-rich samples to 3 10 -12 m 2 s -1 for clay-poor sample. Moreover, HTO hydrodynamic dispersion
coefficients decrease from 10 -9 to 5 10 -11 m 2 s -1 respectively. Then calculations of Peclet number
(advection/hydrodynamic dispersion ratio) indicate that the hydrodynamic regime is slightly (clay-poor
sample) to largely dominated (clay-rich sample) by advection. As a conclusion, the clay content appears to
be an important parameter controlling transfer properties and could influence the RN mobility at the scale of
the sedimentary layer.
In addition, major species concentrations are monitored in the outlet solution. Taken into account HTO and
36 Cl transport parameters, this chemistry/transport modelling allows simulate the evolution of the chemistry
in the outlet solution which expresses mineral/solution interactions.

Relative activity (mL -1 )
0.25
0.20
0.15
0.10
0.05
Exper. Model
HTO
36 Cl
0.00
0 5 10 15 20 25
Eluted volume after injection (mL)
HTO and 36 Cl experimental and modeled breakthrough curves
[1] Montavon G, Sabatié-Gogova A., Ribet S., Bailly C., Bessaguet N., Durce D., Giffaut E., Landesman C., Grambow
B. (2013), “Retention of iodide by the Callovo-Oxfordian formation; an experimental study” submitted in Applied Clay
Science
[2] Huclier-Markai S., Landesman C., Rogniaux H., Monteau F., Vinsot A., Grambow B. (2010), “Non-disturbing
characterization of natural organic matter (NOM) contained in clay rock pore water by mass spectrometry using
electrospray and atmospheric pressure chemical ionization modes”, Rapid Commun. Mass Spectrom, 24: 191-202
PB2-18
POROSITY, DIFFUSIVITY AND HYDRAULIC CONDUCTIVITY IN GRANITIC ROCK
MATRIX: LABORATORY MEASUREMENTS AND NUMERICAL MODELLING
Václava Havlová 1*) , Jan Najser 2) , Libor Gvoždík 3) , Karel Sosna 2) , Petr Večerník 1) , Jiří Záruba 2) and Petr
Dobeš 4)
1) Dept. of Fuel Cycle Chemistry, UJV Rez, a.s., Hlavni 130, 250 68 Řež, Czech Republic, *hvl@ujv.cz;
2) Arcadis Geotechnika Inc., Geologická 4, Prague, Czech Republic;
3) Progeo ltd., Tiché údolí 113, Roztoky u Prahy, Czech Republic;
4) Czech Geological Survey, Geologická 6, Prague, Czech Republic
Crystalline rocks are considered as potential host rock for deep geology repository (DGR) of nuclear waste
in many European countries. In order to define safety functions of host rock and gained performance
assessment relevant input data, detailed examination of migration parameters for relevant inventory
radionuclides has to be performed. DGR is usually planned to be constructed at the depth 400 – 600 m below
ground surface. The concept of deep geological disposal is based on KBS-3 concept. Furthermore, granite is
being considered as the host rock. Therefore, the aim of the current research project was to study relation of
migration properties (porosity, diffusion, hydraulic conductivity) in the context of pore space representation
(considering the pore opening and length).
45 granitic samples from ten different granitic bodies, comprising rock material of different structure and
grain size, were subjected to test in order to determine porosity (ε), effective diffusion coefficients (D e ) and
hydraulic conductivity (K). Porosity was determined using water-saturation method [1]. The effective
diffusion coefficient D e was measured using 3 H as a tracer in through-diffusion experiments. The activities in
both input and output reservoirs were regularly monitored using liquid scintillation spectrometry. Hydraulic
conductivity was measured in pressure cells. A constant pressure difference of Δ = 50 kPa was applied by
pressure controllers and the volume of water that passed through the sample was recorded.
Porosity values differed from 0.3 – 1.2 %; while the effective diffusion coefficient ranged from 4 x 10 -13 to
2.0 x 10 -12 to m 2 .s -1 . The hydraulic conductivities of fresh granite varied from 3.0 x 10 -14 - 3.0 x 10 -11 to m.s -1 .
The trends between diffusivity and hydraulic conductivity, however, were not so clear.

In order to put the measured values of hydraulic conductivity K and effective diffusion coefficient D e into
relation, numerical simulations of both through-diffusion and pressure cell experiments were undertaken
using finite-element code NAPSAC. A discrete fracture network approach with stochastically generated
microcracks was used for constructing granite rock matrix.
The size of microcracks for different types of granites was determined on the basis of digitized figures from
fluorescent microscope where pore network in granitic samples, doped with fluorescent resin, was studied.
The aperture of micropores was defined from the pore distribution measured with mercury porosimetry.
The numerical modelling results showed that reduced microcracks length within conservative pore volume
had an influence on the connectivity of the microcrack network. It led to increasing diffusivity, and
decreasing hydraulic conductivity. The imperfect correlation between experimental “K” and “De” can be
explained by the different microcracks opening and different geometries of the microcrack networks within
the studied granites.
The model of pore space representation of granitic rock matrix would definitely be help for solving the
problems of small scale laboratory experiments transfer into real scale of rock massive. Further development
of the model is ongoing.
[1] Melnyk T. W., Skeet A. (1986): An improved technique for determination of rock porosity. Can. J. Earth
Sci. 23, 1068 - 1074.
PB2-19
MASS TRANSPORT IN SHALE MATRIX UNDER SALINE CONDITIONS
Peter Vilks 1) , Neil H. Miller 1) , Jeffrey G. Miller 1) and Tammy Yang 2)
1) Atomic Energy of Canada Limited, Whiteshell Laboratories, Pinawa, Manitoba, Canada, R0E 1L0
2)
Nuclear Waste Management Organization, 22 St. Clair Avenue East, 6 th Floor, Toronto, Ontario, Canada,
M4T 2S3
The Nuclear Waste Management Organization’s (NWMO) Adaptive Phased Management (APM) Technical
Program is intent on advancing the understanding of solute migration in deep crystalline and sedimentary
groundwater systems. As part of this program, a combination of batch and mass transport sorption tests are
being applied to study sorption in low permeability sedimentary rocks under saline porefluid conditions,
using a reference Na-Ca-Cl brine solution with a TDS of 275 g/L. Mass transport tests are used to examine
sorption considering the specific surface area actually accessible to migrating elements. In this work,
dynamic mass transport tests were conducted for a low permeability (3x10 -21 m 2 ) shale to investigate how the
sorption coefficients (K d ) determined from these tests compare to values measured using batch sorption tests.
Mass transport tests were performed with both advective and diffusive mass transport methods using Ni, Cu,
Pb and U as sorbing tracers and Li as a non-sorbing tracer. An advective transport test with induced
hydraulic flow was performed using the High Pressure Radioisotope Migration (HPRM) apparatus [1] in an
attempt to obtain transport information on a shorter time scale than possible with a diffusion test. While
applying a confining triaxial pressure of 4 to 7 MPa to a core sample (25 mm diameter and 5.5 mm length),
brine and tracers were pumped through the shale (along the core axis) using pressures up to 3.9 MPa for 144
days. Transport was evaluated by monitoring the concentrations of eluted tracers, and by examining post-test
tracer concentration profiles in the shale. In a separate test, diffusive transport was studied using a throughdiffusion
experiment in which tracer concentrations diffusing through a shale core sample (62.5 mm
diameter and 9.75 mm length) were monitored. Tracer diffusion profiles were also measured within the rock
at the end of the diffusion test. These profiles were intended to characterize the diffusion of those strongly
sorbing elements that did not diffuse through the shale sample during the experimental time period of 141
days. The results of both advective and diffusive mass transport tests were interpreted with compartment
models set up within the AMBER modelling environment, and designed to simulate the geometry of both
tests. Batch sorption experiments for elements Ni, Cu, Pb, Zr, U and Li sorbing on shale were performed in
the reference brine solution and the measured K d values were compared to the results from the mass transport
tests. The K d values derived from both the advective transport and the through-diffusion tests were consistent

with K d values determined by batch tests. The element concentration profiles from the diffusion test were the
most reliable for estimating K d values during mass transport due to uncertainties with regard to effective
surface area available to sorption in the advective transport test.
[1] Vilks, P, and N.H. Miller. 2007. Evaluation of experimental methods for characterizing diffusion in
sedimentary rocks. Nuclear Waste Management Organization Technical Report NWMO-TR-2007-11,
Toronto, Canada.
PB2-20
MIGRATION OF KEY RADIONUCLIDES THROUGH HOLLINGTON SANDSTONE
O. Preedy. M. Felipe Sotelo, N.D.M. Evans
Department of Chemistry, Loughborough University, Loughborough, Leicestershire, LE11 3TU, UK
Cement leachate/host rock boundaries are potentially significant interfaces relevant to deep geological
disposal; sandstone has been selected for this research because it comprises a large percentage of the UK’s
overburden rock. Over time the sandstone may encounter groundwater that is highly alkaline due to the
presence of cement pore water. It is important that effects on the structure composition and morphology of
the rock are understood. An important area of concern is to establish whether the alteration of host rock
affects migration of key radionuclides.
Dynamic radial diffusion experiments have been
conducted using Hollington sandstone and simulant
cement leachates. The four solutions used are
representative of different stages relevant to a
geological disposal facility (GDF), ranging from
initially highly alkaline conditions to simulant ground
water.
The experimental procedure consists of 5 x 5 cm
cylindrical sandstone cores which are spiked with a
radiotracer for example 99 Tc, and submerged in the
cement leachate solution. The rate of diffusion through the rock is determined by the increase in
concentration of the tracer in the leachate solution. The experiment proceeds in four consecutive stages
changing the leachate solution to determine the effect on migration. The elements being considered are
Uranium, Thorium and Technetium. Breakthrough for technetium was observed after approximately two
days in the young cement leachate (YCL) and approximately 4 days for the aged material in the intermediate
cement leachate (ICL), no breakthrough has been observed for the other experiments.
99
Tc Diffusion Through Sandstone Block C/C 0
99-Tc YCL
100.00%
90.00%
80.00%
70.00%
60.00%
C/C 0
50.00%
40.00%
30.00%
20.00%
10.00%
0.00%
HTO
99-Tc ICL
0 10 20 30 Time 40 / Days 50 60 70 80

PB4
PB4-1
PB4-2
PB4-3
PB4-4
PB4-5
PB4-6
PB4-7
PB4-8
PB4-9
PB4-10
PB4-11
PB4-12
EFFECTS OF BIOLOGICAL AND ORGANIC
MATERIALS
ASSESSING THE SUITABILITY OF HYDROXYAPATITE BIOMINERALS FOR THE
REMEDIATION AND IMMOBILISATION OF AQUEOUS RADIONUCLIDES
S. Handley-Sidhu, R.A.D Pattrick, J.M. Charnock, J.R Lead,
B. Stolpe, L.E. Macaskie (UK, USA)
BACTERIAL DIVERSITY IN MONT TERRI OPALINUS CLAY AND THE INFLUENCE
OF THE BACTERIAL SPOROMUSA SP. ISOLATE ON PLUTONIUM SPECIATION
H. Moll, L. Lütke, V. Bachvarova, A. Geissler, S. Selenska-Pobell,
G. Bernhard (Germany)
BIODEGRADATION OF CELLULOSE DEGRADATION PRODUCTS
S. Rout, P. Shaw, A. McCarthy, D. Rooks, P. Loughnane,
C. Doulgeris, P. Humphreys, A. Laws (UK)
BIOSORPTION AND MECHANISM OF URANIUM (VI) ON BACILLUS SP ISOLATED
FROM AERATED ZONE SOIL
Xiaolong Li, Congcong Ding, Jiali Liao, Yuanyou Yang, Dong Zhang, Jijun Yang, Jun Tang,
Ning Liu (China)
THE IMPACT OF ALKALIPHILIC AND ALKALITOLERANT MICROORGANISMS
FROM A HYPER-ALKALINE SPRING ON THE TRANSPORT CHARACTERISTICS OF
SANDSTONE
S. Smith, J. Lloyd, J. West (UK)
THE BIGRAD CONSORTIUM - GEOMICROBIOLOGY OF CEMENTITIOUS NUCLEAR
WASTE
A.J. Williamson, K. Morris, G.T.W. Law, S. Shaw, C. Boothman,
J.R. Lloyd (UK)
COMPARISON OF MICROBIOLOGICAL INFLUENCES ON THE TRANSPORT AND
CHEMICAL PROPERTIES OF INTACT SANDSTONE AND ITS RELEVANCE TO
GEODISPOSAL
J. Wragg, K. Bateman, A.E. Milodowski, J.M. West (UK)
MICROBIAL DEGRADATION OF ISA UNDER HIGH PH CONDITIONS
REPRESENTATIVE OF INTERMEDIATE LEVEL WASTE
N.M. Bassil, N. Bryan, J.R. Lloyd (UK)
REDUCTION OF SELENITE BY BIOFILMS OF AN IRON-REDUCING BACTERIUM
Y. Suzuki,* H. Saiki, A. Kitamura, H. Yoshikawa (Japan)
MOLECULAR CHARACTERIZATION OF U(VI) ASSOCIATION WITH U RESISTANT
CLAY BACTERIAL ISOLATE UNDER NEUTRAL CONDITIONS
M. Lopez Fernandez, I. Sanchez Castro , M. Romero Gonzalez,
A. Guenther, P.L. Solari, M.L. Merroun (Spain, UK, Germany)
BIOREMEDIATION OPTIONS FOR REMOVING URANIUM FROM
GROUNDWATER
L. Newsome, K. Morris, D. Trivedi, J. Lloyd (UK)
THE BIGRAD CONSORTIUM - THE HYDROGEN DRIVEN
GEOMICROBIOLOGY OF CEMENTITIOUS NUCLEAR WASTE
M.J.C. Crouch, K. Morris, D. Engelberg, J. Small, J.R. Lloyd (UK)

PB4-1
ASSESSING THE SUITABILITY OF HYDROXYAPATITE BIOMINERALS FOR THE
REMEDIATION AND IMMOBILISATION OF AQUEOUS RADIONUCLIDES
S. Handley-Sidhu (1) , R.A.D. Pattrick (3) , J.M. Charnock (3) , J.R Lead (4) , B Stolpe (1) , L.E. Macaskie (2)
(1) School of Geography Earth and Environmental Sciences, (2) School of Biosciences, The University of Birmingham,
Edgbaston, Birmingham B15 2TT, UK. (3) School of Earth, Atmospheric and Environmental Sciences, The University of
Manchester, Oxford Road, Manchester, M13 9PL, UK. (4) Arnold School of Public Health, University of South Carolina,
USA
Apatites (general formula Ca 5 (PO 4 ) 3 (OH,F,Cl)) have been suggested as a suitable material for the
remediation of heavy metals in groundwater (i.e. permeable reactive barriers; PRB)[1] and as a potential
nuclear waste disposal material[2]. Serratia sp. cells biomanufacture nanophase hydroxyapatite
[Ca 1O (PO 4 ) 6 (OH 2 )] (BHAP) from glycerol 2-phosphate (G2P) and Ca 2+ [3]. Serratia sp. (originally isolated as
a Citrobacter sp. from a contaminated land site[4]) contains high levels of an atypical phosphatase; this
enzyme cleaves glycerol 2-phosphate, producing inorganic phosphate and, in the presence of calcium ions,
the solution will become supersaturated and precipitation of BHAP occurs[3].
Although BHAP has shown promise for radionuclide remediation (up to 15 times higher sorption capacity
than commercial hydroxyapatite for Sr(II) and Co(II))[5], these materials can contain between 5-50%
organic material[6]. This high organic content is not suitable for waste storage/disposal as the bio-degraded
organic material could lead to the remobilisation of adsorbed radionuclides via production of biogenic
organic acids and chelating agents [6].
The aim of the present study is to determine the efficacy of BHAP for the removal of aqueous Sr(II), Co(II),
Eu(III) and U(VI). Heat treating the BHAP materials prior to sorption experiments yields valuable
information on the role of the organic phase and mineral structure in radionuclide
incorporation/immobilisation. Eu(III) is a minor contributor to nuclear waste and an analogue for the highly
active trivalent actinide species, such as Am(III), Np(III) and Pu(III). The radionuclides 90 Sr, 60 Co and
238 U/ 235 U are contributors to radioactivity in nuclear wastes and environmental contamination associated with
nuclear energy and weapons production. Uranium contamination is also specifically associated with the use
of depleted uranium weapons.
1-2 µm
Extracellular Polymeric Substance
Serratia sp. Cell
Periplasm Phosphatase Enzyme
700 °C
Aqueous Radionuclide
Glycerol-2-
Phosphate
HPO 4
2-
Ca 2+
Biogenic Hydroxyapatite
Ca 5 (PO 4 ) 3 (OH)
Maximum loadings of Sr(II), Co(II), Eu(III), and U(VI) onto BHAP were 3.6, 5.7, 23.4 and 31.2 % w/w,
respectively. Analysis by Extended X-ray Absorption Fine Structure (EXAFS) showed that the organic
content and mineral structure influenced the binding mode of radionuclides. For the low heat-treated BHAP
material (< 450 °C), Sr(II) was found to incorporate into the bulk amorphous phase, however, once heattreated
to a more crystalline material (700 °C), the Sr(II) bonding was consistent with substitution at the
Ca(1) or Ca(2) position of the hydroxyapatite crystal structure. Crystalline bound Sr(II) was found to be
more stably incorporated with only 6 % of the incorporated metal remobilised in groundwater, whereas,
amorphously bound Sr(II) was incorporated less stably with up to 24 % remobilized. Further EXAFS and

XRD evidence suggests that Co(II), Eu(III) and U(VI) are preferentially associated within the amorphous
phase of the BHAP in all temperature treatments. Co(II) incorporation was less stable with up to 6.6 %
remobilised in groundwater, whereas, Eu(III) and U(VI) were incorporated most stably with < 2%
remobilised in groundwater. With high metals loadings for the analogue actinides and good stability against
remobilisation, BHAP is a promising material for radioactive waste remediation technologies.
[1] Oelkers, EH. And JP. Montel (2003). Phosphates and Nuclear Waste Storage. Elements, 4, 113-116.
[2] Simon, FG. Biermann, V. Segebade, C. Hedrich, M (2004). Behaviour of uranium in hydroxyapatite-bearing
permeable reactive barriers: investigation using 237U as a radioindicator Sci. Total Environ. 326, 249-256.
[3] Thackray, AC. Sammons, RL. Macaskie, LE. Yong, P. Marquis, P.M (2004). Bacterial biosynthesis of calcium
bone-substitute material. J. Mater. Sci. Mater. Med. 15, 403-406.
[4] Macaskie, LE.; Dean, ACR (1984) Cadmium accumulation by a Citrobacter sp. J. Gen. Microbiol. 130, 1855-1867.
[5] Handley-Sidhu, S. Renshaw, J. C. Moriyama, S. Stolpe, B. Mennan, C. Bagheriasl, S. Yong, P. Stamboulis, A.
Paterson-Beedle, M. Sasaki, K. Pattrick, RAD. Lead, JR. Macaskie, LE (2011). Uptake of Sr(II) and Co(II) into
Biogenic Hydroxyapatite: Implications for Biomineral Ion Exchange Synthesis. Environ. Sci. Technol. 45, 6985–6990.
[6] Brookshaw, DR. Pattrick, R.A.D. Lloyd, J.R. Vaughan, D.J (2012). Microbial effects on mineral radionuclide
interactions and radionuclide solid-phase capture processes. Mineral. Mag. 73, 777-806.
PB4-2
BACTERIAL DIVERSITY IN MONT TERRI OPALINUS CLAY AND THE INFLUENCE OF THE
BACTERIAL SPOROMUSA SP. ISOLATE ON PLUTONIUM SPECIATION
H. Moll (1) , L. Lütke (1) , V. Bachvarova (2) , A. Geissler (1) , S. Selenska-Pobell (1) , G. Bernhard (1)
(1) Institute of Resource Ecology, Helmholtz-Zentrum Dresden-Rossendorf e.V., P.O. Box 510119, 01314
Dresden, Germany
(2)
Department of Developmental Biology / Center for Medical Biotechnology, University of Duisburg-Essen,
45117 Essen, Germany
The concept of geological disposal of nuclear waste comprises a detailed knowledge concerning potential
host rock formations. One of such formations is the Opalinus Clay geologic layer of the Mont Terri
Underground Rock Laboratory (Switzerland). It is well known that bacteria indigenous to such subterranean
soil environments can affect the speciation and hence the mobility of actinides [1, 2]. Dominant bacterial
strains from sites destined for future nuclear waste deposition have to be identified and investigated
regarding their interaction mechanisms with soluble actinide ions [e.g., 3].
For the first time microbial total DNA (tDNA) was isolated from 50 g unperturbed Mont Terri Opalinus
Clay. Analysis of the tDNA revealed that the bacterial community of the unperturbed Opalinus Clay is
dominated by representatives of Firmicutes, Betaproteobacteria, and Bacteriodetes. Representatives of
Firmicutes completely overgrow the other members of the community after treatment of the clay with a
specific medium for oligotrophic microorganisms called R2A (components: protease peptone, yeast extract,
casamino acids, glucose, starch, sodium pyruvate, dipotassium phosphate, magnesium sulfate, and agar).
Bacteria isolated from Mont Terri Opalinus Clay on R2A medium were affiliated with different Sporomusa
spp., Paenibacillus spp., and Clostridium spp..
After isolation, characterization, and cultivation, we studied the unknown interaction between plutonium in
mixed oxidation states (61±5% Pu(VI) and 18±1% Pu(IV)-polymers) and cell-suspensions of one of the
Sporomusa sp. MT-2 isolates. The knowledge concerning the impact of bacterial isolates recovered from
clay on the plutonium speciation is scarce. Accumulation experiments were performed in order to obtain
information about the amount of Pu bound by the bacteria in dependence on the contact time and the initial
plutonium concentration. We used solvent extraction and UV-vis-NIR absorption spectroscopy (see Fig. 1)
to determine the speciation of Pu oxidation states. Sporomusa sp. MT-2 was grown under anaerobic
conditions (N 2 atmosphere) at 30°C using R2A medium. Culture purity was ensured by light microscopy and
molecular analysis (in situ polymerase chain reaction and restriction fragment length polymorphism). Cells
were harvested in the mid-exponential growth phase, washed and suspended in 0.9% NaCl solution. The
experiments were performed by using [dry biomass] of 0.33 ± 0.01 g dry weight /L resuspended in 0.1 M NaClO 4
solution at pH 6.2 and 25 °C under N 2 atmosphere. [ 242 Pu] initial concentration varied between 0.2 and 110
mg/L. The 242 Pu present in blank (no cells added), supernatant, and washed biomass suspension at pH 0 was

analyzed using UV-vis-NIR spectroscopy, solvent extraction, and liquid scintillation counting (LSC) as
described in [4]. In addition experiments were performed by adding an electron donor (Na-pyruvate) in two
concentrations 0.1 mM and 10 mM.
Absorbance / a.u.
0.08
0.07
0.06
0.05
0.04
0.03
0.02
0.01
PuO 2
2+
Pu(VI)-hydroxo species
100 %
60 %
44 %
29 %
Supernatants
[ 242 Pu] initial
110 mg/L
pH 6.2
0.1 M NaClO 4
Contact time:
1.8 h (no cells)
1.8 h
25 h
49 h
73 h
Absorbance / a.u.
0.008
0.007
0.006
0.005
0.004
0.003
0.002
0.001
Pu(V)
1.8 h (no cells)
1.8 h
25 h
49 h
73 h
0.00
820 830 840 850 860 870 880 890
Wavelength / nm
0.000
550 560 570 580 590
Wavelength / nm
Figure 1. Absorption spectra of Pu ([ 242 Pu] initial = 110 mg/L) at pH 6.2 after different contact times with 0.33
g dry weight /L of Sporomusa sp. MT-2.99 after separating the cells by centrifugation.
The amount of Pu sorbed by Sporomusa sp. cells increased with time. Steady state conditions were reached
after approximately 200 h. The data could be successfully fitted to a bi-exponential law. The amount of Pu
associated with Sporomusa sp. cells depends on the initial 242 Pu concentration. In the first step, a fast binding
of the Pu(VI) and Pu(IV)-polymers onto the biomass occurred. Solvent extractions showed that 92 % of the
initially present Pu(VI) was fast reduced to Pu(V) in the presence of the cells within the first 48 h of contact
time (no electron donor added). The corresponding redox potential in the cell suspensions dropped down to
300 mV compared to 780 mV measured in the blanks. This reduction process was much slower in all abiotic
controls. Most of the formed Pu(V) dissolves from the cell envelope back to the aqueous solution due to the
weak complexing properties of Pu(V) (see Fig. 1). Similar observation were made for instance with the
aerobic soil heterotrophs Bacillus spaericus and Pseudomonas stutzeri which were able to reduce one third
of the initially present Pu(VI) to Pu(V) within the first 24 h of incubation [5]. Good binding properties of
Pu(IV)-polymers on functional groups of the Sporomusa sp. cell envelope were found (immobilization). In
contrast to earlier measurements with Pseudomonas fluorescens (CCUG 32456A) cells [6] clear indications
for increased amounts of Pu(IV) and Pu(III) on the Sporomusa biomass were observed. The Pu oxidation
state distributions as a function of time will be discussed in detail and the results of the system without
addition of electron donors will be compared with the electron donor supplemented systems. Here
differences in the Pu interaction mechanism were found.
ACKNOWLEDGMENTS. The authors thank the BMWi for financial support (contract no.: 02E10618) and
the BGR for providing the clay samples.
[1] J.R. Lloyd, G.M. Gadd, Geomicrobiol J 28, 383 (2011).
[2] D.R. Brookshaw, et al., Mineral Magazine 76, 777 (2012).
[3] C. Anderson, et al., Geomicrobiol J 28, 540 (2011).
[4] H. Moll, et al., Radiochim. Acta 94, 815 (2006).
[5] P.J. Panak and H. Nitsche, Radiochim. Acta 89, 499 (2001).
[6] H. Moll, et al., Plutonium Futures – The Science 2012, p. 44.

PB4-3
BIODEGRADATION OF CELLULOSE DEGRADATION PRODUCTS
Simon Rout 1* , Paul Shaw 1 , Alan McCarthy 2 , Dave Rooks 2 , Paul Loughnane 2 , Charalampos Doulgeris 1 , Paul
Humphreys 1 and Andy Laws 1
1 School of Applied Sciences, University of Huddersfield, Huddersfield, West Yorkshire, HD1 3DH
2 Microbiology Research Group, Institute of Integrative Biology, Biosciences Building, University of
Liverpool, Liverpool, L69 7ZB
As part of the NDA/EPSRC funded C14-BIG project a series of batch fed microcosms have been established
to provide stable biogeochemical environments for future experiments on the fate and transport of graphite
associated C-14. These microcosms have been inoculated with near-surface sediments (ca. pH 7.0) obtained
from the canal dock at the University of Huddersfield and fed on Cellulose Degradation Products (CDP)
generated during the alkaline hydrolysis of cellulose. The UK intermediate level waste (ILW) inventory
contains a range of cellulosic materials such as paper and tissue originating from nuclear operations. CDP is
composed of a range of low molecular weight organic compounds including the erythro and threo forms of
2-C-(hydroxymethyl)-3-deoxy-D-pentonic (isosaccharinic) acid (ISA). The isomers of ISA are of interest in
radioactive waste disposal since they are able to complex and mobilise a range of radionuclides including
plutonium.
Consequently the metabolism of CDP in general, and more specifically ISA, by micro-organisms has the
potential to affect the migration of some radionuclides through the near field and the geosphere. The
construction and operational phases of a deep geological disposal facility provide an opportunity for the
colonisation of the facility by microorganisms from the near-surface environment. Therefore, the
establishment of CDP-fed microcosms based on near-surface microbial communities is one potential
representation of a post-closure radioactive waste disposal environment.
In order to characterise these microcosms prior to the inclusion of graphite and associated C-14 a range of
chemical and biological analysis have been performed. Overall CDP removal has been demonstrated by TOC
analysis and the removal of both erythro and threo ISA through HPAEC-PAD. Weekly batch feeding of the
reactors has seen a gradual accumulation of the threo form of ISA and acetate through incomplete
degradation as pH increases. Gas generation profiles indicate methane and carbon dioxide evolution
suggesting the establishment of a stable CDP driven methanogenic community.
Microbial community analysis through 16S rRNA gene cloning and sequencing has shown the presence of a
range of bacterial and methanogenic species capable of generating methane through both H 2 /CO 2 and acetate
pathways across three pH systems at 7.5, 9.5 and 10. Mass balance modelling approaches have been
employed to determine the carbon flow through the system. Data generated to date indicate that microbial
communities able to degrade components of CDP including ISA are present in environments that have not
previously encountered these compounds. This suggests that microbial populations derived from nearsurface
environments may be able to colonise and survive in the post-closure environment of a geological
disposal facility for radioactive wastes.
This work was partially funded by the RCEP/NDA-RWMD Geological Disposal of Nuclear Waste
programme. The Research Councils Energy Programme is a Research Councils UK cross council initiative
led by EPSRC and contributed to by ESRC, NERC and STFC.

PB4-4
BIOSORPTION AND MECHANISM OF URANIUM (Ⅵ) ON BACILLUS SP ISOLATED FROM
AERATED ZONE SOIL
Xiaolong Li (1) , Congcong Ding (2) , Jiali Liao (1)* , Yuanyou Yang (1) , Dong Zhang (3) , Jijun Yang (1) , Jun Tang (1) ,
and Ning Liu (1)
(1)
Key Laboratory of Radiation Physics and Technology, Ministry of Education; Institute of Nuclear Science
and Technology, Sichuan University, Chengdu 610064, P. R. China
(2)
Key Laboratory of Biological Resource and Ecological Environment of the Ministry of Education, College
of Life Sciences, Sichuan University, Chengdu 610064, P.R. China
(3)
Institute of Nuclear Physics and Chemistry, CAEP, Mianyang 621900, P. R. China
As an important environmental media, microorganisms are ubiquity and have good adaptability to hostile
environments of heavy radiation, high salinity and temperature. Moreover, Microbes are well known to have
significant influence on the chemical and migration behaviors of radionuclides via biosorption,
complexation, oxidoreduction and other behaviors. Therefore, it is of great importance to investigate the
interaction between radionuclides with dominant microbes isolated from the media around the radioactive
waste disposal.
In this study, one of the dominant bacteria species was isolated from a potential disposal site of (ultra-) low
uraniferous radioactive waste in Southwest China and then its biosorption behavior and mechanism with
uranium (Ⅵ) was examined via FTIR, SEM, TEM-EDX et al. The preliminary results of macroscopical
biosorption tests indicated that the uranium (Ⅵ) biosorption by bacillus sp was remarkably affected by pH
value. About 80% of total uranium (VI) was absorbed by bacillus sp of 50mg at an initial concentration of 10
mg/L uranium (VI) nitrate solution at room temperature, and the biosorption equilibrium could be obtained
within 12h. The uranium (VI) adsorption capacity was more than 6.0 mg/g bacillus sp (wet weight).
FTIR spectra for control and uranium loaded cells between 4000 cm −1 and 400 cm −1 was implied that a
distinct peak at 910 cm −1 and tiny changes in peak positions and intensity around 550–1000 cm −1 could be
obviously observed in the uranium loaded sample, which resulted from the asymmetric stretching vibration
of UO 2
2+
and stretching vibrations of weekly bonded oxygen ligands with uranium (U–O ligand ).
SEM and TEM were applied to observe the cellular localization of the absorbed uranyl ions. Some floccule
depositions were observed on the cell surface after uranium (Ⅵ) biosorption via SEM (Fig. 1b). These
sediments might be the complex of uranyl ions interaction with the cell surface functional groups.
Furthermore, TEM was employed to verify the distribution and localization of uranium precipitation in the
cells. Acicular depositions were found at both cell wall and cell interior (Fig. 1d). It was possible to confirm
that the needlelike precipitates observed inside the cell wall contained uranium using EDX coupled to TEM
for the elemental characterization of the metal sediments. The presence of specific peak for uranium in U
loaded sample confirmed the presence of absorbed radionuclide (Fig. 1f).
Acknowledgements
This work was financially supported by China National Natural Science Foundation (Grant No. 21071102,
91126013), Joint Funds of China National Natural Science Foundation and China Academy of Engineering
Physics (Grant No. 10476015) and the National Fund of China for Fostering Talents in Basic Science
(J1210004).

a
b
c
d
e
f
Fig. 1. SEM and TEM images of bacillus sp and U-adsorbed bacillus sp.
SEM: bacillus sp (a) and U-adsorbed bacillus sp (b);
TEM: bacillus sp (c) and U-adsorbed bacillus sp (d);
Energy dispersive X-ray spectra of intracellular (f) U precipitates showed by e.
PB4-5
THE IMPACT OF ALKALIPHILIC AND ALKALITOLERANT MICROORGANISMS FROM A
HYPER-ALKALINE SPRING ON THE TRANSPORT CHARACTERISTICS OF SANDSTONE
Sarah Smith 1,2 Jon Lloyd 1 and Julia West 2
1) School of Earth, Atmospheric and Environmental Sciences, University of Manchester, Williamson
Building, Oxford Road, Manchester, M13 9PL
2) British Geological Survey, Environmental Science Centre, Nicker Hill, Keyworth, Nottingham, NG12
5GG
A cementitious repository for geological disposal of radioactive waste will generate hyper-alkaline fluids
which will impact on the engineered barriers and the host-rock. The high-pH environment will also influence
the composition of microbial communities in both in the excavation damaged zone (EDZ) and in the host
rock impacting on biogeochemical processes affecting redox, pH etc. and consequently, radionuclide
migration. Biofilm formation may also affect rock transport processes (Coombs et al, 2010); interactions
between microorganisms and mineral surfaces will influence rock characteristics such as porosity, indirectly
affecting radionuclide transport throughout the host rock. The presence of microorganisms may also directly
impact upon radionuclide mobility. Microbial metabolism under high pH conditions may be severely
inhibited in neutralophilic microorganisms due to an inability to sufficiently regulate cytoplasmic pH;

alkaliphilic and alkalitolerant microorganisms have regulatory mechanisms to enable them to cope, and are
therefore able to grow under highly alkaline conditions (Padan et al. 2005).
This study focusses on microbial activity under hyper-alkaline conditions, and the impacts on environmental
characteristics of the far-field. To investigate some of these microbe/mineral/fluid interactions, slurry
samples were collected from a hyper-alkaline spring at Harpur Hill near Buxton, Derbyshire. Sampling
points varied in pH from 7.5-13. Epifluorescence microscopy was used to carry out total cell counts, which
indicated that cell numbers fell rapidly as pH increased, although viable cells were still present at pH 13. A
total of 45,012 sequences were obtained by amplifying a 311bp region of the16S rRNA gene. Sequence
analysis showed that the microbial communities present at each of the sampling points were highly diverse,
and differed between sampling sites. The pH 8.9 sampling site contained the highest diversity of bacteria,
with a total of 2090 operational taxonomic units (OTU’s) observed. The pH 7.5, 11.5 and 12 sampling sites
also contained diverse communities, and at the pH 13 site 390 OTU’s were observed. The pH 8.9 and 12
sites were found to have complex microbial community structures, with no single species being dominant.
The pH 7.5 and 11.4 sampling sites were dominated by Xanthomonadaceae and Burkholderiales, with a large
proportion of the pH 11.4 site community being Paenbacillaceae. The pH 13 site was found to be a fairly
complex community, with Pseudomonadaceae dominant.
Microcosm experiments were assembled to investigate the biogeochemical activity of the microorganisms
collected from the site, under pH conditions ranging from 7.5-13. Microcosms contained crushed sandstone,
inoculated with sediment and fluid collected from the site. Results showed Fe(III) reduction occurred up to
pH 11.5. Viable microbial cells were present in pH 12 and pH 13 microcosms, but reduced iron was not
observed. These findings are consistent with observations from other studies where the upper pH limits for
microbial activity has been found to be around pH 12 (Rizoulis et al, 2012).
Flow-through column experiments, using sediment and fluids from the pH 11.5 site, then investigated the
impact of the microbial communities on transport characteristics of rock. In brief, fluid was pumped through
PEEK columns packed with crushed sandstone under anaerobic conditions for a total of six weeks. In some
columns, 5mM acetate, 5mM lactate and 50 mgL -1 yeast extract were added as electron donors to stimulate
microbial activity. Total cell counts in the outlet fluid increased rapidly towards the end of the experiment in
the columns with the added electron donors, reaching around 10 9 cells ml -1 , whereas under control conditions
total cell counts remained at approximately 10 5 cells ml -1 throughout. Slight decreases in fluid flow rate were
observed in the biotic columns, suggesting bioclogging. In the biotic experiments, lactate concentrations in
the outlet fluid fell rapidly, and was not present after three weeks. Propionate formation occurred after two
weeks, eventually reaching a concentration of 219 mgL -1 , suggesting lactate fermentation was occurring.
Microbial fermentation leads to the formation of gases, including H 2 , which could have implications in a
repository environment. Ribosomal Intergenic Spacer Analysis on samples taken from along both the biotic
and control columns indicated that microbial diversity, and also total biomass, decreased along the columns
from the inlet end, and that microbial diversity appeared to be higher in the biotic columns. Further analyses
will investigate changes in microbial community composition along the columns. Results from these
experiments indicate bioclogging under alkaline conditions may impact the flow rate through rock,
suggesting the migration of radionuclides may be altered.
The results from these experiments will be used to plan further column experiments with crushed sandstone,
and also with intact sandstone core to investigate how microbial activity impacts upon the transport
characteristics.
Coombs P, Wagner D, Bateman K, Harrison H, Milodowski AE, Noy D and West JM (2010) The role of biofilms in
subsurface transport processes, Quarterly Journal of Engineering Geology & Hydrogeology, 43: 131-139.
Padan E, Bibi E, Ito M and Krulwich TA (2005) Alkaline pH homeostasis in bacteria: New insights, Biochimica and
Biophysica Acta (BBA)- Biomembranes, 1717: 67-88.
Rizoulis A, Steele HM, Morris K and Lloyd JR (2012) The potential impact of anaerobic microbial metabolism during
the geological disposal of intermediate-level waste, Mineralogical Magazine, 76: 3261-3270.

PB4-6
THE BIGRAD CONSORTIUM - GEOMICROBIOLOGY OF CEMENTITIOUS NUCLEAR
WASTE
Adam J. Williamson 1) , Katherine Morris 1) , Gareth T. W. Law 2) , Sam Shaw 1) , Christopher Boothman 1) and
Jonathan R. Lloyd 1)
1) Research Centre for Radwaste and Decommissioning and Williamson Research Centre, School of Earth,
Atmospheric and Environmental Sciences, The University of Manchester, UK
2)
Centre for Radiochemistry Research and Research Centre for Radwaste and Decommissioning, School of
Chemistry, The University of Manchester, UK
It is policy that the UK’s intermediate level radioactive wastes (ILW) will be disposed of in a deep
geological disposal facility (GDF). Here, the current generic GDF design is to mix ILW with cement,
grouting into stainless steel canisters followed by further backfilling possibly with cementitous material.
Post-closure, re-saturation of the cementitious GDF over time will form a hyperalkaline, chemically
disturbed zone (CDZ) within the host rock. Microbial processes, especially metal reduction, may immobilise
redox active radioactive contaminants in the waste either through direct enzymatic reduction of radionuclides
or via interactions with biogenic Fe(II), yet these reactions remain poorly characterised under these extreme
conditions.
Here, we explore the extent of microbiologically-mediated Fe(III) reduction under alkaline conditions in
analogue sediment from a lime workings site. In addition, we consider the impact of these processes on
radionuclide (U and Tc) behaviour. Microcosms were set up using sediments taken from the site, adjusted to
pH 10, augmented with electron donor (organic acids with yeast extract) and to probe for Fe(III) reduction,
some systems were run with added Fe(III) oxyhydroxides. Microcosms were capped and monitored for the
development of microbial reduction processes to assess the biogeochemical behaviour of Fe(III) and U(VI)
or Tc(VII) using X-ray absorption spectroscopy (XAS) to assess the speciation and co-ordination
environment of U/Tc in these systems. Additionally, pre-reduced, Fe(II)-bearing sediments that were reacted
with radionuclide were also analysed.
A cascade of microbial reduction processes occurred at pH 10 – 10.5 in all microbially active systems. In
uranium-supplemented microcosm experiments without added Fe(III), partial removal of U(VI) from
solution occurred whilst with Fe(III)-augmented microcosms, a more complete removal to solids was
apparent early in the experiments. To further explore the mechanism(s) for metal reduction in these systems,
AQDS, an extracellular "electron shuttle", was added to this system and enhanced long term U(VI) removal
from solution was observed during the incubation (200 days). Finally, oxidative perturbations with air and
nitrate were explored in terms of radionuclide behaviour.
T=0
In the technetium microcosm experiments, partial removal of Tc(VII) occurred on development of
bioreduction in sediment systems without added Fe(III) and essentially complete Tc removal from solution
occurred in the ferrihydrite enriched systems. These results and the corresponding XAS spectroscopy data
for end-member samples will be discussed in the context of the microbial reduction of Fe(III), U(VI) and
Tc(VII) in high pH, geological disposal relevant systems.

PB4-7
COMPARISON OF MICROBIOLOGICAL INFLUENCES ON THE TRANSPORT AND
CHEMICAL PROPERTIES OF INTACT SANDSTONE AND ITS RELEVANCE TO
GEODISPOSAL
J. Wragg * , K. Bateman, A. E. Milodowski and J. M. West
British Geological Survey, Nicker Hill, Keyworth, Nottingham, UK, NG12 5GG
The impact of microbes on subsurface transport processes is widely recognised and has been linked to their
ability to form biofilms. Complex biogeochemical processes within and around the biofilm can promote
[1 – 3]
mineral dissolution and precipitation along flow paths, thus changing transport properties. This is
important when considering, for example, landfill and geological repositories for radioactive waste and for
carbon dioxide storage. In the UK, the choice of site, and therefore host geology for the deep geological
disposal facility (GDF) for low and intermediate level waste (L/ILW), has not yet been decided upon.
Consequently, some UK (bio)geochemical research programmes (e.g. NERC BIGRAD consortium, cf.
www.bigradnerc.com)utilisea “generic” rock type with an assemblage of minerals that could be considered
broadly similar to that expected to be found in a potential UK host rock. In this context, a carbonate, clay,
mica (illite)-bearing feldspathic sandstone from the Sherwood Sandstone Group, similar that used by the
BIGRAD project has also been used in recent BGS research for experiments on carbon dioxide storage, as
well as radioactive waste disposal, because this particular formation has the major potential as an aquifer for
carbon capture and storage in the UK.
This study compares the microbiological influences on the physico-chemical properties of Sherwood
Sandstone, potentially impacting on the transport properties and migration of contaminants from the two
geodisposal scenarios described above. These studies used intact rock material, relevant groundwater, a
common generic biofilm-forming bacteria (Pseudomonas aeruginosa) in bespoke flow-through experiments
at realistic pressure and temperature, with and without the presence of CO 2 . [4] The results can be summarised:
1. The soil organism, P. aeruginosa, can survive and thrive in pressurised flow through intact core
column experiments but although able to survive it did not thrive in the presence of CO 2 ;
2. Impacts on permeability were observed in experiments where no CO 2 was present. In these, biofilm
formation was seen together with changes in post inoculation pressure. Under comparable conditions
in the presence of CO 2 , no significant pressure changes were observed. However biofilm formation
was limited in the presence of CO 2 , suggesting that the CO 2 saturated fluid was impacting on the
ability of this particular microbe to form biofilm. A longer experimental time may be required to
allow acclimatisation to the conditions;
3. The effect on the chemistry of the two systems was distinctly different. The fluid chemistry in the
experiment without CO 2 showed a steady decline in element concentrations (Ba, Ca, K, Mo, W and
Zn) throughout - these changes do not appear to be a result of microbial activity. In contrast, in the
presence of CO 2 the effect was profound when compared to the elemental concentration of the
starting fluid. The injection of CO 2 appears to enhance the mobilisation of a number of chemical
species (e.g. Ba, Ca, Mn); and,
4. In both experimental set-ups – with and without CO 2 - changes in the bulk mineralogy were
minimal. Additionally, whilst biofilms formed on mineral surfaces, there was no evidence of
dissolution or alteration effects when compared to the starting materials.
These results suggest that, whilst microbial activity does not appear to impact on overall chemistry results,
their effects on physical transport properties in the geosphere can be profound in terms of fluid flow. The
inclusion of CO 2 saturated fluid impacts on both the ability of the microbes to alter permeability and the fluid
chemistry of the test system.

[1] P. Coombs, D. Wagner, K. Bateman, H. Harrison, A.E. Milodowski, D. Now, and J.M. West, Q J Eng Geol
Hydroge. 43, 131-139 (2010).
[2] A.L. Cunningham, B. Warwood, P. Sturman, K. Horrigan, G. James, J.W. Costerton and R. Hibert, The
microbiology of the Terrestrial Deep Subsurface, Eds P.A. Amy and D.L. Haldeman. 325-344 (1997)
[3] H.L. Ehrlich, Geomicrobiol J. 16, 135-153 (1999)
[4] J.Wragg, H. Harrison, J.M. West and H. Yoshikawa, Mineral Mag. 76, 3251-3259 (2012)
PB4-8
MICROBIAL DEGRADATION OF ISA UNDER HIGH PH CONDITIONS REPRESENTATIVE OF
INTERMEDIATE LEVEL WASTE
Naji Milad Bassil (1,2) , Nicholas Bryan (3) , Jonathan R. Lloyd (1)
1 Research Centre for Radwaste and Decommissioning & Williamson Research Centre for Molecular
Environmental Science, School of Earth, Atmospheric and Environmental Sciences, The University of
Manchester, Oxford Road, Manchester M13 9PL, UK
2 National Council for Scientific Research – Lebanon (CNRS-L), Beirut, Lebanon
3 Centre for Radiochemistry Research, School of Chemistry, The University of Manchester, Oxford Road,
Manchester, M13 9PL, UK
The United Kingdom has accumulated a substantial legacy of radioactive waste, most of which is in the form
of intermediate- and low-level waste, which contains substantial amount of cellulosic material in the form of
contaminated clothes, tissue, wood and paper. The Nuclear Decommissioning Authority (NDA) is
responsible for the safe disposal of this waste, and has proposed the conditioning of these materials in a
cement matrix in a deep cementitious geological facility. Upon resaturation of this facility, the cement
porewater will pass through three stages of evolution, starting at a pH of 13.3, followed by a progressive pH
drops to between 10 & 11 [1].
While the chemical and biological processes in and around this facility remain poorly understood, it is well
known that cellulose is unstable under high pH, and will degrade to short chain organic acids [2]. The
concentration of these degradation products are affected by temperature, pH and the nature of cellulose
present, however, the main degradation product of these reactions, at least during the early porewater
evolution stages, is isosaccharinic acid (ISA) [IUPAC: 2,4,5-Trihydroxy-2-(hydroxymethyl)pentanoic acid]
[2]. ISA is stable under these pH extremes and is able to complex with a number of radionuclides, including
Th(IV), U(IV), Pu(IV), and Np(IV), increasing their mobility [3].
A few published articles and reports have suggested the potential for ISA degradation by microorganisms.
Strand et al. found an Ancylobacter aquaticus species from an environmental sample able to grow at the
expense of ISA under aerobic conditions and neutral to high pH [4]. Bailey also identified two strains of
aerobic bacteria, able to degrade ISA under neutral to low pH [5]. Others have shown a biogeochemical
redox progression from aerobic respiration to Fe(III) reduction in sediments in the presence of ISA [6].
The aim of this work is to determine if microbes present in a high pH environmental sample are able to
degrade ISA under alkali conditions and use it as the sole carbon and energy source under different
biogeochemical conditions. Enrichment cultures were inoculated with a small volume of sediment, acquired
from a high pH lime workings site in the Peak District, UK. A minimal medium was used under aerobic,
nitrate-, Fe(III)- and sulphate-reducing conditions. Appropriate abiotic controls were also prepared and
successful enrichment cultures were subcultured several times.
Results show that bacteria from this environmental inoculum were able to (1) survive under high pH
conditions, and (2) degrade the added ISA under aerobic and nitrate-reducing and to a lesser extent Fe(III)-
reducing conditions. There was no ISA degradation under sulphate-reducing conditions, even after 60 days
of incubation. Collectively these data confirm that alkalitolerant or alkaliphilic bacteria are able to couple the
oxidation of ISA to the reduction of a range of terminal electron acceptors, including oxygen and nitrate.
Future work will use molecular biology techniques to identify the bacteria that are able to degrade ISA in
these samples, and characterise the physiological mechanisms underpinning these reactions. The impact of

these processes on radionuclide complexation and mobility will also be assessed, to help determine the
potential for such processes to control the transport of radionuclides in the near and far field around a
geological disposal facility containing intermediate level waste.
[1] Berner, U.R. 1992. Evolution of porewater chemistry during degradation of cement in a radioactive waste repository
environment. Waste Management 12: 201-219.
[2] Glaus, M.A., Van Loon, L.R. 2008. Degradation of cellulose under alkaline conditions: New insights from a 12
years degradation study. Environmental Science and Technology 42: 2906-2911.
[3] Gaona, X., Montoya, V., Colas, E., Grivé, M., Duro, L. 2008. Review of the complexation of tetravalent actinides
by ISA and gluconate under alkaline to hyperalkaline conditions. Journal of Contaminant Hydrology 102: 217-227.
[4] Strand, S.E., Dykes, J., Chiang, V. 1984. Aerobic microbial degradation of glucoisosaccharinic acid. Applied and
Environmental Microbiology 47: 268-271.
[5] Bailey, M.J. 1986. Utilization of glucoisosaccharinic acid by a bacterial isolate unable to metabolize glucose.
Applied Microbiology and Biotechnology 24: 493-498.
[6] Reinoso-Maset, E., Sidhu, S.H., Fisher, A., Heydon, A, Worsfold, P.J., Cartwright, A.J., Keith-Roach, M.J. 2006.
Effect of organic co-contaminants on Technetium and Rhenium speciation and solubility under reducing conditions.
Environmental Science and Technology 40: 5472-5477.
PB4-9
REDUCTION OF SELENITE BY BIOFILMS OF AN IRON-REDUCING BACTERIUM
Y. Suzuki, 1* H. Saiki, 1 A. Kitamura, 2 H. Yoshikawa 2
1
School of Bioscience and Biotechnology, Tokyo University of Technology,
Katakura-cho 1404-1, Hachioji, Tokyo, 192-0982, Japan
2 Geological Isolation Research and Development Directorate, Japan Atomic Energy Agency,
Muramatsu 4-33, Tokai, Naka, Ibaraki 319-1194, Japan
Selenium-79 is a long-lived radionuclide contained in high-level radioactive waste. It is important to
understand migration behavior of selenium at deep underground. Selenium solubility is largely controlled by
selenium oxidation state. Microbial reduction of highly soluble selenate (Se VI O 4 2- ) and selenite (Se IV O 3 2- ) to
insoluble elemental selenium is one of important phenomena affecting the mobility of selenium. The
microbial reduction of selenate and selenite have been widely studied using planktonic bacteria [1,2]. On the
other hand, it has been known that in natural settings bacteria are predominantly found within surfaceassociated
cell assemblages, or biofilms. However, there are little information on the interaction between
selenium and biofilms. In this study, we examined the reduction of selenite by biofilms of Shewanella
putrefaciens.
Biofilms of S. putrefaciens were made on circular cover glasses in a sterile 24-well microplate containing a 1
mL of nutrient broth in each well at 30°C. After a week incubation, biofilm like membranes were formed on
the cover glasses. The membranes were washed with a deionized water, stained with safranin solution, and
observed by an optical microscope. To investigate the reduction of selenite by the biofilms, a 1 mL of
aqueous solution containing 100 µM sodium selenite as an electron acceptor, 20 mM sodium lactate as an
electron donor, and 20 mM HEPES buffer (pH 7.0) was added to the biofilms formed on the cover glasses in
the microplate. The microplate was placed in an anaerobic plastic sack, where an oxygen absorbent was
placed to attain an anaerobic condition. After 34 h, the biofilms were washed with a deionized water, and
observed by the optical microscope without staining. The selenium concentration in the solution was
measured by ICP-AES. Se K-edge X-ray absorption near edge structure (XANES) spectra of the precipitates
appeared on the biofilms were collected in fluorescens mode at beamline 12C, Photon Factory, KEK
(Tsukuba, Japan). XANES spectra of sodium selenate, sodium selenite, and Se(0) were collected as reference
materials in transmission mode.
The microscopic observation of the cover glass after the incubation with S. putrefaciens revealed that the
cells were heterogeneously distributed on the cover glass indicating the formation of biofilms. After 34 h
incubation of the biofilms with selenite, the red precipitates were observed at the place where the biofilms
were formed. The precipitates were not dissociated from the biofilms by washing with a deionized water
indicating that they associated tightly with the biofilms. The selenium concentration in the solution was
under detection limit. The Se K-edge XANES spectrum of the red precipitates showed that the white line of

the spectrum appeared at the almost same energy of that of Se(0) powder indicating that the precipitates at
the biofilms were elemental selenium. These results suggest that the biofilms with iron-reducing bacteria in
the environment can strongly immobilize the selenite on the biofilms through selenite reduction to elemental
selenium.
[1] C. I. Pearce et al. (2009). “Investigating different mechanisms for biogenic selenite transformations: Geobacter
sulfurreducens, Shewanella oneidensis and Veillonella atypica.” Environ, Technol. 30: 1313-1326.
[2] A. Klonowska, T. Heulin and A. Vermeglio (2005). “Selenite and Tellurite Reduction by Shewanella oneidensis.”
Appl. Environ. Microbial. 71: 5607-5609.
PB4-10
MOLECULAR CHARACTERIZATION OF U(VI) ASSOCIATION WITH BACTERIA ISOLATED
FROM SPANISH CLAY
M. Lopez Fernandez 1)* , I. Sanchez Castro 1) , M. Romero-González 2) , A. Guenther 3) , P.L. Solari 4) , M.L.
Merroun 1)
1) Department of Microbiology, University of Granada, Granada, Spain
2) Cell-Mineral Research Centre, Kroto Research Institute, University of Sheffield, UK
3) Institute of Resource Ecology, Helmholtz-Zentrum Dresden Rossendorf, Dresden, Germany
4) Synchrotron SOLEIL, MARS Beamline, Paris, France
Clay deposits have been studied in recent years as potentially host rock for deep geological disposal of
radioactive wastes [1]. A reliable performance assessment of these systems depends on better knowledge on
the interactions of radionuclides with host rock natural microorganisms [2], [3]. Here, we present a combined
microscopic, spectroscopic, and microbiological approach to address the biomineralization of U(VI) by cells
of a bacterial strain under physiological conditions (using low phosphate growth medium, LPM, as
electrolyte). This strain, called BII-R6r, was isolated from Spanish clay. [4]. Bacterial cells were in contact
with a range of U(VI) concentration from 0.5mM to 4mM. The speciation of U(VI) in LPM medium has
been found to be complex as calculated using geochemical models. The dominant species are mainly
(UO 2 ) 2 CO 3 (OH) 3 - , (UO 2 ) 3 (OH) 5 + and (UO 2 ) 4 (OH) 7 + . Flow cytometry analyses showed that the strain BII-R6r
is resistant to high U concentrations. Infrared spectroscopy analyses indicated that bacterial phosphate
groups are involved in the coordination of U(VI). X-ray absorption spectroscopy (XAS) and time resolved
laser induced fluorescence spectroscopy (TRLFS) proved that uranium is removed from solution as U-
phosphate compounds with a similar structure to that of meta-autunite. The U precipitates were localized at
the cell surface and within the extracellular space, as demonstrated by TEM analysis. The biomineralization
of U(VI)-phosphates on the surface of the bacterium cell may be considered a detoxification mechanism of
the strain BII-R6r to overcome the toxicity of this radionuclide. This is the first study to reveal the U
biomineralization potential of clay natural bacteria under physiological conditions. In the geological disposal
context for the case of Spain, the presence of this microorganism in clay materials contributes to the
immobilization of radionuclides after the long term disposal conditions ended and the potential for natural
remediation of these sites.
[1] Villar et al., Journal of Iberian Geology, 32: 15-36 (2006).
[2] West et al., Interaction of Microorganisms with Radionuclides, Elsevier Science Ltd (2002)
[3] Merroun et al., Journal of Hazardous Materials, 197: 1-10 (2011)
[4] López-Fernández et al. Mineralogical Magazine, 75:3 1356 (2011)

PB4-11
BIOREMEDIATION OPTIONS FOR REMOVING URANIUM FROM GROUNDWATER
Laura Newsome 1 , Kath Morris 1 , Divyesh Trivedi 2 and Jonathan Lloyd 1,2
1 University of Manchester, School of Earth, Atmospheric and Environmental Sciences &
Williamson Research Centre for Molecular Environmental Science, University of Manchester,
Manchester, M13 9PL, UK.
Lead author’s email address laura.newsome@postgrad.manchester.ac.uk
2 National Nuclear Laboratory, Chadwick House, Birchwood Park, Warrington, WA3 6AE
Bioremediation may offer a cheap and effective method of treating radionuclide contamination in situ at UK
nuclear sites. Aqueous uranium can be removed from solution through: bioreduction where soluble U(VI) is
reduced to insoluble U(IV) by metal-reducing bacteria; and biomineralisation where U is co-precipitated
with, for example, phosphate. These techniques have been demonstrated to be effective in the laboratory;
however, little work has been done to determine whether microbial communities present at UK sites are
capable of catalysing these processes under realistic environmental conditions.
Seven sediment samples from Sellafield site boreholes chosen to represent different lithologies have been
characterised. Sediments were dominated by silicate minerals including quartz, feldspar and mica; some
contain calcite and pyrite. To determine whether the extant microbial communities could facilitate
bioreduction of U(VI), sediments were incubated in microcosms with artificial groundwater, an electron
donor of 5 mM acetate and 5 mM lactate, and 12 ppm U(VI) for over three months.
Uranium was completely removed from solution in five sediments (Figure 1), indicating an active microbial
community and the potential for bioreduction to be used as a remediation method at Sellafield. No changes
were observed in two sediments, perhaps due to low concentrations of bioavailable Fe(III) in the sediments,
or low numbers of active Fe(III)-reducing bacteria that are also capable of respiring U(VI), possibly due to
the length of time the materials were stored prior to use.
The long-term stability of bioreduced U(IV) minerals is now being investigated through exposure to
environmental oxidants including: nitrate, at concentrations to represent contamination in Sellafield
groundwater; and oxygen to represent conditions in the unsaturated zone. Results indicate that U(IV)
biominerals are susceptible to reoxidation via oxygen exposure, and are partially susceptible to reoxidation
by nitrate. In the nitrate experiments, sufficient remaining electron donor allows the microbial community to
reduce any reoxidised U(VI). No change in the stability of U(IV) biominerals was observed after 35 days
ageing. Future experiments will address the impact of U(IV) ageing over longer periods of time.
Current work focuses on examining a broad range of bioremediation options for UK nuclear sites, including
biostimulation and bioaugmentation strategies to promote the reductive immobilisation of key mobile
radionuclides on site, alongside alternative biomineralisation strategies.
U(IV) (s)
U(VI) (aq)
Figure 1 – Bioreduced U(IV) microcosm compared to no electron donor control

PB4-12
THE HYDROGEN DRIVEN GEOMICROBIOLOGY OF CEMENTITIOUS NUCLEAR WASTE
M.J.C. Crouch 1) , K. Morris 1) , D. Engelberg 2) , J. Small 3) , J.R. Lloyd 1)
1) School of Earth, Atmospheric and Environmental Science, Research Centre for Radwaste &
Decommissioning and Williamson Research Centre for Molecular Environmental Sciences,
University of Manchester, Manchester,M13 9PL, UK
2) Materials Science Centre, School of Materials, University of Manchester, Manchester,M13 9PL, UK
3) UK National Nuclear Laboratory, Risley, Warrington, WA3 6AE, UK
The construction and operation of the proposed Geological Disposal Facility (GDF) for the UK’s higher
activity nuclear waste legacy [1] poses many questions relating to the actions of native microbes within a
repository. Fe(III)-reducing bacteria are capable of reducing a number of the long-lived “Priority
radionuclides” (e.g. Np(V), Tc(VII) and U(VI)) of most concern to the safe operation of a GDF for the
requisite >100 000a -1 [2]. Microbial reduction of the aforementioned priority radionuclides occurs from the
soluble +5, +6 or +7 redox states to the less soluble +4 state. Reduction will thus retard the mobility of these
radionuclides within a GDF.
The use of cement within a GDF in the form of concrete vault backfill and as a cementitious grout used to
package Intermediate Level Waste (ILW), will create a high pH environment in the near-field. Reactions
between most radionuclides and high pH water will render these radionuclides less soluble. Concrete also
provides sorption sites for radionuclide binding [3]. Moreover; the use of cement and sealing of the GDF
will create an ecological niche favourable towards alkaliphilic, anaerobic microorganisms.
An issue of key concern to the GDF safety case is that of hydrogen production largely from the corrosion of
iron within the steel waste drums, concrete reinforcing rods, rock bolts etc. [4, 5]. Excess hydrogen in a
repository could cause seals to rupture, causing radioactive groundwater to migrate more quickly [4].
Microbial metabolism of hydrogen [6, 7] has the potential to offset the effects of hydrogen pressurisation
and also promote the enzymatic reduction of metals and radionuclides. The focus of the work described in
this poster is the potential for the microbial oxidation of hydrogen under high pH conditions representative
of intermediate level waste forms.
[1] DEFRA et al. (2008) Managing Radioactive Waste Safely: A Framework for Implementing Geological Disposal,
Norwich: TSO (2008).
[2] Lloyd, J.R., Microbial reduction of metals and radionuclides. FEMS Microbiol. Rev., 27(2-3), 411 (2003).
[3] Crossland, I.G. and Vines, S.P., Why a cementitious repository?, Didcot: United Kingdom Nirex Limited (2001).
[4] Geological Disposal Gas status report, Didcot: Nuclear Decommissioning Authority (2010).
[5] Williamson, A.J., Morris, K., Shaw, S., Byrne, J.M., Boothman, C. and Lloyd, J.R., Microbial reduction of Fe(III)
under alkaline conditions relevant to geological disposal. Accepted by Appl. Environ. Microbiol. (2013).
[6] Lovley, D.R., Goodwin, S., Hydrogen concentrations as an indicator of the predominant terminal electron-accepting
reactions in aquatic sediments, Geochim. Cosmochim. Acta 52, 2293 (1988).
[7] Libert, M., Bildstein, O., Esnault, L., Jullien, M. and Sellier, R., Molecular hydrogen: An abundant energy source for bacterial
activity in nuclear waste repositories, Phys. Chem. Earth 36, 1616 (2011).

PB6
PB6-1
PB6-2
NATURAL ANALOGUES
RADIUM NATURAL ANALOGUE REQUIREMENTS FOR ROBUST SAFETY CASE
DEVELOPMENT
W.R. Alexander, I.G. McKinley, E.M. Scourse, S.M.L. Hardie, E. Klein (Switzerland, UK)
NATURAL ANALOGUES FOR RADIONUCLIDES
J.S. Zakharova, A.D. Wheatley, O. Voitsekhovich (UK, Ukraine)
PB6-1
RADIUM NATURAL ANALOGUE REQUIREMENTS FOR ROBUST SAFETY CASE
DEVELOPMENT
W.R. Alexander (1) , I.G. McKinley (2) , E.M. Scourse (3) , S.M.L. Hardie (2) and E. Klein (2)
(1) Bedrock Geosciences, Auenstein, Switzerland
(2) MCM Consulting, Baden-Dättwil, Switzerland
(3) MCM Consulting, Bristol, UK
226 Ra has been identified as a safety-critical radionuclide in many safety assessments [e.g. 1] and recent
reviews [e.g. 2] have recommended that future effort is invested to better support assumptions on the
behaviour of Ra in the engineered barriers, the geosphere and biosphere. In particular for spent fuel, failure
of a long-lived overpack can result in a pulse release of 226 Ra which has grown in from 238 U (at a rate
determined by the half-life of 234 U). Such a release may be constrained by solubility, controlled by either a
pure Ra phase (e.g. RaSO 4 (s)) or co-precipitation as a trace component within minerals such as BaSO 4 (s)
[e.g. 3]. Aqueous phase transport towards the biosphere may be retarded by sorption and further precipitation
processes or facilitated by uptake onto colloids or complexation with mobile organic species. Further dilution
or re-concentration processes in the biosphere can have a large effect (orders of magnitude) on the resultant
dose from this radionuclide and its short-lived daughters.
Ongoing laboratory or URL projects inevitably involve relatively short-term studies with simplified subsystems
and are further complicated by the high radiotoxicity of 226 Ra. However, Ra is not only ubiquitous in
the natural environment, but is also associated with many human industrial activities, with reasonably welldefined
sources in some cases. Despite the fact that analogue data must be treated with great caution when
attempting to derive quantitative data on solution/solid interactions [4, 5], careful tailoring of the analogues
studied to identify key open issues can provide the support needed to provide safety case robustness [6].
Because of the synergies involved, there are advantages in initiating a comprehensive analysis of Ra
partitioning in all relevant repository system components. For example:
• Bentonite (with or without alteration by concrete)
• EDZ crystalline and argillaceous rocks (with or without alteration by concrete)
• Flow paths through undisturbed crystalline and argillaceous rocks
• Soils and biota
• Fresh, brackish and marine sediments
Such a study can utilise the 2 most common naturally-occurring isotopes of Ra (-226 and -228) and use a
range of direct isotope measurements and laboratory studies of selected samples to characterise Ra
distribution between phases:
• that may be mobile (true solution, organic complexed, colloidal, mobile microbes)
• that are immobile (co-precipitated and sorbed on minerals, trapped in matrix pores, encapsulated on
biofilms)

The paper will outline how such a project could identify appropriate locations where natural partitioning of
Ra can be quantified and a range of sites will be examined with the pros and cons of each site assessed and
evaluated. Finally, the possibility of incorporating the produced data, in an integrated manner with
modelling, laboratory and URL data [cf. 7], will be assessed and a project road map to this end laid out.
[1] SKB (2011) Long-term safety for the final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site
project. Main report of the SR-Site project Volumes I-III. SKB TR-11-01, SKB, Stockholm, Sweden.
[2] NEA (2012) The post-closure radiological safety case for a spent fuel repository in Sweden. An international peer
review of the SKB license-application study of March 2011 (Final report) NEA Report NEA/RWM/PEER (2012)2,
NEA/OECD, Paris, France.
[3] U. Berner (2002). Project Opalinus Clay: Radionuclide Concentration Limits in the Near-Field of a Repository for
Spent Fuel and Vitrified High-Level Waste, PSI Report No. 02-22, PSI, Villigen, Switzerland.
[4] I.G. McKinley and W.R. Alexander (1993) Assessment of radionuclide retardation: uses and abuses of natural
analogues. J. Contam. Hydrol. 13, pp249-259.
[5] I.G. McKinley and W.R. Alexander (1996). On the incorrect derivation and use of in situ retardation factors from
natural isotope profiles. Radiochim Acta 74, pp 263-267.
[6] W.R. Alexander, I.G. McKinley and H. Kawamura (2013). The process of defining an optimal natural analogue
programme to support national disposal programmes. Proc. NEA-GRS Workshop on natural analogues for Safety Cases
of repositories in rock salt. 4 – 6 September, 2012, Braunschweig, Germany. NEA/OECD, Paris, France (in press).
[7] W.R. Alexander, A. Gautschi and P. Zuidema (1998). Thorough testing of performance assessment models: the
necessary integration of in situ experiments, natural analogues and laboratory work. Extended abstract in Sci. Basis
Nucl. Waste Manag. XXI, 1013-1014.
PB6-2
NATURAL ANALOGUES FOR RADIONUCLIDES
J. S. Zakharova 1) , A. D. Wheatley 1) , O. Voitsekhovich 2)
1)
School of Civil & Building Engineering, Loughborough University, Epinal Way/Ashby Road,
Loughborough, UK, LE11 3TU.
2) Environment Radiation Monitoring Department, Ukrainian Hydrometeorological Institute, Nauki, 37,
Kyiv, Ukraine, 03028
This paper reports on the idea of using urban metallic pollutants to model the environmental fate of
radionuclides. A potential way to help evaluate and provide evidence on pollution mobility is to carry out
studies on natural analogues. This paper discusses the possible factors affecting radionuclides (uranium and
other U-Th series) mobility, their adsorption and precipitation properties in the conditions of the uranium
tailings dumps, subject to different groundwater flows and rainfalls. This will be gained from complementary
work on the metal removal (Fe, Cu and Zn) from urban runoff using hydraulic interceptor and novel
adsorption surfaces such as biochar (hydrothermal carbonisation of waste organics) and thermally expanded
clays. The impact of temperature, pH, redox, conductivity, organic matter, hardness, particle size and loads
on metal solubility are reported. Understanding and taking these fundamental parameters into consideration
will provide a model to predict radionuclide behaviour, transport and treatment efficiencies. In many natural
environments these complex and antagonistic cross reactions make these basic predictions more difficult.
Few researchers have reported on field data exploring these fundamental interactions which could vary with
location. There is also little previous work attempting establish links between EMC (Event Mean
Concentration), temperature and evaporation with seasonal rainfall characteristics (rainfall intensity and
ADWP (Antecedent Dry Weather Period).
The experimental data is from the analysis of zinc, copper and iron from one of the UKs major roads (M1
J24, peak traffic flow 30,000 vehicles an hour). It then describes for discussion proposed research work on
the fate of radionuclides from one of the Ukraine’s largest uranium production facilities, Pridniprovsky
Chemical Plant (PChP), which was in operation from 1948 to 1991. During operation, nine tailings dumps
were created containing about 42 million tonnes of ore tailings and residues including radioactive waste with
a total activity estimated at about 4x1015 Bq. Five of the tailings dumps are located within the industrial
zone of Dneprodzerzhinsk which is located close to inhabited areas. Run off from the tailings could cause
environmental impacts that extend the risks of exposure to the public living in the local River Dnieper
catchment and contaminate groundwater.

The data from the UK trunk road suggest that pollutants accumulate in stagnant drainage water and are then
diluted and mobilised by rainfall runoff from the site. The solubility of the metal is the most important
mobility factor other than storm characteristics. Analysis of samples from a rural brook confirmed the link to
solubility. Iron in the control brook was bound to the suspended solids and resuspended by increases flow
rates. Copper, on the other hand, was soluble and usually diluted by an increase in flow rate. The research
also showed that solubility and attachment to potential adsorbents (clay, silt, gravel, biomass) was influenced
by other water quality parameters (particle size, pH and total salt content (de-icer).
The full paper will provide more detail information concerning sampling sites (M1 J24; control brook
(Woodbrook) as well as the tailings dumps together with sampling procedure and analysis. Table 1 presents
example data obtained from the M1 during 2 monitored rainfall events compared to a dry weather period in
the same season.
Table 1 Comparison of two rainfall events with dry weather samples
Pollutant,
mg/l
EMC,
mg/l
Rainfall Event 1 (7 samples) Rainfall Event 2 (8 samples) Dry weather (10 samples)
Range
Max
observed
load, g
EMC,
mg/l
Range
Max
observed
load, g
Average
conc,
mg/l
Range
TSS 41.98 6-119.5 102.0 9.44 9-14 89.0 112 12-231.8
TOC 7.55 5.85-8.66 22.6 5.52 3.19-10.31 54.7 11.1 6.95-15.538
Fe tot 1.523 0.659-3.76 3.34 0.679 0.632 – 0.728 6.41 1.934 0.597 – 5.89
Fe dis 0.194 0.07-0.94 0.21 0.073 0.068 – 0.143 0.66 0.058 0.008 – 0.086
Zn tot 0.121 0.081-0.273 0.28 0.092 0.068 – 0.102 0.92 0.117 0.086 – 0.264
Zn dis 0.026 0.024-0.028 0.08 0.052 0.038 – 0.056 0.51 0.026 0.005 – 0.141
Na 57.28 55.8 – 60.49 164 75.56 59.37 – 99.43 878 50.65 22.09 – 164.55
One of the important features of this data is that the values for some pollutants (TSS, Zn tot ) during storms are
comparable with dry weather samples and in some cases are higher during rainfall (TSS, Fe tot , TOC). It can
be explained by interaction of first flush and dilution effect during the rainfall and the accumulation of
pollutants in the stagnant water by evaporation.
The second notable point to evaluating long-term pollutant inputs (described in more details in the full paper)
is that metal solubility increases over time and is influenced by other water quality parameters. Comparing
the proportion of dissolved metals from the M1 and Woodbrook (Table 2), they are different throughout the
seasons and weather conditions. Fe was mainly found in solids which increased in proportional with the
solids during wet weather.
Table 2 Solubility of metals
Solubility of metals, %
Months
April May Jun Jul Nov Jan
Fe M1 10.47 12.55 38.17 38.89 31.07 12.27
6.67 * 14.19
Woodbrook 50.27 29.29 16.8 3.43 15.38 12.09
13.77 1.08
Cu M1 99.95 21.95 20.69 21.43 99.96 traces
traces
traces
Woodbrook traces 16.67 36.84 traces 40 66.67
99.92 3.57
Zn M1 26.83 86.25 29.82 28 15.16 traces
21.66 16.26
Woodbrook 48.72 28 47.73 traces 7.9 43.59
73.42 11.86
* Figures in bold indicate wet weather period

PC1
PC1-1
PC1-2
PC1-3
PC1-4
DATA SELECTION AND EVALUATION
SORPTION VALUES IN SORPTION DATA BASES FOR ARGILLACEOUS ROCKS
AND BENTONITE: A COMPARISON BETWEEN DERIVED AND MEASURED
VALUES
B. Baeyens, M. Marques Fernandes, M.H. Bradbury (Switzerland)
RADIONUCLIDE SOURCE TERM ESTIMATIONS FOR THE PRELIMINARY SAFETY
ASSESSMENT GORLEBEN (VSG)
M. Altmaier, B. Kienzler, Ch. Bube, V. Metz, H. Geckeis (Germany)
AN ATEMPT TO SELECT THERMODYNAMIC DATA AND TO EVALUATE THE
SOLUBILITY OF RADIOELEMENT WITH UNCERTAINTY UNDER HLW DISPOSAL
CONDITIONS
T. Yamaguchi, S. Takeda, Y. Nishimura, Y. Iida, T. Tanaka (Japan)
THEREDA REVISITED - WHAT HAPPENED SO FAR
V. Brendler, F. Bok, M. Altmaier (Germany)
PC1-1
SORPTION VALUES IN SORPTION DATA BASES FOR ARGILLACEOUS ROCKS AND
BENTONITE: A COMPARISON BETWEEN DERIVED AND MEASURED VALUES
B. Baeyens, M. Marques Fernandes, M.H. Bradbury
Paul Scherrer Institut, Laboratory for Waste Management, Villigen PSI, Switzerland
In Switzerland, four argillaceous rock types have been selected as being suitable host rocks for high-level
(HLW) and the low- and intermediate-level (L/ILW) radioactive waste repositories: Opalinus Clay (HLW,
L/ILW) and 'Brauner Dogger', Effingen Member and Helvetic Marls (L/ILW) [1]. Sorption data bases for all
of these host rocks are required for safety analyses, including all of the bounding porewater and
mineralogical composition combinations.
An in-house methodology has been developed for deriving the sorption values for radionuclides at trace
concentrations in sorption data bases (SDBs) for argillaceous rocks and bentonite [2]. The main factors
influencing the sorption in such systems are the phyllosilicate mineral content, particular the 2:1 clay mineral
content (illite/smectite/illite-smectite mixed layers), assumed to be the major sorbing phases, and the
porewater chemistry which determines the radionuclide species in the aqueous phase. The source data used
in this procedure are radionuclide sorption edge measurements, i.e. trace radionuclide sorption as a function
of pH at a fixed ionic strength, on illite and montmorillonite. A series of so called conversion factors are
applied which take into account the 2:1 clay mineral contents and the different radionuclide speciations in
the different porewaters.
Confidence in the validity and correctness of this methodology has been built up in some previous studies.
Firstly, sorption values obtained by the methodology described above have been compared with those in
already existing SDBs for Opalinus Clay and bentonite [3] used in Project Opalinus Clay [4]. Secondly,
blind sorption model predictions of isotherms on Opalinus Clay and MX-80 bentonite in realistic porewater
compositions have been compared with measured sorption isotherms [5]. In all cases the results obtained in
the different comparative studies confirmed and supported the approach in the new in house methodology.
This work reports a third step in providing evidence that the methodology for deriving SDBs as proposed in
[2] is robust and reliable. The aim was to compare the sorption values predicted at trace concentrations with
those values taken from isotherms measured under realistic porewater chemistries for the host rocks being
considered and for MX-80 bentonite for metals with valences from I to VI. The metals chosen were
Co(II)/Ni(II), Eu(III) and Th(IV) as representative for bivalent transition metals, trivalent

lanthanides/actinides and tetravalent actinides, respectively. In addition the elements Cs(I), Np(V) and U(VI)
were included in this study 1 .
The results of this exercise are shown in Figure 1 where the predicted R d values are plotted as function of the
measured R d values. The continuous line represents a 1:1 correspondence between the measured and the
predicted data. Data points falling below or above this line indicate under or over predicted values,
respectively. The majority of the plotted data lies within ±0.6 log units of the 1:1 line. For comparison,
typical errors on R d measurements lie between ±0.2 and ±0.5 log units for medium to strongly sorbing
elements.
Within the scope of this study, and supported by previous work [4, 5], it has been demonstrated that the
procedures developed for deriving SDBs for use in the safety analyses of radioactive waste repositories can
be applied with confidence to a wide variety of different types of argillaceous rocks/bentonite and porewater
chemistry combinations. Thus, the source data sets needed to generate such SDBs are radionuclide sorption
edge measurements on illite and montmorillonite and not specific measurements on the argillaceous
rocks/bentonites themselves in the range of porewater chemistries of interest.
Figure 1. Overview of the comparison of measured and predicted sorption data for Cs(I), Co(II), Ni(II),
Eu(III), Th(IV) and U(VI) at trace concentrations on different argillaceous rock types and
MX-80 bentonite.
[1] Nagra (2008). “Begründung der Abfallzuteilung, der Barrierensysteme und der Anforderungen an die Geologie.
Bericht zur Sicherheit und technischen Machbarkeit.” Nagra Tech. Rep. NTB 08-05
[2] Bradbury, M.H., Baeyens, B. and Thoenen, T. (2010): Sorption data bases for generic Swiss argillaceous rock
systems. Nagra Tech. Rep. NTB 09-03
[3] Bradbury, M.H. and Baeyens, B. (2010). Comparison of the reference Opalinus Clay and MX-80 bentonite sorption
data bases used in the Entsorgungsnachweis with sorption data bases predicted from sorption measurements on illite
and montmorillonite. PSI Bericht Nr. 10-09, Nagra Tech. Rep. NTB 09-07
[4] Nagra (2002). Project Opalinus Clay: Safety Report. Demonstration of disposal feasibility (Entsorgungsnachweis)
for spent fuel, vitrified high-level waste and long-lived intermediate-level waste. Nagra Tech. Rep. NTB 02-05
[5] Bradbury, M.H. and Baeyens, B. (2011). “Predictive sorption modelling of Ni(II), Co(II), Eu(III), Th(IV) and U(VI)
on MX-80 bentonite and Opalinus Clay: A "bottom-up" approach.” Appl. Clay Sci. 52: 27-33

PC1-2
RADIONUCLIDE SOURCE TERM ESTIMATIONS FOR THE PRELIMINARY SAFETY
ASSESSMENT GORLEBEN (VSG)
M. Altmaier, B. Kienzler, Ch. Bube, V. Metz, H. Geckeis
Institute for Nuclear Waste Disposal, Karlsruhe Institute of Technology, Germany
A Preliminary Safety Assessment (VSG) for the potential German waste disposal site Gorleben was
performed during the past three years. VSG was funded by the German Federal Ministry for the
Environment, Nature Conservation and Nuclear Safety (BMU) and coordinated by the Gesellschaft für
Reaktorsicherheit (GRS). Within this large collaborative project, several institutions were contributing
expertise and addressing key questions related to Gorleben salt dome as potential nuclear waste disposal site
for high level radioactive waste. More detailed information on the Preliminary Safety Assessment Gorleben
is available on the web-sites of BMU or GRS (http://www.gorlebendialog.de/hintergrund/vorlaeufigesicherheitsanalyse/
or http://www.grs.de/en/content/gorleben, in German or English). One activity of the
Institute for Nuclear Waste Disposal (INE) at the Karlsruhe Institute of Technology (KIT) within VSG was
focused on deriving radionuclide source terms for radioactive heat producing waste. Two recent publicly
available reports on the radionuclide source term estimations by KIT-INE within the VSG project have been
published at KIT Scientific Publishing 1,2 . They are available from the authors or can be downloaded at
http://www.ksp.kit.edu.
Source terms for potential radionuclide mobilisation from high level waste glass, spent nuclear fuel,
compacted hulls and end-pieces (CSD-C waste) were derived for relevant geochemical conditions of a
nuclear waste repository in the Gorleben salt dome. Further source terms were derived for disposal concepts
including irradiated fuel from prototype, research and education reactors, for waste forms with negligible
heat generation and for uranium tails. For the estimation of radionuclide source terms different scenarios
were adopted which consider NaCl-, MgCl 2 -, and in specific cases CaCl 2 -rich brines in the near-field of the
waste products.
The kinetic and thermodynamic control of radionuclide mobilization and retention and the influence of
temperature on these processes were discussed and analyzed for the adopted scenarios and geochemical
conditions. The maximum expected solubility concentrations of the radionuclides Am, Th, U, Np, Pu, Tc, Zr
and several rare-earth elements were quantified within simplified scenarios for carbonate-free solutions
without explicit treatment of temperature effects. With the exception of adsorption to canister materials,
radionuclide retention on the currently envisaged backfill materials is of minor importance.
Considering a future comprehensive safety assessment for a repository in a rock salt formation, several
recommendations (i.e. on optimized design, layout and backfill concepts) were derived from the perspective
of improved and robust radionuclide source terms. Within the Preliminary Safety Assessment Gorleben
(VSG), KIT-INE has critically analyzed the need for targeted research activities related to (i) relevant open
geochemical issues, (ii) improved process understanding especially for actinide chemistry, and (iii) reliable
thermodynamic description and modeling.
[1] B. Kienzler, M. Altmaier, Ch. Bube and V. Metz, Radionuclide Source Term for HLW Glass, Spent Nuclear Fuel,
and Compated Hulls and End Pieces (CSD-C Waste), KIT Scientific Reports 7624, (2012), KIT Scientific Publishing,
Karlsruhe, Germany.
[2] B. Kienzler, M. Altmaier, Ch. Bube and V. Metz, Radionuclide Source Term for Irradiated Fuel from Prototype,
Research and Education Reactors, for Waste Forms with Negligible Heat Generation and for Uranium Tails, KIT
Scientific Reports 7635, (2013), KIT Scientific Publishing, Karlsruhe, Germany.

PC1-3
AN ATTEMPT TO SELECT THERMODYNAMIC DATA AND TO EVALUATE THE
SOLUBILITY OF RADIOELEMENT WITH UNCERTAINTY UNDER HLW DISPOSAL
CONDITIONS
T. Yamaguchi 1) , S. Takeda, Y. Nishimura, Y. Iida, T. Tanaka
1) Waste Safety Research Group, Nuclear Safety Research Center, Japan Atomic Energy Agency
2-4, Shirakata, Tokai, Ibaraki 319-1195, Japan
An attempt was made to select thermodynamic data with uncertainties and to evaluate the solubility of
radioelements with uncertainties considering variation in groundwater chemistry. The thermodynamic data
were selected by reviewing the JAEA-TDB released in 2012. Data for Nb, Pd and Pa were revised from the
viewpoint of the consistency of the data selection process. Data for Se, U and Pa were revised from the
viewpoint of conservativeness. Up-to-date ternary calcium-metal(IV)-OH complexes were adopted for Zr,
Th, U, Np and Pu. A Monte Carlo-based calculation code, PA-SOL, was used for probabilistic analysis of
the solubility.
Solubility of radioelements is a prioritized parameter used in the analysis of radiological consequences
through groundwater migration scenarios for geological disposal of high-level radioactive waste. The
solubility is evaluated based on thermodynamic data. The major influential factor would be the groundwater
chemistry such as pH, redox potential, and concentrations of aqueous ions. The groundwater chemistry is
variable with time and with space. The thermodynamic data are accompanied by uncertainties. In order to
improve our confidence in the solubility evaluation, it is necessary to evaluate the variations in the estimated
solubility of radioelements due to the variation in the groundwater chemistry and uncertainties of the
thermodynamic data and to quantify the uncertainties in the solubility as an assessment parameter. In this
study, a thermodynamic database [1] was reviewed to select the thermodynamic data with uncertainty to be
used in our solubility evaluation. Then a probabilistic solubility analysis was performed based on a
hypothetical variation in groundwater chemistry by using the selected thermodynamic data. The results help
us specify prioritized parameters to be further studied.
Thermodynamic data selection
Thermodynamic data for Se, Zr, Nb, Tc, Pd, Sn, Pb, Th, Pa, U, Np, Pu and Am used in our probabilistic
safety assessment were selected by reviewing a thermodynamic database [1]. The points of the review were:
- Consistency of the data selection process (Nb, Pd, Pa).
- Data selection and exclusion to prevent underestimation of the solubility (Se, U, Pa).
- Selecting up-to-date data (Zr, Th, U, Np, Pu).
Se: Data for FeSe 2 (cr) were excluded from the dataset used in our solubility estimation.
Zr: Data for Ca 2 [Zr(OH) 6 ] 2+ and Ca 3 [Zr(OH) 6 ] 4+ [2] were added.
Nb: 3Nb 2 O 5 (s) + 4H 2 O = Nb 6 O 19 H 3 5- + 5H + logK 0 = -49.1±2.4 was added for inner consistency.
Pd: Data for Pd(cr) were excluded from the dataset used in our solubility estimation for inner consistency.
Th: Data for Ca 4 [Th(OH) 8 ] 4+ [3] were added.
Pa: Data for PaO(OH) 3 (aq) were excluded from the dataset used in our solubility estimation for inner
consistency (Data for Pa(OH) 5 (aq) are used). Data for PaO(CO 3 ) 5 7- , PaO(CO 3 ) 4 5- and PaO(CO 3 ) 2 (OH) 2
3-
were estimated based on the chemical analogy between Pa(V) and U(IV).
U: Data for USiO 4 (coffinite) were excluded from the dataset used in our solubility estimation. Data for
Ca 4 [U(OH) 8 ] 4+ [3] and U(OH) 6 2- [4] were added for conservativeness.
Np, Pu: Data for Ca 4 [Np(OH) 8 ] 4+ and Ca 4 [Pu(OH) 8 ] 4+ [3] were added.
Tc, Sn, Am: Data compiled by Kitamura et al.[1] were used without any revision.
Solubility estimation
We have developed a Monte Carlo-based calculation code, PA-SOL by incorporating the probabilistic
calculation scheme into widely-used geochemical calculation codes, EQ3/6 and PHREEQC. This PA-SOL
generates large number of parameter sets from the probabilistic distribution of each parameter by Latin
Hypercube random Sampling (LHS). SPOP code [5] is used for stochastic analysis of the calculated
solubilities to generate CDF (Cumulative Distribution Function), CCDF (Complementary Cumulative

Distribution Function) and PRCC (Partial Rank Correlation Coefficient. The result of the probabilistic
analysis of the solubility will be presented and discussed at the Conference.
The authors acknowledge Dr. A. Kitamura for helpful discussion.
[1] A. Kitamura et al., JAEA-Data/Code 2012-006 (2012).
[2] M. Altmaier et al., Radiochim. Acta, 96, 541-550 (2008).
[3] D. Fellhauer et al., Radiochim. Acta, 98, 541-548 (2010).
[4] K. Fujiwara et al., Radiochim. Acta, 93, 347-350 (2005).
[5] T. Homma & T. Kasahara, JAERI-M 93-207, Japan Atomic Energy Agency (1993) [in Japanese].
* Part of this study was funded by the Secretariat of Nuclear Regulation Authority, Nuclear Regulation
Authority, Japan.
PC1-4
THEREDA REVISITED – PROJECT STATUS IN 2013
V. Brendler (1 ), F. Bok (1) , Ch. Marquardt (2) , M. Altmaier (2)
(1) Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology, PO Box 510119,
D-01314 Dresden, Germany
(2) Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal, PO Box 3640,
D-76021 Karlsruhe, Germany
Safety analysis for a geological repository for radioactive waste as well as remediation measures for uranium
mining and processing legacies share an essential: the need for a reliable, traceable and accurate assessment
of potential migration of toxic constituents into the biosphere. The respective computational codes require
site-independent thermodynamic data concerning aqueous speciation, solubility limiting solid phases and
ion-interaction parameters. Such databases, however, show several constraints:
- Incompleteness in terms of major and trace elements
- Inconsistencies between species considered and corresponding formation constants
- Restricted variation ranges of intensive parameters (temperature, density, pressure)
- Limitations with respect to solution compositions (ionic strength)
To overcome these limitations to a significant degree, an ambitious database project – THEREDA – has been
launched in 2006 by institutions leading in the field of safety research for nuclear waste disposal in Germany
[1]:
- Gesellschaft für Anlagen- und Reaktorsicherheit mbH, Braunschweig
- Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology
- Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal
- TU Bergakademie Freiberg, Institute of Inorganic Chemistry
- AF-Consult Switzerland AG, Baden (Switzerland)
The main objective is a centrally administrated and maintained database of verified thermodynamic
parameters for environmental applications in general and radiochemical issues in particular.
This work focuses on the achievements within THEREDA since its previous presentation at the Migration
conference 2011 in Beijing. The most important point is the official release of four more datasets (as of
March 2013; in bold: additions to the hexary system of oceanic salts), all based on the Pitzer model
describing the ion-ion interactions. Additional releases of thermodynamic data for Th(IV), U(IV) and U(VI)
are planned for the near future.
2. Na, Mg, Ca – Cl – Am(III), Nd(III), Cm(III) – H 2 O(l) (2012-05-30)
3. Na, K, Mg, Ca – Cl, SO 4 – HCO 3 /CO 2 (g) – H 2 O(l) (2012-02-21)
4. Na – Cl – Np(V) (2012-12-03)
5. Na, Mg, Ca, K – Cl, SO 4 – HCO 3 /CO 2 (g) – Cs(I) – H 2 O (2013-01-28)

They can all be downloaded as separate files from the project web site www.thereda.de (navigation menu:
THEREDA Data Query Tailored Databases) as generic ASCII type, and in formats specific to the
geochemical speciation codes PhreeqC, EQ3/6, ChemApp and Geochemist’s Workbench. Moreover, access
to data records is now also possible through interactive forms (menu: THEREDA Data Query Single Data
Query // Complex Systems), both with export options as CSV or MS Excel file.
In connection with these data releases, several other measures have been successfully implemented in
THEREDA. Most important from a work efficiency point of view is the new interactive web-based tool for
data entry and editing – though only visible to those permitted to maintain data records, i.e. members of the
editorial board of THEREDA. This tool incorporates also a variety of internal checks for data consistency
and plausibility.
To illustrate the capability of THEREDA the solubility of freshly precipitated Nd(OH) 3 (am) in 3.86 m CaCl 2
solution was calculated and compared to predictions based on the data0.ypf.R1/2 databases from the Yucca
Mountain Project (Fig. 1). Differences in the weak alkaline range might be caused by the differences in the
solubility constants of the Nd(III) hydroxides (ΔlogK Nd(OH)3(am) ≈ 3.3). The missing of the increase of Nd(III)
solubility in higher alkaline range (−log(mH + ) > 10) using the data0.ypf.R1/2 databases is a consequence of
different aqueous speciation, especially for anionic and ternary Ca-Nd-(OH) x complexes. This is of high
relevance as lanthanides regularly serve as analogues for actinides of the same valency type.
0.01
1E-3
THEREDA R2
data0.ypf.R1
data0.ypf.R2
NEC/ALT2009
Nd(OH) 3 (am)
Nd(OH) 3 (cr) suppressed
m(Nd(III)) [molal]
1E-4
1E-5
1E-6
Nd(OH) 3 (am)
Nd(OH) 3 (cr)
1E-7
1E-8
8.5 9.0 9.5 10.0 10.5 11.0 11.5 12.0
-log(mH + )
Fig. 1: Solubility of amorphous/crystalline Nd(OH) 3 in 3.86 m CaCl 2 solution (lines: databases used in
calculations, dots: measured values from literature [2]).
The integration of a new auditing scheme is also a major milestone. It provides an independent measure to
monitor data manipulations and is an essential component of the quality assurance (QA) strategy within
THEREDA. Moreover, QA benefited from an automatic logging system storing all data changes in separate
files, and from the extension of the internal calculation scheme for mutually dependent thermodynamic data.
Another significant improvement to THEREDA was the establishment of a series of benchmark calculations
– all being openly accessible to the public. Each benchmark addresses a specific data release, has its own
documentation file, and covers as many speciation codes as possible. The benchmarks are intended to
monitor whether changes in the database (or the export parsers) affect the results of speciation calculations.
They also serve to detect deviations between the various codes when fed with an identical data input. Test
subjects are the concentrations and pH values of well-defined multiple-salt points or solubility curves of
solids.
The last important advance is the integration of sorption data – namely for the parameterization of surface
complexation models (SCM) – into the existing database management system. The following models will be
supported Diffuse Double Layer Model (DDL), the Constant Capacitance Model (CC), the Non-electrostatic
Model (NE), and the 1pK-Basic Stern Model (1pK-BS). Most of them are implemented in a variety of
speciation codes, namely by PHREEQC and Geochemist’s Workbench (both supported by THEREDA). For
each mineral, the specific surface area (A), the surface site density (N S ), and the equilibrium constants (logK)
of the protolysis reaction(s) will be stored. Additionally mineral-specific capacitance values (C) will be kept
for the CC and the 1pK-BS model. Equilibrium constants (logK) of the surface complex formations are
entered for given combinations of mineral, ligand, and model. All stored SCM data are valid for 298.15K
only. However, the THEREDA database can be adapted to temperature-dependent surface complexation

modeling in the future if required (and when sufficient data sets are available). The extension towards SCM
required the synchronization of various table designs (e.g. for solids and bibliography) and the introduction
of numerous new tables, coupled to new data categories.
Eight issued technical papers (Downloads Documentations) help promoting the transition of THEREDA
into a real information and discussion platform on issues concerning the database but also on geochemical
modeling at large. We are looking forward to receive helpful feedback on THEREDA and are hoping for
intensified interactions of the THEREDA database project with the international user community.
[1] Altmaier, M. et al. (2008). “THEREDA - Ein Beitrag zur Langzeitsicherheit von Endlagern für nukleare und
nichtnukleare Abfälle”. ATW 53, 249.
[2] Neck, V. et al. (2009). “Thermodynamics of trivalent actinides and neodymium in NaCl, MgCl 2 , and CaCl 2
solutions: Solubility, hydrolysis, and ternary Ca-M(III)-OH complexes” Pure Appl. Chem., 81, 1555.
PC3
PC3-1
PC3-2
PC3-3
PC3-4
PC3-5
PC3-6
PC3-7
DEVELOPMENT AND APPLICATION OF MODELS
DISCUSSION OF SOME UNCERTAINTIES IN EVALUATION OF DIFFUSION
EXPERIMENTS WITH COMPACTED BENTONITE
D. Vopálka, A. Vetešník, E. Hofmanová (Czech Republic)
NUCLIDE TRANSPORT OF N-MEMBER DECAY CHAIN IN FRACTURED ROCKS
WITH STAGNANT WATER ZONE- MODEL DEVELOPMENT AND SIMULATIONS
P. Shahkarami, L. Liu, L. Moreno, I. Neretnieks (Sweden)
DISCRETE FRAGMENT MODEL FOR APPARENT FORMATION CONSTANTS OF
METAL COMPLEXES WITH HUMIC SUBSTANCES
T. Sasaki, H. Yoshida, S. Aoyama, T. Kobayashi, I. Takagi, H. Moriyama (Japan)
ANALYSIS OF SENSITIVITY OF MODEL OF TRANSPORTING SUBSTANCES FROM
RADIOACTICE WASTE REPOSITORY
J. Chudoba (Czech Republic)
REACTIVE TRANSPORT MODELING OF 129 I MIGRATION IN GRANITIC ROCKS
E. Orucoglu, B. van den Akker, J. Ahn (USA)
PHREEQC TRANSPORT MODELLING OF CS-137 TO ASSESS THE IMPACT OF A
NEW LEAK TO GROUND INTERACTING WITH EXISTING CONTAMINATION
L. Abrahamsen, J. Graham (UK)
CHARACTERISATION OF GROUNDWATER PATHWAYS AND MODELLING
SOLUTE TRANSPORT THROUGH THE BEDROCK AT OLKILUOTO, FINLAND
L. Hartley, J. Hoek , S. Baxter , L. Koskinen (UK, Finland)
PC3-1
DISCUSSION OF SOME UNCERTAINTIES IN EVALUATION OF DIFFUSION EXPERIMENTS
WITH COMPACTED BENTONITE
D. Vopálka, A. Vetešník, E. Hofmanová
Department of Nuclear Chemistry, Czech Technical University in Prague, CZ-11519 Prague,
Czech Republic
Compacted bentonite has been taken into account as an engineered barrier in high-level radioactive waste
repositories due to its very low permeability and high sorption capacity. The knowledge of values of both
effective (D e ) and apparent diffusion coefficients (D a ) and confidence limits of these values for critical
radionuclides is necessary for proper modelling of their transport in real barrier of waste disposal. The values
of diffusion coefficient obtained by standard type of evaluation of small scale diffusion experiments (e.g. by

time-lag method) represents in reality only characterisation of individual experiments that were performed
under specific conditions and were evaluated by methods that often were not completely consistent with real
experimental conditions. The influences of the presence of filters, evaluation by time-lag method of
experiments during which concentrations in both input and output reservoirs with working solutions were not
constant, and no consideration of the nonlinearity of sorption isotherm in evaluation process are examples of
possible sources of such inconsistencies, which can cause uncertainty of measured diffusion coefficients and
subsequently also determined retardation factor R.
The module prepared in the environment of GoldSim code used in our laboratory [1] can reflect by
evaluation of diffusion experiments real concentration changes in both reservoirs during the experiment,
radioactive decay, influence of separating filters and it also enables to model diffusion in the bentonite plug
in which the porosity is not homogeneous. The nonlinearity of the sorption isotherm was introduced in a
module made in the geochemical code PHREEQC in which both ion-exchange and surface complexation of
species of studied radionuclide could be taken into account.
Both types of mentioned approaches were used for the evaluation of cesium concentration profiles in
bentonite plugs for a set of in-diffusion experiments with two types of bentonite of Czech origin (B75 and
S65). In this case the diffusion module prepared in PHREEQC followed the description of the interaction of
cesium with clay formulated by Bradbury and Baeyens [2]. The parameters of three assumed interaction
processes (ion-exchange and complexation on two types of edge sites) were taken from literature data ([3],
[4]) and were tailored for the studied bentonites.
The obtained parameters of the sorption model were used for the description of interaction of a dissolved
contaminant with the bentonite surface, often formulated in the form of Freundlich and Langmuir isotherms.
The time and spatial variabilities of retardation coefficient R predicted by modelling in PHREEQC were
compared with R obtainable by standard types of evaluation of diffusion experiments, in our case with the
use of the module prepared in GoldSim. A thorough parametric study was completed with the aim to
evaluate the influence of model parameter variabilities on the uncertainty of retardation coefficient R. In the
frame of this study, an advanced sensitivity analysis method was mastered that was used for the assessment
of the developed transport model of sorbing species in the layer of compacted bentonite.
[1] D. Vopálka, H. Filipská, A. Vokál, Mat. Res. Soc.–Symp. Proc. 932, 983 (2006).
[2] M. Bradbury, M., Baeyens, B., J. Cont. Hydrol. 42, 141 (2000).
[2] A.M. Fernández, B. Baeyens, M. Bradbury, P. Rivas, Phys. Chem. Earth 29, 105 (2004).
[4] H. Filipská, K. Štamberg, J. Radioanal. Nucl. Chem. 270, 531 (2006).
PC3-2
NUCLIDE TRANSPORT OF N-MEMBER DECAY CHAIN IN FRACTURED ROCKS WITH
STAGNANT WATER ZONE- MODEL DEVELOPMENT AND SIMULATIONS
P. Shahkarami, L. Liu, L. Moreno, I. Neretnieks
Department of Chemical Engineering and Technology, Royal Institute of Technology (KTH),
S-100 44 Stockholm, Sweden
A model is developed to describe nuclide transport of a decay chain of arbitrary length in a single fracture
situated in a porous rock matrix accounting for hydrodynamic dispersion, matrix diffusion and migration
through stagnant water zone. The model also considers a linear equilibrium adsorption on both the fracture
surface and in rock matrices. The nuclides in the chain can diffuse directly from the flowing channel into the
microspores of the adjacent rock matrix. This allows radionuclides to diffuse into and out of the rock matrix
where they can absorb on the interior surfaces of the matrix and be considerably retarded. They can also
migrate first through the stagnant water zone and then diffuse further into its adjacent rock matrix. This may
result in a significant retardation of nuclide transport in the cases where the volume of stagnant water zone in
the fracture plane is orders of magnitude larger than that of the mobile water in the flowing fractures [1].
In spite of the complexities of the transport mechanisms involved in the model, it is shown that the analytical
solution to the Laplace-transformed concentration at the outlet of the flowing channel is exponential, and it
can easily be included into a 3-D channel network model [2] to describe nuclide transport through a

heterogeneous fractured media consisting of an arbitrary number of channels in rock units with piecewise
constant properties.
More importantly, by numerical inversion of the Laplace-transformed solution, the simulations made in this
study help to gain insight into the relative significance and the different contributions of chain decay,
hydrodynamic dispersion, migration through stagnant water zone, and diffusion into the rock matrices in
determining the fate of nuclide transport through fractured rocks.
Figure 3. Flow in a channel in a fracture from which solute diffuses into rock matrix as well as into stagnant water in
the fracture plane and then further into the rock matrix
The results suggest that migration through the stagnant water zone and its adjacent rock matrix may
considerably contribute to solute retardation and a simplified model that ignores the effect of stagnant water
zone can lead to a significant error in the estimated time of arrival and peak value of the nuclides. The results
also demonstrate that for a two-member decay chain, neglecting the parent and modelling its daughter as a
single species can result in significant overestimation of peak value of the nuclide. Moreover, it is found that
as the dispersion increases, the arrival time and peak time of daughter decrease, while the peak value
increases.
Table 1. Simulations conditions*
Case 1 Case 2 Case 3 Case 4
Matrix adjacent
to channel
✕ ✕ ✕ ✕
Water and its
adjacent matrix
− ✕ ✕ ✕
Adsorption ✕ ✕ ✕ ✕
Chain decay − − ✕ ✕
Dispersion − − − Pe x =100
Inlet Condition
C 10 = 0
C 20 = δ(t)
C 10 = 0
C 20 =δ(t)
C 10 =δ(t)
C 20 = 0
C 10 = δ(t)
C 20 = 0
* (✕)→ Considered
(−)→ Neglected

Figure 4. Comparison of breakthrough curves for cases 1-4
Figure 5. Comparison of breakthrough curves for different Peclet numbers in the channel
[1] Mahmoudzadeh, B., Liu, L., Moreno, L. & Neretnieks, I. 2013. Solute transport in fractured rocks with stagnant
water zone and rock matrix composed of different geological layers—Model development and simulations, Water
Resour. Res., 49, doi :10.1002/wrcr.20132.
[2] Gylling, B., Moreno, L. & Neretnieks, I. 1999. The Channel Network Model - A tool for transport simulation in
fractured media, Groundwater, 37, 367-375.
PC3-3
DISCRETE FRAGMENT MODEL FOR APPARENT FORMATION CONSTANTS OF METAL
COMPLEXES WITH HUMIC SUBSTANCES
T. Sasaki 1) , H. Yoshida 1) , S. Aoyama 1) , T. Kobayashi 1) , I. Takagi 1) , H. Moriyama 2)
1) Department of Nuclear Engineering, Kyoto University, Kyoto daigaku-Katsura, Nishikyo,
Kyoto 615-8540, Japan
2) Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun,
Osaka 590-0494, Japan
Due to the randomly cross-linked heterogeneous and macromolecule properties of humic substances (HSs) in
soils, sediments, and natural waters, the apparent formation constant β app of metal complex with HS
described as below changes depending on the solution conditions such as pH, metal and HS concentrations,
and ionic strength.
[M-L]
βapp
=
[M
n+ ][R
−]

where [R - ] and [M-L] are the concentration of the dissociated functional group and of the metal ion bound
with a ligand in HS, respectively. Numerous data have been collected for a wide range of chemical
conditions, and several empirical models have been proposed such as the well-known NICA/NICCA-Donnan
model and Model VI equipped with a set of free and hypothetical fitting parameters. However, such models
still remain in a development stage for polyvalent actinide ions, where hydrolysis reactions and the
stoichiometry in the presence of competitive ions should be taken into account. In the present study, we
improved a semi-empirical thermodynamic model, called discrete fragment model, which has succeeded in
the Eu-HS system [1], to deal with the β app for the di-, tri-, tetra- and pentavalent metal ions including
Th/Pu(IV) and Np(V). Assuming the hypothetical molecular image containing 9 basic ligands L j (j=1-9) of
aromatic and aliphatic carboxylic acids and connecting by alkyl chain and aromatic ring, 54 combinations of
sites are proposed. The formation constant of each site for the 1:1 and 1:2 species, β 11,k and β 12,k (k=1-54) is
expressed by
where the βs are defined as
log β log (9 )
log β = log β + log β + K (45 sites)
11, k ≡ βj1
sites
12, k j1 j2
where the β j1 of all metal series with nine L j is a known value obtained from many references. The fitting
parameter K implies an intramolecular chelating effect in complexation. Though the K might be depended on
the solution condition, it is assumed to be constant for simplicity. The abundance ratios of binding sites, p 11,k
and p 12,k , were defined as
p11, k = (1 − P) p j (9 sites)
2
12, k j1 j2 j1
p = 2 P ⋅p p ( j1 ≠ j2) or P ⋅ p ( j1 = j2) (45 sites)
where the fitting parameters P and p j are the fraction of 1:1
and 1:2 species, and the abundance ratios of 9 ligands
(j=1-9), respectively. As a result of the fitting analysis
of
the experimental β app values for all metal series, the
parameters were properly converged. For Th(IV) as an
example, the apparent solubility at 1ppm humic acid
and
I=0.1, is shown in Fig.1 with the help of β app s
calculated by the model, the hydrolysis constants, and
the
solubility product.
[1] Sasaki, T., et al., “Discrete fragment model for complex
formation of Eu(III) with humic acid”, J. Nucl. Sci. Technol.
45 (2008) 718.
PC3-4
ANALYSIS OF SENSITIVITY OF MODEL OF TRANSPORTING SUBSTANCES FROM
RADIOACTICE WASTE REPOSITORY
J. CHUDOBA (1)
Fig.1 Th-humic acid solubility by using β app
values calculated by the present model
(1)
Technical University of Liberec, Studentská 2, 461 17 Liberec, Czech Republic
The aim of the contribution is to create sensitivity analysis of the complex model of transporting substances
that release from the underground storage of burnt radioactive fuel. To calculate the transport of substances
the software Flow123D is used. This software allows to set the concentration of substance in each element of
the calculating mesh in a 3D space in time. The software provides a description of geological structure in the
rock, input radioactive decay of the substances to the model. The software includes no constant mass of the
released substance on the border near field / far field.

The 3D model of the hypothetic area (region Rožná – Olší, location Kraví Hora) is described by 9 different
types of rock. The rock has different hydraulic conductivity and porosity, which is changing with depth of
surface area.
Transport of the substances enables to include following parameters. From the results of other projects the
mass of the released substances on the border near field / far field (function of time) was calculated and it is
possible to count with sorption parameter of substances. Model includes radioactive decay of the substances.
On the basis of these input data the flow of underground water in the area and transport of substance from
repository were calculated with the software Flow123D. We found maximal concentration in some elements
and sum of concentration of the substance on surface of the area (this happens at the contact of geosphere
with biosphere).
There was created sensitivity analysis, when we changed parameters of hydraulic conductivity of each type
of rock and porosity of it. Further there were changed sorption parameters of self-transporting substances.
The result of the sensitivity analysis is setting of influence, which parameter most influences the
concentration of substance in the surface of the area.
[1] Chudoba J., Královcová J., Maryška J. Draft methodology for security analysis of underground storage location and
transport modeling of radionuclides in distant fields. Technical university of Liberec, 2009.
[2] Chudoba J., Praks P., Labeau P.E. Modelling of transport of radioactive substances from underground storage for
risk assessment – a case study. ESREL 2010: European Safety & Reliability International Conference, pages 1019-
1025, Rhodos, Greece, 2010.
[3] Vokal A. etc., Aktualizace referenčního projektu hlubinného úložiště radioaktivních odpadů v hypotetické lokalitě,
SURAO, 2010.
[4] Geuzaine C. & Remacle J. F. GMSH: a three-dimensional finite element mesh generator with built-in pre- and postprocessing
facilities. Version 2.4.2; http://www.geuz.org/gmsh/, 2009.
[5] https://dev.nti.tul.cz/trac/flow123d/.
PC3-5
REACTIVE TRANSPORT MODELING OF 129 I MIGRATION IN GRANITIC ROCKS
Esra Orucoglu, Bret van den Akker, and Joohnhong Ahn
Department of Nuclear Engineering, University of California at Berkeley, Berkeley, CA 94720
orucoglu@nuc.berkeley.edu; bpvandenakker@berkeley.edu, ahn@nuc.berkeley.edu
Fission product isotopes, inter alia 129 I, are known to be some of the major contributors to the radiological
risk resulting from the disposal of used nuclear fuels (UNF). The 129 I inventory in the gap between the fuel
cladding and fuel pellets will be released quickly after the breach of the waste package and cladding. If the
UNF is reprocessed, most of 129 I can be separated and solidified, although some would still be included in
intermediate-level wastes. Because of its bioavailability, long half-life, complex chemistry and low sorption
ability, this radioisotope has emerged as a focal point in the performance assessment for the geological
disposal of radioactive wastes. The chemical speciation of iodine in aqueous environments will be affected
by the conditions of the disposal site such as pH, redox reactions, temperature, salinity, all of which show
variations depending on depth of the disposal region. In this study, we have investigated 129 I migration in
granitic rocks by coupling chemical interactions and transport phenomenon with the help of the PHREEQC
and TTBX codes. The PHREEQC geochemical code has been used to model evolution of groundwater
chemistry as well as iodine speciation and sorption equilibrium at various depths in order to provide
parameters necessary for transport modeling. The TTBX code models the transport of iodine in a
heterogeneous one-dimensional pathway with advection, matrix diffusion and sorption retardation. Results
obtained by the PHREEQC code for depth-dependent sorption distribution coefficients for iodine have been
utilized in TTBX calculation.
We have calculated the iodine distribution coefficients (Kd) at different depths by simulating the iodine
speciation in groundwater and iodine sorption with fracture-filling materials and with granite matrix. First,
depth-dependent evolution of groundwater chemistry was simulated by PHREEQC. Groundwater movement

was assumed to be so slow that equilibrium is established locally between groundwater and minerals such as
quarts, albite, calcite and pyrite. Then, sorption reactions were described by the diffuse double layer model.
Surface protonation and deprotonation reactions of aluminum and clay surfaces were also taken into
consideration. Iodine speciation in the evolved groundwater and possible sorption reactions between iodine
species and aluminum surfaces found in rock material and fracture filling material and have been simulated.
The results show that groundwater is in reduced alkaline conditions, and iodine is found as I - form, for which
sorption distribution coefficients onto aluminum minerals in the geological formation have been found very
low, but still dependent of depth. Thus, we have obtained Kd values at respective depths. The Kd values vary
between 5.26x10 -6 - 1.41x10 -5 m 3 /kg for rock materials and fracture filling materials with increasing depth.
TTBX was developed for the release and transport of radionuclides through heterogeneous geological media.
The model accommodates a heterogeneous transport pathway through an arbitrary number of rock types and
admits a decay chain which consists of an arbitrary number of members, subject to an arbitrary timedependent
radionuclide release mode at the entrance point into a series of transport segments. The model
considers the effects of advection, sorption retardation in the transport pathway, matrix diffusion, and
radioactive decay and in-growth. Laplace transformed analytical solutions are inverted numerically by the
Talbot’s method.
In the present study, in addition to the Kd of iodine, the fracture aperture, the water velocity in the fracture,
the diffusion coefficient of iodine in the pore water in the rock matrix, the viscosity and density of water
have been assumed to be depth-dependent. Values of these parameters have been obtained for different
depths. It has been assumed that spent fuel is directly disposed of in a repository at 1,000m depth, and that
3% of the iodine inventory, which exists in the gap between the cladding and the fuel pellet, is released into
the geologic formation within 100 years after the waste package failure, while 97% released congruently
with the fuel matrix dissolution during 1,000,000 years. Granite is assumed to exist at the depth > 100 m.
Sedimentary layer is considered between the surface and the 100 m depth. The granite between 100 m and
1,000 m is divided into four regions. For each layer, groundwater chemistry evolution was simulated, and the
iodine speciation and sorption equilibrium were simulated. The results were compared with in-situ
measurement results at the Tono mine, indicating reasonably good agreement. For the midpoint depth of
each layer, the other transport parameters were evaluated.
The results for the transport simulation by the TTBX code show:
(1) Despite very weak sorption (e.g., the retardation factor slightly greater than 1), iodine would stay
within the deep layers near the repository level because of the retention mechanism by the matrix
diffusion. The fracture aperture is smaller as the depth increases, resulting in significantly smaller
water flow velocity in the fracture due to the cubic law. Therefore, relatively the contribution of
matrix diffusion becomes more significant.
(2) The long tail of iodine is observed in shallower levels, because at shallower levels, advection in the
fracture becomes more prominent. But, the retention by matrix diffusion at the deep layers are so
effective that both 3% and 97% are contained within the far-field geological formation.
Further numerical investigation is being carried out for various scenarios, and its results will be reported at
the conference.
PC3-6
PHREEQC TRANSPORT MODELLING OF Cs-137 TO ASSESS THE IMPACT OF A NEW LEAK
TO GROUND INTERACTING WITH EXISTING CONTAMINATION
L. Abrahamsen, J. Graham
National Nuclear Laboratory, 5 th Floor, Chadwick House, Birchwood Park, Warrington WA3 6AE, UK
Retrievals from legacy ponds and silos is a high priority for hazard reduction at the Sellafield site. Retrievals
from facilities that have undergone historic leaks of active liquor into ground sediments pose a particular
challenge, due to the risk of further leaks. Key to mitigating against further leaks is understanding the

ehaviour of different radionuclides in ground sediments, particular the interaction between existing
contamination and a new leak.
The sorption of Cs-137 onto illite-bearing sediments is assumed to occur via cation exchange onto a range of
sorption sites. Bradbury and Baeyens [1] presented a generalised cation exchange model to describe the
sorption onto three types of illite sorption site:
1. Frayed edge sites representing very high affinity sites for Cs sorption at the edges of illite clay
particles, comprising 0.25% of the total cation exchange capacity (CEC);
2. Type II sites representing higher affinity sites for Cs sorption comprising 20% of the CEC;
3. Planar sites on the layered surface of illites which represent 80% of the CEC. These have a relatively
low affinity for Cs sorption compared to frayed edge and type II sites.
The PHREEQC chemical speciation model has been used to construct a 1-D transport model, representative
of Sellafield sediments and based on the ion exchange sites described above. This model was used to
simulate transport of Cs-137 in three stages: during an initial historical leak; during a period of time
following the end of the historical leak; and during a future leak (assumed to be triggered by retrievals),
which interacts with existing Cs-137 contamination in the ground. The model considers advection by
groundwater, sorption of Cs-137 onto sediments and its radioactive decay (half-life 30.07 years).
The model predicts that Cs-137 from the initial leak migrates very slowly (Figure 1). This is consistent with
measurements taken from around the building, where little Cs-137 contamination has been observed more
than several metres from the building. Frayed edge sorption sites are predicted to become saturated rapidly,
with type II sites approaching saturation at higher concentrations. Planar sites remain at a low level of
saturation at all times.
During the period of time between leaks, the concentration of Cs-137 decreases by radioactive decay.
Migration is extremely slow. Upon the start of a new leak (Figure 2), the high concentrations of competing
ions (K + ) present in the silo liquor cause limited remobilisation of existing Cs-137 contamination. However,
the migration of Cs-137 remains slow compared to other species at all stages.
This initial modelling exercise is the first to consider the interaction of a new leak to ground with existing
Cs-137 contamination present in Sellafield sediments. It predicts that the presence of competing ions (such
as K + ) can cause some remobilisation of existing Cs-137 contamination. However, migration of Cs-137
remains very slow (< 1 m/year) at all times, which supports the case for Monitored Natural Attenuation
(MNA) of this radionuclide, avoiding the need for immediate ground remediation. This simple 1-D transport
model is currently being developed further into a more sophisticated, 3-D model that also considers the
migration of other radionuclides.

Cs-137 concentration in sediment (mol/dm 3 )
1.8E-03
1.6E-03
1.4E-03
1.2E-03
1.0E-03
8.0E-04
6.0E-04
4.0E-04
2.0E-04
0.0E+00
5 years (leak endpoint)
10 years
20 years
50 years
0 1 2 3 4 5 6 7 8 9 10
Distance from leak point (m)
Figure 1 - Cs-137 profile following initial leak, years indicate time elapsed since start of initial leak
Cs-137 concentration in sediment (mol/dm 3 )
1.8E-03
1.6E-03
1.4E-03
1.2E-03
1.0E-03
8.0E-04
6.0E-04
4.0E-04
2.0E-04
50 years (immediately before new leak)
1 year into future leak
3 years into future leak
5 years into future leak
0.0E+00
0 1 2 3 4 5 6 7 8 9 10
Distance from leak point (m)
Figure 2 - Cs-137 profile during new leak
[1] Bradbury, M. H. and Baeyens, B. (2000). A generalised sorption model for the concentration dependent uptake of
caesium by argillaceous rocks. J. Contam. Hydrol. 42. p141-163.

PC3-7
CHARACTERISATION OF GROUNDWATER PATHWAYS AND MODELLING SOLUTE
TRANSPORT THROUGH THE BEDROCK AT OLKILUOTO, FINLAND
L. Hartley (1) , J. Hoek (1) , S. Baxter (1) , L. Koskinen (2)
(1) AMEC, Building 150, Harwell Oxford, Didcot, Oxfordshire, OX11 0QB -UK
(2)
Posiva Oy, Olkiluoto, Eurajoki, 2160FI - Finland
Posiva has responsibility for the management of the disposal of spent fuel from the Loviisa and Olkiluoto
nuclear power plants. According to the three Decision-in-Principles endorsed by the Finnish Parliament in
2001, 2002 and 2010, the spent fuel from these power plants (up to 9000 tU) will be disposed in a geological
disposal facility at Olkiluoto. The Olkiluoto site has been studied for this purpose for over twenty years.
Posiva submitted an application to obtain a construction licence for the Olkiluoto disposal facility in
December 2012.
An essential part of the assessment of long-term safety of a KBS-3 repository is the analysis of groundwater
flow since it is the only natural means of transport of radionuclides to the biosphere. Such analysis requires a
description of details of the groundwater flow in and around the engineered barrier system as well as details
of the groundwater pathway to the biosphere during the next c. 10,000 years [1], as well as considering the
evolution of the system over the long-term. To do so consideration must be given to behaviour under future
climate periods such as glacial conditions. Other safety relevant considerations include [2]: the evolution of
groundwater chemistry in the vicinity of the deposition volumes; the consequences of potential seismic
events causing movement on existing fractures intersecting the repository; and the definition of criteria for
defining volumes suitable for deposition tunnels and deposition holes [3].
Groundwater flow in the crystalline rocks at Olkiluoto takes place in the interconnected fractures in the
bedrock. This, together with the safety assessment requirements listed above, motivates the use of a discrete
fracture network (DFN) modelling concept in describing and modelling groundwater flow and solute
transport at the site. The concept provides a ready framework to capture many of the important details of
fracture geometry, size, connectivity, fracture transmissivity and aperture, mineralogy, and rock matrix
diffusion. The DFN representation can be converted to an equivalent continuous porous medium
representation using flux based upscaling of the fracture system, to facilitate more conventional solute
transport modeling of coupled processes, for example.
This article gives an overview of the methodology used to provide an integrated description of groundwater
flow in the bedrock as the basis for characterising flow-related transport properties along groundwater
pathways. It traces the description of bedrock hydrogeology and solute transport through the synthesis of site
investigation data, parameter estimation of the fracture system, confirmatory testing of the site palaeohydrogeological
model against hydrogeochemical data, uncertainty assessment, and the characterisation of
site flow-related transport and retention properties. The approach begins with the integration of geological
description of bedrock structures, deformation and fracture domains as a framework for the overall site
descriptive model [4] using data from geophysical and mapping surveys together with multi-disciplinary data
from more than 50 deep core drilled holes and the excavation of the ONKALO tunnel research facility. The
drillhole measurements of fracture position, geometry, flow during hydraulic tests and mineral assemblages
are then interpreted to provide the basis for conceptualisation and parameterisation of the water conducting
fracture system through calibration of numerical models [5]. This results in a parameterised bedrock
hydrogeological DFN model of the main hydrogeological structural zones and the sparsely fractured rock
between.
The modelling approach is necessarily stochastic, and necessitates several assumptions and simplifications,
since only limited aspects of the fracture system can be determined directly. Therefore, alternative
assumptions have to be considered to illustrate the sensitivity to recognised uncertainties, and the models
tested against independent site data. Measured pressure profiles and hydraulic disturbances induced by
pumping or tunnel excavation are types of such test that can be used for further calibration of hydraulic
connectivity and properties. However, the chemical composition of groundwater samples taken from both

fractures and porewater is considered to provide the most useful confirmation of the overall site
understanding of groundwater circulation and solute transport. Conditions at the site have changed over the
last 8000 year period (Holocene) including changes to surface topography, sea-level and sea water salinity in
the Gulf of Bothnia. Modelling the consequences of these changes on groundwater composition in the
bedrock tests the overall hydrogeological description and properties and provides insight into the effects of
rock matrix diffusion.
[1] STUK, 2011. Disposal of Nuclear Waste, STUK Guide YVL D.5 (Draft 3).
[2] Safety case for the disposal of spent nuclear fuel at Olkiluoto - Performance Assessment 2012. Eurajoki, Finland:
Posiva Oy. POSIVA 2012-04. ISBN 978-951-652-185-8.
[3] Posiva 2011. Olkiluoto Site Description 2011. Posiva 2011-02. Eurajoki, Finland: Posiva Oy. (ISBN 978-951-652-
179-7).
[4] Posiva, 2012. Rock Suitability Classification, RSC-2012. Posiva 2012-24. Eurajoki, Finland: Posiva Oy.
[5] Hartley L., Appleyard P., Baxter S., Hoek J., Roberts D., Swan D., 2012. Development of a Hydrogeological
Discrete Fracture Network Model for the Olkiluoto Site Descriptive Model 2011. Eurajoki, Finland: Posiva Oy.
Working Report 2012-32.
PC4
PC4-1
PC4-2
MODEL VALIDATION
DIFFUSION AND FLOW IN THIN SLITS WITH VARYING APERTURE:
EXPERIMENTAL STUDIES ON VIZUALIZATION AND EVALUATION OF THE Q-
EQUIVALENT TRANSPORT MODEL
H. Winberg , I. Neretnieks, L. Moreno, L. Liu (Sweden)
BENCHMARKING EXERCISE OF REACTIVE TRANSPORT CODES USING THE
THERMOCHIMIE DATABASE
M. Grivé, A. Idiart, D. García, E. Colàs (Spain)
PC4-1
DIFFUSION AND FLOW IN THIN SLITS WITH VARYING APERTURE:
EXPERIMENTAL STUDIES ON VIZUALIZATION AND EVALUATION OF THE
Q-equivalent TRANSPORT MODEL
H. Winberg , I. Neretnieks, L. Moreno, L. Liu
Department of Chemical Engineering and Technology, Royal Institute of Technology,
S-100 44 Stockholm, Sweden
This paper presents a continuation of the experimental techniques developed by Neretnieks et al. (2011) [1]
and Winberg (2012) [2] to experimentally verify and visualize the Q-equivalent model.
The Q-equivalent concept was developed in 1979 by Neretnieks [3] to assess the amount of radionuclides that
can escape from a damaged copper canister (one of the three barriers in the Swedish repository for disposal
of high level radioactive waste, KBS-3) to the seeping water in rock fractures. The model sums up the
resistances of multiple barriers (e.g. diffusion through the bentonite clay, diffusion into the water, and
convection in the fracture network) as an equivalent flow rate (Q-eq).
The earlier experimental work was mainly focused on the diffusion transport of radionuclides, from the
bentonite clay, into stationary water. This work reports experiments with creeping flow in the fracture. The
variation in aperture in the fractures leads to complex flowpaths in the fracture and to channelling. The
advective mass transport has, in addition to diffusion, to be taken into account to describe the release and
transport.
To assess the diffusion of solute into a rock fracture with seeping water a transparent replica of a real rock
fracture, made of epoxy resin, was made by making positive casts of a rock core from the Äspö Hard Rock
Laboratory accordingly to the procedure of Hakami (1989) [4] . Figure 1 shows how the vertically mounted

fracture replica (1.) was arranged together with a dye chamber (2.) at the bottom. Decompression chambers
at the right (3a.) and left (3b.) side sealed at the upper edges allow the inlet and outlet pressures to be even
along the vertical sides.
Figure 1: Experimental Setup
To determine the fracture aperture the setup was placed in between a camera and a light source. Thereafter it
was filled with a constant concentration carmine solution (used to illustrate the radionuclides) and the light
transmission was measured.
The experiment was then carried out by first separating the fracture from the dye chamber with a filter (4.),
to avoid convection between chamber and fracture, and then filling the setup with de-ionized and degased
water.
The dye chamber was slowly filled with a concentrated carmine solution. It diffuses through the filter and
into the fracture. A pump (5.) is connected to the right decompression chamber to provide a flow with an
even pressure head along the whole height of the slit and a second tube was connected to the left
decompression chamber to enable extraction of the effluent (6.). The effluent concentration was measured by
photo spectrometry at regular intervals to determine how much dye had diffused into the fracture and been
swept away and thereby assess when steady state been obtained. The evolution of the diffusion front within
the fracture was also assessed by photos taken every 24 h.
By using the measured aperture data, simulations were made. They are compared with the experiments. The
diffusion-convection equation and the cubic law were implemented in a two-dimensional model by using
COMSOL Multiphysics (2012) to describe the stationary flow field in the fracture. The predicted result,
obtained from the diffusion-convection simulations, for the concentrations in the slit and effluent showed a
fair agreement with the experimental data. The results support the Q-equivalent concept.
[1] Neretnieks I., Moreno L., Winberg H., Liu L., (2011), Diffusion in a fracture with varying aperture, Migration 2011
Conference, China.
[2] Winberg H., (2012), Diffusion in Thin Slits with Variable Aperture, KTH, Master Thesis CHE Report 2012:9.
[3] Neretnieks I., (1979), Transport Mechanisms and Rate of Transport of Radionuclides in the Geosphere as Related to
the Swedish KBS Concept, International Atomic Energy Agency – SM -243/108, 315-339.
[4] Hakami H., (1989), Water Flow in Single Rock Joints, Technical Report Stripa Project, SKB.

PC4-2
BENCHMARKING EXERCISE OF REACTIVE TRANSPORT CODES USING THE
THERMOCHIMIE DATABASE
M. Grivé 1)* , A. Idiart 1) , D. García 1) , E. Colàs 1)
1) Amphos 21 Consulting S.L. Pg. Garcia i Fària, 49-51, 08019 Barcelona, Spain.
The test and validation of selected thermodynamic data are essential in the process of building a
thermodynamic database. Validation exercises based on geochemical calculations, using different
geochemical codes, should then be performed. Such benchmark exercises can increase confidence in the
application of both the thermodynamic databases and the reactive transport codes. Nevertheless, a general
lack of such exercises is observed in the literature.
The ANDRA‘s thermodynamic database, ThermoChimie, is extracted into the different text format files
needed by geochemical codes such as PhreeqC, Toughreact and Crunchflow (among others) and can then be
diffused to the scientific community. Amphos 21 has assisted ANDRA in further development of the
ThermoChimie database and other applications.
The main objective of the work presented here is the validation of the thermodynamic data files compatible
with the geochemical and reactive transport codes PhreeqC, Crunchflow and Toughreact extracted from
ThermoChimie. This verification has been based on the simulation of several benchmark exercises designed
to test different chemical systems of interest for ANDRA. In the present paper, one of these benchmark
exercises, a reactive transport simulation of clay-cement interaction, is presented.
The cement-clay interaction benchmark has been simulated with PhreeqC, Crunchflow and Toughreact and
considers the following aspects:
• Diffusive transport: testing the behaviour of a conservative tracer (i.e., not related to thermodynamic
modelling).
• Behaviour of the databases in 1D reactive transport simulations: geochemical reactions coupled with
diffusive transport.
• Determine if small changes in the initial systems in the three codes propagate to appreciable
differences during the simulation.
• Quantify the differences between the different codes in terms of mineral dissolution/precipitation,
total aqueous species, pH, Eh, and cation exchanger composition.
The main conclusions of the work presented here are the following:
• The benchmark study among PhreeqC, Toughreact, and Crunchflow showed a very good agreement
when comparing the evolution of the aqueous phase, the Eh and pH, the mineral assemblage, and the
cation exchanger composition.
• Some small differences can still be observed in the mineral phase assemblage and total aqueous
concentrations near the interface. These differences are due to a number of factors analysed here,
such as different time stepping algorithms, or discretization methods used by each code.
• The use and consistency of molar volumes data in Crunchflow and Toughreact has also been
successfully verified and confirmed. Missing molar volume data in the database can pose practical
problems when analysing simulation results. This issue has been exemplified with the present
benchmark exercise.
One of the most important outcomes of the benchmark exercise is that all codes are able to generate
comparable results for a scenario of significant relevance in the field of nuclear waste disposal.
Acknowledgements The authors would like to thank the French National Agency for Radioactive Waste
Management (ANDRA) for funding this work through the ThermoChimie project. E. Giffaut is
acknowledged for fruitful scientific discussions.

PIP-1
PIP-2
PIP-3
PIP-4
PIP-5
PIP-6
PIP-7
EMSL RADIOCHEMISTRY ANNEX: A NEW INTERNATIONAL USER-FACILITY FOR
THE STUDY OF RADIOLOGICAL SAMPLES
N.J. Hess, A.A. Campbell (USA)
FORGE - FATE OF REPOSITORY GASES
R. Shaw (UK)
CROCK: CRYSTALLINE ROCK RETENTION PROCESSES - A 7TH FRAMEWORK
PROGRAMME COLLABORATIVE PROJECT (2011-2013)
T. Rabung, D. Garcia, J. Molinero (Germany, Spain)
FP7 COLLABORATIVE PROJECT FIRST-NUCLIDES
B. Kienzler, V. Metz, L. Duro, A. Valls (Germany, Spain)
THE EUROPEAN NUCLEAR ENERGY FORUM (ENEF) GUIDELINES FOR
ESTABLISHING AND NOTIFYING NATIONAL PROGRAMMES UNDER THE
EURATOM WASTE DIRECTIVE
G. Buckau (EC)
TALISMAN - A LARGE INTERNATIONAL EC FP7 EURATOM FRAMEWORK
PROJECT
Bourg, S., Altmaier, M., Bryan, N., Collings, P., Dacheux, N, Duplantier, B, Ekberg, Ch.,
Grolimund, D., Natrajan, L., Poinssot, Ch., Raison, Ph., Schaefer, Th, Scheinost, A.,
Schimmelpfennig, B.
CINCH–II PROJECT – NEXT STEP IN THE COORDINATION OF EDUCATION IN
NUCLEAR- AND RADIOCHEMISTRY IN EUROPE
Jan John, Václav Čuba, Mojmír Němec, Teodora Retegan, Christian Ekberg,
Gunnar Skarnemark, Jukka Lehto, Teija Koivula, Paul J. Scully, Clemens Walther,
Jan-Willem Vahlbruch, Nick Evans, David Read, Eric Ansoborlo, Bruce Hanson,
Lindis Skipperud, Brit Salbu, Jon Petter Omtvedt
PIP-1
EMSL RADIOCHEMISTRY ANNEX: A NEW INTERNATIONAL USER-FACILITY FOR THE
STUDY OF RADIOLOGICAL SAMPLES
Nancy J. Hess and Allison A. Campbell
EMSL, Pacific Northwest National Laboratory, Richland WA USA
The Radiochemistry Annex, a new state-of-the-art laboratory to facilitate the application of advanced
analytical methods to the study of samples containing radionuclides, has been established at EMSL, the
Environmental Molecular Sciences Laboratory, a U.S. Department of Energy Office of Biological and
Environmental Research user facility located at Pacific Northwest National Laboratory in Richland,
Washington. It supports world-class research in the biological, chemical and environmental sciences to
provide solutions to the U.S.’s environmental challenges. EMSL's distinctive focus on integrating
computational and experimental capabilities, as well as collaborating among disciplines, yields a strong
synergistic scientific environment.
Development of accurate and validated predictive models of radionuclide fate and transport in the
environment requires mechanistic understanding of the chemistry of the radionuclides in their host
environment at the molecular level. Interfacial molecular science has been a rapidly growing area that has
contributed enormously to the understanding of the molecular processes that can control the geochemistry
and biogeochemistry of natural systems. It is widely recognized that the critical determinants of radionuclide
mobility are oxidation state, chemical speciation, and formation of surface and aqueous complexes. How
environmental conditions impact these determinants is key to the understanding of the fundamental processes

controlling the fate and transport of radionuclides in environmental systems and to unravelling the
fundamental chemical bonding of actinides and other radionuclides at the interface of novel materials for
selective extraction or materials fabrication. Unfortunately, the application of important advances in
molecular science to radiochemistry, particularly in the area of evaluating the importance of interfacial
processes, has lagged behind other areas of environmental science owing to the expertise required to perform
the radiological work, the need for dedicated equipment and facilities for such studies, and the fundamental
difficulties of observing molecular level processes for radionuclides that are often present in very minor
amounts in the interfacial region of bulk wastes or geologic materials.
A major objective of EMSL’s Radiochemistry Annex is to provide a specialized environment where users
from around the world can apply advanced experimental resources for microscopy and interfacial molecular
science to studies of radionuclides in environmental samples and waste forms. The radiochemistry facility
consists of approximately 6000 sq ft of lab space. The surface analysis-imaging suite contains a Focused Ion
Beam – Scanning Electron Microscope, Transmission Electron Microscope with an Electron Energy Loss
Spectrometer, Scanning Probe Microscope (Atomic Force Microscope), X-ray Photoelectron Spectrometer
and an Electron Microprobe with several different detector systems. The Magnetic Resonance facility houses
wide bore 100 and 750 MHz NMR spectrometers and an EPR spectrometer. The sample preparation and
analytical suite houses several instruments including Inductively Coupled Plasma – Mass Spectrometry and
Quadrupole Time of Flight Mass Spectrometry for the analysis of solution samples.
These capabilities will provide the spatially resolved elemental analysis, oxidation state determination,
chemical speciation, mineral identification, and microbe-mineral associations necessary for understanding
the chemical fate and mobility of radionuclides in the biogeochemical environment. Together with
NWChem, EMSL’s premier computational modeling code, users are able to address radionuclide systems
from both experimental and computational vantage points. User research conducted at the Radiochemistry
Annex will provide the scientific foundation required to inform and guide the development of engineering
and remediation solutions to critical challenges faced in today’s nuclear waste management and advanced
nuclear fuel cycle. EMSL’s new radiochemistry annex is currently open to users through EMSL’s existing
peer-review proposal process
PIP-2
FORGE – FATE OF REPOSITORY GASES
Richard Shaw (FORGE Co-ordinator)
British Geological Survey, Keyworth, Nottingham, NG12 5GG, UK
The FORGE project was a pan-European project with links to international radioactive waste management
organisations, regulators and academia, specifically designed to tackle the key research issues associated
with the generation and movement of repository gasses with partners from 24 organisations in 12 European
countries. It is supported by funding under the European Commission FP7 Euratom programme and started
in 2009 and ended in 2013. Of particular importance are the long-term performance of bentonite buffers,
plastic clays, indurated mudrocks and crystalline formations. FORGE has provided experimental data to
reduce uncertainty relating to the quantitative treatment of gas in performance assessment. This has been
achieved through a series of laboratory, field-scale experiments and modelling, including the development of
new methods for up-scaling allowing the optimisation of concepts through detailed scenario analysis.
Various gases will be generated in a repository including hydrogen (from metal corrosion), methane and
carbon dioxide (both from decomposition of organic materials contained in some wastes). Understanding
where and how these gases form and how they move through the repository and surrounding rocks was the
focus of the FORGE project. By using small scale laboratory experiments, large scale field tests (performed
at a number of underground research laboratories throughout Europe), data and numerical modelling the
results from FORGE is providing information to help guide repository design and predict future radionuclide
migration.

The understanding and prediction of the evolution of repository systems over geological time scales requires
a detailed knowledge of a series of highly-complex coupled processes. There remains uncertainty regarding
the mechanisms and processes governing gas generation and migration in natural and engineered barrier
systems. It is important to understand a system to an adequate level of detail to allow confidence in the
assessment of site performance, recognising that a robust treatment of uncertainty is desirable. Of particular
importance to the European radioactive waste management programmes are the long-term engineering
performance of bentonite buffers, plastic clays, indurated mudrocks and crystalline formations.
Results of a series of long-term laboratory experiments to examine the mechanisms controlling gas flow and
pathway sealing in the Callovo-Oxfordian Claystone (COx), the proposed host rock for the French
repository, demonstrate that advective gas flow is accompanied by dilation of the samples (i.e. the formation
of pressure induced micro-fissures) at gas pressures significantly below that of the minimum principal stress.
Flow appears to occur through a local network of inherently unstable pathways, whose properties vary
temporarily and spatially within the claystone. The coupling of parameters results in the development of
significant time-dependent effects, impacting many aspects of COx behaviour, from gas breakthrough time,
to the control of deformation processes. Variations in gas entry, breakthrough and steady-state pressures are
indicative of microstructural heterogeneity which may exert an important control on the movement of gas.
As data continues to be acquired and our understanding of these processes improves, a new conceptual
model for advective gas flow in COx is beginning to emerge, one in which the onset of gas flow and the
hydromechanical response of the material are integrally linked. The importance of pathway dilatancy during
gas flow has now been clearly demonstrated and is driving the development of new experimentation to
continue the development of the conceptual model for gas flow in COx.
Data collected during from a study of gas migration in bentonite clearly demonstrate a strong coupling
between total stress, pore-pressure and applied gas pressure. In both tests so far completed, the evidence is
for gas migration through the saturated bentonite by way of dilational pathways. This provides more
evidence that, in some circumstances, gas flow through clay rich materials is at least partially through
dilatant pathways.
Experimental validation of critical stress theory applied to repository concepts has greatly increased our
understanding and database of fracture flow properties. This study has highlighted the importance of stresshistory
on the flow properties of fractures because these systems display considerable hysteresis. Shearing
has been seen to be a very effective self-sealing mechanism. Repeat gas injection testing has shown
repeatability in “gas entry” values, but considerable differences have been seen in gas peak pressures.
The effect of healing of the interfaces between manufactured bentonite blocks has been demonstrated by
measuring the shear strength properties of the healed interface. The observation of significant cohesion
confirms the “real” healing of the interface.
Acknowledgement
With thanks to all FORGE partners for their contribution in developing the FORGE project and their effort
over the last four years working towards the successful outcome of the project.
Further details on the FORGE project and its outcomes can be accessed at www.FORGEproject.org. All
reports arising from the project are published here and will remain accessible until at least 2019.
For more information on the project contact the co-ordinator at rps@bgs.ac.uk
FORGE receives funding from the European Commission FP7 Euratom Programme under grant agreement
230357

PIP-3
CROCK: CRYSTALLINE ROCK RETENTION PROCESSES
A 7 TH FRAMEWORK PROGRAMME COLLABORATIVE PROJECT (2011-2013)
Th. Rabung, D. Garcia * , J. Molinero *
Institut für Nukleare Entsorgung, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344
Eggenstein-Leopoldshafen, Germany
* Amphos 21, Passeig de Garcia i Faria 49-51 – Barcelona, Spain
The EURATOM FP7 Collaborative Project “Crystalline Rock Retention Processes” (CP CROCK) is
established with the overall objective to develop a methodology for decreasing the uncertainty in the longterm
prediction of the radionuclide migration in the crystalline rock far-field. The project is launched in
response to the need identified in conjunction with selection of retention data for the forthcoming crystalline
host-rock HLW disposal Safety Case. The process of selecting a set of data for this purpose showed that the
spread in data is broad and that this spread in data cannot presently be related to material properties or
processes. Consequently, very conservative numbers need to be used in order to be defendable within the
Safety Case. This does not lead to unacceptable dose predictions, but remains highly unsatisfactorily.
The project makes use of the broad set of existing analytical approaches, methodologies, and general
knowledge from decades of past investigations. It builds on the output and main conclusions of the 6 th FP IP
FUNMIG project and the Swedish site selection program. The experimental program reaches from the nanoresolution
to the Performance Assessment (PA) relevant real site scale, delineating physical and chemical
retention processes. Existing and new analytical information provided within the project is used to set up
step-wise methodologies for up-scaling of processes from the nano-scale through to the PA relevant kmscale.
Modeling includes testing up-scaling process and parameters for the application to PA and in
particular, the reduction of uncertainty.
The scientific-technical work program of the project is structured along six RTD (Research and
Technological Development) workpackages (WP1-6). Workpackage 1 started at the very beginning of the
project providing new drill core fracture samples and characterizing the experimental materials.
Workpackage 2 focuses on radionuclide transport and sorption studies. Workpackage 3 deals with matrix
diffusion and natural chemical homologue analysis. The general objective of workpackage 4 is to
conceptualize and model radionuclide transport processes on systems at different scales. In workpackages 5
is described how the outcome of the other WPs can contribute to decrease the uncertainty in PA related with
transport treatment. Workpackage 6 is a cornerstone of the project, since its first objective is to establish a
state-of-the-art of the current knowledge on retention processes in crystalline rocks, then to continuously
collect the results obtained in the other workpackages, and finally to deliver a report summarizing the major
advances which will have been accomplished at the end of the project. There is also one workpackage on
knowledge management, dissemination and training (WP7). The last workpackage is on administrative and
financial project management (WP8).
The project started on 1 st January 2011 and will last 2 years and a half (2011-2013).The project is
implemented by a consortium with 10 Beneficiaries consisting of large European Research Institutions,
Universities and SME’s and from countries with dedicated crystalline host-rock disposal programs and
particular competence in this field. National Waste Management organizations participate as associated
groups, contributing with co-funding to beneficiaries, infrastructure, knowledge and information. They also
contribute together with national regulators to guidance with respect to application of the project to the
disposal Safety Case and scientific-technical review. The project is open for additional organizations
entering into formal cooperation and participation via Associated Group agreement.
More information about the project and the scientific output can be found on the project webpage:
http://www.crockproject.eu/
The research leading to these results has received funding from the European Atomic Energy Community's
Seventh Framework Programme (FP7/2007-2011) under grant agreement No. 269658 (CROCK).

PIP-4
FP7 COLLABORATIVE PROJECT FIRST-NUCLIDES
Bernhard Kienzler 1) , Volker Metz 1) , Lara Duro 2) and Alba Valls 2) ,
(1) Karlsruhe Institute of Technology, Karlsruhe
(2) Amphos 21, P. García Faria 49-51, 1-1, E-08019-Barcelona, Spain
* Corresponding author: lara.duro@amphos21.com
The FP7 Collaborative Project FIRST-Nuclides is focused on the study of the fast release of safety relevant
radionuclides from the spent nuclear fuel disposed in an underground facility once groundwater intrudes the
geological repository and contacts the spent fuel. Both, experiments and modelling studies are planned
within the 3 years project. The project consortium includes 10 partners from large Research Institutions and
SME’s (small and medium enterprises) from 7 countries, as well as the European Institute for Transuranium
Elements.
The research of the project aims at contributing to answer some questions still open regarding the fast release
of radionuclides of relevance for safety analyses. In this regard, the release mechanisms and the speciation of
anionic elements, such as iodine, carbon, chloride and selenium are not completely understood. The
understanding of the behaviour of these elements deserves special attention, as they are not very efficiently
retained in the engineering barriers of the repository. A combination of experimental methods with irradiated,
high burn-up UOx fuel samples in hot-cells and advanced modelling studies with existing data and data
gathered along the project is planned. The difficulty of the work with hot material requires the involvement
of specialized institutions, and training and information exchange among the consortium members is an
important part of the project.
The collaborative project is organized in four research, technical and development workpackages. WP 1
deals with the selection, characterization and preparation of materials and the set-up of the corresponding
tools. WP 2 “Gas release and rim and grain boundary diffusion” and WP 3 “Dissolution based release” cover
the experimental determination of fission gases release, rim and grain boundary diffusion processes and the
dissolution based fast/instant radionuclide release including the gap and grain boundary inventories and the
dependence of the fast release on the UO 2 fuel, burn-up, linear power and fuel temperature, ramping, and
storage time. WP 4 “Modeling” deals with numerical treatment of migration/retention processes of fission
products in the spent fuel structure, a realistic relationships between FGR and release of iodine, and other
RNs. Furthermore, migration of gaseous and non-gaseous fission products in the microstructures (grains,
grain boundaries, gas bubbles or pores), are modelled and the up-scaling from laboratory sample sizes to the
fuel rod scale.
PIP-5
THE EUROPEAN NUCLEAR ENERGY FORUM (ENEF) GUIDELINES FOR ESTABLISHING
AND NOTIFYING NATIONAL PROGRAMMES UNDER THE EURATOM WASTE DIRECTIVE
Gunnar Buckau
European Commission, Joint Research Centre – Institute for Transuranium Elements (JRC-ITU),
Herrmann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany
The present paper describes guidance established within the context of implementing the European “Waste
Directive”. The EURATOM Directive establishing “Safe and Responsible Management of Spent Fuel and
Radioactive Waste” 2 (Council Directive 2011/70/EURATOM), entered into force 22 August 2011. The
Directive was negotiated between the EU Member States and adopted by the Council. The Member States
commit to adopting National Policies in agreement with the best international standards and ensure that a
National Framework is available. A key implementation requirement for the Member States is to establish
and implement a National Programme where the policy is turned into practice. Four years after entering into

force, Member States need to notify these National Programmes to the European Commission (DG ENER).
The Commission will assess the National Programmes and request further information or action, as
necessary.
A Guide on the expectations for National Programmes is useful both for the Member States in establishing
and implementing National Programmes, and for the Commission’s assessment. The European Nuclear
Energy Forum (ENEF) established such a Guide 2 . The Members of ENEF are especially representatives from
industry and national waste management organizations. The European Nuclear Safety Regulators Group
(ENSREG) was also consulted for comments. Member States were also consulted for feedback and
comments, including invitation to a Workshop where the draft Guide was presented. By this overall process,
the Guide, having no legal implications, has a Peer character with broad involvement of the organizations
responsible for establishing and implementing the National Programmes.
The Guide discusses the directly link between the Policy, the Framework and the National Programme. The
waste inventory with the estimates of future arisings establishes the actual problem to be tackled. The
inventory is also an instrument for ensuring that disposal solutions are allocated to all waste remaining after
clearance, processing and treatment. The bulk of the Guide treats the provisions in Article 12.1 where the
elements of National Programmes are given. The Guide also provides a proposal for a possible structure of a
lead document for documenting, disseminating and communicating the National Programme to different
stakeholders, but also for notifying the National Programme to the Commission.
1 OJ L 199 of 2.8.2011: Council Directive 2011/70/EURATOM establishing a Community framework for the
responsible and safe management of spent fuel and radioactive waste resulting from civilian activities, adopted 19 July
and entered into force on 22 August 2011.
2 Guidelines for the Establishment and Notification of National Programmes under the Council Directive
2011/70/EURATOM of 19 July 2011 on the Responsible and Safe Management of Spent Fuel and Radioactive Waste,
European Nuclear Energy Forum (ENEF), 2013.
PIP-6
TALISMAN - A LARGE INTERNATIONAL EC FP7 EURATOM FRAMEWORK PROJECT
Bourg, S. 1 , Altmaier, M. 2 , Bryan, N., Collings, P. 4 , Dacheux, N. 5 , Duplantier, B. 6 , Ekberg, Ch. 7 , Grolimund,
D. 8 , Natrajan, L. 2 , Poinssot, Ch. 1 , Raison, Ph. 9 , Schaefer, Th. 2 , Scheinost, A. 10 , Schimmelpfennig, B. 2 ,
1 Commissariat à l’Energie Atomique, France
2 Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal, Germany
3 Centre for Radiochemistry Research, University of Manchester, UK
4 National Nuclear Laboratory, UK
5 Centre National de la Recherche Scientifique, France
6 LaGrange sarl (LGI Consulting), France
7 Chalmers University of Technology, Sweden
8 Paul Scherrer Institut, Switzerland
9 Institute for Transuranium Elements, European Commission
10 Helmholtz-Zentrum Dresden-Rossendorf, Germany
TALISMAN is a large international project funded within the European Commission FP7 EURATOM
framework. The aim of TALISMAN is to offer transnational access to large infrastructures for a safe
management of actinides. TALISMAN project is coordinated by CEA (contact: stephane.bourg@cea.fr).
Safety issues are of fundamental importance for the acceptance and sustainable application of nuclear energy
as was strongly reinforced following the Fukushima accident. Actinides play a central role in the nuclear fuel
cycle from mining, fuel fabrication, energy production, up to reprocessing, partitioning and transmutation
treatment of used fuel, and finally the management and disposal of radioactive waste. Fundamental
understanding of actinide properties and behaviour in fuel materials during the separation processes and in
geological repositories is an imperative prerequisite to tackle all the related safety issues.

Unravelling the complexity of the actinide components of used nuclear fuel certainly represents one of the
great challenges in nuclear science. To meet the needs of safe and sustainable management of nuclear
energy, it is essential to maintain a high level of expertise in actinide sciences in Europe and to prepare the
next generation of scientists and engineers who will contribute to developing safe actinide management
strategies. Because actinides are radioactive elements, their study requires specific tools and facilities that are
only available to a limited extent in Europe.
Only a few academic and research organisations have the capabilities and licenses to work on actinide
elements. It is therefore strategic to coordinate the existing actinide infrastructures in Europe and to
strengthen the community of European scientists working on actinides. Within TALISMAN we offer (for
positively evaluated scientific research proposals submitted in reply to a specific TALISMAN call) access to
the previous ACTINET Pooled Facilities (CEA Atalante and CEA DPC, France; ITU Laboratories & hotcells,
European Commission; KIT-INE laboratories and KIT-INE beamline, Germany; HZDR-IRE & ROBL,
Germany; PSI microXAS Beamline, Switzerland) to which two new facilities have been added: NNL Central
Lab in the UK and CHALMERS in Sweden.
TALISMAN leads and coordinates a network of actinide facilities across Europe, but also manages a
network between facilities and users to increase the knowledge for a safer management of actinides.
TALISMAN also enhances the efforts made to support education and training issues by continuing the
former ACTINET Summer School series and travel grant attributions for attending international conferences.
The TALISMAN project website http://www.talisman-project.eu offers detailed information on all
TALISMAN activities, including contact addresses, TALISMAN newsletters, announcements and
description of open and forthcoming calls for transnational user access and indicates several other options to
perform actinide research within the TALISMAN context.
PIP-7
CINCH–II PROJECT – NEXT STEP IN THE COORDINATION OF EDUCATION IN NUCLEAR-
AND RADIOCHEMISTRY IN EUROPE
Jan John 1 , Václav Čuba 1 , Mojmír Němec 1 , Teodora Retegan 2 , Christian Ekberg 2 , Gunnar Skarnemark 2 , Jukka
Lehto 3 , Teija Koivula 3 , Paul J. Scully 4 , Clemens Walther 5 , Jan-Willem Vahlbruch 5 , Nick Evans 6 , David
Read 6 , Eric Ansoborlo 7 , Bruce Hanson 8 , Lindis Skipperud 9 , Brit Salbu 9 , Jon Petter Omtvedt 10
1 Czech Technical University in Prague, Czech Republic
2 Chalmers University of Technology, Gothenburg, Sweden
3 University of Helsinki, Finland
4 National Nuclear Laboratory Ltd., Warrington, United Kingdom
5 University of Hanover, Germany
6 Loughborough University, Great Britain
7 Commissariat à l’énergie atomique et aux énergies alternatives, France
8 Leeds University, United Kingdom
9 Norwegian University of Life Sciences, Aas, Norway
10 University of Oslo, Norway
Any of the potential options for the nuclear power – both the renaissance, if any, or the phase out – will
require significant numbers of the respective specialists, amongst others the nuclear and/or radiochemists. In
parallel, a significant demand exists for these specialists in non-energy fields, such as environmental
protection, radiopharmacy, nuclear medicine, biology, authorities, etc. Since the numbers of staff in teaching
and the number of universities with facilities licensed for the work with open sources of ionizing radiation
has decreased on or sometimes even below the critical level, coordination and collaboration are required to
maintain the necessary teaching and training capabilities.
The CINCH-II project, aiming at the Coordination of education and training In Nuclear CHemistry in
Europe, will be a direct continuation of the CINCH-I project which, among others, identified the EuroMaster
in Nuclear Chemistry quality label recognized and guaranteed by the European Chemistry Thematic Network

Association as an optimum common mutual recognition system in the field of education in Nuclear
Chemistry in Europe, surveyed the status of Nuclear Chemistry in industry / the needs of the end-users,
developed an efficient system of education/training compact modular courses, or developed and tested two
electronic tools as a basis of a future efficient distance learning system.
In the first part of this paper, the achievements of the CINCH-I project will be described. This description
will cover both the status review and the development activities of this Collaboration. In the status review
field, the results of a detailed survey of the universities and curricula in nuclear- and radiochemistry in
Europe and Russia will be presented. Another survey mapped the nuclear- and radiochemistry in industry –
specifically the training and education needs of the end users. In the development activities field, the main
achievements of the CINCH project will be presented. They are particularly the NukWik – an open platform
for collaboration and sharing teaching materials in nuclear- and radiochemistry based on a wiki engine.
Further outputs are a set of compact joint modular courses in different branches of modern nuclear- and
radiochemistry, or an e learning platform (CINCH Moodle) available for both education and training
(applicable at the Ph.D., life-long learning, and MSc. levels).
The expected outcomes of the follow-on CINCH-II project will be described in detail. The CINCH-II
project is built around three pillars - Education, Vocational Education and Training (VET), and Distance
Learning - supported by two cross-cutting activities – Vision, Sustainability and Nuclear Awareness that
includes also dissemination, and Management. Its main objectives, expected to have the broadest impact to
the target groups, are further development and implementation of the EuroMaster in Nuclear Chemistry,
completion of a pan-European offer of modular training courses for the customers from the end users,
development of a Training Passport in Nuclear Chemistry and preparing the grounds for the European Credit
system for Vocational Education and Training (ECVET) application in nuclear chemistry, implementation of
modern e-learning tools developed in CINCH-I and further development of new tools for the distance
learning, laying the foundations of a Nuclear Chemistry Education and Training Platform as a future
sustainable Euratom Fission Training Scheme (EFTS) in Nuclear Chemistry, development of a Sustainable
Systems for Mobility within the Nuclear Chemistry Network, or development of methods of raising
awareness of the possible options for nuclear chemistry in potential students, academia and industry. The
CINCH-II project will mobilize the identified existing fragmented capabilities to form the critical mass
required to implement the courses and meet the nuclear chemistry postgraduate education and training needs,
including the high-level training of research workers, of the European Union. Networking on the national
level and with existing European as well as international platforms will be an important feature of the project.

SESSION 6
C2: COUPLING CHEMISTRY AND TRANSPORT
MODELING REACTIVE TRANSPORT PROCESSES IN POROUS MEDIA WITH
OPENGEOSYS
H. Shao, O. Kolditz (INVITED) (Germany)
LONG-TERM REACTIVE TRANSPORT SIMULATION OF THE INTERACTIONS OF
CORROSION PRODUCTS AND COMPACTED BENTONITE IN A HLW REPOSITORY AND
SENSITIVITY ANALYSES TO KEY PARAMETERS
J. Samper, A. Naves, L. Montenegro, A. Mon, B. Pisani (Spain)
NUMERICAL ANALYSIS OF MULTI-MINERALS TRANSFER BETWEEN THE HOST
ROCK AND GROUNDWATER AGAINST EXPERIMENT FOR NUCLEAR WASTE
DISPOSAL
Xiao Hui Chen, S. Thornton, J. Small, E. Moyce, S. Shaw, T. Milodowski, C. Rochelle (UK)
HIGH-PERFORMANCE REACTIVE TRANSPORT MODELLING OF RADIONUCLIDE
MIGRATION. THE COMSOL-PHREEQC APPROACH
J. Molinero, A. Nardi , D. García, C. Domènech, M. Grivé, B. Cochepin (Spain, France)
C2-1
C2-2
C2-3
C2-4
C2-1
MODELING REACTIVE TRANSPORT PROCESSES IN POROUS MEDIA WITH OPENGEOSYS
Haibing Shao 1 , Olaf Kolditz 1,2 , Georg Kosakowski 3 , Christof Beyer 4 , Sebastian Bauer 4
1 UFZ – Helmholtz Center for Environmental Research, Permoserstrasse 15, 04318 Leipzig, Germany.
2 Applied Environmental System Analysis, TU Dresden, Germany.
3 Paul-Scherrer-Institute, Switzerland.
4 Christian-Albrechts-Universität Kiel, Germany.
The numerical modeling of reactive transport processes has wide applications in different disciplines of earth
sciences. In contaminant hydrogeology, it was often utilized to predict the degradation of chemical pollutants
in the groundwater. For deep geothermal sites, it is helpful in understanding the water-rock interactions, and
accessing its impact on the performance of reservoirs. Last but not least, it is one of the tools to evaluate long
term behaviors of radionuclides’ migration in waste repositories.
Recently, multiple geochemical solvers have been coupled with the OpenGeoSys software in order to
simulate the above mentioned reactive transport problems, e.g. PhreeqC, ChemApp, BRNS. OGS itself
provides a large variety for different types of kinetic reactions. Among them, the OGS#GEMS code is
capable of simulating complex geochemical reactions, including mineral precipitation/dissolution, cation
exchange and ideal/non-ideal solid solution formations. The GEMS chemical solver will also update the
porosity values by counting the volume change of mineral phases, making the code capable of handling
clogging processes. On the technical side, the Massage-Passing-Interface (MPI) library has been
implemented in the OGS#GEMS code, to fully exploit the power of parallelized computation platform. For
realistic reactive transport problems, which typically involve dozens of chemical reactions and hundreds of
components, the majority of the computational resources are consumed on the chemistry part. Thus, the
speed-up of the code is close to linear.
In this presentation, we will demonstrate the capability of OGS#GEMS through the Maqarin site, which
serve as a natural analogue for the long term changes of clay rock in contact with hyper-alkaline solutions
leaching from the cement formation. The geochemical setup for the rock mineralogy and the pore water was
calibrated to match measurements from the site investigation. The setup includes several clay and zeolite
minerals, considers cation exchange processes, and employs the latest solid solution model to represent C-S-

H phases in the cement. The numerical model predicts that the marl rock will be clogged after several
hundred years at a distance of 5–10 mm from the contact to the hyper-alkaline solution. Sensitivity analysis
has also been performed to reveal the factors that influence the final clogging profile.
C2-2
LONG-TERM REACTIVE TRANSPORT SIMULATION OF THE INTERACTIONS OF
CORROSION PRODUCTS AND COMPACTED BENTONITE IN A HLW REPOSITORY AND
SENSITIVITY ANALYSES TO KEY PARAMETERS
J. Samper, A. Naves, L. Montenegro, A. Mon & B. Pisani
Civil Engineering School. University of Coruña, Coruña, 15071-Spain
The assessment of the long-term performance of the engineered barrier system of a high-level radioactive
waste repository requires the use of reactive transport models. Samper et al. (2012) [1] presented a 1-D
axisymmetric model for the long-term hydrochemical evolution of the bentonite porewater in the bentonite
barrier of a spent-fuel, carbon-steel canister repository in granite. The model accounts for canister corrosion,
the chemical interactions of corrosion products and bentonite, mineral dissolution/precipitation, Fe 2+ and H +
surface complexation reactions on three types of sorption sites and cation exchange reactions of Ca 2+, Mg 2+ ,
Na + , K + and Fe 2+ . The model considers also the generation of H 2 (aq) which is allowed to diffuse through the
bentonite [2,3,4]. Here we report the improvements of such model which include: 1) Reviewing and updating
the chemical compositions of the bentonite and granite; 2) Adopting carbon-steel corrosion rates derived
from recent experimental data [5]; 3) Extending the time span of the simulations to 1 Ma; 4) Performing
simulations for the reference scenario and for a set of variant scenarios or sensitivity runs. Sensitivity
analyses have been performed for: 1) The corrosion rate; 2) The effective diffusion coefficient, D e, of the
dissolved species in the bentonite; 3) The water flow through the granite at the bentonite/granite interface; 4)
The cation selectivities and 5) The chemical compositions of the bentonite and granite porewater.
Model results indicate that canister corrosion causes a marked increase in pH and the concentration of
dissolved Fe 2+ and a decrease in Eh. Most of the released Fe 2+ diffuses from the canister into the bentonite
where it precipitates mainly as magnetite and to a lesser extent as siderite. Fe 2+ sorbs by surface
complexation on weak sorption sites and undergoes cation exchange. Sorption plays a relevant role in the
geochemical evolution of bentonite. The competition of Fe 2+ and H + for the sorption sites near the
canister/bentonite interface causes several sorption fronts which induce fronts on pH, Eh, the concentration
of dissolved Fe 2+ and mineral dissolution/precipitation. Model results lead to significantly high H 2 (g)
pressures. The reduction of bentonite porosity due to mineral precipitation near the canister/bentonite
interface could be large and result in the clogging of the bentonite pores.
The main conclusions of the sensitivity analyses include: 1) The larger the corrosion rate, the larger the pH,
the larger the concentration of precipitated magnetite near the canister/bentonite, the larger the zone where
corrosion products precipitate (Figure 1) and the larger the H 2 (g) partial pressure; 2) The D e of bentonite
affects the concentration of the dissolved Fe 2+ and the precipitation of the corrosion products (Figure 1); 3)
The computed concentrations of dissolved species, the sorption fronts and the concentrations of precipitated
magnetite and siderite are very sensitive to the water flow in the granite; 4) The thickness of bentonite
affected by pore clogging is sensitive to all the investigated parameters; 5) The cation selectivities affect
mostly the concentration of exchanged cations. However, the computed pH, Eh and the concentrations of
dissolved and precipitated species lack sensitivity to the selectivities; and 6) Model results are sensitive to
the chemical compositions of the bentonite and granite porewaters. The general patterns of pH, Eh, H 2 (g)
pressure and magnetite precipitation, however, are similar to those of the reference run.

80
90
70
Magnetite after 10 6 y
80
Magnetite at 10 6 y
Cumulative precipitation (mol/L)
60
50
40
30
20
0.1 μm/y
0.5 μm/y
1 μm/y
2 μm/y
3 μm/y
5 μm/y
Cumulative precipitation (mol/L)
70
60
50
40
30
20
0.5·De
De
2·De
10
10
0
4.5 4.6 4.7 4.8 4.9 5 5.1 5.2 5.3 5.4 5.5 5.6 5.7
r (dm)
0
4.5 4.75 5 5.25 5.5 5.75 6
r (dm)
Figure 1. Sensitivity of the radial distribution of the concentration of cumulative precipitated magnetite after
t = 10 6 years to the corrosion rate (left) and the effective diffusion, D e , of the bentonite (right).
[1] Samper, J., A. Naves, L. Montenegro, A. Mon and B. Pisani (2012). Long term reactive transport simulations of the
interactions of corrosion products and compacted bentonite in a HLW repository. In: 5 th International Meeting Clays in
Natural and Engineered Barriers for Radioactive Waste Confinement, Montpellier (France).
[2] ENRESA (2005). NF-PRO project. Phenomenological description. Reference Concept (Spent Fuel-Carbon Steel
Canister–bentonite-Granite). Deliverable D5.1.1. Part 1.
[3] Samper, J., C. Lu, and L. Montenegro (2008). Coupled hydrogeochemical calculations of the interactions of
corrosion products and bentonite. Phys Chem Earth 33, S306-S316.
[4] Lu, C., J. Samper, B. Fritz, A. Clement & L. Montenegro (2011). Interactions of corrosion products and bentonite:
An extended multicomponent reactive transport model. Phys Chem Earth 36, 1661–1668.
[5] King, F. (2008). Corrosion of carbon steel under anaerobic conditions in a repository for SF and HLW in Opalinus
Clay. Nagra Technical Report 08-12.
Acknowledgements: The research leading to these results has received funding from the European Atomic
Energy Community’s Seventh Framework Programme (FP7/2007-2011) under grant agreement 232598. This
work was partly funded by ENRESA (Spain), Xunta de Galicia (project 10MDS118028PR) and the CICYT
Project of the Spanish Ministry of Economy and Competitiveness (Project CGL2012-36560). We
acknowledge the contributions of Juan Carlos Mayor from ENRESA.
C2-3
NUMERICAL ANALYSIS OF MULTI-MINERALS TRANSFER BETWEEN THE HOST ROCK
AND GROUNDWATER AGAINST EXPERIMENT FOR NUCLEAR WASTE DISPOSAL
XiaoHui Chen (1) , Steve Thornton (1) , Joe Small (2) , Elizabeth Moyce (3) , Sam Shaw (3) , Toni Milodowski (4) ,
Chris Rochelle (4)
(1) Kroto research institute, Civil and structural Engineering Department, The University of Sheffield, UK, S3
7HQ
(2)
National Nuclear Laboratory, UK
3)
School of Earth, Atmospheric and Environmental Sciences, The University of Manchester, Manchester
M13 9PL
4
British Geological Survey, UK.
In the near field of intermediate and low level waste (I/LLW), the geochemical gradient of the chemical
disturbed zone may strongly affect radionuclide behaviour. Based on UK I/LLW disposal concept, a Portland
Cement based backfill will be placed around the nuclear waste container, and the high alkalinity pore fluid
may leach into the host rock, and then strongly affect the chemistry. Experiments were then established to
investigate the chemical mass transfer between the Borrowdale Volcanic Group (BVG), the lithology of
interest during the NIREX studies of a site at Sellafield, Cumbria, and young near-field porewater (YNFP),
which is from the cement pore fluid.

In this paper, numerical analysis has been conducted to analyze the experiment. BVG rock has 6 kinds of
minerals, with different dissolution rates and kinetics coefficients, meanwhile, some of which is not possible
to be obtained from the experiment. In the modelling process, the modelling of kinetics is strongly limited by
the database of geochemical software. For example, in the widely used geochemical software phreeqc 6 , the
kinetics database includes only 6 rates for kinetics information such as calcite, k-field spar etc., which are
absolutely not enough for modelling of a complex system. To overcome such challenge, a new modelling
approach, ‘mix kinetic-finite equilibrium (MK-FE)’ approach has been developed in this paper.
To model the mineral alteration transfer, some currently known geochemical models assume local
thermodynamic equilibrium 1-4 , which is good for the bulk chemistry of most natural water dominated by fast
reactions. However, some precipitation and dissolution may not be in equilibrium and will be kinetically
controlled 5 . Such processes may be expressed in terms of first or second order kinetic formulations, which
lead to a kinetic approach. The problem associated with kinetics approach is that not all the information of
parameters for kinetics can be obtained such as surface area, initial moles and final moles, reaction rates.
Another very attractive approach is mix kinetic-equilibrium approach, which combines kinetic and
equilibrium laws together to allow modelling for very complex multi-minerals and multi-components
reaction system. Some research has been conducted by using this approach 7,8 . In this approach, some of the
faster reactions can be implemented by using equilibrium approach and some slower reactions are controlled
by kinetics. This is also very good to reuse existing models and solution methods which are developed for
pure thermodynamic equilibrium approach. The challenge of this approach is that a slower reaction’s kinetics
must be available. However, if the reaction is slow, but the kinetics is not known in a mix kinetic-equilibrium
approach, a further development, MK-FE (Figure 1) must be taken, especially for modelling of a complex
reaction system.
The idea of Finite equilibrium comes from the concept of finite element, which has been widely used for
modelling of non-linear solid mechanics system 9-11 . This idea shows that no matter how slowly the reaction
is, in an enough discrete finite equilibrium step (for example 0.001s), the slow chemical reaction can be
viewed as react very fast and in equilibrium. The whole non-equilibrium systems then can be assumed to be
built on numbers of equilibrium reactions. This approach allows any non-equilibrium reaction to be modelled
based on the current models for equilibrium reactions in lots of software, which is useful for multi-minerals
and multi-reactions in which the kinetics of minerals are unknown. It may be also used for testing the
kinetics information from experiment. If the key tracer (one of the dissolution element which is not involved
in precipitation of the secondary phases or other reactions) can be identified, the kinetics of the congruent
dissolution of the mineral can be obtained by MK-FE approach even if there is not actual experiment for
determination of the kinetics only, which happens in most of the multi-minerals dissolution experiment.
The numerical simulation has been compared with the experimental results. The good agreed results show
the capability of the MK-FE approach. In the modelling analysis, CSH phases have been proved to be the
most important secondary phases around CDZ. pH (Figure 2), Mg, Ca changes against time have been
modelled and analyzed. The modelling results all agree well with the experimental results.

Figure1. MK-FE program structure Figure 2 comparison of modelling with experimental pH
[1] J. Westall, J. Zachary, F. Morel, MINEQL-A computer program for the calculation of the chemical equilibrium
composition of aqueous systems, in: Technical Report 86-01, Department of Chemistry, Oregon State University, 1986.
[2] C. Bethke, The Geochemist's Workbench. A User's Guide to Rxn, Act2, Tact, React and Gtplot, in, University of
Illinois, Urbana-Champaign, 1994.
[3] C.M. Bethke, Geochemical reaction modeling. Concepts and applications, Oxford Univ., New York, 1996.
[4] J. van der Lee, HYTEC, un modèle couplé hydro-géochimique de migration de pollutants et de colloïdes, in:
Technical Report LHM/RD/97/02, CIG, École des Mines de Paris, Fontainebleau, 1997a.
[5] K. Soetaert, P.M.J. Herman, J.J. Middelburg, A model of early diagenetic processes from the shelf to abyssal depths,
Geochim. Cosmochim. Acta, 60 (1996) 1019-1040.
[6] D.L. Parkhurst, C.A.J. Appelo, User's Guide to PHREEQC (Version 2)-A Computer Program for Speciation, Batch-
Reaction, One-Dimensional Transport, and lnverse Geochemical Calculations,
http://web.inter.nl.net/users/pyriet/bijlage%206.pdf, (2010).
[7] C. Bethke, Geochemical Reaction Modeling, Oxford University Press, New York, 1996.
[8] J. van der Lee, Thermodynamic and mathematical concepts of CHESS, in: Technical Report LHM/RD/98/39, CIG,
École des Mines de Paris, Fontainebleau, France, 1998.
[9] O.C. Zienkiewicz, The Finite Element Method, 3 ed., McGraw Hill, London, 1977.
[10] Y. Sun, X. Chen, W. Zhong, Symplectic Conservative Integration for Nonlinear Differential Equation, Chinese
quarterly of mechanics, Chinese quarterly of mechanics, 27 (2006).
[11] W. Zhong, X. Chen, Solving shallow water waves with the displacement method, Journal of hydrodynamics
(China), 21 (2006).
C2-4
HIGH-PERFORMANCE REACTIVE TRANSPORT MODELLING OF RADIONUCLIDE
MIGRATION. THE COMSOL-PHREEQC APPROACH.
J. Molinero (1) , Albert Nardi (1) , David García (1) , Cristina Domènech (1) , Mireia Grivé (1) , Benoit Cochepin (2)
(1) AMPHOS 21 Consulting. Passeig Garcia i Faria 49-51. 08019 Barcelona - Spain
(2)
ANDRA. 1-7, rue Jean-Monnet - 92298 Châtenay-Malabry cedex - France
Andra envisages the safe disposal of high-level waste (HLW) and intermediate-level long-lived waste (IL-
LLW) in deep geological storage using a multibarrier system. After concluding in 2005 the feasibility of
deep geological disposal for HLW and IL-LLW, Andra was commissioned by the Planning Act of June 28
2006 to design and implement a storage facility for these wastes. The repository is called Cigéo (Centre
industriel de stockage géologique).
Accurate description of physics and chemical processes taking place in the near-field of HLW cells, and
quantitative evaluation of chemical interactions and migration of radionuclides are two priorities for Andra
R&D activities. In this framework, a new numerical tool for high-performance simulation of coupled
reactive transport processes has been applied, named COMSOL-PHREEQC.
PHREEQC is a freely available computer program for simulating chemical reactions and 1D transport
processes in aqueous systems [1]. It is perhaps the most widely used geochemical code in the scientific
community and is openly distributed. The program is based on equilibrium chemistry of aqueous solutions
interacting with minerals, gases, solid solutions, exchangers, and sorption surfaces, but also includes the
capability to model kinetic reactions with rate equations that are user-specified in a very flexible way by
means of BASIC statements directly written in the input file.
Comsol Multiphysics (COMSOL, from now on) is a powerful Finite Element software environment for
modelling and simulation of a large number of physics-based systems [2]. COMSOL is widely used in
several scientific and technological applications for the modelling of processes based on partial/ordinary
differential equations and algebraic differential equations, which can derive in different coupled and highly
non-linear systems.
The COMSOL-PHREEQC tool is driven by a JAVA interface which uses the COMSOL Java API and the
IPhreeqC dynamic library [3], accessed through the JNA wrapping library, which is able to communicate
both simulators following a Sequential Non-Iterative Approach. Furthermore, some optional coupling

etween both codes like temperature effects on chemistry and density or porosity changes due to chemical
reactions has been enabled. The chemical step has been parallelized using shared memory to optimize the
simulations on multi-core processors. The numerical tool has been extensively verified by comparison of
simulation results of 1D, 2D and 3D benchmark examples solved with other reactive transport simulators.
In this work, we show a full-scale simulation of the geochemical evolution of a HLW cell during 100,000
years. The migration of 4 radionuclides (Uranium, Caesium, Selenium and Technetium) has been also
simulated, fully coupled with the physical and chemical evolution of the system. The main retention
mechanisms of those radionuclides have been studied in the different components of the model domain
(vitrified glass, stainless steel, bentonite, cement and geological formation). Figure 1 shows the geometrical
domain included in the simulation, and Figure 2 shows an example of exchanged (retained) Caesium
evolution in the geological formation during 100,000 years.
Figure 1. Geometrical domain included in the reactive transport simulations
Figure 2. Evolution of exchanged Caesium in the geological formation during 100,000 years
[1] D.L. Parkhurst, K.L. Kipp, Peter Engesgaard, and S.R. Charlton. Phast, a program for simulating
ground-water flow, solute transport, and multicomponent geochemical reactions. U.S. Geological
Survey Techniques
[2] COMSOL Multiphysics Version 4.2a (www.comsol.com).

[3] S.R Charlton and D.L. Parkhurst. Modules based on the geochemical model PhreeqC for use in
scripting languages and reactive-transport calculations. Environmental Modelling & Software, 2010.
SESSION 7
B5: FIELD AND LARGE SCALE EXPERIMENTS
TEASING OUT BIOGEOCHEMICAL CONTROLS ON THE TRANSPORT OF
PLUTONIUM: EXAMPLES FROM FIELD AND LABORATORY STUDIES
A.B Kersting, M. Zavarin, B.A. Powell , P. Zhao, J. Begg, Z. Dai, M. Boggs (USA)
LONG TERM DIFFUSION EXPERIMENT LTD PHASE I: EVALUATION OF
RESULTS AND MODELLING
V. Havlová, A. Martin, J. Landa, F. Sus, M. Siitari-Kauppi, J. Eikenberg, P. Soler, J.
Miksova (Czech Republic, Switzerland, Finland, Spain)
TRACERS BEHAVIOR INTO THE GROUNDWATER OF A FRENCH NUCLEAR
WASTE DISPOSAL
O. Péron, S. Razafindratsima, A. Piscitelli, C. Gégout, V. Schneider, F. Barbecot, E.
Giffaut, J-C. Robinet, G. Montavon (France, Canada)
DENITRIFICATION PROCESSES IN THE MONT TERRI IN SITU BITUMEN
NITRATE CLAY INTERACTION EXPERIMENT AND MODELLING EFFECTS ON
Eh AND RADIONUCLIDE MIGRATION
J. Small, L. Abrahamsen, A. Albrecht, N. Bleyen, E. Valcke (UK, France, Belgium)
B5-1
B5-2
B5-3
B5-4
B5-1
TEASING OUT BIOGEOCHEMICAL CONTROLS ON THE
TRANSPORT OF PU: EXAMPLES FROM FIELD AND LABORATORY STUDIES
A.B Kersting (1) , M. Zavarin (1) , B.A. Powell (2) , P. Zhao (1) , J. Begg (1) , Z. Dai (1) , and M. Boggs (1)
(1) Glenn T. Seaborg Institute, Physical & Life Sciences, Lawrence Livermore National Laboratory, PO Box
808, L-231, Livermore, CA 94550
(2) Dept. of Environmental Engineering and Earth Sciences, Clemson University, SC, 29625
In 2010, the worldwide inventory of plutonium (Pu) was estimated at ~1900 metric tons with increases
averaging approximately 70-90 metric tons/year from the nuclear fuel industry 1 . In 2012, further accounting
estimated that this global inventory consisted of about 500 metric tons of separated Pu, produced nearly
equally between the civilian nuclear power industry and weapons related activities 2 . Despite this sizable
inventory, significant uncertainty remains as to how to safely store and isolate long-lived, toxic
radionuclides, such as Pu, from the biosphere for the thousands of years necessary for its decay. Once the
engineered barrier system is breached, the natural geologic media will need to provide an additional longterm
barrier to prevent the radioactivity from reaching the biosphere. Developing a credible strategy for
isolating and safely storing high-level nuclear waste is one of the most pressing environmental challenges of
the 21 st century.
Currently, scientists cannot reliably predict under what conditions Pu will or will not migrate once it
breaches an engineered barrier system preventing accurate long-term assessment of risk to human health.
Although several field studies have shown Pu to migrate associated with colloidal particles 3-5 , other field
studies suggest that colloidal transport is not the dominant mechanism for Pu migration 6 . In addition, at high
concentrations, under conditions expected in the near-field of a high-level waste repository, Pu can
hydrolyze and form a Pu-oxide (PuO 2 ) colloid; yet their importance in fate and transport of Pu is not well
understood. Without a more complete understanding of the dominant biogeochemical processes that may
inhibit or facilitate the Pu transport, we cannot adequately model its transport behavior and achieve
confidence in ultimately designing repositories that minimize its migration.

Despite the gaps in our understanding, recent field and laboratory experiments are helping to shape our
conceptual knowledge and significant progress has been made in understanding the behavior of Pu in the
subsurface. In this paper, we will summarize our current understanding of the dominant biogeochemical
processes controlling Pu transport by discussing recent field and laboratory studies. Field studies include a
review of weapons facilities where contamination of large quantities of Pu have been deposited and migrated
in the subsurface and where the geologic and source term environments are different (e.g. Nevada Test Site,
Rocky Flats, Mayak Russia, and Hanford Site). Laboratory studies will include experiments evaluating how
Pu behaves in both high concentration and low concentration environments with inorganic and organic
colloids. A further focus will be on how to apply this understanding to help design strategies to minimize Pu
mobility in high-level waste repositories. It is becoming clear that the fate and transport of Pu depends not
only on the initial chemical form at the source, but the geochemistry and geohydrology of the source location
and subsequent geochemistry along the transport pathways.
[1] D. Albright, K. Kramer, Bulletin Atomic Scientist 2004, 14-16;
[2] R. Ewing, W. Runde, T. E. Albrecht-Schmitt, MRS Bulletin 2010, 35, 859-866. International Panel on Fissile
Materials, in Sixth annual report of the International Panel on Fissile Materials, IPFM, Princeton, NJ, 2011, p.
42.
[3] A. B. Kersting, D. W. Efurd, D. L. Finnegan, D. J. Rokop, D. K. Smith, J. L. Thompson, Nature 1999, 397, 56-
59;
[4] A. P. Novikov, S. N. Kalmykov, S. Utsunomiya, R. C. Ewing, F. Horreard, A. Merkulov, S. B. Clark, V. V.
Tkachev, B. F. Myasoedov, Science 2006, 314, 638-641;
[5] P. H. Santschi, K. A. Roberts, L. D. Guo, Environmental Science & Technology 2002, 36, 3711-3719.
[6] P. Zhao, M. Zavarin, R. N. Leif, B. A. Powell, M. J. Singleton, R. E. Lindvall, A. B. Kersting, Appl Geochem
2011, 26, 308-318.
B5-2
LONG TERM DIFFUSION EXPERIMENT LTD PHASE I: EVALUATION OF RESULTS AND
MODELLING
V. Havlova 1)* , A. Martin 2) , J. Landa 1) , F. Sus 1) , M. Siitari-Kauppi 3) , J. Eikenberg 4) , P.Soler 5) ,
J. Miksova 6)
1) Dept. of Fuel Cycle Chemistry, UJV Rez, a.s., Hlavni 130, 250 68 Řež, Czech Republic,
2) NAGRA, Hardstrasse 73, 5430 Wettingen, Switzerland,
3) Laboratory of Radiochemistry, University of Helsinki, A.I. Virtasen aukio 1, 00014 Helsinki, Finland,
4) Paul Scherrer Institute, 5232 Villigen, Switzerland,
5) IDAEA-CSIC, Jordi Girona, 18-26, 08034 Barcelona, Spain,
6) RAWRA, Dlažděná 6, 110 00 Prague, Czech Republic
Radionuclide diffusion into crystalline rock matrix is considered to be one of the most important processes
that can significantly reduce the amount of non-sorbing radionuclides reaching the biosphere from a deep
geological repository of radioactive wastes. A systematic approach, including in-situ experiments in the
Grimsel Test Site (GTS) (www.grimsel.com), supporting laboratory experiments and modelling was carried
out within Phase I of the Long Term Diffusion project (LTD) in order to better understand the processes that
control matrix diffusion. This paper presented will focus mainly on techniques and procedures used in the
frame of the in-situ experiment and supporting laboratory programme.
The in-situ experiment was performed in crystalline rock massive within the radiation controlled zone of the
GTS. A cocktail of radioactive tracers consisting of 3 H, 22 Na, 131 I and 134 Cs was injected into a packed-off
test-interval in a 56mm diameter borehole drilled up to 8 m depth into the undisturbed granitic matrix. The
radionuclides represented different types of tracers ranging from conservative, slightly sorbing and sorbing
as well as the inclusion of an anion. The solution was continuously circulated in the experimental system for
26 months. Small samples of the circulation solution were regularly taken in order to determine the temporal
pattern of radionuclide activity decrease in the test interval. Short-lived 131 I, was used to try to observe the
concentration decrease due to diffusion during the period shortly after the injection by online γ-spectrometry.
However this could not be carried out continuously for more than a few weeks owing to noise interference
from the coupled cooling device when used in the tunnel.

A supporting laboratory program was also carried out in parallel. Sorption batch experiments and throughdiffusion
experiments were performed with core material from the experimental borehole and with the same
tracers as used in the in-situ experiment. From this, sorption distribution coefficients K d (m 3 /kg) and effective
diffusion coefficients D e (m 2 /s) were determined.
After 26 months the borehole was overcored. The rock was divided into segments and carefully sealed for
further analysis. Some of the rock segments were analysed for radionuclide concentration at UJV Rez, a.s.
and some at Helsinki University. 22 Na and 134 Cs were analysed, using γ-spectroscopy. A special technique
was applied for 3 H analyses involving a combination of distillation and subsequent liquid scintillation
spectroscopy. Diffusion profiles within the rock segment were thus obtained. The profiles showed that 3 H
reached a distance of 17 cm from the borehole contact surface. 22 Na migrated up to 7 cm and 134 Cs up to 1.5
cm distance respectively.
Post-mortem modelling was carried out to evaluate the results of both activity decrease in the in-situ
experimental tank and activity profiles determined in the rock profiles. UJV Rez, a.s. used its own code. The
concentration change rate at a point in a one dimensional representation of the porous rock layer was
described according to Fick´s second law. Calculations considered symmetry around the borehole axis (1D
cylindrical coordinates) and included radioactive decay. The results from the code were compared with other
codes (CrunchFlow, Comsol and GoldSim).
The activity measurements in the rock profiles revealed that it was not possible to recover the full tracer
activity in the rock. It was also evident that a borehole disturbed zone (BDZ) had to be taken into account to
fit the modelled curves with the experimental observations. Generally, effective diffusivity D e and sorption
distribution coefficients K d for the tracers in the BDZ were larger than the respective values in the bulk rock.
K d values in the bulk rock were largest for Cs due to sorption and smallest for HTO where no sorption takes
place. However, 3 H seemed to display higher K d values in the BDZ than expected due to a process not fully
understood. This phenomenon should be investigated in more detail in future.
B5-3
TRACERS BEHAVIOR INTO THE GROUNDWATER OF A FRENCH NUCLEAR WASTE
DISPOSAL
O. Péron (1) , S. Razafindratsima (1) , A. Piscitelli (1) , C. Gégout (1) , V. Schneider (2) , F. Barbecot (3) , E. Giffaut (4) , J-
C. Robinet (4) , G. Montavon (1)
(1) Laboratoire SUBATECH, IN2P3/CNRS/EMN/Université de Nantes, 4, rue Alfred Kastler, 44307 Nantes,
France
(2)
ANDRA, DI/CA/QSE, service « Qualité Sûreté et Environnement », Centre de l’Aube BP7, 10200
Soulaines-Dhuys, France
(3)
GEOTOP, Université du Québec à Montréal, Montréal, Québec, Canada
(4)
ANDRA, 1-7 rue Jean Monnet, 92298 Châtenay-Malabry Cedex, France
Within the context of nuclear waste disposal ageing, the potential chemical spillage of radionuclides into the
environment must be evaluated. Indeed, a release of radioactive species in geological media would be a
threat to environment and human health. The CSA (Centre de Stockage de l’Aube) of Andra (The French
National Radioactive Waste Management Agency) is one of the French Nuclear Waste disposals, dedicated
to wastes of low and intermediate activity with short life, where the geological media is one of the
containment dispositive. The waste disposal is built upon a sandy aquifer which has a limited extent and
laying other a thick clayey layer protecting the underlying aquifers. Moreover, this geological barrier has the
advantage of having a single and well controlled outlet. This paper reports on a field experiment to study the
behavior of D 2 O, Br - , I - and Li + species into the CSA’s groundwater. D 2 O is used to probe the whole pore
volume, Br - to get information about the anionic exclusion volume, and I - and Li + to study the potential of the
barrier to retain reactive species.
A hydrogeological model was built in order to simulate the behavior of species thanks to hydraulic heads and
drawdowns recorded from the site. To perform the model, two in-situ forced gradient tracings with Dirac

injection were considered: case (i) D 2 O (water tracer), Li + and Br - (anionic water tracer) used in 2003 1 and
case (ii) D 2 O, Br - and I - used in 2012. The first one (2003), with D 2 O and Br - , allowed us to calibrate the
breakthrough curve of the model output. The latter was then used to predict the behavior of tracers in case
(ii) and was validated through the in-situ experiment.
The Figure 1 (a) presents the study zone with the location of the boreholes dedicated to the injection and the
sampling during each in-situ tracing. The groundwater flow direction (N320) was also indicated with a
calculated hydraulic gradient from 1% to 1.4%. Here, the essential information of the tracing zone is that in
the case (i) the species were injected in DS43 and sampled in TS05 and in case (ii) the species were injected
in DS70 and sampled in DS69 and TS05. In both cases, a pumping was performed in TS05 (0.9 m 3 /h) with,
in addition, a pumping in DS69 (0.1 m 3 /h) for case (ii).
In the Figure 1 (b), breakthrough curves of the calculated and observed concentrations for D 2 O, Br - and I -
were shown. For both cases, the modeling of the deuterium transport led to the hydrodynamic parameters of
the aquifer. A permeability K = 1.8×10 -5 m/s and an effective porosity we = 38 % with a longitudinal
dispersivity α L = 0.01 m were obtained. Thus, the model allows us to reproduce accurately the behavior of
D 2 O whatever the injection borehole (either DS43 or DS70), the sampling well (either DS69 or TS05) and
the distance between them. Therefore, the determination of hydrodynamic properties for the other tracers (Br -
and I - ) was relevant. A maximum concentration of the conservative tracer (Br - ) was observed around one day
before D 2 O one. Such a difference in the retention time could be explained through anionic exclusion.
Indeed, fitting the model curves with Br - breakthrough data gave a permeability K = 1.8×10 -5 m/s and an
effective porosity we = 35.5 %. Such a result was attributed to an exclusion pore volume of 2.5 % which
would correspond to the negatively charged clay surface 2 . Otherwise concerning I - tracer, same retention
time and effective porosity were found with respect to Br-. This indicates a Kd (coefficient distribution)
value of zero.
A slight but significant sorption was observed in the case of Li + ; the difference in restitution time with
respect to D 2 O is quantitatively described by a Kd value of 0.18 mL/g. This retention parameter could be
recovered by geochemistry modelling considering an exchange reaction process with the clay mineral phase
present in the formation (clayey part: 5-15% 3 ). This result illustrates the importance of the clay fraction in
the retention of cationic species.
(a)
(b)
Figure 6. (a) Tracing zone with the location of the boreholes. (b) Calculated (line) and observed
concentrations of deuterium, bromide and iodide in cases (i) and (ii).

B5-4
DENITRIFICATION PROCESSES IN THE MONT TERRI IN SITU BITUMEN-NITRATE-CLAY
INTERACTION EXPERIMENT AND MODELLING EFFECTS ON E h AND RADIONUCLIDE
MIGRATION
J. Small (1) , L. Abrahamsen (1) , A. Albrecht (2) , N. Bleyen (3) , E. Valcke (3)
(1) UK National Nuclear Laboratory, Chadwick House, Birchwood Park, Warrington WA3 6AE, UK
(2) ANDRA, 1-7 rue Jean-Monnet, FR-92298 Châtenay-Malabry Cedex, France
(3) W&D Expert Group, SCK•CEN, Boeretang 200, BE-2400 Mol, Belgium
In several countries, such as France and Belgium, bituminised waste containing high quantities of NaNO 3
has been manufactured to immobilise intermediate-level long-lived radioactive waste (ILW) that is planned
for disposal in deep geological clay formations. The Bitumen-Nitrate-Clay Interaction experiment (BN) [1]
currently being performed at the Mont Terri Rock Laboratory (Switzerland) provides an in situ examination
of redox reactions expected between nitrate released from the waste and mineral and organic electron donors
present in clay host rocks or as waste degradation products. Nitrate has the potential to affect the chemical
conditions and the speciation and transport behaviour of redox sensitive radionuclides (e.g. Se, Tc, U, Np,
Pu) within the engineered barrier and in part of the host rock, which has initially reducing conditions.
BN consists of a vertical borehole rigged with downhole equipment containing three packed-off intervals,
each lined with a cylindrical sintered stainless steel filter screen to allow contact with the surrounding clay.
Each interval is connected to a stainless steel water circulation unit, equipped with water sampling
containers, circulation pumps, and flow meters. Continuous monitoring of the nitrate and nitrite
concentrations using a UV-Vis spectrophotometer, and pH and E h can be carried out [1]. The circulating
water is also sampled periodically for further geochemical analysis and microbial characterisation.
To aid the interpretation of the experiments the Generalised Repository Model (GRM) [2] has been used to
simulate the microbial processes and the resulting effect on E h . The GRM is a 2-dimensional biogeochemical
reactive transport model that considers the main anaerobic microbial growth processes of: nitrate reduction
(denitrification, including formation of nitrite), Fe(III) reduction, sulphate reduction, acetogenesis
(fermentation), and methanogenesis. The GRM considers a range of organic substrates for these processes
including acetate as a source of carbon for biomass growth and as an electron donor. Reduced Fe(II) and
sulphide may also act as electron donors for the denitrification processes. The GRM considers the effect of
these anaerobic processes in mediating reducing conditions and calculates an E h based on the
concentration/presence of electron acceptor species (e.g. NO 3 - , Fe(III), SO 4 2- ). The E h is used as input to a
chemical speciation and mineral reaction (equilibrium) calculation that simulates changes in pH, mineral
reaction, and radionuclide speciation.
The experimental data for a nitrate and acetate case (Figure 1) show rapid reaction of nitrate with acetate that
is followed by a slower decrease in nitrate concentration (upper figure). The initial reaction with acetate
produces nitrite (lower figure). Modelling is based on a hypothesis that during the second phase of reaction
mineral electron donors such as FeCO 3 , FeS 2 or dissolved organics present in the clay are used. The model
simulates that the majority of mineral oxidation occurs on the surface of the borehole, with some nitrate and
nitrite diffusing to a distance of 15 mm into the excavation damaged zone (EDZ) of the clay. The model also
indicates that the denitrification reactions involving Fe(II) phases result in the precipitation of Fe(OH) 3 . The
model simulates an E h of around 650 mV when nitrate is present and an E h of -100 mV for the zone where
Fe(OH) 3 is precipitated. The undisturbed clay has a modelled E h of -200 mV controlled by the
sulphate/sulphide couple.
The model has been used to simulate the longer term chemical response to the nitrate injection and the
potential effect on radionuclide speciation. After nitrate has dropped below a threshold value, the
precipitated Fe(OH) 3 that resulted from denitrification is again reduced by electron donors that slowly diffuse
from the Opalinus Clay. Under the transient nitrate and Fe(III) controlled redox conditions the modelled
solubility of Tc and U increases, leading to enhanced migration into the EDZ. However, the effect is limited
by the slow transport rates through the undisturbed clay during the short period of nitrate release.

1
Nitrate concentration (mM)
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
Br diffusion De 2e-11m 2 /s
D 2 O diffusion De 1e-10m 2 /s
"EDZ" model 15mm porosity 0.3,
De 1e-10m 2 /s
Reaction with co-injected
Acetate
Reaction with
FeCO 3 and FeS 2
0
0 2 4 6 8 10 12 14 16 18 20
days after injection
Nitrite concentration (mM)
1
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
Br diffusion De 2e-11m 2 /s
D 2 O diffusion De 1e-10m 2 /s
0 2 4 6 8 10 12 14 16 18 20
days after injection
Figure 1 Nitrate and nitrite concentration (symbols) during in situ denitrification in a borehole in the
Opalinus Clay, solid lines represent modelled nitrate and nitrite concentration for different diffusive transport
models.
Acknowledgements
Financial support was provided by the Mont Terri consortium (in particular by the project partners: ANDRA
(France), IRSN (France), NAGRA (Switzerland), and SCK•CEN (Belgium)), the UK National Environment
Research Council BIGRAD Consortium Research Grant and the National Nuclear Laboratory.
[1] Bleyen, N., Smets, S., and Valcke, E. BN experiment: status and raw data report of phases 15 and 16. Mont Terri
Project Technical Note 2009-49. Mont Terri Consortium, St. Ursanne, Switzerland, 2012.
[2] Small J., Nykyri M., Helin M., Hovi U., Sarlin T. and Itävaara M., (2008). Experimental and Modelling
Investigations of the Biogeochemistry of Gas Generation from Low and Intermediate Waste. Applied
Geochemistry, 23(6): 1383–1418.

SESSION 8
C5: SAFETY ASSESSMENT AND REPOSITORY CONCEPTS
PRELIMINARY SAFETY ASSESSMENT OF THE GORLEBEN SITE
G. Bracke K. Fischer-Appelt (INVITED) (Germany)
C5-1
COMMON CHALLENGES FOR THE UK AND JAPANESE GEOLOGICAL DISPOSAL
PROGRAMMES
E.M. Scourse, T. Beattie, S.M.L. Hardie, I.G. McKinley (Switzerand, UK)
CHALLENGING THE FUNCTION OF ENGINEERED BARRIERS IN THE REPOSITORY
FOR STORAGE OF SPENT NUCLEAR FUEL
W. Forsling (Sweden)
C5-1
C5-2
C5-3
PRELIMINARY SAFETY ASSESSMENT OF THE GORLEBEN SITE
G. Bracke
GRS mbH, Schwertnergasse 1, 50667 Cologne, Germany
The safety requirements governing the final disposal of heat-generating radioactive waste in Germany
implemented by the federal ministry of environment, natural conservation and nuclear safety (BMU) from
2010 consider the fundamental objective to protect man and environment against the hazard from radioactive
waste [1]. Unreasonable burdens and obligation for future generations shall be avoided. The main safety
principles are concentration and inclusion of radioactive and other pollutants in a containment providing rock
zone. Any release of radioactive nuclides may increase the risk for men and the environment only negligibly
compared with natural radiation exposure. No intervention or maintenance work shall be required during the
post-closure phase. The retrieval / recovering of the waste shall be possible.
The objective of the project “Preliminary Safety Analysis of the Gorleben site” (German acronym: VSG)
was to compile and assess the available data on exploration of the Gorleben site and research on disposal in
salt rock. An important objective was also to identify the needs for future R&D-work and further Gorleben
site investigations. In addition, the feasibility of the methodology for use within a future site selection
procedure was assessed.
The VSG was composed of four main working topics:
Basic data: This included the geological site description [2], [3] and the geological evolution over the next
million years [4]. The radioactive waste was compiled and classified which could presumably be emplaced
in a repository at the Gorleben site according to the current situation in Germany with its phase out of
nuclear energy (June 2011) [5]. A safety and demonstration concept was developed [6] based on the safety
requirements which were applied accordingly.
Repository concepts: Repository concepts were developed considering operational safety, long-term safety
and retrieval / recovering of the waste. Several alternatives, such as storage in drifts or boreholes and
different types of canisters, were projected. The retrieval of disposed waste from boreholes during the
operational phase was projected for the first time [7].
System analysis: The features, events and processes were described, compiled and used to derive scenarios
for the evolution of the systems and linked to their probability [6]. Geomechanical analyses showed that
integrity of the geological barrier (containment providing rock zone) can be demonstrated for 1 million years
[8], [9]. This applies for external (e.g. glaciation) and for internal events and processes such as decay heat or
gas generation. Similarly, the seals for shafts and drifts were designed and analysed as a concept. The
radiological consequences were analysed using numerical models for the transport of the liquid and gas
phase (two phase transport) in the long-term safety analysis [9].

Synthesis: The feasibility of repository concepts for containment of radionuclides according to the safety
requirements is assessed. Uncertainties and additional R&D is shown [11].
The safety and demonstration concept, which was generated during the course of the project, was suitable to
ensure the compatibility of a repository concept with these safety requirements. The generated design of the
repository is technically feasible. Nevertheless, some necessary assumptions referred to the status of
geological exploration, the reliability of the technical constructions and some inherent uncertainties, which
cannot further minimized presently. If these assumptions are met, the repository systems are assessed to be
robust.
Further optimizing strategies regarding the repository systems and design are conceivable. A repository
layout such as placing the structural components farther away from the drift seals would likely result in a
lower C-14 flow through the drift seals. Furthermore, implementing a void volume as a sink (infrastructure
area backfilled with gravel) might hinder gas flow through the shaft seals. The use of gas tight casks for the
structural components (like POLLUX®) could confine volatile radionuclides for decades to centuries.
Lessons learnt are:
• The possible release of gaseous radionuclides, the two phase-flow processes and the subsequent
model for radiation exposure require additional research and development.
• The containment providing rock zone can be minimized by an iterative process but has to be assessed
initially as an guess.
• The handling of less probable but interdependent FEP has to be improved.
• Requirements concerning the mobilization of other pollutants and groundwater flow were missing.
• The preliminary safety analysis for the Gorleben site identified important tasks for research and
development. This would have not been possible by a generic safety analysis. A safety analysis
should be repeated in time intervals and for different geological conditions.
[1] BMU (2010): Sicherheitsanforderungen an die Endlagerung wärmeentwickelnder radioaktiver Abfälle (Stand:
30.09.2010). Bundesministerium für Umwelt, Naturschutz und Reaktorsicherheit (BMU), Bonn, Germany.
[2] Weber J.R., Mrugalla S., Dresbach C. and Hammer J. Preliminary Safety Analysis of the Gorleben Site:
Geological Database – 13300, in: WM2013. 2013, Phoenix.
[3] Bornemann O., Behlau J., Fischbeck R., Hammer J., Jaritz W., Keller S., Mingerzahn G. & Schramm M.:
Description of the Gorleben Site, part 3: Results of the geological surface and underground exploration of the
salt formation. Bundesanstalt für Geowissenschaften und Rohstoffe (2011), 223 pp., 50 fig., 7 tab., 5 app.,
Hannover, ISBN 978-3-9813373-6-5.
[4] Mrugalla S., Geowissenschaftliche Langzeitprognose, Bericht zum Arbeitspaket 2, GRS-275, ISBN 978-3-
939355-51-9, 2011.
[5] Peiffer F., McStocker B., Gründler D., Ewig F., Thomauske B., Havenith A., Kettler J., Abfallspezifikation
und Mengengerüst, Bericht zum Arbeitspaket 3, GRS-278, ISBN 978-3-939355-54-0, 2011.
[6] Mönig J., Beuth T., Wolf J., Lommerzheim A., Mrugalla S., Preliminary Safety Analysis of the Gorleben Site:
Safety Concept and Application to Scenario Development Based on a Site-specific Features, Events and
Processes (FEP) Database – 13304, in: WM2013. 2013, Phoenix.
[7] Bollingerfehr W., Filbert W., Herold P., Lerch C., Müller-Hoeppe N., Charlier F., Kilger R., Technical Design
and Optimization of a HLW-Repository in the Gorleben Salt Dome including Detailed Design of the Sealing
System in: WM2013. 2013, Phoenix.
[8] Eickmeier R., Heusermann S., Knauth M., Minkley W., Nipp H.-K. and Popp T., Preliminary Safety Analysis
of the Gorleben Site: Thermo-mechanical Analysis of the Integrity of the Geological Barrier in the Gorleben
Salt Formation - 13307, in WM2013. 2013, Phoenix.
[9] Nipp H.-K., Heusermann S. (2000): Erkundungsbergwerk Gorleben, Gebirgsmechanische Beurteilung der
Integrität der Salzbarriere im Erkundungsbereiche EB1 für das technische Endlagerkonzept 1
(Bohrlochlagerung, BSK3). Bericht, Bundesanstalt für Geowissenschaften und Rohstoffe (BGR), Hannover,
Germany.
[10] Kock I., Larue J., Fischer H., Frieling G., Navarro M., Seher H., Results from one and two phase fluid flow
calculations within the Preliminary Safety Analysis of the Gorleben Site – 13310, in: WM2013. 2013, Phoenix.
[11] Fischer-Appelt K., Baltes B., Buhmann D., Larue J., Mönig J.: Synthesebericht für die VSG, Bericht zum
Arbeitspaket 13, Vorläufige Sicherheitsanalyse für den Standort Gorleben, GRS 290, 2013 (in preparation).

C5-2
COMMON CHALLENGES FOR THE UK AND JAPANESE GEOLOGICAL DISPOSAL
PROGRAMMES
Ellie M. Scourse (1) , Tara Beattie (1) , Susie M.L. Hardie (2) , Ian G. McKinley (2)
(1) MCM Consulting, Bristol, UK
(2) MCM Consulting, Baden-Dättwil, Switzerland
Historically, the Japanese geological disposal programme has been most closely linked to that in Switzerland due
to a focus on vitrified HLW and consideration of a wide range of siting environments in tectonically active
settings. The UK programme has never had a specific national partner, despite being involved in a wide range of
EU initiatives.
More recently, however, there has been a convergence between the Japanese and UK programmes with the former
expanding its inventory to include spent fuel and long-lived ILW and the latter moving towards a volunteering
approach to siting. Although the UK is much less tectonically active, potentially more relevant is the likely
complex geology and coastal settings of volunteer sites in these island nations. Especially in terms of developing
post-closure safety cases, the challenges in developing more realistic and robust models of radionuclide release
and migration are very similar and, given complementary resources and infrastructure, could form the basis of
more extensive future collaboration.
The paper will expand on the key issues associated with developing effective disposal systems for the waste
inventories in both countries and identify commonalities and contrasts. This will take into account also the
resources, infrastructure and intellectual property available to the implementing organisations – NUMO in Japan
and NDA RWMD in the UK – and also the likely timescales of project implementation which will be used as a
basis for highlighting synergies and potential for mutually beneficial collaboration. Specific topics covered will
include:
• Next generation disposal concepts which tailor co-disposal of a range of waste types to specific volunteer
site conditions
• Testing of safety, practicality and QA of novel designs, especially in underground research facilities
• Development of tools and databases for realistically assessing post-closure performance, with special
consideration of explicitly including site heterogeneity and its evolution with time (e.g. due to climate
change)
• Validation of models and databases using focused field and analogue studies
• Holistic management of the massive, multidisciplinary knowledge base that supports such projects by
integrating state-of-the-art methodology for managing Knowledge, Requirements, Issues, Quality,
Change, etc.
• Communication with stakeholders, with special emphasis on advance, web-based communication
platforms.

C5-3
CHALLENGING THE FUNCTION OF ENGINEERED BARRIERS IN THE REPOSITORY FOR
STORAGE OF SPENT NUCLEAR FUEL
Willis Forsling
Swedish National Council for Nuclear Waste, 103 33 Stockholm, Prof. Emeritus of Inorganic Chemistry at
Lulea University of Technology, 97187 Lulea
High-level nuclear waste primarily consists of spent (used) fuel from the nuclear power plants’ reactors. The
spent fuel is long-lived and has a very high level of radiation. It takes about 100 000 years for the
radioactivity to approach the level in natural uranium deposits.
The main alternative for the final storage of spent nuclear fuel in Sweden is based on three protective
barriers: a copper canister, a buffer of natural clay and the surrounding rock. The spent fuel rods are encased
in copper canisters with an inner container of cast iron, and the canisters are emplaced, surrounded by blocks
of bentonite, in a tunnel system at a depth of about 500 meters in the bedrock.
The method involves a number of scientific and technical challenges with respect to the long period of time,
the temperature variations in the repository due to long-term global climate changes, the absorption of
ground water and swelling of the clay, the initial temperature gradient between the canister and the
surrounding rock and the change of the environment from aerobic to anaerobic conditions. These events may
cause specific complications to the barrier system implying e.g. copper corrosion, cementation and erosion of
the clay buffer and leakage of radioactive nuclides.
In my presentation I will describe the function of the barrier system of the Swedish KBS-3 method in detail
and report on the impact of the threats from a scientific (thermodynamic) point of view and how these threats
are supposed to be managed. There is a comprehensive research program in progress to secure the long-term
safety of the repository and to keep up the confidence of the method in the public opinion and the political
system in Sweden.
Furthermore, I will report on some challenges and recent observations in the scientific and technical research
activities in Sweden focusing on the technical barriers to protect the high-level radioactive spent fuel.
In March 2011 SKB has applied to construct a repository for storage of spent nuclear fuel in Sweden. The
application is now reviewed by the Swedish Radiation Safety Authority and the Environmental Court with
respect to environmental impact and technical durability.
Quite recently the Swedish National Council for Nuclear Waste has sent a report to the Environmental Court
on the need for supplementing the application from SKB regarding a number of critical issues.
The presentation will include detailed justifications of the Council´s statements with reference to the
engineered barrier system in the KBS-3 method.

SESSION 9
A3: COMPLEXATION WITH INORGANICAND ORGANIC
LIGANDS
INVESTIGATION OF THE SYSTEM Ln(III)/An(III)-B(OH) 3 -ORGANICS
J. Schott, M. Acker, J. Kretzschmar, A. Barkleit, S. Taut, V. Brendler, G. Bernhard (Germany)
A3-1
INFLUENCE OF THIOL COMPOUNDS ON STABILIZING BULK U(IV) SPECIES AND
SPECTROSCOPIC DETERMINATION OF U(VI) AND U(IV) UNDER THIOL-RICH
CONDITIONS
W. Cha, E. C. Jung, H-R. Cho, S. Y. Lee , M. H. Baik (Korea)
COMPLEXION OF Nd(III)/Cm(III) WITH BORATE IN DILUTE TO CONCENTRATED
ALKALINE NaCl, MgCl 2 AND CaCl 2 SOLUTIONS: SOLUBILITY AND TRLFS STUDIES
K. Hinz, M. Altmaier, X. Gaona, T. Rabung, D. Schild, C. Adam, H. Geckeis (Germany)
A3-2
A3-3
A3-1
Investigation of the System Ln(III)/An(III)-B(OH) 3 -Organics
J. Schott 1 , M. Acker 2 , J. Kretzschmar 1 , A. Barkleit 1 , S. Taut 2 , V. Brendler 1 , G. Bernhard 3
1 Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology, Dresden, Germany
2 Technische Universität Dresden, Central Radionuclide Laboratory, Dresden, Germany
3 Technische Universität Dresden, Department of Chemistry and Food Chemistry, Radiochemistry, Dresden, Germany
Actinides, such as Am, Cm and Pu, will define the long-term radiotoxicity of spent nuclear fuel in a
nuclear waste repository. Concerning the required long-term safety and risk assessment of the storage of
high-level radioactive waste the understanding of fundamental physicochemical reactions (sorption,
complexation, solid/colloid formation) between actinides and components of a repository (host rock,
container material, decomposition/corrosion products, dissolved organic/inorganic matter, …) is essential.
Concerning the migration behavior of trivalent actinides boric acid (B(OH) 3 ) and (poly)borates are of great
interest, particularly in host rock formations based on salt, because they can appear in considerable amounts
due to naturally occurring deposits and anthropogenic sources (waste containers, borosilicate glasses) [1].
Until now, only little is known about the system B(OH) 3 /borates and trivalent actinides (or lanthanides, like
Nd(III) or Eu(III), as chemical analog). A first work concerning this system was carried out by Borkowski et
al. [1]. They investigated the complexation in the Nd(III)/tetraborate system due to solubility experiments
and found a moderate Nd(III) borate complex competing with the carbonate complexation.
The present work describes a new approach to determine the formation constant in the An(III)/Ln(III)
borate system. It is known that B(OH) 3 forms esters with small organic ligands, like salicylate [2]. The borate
ester structure, occurring in the acidic pH range, is used as analog structure for the monoborate ion B(OH) 4 - ,
which only exists in the alkaline pH range. Thus, the strong hydroxide complexation or precipitation of
trivalent actinides/lanthanides can be avoided.
For the first time the complex formation between a borate ester of salicylate and Eu(III) was investigated.
The determination of the formation constant of the Eu(III) borate ester bases on a spectrophotometric
(TRLFS, time-resolved laser-induced fluorescence spectroscopy) titration. A formation constant of log =
2.07 ± 0.15, I = 0.1 M, was calculated (under validation).
Furthermore, in presence of polyborates (detected in solution due to 11 B-NMR) a solid Eu(III) borate
species can be documented, which already occurs in slightly acidic (pH > 5.5) salt solutions (0.1 - 3 M
NaClO 4 /NaCl). This solid formation was investigated by filtration experiments and TRLFS (c Eu(III) = 3∙10 -
5 M). The filtration proves the solid formation. A characteristic band splitting of the Eu luminescence

transitions (Fig. 1) and an extensive increase of the Eu luminescence lifetime (Fig. 2)
luminescence properties of the Eu(III) borate precipitate in comparison with the dissolved Eu(III) species.
The solid formation depends on the polyborate concentration. The smaller the initial B(OH) 3 (polyborate)
concentration the slower the formation of the solid species (Fig. 2). Furthermore, the solid formation depends
on the salt concentration and type (NaClO 4 /NaCl). In the NaClO 4 medium a minimum of the formation rate
at ~ 0.5 M NaClO 4 can be detected. For the NaCl medium, the formation rate decreases with increasing salt
concentration, probably due to a competing Eu(III) chloride complexation.
Normalized intensity [a.u.]
0.024
0.021
0.018
0.015
0.012
0.009
0.006
0.003
0.000
day:
20
8
0
spectrum of
the dissolved
Eu polyborate
species
spectrum of the
solid Eu borate
species
mixed spectrum
of solid and
dissolved
Eu (poly)borate
species
570 580 590 600 610 620 630 640
Wavelength [nm]
Fig. 1: Eu luminescence spectra of the dissolved
and
solid Eu borate (c Eu(III) = 3∙10 -5 M)
Luminescence lifetime τ [µs]
700
600
500
400
300
200
solid
Eu borate
species
[B(OH) 3
]:
0.7 M
0.6 M
0.5 M
0.4 M
0.3 M
0.2 M
100
dissolved
Eu polyborate
0 10 20 30 40 50 60 70 80 90
species
100
Time [d]
Fig. 2: Eu luminescence lifetime
on
initial [B(OH) 3]; 0.1 M NaClO 4, pH 6, 3∙10 -5 M Eu(III)
[1] M. Borkowski et al., Radiochim. Acta 98, 577 (2010).
[2] L. H. J. Lajunen et al., Pure & Appl. Chem. 69, 329 (1997).
A3-2
INFLUENCE OF THIOL COMPOUNDS ON STABILIZING BULK U(IV) SPECIES AND
SPECTROSCOPIC DETERMINATION OF U(VI) AND U(IV)
UNDER THIOL-RICH CONDITIONS
W. Cha (1) , E. C. Jung (1) , H-R. Cho (1) , S. Y. Lee (2) , M. H. Baik (2)
(1) Nuclear Chemistry Research Division, (2) Radioactive Waste Disposal Research Division
Korea Atomic Energy Research Institute
P.O. Box 105, Yuseong, Daejeon 305-600, Republic of Korea
Under anoxic and reducing conditions similar to those common in deep geological repositories, uranium (U)
species are present at lower oxidation states. It is generally known that U(IV) is a major reduced form in
aquatic environments and that it is much less mobile than U(VI) due to the low solubility of U(IV) hydrous
oxide (UO 2 •xH 2 O)(am). Microbe-mediated metal reduction processes play a role in the immobilization of
uranium. The anoxic growth of bacteria results in a reduction of U caused by the formation of U(IV)
aggregates 1,2 . However, in the presence of inorganic or organic ligands U(IV) may become mobile due to the
formation of soluble complexes or nanoparticles. Recent studies have reported the formation of extracellular
uranite nanoparticles [1], which may be suspended in the resultant filtrate and supernatant solutions obtained
for an analysis of dissolved species. In previous studies of the biotic transformation U using sulfate-reducing
bacteria, it was demonstrated that U(IV) is present in a dissolved form in a bulk culture media, particularly at
a later stage of cell incubation 3 . In a separate study of U(VI) complexation with a model ligand bearing thiol
functionality (i.e., thiosalicylate), we found that with an excess amount of ligand the solubility of U(VI) is
greatly enhanced at higher pH levels (< ~9) 4 . Thus, we speculate that the accumulation of hydrogen sulfide
or thiols as metabolic products of bacteria is related to the dissolution process of U(IV) during the course of
bacterial growth.
In this study, we first developed a TRLFS (time-resolved laser-induced fluorescence spectroscopy)-based
analytical method capable of determining both U(VI) and U(IV) concentrations dissolved in aqueous
samples 5 . To eliminate the strong luminescence quenching effects the method utilizes thiol oxidation
reactions using the selected oxidant, monopersulfate. It was shown that the method improves the limit of
detection (LOD) of the total U analysis (~ nM levels) without requiring conventional sample pre-processing
(ashing) procedures for the removal or decomposition of the quenchers (i.e., thiols). Further, it enables an

indirect estimation of dissolved U(VI) and U(IV) concentrations (LOD, ~ 1 µM) by comparing the U(VI)
concentration before and after the oxidation reaction using a selected oxidant, monopersulfate. Using this
method, it is shown that the concentration of dissolved U(IV) in a bacterial culture media increases to
approximately 20 μM as illustrated in Fig. 1.
The objective of this study is to identify the major factor influencing the bulk concentration of U(IV) at a low
U(IV) concentration (< 0.15 mM). Because the pK a values of thiols (e.g., cysteine, cysteamine and
thiosalicylate) are between 8.0 and 8.5, certain effects resulting from complexation or colloid stabilization by
thiol groups may arise and induce an increase in the bulk U(IV) concentrations. Thus, the effect of thiol
species is our primary interest. Typically, after mixing U(IV) with thiols, the solution pH was slowly
adjusted up to a target pH between 3 and 7 (i.e., a weakly acidic or neutral pH). The equilibrium U
concentration in the filtrate obtained using a membrane filter (0.2-μm pore size) was then measured using the
above method. Overall, the measured concentrations of dissolved U species increase with an increase in the
concentration of the thiols. The thiol-dependent changes and the effects of the presence of U(VI) and
carboxyl ligands will be presented in detail. However, once the U(IV) aggregate forms at a higher pH (e.g.,
8.0), the U(IV) concentration in the filtrate solution is found to be about 1 μM or less, even with an excess
amount of thiol, e.g., 20 mM. This may indicate that the formation of nano-size UO 2 colloids is blocked or
that other forms of insoluble thiol-U(IV) aggregates are created at a higher pH. The elemental compositions
of the solid precipitates will be characterized and reported in the presentation. We believe that the results
from this macroscopic approach will extend the knowledge of the migration behavior of U(IV) in nearneutral
aqueous environments.
40
35
30
[U total
]
[U(IV)]
25
[U] (µM)
20
15
10
5
0
D3 D4 D5 D6 D13 D18
Media Sampling Point
Figure 1. The measured concentrations of dissolved U total and U(IV) in filtered samples periodically collected
from a bacterial (Desulfovibrio desulfuricans) culture. The initial concentrations of U, Cys, and FeSO 4 were
50 µM, 2 mM, and 2 mM, respectively.
[1]. Suzuki, Y., Kelly, S. D., Kemner, K. M., Banfield, J. F. (2002). "Nanometre-size products of uranium
bioreduction." Nature 419: 134.
[2]. Lovely, D. R., Phillips, E. J. P., Gorby, Y. A., Landa, E. R. (1991). "Microbial reduction of uranium." Nature 350:
413-416.
[3]. Lee, S. Y., Baik, M. H., Cho, H.-R., Jung, E. C., Jeong, J. T., Choi, J. W., Lee, Y. B., Lee, Y. J. (2013). "Abiotic
reduction of uranium by mackinawite (FeS) biogenerated under sulfate-reducing condition." J. Radioanal. Nucl. Chem.
DOI 10.1007/s10967-013-2438-6.
[4]. Cha, W., Cho, H.-R., Jung, E. C. (2013). "Studies of aqueous U(VI)-thiosalicylate complex formation via UV-Vis
absorption spectrophotometry, TRLFS, and potentiometry." Polyhedron http://dx.doi.org/10.1016/j.poly.2013.02.080.
[5]. Cha, W., Lee, S. Y., Jung, E. C., Cho, H.-R., Baik, M. H. (2013). "Determination of U(VI) and U(IV)
concentrations in aqueous samples containing strong luminescence quenchers using TRLFS and quencher oxidation
reactions." Analytica Chimica Acta submitted.

A3-3
COMPLEXION OF Nd(III)/Cm(III) WITH BORATE IN DILUTE TO CONCENTRATED
ALKALINE NaCl, MgCl 2 AND CaCl 2 SOLUTIONS: SOLUBILITY AND TRLFS STUDIES
K. Hinz, M. Altmaier, X. Gaona, Th. Rabung, D. Schild, C. Adam, H. Geckeis
Institute for Nuclear Waste Disposal (KIT-INE), Karlsruhe Institute of Technology, Karlsruhe, Germany
The prediction of the long-term safety of a nuclear waste repository requires an adequate knowledge of the
chemical behavior of actinides in aqueous solutions. This understanding must be extended to concentrated
salt brines when considering repositories in rock salt formations or certain sedimentary bedrocks with high
salinity. The An(III) and An(IV) oxidation states are favored under the reducing conditions expected in a
deep underground repository. Boron can be present in high concentrations in the intruding brines of saltbased
repositories as well as a part of the emplaced waste. In view of complexation/solubility phenomena
that may influence the mobilization of actinides into the biosphere, the interaction of trivalent actinides with
borate is a relevant research topic.
A comprehensive experimental study was conducted under inert gas (Ar) atmosphere at 22 ± 2°C. Samples
were prepared in 0.1 - 5.0 M NaCl, 0.25 - 3.5 M MgCl 2 and 0.25 - 3.5 M CaCl 2 with 4 mM ≤ [B] tot ≤
400 mM and 7 ≤ pH c ≤ 13. Nd(III) solubility was investigated in independent batch experiments from
undersaturation with ~ 10 mg Nd(OH) 3 (am) solid phase per sample. Samples were equilibrated for up to 142
days; pH c and [Nd] (ICP-MS) were monitored at regular time intervals. After reaching equilibrium
conditions, selected solid phases were characterized by XPS, XRD and Raman spectroscopy. The aqueous
phase was further investigated by TRLFS with 10 -7 M Cm(III) in 0.1-5.0 M NaCl, 0.25-3.5 M MgCl 2 and
0.25-3.5 M CaCl 2 with 4 mM ≤ [B] tot ≤ 400 mM and pH c = 8. Additional Cm(III) TRLFS spectra were
collected at pH c = 12 for CaCl 2 systems. Provided the very limited knowledge on borate speciation under
highly saline conditions, aqueous borate speciation was investigated by 11 B-NMR in 0.1/5.0 M NaCl,
0.25/3.5 M MgCl 2 and 0.25/3.5 M CaCl 2 solutions at pH c = 8 and 12 with 40 mM ≤ [B] tot ≤ 400 mM.
The solubility of Nd(OH) 3 (am) in NaCl, MgCl 2 and CaCl 2 solutions with 4 mM ≤ [B] tot ≤ 400 mM shows no
significant increase of Nd(III) concentration compared to borate-free systems [1]. The thermodynamic model
derived in [1] for borate-free systems can therefore be used to establish upper-limit concentrations for the
systems investigated in the present work. Most interestingly, a clear solubility decrease at pH c < 9 and [B] tot
> 40 mM was observed in several samples independent of the background electrolyte and salt concentration.
The decrease of Nd(III) solubility with increasing [B] tot as well as the clear change of the slope of the
solubility curve indicate the formation of a secondary borate-bearing Nd(III) solid phase. This solid phase
transformation constitutes a potential An(III) retention mechanism and is currently under further
investigation.
Significant peak shifts in the Cm(III) fluorescence spectra under mildly alkaline conditions (pH c < 9) with
increasing [B] tot , indicate the formation of Cm(III)-borate complexes. The collected TRLFS spectra do not
allow to resolve the number and stoichiometry of the Cm(III)-borate species forming, mostly due to the
broad fluorescence bands and the complex and largely unclear boron speciation in the conditions of this
study. No effect of borate is observed in the Cm(III) spectra at pH c = 12, indicating that Cm(III)-borate
complexes are outcompeted by the strong hydrolysis of Cm(III) under hyperalkaline conditions. 11 B-NMR
studies in dilute to concentrated NaCl, MgCl 2 and CaCl 2 solutions show the dependence of the aqueous
borate speciation on pH c , boron concentration and ionic strength.
The present study indicates that borate species are only weakly coordinating ligands for trivalent actinides
under the investigated experimental conditions. Solubility experiments do not show any pronounced
solubility increase due to borate complexation but indicate a significant decrease in An(III) concentrations
likely related to solid phase transformation processes. Given the high uncertainties regarding borate
speciation and the composition of the solubility limiting An(III) solid phase, no attempt is made to propose a
comprehensive thermodynamic model at present.

[1] Neck, V., Altmaier, M., Rabung, Th., Lützenkirchen, J., Fanghänel, Th. (2009) Thermodynamics of trivalent
actinides and neodymium in NaCl, MgCl 2 and CaCl 2 solutions: solubility, hydrolysis and ternary Ca-M(III)-OH
complexes. Pure Applied Chemistry, 81, 1555-1568.
[2] Rabung, T., Altmaier, M., Neck, V., Fanghänel, Th. (2008). A TRLFS study of Cm(III) hydroxide complexes in
alkaline CaCl 2 solutions. Radiochimica Acta, 96, 551-559.

SESSION 10
POSTER SESSION II
SPONSORED BY AWE
PA2 SOLID SOLUTION AND SECONDARY PHASE
FORMATION
PA2-1
PA2-2
PA2-3
PA2-4
PA2-5
RECRYSTALLIZATION OF BARITE IN THE PRESENCE OF RADIUM
- A MICROSCOPIC AND SPECTROSCOPIC STUDY
M. Klinkenberg, F. Brandt, U. Breuer, D. Bosbach (Germany)
ISOTOPIC EXCHANGE OF 45 Ca AND 14 C ON CALCITE
J. Lempinen, S. Kallio, M. Hakanen, J. Lehto (Finland)
ACTINIDE BORATES FORMATION AT NORMAL AND EXTREME CONDITIONS
E.V. Alekseev, S. Wu, S. Wang, W. Depmeier, T.E. Albrecht-Schmitt (Germany, China, USA)
MECHANISM OF ENHANCED INCORPORATION OF STRONTIUM IN CALCITE
VIA AMORPHOUS CALCIUM CARBONATE PRECIPITATION
J. Littlewood, I.T. Burke, S. Shaw, C.L. Peacock, D. Trivedi (UK)
THERMODYNAMIC AND STRUCTURAL DATA FOR THE RADIUM AND BARIUM
SULFATE SYSTEM
N. Torapava, H. Hedström, C. Ekberg (Sweden)
PA2-1
RECRYSTALLIZATION OF BARITE IN THE PRESENCE OF RADIUM – A MICROSCOPIC
AND SPECTROSCOPIC STUDY
M. Klinkenberg 1) , F. Brandt 1) , U. Breuer 2) , D. Bosbach 1)
1) Institute of Energy and Climate Research, IEK-6 Nuclear Waste Management, Forschungszentrum
Jülich GmbH, 52425 Jülich, Germany
2) Central Institute for Engineering, Electronics and Analytics, ZEA-3 Analytics,
Forschungszentrum Jülich GmbH, 52425 Jülich, Germany
The possible solubility control of Ra by coprecipitation of a Ra x Ba 1-x SO 4 solid solution has been
demonstrated in many cases e.g. Doerner & Hoskins, 1925 [1]. However, an open question is whether a Ra
containing solution will equilibrate with solid BaSO 4 under repository relevant conditions due to barite
recrystallization. Here, Ra enters a system in which barite is in equilibrium with the aqueous solution.
Previous studies have indicated that uptake of Ra is not limited by pure adsorption at close to equilibrium
conditions but involves a significant fraction of the bulk solid [2,3]. So far only macroscopic data indicate
that barite may fully recrystallize to radiobarite. Here we present new microscopic (SEM) and spectroscopic
(ToF-SIMS) data to investigate the recrystallization of barite in the presence of Ra. The study is focused on
the spatial distribution of Ra in the newly formed radiobarite and the effect of Ra on the barite
recrystallization, especially the morphology and grain size evolution.
Microscopic and spectroscopic data from batch recrystallization experiments at room temperature (RT) are
presented. A pure barite solid was put into contact with an aqueous solution (0.1 n NaCl) with an initial
Ra/Ba ratio of 0.3 (5⋅10 -6 mol/L Ra) and neutral pH. The solid/liquid ratio was 5 g/L and 0.5 g/L. A
Sachtleben barite (SL) which consists of blocky crystals with a particle size of > 10 µm and a narrow grain
size distribution was used for the recrystallization experiments. The SL has a specific surface area of

0.17 m²/g. Furthermore, Aldrich barite (AL) was used, which consisted of rounded particles (< 2 µm)
forming agglomerates. The particles showed smooth crystal surfaces with small pores. The specific surface
area was 1.7 m²/g. The SEM and ToF-SIMS data were combined with Gamma spectrometry and ICP-MS
analysis of Ra and Ba concentration in solution with time.
Based on gamma spectrometry, it was observed that SL incorporates Ra faster than AL. The final
concentration plateau is similar, which may indicate the approach of equilibrium.
A wide grain size distribution with a range of 200 nm to 3000 nm and a mean particle size of 500 nm was
determined via SEM observation for AL. A coarsening of AL crystals due to Ostwald ripening and an
additional effect due to the presence of Ra was detected. The mean particle size of AL increased to a value of
> 800 nm after 443 days with Ra.
The grain size analysis of the SL barite didn’t show significant changes with time and no influence of the
presence of Ra. The morphology of the barite crystals changed slightly. Grains were grown together forming
large chains and the crystal surfaces became smoother.
TOF-SIMS was carried out on SL samples taken after 443 days. Fig. 1 shows an overlay of the integrated
elemental signal of Ba (top) and Ra (bottom) with the complementary electron microscopy image. The
integrated Ra concentration corresponds with the size of the barite particles, i.e. all particles contain Ra in
similar concentrations. A depth profile of the respective Ba and Ra concentrations was reconstructed from
the TOF-SIMS data (Fig. 1b). A homogenous Ra concentration distribution was observed for the large barite
crystal in the middle of the SEM image (Fig. 1a).
The Ba/Ra ratio was calculated from several TOF-SIMS measurements, using the integrated elemental
signals (Fig. 1c). Mass balance calculations for the SL suggest a mole fraction of X RaSO4(s) = 2.3 . 10 -3 . The
Ra/Ba intensity distribution of Fig. 1c has its maximum between 2 and 4 . 10 -3 , corresponding well with the
macroscopic results.
Thus, the different results of Ra uptake kinetics may be a result of different uptake and recrystallisation
mechanisms.
a) b) c)
Fig. 2: a) Combined SEM and TOF-SIMS image of SL 0.5 g/L after 443 days of recrystallization. The blue
colour indicates the integrated TOF-SIMS signal of the respective element; b) Depth profiles reconstructed
for the indicated X-Y areas of the left images (a). c) Ra/Ba intensity ratio as calculated from the TOF-SIMS
measurements.
[1] Doerner, H. A. & Hoskins, W. M., Journal of the American Chemical Society, 1925, 47, 662-675
[2] Bosbach, D.; Böttle, M. & Metz, V., SKB Technical Report TR-10-43 Waste Management, Svensk
Kärnbränslehantering AB, 2010

[3] Curti, E.; Fujiwara, K.; Iijima, K.; Tits, J.; Cuesta, C.; Kitamura, A.; Glaus, M. & Müller, W. Geochimica et
Cosmochimica Acta, 2010, 74, 3553-3570
PA2-2
ISOTOPIC EXCHANGE OF 45 Ca AND 14 C ON CALCITE
J. Lempinen, S. Kallio, M. Hakanen, J. Lehto
Laboratory of Radiochemistry, Department of Chemistry, University of Helsinki, P.O. Box 55, 00014
University of Helsinki, Finland
Radiocarbon ( 14 C) is formed in nuclear fuel and surrounding metal materials by neutron activation reaction
of stable nitrogen 14 N(n,γ) 14 C. The speciation of radiocarbon in spent nuclear fuel is unknown, but it is
assumed that radiocarbon will be in the form of elemental carbon and/or insoluble carbides. These forms of
carbon may be oxidized to more soluble species such as carbonate due to the radiolysis of water caused by
the radiation of the spent nuclear fuel. Because of the strongly reducing conditions in the deep bedrock it is
possible that the soluble carbonates are reduced and methane is formed. The methane dissolved in
groundwater is not retained by minerals and can thus passively migrate to the upper layers of the bedrock
with the groundwater flow. Because of the higher redox potential or microbial activity of the upper bedrock
the methane may be again oxidized to carbonate.
Radiocarbon in carbonate species can be retained by calcite mineral (CaCO 3 ) because of isotopic exchange.
In equilibrium state calcite is dissolved and precipitated continuously. Thus radiocarbon in contact with
calcite-bearing rock fractures can be incorporated into calcite by precipitation. Radiocarbon can thus form a
solid solution with the calcite mineral in bedrock through isotopic exchange. The isotopic exchange of 14C
between calcite and groundwater has been reported in karst environment. [1] In the present study the isotopic
exchange in calcite was investigated using 45 Ca 2+ and 14 CO 3 2- tracers.
Experiments were carried out in deionized water and 0.1M NaCl solution. 1.0 g of reagent-grade CaCO 3 was
placed in centrifuge tubes and 29 to 35 mL water or NaCl solution was added. The samples were left to stand
until equilibrium between dissolved and solid CaCO 3 was attained. The solutions were then spiked with the
45 Ca or 14 C tracer and tubes were closed with caps. The solution was allowed to equilibrate for 1 to 42 days.
During this time the centrifuge tubes were gently shaken every few days. For sampling the tubes were
centrifuged to separate the liquid phase and 0.25 or 0.50 mL subsamples of the supernatant were taken for
14 C and 45 Ca activity measurements. The centrifuge tubes were then closed with caps again and shaken to
continue the experiment.
For the activity measurements the samples were mixed with Optiphase HiSafe 3 liquid scintillation cocktail.
Activity standards were prepared using deionized water in equilibrium with CaCO 3 and the radioactive
tracers. The activity standards and the samples were then measured using Perkin Elmer Tri-Carb liquid
scintillation spectrometer. From the results of the activity measurements the fraction of both radionuclides in
the solid phase was calculated.
Figure 1 shows the fraction (%) of initial 45 Ca transferred to calcite phase as a function of time. From Figure
1 it can be seen that the increase rate of 45 Ca in solid phase is rather slow after 20 days but no steady-state
has been attained even after 42 days (experiments are being continued). The fractions of 45 Ca in solid phase
at 42 days were 17 % and 15 % for 0.1M NaCl solution and deionized water, respectively.
Experiments are being conducted to determine the rate of the isotopic exchange reaction of 14 C between
groundwater and calcite. Isotope exchange reactions in dissolution-precipitation systems are assumed to be
first-order reactions [1]. This assumption will be verified considering this particular case and the rate
constant of the reaction will be defined. Experiments will be done in different temperatures as temperature
greatly affects reaction rate. This data can be used for modelling and safety analysis of the final disposal of
spent nuclear fuel.
In addition to water and NaCl solutions the isotopic exchange of 14 CO 3 2- on calcite will be investigated in
different ground water model solutions (e.g. Allard, fresh, saline and brine). The aim of the study is to

provide information on whether the transport of radiocarbon in carbonate form in bedrock will be delayed
with respect to groundwater flow due to the isotopic exchange between the liquid and solid phase.
20
15
k [%]
10
5
0,1 M NaCl
H2O
0
0 10 20 30 40
t [d]
Figure 1. The fraction of 45 Ca in calcite phase (k) as a function of time (t).
[1] R. Gonfiantini, G. M. Zuppi, Chem. Geol. 197, 319 (2003).
PA2-3
ACTINIDE BORATES FORMATION AT NORMAL AND EXTREME CONDITIONS
Evgeny V. Alekseev 1,2) , Shijun Wu 1,3,4) , Shuao Wang 5,6) , Wulf Depmeier 3) , Thomas E. Albrecht-Schmitt 7)
1 Institute for Energy and Climate Research (IEK-6), Forschungszentrum Jülich GmbH, 52428 Jülich,
Germany
2 Institut für Kristallographie, RWTH Aachen University, 52066 Aachen, Germany
3 Institute of Geosciences, Kiel University, 24118 Kiel, Germany
4 Guangzhou Institute of Geochemistry, Chinese Academy of Sciences, 511 Kehua Street, 510640 Guangzhou,
P. R. China
5 Actinide Chemistry Group, Chemical Sciences Division, Lawrence Berkeley National Laboratory, Berkeley,
CA 94720, USA
6 Department of Chemistry, University of California, Berkeley, Berkeley, California 94720, USA
7 Department of Chemistry and Biochemistry, Florida State University, 102 Varsity Way, Tallahassee,
Florida 32306-4390, USA
One nuclear waste management strategy is the vitrification of high level radioactive wastes using borosilicate
glasses. Although most of the uranium is recovered from wastes during reprocessing, the produced
borosilicate glasses still contain some uranium and plutonium as well as minor transuranium elements such
as neptunium and americium as well as up to 16.9% B 2 O 3 . Some nuclear waste repositories possess strongly
reducing conditions, with U(IV), Pu(III/IV) and Np(IV) as the dominant oxidation states. Therefore, it could
be beneficial to obtain information about actinide borate species in certain scenarios.
We have extended the knowledge of actinide borate chemistry significantly using boric acid as a reactive
flux over the course of the past three years. The An n+ cations coordination generated by oxo-borate polyhedra
in the phases obtained from boric acid flux (normal conditions experiments) is quite similar. All phases (with
only one exception, β-UO 2 B 2 O 4 ) possess six oxygen atoms donated by BO 3 flat triangles and/or BO 4
tetrahedra to an equatorial plane of the An n+ coordination environment. The An n+ cations are located in
triangular holes of the polyborate 2D or 3D frameworks. This coordination keeps in different actinide borates
obtained from low/mild temperature conditions, even with different valence of An atoms (from +3 to +6). It
is noteworthy that in all these phases the boron to actinides atomic ratios are equal or larger than 1.

Actinide borates, obtained from extreme conditions of pressure and temperature (high-temperature/highpressure,
HT/HP), demonstrate significant differences in chemistry and structure. First of all, the chemical
composition is different. We have synthesized several phases under extreme conditions and in all pure
uranium borates the B/U ratio was less than 1. This observation shows a significant difference in
thermodynamic stability of actinide borates with excess of the boron atoms at HT/HP conditions. The second
point is the dramatic increase of the structural complexity of uranium borates from HT/HP reactions. The
structure of the layers in HT/HP phases is totally different from the phases obtained under normal and mild
conditions. U polyhedra are edge- or corner-sharing to form infinite 2D sheets. All boron oxo-polyhedra are
BO 3 triangles and this is quite different from the phases obtained at normal conditions where BO 4 groups are
the major content of the polyborate nets. The number of possible UO 2 2+ to BO 3 coordination is quite high in
HT/HP phases. For example, 10 different coordination schemes were observed in only three new HT/HP
uranyl borates, viz. K 12 [(UO 2 ) 19 (UO 4 )(B 2 O 5 ) 2 (BO 3 ) 6 (BO 2 OH)O 10 ] ⋅nH 2 O, K 4 [(UO 2 ) 5 (BO 3 ) 2 O 4 ]⋅H 2 O, and
K 15 [(UO 2 ) 18 (BO 3 ) 7 O 15 ]. The structural and chemical complexity of uranyl borates obtained from HT/HP
conditions is quite unusual for actinide borates and for actinide oxo-salts chemistry in general.
PA2-4
MECHANISM OF ENHANCED INCORPORATION OF STRONTIUM IN CALCITE
VIA AMORPHOUS CALCIUM CARBONATE PRECIPITATION
Janice Littlewood 1 , Dr IT Burke 1 , Dr S Shaw 2 , Dr CL Peacock 1 , Dr D Trivedi 3
1 University of Leeds, 2 University of Manchester, 3 National Nuclear Laboratory
Radioactive liquid wastes and sludges have leaked from surface storage ponds at nuclear sites, resulting in
extensive ground contamination. For example, at Sellafield, approximately 20 million m 3 of soil is thought to
be contaminated with 90 Sr, 137 Cs, 238 U and 99 Tc identified as the most important contaminants 1 . The long halflives
and mobility of these radionuclides within ground water make them particularly problematic. Hence,
there is a need for low cost, non-invasive, remediation techniques which will prevent the spread of
radionuclides in the environment, and/or produce stable waste forms suitable for geological disposal within
cement grout filled containers (e.g. during pump and treat remediation).
This work presents a novel in situ remediation technique to immobilise Sr-90 into calcium carbonates, via
precipitation of an amorphous calcium carbonate (ACC) precursor.
Group II metals such as Sr and Mg are known to substitute for Ca within the calcium carbonate crystal.
However the substitution of Sr for Ca is limited by size constraints as Sr has a larger ionic radius than Ca
(1.3 Å vs. 1.0 Å). To accommodate Sr, the calcium carbonate crystal distorts resulting in 6-fold Sr
coordination in calcite 2 . As such the amounts of Sr incorporated into calcium carbonates are traditionally
low, only ~ 0.2% Sr has been incorporated during the formation of calcite 3 . These levels of trace metal
incorporation into calcium carbonates are understood in terms of a distribution coefficient (k d ) where the
trace metal/Ca ratio in the precipitated calcium carbonate mineral is controlled by the trace metal/Ca ratio in
solution 3 .
This work considers the effects of the reaction pathway on total Sr-substitution into calcium carbonates (K d
values) and the bonding environment of Sr within both the ACC precursor and subsequent crystalline phases.
Sr-containing calcium carbonates were prepared via the ‘Leeds method’ 4 where CaCl 2 is rapidly reacted with
Na 2 CO 3 in solution. By partially substituting SrCl 2 for CaCl 2 , a range of Sr-calcium carbonates were formed,
via an ACC precursor. Briefly, equal volumes of SrCl 2 /CaCl 2 and Na 2 CO 3 were rapidly stirred in a beaker at
room temperature (RT). A white gelatinous ACC precursor formed within seconds. In the case of 10% Sr
substitution for Ca, ACC transformed to metastable vaterite within minutes and after 3 hours vaterite fully
transformed to calcite. (Figure 1).

Figure 1 – XRD patterns showing the crystallisation of Sr (10%)-ACC to vaterite (v) and calcite (c)
Samples were taken at regular time intervals and vacuum-filtered through 0.2μm polycarbonate filters,
quenched with isopropyl alcohol (IPA), dried at room temperature and stored in sealed containers. The solid
phases were analysed by a number of complimentary techniques, namely XRD and SEM.
The presence of Sr in the starting solutions produced the same reaction pathway as was anticipated for
crystallisation in the absence of Sr (Rodriguez-Blanco et al., 2011), namely:
ACC → vaterite → calcite
As the aqueous Sr concentration decreased, the phase transitions, although unchanged, were achieved more
quickly.
In samples crystallised via the ACC precursor, the K d values were calculated to be 5 mol % Sr which is
considerably higher than values for Sr incorporation by inorganic, non-ACC, growth mechanisms of < 1mol
% Sr-substitution (Tesoriero and Pankow, 1996).
The mechanism for the incorporation of trace metals, via the ACC route to crystalline polymorphs such as
vaterite and calcite will only be a viable remediation strategy for radionuclides, such as 90 Sr, if the
incorporation is into stable crystallographic sites within the crystal lattice. If incorporation occurs at
interstitial sites, there is a danger the radionuclides could be released back into the environment during
crystal ripening. It is therefore important to ascertain the bonding environment of Sr incorporated into
vaterite and calcite via the ACC route. Proposals have been granted for beamtime at Diamond and ESRF to
determine the Sr incorporation mechanism using Sr K-edge EXAFS. It is anticipated that uptake to ACC will
be via incorporation into the poorly ordered structure. Incorporation of Sr into vaterite and calcite phases is
likely to be as a result of isomorphic substitution of Sr for Ca within the calcite lattice.
If this is the case, incorporation of trace metals into calcium carbonates, via an amorphous calcium carbonate
(ACC) precursor, offers a potential in situ remediation treatment for radionuclides such as 90 Sr.
1 HUNTER, J. 2004. SCLS Phase 1 - Conceptual model of contamination below ground at Sellafield, BNFL.
2 PARKMAN R.H., C., J.M. LIVENS, F.R., VAUGHAN, D.J 1998. A study of the interaction of strontium ions in
aqueous solution with the surfaces of calcite and kaolinite. GEOCHIMICA ET COSMOCHIMICA ACTA 62 62, 1481-
1492.
3 RODRIGUEZ-BLANCO, J. D., SHAW, S. & BENNING, L. G. 2011. The kinetics and mechanisms of amorphous
calcium carbonate (ACC) crystallization to calcite, via vaterite. Nanoscale, 3, 265-271.
4 TESORIERO, A. J. & PANKOW, J. F. 1996. Solid solution partitioning of Sr 2+ , Ba 2+ , and Cd 2+ to calcite. Geochimica
Et Cosmochimica Acta, 60, 1053-1063.

PA2-5
THERMODYNAMIC AND STRUCTURAL DATA FOR THE RADIUM AND BARIUM SULFATE
SYSTEM
N. Torapava, H. Hedström, C. Ekberg
Department of Chemical and Biological Engineering, Nuclear Chemistry, Chalmers University of
Technology, SE-41296, Gothenburg, Sweden
Radium chemistry accounts for little bit more than one century and radium, being a naturally occurring
element, was not discovered until 1898 in trace amounts by M. and P. Curie. Due to its radioactive nature
few data and reliable experiments exist in the literature and the number of research groups at universities
able to work with relevant amounts are few these days. Radium is one of the radioactive elements that
dominate the naturally occurring radioactive material (NORM) and its technically enhanced equivalent
(TNORM). The main anthropogenic sources of radium are oil extraction and natural gas production,
manufacturing of phosphoric acid, uranium mining and the storage of spent nuclear fuel [1-3]. When
performing a safety analysis of a final repository for spent nuclear fuel of the underground type planned for
e.g. in Sweden, radium will dominate the dose to man in the long term perspective. In all the cases
mentioned above modelling of the radium behaviour is one of the paths to understand and minimise the
doses to the public either living in areas where radium contamination exists or working with radium
containing material.
In this work the radium sulphate crystal structure has been investigated using XRD and, for the first time,
EXAFS. The unit cell parameters, a = 9.07 Å, b = 5.52 Å, c = 7.28 Å and V = 364.48 Å 3 belonging to
orthorhombic space group were determined and support the fact that radium sulphate is isostructural with
barium-, strontium- and lead sulfates. The study of coprecipitation and recrystallization of the radium and
barium sulphate systems has been done. The reactions of the radium, barium and strontium co-precipitation
systems in sulfate media are showing an Arrhenius behavior, i.e. the relationship between ln(k) and 1/T is
linear in the temperature range of 10 – 30 ºC. These findings show that it is possible for radium to coprecipitate
with barium, strontium and lead in sulphate media to form a substitution solid solution. The
activation energy of the pure radium, barium and strontium sulphates decreases in the order Sr>Ba>Ra,
which may be correlated to their ionic size, since this effects the ability to lose their hydration water. The
activation energy (E a ) is lower for the precipitation of pure radium than for radium in a mixture of radium
and barium.
System E a /R [K] ln(A)
100% Ba 7524±206 21.22±0.74
50% Ba 50% Ra
100% Ra
6182±180
5623 ±228
16.64±0.64
15.00±0.68
The study of the kinetics of 223 Ra/ 133 Ba recrystallization on the surface of synthesized barium sulfate crystals
has been studied. The use of short-lived 223 Ra isotope (half-life 11.4 days) allows reaching concentrations
below 10 -13 M as found in nature and avoids generation of long-lived radioactive waste. The concentrations
of 223 Ra 2+ and 133 Ba 2+ spikes are far below the calculated solubility of corresponding radium and barium
sulphates. A decrease of simultaneously added spikes of 223 Ra 2+ and 133 Ba 2+ is fast within 400 hours and
slows down at longer time, figure 1. The addition of the second 223 Ra 2+ spike after 42 days allows detect a
mechanism of recrystallization, i.e. while radium concentration of the second spike decreases, 133 Ba 2+
concentration (no additional barium added) first, increases before it starts to decrease again. This may
indicate the dissolution of a previously formed radiobarite, which releases barium and the formation of a new
radiobarite after addition of the second radium spike.

7.E-11
3.E-13
[ 133 Ba], mol∙dm -3
6.E-11
5.E-11
4.E-11
3.E-11
2.E-11
1.E-11
System 1'
pH = 1.3
3.E-13
2.E-13
2.E-13
1.E-13
5.E-14
[ 223 Ra], mol∙dm -3
1.E-13
0 1000 2000 3000
Ba-133
t, h
Ra-223
1.E-15
Figure 1: 223 Ra 2+ and 133 Ba 2+ concentrations change with time in the system containing BaSO 4 (s), after 42
days spiked with more 223 Ra 2+
[1] M. S. Angus, J. B. Raynor and M. Robson, Accred. Qual. Assur. 12, 249 (2007).
[2] M. S. Halmlat, H. Kadi, H. Fellag, Appl. Radiat. Isotopes 59, 95 (2003).
[3] C. R. Paige, W. A. Kornicker, O. E. Hileman Jr, W. J. Snodgrass 178 2, 261 (1993).

PA5 SOLID-WATER INTERFACE REACTIONS
PA5-1
PA5-2
PA5-3
PA5-4
PA5-5
PA5-6
PA5-7
PA5-8
PA5-9
PA5-10
PA5-11
PA5-12
PA5-13
PA5-14
60 CO SORPTION ON LAYERED DOUBLE HYDROXIDES FROM AQUEOUS
SOLUTIONS
N. A. Konovalova, S. A. Kulyukhin, E. P. Krasavina, I. A. Rumer (Russia)
APPROACHES TO SURFACE COMPLEXATION MODELING OF Ni(II) ON
CALLOVO-OXFORDIAN CLAY STONE
Z. Chen, Z. Guo, X. Wang, S. Razafindratsima, J.C. Robinet, G. Montavon, C.
Landesman (France, China)
EFFECT OF B-SITE VACANCY ON CESIUM ADSORPTION TO LAYERED
PEROVSKITE KCa 2 Nb 3 O 10
Zhu-Ling Jiang, TsingHai Wang, Chu-Fang Wang (Taiwan)
ENTHALPY MEASUREMENT OF THE PROTONATION OF GAMMA-ALUMINA
BY CALORIMETRIC TITRATION
Y. Saito, A. Kirishima, N. Sato (Japan)
RADIOTRACER EXCHANGE STUDIES ON THE REVERSIBILITY OF
INTERACTION PROCESSES RELATED TO HUMIC-BOUND METAL
TRANSPORT
H. Lippold, J. Lippmann-Pipke (Germany)
REDISTRIBUTION OF FORMS OF Cs137, Sr90 AND Co60 IN PROCESS OF
LONG-TIME INTERACTION WITH BACKFILL MATERIALS
E.E. Ostashkina, O.A. Burlaka, Z.I. Golubeva, G.A. Varlakova (Russia)
REVERSIBILITY IN RADIONUCLIDE/BENTONITE TERNARY SYSTEMS
N. Sherriff, R. Issa, P. Ivanov, T. Griffiths, N. Grieg, G. Bozhikov, S. Dmitriev, K.
Morris, N.D. Bryan (UK, Russia)
SORPTION BEHAVIOR OF HYDROGEN SELENIDE (HSe-) ONTO IRON-
CONTAINING MINERALS
Y. Iida, T. Yamaguchi, T. Tanaka (Japan)
SORPTION OF 99 TC TO CEMENTITIOUS MATERIALS UNDER REDUCING AND
OXIDIZING CONDITIONS
S.L. Estes, D.I. Kaplan, B.A. Powell (USA)
SORPTION OF CESIUM ON PEAT
M. Lusa, S. Virtanen, J. Lempinen, A.T.K. Ikonen, A.-M. Lahdenperä, J. Lehto (Finland)
SORPTION OF THORIUM FROM AQUEOUS SOLUTIONS USING GRAPHENE
OXIDE ANG IRRADIATED GRAPHENE OXIDE
Li Yan, Zhong Zheng, Chunli Wang, Chunli Liu, Wangsuo Wu (China)
SORPTION OF U(VI) ON LAYERED DOUBLE HYDROXIDES OF Mg, Al, AND
Nd FROM COMPLEX CHEMICAL COMPOSITION SOLUTIONS
N. A. Konovalova, S. A. Kulyukhin, E. P. Krasavina (Russia)
SYNTHESIS AND CHARACTERIZATION OF Ba 2 T iSi 2 O 8 FOR SORPTION
STUDIES
A.S. Kar, M. Sahu, M. Keskar, B.S. Tomar (India)
URANYL IONS SORPTION TO TiO 2 AND INTERACTIONS WITH FA SORPTION:
EXPERIMENTS AND MODELING
Y. Ye, X. Wang, N. Guo, Z. Guo, W. Wu (China)

PA5-15
PA5-16
PA5-17
PA5-18
PA5-19
PA5-20
PA5-21
PA5-22
PA5-23
PA5-24
PA5-25
PA5-26
PA5-27
PA5-28
MODELING THE ADSORPTION OF Eu(III) AND Am(III) ON GRANITE USING A
GENERALIZED COMPOSITE APPROACH
Z. Guo, Z. Chen, Q. Jin, W. Wu (China)
THE BIGRAD CONSORTIUM - THE INFLUENCE OF IRON OXIDE
CRYSTALLIZATION ON THE MOBILITY OF NEPTUNIUM
P. Bots, S. Shaw, G.T.W. Law, T. Marshall, J.F.W. Mosselmans, F.R. Livens, M.A.
Denecke, J. Rothe, K. Dardenne, K. Morris (UK, Germany)
MECHANISM OF STRONTIUM SORPTION AND INCORPORATION DURING
THE ALKALINE ALTERATION OF SEDIMENTS BY KOH DOMINATED
CEMENT LEACHATE
S.H. Wallace, S. Shaw, K. Morris, J.S. Small, I.T. Burke (UK)
RETENTION OF SELENIUM ON GRANITE: ANALYSIS OF GRANITIC
SURFACE FOR THE PURPOSE OF DETERMINATION OF SELENIUM SPECIES
K. Videnská, V. Havlová, M. Vašinová Galiová, V. Kanický, P. Sajdl (Czech Republic)
SELENIUM FORMS ON ÄSPÖ DIORITE: ANALYSIS OF ROCK SURFACE USING
LA-ICP MS AND XPS
K. Videnská, V. Havlová, M. Vašinová Galiová, V. Kanický, P. Sajdl (Czech Republic)
A NANO TO MACRO SCALE INVESTIGATION OF MULTI-SITE CAESIUM
SORPTION TO ILLITE AND SEDIMENT
A.J. Fuller, S. Shaw, C.L. Peacock, D. Trivedi. I.T. Burke (UK)
Tc(VII) IMMOBILIZATION ON GRANITOID ROCKS FROM ÄSPÖ (SWEDEN)
Y. Totskiy, F. Huber, T. Schäfer, H. Geckeis (Germany)
INVESTIGATION OF ACTINIDE AND LANTHANIDE SORPTION ON CLAY
MINERALS UNDER SALINE CONDITIONS
A. Schnurr, R. Marsac, T. Rabung, J. Lützenkirchen, H. Geckeis (Germany)
SORPTION OF Np(V) ONTO Na -BENTONITE AND GRANITE: EFFECT OF
EQUILIBRIUM TIME, pH, IONIC STRENGTH AND TEMPERATURE
P. Li, Z. Liu, Z. Guo, W. Wu (China)
COMPARISON OF URANYL ADSORPTION ON IRON(III) OXYHYDROXIDES
K. Niida, T. Saito, S. Tanaka (Japan)
URANYL COORDINATION CHEMISTRY ON Mg-RICH MINERALS:
POLARISATION DEPENDENT EXAFS
A. van Veelen, R. Copping, G. La 3 , A.J. Smith, J.R. Bargar, D.K. Shuh, R.A. Wogelius
EFFECT OF AGING ON THE REVERSIBILITY OF Pu(IV) SORPTION TO
GOETHITE
J.C. Wong, M. Zavarin, J.D.C. Begg, A.B. Kersting, B.A. Powell (USA)
INVESTIGATIONS OF THE SORPTION OF U(VI) ONTO SiO 2 IN THE PRESENCE
OF PHOSPHATE: IN SEARCH OF A TERNARY SURFACE COMPLEX
M.J. Comarmond, H. Foerstendorf, R. Steudtner, E. Chong, K. Heim, K. Müller, K.
Gückel, V. Brendler, T.E. Payne (Australia, Germany)
U(VI) SURFACE DISTRIBUTION ON ÄSPÖ DIORITE UNDER ANOXIC
CONDITIONS
U. Alonso, T. Missana, A. Patelli, D. Ceccato, M. García-Gutiérrez, V. Rigato (Spain,
Italy)

PA5-29
PA5-30
PA5-31
PA5-32
PA5-33
PA5-34
PA5-35
PA5-36
PA5-37
PA5-38
PA5-39
PA5-40
PA5-41
IMMOBILISATION OF TECHNETIUM-99 ON BACKFILL CEMENT: SORPTION
UNDER STATIC AND SATURATED FLOW CONDITIONS
C.L. Corkhill, J.W. Bridge, P. Hillel, L.J. Gardner, M.C. Stennett, R. Tappero, N.C.
Hyatt (UK)
SPECIATION OF PLUTONIUM DURING DIFFUSION IN OPALINUS CLAY
S. Amayri, U. Kaplan, J. Drebert, J. Rosemann, D. Grolimund, T. Reich (Germany,
Switzerland)
INFLUENCE OF SOIL PROPERTIES AND PH CHANGES IN AMERICIUM
SORPTION-DESORPTION ON SOILS
O. Ramírez-Guinart, M. Vidal, A. Rigol (Spain)
THE INFLUENCE OF DIFFERENT MINERAL SURFACE PROPERTIES AND THE
PRESENCE OF NICKEL(II) ON EUROPIUM(III) RETENTION AT VARIOUS
OXIDE MINERALS
S. Virtanen, J. Knuutinen, N. Huittinen, T. Rabung, H. Geckeis, J. Lehto (Finland,
Germany)
STUDY OF THE Cs SORPTION IN KAOLINITE AND APPLICATION TO Cs
SORPTION MODELLING IN MIXED CLAY SYSTEMS
A. Benedicto, T. Missana, M. Garcia-Gutierrez (Spain)
CHARACTERIZING URANIUM AND THORIUM IN SOILS: COMPLEMENTARY
INSIGHTS FROM ISOTOPIC EXCHANGE AND SINGLE EXTRACTIONS
H. Ahmed, S. Young, G. Shaw (UK)
PREDICTIONS OF NpO 2 + IONIC EXCHANGE ON MONTMORILLONITE IN
NATURAL WATERS
A. Benedicto, J. D. Begg, P. Zhao, A. B. Kersting, M. Zavarin (USA, Spain)
EXAFS INVESTIGATION ON Eu(III)-SILICA-HUMIC ACID SORPTION SYSTEM:
EFFECT OF ADDITION ORDER
S. Kumar, S. Kasar, A.S. Kar, B.S. Tomar (India)
STUDIES ON URANIUM(VI) SORPTION ONTO MONTMORILLONITE IN
HIGHLY CONCENTRATED BACK-GROUND ELECTROLYTES
K Fritsch, K Schmeide, G Bernhard (Germany)
STUDY OF EUROPIUM AND NICKEL INTERACTION WITH CALCITE - BATCH
EXPERIMENTS AND SPECTROSCOPIC CHARACTERIZATION
A. Sabau , N. Jordan , C. Lomenech , N. Marmier, V. Brendler ,
A. Barkleit , S. Surblé , N. Toulhoat , Y. Pipon , N. Moncoffre ,
E. Giffaut (France, Germany)
SELECTIVE REMOVAL OF METALS FROM AQUEOUS SOLUTIONS USING
SILICA ATTACHED LIGANDS
J. Holt, S. Christie, N. Evans, S. Edmondson (UK)
EFFECTS OF CEMENT SUPERPLASTICIZERS ON Eu SORPTION ONTO
KIVETTY GRANITE
S. Holgersson (Sweden)
URANIUM (VI) SORPTION ONTO ROCK SAMPLES FROM AREAS OF THE
PROPOSED HLW AND SNF REPOSITORY IN RUSSIA (NIZHNEKANSKY
MASSIVE)
N.V. Kuzmenkova, V.G. Petrov, I.E. Vlasova, V.A. Petrov,
V.V. Poluektov, S.N. Kalmykov (Russia)

PA5-42
PA5-43
PA5-44
PA5-45
PA5-46
PA5-47
PA5-48
PA5-49
SORPTION OF SELENIUM OXYANIONS ONTO HEMATITE
N. Jordan, S. Domaschke, H. Foerstendorf, A.C. Scheinost, S. Weiß,
K. Heim (Germany)
EFFECT OF NATURAL ORGANIC LIGANDS ON PLUTONIUM SORPTION TO
MONTMORILLONITE: OBSERVATIONS ON DIFFERENCES DUE TO ORDER OF
ADDITION
M.A. Boggs, M. Zavarin, B.A. Powell, A.B. Kersting (USA)
SORPTION AND SPECIATION OF MOLYBDENUM ON BOREAL FOREST SOIL
SAMPLES
M. Söderlund, J. Lehto (Finland)
Sr-85 AND Eu-152 SORPTION ON MX-80 BENTONITE COLLOIDS
P. Hölttä, O. Elo, S. Jortikka, S. Niemiaho, M. Lahtinen, J. Lehto (Finland)
Eu AND Cm SORPTION ONTO UNPURIFIED ILLITE: BATCH-TYPE
EXPERIMENTS AND TIME RESOLVED LASER FLUORESCENCE
SPECTROSCOPY (TRLFS)
T. Kupcik, R. Marsac, M. Hedde, T. Rabung, T. Schäfer, M. Marques Fernandes, B.
Baeyens, H. Geckeis (Germany, Switzerland)
INTERACTION OF Ni(II) WITH CLAY MINERALS STUDIED BY
MACROSCOPIC AND MICROSCOPIC APPROACH
Shitong Yang, Guodong Sheng, Xiaoli Tan, Jun Hu, Xiangke Wang (China)
SURFACE MODIFIED MINERALS FOR RADIONUCLIDE SEQUESTRATION
H. Gillings, S.E. Dann, D. Read (UK)
RADIONUCLIDE INTERACTIONS WITH THE Fe 1-x O SURFACE UNDER GDF
CONDITIONS
O. Preedy, A van Veelen, M.P. Ryan, R.A. Wogelius, K. Morris, G.T.W. Law, N.A.
Burton, F. Mosselmans, N.D.M. Evans (UK)

PA5-1
60 Co SORPTION ON LAYERED DOUBLE HYDROXIDES FROM AQUEOUS SOLUTIONS
N. A. Konovalova, S. A. Kulyukhin, E. P. Krasavina, I. A. Rumer
Frumkin' Institute of Physical Chemistry and Electrochemistry, Russian Academy of Science, Leninskii
prospekt, 31, 119071 Moscow, Russia, E-mail: kulyukhin@ipc.rssi.ru
The development of the nuclear power industry resulting in the accumulation of large amounts of radioactive
waste including liquid waste has been urging the development of new express methods of liquid radioactive
waste (LRW) reprocessing that would allow LRW handling without its long-term pre-storage. This requires
developing methods of separation of highly toxic radionuclides including 60 Co for their subsequent
concentration and burial. One of the methods that has been widely used for separating radionuclides in the
ionic state is sorption.
This work studies 60 Со sorption from aqueous solutions of varied compositions on layered double hydroxides
(LDH) of the composition M 3 2+ M 3+ (OH - ) 6 . (A n- ) 1/n . mH 2 O, where M 2+ = Mg, Zn, or Ni; M 3+ = Al, Fe, and Nd;
A n- = CO 3 2- , Cl - , SO 4 2- , EDTA 4- , NO 3 - , OH - , or C 2 O 4 2- , as well as on layered double oxide (LDO) of Mg and
Al.
The research study has found that 60 Со sorption on LDH-Mg-Al-Cl, LDH-Mg-Al-NO 3 , LDH-Zn-Fe-CO 3 ,
LDH-Ni-Fe-CO 3 , and LDH-Mg-Nd-CO 3 from 10 -15 -10 -9 mol/l aqueous solutions of Co(NO 3 ) 2 and on LDO-
Mg-Al from 10 -3 -10 -5 mol/l aqueous solutions of Co(NO 3 ) 2 is poor. At a liquid-solid phase contact time of 24
h and V/m = 50 ml/g, the distribution coefficients (К d ) for 60 Со are ~10 2 -10 3 ml/g.
On the other hand, 60 Со displays good sorption on LDH-Mg-Al-CO 3 , СLDH-Mg-Al-SO 4 , LDH-Mg-Al-
EDTA, LDH-Mg-Al-C 2 O 4 , LDH-Zn-Al-CO 3 , LDH-Ni-Al-CO 3 , LDH-Mg-Fe-CO 3 , and LDH-Mg-Nd-CO 3
from aqueous nitrate solutions. The K d values for 60 Со for the above-mentioned compounds are as high as
10 4 -10 5 ml/g at a liquid-solid phase contact time of 24 h and V/m = 50 ml/g in the Co(NO 3 ) 2 concentration
range 10 -15 -10 -9 mol/l.
A study of 60 Со desorption from LDH-Mg( 60 Со)–Nd-CO 3 in 0.05-0.2 mol/l solutions of Na 2 CO 3 , NaNO 3 ,
(NH 4 ) 2 C 2 O 4 , and Na 2 Н 2 EDTA and distilled water has found that the highest desorption ability is
characteristic of Na 2 Н 2 EDTA. After 15-min contact of LDH-Mg( 60 Со)-Nd-CO 3 with 0.1 and 0.05 mol/l
solutions of Na 2 Н 2 EDTA, the desorption efficiency for 60 Со is ∼100% and ∼99%, respectively.
The research findings allow us to suggest that LDH-Mg-Al-CO 3 , LDH-Mg-Al-SO 4 , LDH-Mg-Al-EDTA,
LDH-Mg-Al-C 2 O 4 , LDH-Zn-Al-CO 3 , LDH-Ni-Al-CO 3 , LDH-Mg-Fe-CO 3 , and LDH-Mg-Nd-CO 3 may be
used for separating 60 Со radionuclide, as well as U(VI) and also Sr and Y radionuclides (as analogs of
trivalent f-elements) from aqueous solutions in various technological processes.
PA5-2
APPROACHES TO SURFACE COMPLEXATION MODELING OF Ni(II) ON CALLOVO-
OXFORDIAN CLAY STONE
Z. Chen (1,2) , Z. Guo (2) , X. Wang (3) , S. Razafindratsima (1) , J.C. Robinet (4) , G. Montavon (1) , C. Landesman (1)
1) SUBATECH, 4 rue A. Kastler, 44307 Nantes Cedex, France
2) School of Nuclear Science & Technology, Lanzhou University, 730000, China
3) Institute of Plasma Physics, P O Box 1126, Hefei, 230031 PR China
4) ANDRA, Research and Development Division, 1/7 rue Jean Monnet - 92298 Châtenay-Malabry cedex,
France
Sedimentary clay-rich formations are under investigation in Europe (Callovo-Oxfordian formation (COX) in
France, the Boom clay in Belgium and the Opalinus clay in Switzerland) for long term nuclear waste

epositories. Thus, it is a main concern and pressing need to predict the uptake of radionuclide on host
formations under “in situ” conditions. This requires the measurement of sorption data in conditions relevant
to the real system and the development of sorption models explaining notably how the retention parameters
are affected by the large scale natural heterogeneities in the formation (mineral, pore water composition) as
well as perturbation close to the vicinity of the waste (e.g. temperature, alkaline plume). These models may
be used in a second step to derive reliable operational for the safety case analysis.
The ideal case would be to consider a bottom-up approach including all processes occurring in solution or at
the solid/liquid interface with parameters proposed in the thermodynamic databases for model
phases/compounds in order to predict macroscopic sorption data. This “bottom-up” approach may be realist
when two simplifications are performed, i.e. (i) to consider that the sorption behavior is only governed by
clay fraction and (ii) to consider mixed-layer illite-smectite minerals (I/S) as a mixture of illite and
montmorillonite [1]. These simplifications were successfully applied to the sorption of Cs on Opalinus [2]
and Boom [3] clay formations, where the sorption is governed by cation-exchange. However, the
applicability of the bottom-up approach for metal ions sensitive to both cation exchange and surface
complexation reactions is still being discussed [4, 5].
The objective of this work was to provide a range of Kd values describing Ni(II) sorption on Callovo-
Oxfordian clay rock (COX) and to assess whether a simplified “bottom-up” approach may explain the
retention considering two levels (i) from clay surfaces to clay fraction and (ii) from clay fraction to COX
samples. To this end, Ni(II) sorption was investigated by batch, XPS, and EXAFS techniques on several
COX samples having various clay contents and their enriched clay fraction.
The results show that ‘bottom-up’ approach with published models fails to explain the sorption properties of
Ni(II) on the enriched clay fraction of the Callovo-Oxfordian claystone when the retention is governed by
surface complexation on the reactive edge sites (pH>5). In order to solve this issue, further works are needed
to define a generalized model for clay materials of interest (illite and smectite) and data are needed for I/S. A
better agreement between model and sorption data was however obtained at pH < 5 when the sorption is
dominated by cation exchange processes. The Generalized Composite Model (GCM) was used in a second
approach to explain the Ni(II) sorption data obtained for enriched clay fraction considering an interaction
between Ni 2+ and NiOH + with one type of sites. Integrating this model into an additive approach taking into
account clay mineralogy allows to predict the behavior of Ni(II) on a COX sample presenting an important
clay fraction (53%). The crucial role of this clay fraction in the retention was confirmed at the molecular
level by XPS and EXAFS. However, the model overestimates the retention when the clay content decreases,
i.e. when the content in carbonate phases increases. Carbonate phase appears to be the reactive one. For a
safety case analysis, one can finally give a Kd range in the formation from 60 to 300 L/Kg for trace
concentrations of Ni(II) (~5.10 -8 -10 -5 M) depending on the clay content.
[1] Tournassat, C.G., et al. (2009), "Cation exchange selectivity coefficient values on smectite and mixedlayer
illite/smectite minerals". Soil Science Society of America Journal. 73(3): 928-942.
[2] Van Loon, L.R., B. Baeyens, and M.H. Bradbury (2009), "The sorption behaviour of caesium on
Opalinus clay: A comparison between intact and crushed material". Applied Geochemistry. 24(5): 999-1004.
[3] Maes, N., et al. (2008), "Retention of Cs in boom clay: Comparison of data from batch sorption tests and
diffusion experiments on intact clay cores". Physics and Chemistry of the Earth 33: S149-S155.
[4] Bradbury, M. and B. Baeyens (2011), "Predictive sorption modelling of Ni(II), Co(II), Eu(IIII), Th(IV)
and U(VI) on mx-80 bentonite and Opalinus clay: A "bottom-up" approach". Applied Clay Science. 52: 27-
33.
[5] Davis, J.A., et al. (2004), "Approaches to surface complexation modeling of uranium (VI) adsorption on
aquifer sediments". Geochimica et Cosmochimica Acta. 68(18): 3621-3641.

PA5-3
EFFECT OF B-SITE VACANCIES ON CESIUM ADSORPTION TO LAYERED PEROVSKITE
KCa 2 Nb 3 O 10
Zhu-Ling Jiang1), TsingHai Wang1), Chu-Fang Wang1)*
1)Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Hsinchu, 300,
Taiwan
Layered minerals such as the family of clay, mica and zeolite are increasingly applied for waste management
purposes because of their low cost. In depth knowledge on the interaction of waste components with these
layered minerals is required for performance assessment of a waste repository plan. Weathering of minerals
in natural environment may significantly change surface properties and thus influence the capability of
pollutant uptake. Such potential variations must be considered to reduce uncertainties in transport
simulations. In this study, we intentionally manipulated the surface basicity of layered perovskite KCa 2 Nb 3 -
xO 10 by introducing B-site vacancies where x = 0, 0.1, 0.2, 0.3, and 0.4. The influence of B-site vacancies on
the surface chemistry was monitored by analyzing the cation exchange capacity (CEC) and Cs uptake.
Results from XRD showed that modified KCa 2 Nb 3 -xO 10 specimen kept their layered perovskite structure and
no significant structural distortion was observed due to the introduction of B-site vacancies. Triplicate CEC
analyses using the ammonium acetate method showed that the CEC decreased inversely proportional to the
amount of introduced B-site vacancies (111.3 for x=0, 102.1 for X=0.1, 99.9 for x=0.2, 93.3 for x=0.3, 91.2
for x=0.4, mEq/100g respectively). These values were equivalent to 6.22 for x=0, 5.48 for X=0.1, 5.40 for
x=0.2, 5.04 for x=0.3, 4.75 for x=0.4, mmole Na per mole of KCa 2 Nb 3 -xO 10 entities. Further Cs adsorption
experiments conducted at an ionic strength of 0.01 N of NaCl clearly revealed pH independent sorption
behavior. This suggested that Cs uptake by KCa 2 Nb 3 -xO 10 occurs predominantly by cation exchange.
Interestingly, unlike the trend observed from CEC results, the Cs uptake of KCa 2 Nb3-xO10 increased along
with the amount of B-site vacancies. At an initial concentration of 100 ppm, the average Cs uptake is 11.5 %
for x=0, 14.3 % for x=0.1, 18.2 % for x=0.2, 22.6 % for x=0.3 and 25.1 % for x=0.4, respectively. Based on
those results, it seems that the introduced B-site vacancies increase the basicity of KCa 2 Nb 3 -xO 10 . To
compensate the negative surface charge induced by B-site vacancies, KCa 2 Nb 3 -xO 10 apparently tends to
preferentially sorb K+ over Na+. As a result, CEC decreases as the number of B-site vacancies increases. On
the other hand, since cesium ions have a higher charge/volume ratio than sodium, KCa 2 Nb 3 -xO 10 specimen
with more B-site vacancies are more favorable to Cs uptake than those with less B-site vacancies. Further
qualitative and quantitative analyses of surface basicity by studying pyrrole adsorption as an IR
spectroscopic molecular probe are conducted.
* corresponding author: Chu-Fang Wang, Professor, cfwang@mx.nthu.edu.tw
[1] R. E. Schaak, T.E. Mallouk, J. Am. Chem. Soc. 122, 2798 (2000).
[2] K. Maeda, et al., Chem. Mater. 20, 6770 (2008).
[3] J.A. Schottenfeld, et al., Chem. Mater. 20, 213 (2008)
PA5-4
ENTHALPY MEASUREMENT OF THE PROTONATION OF γ-ALUMINA BY
CALORIMETRIC TITRATION
Y. Saito, A. Kirishima and N. Sato
Institute of Multidisciplinary Research for Advanced Materials, Tohoku University,
2-1-1, Katahira, Aoba-ku, 980-8577 Sendai, Japan
In the geological disposal system, high-level radioactive waste is isolated from the living area by the multiple
barrier system composed of the engineered barrier and natural barrier by surrounding minerals. There is a
possibility that long life nuclides such as minor actinides would be eluted in groundwater as metal ions, and
reach the living area by migrating underground interacting with the dissolved materials and surrounding

minerals. So, in order to evaluate the long-term safety of geological disposal system, understanding of the
reaction mechanism of sorption between metal ions and minerals is essential. In this study, we use a surface
complexation model for the understanding of the reaction mechanism between the metal ions and minerals.
This model describes mineral surface reaction separately into the complex formation of surface functional
groups and the electrostatic bound by the confluence of the surface charges. Therefore, in this model, Gibbs
free energy (ΔG total ) is decomposed into complex term (ΔG complex ) and the electrostatic term (ΔG coulomb ) as
follows[1],
∆ G = ∆G
+ ∆G
. (1)
total
complex
coulomb
when we substitute species concentrations for ΔG, we can rewrite (1) as
log
[ ≡ XOM][ H]
[ ≡ XOH][ M]
= log K
(M)
int .
+ loge
( −Fψ
/ RT )
where XO is surface functional group of mineral, H is proton, M is metal ion, K is equilibrium constant, ψ is
surface potential, R is gas constant and T is the temperature. We can observe only ΔG total from the normal
sorption experiments. So it is difficult to evaluate that the decomposition of the complex term and the
electrostatic term is being carried out correctly in the sense of thermodynamics. Focusing on the reaction
heat of each term, it is expected that there is a big difference between complex term that derives new
chemical bond formation and the electrostatic term that is originated only from the electrostatic attraction.
Therefore, it is considered that we can evaluate whether two terms can be decomposed properly from the
measurement of reaction heat of metal ion sorption to mineral by using the calorimetric titration technique.
As a first step of this attempt, the protonation enthalpy of γ-alumina was measured in this study.
First, the equilibrium constants of the protonation and de-protonation of the γ-alumina were determined in
0.1 or 1.0 M NaClO 4 solution with 2.5, 7.5 and 22.5 g/L of γ-alumina suspension concentration by
potentiometric titration at 25.0 ± 0.2 o C. From the values of these equilibrium constants, the effect of ionic
strength and suspension concentration were discussed. Then, the enthalpy of the protonation and deprotonation
of γ-alumina were determined in 1.0 M NaClO 4 solution with 22.5 g/L of γ-alumina suspension
concentration by iso-thermal calorimetric titration at 25.000000 ± 0.000004 o C. In the calorimetric titration,
ultra-sensitive heat flow measurement system TAM-III (TA instruments, USA) was used for the protonation
and de-protonation reaction heat measurement with a module of nano-calorimeter, which is specially
designed for nano-watt (nW) scale heat flow measurement. The ΔH of each reaction was determined by a
fitting of the equation (3) to the obtained calorimetric titration results.
, (2)
− − + +
− −
OH OH XOH2 XOH2
XO XO
( ν ν
−1 ) ( ν ν
−1 ) ( ν ν
−1
)
dQ =∆H − +∆H − +∆H
− (3)
i −
OH i i +
XOH i i −
2
XO i i
In the equation (3), dQ i denotes observed reaction heat at i-th titration step, and ν i X is the mole amount of
species ‘X’ existing in the reaction vessel after the i-th titration step. ν i X was calculated using solution
volume and the stability constants previously determined by the potentiometric titration. The reported
dissociation enthalpy of H 2 O (=ΔH OH- ) in NIST database [2] was employed for the evaluation of protonation
enthalpy (=ΔH XOH2+ ) and de-protonation enthalpy (=ΔH XO- ). The determined enthalpies are as follows.
protonation (XOH + H + ↔ XOH 2 + ), (4)
H XOH2+ = -20.9 ± 0.7 kJ/mol at I= 1.0 with 22.5 g/L of γ-alumina
de-protonation (XOH ↔ XO - + H + ), (5)
H XO- = 28.0 ± 0.7 kJ/mol at I= 1.0 with 22.5 g/L of γ-alumina
From the obtained enthalpy and stability constant values, TΔS was also calculated. The result shows that ΔH
is independent on the electro static potential of the solid surface while TΔS is proportional to the electrostatic
potential. This implies that the same bonds are formed in wide pH range between proton and the surface
function groups of γ-alumina, while the change of the reaction entropy compensates the ΔG coulomb which
depends on the charge state of solid surface.

[1] D. Dzombak, Surface Complexation Modeling, WILEY-INTERSCIENCE, New York, (1990)
[2] Martell, A. E., Smith, R. M., Motekaitis, R. J.: NIST Critically Selected Stability Constants of Metal Complexes,
ver. 8.0, Texas A&M University, Texas, (2004)
PA5-5
RADIOTRACER EXCHANGE STUDIES ON THE REVERSIBILITY OF INTERACTION
PROCESSES RELATED TO HUMIC-BOUND METAL TRANSPORT
H. Lippold, J. Lippmann-Pipke
Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology (Research Site Leipzig),
04318 Leipzig, Germany
The mobility of actinides or other contaminants in the subsurface hydrosphere is considerably influenced by
their interaction with natural colloids. Besides inorganic particles, aquatic humic substances are ubiquitous in
natural waters, and their complexing ability can dominate the speciation of toxic or radioactive metals [1, 2].
Since humic carriers are subject to a solid-liquid distribution depending on geochemical parameters, an
adequate assessment of migration processes needs thorough consideration of all interactions within the
ternary system metal – humic substance – solid surface, including adsorption / retardation of humic colloids.
Reactive transport models have been developed, taking all these processes into account [3-5]. As a
prerequisite, reversibility is commonly assumed.
There is, however, a lack of clarity as to whether full reversibility is actually given for the whole ternary
system, especially concerning interactions of humic matter with mineral surfaces and metals. For adsorption
of humic substances, strong hysteresis has been observed (hardly any desorption upon dilution) [6-8], and
recoveries in column experiments have been found to be far from complete [9, 10]. Regarding metal-humic
interaction, it has been reported that complexation of higher-valent metals is accompanied by slow processes
leading to an increase in complex inertness, i.e., a growing resistance towards dissociation in the presence of
competing ligands or metals [11-13].
In view of these uncertainties, the aim of the present study was to elucidate the reversible / irreversible
character of processes controlling humic-bound transport. For this purpose, the principle of tracer exchange
was employed to gain insight into the dynamics of equilibria within the ternary system. In case of
reversibility, a dynamic equilibrium exists, i.e., a permanent run of adsorption and desorption (or complex
formation and dissociation) at equal rates. Such an exchange can be detected by introducing a radiotracer
into pre-equilibrated systems where all binding sites are occupied.
The chosen model system for these experiments consisted of terbium(III) (as an analogue of trivalent
actinides), humic acid (Aldrich) or fulvic acid (isolated from bog water), and kaolinite (KGa-1b standard
material). 160 Tb as a radioisotope was produced by neutron activation of 159 Tb at the TRIGA Mark II reactor
of the University of Mainz. Humic and fulvic acids were radiolabelled by an azo coupling reaction with
[ 14 C]aniline [14].
To investigate the dynamics of adsorption equilibria, kaolinite suspensions were first contacted with nonradioactive
Tb(III) or humic / fulvic acid at a range of concentrations, covering an adsorption isotherm
including the plateau region. Subsequent to a pre-equilibration phase, a small amount of the radiotracer
( 160 Tb(III) or 14 C-labelled humic / fulvic acid, respectively) was added. After admitting different time periods
for equilibration, tracer exchange was evaluated from the concentration decrease in the supernatant. In
additional batch experiments, desorption of humic and fulvic acid upon dilution was examined within a
comparable time frame.
Reversibility of Tb(III)-humate complexation was investigated in a similar way. Since humic acid
precipitates completely on loading with Tb(III), adsorption systems were generated. Here, times for tracer
exchange were kept constant, and pre-equilibration times were varied instead.

As expected, adsorption of Tb(III) onto kaolinite was found to be a fast dynamic equilibrium process.
Identical adsorption isotherms were obtained regardless of whether the radiotracer was introduced
instantaneously together with the non-radioactive metal or subsequently after 2 days of pre-equilibration. For
humic and fulvic acid, such dynamic exchange was proven to exist as well, but at considerably lower rates.
In case of subsequent tracer addition, the plateau sections of the isotherms were significantly lowered
(notably, not to zero). When equilibration times were increased (from 6 hours to 4 weeks), the plateaus
approached the respective isotherm for instantaneous tracer addition. Finally, both isotherms coincided, i.e.,
the dynamic equilibrium was quantitatively represented by the tracer.
In desorption experiments with humic or fulvic acid, initiated by diluting the supernatant after an adsorption
phase, no release was observed in the course of 4 weeks, which seems to be contradictory to the above
results. One may conclude that the absence of desorption upon dilution is not necessarily indicative of a
static equilibrium without any exchange. Thus, models for humic-bound transport are certainly applicable
under appropriate conditions. Nonetheless, when comparing the kinetics of exchange to the kinetics of
adsorption for humic / fulvic acid, rate constants differ by one order of magnitude [15]. This discrepancy
must be taken into account when conditions of a steady local equilibrium are assigned to a maximum flow
velocity.
For complexation of Tb(III) with humic acid, we did not find any indications of stabilisation processes
affecting the reversibility. Increasing complex inertness has been observed for a variety of metals such as
Al(III), Eu(III), Am(III), Th(IV) or U(VI), on time scales ranging from 2 days up to several months (see [13]
for a review). In our tracer exchange experiments, Tb(III)-humate complexes were pre-equilibrated for 1 to
90 days before 160 Tb(III) was added and a subsequent equilibration period of 1 day was admitted. In all
cases, the binding isotherms were indistinguishable from the binding isotherm obtained for instantaneous
tracer addition, i.e., an unresisted dynamic exchange was indicated by the tracer, even after long contact
times prior to its introduction. Obviously, increasing complex inertness is not a general phenomenon
occurring across all higher-valent metals.
[1] J. P. L. Dearlove et al., Radiochim. Acta 52/53, 83 (1991).
[2] J. I. Kim et al., Radiochim. Acta 58/59, 147 (1992).
[3] L. Lührmann et al., Water Resour. Res. 34, 421 (1998).
[4] N. D. Bryan et al., J. Environ. Monit. 7, 196 (2005).
[5] M. Kim and S. B. Kim, Environ. Technol. 28, 205 (2007).
[6] E. M. Murphy et al., Sci. Total Environ. 117/118, 413 (1992).
[7] B. Gu et al., Environ. Sci. Technol. 28, 38 (1994).
[8] M. J. Avena and L. K. Koopal, Environ. Sci. Technol. 32, 2572 (1998).
[9] F. M. Dunnivant et al., Soil Sci. Soc. Am. J. 56, 437 (1992).
[10] H. Weigand and K. U. Totsche, Soil Sci. Soc. Am. J. 62, 1268 (1998).
[11] R. Artinger et al., J. Contam. Hydrol. 35, 261 (1998).
[12] H. Geckeis et al., Environ. Sci. Technol. 36, 2946 (2002).
[13] H. Lippold et al., Appl. Geochem. 27, 250 (2012).
[14] A. Mansel and H. Kupsch, Appl. Radiat. Isot. 65, 793 (2007).
[15] H. Lippold et al., in preparation (2013).
PA5-6
REDISTRIBUTION OF FORMS OF CAESIUM-137, STRONTIUM-90 AND COBALT-60 IN
PROCESS OF LONG-TIME INTERACTION WITH BACKFILL MATERIALS
E.E. Ostashkina 1) , O.A. Burlaka 1) , Z.I. Golubeva 1) , G.A. Varlakova 2)
1) Scientific and Industrial Association "Radon", 119121 Moscow, Russia, the 7-th Rostovsky Lane, 2/14
2) Joint Stock Company "A.A. Bochvar High-technological Research Institute of Inorganic Materials ",
123098, Moscow, Russia, the Rogov Str, 5a.
Nature barrier materials are very important structure elements of backfill for near surface repositories for
radioactive wastes. They retain radionuclide migration from repositories and their expansion into the
biosphere. These backfill materials must be cheap and easy for getting, these are such as sands, excavated
rocks and their mixtures.

The radionuclide migration depends on sorption properties of backfill materials. Properties of backfill
materials determine the forms of stored radionuclide, especially in long-time contact between them.
Physicochemical forms of radionuclides penetrated into nature barrier materials change in time [1].
This paper presents evaluation of Cs 137 , Sr 90 and Co 60 forms and their redistribution in barrier materials in
time.
The experiments were carried on with covering silt and sand. The covering silt contained quartz, feldspar and
clay minerals (kaolinite group, illite, vermiculite, montmorillonite). The sand of glaciolacustrine origin
contained quartz and small amounts of clay minerals, feldspar, ferrum hydroxides and manganese
compounds [2-4]. In the experiments, the samples of above mentioned materials were in contact with
radionuclides solutions during 2 weeks, 1, 2, 4 and 6 months accordingly. After these they were treated by
solutions of different compositions [1, 5].
The experiments showed that Cs 137 , Sr 90 and Co 60 were in different forms and proportions in tested materials.
Redistribution of these forms can be observed in time. The main feature for all radionuclide was the
decreasing of ratio of mobile forms during their contact with materials increase. Simultaneously increased
tightly bound forms of radionuclides. The time of transformations processes in the materials differed, which
could be connected with mechanism of radionuclide absorption by the solid phase.
The time of transformation of radionuclide forms increases in such way in line Sr < Co < Cs and in line
sand< covering silt. A great part of cesium ran to tightly bound form in a first weeks of contact. The ratio of
cesium in tightly bound form achieved its maximum (95%) already after two weeks of contact. Strontium
and cobalt compared to cesium were more mobile and their transformation into tightly bound form occurred
gradually during the experiment. The maximal abundance of these radionuclides in tightly bound form was
68–71 % and 74–85 % respectively.
The results of experiments allowed to estimate hardness of radionuclide fixation by barrier materials for near
surface repositories and to conclude that radionuclide migration in backfill increases in operational period.
[1] F.I. Pavlotskaya (1979). Moscow Atomizdat: 215 p.
[2] S.A. Dmitriev, A.S. Barinov, G.A. Varlakova et al. (2005). Migration’05 Avignon, France: P. 201.
[3] A.S. Barinov, G.A. Varlakova, I.V. Startceva, Z.I. Golubeva, O.A. Burlaka. (2006) Radiochemistry-2006, Dubna:
pp. 289-290.
[4] G.A. Varlakova, Z.I. Golubeva, S.V. Roschagina et.al. (2009) Radiochemistry-2009, Moscow: pp. 253-254.
[5] F.I. Pavlotskaya, A.P. Novicov, T.A Goryachenkova et.al. (1998). Radiochemistry, 40, 5: pp. 462-467.
PA5-7
REVERSIBILITY IN RADIONUCLIDE/BENTONITE TERNARY SYSTEMS
N. Sherriff, (1) R. Issa, (1) P. Ivanov, (1) T. Griffiths, (1) N. Greig, (1) G. Bozhikov, (2) S. Dmitriev, (2)
K. Morris (1) and N.D. Bryan (1)
(1) Centre for Radiochemistry Research and Centre for Radwaste and Decommissioning, University of
Manchester, Manchester, M13 9PL, United Kingdom
(2)
Flerov Laboratory of Nuclear Reactions, JINR, 141980 Dubna, Russia
In order for colloids to be of significance in the transport of radionuclides, certain conditions must be met
[1]: they must be present, mobile and stable and radionuclide binding must be possible and ‘irreversible’. If
any of these conditions is not met, then the effect of colloids on radionuclide migration will probably be
insignificant. Although radionuclide uptake onto bentonite has been studied extensively, the critical
dissociation kinetics have been studied barely at all [2]. Previous experiments with actinides and technetium
with bentonite colloids in the presence of fracture filling material showed two distinct types of behaviour:
U(VI), Np(V) and Tc(VII) did not associate with the bentonite colloids very strongly, but Th(IV), Pu(IV) and
Am(III) did [3,4]. Huber et al have provided dissociation rate constants for these experiments: for Am(III),
the values were in the range 1 - 2.5 x10 -6 s -1 , whilst for Pu(IV), the range was 3.9 x10 -7 - 2.4 x10 -6 s -1 [4].

Wold has also estimated representative first order dissociation rate constants for: Pu(IV) (1.2x10 -6 s -1 );
Am(III) (5.6x10 -7 s -1 ); Cm(III) (1.7x10 -6 s -1 ) [2].
This study has studied the reversibility of radionuclide binding to bulk bentonite in ternary systems. The
effect of humic acid (HA) on U(VI) sorption on bentonite has been studied in batch experiments at room
temperature at a 237 U(VI) concentration of 10 -10 M and HA concentration of 100 mg∙L -1 . The distribution of
U(VI) between the liquid and solid phase was studied as a function of pH and ionic strength, both in the
absence and presence of HA. The uranyl sorption on bentonite is dependent on pH and the presence of
humics. In the absence of HA an enhancement in the uptake with increasing pH was observed, and a sharp
sorption edge was found between pH 3.2 and 4.2. The presence of HA slightly increases uranium(VI)
sorption at low pH and curtails it at moderate pH, compared to the behaviour in the absence of HA. In the
basic pH range for both the presence and absence of HA, the sorption of uranium is significantly reduced,
which could be attributed to the formation of soluble uranyl carbonate complexes. Most importantly, it was
found that the effect of the addition order was negligible, and there was no evidence for slow dissociation or
(pseudo-)irreversible binding to the bentonite for uranyl (Figure 1).
Batch experiments have also been performed for ternary systems of 152 Eu(III), bentonite and humic acid and
also 152 Eu(III), bentonite and EDTA. The Eu(III) concentration was 7.9 10 -10 M and the pH was in the range
of 6.0 – 7.0. In the absence of a competing ligand, there was evidence for some slow uptake in a two stage
process, with initial rapid sorption of Eu(III), followed by slower uptake of a much smaller fraction. In the
humic acid ternary system, it was found that on a timescale of several months, there were important kinetic
effects in the interactions between: bentonite and humic acid; bentonite and Eu(III); humic acid and Eu(III).
In a separate series of experiments, the reversibility of Eu(III) binding was tested by allowing Eu(III) to sorb
to bentonite (50 g/l) for times in the range of 1 – 65 days. EDTA was added to the pre-equilibrated Eu
bentonite mixtures at a concentration (0.01 M) that was sufficient to suppress sorption in a system where
EDTA is present prior to the contact of Eu(III) with bentonite. It was found that some fraction of the sorbed
Eu was released by the bentonite instantaneously, but that a significant fraction remained bound. With time,
the amount of Eu(III) remaining bound to the bentonite reduced, with a dissociation rate constant of
approximately 3.8 x 10 -7 s -1 with values in the range of 8.2 x 10 -8 - 8.6 x 10 -7 s -1 . However, after an EDTA
contact time of approximately 100 days, the amount of Eu(III) remaining bound to the bentonite was within
error of that when EDTA was present prior to contact (4.5% + 0. 6). It was found that the amount of slowly
dissociating Eu increased with increasing Eu(III)/bentonite pre-equilibration time.
Although slow dissociation from bentonite has been observed in the case of Eu(III), no ‘irreversible’ uptake
of either Eu(III) or uranyl(VI) by bentonite has been observed. Interestingly, the rate constants determined
here are in the range observed by Huber et al [4] and predicted by Wold [2].
Figure1. Effect of the addition order and pre-equilibration time on uranyl sorption onto 1 g L bentonite at pH
5.0 and I = 0.01; 100 ppm HA added after: 0 days; 1 day; 7 days after uranyl was put in contact with the
bentonite

[1] W.M. Miller, W.R. Alexander, N.A. Chapman, I.G. McKinley, J.A.T. Smellie (2000) Geological Disposal of
Radioactive Wastes and Natural Analogues. Waste Management Series, Vol. 2, Pergamon, Amsterdam.
[2] S. Wold (2010) Sorption of prioritized elements on montmorillonite colloids and their potential to transport
radionuclides, SKB Technical Report, TR-10-20.
[3] M. Bouby, A. Filby, H. Geckeis, F. Geyer, R. Götz, W. Hauser, F. Huber, S. Keesmann, B. Kienzler, P. Kunze, M.
Küntzel, J. Lützenkirchen, U. Noseck, P. Panak, M. Plaschke, A. Pudewills, T. Schäfer, H. Seher, C. Walther (2010)
Sorption reversibility studies: Comparison of fracture filling material (FFM) from Grimsel (Switzerland) and Äspö
(Sweden) in Colloid/Nanoparticle formation and mobility in the context of deep geological nuclear waste disposal
(Project KOLLORADO-1; Final report), Thorsten Schäfer & Ulrich Noseck (Eds.), FZKA Wissenschaftliche Berichte,
FZKA 7515, pg 25 - 36.
[4] F. Huber, P. Kunze, H. Geckeis, T. Schafer, Sorption reversibility in the ternary system radionuclide- bentonite
colloids/ nanoparticles- granite fracture material. Applied Geochemistry; volume 26; issue 12; page 2226- 2237; 2011
PA5-8
SORPTION BEHAVIOR OF HYDROGEN SELENIDE (HSE – ) ONTO IRON–CONTAINING
MINERALS
Y. Iida, T. Yamaguchi, T. Tanaka
Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki, 319-1195, Japan
Retardation of radionuclide migration by sorption onto a host rock is one of the main geologic factors that
influence performance of a radioactive waste disposal system. Performance assessment calculations for
hypothetical high-level radioactive waste repositories show that 79 Se, a long-lived fission product, is one of
the radionuclides that dominate the long-term radiological hazard. Selenium is likely to be stable as Se(–II)
under anoxic and reducing conditions in deep subsurface environments, and hydrogen selenide, HSe – , is
considered to be dominant in groundwater. The dominant sorbent minerals for HSe – have been determined to
be biotite and iron oxide such as goethite in granitic rock [1].
Sorption behaviors of HSe – onto iron–containing minerals, magnetite, goethite, ferrous oxide and biotite,
were investigated by batch sorption experiments under reducing conditions in a controlled atmosphere glove
box under Ar (O 2 < 1 ppm). A stock solution of HSe – was prepared by reducing Na 2 SeO 3 solution including
radioactive 75 Se by hydrazine monohydrate. The stock solution was spiked into 0.01, 0.1 and 1 mol dm -3
NaCl solutions. The pH of the solution was adjusted to 7 to 13 with HCl or NaOH solution. One or 0.1 g of
granulated mineral sample was immersed in a 10 cm 3 volume of the solution in a polypropylene test tube.
After 2 weeks, the suspension was filtered through a 10,000 NMWL ultrafilter and the radioactivity of the
solution was measured by γ- spectrometry.
The sorption ratio, R s (%), was determined by the following equation
R
c
– c
c
ini eq
s = ×
ini
100
where, c ini is the initial concentration and c eq is the equilibrated concentration of Se. The R s values for
goethite are shown in figure 1 as representative examples. They show negative dependence on pH and
independence of the NaCl concentration. The sorption mechanism of HSe – onto goethite has been estimated
to be surface complexation with ferrol sites (≡FeOH) [1]. The negative pH dependence of the R s supports the
surface complexation of HSe – with hydroxyl groups. In addition, HSe – hardly competed against Cl – ion
which sorbed as outer sphere surface complexation [2], indicating that the sorption behavior of HSe – is inner
sphere surface complexation. The inner sphere surface complex of hydrogen sulfide ion (HS − ), which is
chemically analogous to HSe − , with ferrol sites was reported as ≡FeS − [3]. The inner sphere surface
complexation of HSe − can be described as
≡FeOH + HSe − = ≡FeSe − + H 2 O.
The triple layer model [4] was used in the analysis of the sorption behavior of HSe – with Visual Minteq
computer program. The intrinsic equilibrium constant

K
−
int
HSe =
−
b
[ ≡ FeSe ] ⎛ − Fψ0
⎞
exp⎜
⎟
[ ≡ FeOH][HSe ] ⎝ RT ⎠
was determined by the fitting of model calculations to the experimental results. In this equation, F is the
Faraday constant, ψ 0 is electric potential at 0 surface, R is gas constant and T is temperature. The obtained
value, log K int HSe = 5.5, was close to the value of HS −
≡FeOH + HS − = ≡FeS − + H 2 O log K int HS = 5.3
derived from the equilibrium constants of the reactions
and
≡FeOH + H 2 S(g) = ≡FeS − + H + + H 2 O log K int H2S = −2.7 [3]
H 2 S(g) = HS − + H + log K = −8.0 [3].
100
80
R s (%)
60
40
20
0
0.01
0.1
1
0.01
0.1
1
7 8 9 10 11 12 13
pH
Figure 1. The sorption ratio (R s ) of HSe – onto goethite. The circles, squares and triangles represent the data
obtained at 0.01 mol dm -3 , 0.1 mol dm -3 and 1 mol dm -3 , respectively. The curves represent the results of the
triple layer model calculations.
[1] Y. Iida et al., J. Nucl. Sci. Technol., 48[2], 279–291 (2011).
[2] S. A. Carroll et al., Geochem. Trans., 9, 2 (2008).
[3] M. D. S. Afonso and W. Stumm, Langmuir, 8[6], 1671–1675 (1992).
[4] K. F. Hayes et al., J. Colloid Interf. Sci., 125[2], 717–726 (1988).
* Part of this study was funded by the Secretariat of Nuclear Regulation Authority, Nuclear Regulation
Authority, Japan.

PA5-9
SORPTION OF 99 Tc TO CEMENTITIOUS MATERIALS UNDER REDUCING AND OXIDIZING
CONDITIONS
Shanna L. Estes 1 , Daniel I. Kaplan 2 , Brian A. Powell 1
1 Environmental Engineering and Earth Sciences, Clemson University, 342 Computer Court, Anderson, South
Carolina 29625, United States
2 Savannah River National Laboratory, Aiken, South Carolina 29808, United States
Cementitious waste forms offer a possible alternative to vitrification for the long-term
immobilization of low-level radioactive waste. The fission product technetium ( 99 Tc) represents one of the
major risk drivers present in these wastes due to its formation of the highly mobile pertechnetate (TcO 4 - ) ion
over the pH range relevant for environmental systems. However, few studies have investigated the extent and
mechanism of 99 Tc retention by cementitious materials. Therefore, we examined 99 Tc sorption under
reducing conditions to cementitious materials simulating those in use at the Savannah River Site Saltstone
Disposal Facility. After 319 days of equilibration, we transferred samples to oxidizing conditions and
monitored the desorption of 99 Tc.
Sorption experiments were conducted for 319 days in an inert atmosphere with four cementitious
materials. These materials contained varying amounts of blast furnace slag, a highly reducing cement
additive, ranging from ~0 to 90%. The samples included: Aged Cement (unknown, low % slag, in a 50 year
old, field-aged cement), Vault 2 grout (17% slag), TR547 saltstone (45% slag), and TR545 saltstone (90%
slag). Two initial 99 Tc concentrations (2.5 × 10 -8 , 1.0 × 10 -7 M) were used in triplicate batch reaction systems
with pH values ranging from ~11.3 to 12.2 and Eh values ranging from approximately -300 to -500 mV.
Here we describe the change in 99 Tc aqueous concentrations, measured distribution coefficients (K d ), and
apparent solubility values as a function of time.
Technetium sorption to the cementitious materials tended to reach equilibrium between 154 and 319
days, with K d values ranging from approximately 100 to >8,000 mL∙g -1 . These K d values were not correlated
with percent slag content. The batch systems also approached a final aqueous 99 Tc concentration of 10 -9 M
for Vault 2 and TR545, and 10 -8 M for TR547, which is very close to the expected solubility limit for
99 TcO 2 ∙1.6H 2 O under these high alkaline conditions. No removal of 99 Tc was detected in the no-solids
control systems. Additionally, oxidation state measurements determined that 99 Tc(VII) was the dominant
aqueous oxidation state remaining at the end of the sorption experiment. These data therefore provided
strong evidence that solubility and reductive precipitation, not adsorption (K d ), was the primary factor
controlling 99 Tc removal from the aqueous phase. Furthermore, comparison of aqueous speciation modeling
with our experimental results also supported this conclusion. After the 319 day sorption period, the
suspensions were removed from the inert atmosphere, and a desorption experiment under oxidizing
conditions was conducted for 20 days. Aqueous 99 Tc concentrations and Eh increased significantly within 24
hours after removal due to the introduction of oxygen, and continued to increase over the remaining time
period of the desorption experiment. Conditional K d values at 20 days ranged from approximately

PA5-10
SORPTION OF CESIUM ON PEAT
M. Lusa 1) , S.Virtanen 1) , J. Lempinen 1) , A.T.K. Ikonen 2) , A.-M. Lahdenperä 3) , J. Lehto 1)
1) Laboratory of Radiochemistry, Department of Chemistry, University of Helsinki, P.O. Box 55, 00014
University of Helsinki, Finland
2) Posiva Oy, Olkiluoto, 27160 Eurajoki, Finland
3) Saanio & Riekkola Oy, Consulting Engineers, Laulukuja 4, 00420 Helsinki, Finland
In Finland spent nuclear fuel is planned to be disposed of into a deep bedrock repository in the crystalline
bedrock at the Olkiluoto Island locating in south-western Finland. The dissolution of radionuclides into
ground water from the nuclear waste and the potential migration of these substances into the overburden may
finally result in the contamination of biosphere. Among the most important radionuclides in the long-term
safety assessments is 135 Cs, due to its large inventory in spent nuclear fuel and a very long half-life (2.3 My).
In this study the sorption of cesium on peat from the Lastensuo mire situated in the western coast of Finland
was examined. Peat samples from six different layers between the depths of 0.5 and 6.0 m were studied using
model laboratory experiments with added radionuclides to determine distribution coefficients (K d ) of cesium.
In this study batch model experiments were performed by first stabilising 0.2 g of air-dried peat samples and
25 ml of bog water model solution for two weeks, where after 134 Cs tracer was added. The solution - peat
mixture was left to equilibrate for 1 to 84 days under constant stirring. After incubation the samples were
centrifuged, filtered through a 0.2 µm syringe filter, pH of the solution was measured and the solution was
used for the gamma spectrometric determination of 134 Cs activity. K d values were calculated from the
difference between the initial and final 134 Cs activity. In addition, the cation exchange capacity (CEC) of the
peat layers was determined using 1M ammonium acetate (pH 4.5) extraction. The extracted Na, K, Ca, Mg,
Fe and Al concentrations were determined using ICP-MS and the CEC values were calculated as a sum of
these cations.
Figure 1 presents the distribution coefficients of cesium as a function of depth and CEC while in Figure 2 the
distribution coefficients of cesium as a function of pH and equilibrium time are shown. As seen in Figure 1A
the distribution coefficients of cesium increase as a function of depth. From the Figure 1B similar increasing
trend between the distribution coefficients and CEC can be seen. The CEC values of the peat layers
increased as a function of depth from 11 meq/kgDW in the upper layer to the 34 meq/kgDW in the lowest
peat layer. This is likely a consequence of the difference in the organic matter and mineral matter content of
the different peat layers. In the upper layers from 0.5 m to 4.5 m the average organic matter content was 99.2
%, but in the lowest layer at 5.5 – 6.0 m the organic matter content decreased to an average of 91.1 %. In the
lowest peat layer there is evidence of clay minerals, such as illite, in the XRD analyses of the samples. The
clay minerals serve as more selective sorbents for cesium than organic matter. Although the cations with
similar ionic radii and charge e.g. K and Na predictably compete with cesium for adsorption sites, clay
minerals generally show high selectivity for cesium. The difference in the mineral matter content and clay
content between upper and lower layers presumably results in the difference between both the K d values of
cesium and the CEC values. From the Figure 2A it can be seen that the K d values of cesium increase as a
function of pH. This is presumably because in the upper layers, where the organic matter content is higher
also the pH values are lower due to organic acids from the degradation of the organic matter. As the organic
matter content decreases in the lowest layer the pH of the layer increases. From Figure 2B it can be seen that
the K d values of cesium increase as the incubation time increases from 1 day to 28 days. The increase is
however relatively low from 55 ml/gDW to 77 ml/gDW. The laboratory batch experiments with relatively
short incubation times, may not allow complete attainment of equilibrium of cesium sorption. The results
from the tests with longer incubation times up to 84 days will clarify this trend.

A
Kd (Cs-134) ml/g DW
65
60
55
50
45
40
35
30
0 2 4 6 8
Depth of the peat layer (m)
Figure 1. The experimental K d values for 134 Cs as a function of depth (A) and CEC (B) for the 7 days
equilibrium time.
B
Kd (Cs-134) ml/gDW
65
60
55
50
45
40
35
30
10 15 20 25 30 35
CEC (meq / 100g D.W.)
A
B
Kd (Cs-134) ml/g DW
65
60
55
50
45
40
35
30
3.3 3.5 3.7 3.9 4.1 4.3 4.5
pH
Kd (Cs-134) ml/kg DW
90
80
70
60
50
40
30
20
10
0
0 4 8 12 16 20 24 28 32
Equilibrium time (days)
Figure 2. The experimental K d values for 134 Cs as a function of pH (A) for the 7 days equilibrium time and as
the function of equilibrium time (B) for the layer from 5.5 to 6.0 m.
PA5-11
SORPTION OF THORIUM FROM AQUEOUS SOLUTIONS USING GRAPHENE OXIDE ANG
IRRADIATED GRAPHENE OXIDE
Li Yan 1,2 , Zhong Zheng 1 ,Chunli Wang 1 , Chunli Liu 1,* Wangsuo Wu 2,*
1. Beijing National Laboratory for Molecular Science, Radiochemistry & Radiation Chemistry
Key Laboratory for Fundamental Science, College of Chemistry and Molecular Engineering,
Peking University, Beijing 100871, P.R. China
2. Radiochemistry Laboratory, School of Nuclear Science and Technology, Lanzhou University,
Lanzhou, 730000, P. R. China
Author for correspondence liucl@pku.edu.cn,wuws@lzu.edu.cn
Graphene oxide [1,2] is one of the most concerning carbon nano-material in scientific communities recently.
Because of its special surface structure[3,4], such as large specific surface area and crumples, it can be
concerned as a kind of suitable sorbents to extraction of radionuclides from aqueous solutions. Radiationinduced
grafting technique is one of remarkable achievements in the field of graphene-based
functionalization[5]. Because of the feature of scalable, simple and cost-effective, γ-rays [6] have been
widely used to synthesis adsorbents for the removal of heavy metal ions from waste water.
In this paper, graphene oxide(GO) was synthesized from flake graphite and characterized by a series of
methods (such as XRD, TEM, AFM, FT-IR, Raman, XPS, TGA). Irradiated graphene oxide was prepared
under the absorbed dose 200kGy with 160Gy/s. The sorption of Th(IV) on GO-based materials was
investigated as a function of contact time, solid-to-liquid ratio (m/V), pH, ionic strength, FA and HA using
batch technique. The results show that GO-based materials is one kind of promising sorption materials which

has a higher sorption capacity for Th(IV). The sorption processes of Th(IV) can be described accurately by
the pseudo-second order rate model.
The aims of this studies are (a) to prepare few-layered graphene oxide nanosheets from nature flake graphite
using modified Hummers method; (b) to obtain irradiated graphene oxide using radiation-induced grafting
technique (c) to investigate the influence of pH, ionic strength, m/V ratio, foreign ions and humic acid in the
sorption of Th(IV) onto graphene oxide and irradiated graphene oxide; (c) to determine the sorption isotherm
and to simulate data with Freundlich and Langmuir models; (d) to highlight the possible application of fewlayered
GO-based materials as suitable adsorbents for the removal of Th(IV).
The results show that: graphene oxide based materials is a promising sorption material and has high sorption
capacity for Th(IV). The sorption processes of Th(IV) can be described accurately by the pseudo-second
order rate model. The capacities of Th(IV) on graphene oxide and irradiated graphene oxide are 5.86×10 -
4 mol/g and 2.12×10 -4 mol/g, respectively. Compared to Langmuir model, Freundlich model could fit the
experimental data better, according to the high relative coefficients. The negative value of ΔG ѳ and positive
value of ΔS ѳ indicates that the sorption of Th(IV) onto graphene oxide-based materials is a spontaneous and
endothermic nature process.
Figure1. C 1S XPS spectra of graphene oxide(a) and Irradiated graphene oxide(b)
[1] Z.H. Huang, X. Zheng, W. Lv, M. Wang, Q.H. Yang, F. Kang, (2011) Adsorption of Lead (II) Ions from Aqueous
Solution on Low-Temperature Exfoliated Graphene Nanosheets, Langmuir : the ACS journal of surfaces and colloids.
[2] M. Hirata, T. Gotou, S. Horiuchi, M. Fujiwara, M. Ohba, (2004)Thin-film particles of graphite oxide 1: High-yield
synthesis and flexibility of the particles, Carbon 42 2929-2937.
[3] S. Park, R.S. Ruoff, (2009)Chemical methods for the production of graphenes, Nat Nanotechnol 4 217-224.
[4] J.T. Yang, M.J. Wu, F. Chen, Z.D. Fei, M.Q. Zhong, (2011)Preparation, characterization, and supercritical carbon
dioxide foaming of polystyrene/graphene oxide composites, J Supercrit Fluid 56 201-207.
[5] L. Chen, Z. Xu, J. Li, Y. Li, M. Shan, C. Wang, Z. Wang, Q. Guo, L. Liu and G. Chen (2012) Journal of materials
chemistry
[6]. Y. Zhang, L. Xu, L. Zhao, J. Peng, C. Li, J. Li and M. Zhai (2012) Carbohydrate Polymers

PA5-12
SORPTION OF U(VI) ON LAYERED DOUBLE HYDROXIDES OF Mg, Al, AND Nd FROM
COMPLEX CHEMICAL COMPOSITION SOLUTIONS
N. A. Konovalova, S. A. Kulyukhin, E. P. Krasavina
Frumkin' Institute of Physical Chemistry and Electrochemistry, Russian Academy of Science, Leninskii
prospekt, 31, 119071 Moscow, Russia, E-mail: kulyukhin@ipc.rssi.ru
The development of novel materials for localizing radionuclides from different media including liquid media
is currently an acute issue. These materials can be various minerals (bentonite, clinoptilolite, and
montmorillonite), as well as layered double hydroxides - synthetic analogs of hydrotalcite, supramolecular
dimeric systems, and compounds of the composition [(M 2+ ) 1-x·(M 3+ ) x (OH) 2 ][(A n- ) x/n·mH 2 O], where M 2+ and
M 3+ are cations in the 2+ and 3+ oxidation states, respectively, and A n– is virtually any anion or anionic
complex.
We studied the sorption of U(VI) on layered double hydroxides (LDH) of Mg, Al, and Nd, аs well as on
layered double oxides (LDO) of Mg and Al from aqueous solutions of different compositions. We found that
the sorption of the carbonate complex [UO 2 (СО 3 ) 3 ] 4- by LDH-Mg-Al, which arose upon contact of LDO-Mg-
Al with water, was poor. For 3.3⋅10 -3 mol/l solutions of [UO 2 (СО 3 ) 3 ] 4- , K d were not higher than 1.0 ml/g at
V/m = 50 ml/g after 24-hour contact between the liquid and solid phases (τ). The sorption of U(VI) on LDH-
Mg-Al-NO 3 from aqueous nitrate solutions was poor (K d for U(VI) were not higher than 1.0 ml/g at τ = 24
hours and V/m = 50 ml/g). On the other hand, U(VI) effectively sorbed on LDH-Mg-Al and LDH-Mg-Nd,
with CO 3 2- and OH - being in the interlayer space, from 10 -3 mol/l aqueous solutions of UO 2 (NO 3 ) 2 (K d for
U(VI) were >5⋅10 3 ml/g at τ = 15 min). Increasing the UO 2 2+ concentration to 10 -1 mol/l led to a significant
decrease in K d (to 5⋅10 3 ml/g to 70 ml/g for LDH-Mg-Al-СО 3 and from 170 ml/g to ∼0 ml/g for LDH-
Mg-Nd-СО 3 . In the aqueous solution containing 10 -3 - 5⋅10 -2 mol/l CO 2- 3 , hardly any sorption of U(VI) on
LDH-Mg-Nd-СО 3 took place (К d were not higher than 16 ml/g at V/m = 50 ml/g). For LDH-Mg-Al-СО 3 , the
sorption of U(VI) abruptly decreased (К d dropped from >5⋅10 3 to ∼0 ml/g at V/m = 50 ml/g).
We also studied the influence of Sr 2+ ions on the sorption of U(VI) microquantities, as well as the influence
of UO 2 2+ ions on the sorption of 90 Sr and 90 Y microquantities from aqueous solutions on LDH-Mg-Nd-CO 3 .
In addition, we investigated the effects of Na + , Ca 2+ , Cl - , and SO 4 2- ions, as well as of the рН of the initial
solution on the sorption of U(VI) on LDH-Mg-Al-CO 3 from aqueous solutions of UO 2 (NO 3 ) 2 .
The obtained results allow us to conclude that natural minerals referring to hydrotalcites of the composition
Mg n Al(OH) 6 [(CO 3 ) 1/2 ∙mH 2 O] can accumulate U(VI) by sorbing UO 2+
2 from aqueous media, and LDH-
Mg-Nd-CO 3 may find use for withdrawing 90 Sr and radionuclides of the trivalent f-elements, and also U(VI)
from aqueous solutions of complex chemical compositions.

PA5-13
SYNTHESIS AND CHARACTERIZATION OF Ba 2 TiSi 2 O 8 FOR SORPTION STUDIES
Aishwarya. S. Kar 1) , M. Sahu 1) , M. Keskar 2) , B. S.Tomar 1)*
1) Radioanalytical Chemistry Division, Bhabha Atomic Research Centre, Mumbai-400085
2) Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai-400085
Titanosilicates are a versatile class of compounds having potential application in removing long-lived
radionuclides from nuclear waste effluents owing to their high ion exchange capacity and radiation stability.
They can be transformed into durable waste forms after sorption of radionuclides [1-2]. The discovery of
titanosilicates initiated the synthesis of molecular sieves containing titanium and their evaluation for sorption
of large number of radionuclides [3-4]. In order to understand the sorption mechanism of titanosilicates,
relatively simpler titanosilicate Ba 2 TiSi 2 O 8 , (BTS) the fresnoite mineral, was synthesised using gel
combustion method.
Titanyl acetyl acetonate, tetra ethyl ortho silicate and barium carbonate were dissolved in nitric acid. Glycine
was added as complexing agent and the resultant solution was heated at 100 0 C until homogenous gel was
formed which was burnt in a furnace preheated at 500 0 C.The resultant carbonaceous powder was heated at
600 0 C.The resultant white powder was pelletised with a hardened steel die with pressure 500MPa. The pellet
was heated in a furnace at 900 0 C for 19 hours. The XRD of the resultant powder was taken which
corroborated with the XRD pattern of Ba 2 TiSi 2 O 8 in PC-PDF (22-0513) as shown in figure 1. The surface
area of Ba 2 TiSi 2 O 8 determined by BET was found to be 9 m 2 /g. . The surface charge density of the colloidal
suspension of Ba 2 TiSi 2 O 8 in 0.001M NaClO 4 was measured by dynamic light scattering (DLS) technique.
The pH pzc was found to be 2.68.
Kinetic studies on sorption of 244 Cm by Ba 2 TiSi 2 O 8 were carried out at pH 7.0. The equilibration sorption ~
98% was reached in 24 hours as shown in Figure 2. The sorption data under varying conditions of pH
revealed almost quantitative sorption above pH 6. The solubility of the mineral oxide in water was also
studied at pH 4.8 and 6.6, by measurement of elemental concentration by ICP-AES and ICP-MS in the
supernatant of the suspensions. Concentration of Ba and Si in the supernatant was found to be much higher
than that of Ti.
1000
800
28.98
100
80
Intensity
600
400
200
16.98
20.82
22.52 23.28
27
33.22
34.36
37.52
40.52 41.9
43.7
45.02
45.98 47.2248.5
56.8
58.2
% Sorption
60
40
20
0
10 20 30 40 50 60
2θ
0
0 10 20 30 40 50
Time (hrs)
Figure 1. : XRD pattern of Ba 2 TiSi 2 O 8
Figure 2 : Kinetics of sorption of 244 Cm by Ba 2 TiSi 2 O 8
1. K. Popa, C. C. Pavel, Desalination 293, 78 (2012).
2. L.M. Wang, J. Chen, R.C. Ewing, Current Opinion in Solid State and Materials Science 8, 405(2004)
3. L. Al-Attar, A. Dyer, R. Blackburn, J. Radioanal. Nucl. Chem. 246,451(2000).
4. P. Misaelidis, A. Godelitsas, A. Filippidis, D. Charistos, I., Sci. Total Environ. 173/174,237(1995).

PA5-14
URANYL IONS SORPTION TO TIO 2 AND INTERACTION WITH SORBED FA: EXPERIMENTS
AND MODELING
Y. Ye, X. Wang, N. Guo, Z. Guo, W. Wu
Radiochemistry Laboratory, School of Nuclear Science and Technology, Lanzhou University, Lanzhou,
730000, China
In the environment, an important factor of radionuclide mobility is their interaction with mineral–water
interfaces. To predict radionuclide mobility, it is necessary to understand fundamental processes such as
surface precipitation and surface complexation. Studies of uranium sorption onto mineral surfaces have great
practical importance for risk assessment 1-3 .In recent years it turned out that many groundwaters in
granitoidic environment contain excessive amounts of dissolved U 4 .
In this work, experiments and modeling studies are performed to elucidate the interaction of fulvic acid (FA)
with uranyl ions in the presence of TiO 2 surfaces. FA is strongly bound to TiO 2 , and has a very strong effect
on the U(VI) sorption. U(VI) sorption to TiO 2 in the presence and absence of sorbed FA can be well
predicted with the SCD model (surface and complex distribution). According to the model calculations, the
nature of the interaction between FA and U(VI) at the TiO 2 surface is mainly surface complex formation.
This is the first time that effects of natural organic matter (NOM) on the sorption of a cation are predicted
successfully using an integrated ion-binding model for oxides and for humics that accounts for the chemical
heterogeneity of NOM.
1] Van Loon, L. R.; Baeyens, B.; Bradbury, M. H. Appl. Geochem. 2009, 24, 999.
[2] Rabung, T.; Pierret, M. C.; Bauer, A.; Geckeis, H.; Bradbury, M. H.; Baeyens, B. Geochim. Cosmochim. Acta,
2005, 69, 5393.
[3] Bradbury, M. H.; Baeyens, B. Geochim. Cosmochim. Acta, 2005, 69, 5391.
[4] Gustafsson, J. P.; Dassman, E.; Backstrom, M. Appl. Geochem. 2009, 24, 454.
100
80
U(VI) sorption %
60
40
no FA
[FA]=20mg/l
20
0
2 3 4 5 6 7 8
pH
Fig. 1 Sorption of U(VI) onto TiO 2 vs. pH in the absence and presence of FA, C [U(VI)] =8.024E-5 mol/ L, C [FA] =20 mg/L,
m/ V = 20 g/ L , T = 22±1 ℃, I =0.1 mol/ L (NaCl).

100
80
FA sorption %
60
40
20
no U(VI)
C(U(VI))=4.012E-4mol/L
C(U(VI))=8.024E-5mol/L
0
2 3 4 5 6 7 8 9
Fig. 2 Sorption of FA onto TiO 2 vs. pH in the absence and presence of FA, C [FA] =20 mg/L, m/ V = 20 g/ L , T = 22±1 ℃,
I =0.1 mol/ L (NaCl).
pH
PA5-15
MODELING THE ADSORPTION OF EU(III) AND AM(III) ON GRANITE USING A
GENERALIZED COMPOSITE APPROACH
Z. Guo, Z. Chen, Q. Jin, W. Wu
School of Nuclear Science and Technology, Lanzhou University, Lanzhou, 730000, China.
According to the guidelines for the R&D for geological disposal of high-level radioactive waste in
China [1], Beishan (in Gansu province) granite has been considered as a preliminary selection of the host
rock in Chinese case. Therefore, it is necessary to evaluate adsorption of radionuclides on the granite, but
this is difficult using Component Additivity (CA) modeling approach because effective surface area of each
dominant mineral in the igneous rock is usually unknown due to mutual coating of the components. But on
the other hand, Generalized Composite (GC) approach has been demonstrated to be a powerful tool for some
relatively complicated systems [2-3]. The GC approach assumes that surface complexation reactions occur
on “general” sites and ignores electrostatic effects.
In the present study, the adsorption of Eu(III) and Am(III) on crushed Beishan granite (which was from
the borehole BS03 at a depth of 600 m) was investigated and interpreted by a surface complexation model
(SCM) using GC approach. First, the SCM was set up based on experimental adsorption data for Eu(III) in
NaCl solutions [4]. The model involves a cation exchange reaction and two inner-sphere surface
complexation reactions (see Table 1). Second, the SCM was verified by experimental data collected in CaCl 2
solutions and those collected in synthesized underground water by considering additional cation exchange
reaction between Na + and Ca 2+ (reaction (4) in Table 1). Finally, the SCM was used to predict Am(III)
adsorption on Beishan granite by assuming that the modeling parameters for Am(III) were identical to those
for Eu(III). It was found that the prediction and the experimental data were in good agreement.
Table 1 The site density of Beishan Granite and the modeling parameters for Eu(III) adsorption on Beishan
granite (BS03, 600m). Nagra/PSI data base [5] was used for Eu(III) reactions in the aqueous solution.

Sites
Site densities
X - 4.04×10 -5 mol/m 2
≡SOH 7.35×10 -6 mol/m 2
Reactions
logK
3XNa + Eu 3+ X 3 Eu + 3Na + (1) 2.2±0.3
≡SOH + Eu 3+ ≡SOEu 2+ + H + (2) -0.6±0.2
≡SOH + Eu 3+ + 2H 2 O ≡SOEu(OH) 2 + 3H + (3) -17.2±0.2
XNa + 0.5Ca 2+ XCa 0.5 + Na + (4) 0.3±0.1
[1] Guidelines for the R&D for geological disposal of high-level radioactive waste. Beijing: China Atomic Energy
Authority (CAEA), Ministry of Environmental Protection, Ministry of Science and Technology, 2006 (in Chinese).
[2] Davis J A, Meece D E, Kohler M, Curtis G P. Approaches to surface complexation modeling of Uranium(VI)
adsorption on aquifer sediments. Geochimica et Cosmochimica Acta, 2004, 68: 3621-3641.
[3] Tertre E, Hofmann A, Berger G. Rare earth element sorption by basaltic rock: Experimental data and modeling
results using the “Generalised Composite approach”. Geochim Cosmochim Acta, 2008, 72: 1043-1056.
[4] Guo Z, Chen Z, Wu W, Liu C, Chen T, Tian W, Li C, The adsorption of Eu(III) on Beishan granite, SCIENTIA
SINICA Chimica, 2011, 41: 907-913.
[5] Hummel W, Berner U, Curti E, Pearson F J, Thoenen T. (2002): Nagra/PSI Chemical Thermodynamic Data Base
01/01. Universal Publishers/uPUBLISH.com USA. Also issued as Nagra Technical Report NTB 02-16, Nagra,
Wettingen, Switzerland.
PA5-16
THE BIGRAD CONSORTIUM - THE INFLUENCE OF IRON OXIDE CRYSTALLIZATION ON
THE MOBILITY OF NEPTUNIUM
Pieter Bots 1 , Sam Shaw 1 , Gareth T.W. Law 2 , Timothy Marshall 1 , J. Frederick W. Mosselmans 3 , Francis R.
Livens 2 , Melissa A. Denecke 4 , Jörg Rothe 4 , Kathy Dardenne 4 , Katherine Morris 1
1) School of Earth, Atmospheric and Environmental Sciences, The University of Manchester, Manchester,
M13 9PL, UK
2) School of Chemistry, The University of Manchester, Manchester, M13 9PL, UK
3) Diamond Light Source, Didcot, OX11 0DE, UK
4) Institut für Nukleare Entsorgung (INE), Forschungszentrum Karlsruhe, D-76021 Karlsruhe, Germany
The current preferred management strategy for the long-lived, higher activity radioactive wastes is
containment in a Geological Disposal Facility (GDF) in the deep sub-surface. The GDF will be designed for
the waste form: the Intermediate Level Waste (ILW) will be grouted in steel drums and backfilled with a
cementitious material in a system that will also use cement and iron rock bolts as part of the engineering.
Over time, groundwater will re-saturate the GDF; this will result in the formation of a hyperalkaline plume
of cement leachate (pH 13 - 10) which will move through, and react with, the geosphere over time 1 . The
interaction of the cement leachates with the surrounding host rock will result in the formation of a
Chemically Disturbed Zone (CDZ). Initially, the GDF will be open to the atmosphere and hence the CDZ
will be aerobic: Corrosion of iron, reoxidation of Fe(II) from the deep subsurface and iron flocs in the wastes
themselves will contribute to ferric and ferrous iron oxides being ubiquitous within the engineered barrier
system. Initially, meta-stable iron(III)(hydr)oxide minerals, such as ferrihydrite (Fe 5 O 7 (OH)·4H 2 O), will
transform to more stable goethite (α-FeOOH) and hematite (α-Fe 2 O 3 ). Latterly chemically reducing
conditions are expected to occur as the oxidation of reduced components within the GDF (i.e. Fe-metal)
begins to dominate the system. This will induce formation of ferrous iron bearing minerals such as magnetite
(Fe 3 O 4 ).
Neptunium-237 is a long lived transuranic element with a half-life of 2.14 x 10 6 y and is a key radionuclide
to consider in the assessment of geological disposal facilities throughout their evolution. Neptunium is also a
redox active element with pentavalent neptunium (Np(V) as NpO 2 + ) dominant in oxic environments and
tetravalent neptunium (Np(IV) as Np 4+ ) dominant in reducing environments. Previous research has shown

that neptunium is scavenged by iron oxide minerals, can be reduced on contact with magnetite and as a result
of microbially induced reduction in sedimentary systems 2,3 . However, the details of the interactions of
neptunium and its fate during the crystallisation of iron oxide minerals remain poorly constrained. This is a
critical lack of knowledge when considering any safety case for high pH conditions relevant to geological
disposal of intermediate level wastes.
Here, we present the results of experiments on the fate of neptunium during the transformation of ferrihydrite
to hematite and magnetite. Np(V) was adsorbed to ferrihydrite in a Ca(OH) 2 solution that was representative
of a pore water in an evolved repository (Old Cement Leachate; OCL) with a pH ~10.5. To determine the
fate of neptunium during the transformation of ferrihydrite to hematite, Np(V)-labelled ferrihydrite was aged
at 60°C for two weeks 4 . To determine the fate of neptunium during the crystallization of ferrihydrite in
reducing conditions, Np(V)-labelled ferrihydrite was converted to magnetite by stochiometric addition of
Fe(II) and the subsequent aging of the freshly crystallized magnetite for several days. During all
manipulations, the pH was monitored and adjusted where necessary. Additionally, neptunium concentrations
were tracked during all experiments and after each manipulation, solid samples were collected and analysed
using XRD and TEM, from parallel non-active experiments, for solid phase characterization. The solid
samples were also analysed using Extended X-ray Absorption Fine Structure (EXAFS) spectroscopy at the
INE beamline, ANKA synchrotron at the Karlsruhe Institute of Technology (Germany) to determine the
neptunium speciation and local bonding environment. In this work, we used the aqueous and solid phase
geochemistry and EXAFS analyses to understand the mechanisms of neptunium interactions during mineral
alteration. We also examined the final fate, speciation and redox state of the radionuclide in the reacted
mineral phases. During all experiments performed, > 99% of the added Np was removed from solution.
Furthermore, the Np XANES and EXAFS revealed that during the transformation to magnetite, neptunium
was reduced to Np 4+ . Additionally, the EXAFS spectra from Np adsorbed to the iron oxide surfaces are
distinctly different than those from Np associated with the crystalline end products of the experiments,
suggesting that Np may have become incorporated into the structure.
1. J. S. Small, O. R. Thompson, in Scientific Basis for Nuclear Waste Management Xxxii, N. C. Hyatt, D. A. Pickett,
R. B. Rebak, Eds. (2009), vol. 1124, pp. 327-332.
2. G. T. W. Law et al., Environmental Science & Technology 44, 8924 (2010).
3. K. Nakata et al., Radiochimica Acta 90, 665 (2002).
4. S. Shaw, S. E. Pepper, N. D. Bryan, F. R. Livens, American Mineralogist 90, 1852 (2005).
PA5-17
MECHANISM OF STRONTIUM SORPTION AND INCORPORATION DURING THE ALKALINE
ALTERATION OF SEDIMENTS BY KOH DOMINATED CEMENT LEACHATE.
Sarah H. Wallace 1 , Samuel Shaw 1,2 , Katherine Morris 2 , Joe S. Small 3 and Ian T. Burke 1
1 Earth Surface Science Institute, School of Earth and Environment, University of Leeds, Leeds, LS2 9JT, UK.
2
Research Centre for Radwaste and Decommissioning and Williamson Research Centre, School of Earth,
Atmospheric and Environmental Sciences, University of Manchester, Manchester, M13 9PL, U.K.
3
National Nuclear Laboratory, Risley, Warrington, Cheshire, WA3 6AE, U.K.
The use of cementitious materials is ubiquitous at nuclear facilities around the world, where it is utilised in
buildings, storage facilities and also the packaging of radioactive waste. When fresh cement and concrete
comes into contact with water, a highly alkaline (0.3 – 0.7 mol L -1 OH - ) K- and Na-rich fluid forms,
commonly known as young cement water (YCW). Interaction of such a hyperalkaline solution with
aluminosilicate minerals in natural soils (e.g. clay minerals) and rock promotes the dissolution and
recrystallisation of these minerals to neoformed phases (e.g. zeolites). These reactions lead to the formation
of an alkaline disturbed zone in the geosphere surrounding the cement structures, which significantly alters
the chemical (e.g. adsorption capacity) and physical (e.g. porosity) properties of the soils or sediment.
Results are presented from one year batch experiments [1] where K-rich hyperalkaline pH 13.5 young
cement water was reacted with sediments to investigate the effect of high pH, mineral alteration and
secondary mineral precipitation on 90 Sr sorption. At circumneutral Sr2+ adsorbs to sediments via formation
of simple outer-sphere complexes (Figure 1; LHS). These complexes are susceptible to desorption via cation
exchange. At a pH of 10 the adsorption of 90 Sr was low (K d ~1 L Kg -1 ) [2], and if this trend were to continue

it predicts that at pH 13.5 90 Sr adsorption would be extremely low due to high ionic strength (>0.1 mol L -1 in
YCW). After reaction with YCW, However, Sr sorption was found to be >75 % in all samples up to 365
days, and 98 % in a sample reacted for 365 days at 70 °C. SEM analysis of sediment samples reacted at room
temperature showed surface alteration and precipitation of a secondary phase, likely a K-rich aluminosilicate
gel. The presence of Sr-Si(Al) bond distances in Sr K-edge EXAFS analysis suggested that the Sr was
present as an inner-sphere adsorption complex (Figure 1; centre). Sequential extractions, however, found the
majority of this Sr was still exchangeable with Mg 2+ at once the was lowered to pH 7, suggesting that he
sorption mechanism reverted to outer-sphere at lower pH.
For the sample reacted for one year at 70 °C, EXAFS analysis revealed clear evidence for ~6 Sr-Si(Al)
backscatters at 3.45 Å, consistent with Sr incorporation into the neoformed K-chabazite phase that was
detected by XRD and electron microscopy (Figure 1; RHS). Once incorporated into chabazite, 90 Sr was not
exchangeable with Mg 2+ and chemical leaching with pH 1.5 HNO 3 was required to remobilise 60 % of the
90 Sr. These results indicate that in high pH cementitious leachate there is significantly enhanced Sr retention
in sediments due to changes in the adsorption mechanism and incorporation into secondary silicate minerals.
This suggests that Sr retention may be enhanced in this high pH zone and that the incorporation process may
lead to irreversible exchange of the contaminant over extended time periods.
[1] Wallace S.H., Shaw S., Morris K., Small J. S. and Burke I.T. Alteration of Sediments by Hyperalkaline K-Rich
Cement Leachate: Implications for Strontium Adsorption and Incorporation. Environmental Science and Technology
(2013). In press, DOI: 10.1021/es3051982.
[2] Wallace S.H., Shaw S., Morris K., Small J. S., Fuller A. J. and Burke I.T. Effect of groundwater pH and ionic
strength on strontium sorption in aquifer sediments: Implications for 90Sr mobility at contaminated nuclear sites,
Applied Geochemistry. (2012) 27 (8), 1482-1491.
Figure 1. Mechanisms of Sr-sorption observed during alkaline alteration of aquifer sediments with young cement water
leachate.

PA5-18
RETENTION OF SELENIUM ON GRANITE: ANALYSIS OF GRANITIC SURFACE
FOR THE PURPOSE OF DETERMINATION OF SELENIUM SPECIES
K. Videnská (1),(2) , V. Havlová (1) , M. Vašinová Galiová (3),(4) , V. Kanický (3) , P. Sajdl (5)
(1) Fuel Cycle Chemistry Department, ÚJV Řež, a.s., 250 68 Husinec-Řež, Czech Republic
(2) Department of Analytical Chemistry, Institute of Chemical Technology, Prague 6, 166 28 Prague, Czech
Republic
(3) Masaryk University, Faculty of Science, Department of Chemistry, Kotlářská 2, 611 37 Brno, Czech
Republic
(4) Masaryk University, Central European Institute of Technology (CEITEC), Kamenice 5, 625 00 Brno,
Czech Republic
(5) Institute of Chemical Technology, Department of Power Engineering, Technická 5, 166 28 Prague 6,
Czech Republic
Selenium ( 79 Se, T 1/2 = 3.77·10 5 yrs) belongs among seven long-lived fission products present in spent nuclear
fuel (SNF) and in the wastes resulting from the spent fuel reprocessing. Because of the long lifetime of this
radioisotope, the complex selenium chemistry, its high mobility and mainly anionic character, selenium
contributes significantly to the risk associated with the storage of radioactive waste in deep underground
repositories in geological environment. Therefore, knowledge of migration behaviour under different
conditions can significantly improve input into performance and safety assessment models. Granite is
considered as a potential host rock of deep underground repositories for SNF in many countries. Granitic
rocks consist usually of quartz, feldspar, plagioclase (main components), mica, chlorite, kaolinite (minor
components).
The presented work is focused on interaction of selenium species (SeO 3 2- , SeO 4 2- ) with granitic rock. Granite
was sampled from the bore core, originating from the depth 97.5-98.7 m (Melechov massive, Centre
Bohemian Massive, Czech Republic). Two forms of granite were used for sorption experiments, crushed
granite and granitic cubes. The crushed granite was sieved into five defined fractions. Grain size of analysed
crystalline rock was in the range of 0.063 – 1.25 mm. Each fraction was analysed by X-ray diffraction.
Results showed inhomogeneous mineral composition through the fractions. Some granitic fractions
contained the highest concentration of orthoclase and also contained the lowest concentration of minor
minerals as chlorite and mica. On the contrary, some fractions contained in comparison lower concentrations
of orthoclase and plagioclase and higher concentration minor minerals.
Synthetic granitic water was used as a solution for selenium solution preparation. Concentration of selenite
and selenate respectively used was 2·10 -5 mol/L. Concentration of selenium was measured by Inductively
Coupled Plasma Mass Spectrometry (ICP-MS).
Stability diagrams of selenium in granitic water, modelled using The Geochemist´s Workbench programme,
showed that selenium will be present at oxidation state +IV and +VI in studied system under oxic conditions.
The static batch experiments were conducted on all granitic fractions under aerobic conditions and lasted
seven days. Selenite showed higher retention than selenate. Selenate was not almost sorbed at all. The results
showed influence of granite composition on adsorption of selenite. Values of distribution coefficient K d ,
describing sorption of selenite, was higher in case of granitic fraction with higher content of minor minerals.
Fe containing minerals are pronounced as a selective Se sorbents that were presumable reduced on their
surfaces. Dark micas in granite are minerals, rich in Fe, increased sorption of anionic species onto mica
enriched fractions can explain such behaviour. On the other hand, fractions enriched in feldspar did not show
increased sorption affinity to selenium at all.
Within experiments with granitic cubes, solution of selenite or selenate at synthetic granitic water was in
contact with the granitic samples for seven days and surface of granitic cubes was analysed by Electron
Spectroscopy for Chemical Analysis (ESCA) and Laser Ablation Inductively Coupled Plasma Mass
Spectrometry (LA-ICP-MS) which provide information about retention of selenium on granite. The aim of
using these methods was identification of potential retention mechanism of selenium. Results of ESCA and

LA-ICP-MS methods were interpreted as surface maps which illustrated distribution of elements (Se, Fe, Na,
Al, etc.) on granite. The results showed that retention mechanism of selenium species was in direct
correlation to the presence of Fe minerals. Selenium was sorbed on granite in place with high content of iron.
PA5-19
SELENIUM FORMS ON ÄSPÖ DIORITE: ANALYSIS OF ROCK SURFACE USING LA-ICP MS
AND XPS
K. Videnská (1),(2) , V. Havlová (1) , M. Vašinová Galiová (3),(4) , V. Kanický (3) , P. Sajdl (5)
(1) Waste Disposal Department Chemistry of Fuel Cycle and Waste Management Division, ÚJV Řež, a.s., 250
68 Husinec-Řež, Czech Republic
(2) Department of Analytical Chemistry, Institute of Chemical Technology, Prague 6, 166 28 Prague, Czech
Republic
(3) Masaryk University, Faculty of Science, Department of Chemistry, Kotlářská 2, 611 37 Brno, Czech
Republic
(4) Masaryk University, Central European Institute of Technology (CEITEC), Kamenice 5, 625 00 Brno,
Czech Republic
(5) Institute of Chemical Technology, Department of Power Engineering, Technická 5, 166 28 Prague 6,
Czech Republic
Selenium ( 79 Se, T 1/2 = 3.77·10 5 yrs) belongs among long-lived fission products present in spent nuclear fuel
(SNF). Due to the long lifetime of the radioisotope, the complex chemistry, high mobility and prevailing
anionic character, selenium contributes significantly to the risk associated with the storage of radioactive
waste in deep geological repositories.
The presented work is focused on study of Äspö diorite surface after contact with solutions of selenite
Se(+IV) and selenate Se(+VI). The aim of the experiments was to identify potential retention mechanism of
selenium on diorite surface and to observe correlation between sorbed selenium and mineral components
present in Äspö diorite.
Diorite rock was sampled from the borehole KA2368A-01 from Äspö hard rock laboratory within FP7 EC
CROCK project (No. 269658). The samples originated from depth 10.65 – 11.12 m. The rock consists of
plagioclase, K-feldspar, quartz, biotite, calcite (major phases); muscovite, titanite, apatite, magnetite (minor
phases). The rock core was cut into thin slices (10102 mm).
Sorption experiments were performed by inserting rock cubes (dimension 1×1×0.2 cm) into 10 mL of Äspö
synthetic groundwater, containing defined amount of selenium. Concentrations of Na 2 SeO 3 or Na 2 SeO 4 in
synthetic groundwater were 2·10 -3 mol/L. All the experiments were held under aerobic conditions.
The spectroscopic techniques Laser Ablation Inductively Coupled Plasma Mass Spectrometry (LA-ICP-MS)
and X-ray Proton Induced Spectroscopy (XPS, also called Electron Spectroscopy for Chemical Analysis,
ESCA) were used to observe Se sorption over the diorite surface. The results were interpreted as surface
maps which illustrated distribution of elements (Se, Fe, Na, Al, etc.) on diorite.
The LA-ICP MS and XPS results partially confirmed the initial assumptions that retention mechanism of
selenium species was in correlation to the presence of Fe minerals. Moreover, positive correlation of
selenium and calcium content on the surface was observed (see Fig. 1). As the methods did not confirm
presence of reduce Se species it can be assumed that mica (as Fe bearing phase) and calcite (Ca bearing
phase) play the important role in formation of Se surface complexes.

Fig. 1: LA ICP-MS surface maps (from the left): a) Fe content, b) Ca content c) Se content. Na 2 SeO 3 solution
(2·10 -3 mol/L) in contact with Äspö diorite samples.
The research leading to these results has received funding from the European Atomic Energy Community's
Seventh Framework Programme (FP7/2007-2011) under grant agreement No. 269658 (CROCK).
PA5-20
A NANO TO MACRO SCALE INVESTIGATION OF MULTI-SITE CAESIUM SORPTION TO
ILLITE AND SEDIMENT
Adam J. Fuller 1 , Samuel Shaw 2 , Caroline L. Peacock 1 , Divyesh Trivedi 3 and Ian T. Burke 1
1 School of Earth and Environment, University of Leeds, Leeds, LS2 9JT, UK
2 School of Earth, Atmospheric and Environmental Sciences, Williamson Building, University of Manchester,
Manchester, M13 9PL, UK
3 National Nuclear Laboratory, Chadwick House, Birchwood Park, Warrington, WA3 6AE, UK
As a high yield fission product 137 Cs is ubiquitous in nuclear wastes. Catastrophic releases of nuclear
material such as those which occurred at Chernobyl, Ukraine and Fukushima, Japan have left large areas of
land contaminated with radionuclides. Leaks and spills at civil and military nuclear sites, such as Sellafield,
UK, also contaminate the subsurface environment. 137 Cs is one of the key radionuclides associated with this
contamination, and a large contributor to external radiation dose. The environmental mobility of 137 Cs is
strongly controlled by its sorption to clay minerals. It is therefore essential to understand the sorption of 137 Cs
to minerals and mixed sediments.
Here we present results from experimental and modelling studies of Cs sorption onto both mixed sediments
(representative of the Sellafield site, UK) and pure illite, which is thought to be the main mineral which
controls Cs sorption [1].
We investigated the effect of Cs concentration, pH and cationic competition on the sorption of Cs to mixed
sediments. This was done through batch sorption experiments and confirmed by multi-site cation exchange
modelling. It was found that Cs was sorbing to three distinct sorption sites within the sediment matrix (see
figure 1). Cs sorbs to the illite Frayed Edge Sites (FES) at very low concentrations (

nanoscale, the process of Cs sorption to the FES. We investigated the structural properties of the FES and
sorption of Cs to them. Although the theory of Cs sorption to the FES has been proposed for many decades
[5] this represents the first direct evidence of this process.
This work represents a strongly coherent model of the environmental behaviour of Cs. It provides both a
mechanistic understanding of sorption to the FES and applied predictions of sorption under a wide range of
likely environmental conditions. It is therefore a major advancement in our understanding of Cs sorption.
Figure 1. Theoretical model of multi-site Cs sorption to illite.
[1] Sawhney, B.L. (1972) Selective Sorption and Fixation of Cations by Clay-Minerals – A Review. Clays & Clay Min.,
20(2): p. 93-100.
[2] Jackson, M.L., et al. (1952) Weathering Sequence of Clay-size Minerals in Soils and Sediments: II. Chemical
Weathering of Layer Silicates. Soil Sci. Soc. Am. Proc., 16(1): p. 3-6.
[3] Bradbury, M.H. and B. Baeyens (2000) A generalised sorption model for the concentration dependent uptake of
caesium by argillaceous rocks. J. Contam. Hydrol., 42(2-4): p. 141-163.
[4] Steefel, C.I., et al., (2003) Cesium migration in Hanford sediment: a multisite cation exchange model based on
laboratory transport experiments. J. Contam. Hydrol., 67(1-4): p. 219-246.
[5] Jacobs, D.G. and T. Tamura, (1960) The Mechanism of Ion Fixation using Radio-isotope Techniques, Trans.7th Int.
Congr. Soil Sci., Madison, p. 206-214.
PA5-21
Tc(VII) IMMOBILIZATION ON GRANITOID ROCKS FROM ÄSPÖ (SWEDEN)
Y. Totskiy, F. Huber, T. Schäfer, H. Geckeis
Institute for Nuclear Waste Disposal (INE), Karlsruhe Institute of Technology (KIT), Karlsruhe, Germany
The generally accepted concept of spent nuclear fuel (SNF) and high-level waste long-term isolation from
the biosphere is disposal in deep geological formations at depths of more than 300-500 meters. Deep
geological repositories are designed with a multi-barrier system, where the host rock formation as a
geological barrier will provide (a) mechanical stability for the storage of waste containers, (b) protection of
the waste from direct human access and (c) isolation of the radiotoxic waste from the biosphere until they
have decayed to approximately the level of natural uranium. Thus, site selection with regard to appropriate
geochemical conditions and geological properties of the respective host rock is an important issue [1]. 99 Tc
with a half-life of 2.14∙10 5 years and a fission yield of ca. 6.14 % is one of the relevant components of HLW
after reprocessing of SNF in the long term. Therefore, information on speciation and mobility in a given
repository system is required for performance assessment.
This work deals with the interaction of Tc(VII) with crystalline rock materials from the Äspö Hard Rock
Laboratory (Sweden). New core material drilled under anoxic conditions was obtained within the EU project
CP CROCK and transferred to KIT-INE. Batch-type sorption studies on Äspö diorite (ÄD) in contact with
Äspö groundwater show that low Tc concentrations (< 10 -8 M) are almost fully retained by un-oxidized
granite after 1-2 months contact time, whereas considerably less retention is found in the case of elevated
concentration (10 -5 M). Immobilization of Tc(VII) strongly depends on sample preservation conditions. Tc
sorption onto oxidized material shows approximately 2.4 times less adsorption compared to samples drilled,

transported and stored under anoxic conditions. This results can be explained by Tc(VII) reduction to poorly
soluble and strongly sorbing Tc(IV) by the ion–exchangeable ferrous iron pool available in the ÄD. The
measured Eh (SHE) values of -80 to -150mV after one month equilibration time of the anoxic samples
corroborate the proposed Tc oxidation state and predict the stability of TcO 2 ∙1.6H 2 O(s) for the 10 -5 M and 10 -
8 M samples and TcO(OH) 2 for the 10 -9 M samples. Based on this thermodynamic calculations a Tc(IV)
species sorption in the low concentration samples and a reduction to poorly soluble TcO 2 ∙1.6H 2 O(s) solid
phase in the higher concentrated samples can be proposed.
Tc(VII) reduction was proven by XPS analysis of granite surfaces after sorption. Exclusively Tc(IV) was
found. Surface ion-exchangeable Fe(II) was quantified [ 2] as approx. 1-3 μg/g for oxidized material, whereas
for the un-oxidized samples higher values of around 4-6 μg/g were obtained. Experimentally determined
solid/liquid distribution ratios for experiments with the lowest Tc concentration are at ~10 3 mL/g under
reducing conditions and at ~3 mL/g for partially oxidized material. These values are in good agreement with
previous work on comparable materials of 10 3 mL/g for reducing and 0 mL/g for oxidizing conditions [3].
Only small Tc desorption from ÄD was found under reducing conditions (< 6%) and was independent of
contact time (from 1 day to 1 month) and groundwater composition (natural Äspö: pH = 7.8, I = 0.18 M and
Grimsel: pH = 9.6, I = 1.3∙10 -3 M groundwater were used). Colloidal phase formation was not detected
during both sorption and desorption studies via ultracentrifugation experiments (90,000 rpm for 60 min).
Furthermore, migration studies with Äspö core material containing a natural fracture, which has been
preserved under anoxic conditions are performed. The core was sealed into a Plexiglas container and
characterized by micro-computer tomography (μCT) for 3D reconstruction of the fracture geometry (voxel
size 15µm) and simulation of fluid dynamics. After hydraulic characterization using the conservative tracer
tritium (HTO) and 36 Cl - to evaluate matrix diffusion (anion exclusion) under different fluid velocities, the
technetium migration is studied using 99 Tc(VII) with 95m Tc(VII) isotope admixture for detection of low
concentrations. The results of the Tc migration experiments at variable flow rates will be discussed in the
light of sorption kinetics studied in batch experiments.
The research leading to these results has received funding from the EURATOM 7 th Framework Programme
FP7/2007- 2011 under grant agreement n° 269658 (CROCK project) and was supported by the German
Federal Ministry of Economics and Technology (BMWi), project VESPA (02E10800).
[1] E.B. Anderson et al., Proceeding of the Khlopin Radium Institute 11, 8 (2006)
[2] G. Heron et al., Environ. Sci. Technol. 28, 1698 (1994)
[3] P. Carbol, I. Engkvist, SKB reports R-97-13 (1997)
PA5-22
INVESTIGATION OF ACTINIDE AND LANTHANIDE SORPTION ON CLAY MINERALS
UNDER SALINE CONDITIONS
A. Schnurr, R. Marsac, Th. Rabung, J. Lützenkirchen, H. Geckeis
Institute for Nuclear Waste Disposal (KIT-INE), Karlsruhe Institute of Technology, Karlsruhe, Germany
The final disposal in deep geological formations is considered as the safest way to keep high level
radioactive waste isolated from the biosphere. Due to their high sorption capacity, their swelling properties
and their low water permeability, clay minerals are of great interest as suitable components of geotechnical
and geological barriers. Therefore, several countries have selected clay formations for their deep geological
disposal projects e.g. Opalinus clay (Switzerland), Callovo-Oxfordian (France) or Boom clay (Belgium). In
other concepts compacted clays have been proposed as backfill and sealing materials. For the safety
assessment of nuclear waste concepts, the potential contact of waste forms with groundwater has to be
considered. Sedimentary clay formations discussed as potential repository host rocks can be in contact with
highly saline (up to 5 molal) groundwater [1]. Only a few studies so far addressed radionuclide sorption
under high ionic strength conditions [2]. The non-electrostatic 2SPNE SC/CE sorption model developed by
Bradbury and Baeyens to describe sorption onto illite and smectite is only valid for relatively low ionic
strength systems (I < 1 M) [3]. The present work focuses therefore on the sorption of trivalent metal cations
(Eu(III)/Cm(III)) and hexavalent uranium onto different natural relevant clay minerals (illite, smectite and
kaolinite) under saline conditions and exclusion of CO 2 .

Batch sorption edge experiments were carried out with three different background electrolytes (NaCl, CaCl 2
and MgCl 2 ) at different ionic strengths ([NaCl] up to 4 M, [CaCl 2 ] and [MgCl 2 ] up to 2 M), at fixed metal
concentration ([Eu] total = 2 x 10 -7 M labeled with 152 Eu for γ-counting or [U(VI)] total = 4 x 10 -7 M) and at
constant solid to liquid ratios (S:L = 2 g/L) over a wide pH c range (3-12). The extent of metal ion adsorption
onto the different clay minerals approaches 100% at pH c > 7 for all investigated electrolyte solutions. In
general, sorption edges shift to higher pH c -values with increasing ionic strength (confirmed for Eu(III), first
hints for U(VI)). In presence of divalent cations, however, actinide/lanthanide sorption is much stronger
suppressed than observed for NaCl solutions. This is explained by the stronger competition of divalent
cations with actinide/lanthanide sorption to cation exchange sites but also to inner sphere sorption sites at the
edges of the clay particles.
The applicability of the 2SPNE SC/CE model [3] was tested to describe Eu(III) sorption at high ionic
strength. A good agreement of experimental data and model prediction was achieved in the NaCl systems at
pH c < 7 where metal ion activities were calculated using the Pitzer approach. However, the model
underestimates sorption at higher pH c . Agreement for Ca/MgCl 2 systems is much better. Deviations might be
due to uncertainties in the Pitzer parameter data set for An(III) in NaCl solutions.
Time resolved laser fluorescence spectroscopy (TRLFS) studies confirm the existence of three distinguished
Cm(III) surface complexes which have been described already earlier for low ionic strength conditions [4]
also at high salinity. This finding demonstrates that the coordination of adsorbed actinide species does not
change noticeably under highly saline conditions and validates speciation assumptions underlying the
2SPNE SC/CE model. However, with an improved detection system, an additional fourth inner-sphere
surface species could be identified at higher wavelengths (λ ~ 610 nm) being comparable to the findings of a
previous study on curium interaction with kaolinite. This species was interpreted as a clay/curium/silicate
ternary complex [5].
Sorption %
100
80
60
40
20
[NaCl] = 0.1 M
[NaCl] = 1.0 M
[NaCl] = 4.0 M
[Eu] initial
= 2x10 -7 M
Illite du Puy:
solid to liquid = 2 g/L
log K D
7
6
5
4
3
[NaCl] = 0.1 M
[NaCl] = 1.0 M
[NaCl] = 4.0 M
[Eu] initial
= 2x10 -7 M
Illite du Puy:
solid to liquid = 2 g/L
0
4 5 6 7 8 9 10 11 12
pH c
2
4 5 6 7 8 9 10 11 12
pH c
Fig. 1: Eu(III) sorption onto illite in percentage uptake (left) and as the logarithm of the distribution
coefficient (right) vs. pH c at different ionic strengths.
The present work clearly demonstrates that clay minerals represent strong retardation barriers for tri- and
hexavalent metal ions even under highly saline conditions.
[1] Brewitz, W.: Eignungsprüfung der Schachtanlage Konrad für die Endlagerung radioaktiver Abfälle. GSF-T136,
Neuherberg (1982).
[2] Vilks, P.: Sorption of Selected Radionuclides on Sedimentary Rocks in Saline Conditions – Literature Review.
Nuclear Waste Management Organization, Toronto, Ontario, Technical Report NWMO TR-2011-12 (2011).
[3] Bradbury, M.H., Baeyens, B.: Sorption of Eu on Na- and Ca-montmorillonites: Experimental investigations and
modeling with cation exchange and surface complexation, Geochim. Cosmochim. Acta, 66, 2325-2334 (2002).
[4] Rabung, Th., Pierret, M.C., Bauer, A., Geckeis, H., Bradbury, M.H., Baeyens, B.: Sorption of Eu(III)/Cm(III) on
Ca-montmorillonite and Na-illite. Part 1: Batch sorption and time-resolved laser fluorescence spectroscopy
experiments, Geochim. Cosmochim. Acta, 69, 5393-5402 (2005).
[5] Huittinen, N., Rabung, Th., Schnurr, A., Hakanen, M., Lehto, J., Geckeis, H.: New insight into Cm(III) interaction
with kaolinite – Influence of mineral dissolution, Geochim. Cosmochim. Acta, 99, 100-109 (2012).
Acknowledgements
We are grateful to M. Marques Fernandes from the Labor für Entsorgung (LES) of the Paul-Scherrer-
Institute (PSI), Switzerland for providing the purified illite de Puy and Na-SWy-2. This research has received

partially funding from the German Federal Ministry of Economics and Technology (BMWi) under contract
no. 02 E 10961.
PA5-23
SORPTION OF NP(V) ONTO NA-BENTONITE AND GRANITE: EFFECT OF EQUILIBRIUM
TIME, PH, IONIC STRENGTH AND TEMPERATURE
P. Li, Z. Liu, Z. Guo, W. Wu
School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000, China
It has been estimated that 237 Np will be a major contributor to environmental radioactivity from the disposal
of high- or intermediate-level radioactive waste [1]. Neptunium(V), which is dominant under oxidizing
conditions, shows weak interactions with mineral surfaces and is therefore regarded as a rather mobile
species [2]. Therefore, it is very important to study about the sorption/migration behavior of neptunium in
repositories of nuclear waste. The aim of the present study is to obtain the sorption and diffusion data of
long-lived neptunium(V) in the back-filled materials and natural wallrock.
The sorption of Np(V) onto Na-bentonite and granite was studied as a function of equilibrium time, pH, ionic
strength and temperature under anaerobic conditions using a batch technique. The results showed that the
Np(V) sorption onto Na-bentonite and granite is strongly depended on solution pH, sorbent dose, Np(V)
concentration, and weakly influenced by ionic strength. The kinetic process of Np(V) sorption indicated that
the sorption achieves equilibrium within about 10 h (Fig.1), thereby 3 days was selected for the experiments
to assure that the sorption equilibration was reached. From Fig. 2 we can see that the sorption ratio increases
with increasing sorbent dose. The calculated distribution coefficient (K d ) is weakly dependent on the sorbent
dose, which is consistent with the physicochemical properties of K d value, i.e., K d is independent of solid-toliquid
ratio at very low solid content. The dependence of Np(V) sorption on ionic strength at different pH is
presented in Fig.3. A large ionic strength would restrict the sorption of Np(V) onto Na-bentonite/granite at
pH 6.5 while the sorption is independent of ionic strength at pH 8.5, which indicates that the sorption maybe
of an outer-sphere complexation type at pH 6.5 and of an inner-sphere complexation type at pH 8.5 [3]. Fig.
4 shows the sorption isotherm of Np(V) on Na-bentonite/granite. At low Np(V) concentrations, Na-bentonite
exhibited almost the same sorption capacity as granite while more Np(V) would uptake onto Na-bentonite at
higher Np(V) concentrations. Furthermore, the sorption and diffusion of neptunium on/through compacted
bentonite will be conducted.
100
80
Bentonite
Granite
80
Sorption (%)
60
40
Sorption (%)
60
40
Bentonite, pH=6.5
Bentonite, pH=8.5
Granite, pH=6.5
Granite, pH=8.5
20
20
0
0 50 100 150 200 250
t (h)
0
0 10 20 30 40 50
m/V (g/L)
Fig. 1. Sorption of Np(V) onto Na-bentonite
and granite as a function of sorbent dose.
C 0 (Np(V))=4×10 -7 mol/L, I(NaCl)=0.1 mol/L.
Fig. 2. Sorption of Np(V) onto Na-bentonite
and granite as a function of sorbent dose.
C 0 (Np(V))=4×10 -7 mol/L, I(NaCl)=0.1 mol/L.

100
80
Bentonite, pH=6.5
Bentonite, pH=8.5
Granite, pH=6.5
Granite, pH=8.5
2.5x10 -7 4.0x10 -9
2.0x10 -7
2.0x10 -9
1.5x10 -7 0.0
0.0 2.0x10 -8 4.0x10 -8 6.0x10 -8 8.0x10 -8
1.0x10 -7
60
Sorption (%)
40
20
Cs (mol/g)
0
5.0x10 -8
0.0
Bentonite
Granite
0.00 0.05 0.10 0.15 0.20 0.25 0.30
I (NaCl) / mol/L
Fig. 3. Influence of ionic strength on sorption of
Np(V) onto Na-bentonite and granite.
C 0 (Np(V))=4×10 -7 mol/L, m/V=10 g/L.
-5.0x10 -8
-5.0x10 -7 0.0 5.0x10 -7 1.0x10 -6 1.5x10 -6 2.0x10 -6 2.5x10 -6
Ce (mol/L)
Fig. 4. Sorption isotherm of 237 Np(V) onto Nabentonite
and granite.
m/V=10 g/L, pH=8.5, I(NaCl)=0.1 mol/L.
[1] P. Thakurn, G.P. Mulholland (2012). “Determination of 237 Np in environmental and nuclear samples: A review of
the analytical method.” Appl Radiat Isotopes. 70: 1747–1778.
[2] K. Schmeide, G. Bernhard (2010). “Sorption of Np(V) and Np(IV) onto kaolinite: Effects of pH, ionic strength,
carbonate and humic acid.” Appl Geochem. 25: 238–1247.
[3] K. Nakata, T. Fukuda, S. Nagasaki, S. Tanaka and A. Suzuki (1999). “Sorption of neptunium on iron-containing
minerals.” Czech. J. Phys. 49/S1: 159–166.
PA5-24
COMPARISON OF URANYL ADSORPTION ON IRON(III) OXYHYDROXIDES
Keigo Niida (1) , Takumi Saito (2) and Satoru Tanaka (1)
(1) Department of Nuclear Engineering and Management, School of Engineering, The University of
Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656, Japan
(2) Nuclear Professional School, School of Engineering, The University of Tokyo, 2-22 Shirakata-shirane,
Tokai-mura, Ibaraki 319-1188, Japan
In the performance assessment of geological disposal of nuclear wastes, colloids in subsurface environments
may act as potential carriers for radionuclides, which can cause retardation or acceleration of their migration
[1]. Especially, at a transition zone from anoxic to oxic environments near surface, hydrous ferric oxide
colloids are generated by oxidation of dissolved Fe 2+ . Ferrihydrite is one of the most prominent hydrous
ferric oxides with a large specific surface area and high affinity for various metal ions. Previous studies have
revealed an important role of ferrihydrite as a transportation carrier for radionuclides [1]. Ferrihydrite is a
metastable mineral and an intermediate of other more stable iron oxyhydroxides such as goethite and
hematite [2]. The effects of this transformation on metal uptakes are still under extensive research [3].
Ferrihydrite exhibits two different crystalline phases, namely low-crystalline 2-line ferrihydrite (2L-Fh) and
high-crystalline 6-line ferrihydrite (6L-Fh). An ideal crystal structure of 2L-Fh is reported to consist of about
80% octahedrally coordinated iron and 20% tetrahedrally coordinated iron. Previous studies showed that
UO 2+
2 adsorbed on 2L-Fh via bidentate edge-sharing with an octahedrally coordinated iron site at the
relatively high concentration (~1.04 µmol∙m -2 ) by extended X-ray absorption fine structure (EXAFS)
analysis[4].The purpose of this study is to understand the adsorption behaviors of UO 2+ 2 at relatively low
concentrations on iron(III) oxyhydroxides (2L-Fh, 6L-Fh, goethite (Gt), and maghemite (Mh)) thorough
comparing UO 2+
2 adsorption isotherms, pH edges and attenuated total reflectance Fourier transformed
infrared (ATR-FTIR) spectra. The comparison will be interpreted via differences in their crystal structures;
namely, Gt has only octahedrally coordinated irons and Mh is composed of 50 % tetrahedrally coordinated
irons.
2L- and 6L-Fh were synthesized as described by Cornel and Schwertmann [2]. Gt was synthesized,
according to Saito, et al [5]. Mh nano powder was purchased from the Sigma Aldrich and used without any

pretreatments. Adsorption experiments with UO 2+
2 were performed at room temperature under Ar
atmosphere with 0.5 g L -1 iron(III) oxyhydroxides in the presence of 0.1 M NaClO 4 .The concentrations of
UO 2+ 2 were 1, 10, and 50 μmol∙L -1 . The pH values of the samples were adjusted from 3 to 10 with 0.1 mol∙L -
1 HClO 4 and 0.1 mol∙L -1 NaOH. The samples were shaken for 24 h at 25°C, followed by solid/liquid
separation by centrifugal ultrafiltration with 3000 Da cut-off membranes at 9,500 g for 60 min. The
concentrations of UO 2+ 2 in the filtrates were determined by ICP-MS after acidification. Adsorption isotherm
measurements of UO 2+ 2 on 0.5 g∙L -1 of the iron(III) oxyhydroxides were performed in the presence of 0.1
mol∙L -1 NaClO 4 at room temperature under Ar condition. The concentrations of UO 2+ 2 were varied from 0.1
to 600 μmol∙L -1 . The pH values of the samples were adjusted 4 ± 0.05. The samples were shaken for 24
hours at 25°C. After shaking, the samples were treated in the similar way to the adsorption pH-edge
measurements for the determination of UO 2+ 2 adsorption amounts as a function of free UO 2+ 2 concentration.
ATR-FTIR measurements were performed, using a flow-through ATR cell with Ge crystal and a mercurycadmium-telluride
(MCT) detector equipped in a FTIR spectrometer (FTIR-8400s, Shimadzu). The Ge
crystal surface was coated with thin layer of the iron(III) oxyhydroxides. The ATR-FTIR spectra of adsorbed
UO 2+ 2 were measured by applying UO 2+ 2 with the suspensions of iron(III) oxyhydroxide, while varying the
pH.
The pH dependence of UO 2+ 2 adsorption (i.e. pH edges) were similar among the iron(III) oxyhydroxides
investigated. The adsorption of UO 2+
2 started to increase around pH 3.0and reached to the complete
adsorption around pH 5.9. The adsorption isotherm of UO 2+ 2 on the iron(III) oxyhydroxides are shown in Fig.
1. The slopes of the isotherms in log-log plots were smaller than 1 at the low UO 2+ 2 concentration and larger
than 1 at the high UO 2+ 2 concentration for all iron(III) oxyhydroxides. The slopes for 2L-Fh, 6L-Fh, and Mh
at the high UO 2+ 2 range were close to each other (~1.75) and different from that of GT (1.00); however, those
at the low UO 2+ 2 concentration range were different among 2L/6L-Fhs (0.86 and 0.87, respectively), Mh
(0.74), and Gt (0.57). The variations of the slopes of the isotherms should reflect changes in underlying
surface complexation mechanisms, which can be tackled by molecular-scale information from ATR-FTIR
measurements.
Fig 1. Comparison of UO 2+ 2 adsorption isotherms on the iron(III) oxyhydroxides in log-log plots.
[1] A.B. Kersting, D. W. Efurd, D. L. Finnegan, D. J. Rokop, D. K. Smith and J. L. Thompson. (1999).” Migration of
plutonium in groundwater at the Nevada Test Site” Nature 397 (6714):56-59.
[2] R.M. Cornell and U. Schwertmann(2003). “The Iron Oxides”, Wiley-VCH, Weinheim).
[3] F. Marc Michel, Lars Ehm, Sytle M. Antao, Peter L. Lee, Peter J. Chupas, Gang Liu, Daniel R. Strongin, Martin A.
A. Schoonen, Brian L. Phillips, John B. Parise (2007). “The Structure of Ferrihydrite, a Nanocrystalline Material”
Science 316 (5832) :1726-1729
[4] T. D. Waite, J. A. Davis, T. E. Payne, G. A. Waychunas, N. Xu. (1994). ” Uranium(VI) adsorption to ferrihydrite:
Application of a surface complexation model” Geochimica et Cosmochimica Acta 58.(24) :5465-5478

[5] T. Saito, L.K. Koopal, W.H. van Riemsdijk, S. Nagasaki, S. Tanaka. Langmuir (2004) “Adsorption of Humic Acid
on Goethite: Isotherms, Charge Adjustments, and Potential Profiles” Langmuir 20:689-700.
PA5-25
URANYL COORDINATION CHEMISTRY ON Mg-RICH MINERALS: POLARISATION
DEPENDENT EXAFS
A. van Veelen 1 , R. Copping 2 , G. Law 3 , A.J. Smith 1 , J. R. Bargar 4 , D.K. Shuh 2 and R.A. Wogelius 1 *
1 University of Manchester, School of Earth, Atmospheric and Environmental Sciences, Oxford Road,
Manchester, M13 9PL, United Kingdom (*correspondence: roy.wogelius@manchester.ac.uk)
2 Chemical Sciences Division, Lawrence Berkeley National Laboratory, MS70A1150, One Cyclotron Road,
Berkeley, CA 94720, USA
3 University of Manchester, Centre for Radiochemistry Research, School of Chemistry, Oxford Road,
Manchester, M13 9PL, UK
4 Stanford Synchrotron Radiation Lightsource, PO Box 4349, Stanford, CA 94309, USA
Disposal of intermediate-level nuclear MAGNOX (MAGnesium Non-OXidising) waste poses major
scientific and social challenges. In the UK, poor waste management created complex radiological
remediation and clean-up challenges. Chemical understanding of uranium uptake in these Mg-rich sludges is
vital. By combining various EXAFS techniques, we determined: (1) where uranyl (UO 2 2+ ) is adsorbed onto
Mg-rich minerals, and (2) how uranyl is attached to the mineral surface. The first set of powder adsorption
experiments of U(VI) were performed with magnesite [MgCO 3 ], brucite [Mg(OH) 2 ], nesquehonite
[MgCO 3·3H 2 O] and hydromagnesite-[Mg 5 (CO 3 ) 4 (OH) 2·4H 2 O]. The second set of experiments consisted of
single crystals of magnesite (10.4) and brucite (0001). The powders were reacted in solution pH ~8.5 with
uranyl nitrate for 48 hrs under ambient PCO 2 = -3.5. The set of single crystals were reacted under ambient
and reduced PCO 2 ~ -4.5 for 48 hrs with concentrations of uranyl chloride above (500; 50 ppm) and below (5
ppm) solubility boundaries of schoepite [UO 2 (OH) 2·H 2 O]. The EXAFS measurements of the single crystals
were made at χ = 0˚ and χ = 90˚ where the polarisation of the synchrotron beam was used to determine the
adsorbate and coprecipitate structures. K d values for the Mg carbonate powders were comparable to or
exceeded those published for Ca carbonates. Polarisation dependent EXAFS clearly showed polarisation for
both ambient and reduced PCO 2 . Spectra demonstrated consistently that the uranyl molecule is preferentially
oriented with the axial oxygens perpendicular to the mineral surface. The EXAFS structural model
corroborates uranyl-triscarbonate (Table 1). This implies the creation of local rutherfordine-like
[UO 2 (CO 3 ) 3 )] regions which may polymerise at high uranyl activities into a thin film. These results give us
information about how uranium behaves during various processing scenarios of spent MAGNOX-fuel and
transport through Mg-bearing backfill materials.
Table 1. Shell-by-shell EXAFS fit results
Sample name U-Oax U-Oeq(1) U-Oeq(2) U-C
Theta N R 2σ 2 N R 2σ 2 N R 2σ 2 N R 2σ 2
(θ) (Å) (Å) (Å) (Å)
MgO + MgCl + 500
ppm U(VI)
0 2.35 1.747 0.006 3.83 2.238 0.005 1.14 2.467 0.001 2.65 2.945 0.004
90 2.38 1.804 0.001 3.21 2.195 0.004 3.08 2.42 0.008 3.62 2.851 0.003
MgO + 500 ppm U(VI) 0* 2.11 1.72 0.001 2.96 2.327 0.005 2.66 2.51 0.005
90 2.26 1.793 0.005 5.01 2.317 3.72 2.92 0.001
MgCO 3 + 500 ppm
U(VI)
90 3.14 1.798 0.001 4.44 2.284 0.009 2.03 2.557 0.014 3.11 2.93 0.003
MgO + 50 ppm U(VI) 0 2.12 1.78 0.002 6.66 2.33 0.015 2.51 2.843 0.003
90 2.53 1.83 0.002 5.21 2.44 0.012 2.04 2.89 0.002
MgO + MgCl + 50 ppm
U(VI)
90* 2.95 1.94 0.001 6.55 2.56 0.028 2.1 3.01 0.009
MgO + 5 ppm U(VI) 0 2.02 1.87 0.005 5.74 2.5 0.01 3.7 2.98 0.001
90 2.26 1.82 0.002 6.05 2.41 0.018 2.06 2.86 0.004

* Data quality not optimal
PA5-26
EFFECT OF AGING ON THE REVERSIBILITY OF PU(IV) SORPTION TO GOETHITE
J. C. Wong 1 , M. Zavarin 2 , J. D. C. Begg 2 , A. B. Kersting 2 , B. A. Powell 1
1 Department of Environmental Engineering and Earth Sciences, Clemson University, 342 Computer Court,
Anderson, South Carolina 29625 USA
2 Glenn T. Seaborg Institute, Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore,
California 94551 USA
In order to develop a predictive transport model for plutonium (Pu), surface stability of Pu on
goethite was studied with respect to aging by using batch sorption experiments. Distribution ratios and
sorption curves indicate that surface stability increases with aging and does not reach equilibrium by 116
days. Sorption curves were modeled in FITEQL to determine logK values.
Sorption of Pu(IV) to iron oxides has been observed to be rapid, and possibly irreversible or
hysteretic. One explanation for this behavior is that Pu becomes more strongly bound to mineral surfaces
over time. This could be explained by formation of a stronger surface complex by dehydration, multi-dentate
bond formation, or a two-site surface site model which considers a higher density of weak sorption sites and
a lower density of strong sorption sites. These processes are expected to result in greater overall Pu surface
stability with aging time. Aging can be defined as a surface chemical process happening after initial sorption
which causes a change in contaminant surface speciation over time [1].
Batch experiments were run in goethite suspensions (0.1 g/L goethite) in 10 mM NaCl which
spanned the pH range 4 to 7. In the adsorption step, Pu(IV) was reacted with goethite in ligand-free batch
samples for various lengths of time (1, 6, 15, 34 and 116 days). In the desorption step, supernatant was
quantitatively replaced by a 1.7 µM desferrioxamine B (DFOB) solution and equilibrated for 34 days.
Aqueous Pu was measured following both the adsorption and desorption steps by liquid scintillation
counting.
DFOB was used as a complexant capable of competing with the proposed strong surface complexes.
DFOB also has the advantage of stabilizing Pu(IV) as the dominant aqueous oxidation state by forming
strong Pu(IV)-DFOB complexes; the complexes Pu(IV)(DFOB) 2+ and Pu(IV)H 2 (DFOB) 2(aq) have logK
values of 34.0 and 62.3, respectively, at 0.10 M ionic strength [2]. With DFOB in excess, virtually all
aqueous Pu is expected to be complexed with DFOB and in the tetravalent state. Therefore, oxidative
leaching of Pu(IV) from goethite as Pu(V) should be minimal. Additionally, DFOB has very weak
interactions with goethite surface sites. In our experiments, dissolution of goethite in DFOB solutions
resulted in iron concentrations near to ICP-MS detection limits, and DFOB sorption to goethite is expected to
be less than 2 µmol/g [3,4]. Consequently, ternary surface complexes are not expected to form with DFOB,
Pu and goethite. When Pu sorption to goethite is studied in a DFOB solution, virtually all the Pu is expected
to be tetravalent, either as binary ≡FeOH-Pu(IV) surface complexes or as aqueous complexes with DFOB.
Distribution ratios increased with pH and aging. This effect was more pronounced in low pH
systems. Distribution ratios on the order of 10 4 L/kg were observed for samples above pH 6 aged >1 day. A
comparison of sorption curves (Fig. 1) indicates that less Pu desorption occurred with increased aging,
suggesting that the stability of Pu surface complexes on goethite increases with aging and does not reach
equilibrium by 116 days.
In order to compare the stability of surface complexes of different ages, a single double-layer surface
complexation reaction was chosen to model the sorption curves in FITEQL. Equilibrium speciation modeling
yields logK values which increase from 0.078 to 0.953 over the course of 116 days.
≡FeOH + Pu 4+ + 3 H 2 O ↔ ≡FeOPu(OH) 3 + 4 H +
245

Good model fits are achieved for data aged >15 days. For the 1 and 6 day data, the pH of sorption edges
predicted by the model is too low, indicating the surface complexes did not reached equilibrium. The
increasing distribution ratios, sorption curves, and modeled logK values with adsorption time demonstrate
that aging can play an important role in the reversibility of Pu(IV) sorption to goethite on a time scale of
days to months.
Sorbed Pu Concentration (mol/g)
8.0E-10
7.0E-10
6.0E-10
5.0E-10
4.0E-10
3.0E-10
2.0E-10
1.0E-10
0.0E+00
4.0 4.5 5.0 5.5 6.0 6.5 7.0
-1.0E-10
pH
Modeled
Data
Not Modeled
Model
Sorption
Time
Desorption
Time
1 day 34 days
6 days 34 days
15 days 34 days
34 days 34 days
116 days 34 days
Figure 1. Sorbed Pu concentrations resulting from various aging times (indicated in the legend) and 34 days
desorption in 1.7 µM DFOB solution. Open symbols represent data points which were not included in
equilibrium speciation modeling.
1. Tinnacher, R. M., Zavarin, M., Powell, B. A., Kersting, A. B.: Geochim. Cosmochim. Acta 75, 21 (2011).
2. Boukhalfa, H., Reilly, S. D., & Neu, M. P.: Inorg. Chem. 46, 3 (2007).
3. Cheah, S.: Chem. Geol. 198, 1-2 (2003).
4. Kraemer, S. M., Cheah, S.-F., Zapf, R., Xu, J., Raymond, K. N., Sposito, G.: Geochim. Cosmochim. Acta 63, 19-
20 (1999).
246

PA5-27
INVESTIGATIONS OF THE SORPTION OF U(VI) ONTO SiO 2 IN THE PRESENCE OF
PHOSPHATE: IN SEARCH OF A TERNARY SURFACE COMPLEX
M.J. Comarmond 1 , H. Foerstendorf 2 , R. Steudtner 2 , E. Chong 1 , K. Heim 2 , K. Müller 2 , K. Gückel 2 , V.
Brendler 2 , T.E. Payne 1
1) Australian Nuclear Science and Technology Organisation, Locked Bag 2001, Kirrawee DC, NSW 2232,
Australia
2) Helmholtz-Zentrum Dresden-Rossendorf e.V., Institute of Resource Ecology, P.O. Box 510119, 01314
Dresden, Germany
The adsorption of U(VI) in the environment is known to be influenced by the presence of both inorganic and
organic ligands in the aqueous phase [1]. In the present study, the ternary system containing U(VI),
phosphate and SiO 2 is investigated using a batch sorption technique, in situ attenuated total reflection
Fourier-transform infrared (ATR FT-IR) spectroscopy and time-resolved laser-induced fluorescence
spectroscopy (TRLFS). Whilst SiO 2 is a predominant component of soils and host rock and the speciation of
uranium with phosphate is complex [2], limited studies of the U(VI)/PO 4 /SiO 2 ternary system [3] exist.
Nevertheless, such studies have significant relevance in understanding radionuclide migration in geological
systems.
Batch sorption studies of U(VI) on SiO 2 were conducted with and without phosphate in ambient atmosphere.
The effect of phosphate on U(VI) sorption at the SiO 2 surface was found to be minimal for equimolar U(VI)-
phosphate concentrations of 20 µM. In both cases a sorption edge between pH 3 and 5 with a pH 50 of ca. 3.7
was observed. However, an increase in U(VI) uptake was observed in the presence of higher concentrations
of phosphate.
The in situ IR spectroscopic sorption experiments provide an on-line monitoring of the spectral changes of a
stationary solid SiO 2 phase occurring during subsequent or simultaneous sorption of aqueous U(VI) and
phosphate ions. Additionally, the spectra of the aqueous solutions allow the identification of aqueous U(VI)
phosphate species. All spectra of the sorption complexes – and of the aqueous U(VI)/PO 4 solutions as well –
recorded at pH 4 showed bands which could be assigned to phosphate modes significantly shifted compared
to the aqueous HPO 4 2− species [4] indicating the presence of sorbed and dissolved U(VI) phosphate species,
respectively. The band representing the ν 3 (UO 2 ) mode was observed at 919 cm −1 for the aqueous U(VI)/PO 4
system irrespective of the absence or presence of a precipitate in solution. Similar IR spectra were obtained
for the ternary U(VI)/PO 4 /SiO 2 system which is possible evidence of the formation of U(VI) phosphate
species both in solution and on the SiO 2 surface. However, from the relatively small band widths observed in
the spectra of prolonged sorption, the formation of a surface precipitate is strongly suggested [5].
The TRLFS spectra at pH 4 confirm the results obtained from IR spectroscopy. The luminescence spectra of
the U(VI)/PO 4 solutions clearly indicate the formation of U(VI) phosphate species, most probably solid or
colloidal (UO 2 ) 3 (PO 4 ) 2 ∙4H 2 O, which was supported by photon correlation spectroscopy analysis and is
consistent with literature reports and speciation modelling of this experimental system using EQ3/6 and
Geochemical Workbench. From the modelling, solid precipitation is predicted to occur throughout the pH
range of interest. The luminescence spectrum for sorption samples prepared in the presence of phosphate
shows lower luminescence intensity compared to the spectrum in the absence of phosphate, which might be
due to surface precipitation.
At slightly acidic pH, (~ pH 5.5), the luminescence spectrum indicates the presence of two surface species.
This is derived from the sorption samples prepared with and without phosphate. Although similar intensities
are observed in the luminescence spectra, small variations in the band positions can be possibly assigned to
an additional U(VI) surface species.
The in situ ATR FT-IR and TRLFS studies provide valuable insight towards understanding the interactions
of the U(VI)/PO 4 /SiO 2 ternary system. The results indicate that the sorption of U(VI) on SiO 2 in the presence
247

of inorganic phosphate possibly involves surface-sorption species (particularly in early stages) and evolves
towards surface precipitation. Further investigations of these surface species are in progress.
[1] Payne, T.E., Lumpkin, G.R. and Waite, T.D. (1998). ‘Uranium VI adsorption on model minerals” in Adsorption of
Metals by Geomedia, Chapter 2, p. 75-97, Jenne, E.A., Ed.; Academic Press, San Diego, U.S.A.
[2] Sandino, A. and Bruno, J. (1992). “The solubility of (UO 2 ) 3 (PO 4 ) 2 .4H 2 O(s) and the formation of U(VI) phosphate
complexes: Their influence in uranium speciation of natural waters.” Geochimica et Cosmochimica Acta, 56,
4135-4145.
[3] Zhang, H. and Tao, Z. (2002). “Sorption of uranyl ions on silica: Effects of contact time, pH, ionic strength,
concentration and phosphate.” Journal of Radioanalytical and Nuclear Chemistry 254(1), 103-107.
[4] Tejedor-Tejedor, M.I. and Anderson, M.A. (1990). “Protonation of phosphate on the surface of goethite as studied
by CIR-FTIR and Electrophoretic Mobility.” Langmuir, 6, 602-611.
[5] Gückel, K., Rossberg, A., Brendler, V. and Foerstendorf, H. (2012). “Binary and ternary surface complexes of
U(VI) on the gibbsite/water interface studied by vibrational and EXAFS spectroscopy.” Chemical Geology, 326-
327, 27-35.
PA5-28
U(VI) SURFACE DISTRIBUTION ON ÄSPÖ DIORITE UNDER ANOXIC CONDITIONS
U. Alonso 1) , T. Missana 1) , A. Patelli 2) , D. Ceccato 3) , M. García-Gutiérrez 1) , V. Rigato 3)
1) CIEMAT, Avda. Complutense 40, 28040 Madrid, Spain
2) CIVEN, Via delle Industrie 9, 30175 Venezia-Marghera, Italy
3)
INFN-LNL Viale dell’ Università 2, 35020 Legnaro-Padova, Italy
The identification and reduction of uncertainties related to the long-term prediction of radionuclide (RN)
migration within the host rock barrier or a radioactive waste repository are fundamental in radioactive
disposal management programmes. For example, heterogeneity, up-scaling and redox effects are issues to be
addressed.
Radionuclide retention in rock is mainly evaluated obtaining average distribution coefficients (Kd), which
account for the RN distribution between the liquid and the solid phase, in homogenised powder or crushed
material. This approach applied to a highly heterogeneous material, as granite or diorite, generally implies
wide variability of experimental values and lack of awareness on relevant retentive minerals [1]. The
identification of main retentive minerals contributes to give weight to heterogeneity on Kd selection, but also
contributes to a better understanding of retention underlying mechanisms and to the application of surface
complexation models.
This study presents an experimental approach that aims to provide sound sorption parameters, through the
determination of radionuclide surface distribution coefficients (Ka) directly measured on intact rocks, at a
mineral micro-scale, by micro-Particle induced X-ray Emission (µPIXE) analyses. Uranium was selected as
relevant redox sensitive radionuclide and as a heavy element, easily detected by PIXE analyses.
Sorption experiments were carried out on unaltered diorite samples from the Äspö underground research
laboratory, which were extracted, handled, transported and stored under anoxic conditions. Notably, the
maintenance of redox conditions, from extraction to final sorption measurements, represents an additional
challenge, not previously considered but, deemed essential to reduce uncertainties associated to sorption data
obtained for redox sensitive radionuclides.
µPIXE technique allows mapping of a studied area on diorite samples and main minerals and tracers
distribution could be visualised. Typical areas analysed showed major presence of plagioclase, biotite,
titanite and quartz. Minerals identified as hornblende, apatite or zircon were as well detected. The natural
presence of U in the studied samples was undetectable under experimental conditions.
Results showed that both uranium and selenium distribution on Äspö diorite surface was heterogeneous, and
the most retentive minerals were identified.
248

By the analyses of the individual PIXE spectra obtained on selected regions, surface distribution coefficients
(Ka) can be obtained, as described in [2]. The Ka distribution values measured on specific minerals on the
rock surface were used to obtain average surface distribution coefficients accounting for the mineral
occurrence (%) of diorite samples.
The average surface distribution coefficient measured for uranium was K a = (4.4·± 0.1)·10 -4 m, with higher
values on Fe minerals (K a = 8·10 -4 m). These values are higher than those measured for U on granite under
oxic conditions.
The obtained values were compared to the bulk distribution coefficients (K d ) determined on same samples by
batch experiments, considering the surface area.
Finally, results obtained under anoxic conditions are compared to previous studies carried out under oxic
conditions.
The research leading to these results has received funding from EU Seventh Framework Programme
(FP7/2007-2011) under the grant agreements Nº 269658 (CROCK, Crystalline Rock Retention processes and
Nº 2620109 (ENSAR, European Nuclear Science and Applications Research).
[1] T. Missana et al. J. of Iberian Geology 32 (1) 55-77 (2006).
[2] U. Alonso et al., Coll. & Surf. A: Phys. and Eng. Asp. 347, 230-238 (2009).
PA5-29
IMMOBILISATION OF TECHNETIUM-99 ON BACKFILL CEMENT: SORPTION UNDER
STATIC AND SATURATED FLOW CONDITIONS
C. L. Corkhill 1)* , J. W. Bridge 2) , P. Hillel 3) , L. J. Gardner 1) , M. C. Stennett 1) , R. Tappero 4)
and N. C. Hyatt 1) .
1 The Immobilisation Science Laboratory, Department of Materials Science and Engineering,
The University of Sheffield, UK
2
Centre for Engineering Sustainability, School of Engineering, University of Liverpool, UK
3 Department of Nuclear Medicine, Hallamshire Hospital, Sheffield, UK
4 National Synchrotron Light Source, Brookhaven National Laboratory, Upton, New York, USA
Technetium-99, a β-emitting radioactive fission product of 235 U, formed in nuclear reactors, presents a major
challenge to nuclear waste disposal strategies. Its long half-life (2.1 x 10 5 years) and high solubility under
oxic conditions as the pertechnetate anion [Tc(VII)O 4 ] is particularly problematic for long-term disposal of
radioactive waste in geological repositories.
In this study, we investigate the effectiveness of the backfill cement, Nirex Reference Vault Backfill
(NRVB) and other cements commonly used in nuclear waste disposal scenarios (crushed Ordinary Portland
Cement (OPC) and OPC combined with Blast Furnace Slag (BFS) or Pulversised Fly Ash (PFA)) for their
effectiveness towards immobilisation of Tc(VII). Sorption in batch experiments was shown to be dependent
on cement type; NRVB, OPC and OPC/PFA weakly sorbed pertechnetate, while the BFS-containing OPC
cement sorbed ~50% of the injected Tc(VII). Oxidation state µ-XRF mapping, combined with µ-XANES
performed on a BFS-containing cement reacted with Tc(VII) showed that immobilisation in this cement was
due to a rapid reductive-precipitation mechanism, with Tc(IV) precipitates localised on the surface of BFS
particles. The cements that displayed poor sorption of technetium were found to contain only Tc(VII). Timelapse,
non-invasive, quantitative radiographic imaging of a 99m Tc radiotracer through the different cement
compositions was performed to investigate Tc(VII) immobilisation under dynamic conditions. A standard
medical gamma camera was used to monitor pulse-inputs of ~15MBq 99m Tc under saturated conditions and at
a constant flow of 0.33ml/min. Dynamic gamma imaging was conducted every 30s for 2 hours. Spatial
moments analysis of the resulting 99m Tc plume provided information about the relative changes in mass
distribution of the radionuclide in the various cement materials. 99m Tc advected through NRVB demonstrated
typical conservative transport behaviour, while OPC and OPC/PFA produced a slight reduction in 99m Tc
249

centre of mass transport velocity over time. BFS-containing cement was shown to be most effective at
immobilising 99m Tc under dynamic, rapid-flow conditions, with up to 50% of the injected activity retained
irreversibly by the cement, indicating that the determined sorption mechanism has a significant effect on the
transport of technetium.
PA5-30
SPECIATION OF PLUTONIUM DURING DIFFUSION IN OPALINUS CLAY
S. Amayri 1)* , U. Kaplan 1) , J. Drebert 1) , J. Rosemann 1) , D. Grolimund 2) , T. Reich 1)
1)
Institute of Nuclear Chemistry, Johannes Gutenberg-Unversität Mainz, 55099 Mainz, Germany
2) Swiss Light Source, Paul Scherrer Institut, 5232 Villigen, Switzerland
With regard to the safe disposal of heat-generating radioactive waste in deep geological formations, detailed
information on the interaction between the radiotoxic, long-lived radionuclides such as 239 Pu (t 1/2 = 24,110 a)
and the clay formations, which are considered as a possible host rock, are required. A combination of
spatially-resolved synchrotron based techniques has been used to determine the chemical speciation of the
redox-sensitive Pu sorbed on heterogeneous Opalinus Clay (OPA, Mont Terri, Switzerland). Different
sorption thin sections and diffusion samples (OPA bore cores) were contacted with 20 µM Pu(V) or Pu(VI)
under aerobic conditions at pH 7.6. The contact time of the sorption samples was at least three days and one
month for the diffusion samples. Spatially-resolved molecular-level investigations were performed at the
microXAS Beamline of the Swiss Light Source (SLS, Paul Scherrer Institut, Villigen, Switzerland). Micro-
X-ray fluorescence (μ-XRF) mappings of both sorption and diffusion samples showed a heterogeneous
distribution of Pu and other elements (e.g., Ca, Mn, Fe, Sr) contained in OPA. Regions with high
concentrations of Pu and Fe have also been found and the oxidation states were determined by micro-X-ray
absorption fine structure spectroscopy (µ-XAFS) (see Fig.1).
Fe
1
Pu
High
Absorption (a.u.)
2.0
1.5
1.0
0.5
Spot 1
Reproduction
Pu(III) 30 %
Pu(IV) 57%
Pu(V) 13 %
Low
0.0
18000 18100 18200
Energy (eV)
Figure 1: Left: µ-XRF elemental distribution maps of Pu and Fe (1×1 mm; 10 μm step size; E = 18,07 keV)
on one aerobic diffusion sample prepared by contacting OPA bore core with 20 µM 242 Pu(V) solution at pH
7.6 for about one month; right: Normalized Pu L III -edge XANES spectrum measured on spots 1 marked in
the Pu distribution map.
Generally, in all investigated sorption and diffusion samples Pu L III -edge µ-XANES spectra on Pu hot spots
confirmed that Pu(IV) is the dominating species on OPA, i.e., the highly soluble Pu(V)/Pu(VI) was retained
by OPA in the reduced and less mobile tetravalent oxidation state. In addition, for the first time a diffusion
profile of Pu in OPA was measured using μ-XRF. The speciation of Pu with μ-XANES showed that Pu(V)
was reduced progressively along its diffusion path to Pu(IV). To gain further information about the
mineralogical environment of Pu, micro-X-ray diffraction (μ-XRD) measurements were employed in areas
of interest. In all investigated thin sections, a correlation of Pu(IV) with the Fe(II)-bearing mineral siderite
and the clay mineral illite was observed by µ-XRD. The combination of these spatially-resolved methods is
powerful tool to determine the transport mechanisms of actinides in heterogeneous systems. The obtained
information is an essential part of the evaluation of the long-term safety of a repository and increases also the
confidence in clay as option for the geological disposal of radioactive waste.
250

This work was financially supported by the BMWi (contract number 02E10981).
PA5-31
INFLUENCE OF SOIL PROPERTIES AND pH CHANGES IN AMERICIUM SORPTION-
DESORPTION ON SOILS
O. Ramírez-Guinart, M. Vidal, A. Rigol
Departament de Química Analítica, Universitat de Barcelona, SPAIN
The main approach for the final disposal of High Level Radioactive Wastes (HLRW) is based on its longterm
storage in underground facilities located in geological stable sites with a multi-barrier system, the so
called Deep Geological Repository (DGR). However, as the DGR barriers may lose efficiency before the
radioisotopes present in the HLRW completely decay, it is possible that radioactive leachates may escape
from the DGR and reach the soil and water compartments in the biosphere. Therefore, it is required to
examine the interaction and mobility of radionuclides present in the HLRW to predict the impact of their
eventual incorporation in the biosphere and to assess the derived risk.
The main process governing the mobility of radionuclides in soils is their sorption to soil solid phase, and the
related reversibility of this process [1]. Most of the models used to describe the radionuclide sorption are
based on empirical solid-liquid distribution coefficient (K d ). Although relevant data have been recently
obtained for a few radionuclides in soils [2], there are still some important gaps for radionuclides which are
present in the HLRW, such as americium (Am). The main objective of this work was to obtain Am sorptiondesorption
data (K d and reversibly sorbed fraction) gathered from a set of 20 soils with contrasting values of
edaphic properties such as pH, texture, carbonate and organic matter contents.
Sorption assays were done at soil pH by means of batch experiments, which consisted in a pre-equilibration
step of 2 g of each soil with 50 ml of double deionised water for 16 h, and a subsequent equilibration step for
24 h with the same solution, but labelled with 241 Am. The K d (Am) values were determined by measuring
241 Am in the supernatant before and after equilibration. Blank assays were carried out in parallel without
labelling the contact solution to characterize the supernatants in terms of dissolved organic and inorganic
carbon as an estimation of the solubilized organic matter and carbonates that can act as ligands and may
influence Am sorption [3]. Additionally, the reversibly sorbed fraction was estimated by the application of a
single extraction test, with double deionised water, to soil residues coming from all previous sorption assays.
Desorption percentages were calculated by measuring 241 Am in the supernatants after shaking 16 h and
referring the extracted 241 Am amount to the initial 241 Am sorbed in soil residues.
On the other hand, since previous works in the literature had pointed out that Am sorption in porous media is
a pH-dependent process [4], the influence of pH changes was checked in seven soils selected from the
aforementioned collection. Batch experiments at three additional different pH values, lower and higher than
the initial soil pH, were carried out for each soil sample following the previous procedure but adding to the
initial contact solution the necessary acid or base amount to reach a given pH at equilibrium.
The results of this work showed a high retention of americium in soils. K d values for the 20 soils ranged from
1380 to 277000 L kg -1 and desorption percentages were < 3.5% in all cases. From K d values it was observed
that Am has a lower sorption capacity in soils with greater organic matter content and much higher in those
with greater carbonate content. With respect to the influence of changes in pH in Am sorption, a general
trend was observed in the soils tested, as the maximum K d value was obtained in the sorption assay carried
out at soil pH, whereas K d values progressively decreased when sorption was tested at lower pH values and
sharply decreased at higher pH values. The difference between the minimum and maximum K d values due to
changes in pH ranged from 1 to 3 orders of magnitude depending on the soil sample.
In order to elucidate the role that soil properties may play in Am sorption, univariate regressions between K d
values and the soil properties which seemed to be more relevant (pH, carbonate and organic carbon content)
were performed as well as multivariate regressions including other characterization data of the soil samples
251

(e.g., cationic exchange capacity, clay, carbonate and metallic oxides content, etc.) and supernatants (e.g.,
organic and inorganic carbon dissolved).
[1] C. Gil-García et al, Applied Radiation and Isotopes 66, 126 (2008).
[2] IAEA, TECDOC-1616. Quantification of Radionuclide Transfer in Terrestrial and Freshwater Environments for
Radiological Assessments (2009).
[3] P.N. Pathak and G.R. Choppin, Journal of Radioanalytical and Nuclear Chemistry 274, 517 (2006).
[4] A. Kitamura et al, Journal of Radioanalytical and Nuclear Chemistry 239, 449 (1999).
PA5-32
THE INFLUENCE OF DIFFERENT MINERAL SURFACE PROPERTIES AND THE PRESENCE
OF NICKEL(II) ON EUROPIUM(III) RETENTION AT VARIOUS OXIDE MINERALS
S. Virtanen 1 , J. Knuutinen 1 , N. Huittinen 1 , Th. Rabung 2 , H. Geckeis 2 and J. Lehto 1
1) Laboratory of Radiochemistry, Department of Chemistry, University of Helsinki, P.O. Box 55, 00014
University of Helsinki, Finland
2) Institut für Nukleare Entsorgung, Karlsruhe Institute of Technology, P.O. Box 3640, 76021 Karlsruhe,
Germany
Studies on the competitive sorption of radionuclides and other metal ions are usually conducted using
elements with similar chemical properties, such as the oxidation state [1-3]. A study by Bradbury and
Baeyens [4] suggests that metal ion sorption on clay minerals is non-competitive when the metal ions exhibit
significantly different chemical properties. This would imply that competition is selective. For example,
competition for surface sorption sites would occur for Eu(III) and Am(III) but not for Eu(III) and Th(IV) or
U(VI). In performance assessment calculations for nuclear waste repositories, the effect of competitive and
non-competitive sorption has to be considered correctly in order to have a reliable description of the system
and not to overestimate the retention of radionuclides due to sorption processes.
In the present study, the competitive sorption of metal cations on various oxide minerals with different
surface charge properties is investigated in batch sorption studies. Thus, we aim at obtaining information on
both the influence of the pH-dependent surface charge of oxide minerals on the sorption behaviour of metal
cations as well as the competitive (or non-competitive) sorption of trivalent actinides or their lanthanide
analogues with metals of different oxidation states.
The oxide minerals TiO 2 , ZrO 2 and α-Al 2 O 3 were chosen for the study based on their different surface and
surface charge properties. The zeta potentials of the three minerals were measured in three different ionic
strengths (MilliQ-water, 0.01M NaClO 4 and 0.1M NaClO 4 ) as a function of pH. The mineral concentration
was kept constant at 1.0 g/L. The pH IEP was found to be at 3.5-4.0 for TiO 2 , at 6.0-7.2 for ZrO 2 and at 9.0-9.5
for α-Al 2 O 3 . pH IEP values reflect the acid/base properties of surface hydroxyl groups, which are considered as
ligands for metal ion surface complexation reactions. They consequently have – amongst other inherent
properties of the metal oxides – a strong impact on the pH dependent sorption of metal ions. First batch
sorption studies are conducted using Ni(II) and Eu(III) that have different oxidation states and hydrolysis
behaviour. The hydrolysis of Eu(III) and Ni(II) starts after pH 6 and pH 7.5, respectively, in 10 -6 M
solutions. Eu(III) is a frequently used chemical analogue for the trivalent actinides and 59 Ni is an important
long-lived activation product occurring in spent nuclear fuel. At present, only individual Eu(III) and Ni(II)
sorption data on TiO 2 and α-Al 2 O 3 are available (Figure 1). As expected Eu(III) sorption is stronger than that
of Ni(II) and stronger complexation strength of both metal ions is observed for TiO 2 compared with α-Al 2 O 3 .
The competitive sorption experiments are currently on-going and results are expected briefly. These
experiments will be done with and without the competing element in buffered solutions as pH isotherms. The
competing element concentrations and solid/liquid ratios are kept constant, whereas the concentration of the
other metal ion is widely varied. All the experiments will be done in a glove box under N 2 atmosphere. The
possible influence of the NaClO 4 background electrolyte cation Na + on the sorption of Ni(II)/Eu(III) is
studied in addition. If the hypothesis of Bradbury and Baeyens [4] is correct and only the elements with
similar chemical properties compete, there will be no competition found between Eu(III) and Ni(II) in the
batch experiments. Future studies will include spectroscopic experiments with the time resolved laser
252

fluorescence spectroscopy (TRLFS), sorption modelling and batch sorption experiments also with Ca(II) and
the lanthanides La(III) and Lu(III) as the competing elements.
4
4
3
2
3
2
Ni sorption on TiO 2
Ni sorption on α-Al 2
O 3
log(K d
/A s
)
1
0
-1
log(K d
/A s
)
1
0
-1
-2
Eu sorption on TiO 2
-2
-3
Eu sorption on α-Al 2
O 3
-3
3 4 5 6 7 8 9 10 11
pH
3 4 5 6 7 8 9 10 11
pH
Figure 1. Specific surface area normalized distribution values K d /A s (L/m 2 ) for Eu and Ni (2∙10 -6 mol/l) on
TiO 2 and α-Al 2 O 3 (2 g/l) in 0.01M NaClO 4
[1] Benjamin, M. M. and Leckie J. O., (1981) Competitive adsorption of Cd, Cu, Zn, and Pb on amorphous iron
oxyhydroxides, Journal of Colloid and Interface Science, 83, 410-419
[2] Echeverría J. C., Morera, M.T., Mazkiarán, C. and Garrido J. J., (1998) Competitive sorption of heavy metals by
soils. Isotherms and factorial experiments, Environmental Pollution, 101, 275-284
[3] Vidal, M., Santos, M. J., Abrão, T., Rodríguez, J. and Rigol, A., (2009) Modeling competitive metal sorption in a
mineral soil, Geoderma, 149, 189-198
[4] Bradbury M. H. and Baeyens B., (2005) Experimental measurements and modelling of sorption competition on
montmorillonite, Geochimica et Cosmochimica Acta, 69, 4187-4197
PA5-33
STUDY OF THE Cs SORPTION IN KAOLINITE AND APPLICATION TO Cs SORPTION
MODELLING IN MIXED CLAY SYSTEMS
A. Benedicto, T. Missana, M. Garcia-Gutierrez
CIEMAT, Department of Environment, Avenida Complutense, 40, 28040 Madrid, Spain.
Caesium is a widely studied element in environmental science due to the high toxicity of its isotope 137 Cs,
which is a major radionuclide in spent nuclear fuel and hence in the global radioactive waste inventory.
Furthermore, 137 Cs has been introduced to soils and groundwater over the past decades by nuclear accidents
and from nuclear weapons testing [1].
Cs migration-retention phenomena in the environment are highly influenced by the Cs adsorption on clay
minerals which are ubiquitous in natural systems and have a large specific surface area. Clays as
montmorillonite, illite and kaolinite are typical components of argillaceous rocks. A large amount of studies
about Cs sorption on montmorillonite and illite exists in the literature: illite and montmorillonite present
higher sorption capacity respect to kaolinite [2,3], given the lower permanent charge of kaolinite, possibly
coming from the Al or Si isomorphic substitution in the crystalline network.
For this reason, Cs sorption behavior in kaolinite is not well known and still the sorption reactions need to be
determined conveniently, for their introduction in the geochemical computer codes, and to adequately model
the adsorption behavior of the radionuclide in complex clayey systems.
In this study, Cs sorption behavior on kaolinite was specifically analyzed first. Then, the results obtained in
kaolinite, jointly with previous knowledge in illite and montmorillonite, were applied to study Cs sorption in
complex mixed systems. In particular, the clay used as final lid in the Spanish low level radioactive waste
repository (El-Cabril) was analyzed. In this clay, kaolinite represents the 18% of the argillaceous fraction.
253

Cs sorption on kaolinite exchanged with Na, K and Ca was determined experimentally in a wide range of Cs
loadings (10 -9 to 10 -2 M), ionic strength (0.01-0.2M) and pH (2.6-10.5) in the corresponding electrolytes
(NaClO 4 , KCl and CaCl 2 ). In view of the ionic strength behavior, Cs sorption in Na-kaolinite could be
explained as a pure ionic exchange mechanism (Figure Left) but sorption edges show little dependence on
pH, which could indicate the presence of additional mechanisms for Cs retention. On the other hand, Cs
sorption in K-kaolinite and Ca-kaolinite was unexpectedly low (considering ionic exchange) in low ionic
strength solutions, indicating that additional processes are implicated in Cs sorption in these cases. Sorption
isotherms in K-kaolinite are substantially flat, and a slight non linearity of Cs sorption is observed in Na and
Ca-kaolinite. This suggests that two environments in the kaolinite surface for the Cs sorption, given that Cs +
is the sole aqueous specie, being analysed in the present work.
The simulation in mixed clay systems (ex. Figure Right) indicates that the combined use of single models for
illite, kaolinite and montmorillonite is valid to predict Cs sorption in complex systems. The estimated
contribution of kaolinite to the Cs sorption is very low and hence could be obviated when montmorillonite or
illite are present.
Log [Cs]ads (mol Kg -1 )
0
-2
-4
-6
Cs sorption in Kaolinite:
[Na-Kaolinite] = 10 g L -1
[Cs] tot
= 1.2e-8 M
Electrolyte: NaClO 4
Simulation I=0.01M
Experimental data I=0.01 M
Simulation I=0.2M
Experimental data I=0.2 M
-8
2 4 6 8 10 12 -12 -10 -8 -6 -4 -2
pH
Log [Cs] sol
Acknowledgments: This work has been partially financed by the Spanish Government under the project NANOBAG
(CTM2011-2797).
[1] IAEA. Estimation of global inventories of radioactive waste and other radioactive materials. IAEA, Vienna, 2008,
pp. 44.
[2] R. B. Ejeckam; B. L. Sherriff, Can. Mineral. 2005, 43, 1131-1140.
[3] T. Shahwan; H. N. Erten, J. Radioanal. Nucl. Chem. 2002, 253, (1), 115-120.
PA5-34
Characterizing uranium and thorium in soils: complementary insights from isotopic exchange and
single extractions
H. Ahmed (1) , S. Young (1) , and G. Shaw (1)
(1) University of Nottingham, School of Bioscience, Agricultural and Environmental Science, Nottingham,
United Kingdom
Assessing Th and U solubility, speciation and the amount that participates in sorption/desorption in soils is
important when investigating Th and U migration in the environment. Thirty-seven soils with a wide range
of physico-chemical characteristics were obtained from locations with a variety of land uses including an
acidic moorland, loam woodland, loam-arable, loam-biosolid (sewage-sludge amended arable soils) and
calcareous soils. Pore water samples were taken from soil columns held at close to field capacity to measure
U and Th solubility and speciation; the effects of time and temperature on Th and U solubility and speciation
were also studied. Soils were extracted with ammonium acetate, EDTA, 0.43 M nitric acid and tetramethyl
254
Cs sorption in mixed system:
[Na-Illite] = 0.6 g L -1
[Na-Montmorillonite] = 0.2 g L -1
[Na-Kaolinite] = 0.2 g L -1
Electrolyte: 0.2M NaClO 4
pH = 7
Simulation in Kaolinite
simulation
in illite
Simulation
in montmorillonite

ammonium hydroxide (TMAH), with an isotope exchange method using 233 U and 230 Th. Solubility and
speciation of Th and U were modelled using the Windermere Humic Aqueous model (WHAM(VII)).
Solubility of Th and U varied within and between soils, influenced by pH, DOC, DIC and phosphate
concentrations. In acidic soils Kd values for U were 1.2 to 5.2 times greater than for Th. Solubility of Th and
U decreased with soil depth and differences between Th and U mobility were greater in the upper soil layers.
Values of Th and U Kd ranged from 707 to 550,091 L kg -1 for Th and 3214 to 640,061 L kg -1 for U; Kd was
negatively correlated with DOC. In mildly acidic and calcareous soils (pH > 5 < 8), Kd values for Th were
significantly higher than for U. The formation of soluble uranyl carbonate complexes give rise to a strong
positive correlation between U and DIC concentrations in soil solution, especially under anaerobic
conditions at high CO 2 partial pressure and also at high temperatures which encouraged microbial activity.
Values of Kd for Th were 18 to 1046 times greater than for U; Th and U Kd values diverged with soil pH,
due to enhanced U solubility as carbonate complexes and stronger Th adsorption. Including both DOC and
DIC in an empirical model significantly improved the prediction of Th and U Kd values.
The speciation of Th and U in soil pore water was investigated using size-exclusion chromatography and
ICP-MS (SEC-ICP-MS). Th solubility in all soils was dependent on fulvic acid complex formation.
Dissolved Th and Th-fulvic acid complexes were highly correlated. By contrast, U complexation with fulvic
acid was more variable, especially with increasing soil pH.
Isotopically exchangeable 238 U (E-values) in the soils studied varied from 2.7 to 14.5% of total U. A resin
purification step was used to assess non-labile colloidal 238 U (NLC) formed during E-value assay; significant
amounts of NLC- 238 U were associated with colloids < 0.2 µm in the solution phase and comprised 20% of
the labile 238 U in loam-biosolid soils. This was significantly related to the high phosphate and DOC
concentrations in filtered solutions from the biosolid soils suggesting an association with fixed forms of 238 U.
For the remaining soils NLC- 238 U comprised only 2.4% of the labile 238 U and was significantly correlated
with pH and DIC, suggesting that carbonate colloids in soils with high pH may have been responsible. The
labile 238 U was higher than the amount of U extracted by CH 3 COONH 4 and EDTA in all soils, and lower
than that extracted by HNO 3 . U extracted by HNO 3 was similar to labile 238 U below pH 5 and higher than in
loam-biosolid and loam-arable soils and was significantly correlated with extracted phosphate; nitric acid
may dissolve uranyl phosphate minerals in these soils. The labile 238 U and U extracted by the TMAH
solution were not significantly different; TMAH-extracted U was slightly higher than, or equal to, labile U in
acidic, organic woodland and loam biosolid soils, whereas in calcareous and loam-arable soils TMAH was
lower than labile U. It appears that TMAH is able to dissolve exchangeable and organically complexed, but
not carbonate-associated, U. Overall, CH 3 COONH 4 and EDTA underestimated E value by a factor of 15 and
3, while HNO 3 and TMAH overestimated E value by a factor of 2.2 and 1.1, respectively. TMAH gives the
best estimate of E value compared to other extractants, suggesting that labile U and Th are predominantly
associated with humus. However, there was no consistent match between E value and any single extractant.
Soil pH must be taken into account if reactive U(VI) concentration is to be estimated by a single extraction
method.
Including isotopically exchangeable U as an input parameter improved the prediction of U solubility in the
geochemical model WHAM(VII) compared to the total soil U concentration or the pools extractable by
CH 3 COONH 4 , EDTA, HNO 3 and TMAH. The best predictions for U solubility were for soils with pH < 7.
Measuring the isotopically exchangeable Th is still in progress, but preliminary results are promising.
PA5-35
PREDICTIONS OF NpO 2
+
IONIC EXCHANGE ON MONTMORILLONITE IN NATURAL
WATERS
A. Benedicto 2,1) , J. D. Begg 1) , P. Zhao 1) , A. B. Kersting 1) , M. Zavarin 1)
1)
Glenn T. Seaborg Institute, Physical & Life Sciences, Lawrence Livermore National Laboratory, 7000 East
Avenue, Livermore, CA 94550, USA. 2) CIEMAT, Department of Environment, Avenida Complutense, 40,
28040 Madrid, Spain.
255

Neptunium (Np) is a transuranic element of particular environmental concern. The dominant isotope 237 Np,
has a long half-life (2.13 × 10 6 years) and high biological toxicity. Under a wide range of oxic conditions the
pentavalent oxidation state, in the form of the NpO 2 + , is expected to dominate aqueous speciation [1]. NpO 2
+
is generally poorly adsorbed by a range of minerals and rocks [2], resulting in high environmental mobility.
Given its existence over long timescales and its predicted mobility, understanding Np interaction with the
natural environment is important for ultimately determining its behavior. Clay minerals are both ubiquitous
in the geosphere and currently proposed for use in nuclear repository scenarios, making understanding Np(V)
sorption to clays such as montmorillonite of particular interest. Moreover, insight into the sorption behavior
of Np(V) may provide insight into the sorption behavior of other pentavalent actinide, Pu(V), which is more
redox sensitive and experimentally more difficult to study [3].
This study focuses on Np(V) sorption by ionic exchange, on Na, K, Ca and Mg-homionized montmorillonite.
Sorption experiments isolating the ionic exchange mechanism were carried out at pH 4.5, where surface
complexation at the variably charged adsorption sites is not expected to occur and cation exchange is
expected to be the controlling Np(V) adsorption mechanism. Np(V) ionic exchange was examined in a wide
range of Np loading (5×10 -8 M - 5×10 -6 M) and electrolyte conditions not previously studied (NaCl, KCl,
CaCl 2 or MgCl 2 ; ionic strength from 0.001M to 0.1M).
The data are modeled using the published selectivity coefficients of Fletcher and Sposito (1989) [4] and
Bradbury and Baeyens [5], for exchange reactions between major cations on montmorillonite. Both
Vanselow and Gaines-Thomas conventions were analyzed. The resulting logarithms of the Np-Na selectivity
coefficients were -0.26±0.03 and -0.14±0.04, respectively, which are in the range of coefficients calculated
from limited Np sorption data available in the literature (LogK from -0.75 to 0.06) [3-5].
The estimated selectivity coefficients were applied to simulate the independent effect of each major cation
(Na, K, Ca or Mg) on Np exchange on montmorillonite without interference of other cations in the
electrolyte. Further, the selectivity coefficients were used to determine the distribution coefficient (K d ) values
for Np(V) adsorption to montmorillonite for a range of natural waters representative of different
environmental scenarios. The K d values in natural waters were simulated as an index of Np sorption by ionic
exchange mechanism, i.e., aqueous Np(V) complex formation was not considered.
Results indicate that, at equal cation concentration in the electrolyte, the divalent cations Ca 2+ and Mg 2+
highly limit Np(V) exchange on montmorillonite compared with the monovalent cations (K + and Na + ). The
highest Np ion exchange on montmorillonite (model conditions: [Np] = 5×10 -7 and [Clay] = 2 g L -1 ) is
predicted in natural waters with low ionic strength such as rainwater (distribution coefficient K d = 48.1 mL g -
1 ) or streams draining igneous rocks (K d = 31.3 mL g -1 ). Conversely, Np ion exchange on montmorillonite in
seawater is predicted to be null given the high salinity. Instead, surface complexation to montmorillonite
edge sites will control its sorption behaviour. The Np(V) ionic exchange predicted in groundwater is strongly
variable depending on different factors in the bedrock, residence time or sea water intrusion. K d values
between 21.8 and 0.8 mL g -1 were predicted in the studied groundwater cases.
Importantly, for seawater or other high pH situations, ionic exchange may not be significant given these
solution conditions. However, surface complexation to clay edges may dominate the Np(V) sorption
behavior in high pH systems [4].
[1] J. P. Kaszuba; W. H. Runde, Environ. Sci. Technol. 1999, 33, (24), 4427-4433.
[2] I. R. Triay; B. A. Robinson; R. M. Lopez; A. J. Mitchell; C. M. Overly in: Neptunium retardation with tufts and
groundwaters from Yucca Mountain, Proc 4th Annu Int Conf on High-Level Radioactive Waste Management., La
Grange Park, IL, 1993; Am. Nuclear Soc.: La Grange Park, IL, 1993; pp 1504-1508.
[3] M. Zavarin; B. A. Powell; M. Bourbin; P. Zhao; A. B. Kersting, Environ. Sci. Technol. 2012, 46, (5), 2692-2698.
[4] D. R. Turner; R. T. Pabalan; F. P. Bertetti, Clays Clay Miner. 1998, 46, (3), 256-269.
[5] M. H. Bradbury; B. Baeyens, Radiochim. Acta 2006, 94, (9-11), 619-625.
[6] P. Fletcher; G. Sposito, Clay Miner. 1989, 24, (2), 375-391.
[7] M. H. Bradbury; B. Baeyens, J. Contam. Hydrol. 1997, 27, 223-248.
PA5-36
256

EXAFS INVESTIGATION ON EU(III)-SILICA-HUMIC ACID SORPTION SYSTEM: EFFECT OF
ADDITION ORDER
S. Kumar, S. Kasar, A. S. Kar, B. S. Tomar
Radioanalytical Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085, India
Humic substances (HS) are naturally occurring polydisperse and heterogeneous mixtures of organic
molecules that exhibit high affinity for metal ions [1]. Although the effects of humic acid on the sorption of
metal ions on mineral surfaces has been widely studied [2], there remains ambiguity in the behaviour of
ternary systems, especially the molecular level descriptions of metal ion-HS and HS-mineral surfaces
interactions are yet to be fully understood [1]. The general observation regarding the role of HS in literature
is enhanced sorption of metal ions at lower pH and decreased sorption at higher pH compared to the binary
metal ions-mineral surface sorption system [3]. The magnitude of enhancement/suppression, however,
depends on solution chemical conditions, types of (i) adsorbing metal ion, (ii) surface and (iii) HS, and the
magnitude of the HS loading on the surface. In this study Eu(III) sorption on silica was studied by EXAFS
spectroscopy with different addition order of metal ion and Humic acid (HA) in order to extract the structural
information of the species involved in the particular sorption trend.
Three sorption systems studied included, binary Eu(III0-silica system and ternary system of Silica-
HA-Eu(III) and Silica-Eu(III)-HA, indicating the order of addition of HA and Eu(III) to the silica
suspension. Sorption of Eu(III) on amorphous silica particles with and without HA was studied by both batch
sorption and EXAFS spectroscopy measurements. The sorption experiment was carried out with 1 g/l silica
suspension (0.1 M NaClO 4 ) and 5 × 10 -5 M Eu(III) concentration as a function of pH. The Eu L III EXAFS
measurements were performed in the fluorescence mode at the XAFS beam line in Elettra synchrotron
facility with sorption systems prepared at pH ~ 6.0. Suspensions were centrifuged after equilibration; solid
paste was dried at 40 0 C and mixed with polyvinyl pyrollidone to make a pallet which was used for EXAFS
measurement.
Figure 1 shows that compared to binary system, percentage sorption of Eu(III) in presence of HA is
enhanced at lower pH while no difference is observed at higher pH values. Significantly, there was not much
of difference in the sorption profile with the change of addition order of metal ion and HA. At lower pH
value higher sorption of HA on the silica surface may enhance the sorption of Eu(III) in the ternary system
[2]. Fig. 2 shows the Fourier transformed and k-space EXAFS spectra of the ternary system Silica-Eu(III)-
HA. Analysis of the EXAFS spectra was restricted to the R range of 1.0-3.0 A 0 and was carried out using
IFEFFIT suite of software. Back-FT of the restricted R range spectra indicates oscillations characteristic of
the presence of a single shell. Self-absorption correction of the spectra was carried out. First shell modelling
of the spectra was carried out using the theoretical spectrum produced using Eu 2 O 3 crystallographic data.
Table 1 gives the analysis data for EXAFS spectrum. Eu-O shell R value indicates the similarity of
binding in binary and silica-Eu(II)-HA system. However, N value (considering S 2 0 =1) is higher for the
ternary system. The higher Debye Waller factor for the silica-Eu(III)-HA system indicates higher
heterogeneity in Eu-O shell. It can be inferred that apart from binding similar to the binary system, in this
ternary system, binding characteristics are also influenced by the presence of HA. In the ternary system,
silica-HA-Eu(III), R value for the Eu-O shell is higher than the binary and other ternary system (cf. Table 1).
The N and R value compares well with that reported
for Eu(aq) [4]. This concludes the formation of type
100
A of ternary surface complex, i.e., Eu binding to the
90
HA anchored on the silica surface. Analysis of
higher R-space data is in progress.
% Sorption
80
70
60
50
40
30
20
10
0
Eu-Silica
Eu-Silica-HA
HA-Silica-Eu
5 6 7 8
pH
257

Fig 1. Sorption profile
Table 2.Metric data from the analysis of the EXAFS data
R(Eu-O), Ǻ N σ 2
Silica-Eu 2.406 7.43 0.012
Silica-HA-Eu 2.422 8.62 0.011
Silica-Eu-HA 2.404 11.55 0.017
1.2
1.0
1.5
1.0
data
Fit
0.8
0.5
|χ(R)|, (A 0 ) -3
0.6
0.4
k 2 *χ(K)
0.0
-0.5
0.2
-1.0
0.0
0 2 4 6 8
Radial Distance (A 0 )
-1.5
2 4 6 8 10
k
Fig 2.
1. N. D. Bryan, L. Abrahamsen, N. Evans, P. Warwick, G. Buckau, L. Weng, W. H. Van Riemsdijk, Appl. Geochem.
27 (2012) 378 -389.
2. A.J. Fairhurst, P. Warwick, S. Richardson, Coll. Surf. A 99 (1995) 187 -199.
3. A.S. Kar et al. J. Hazardous Materials 186, (2011)1961-1965.
4. S. Kumar, A. S. Kar, B. S. Tomar, D. Bhattacharya, Polyhedron 33 (2012) 33-40.
PA5-37
STUDIES ON URANIUM(VI) SORPTION ONTO MONTMORILLONITE IN HIGHLY
CONCENTRATED BACKGROUND ELECTROLYTES
K Fritsch, K Schmeide, G Bernhard
Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology, PO Box 510119, 01314 Dresden,
Germany
Argillaceous rock is considered as one of the possible host rock types for radioactive waste repositories.
Therefore, it is necessary to examine the retention behaviour of argillaceous rock towards long-lived
radionuclides, such as uranium. In this study, the 2:1 clay montmorillonite, consisting of an octahedral
alumina sheet sandwiched by two tetrahedral silicate sheets, is used. Montmorillonite is a bentonite, a class
of clay that is favoured for sealing radioactive waste repositories in argillaceous and crystalline rock due to
its excellent swelling properties. Furthermore, montmorillonite serves as model clay for the smectites
fraction in natural argillaceous rock.
258

The clay deposits in Germany that are eligible to become radioactive waste repositories are situated in South
Germany and North Germany.[1] The South German clay deposits are comprised of the well-researched
Opalinus clay (OPA), but the area is rather densely populated. The North German clay deposits have not
been researched in comparable depth yet. Nonetheless, it is known that the pore waters in the North German
deposits have a salt concentration of up to 4 mol l −1 .[2]
The SWy-2 montmorillonite was purchased from the Clay Minerals Society source clays repository. It was
cleaned according to the procedures of Poinssot [3] and Bradbury and Baeyens [4]. Subsequent analysis
showed that its composition and its cation exchange capacity (CEC) and specific surface area (N 2 -BET) are
in excellent agreement with the literature. [5-7]
The sorption of U(VI) onto montmorillonite is studied in NaCl, KCl, CaCl 2 , and MgCl 2 under the following
conditions:
• S/L = 4.00 ± 0.01 g l −1
• background electrolytes with b cation = 1 … 3 mol kg −1
• pH = 4 … 10
• room temperature, atmospheric conditions
• c U(VI) = 1×10 −6 mol kg −1
• sorption time: 5 to 6 days (depending on the stability of the system).
Figure 1. Comparison between the relative uranium uptake in NaCl (left) and in CaCl 2 (right).
Figure 1 shows the results for the U(VI) sorption in NaCl and CaCl 2 . These two experimental series stand as
examples for the fact that the background electrolyte has a great influence on the sorption behaviour of
U(VI) on montmorillonite.
In NaCl, there is no influence of the ionic strength visible in the pH range above pH 6. In the acidic pH range
on the other hand, there is a difference in the relative uranium uptake, but only between low ionic strengths
(a NaCl concentration of 1 mol kg −1 is a low ionic strength for montmorillonite) and high ionic strengths.
The sorption maximum for all electrolyte concentrations is at pH 6 & 7, while the minimum is at pH 9 & 10.
Whereas the results obtained for KCl are quite similar to those found for NaCl, things could not be more
different in CaCl 2 – there are two sorption maxima (at pH 5 and 6 and pH 9 and 10) and the relative uranium
uptake always is above 50%. At pH 7 to 9 there is a clear correlation between ionic strength and uranium
uptake, with more uranium being removed from the solution at higher calcium concentrations. A stationary
carbonate content in the solution above pH 7 indicates that there is a precipitation of CaCO 3 in this pH range.
Further experiments are under way to study the precise nature of the influence the CaCO 3 precipitation has
on the U(VI) uptake and the cause of its correlation with the calcium content of the solution.
Sorption is influenced by the uranium speciation in solution. Hence, speciation calculations are used to
interpret the experimental data. The results of the presented sorption experiments form the basis for further
259

sorption as well as diffusion experiments with U(VI) on/in model clays and Opalinus clay under saline
conditions.
[1] “Endlagerung radioaktiver Abfälle in Deutschland“, BGR (2007).
[2] Brewitz, W, “Eignungsprüfung der Schachtanlage Konrad für die Endlagerung radioaktiver Abfälle.
Abschlussbericht.“, Gesellschaft für Strahlen- und Umweltforschung GSF-T 136 (1982).
[3] Poinssot, C et al. “Experimental studies of Cs, Sr, Ni, and Eu sorption on Na-illite and the modelling of Cs
sorption“, Nagra Technical Report 99-04 (1999).
[4] Bradbury, M et al., Geochim. Cosmochim. Acta, 73, 990-1003 (2009).
[5] Olphen, HV et al., “Data handbook for clay materials and other non-metallic minerals”, Pergamon, Oxford (1979).
[6] Mermut, A et al., Clays Clay Miner, 49, 381-386 (2001).
[7] Vogt, K et al., Clay Minerals, 13, 25-43 (1978).
PA5-38
STUDY OF EUROPIUM AND NICKEL INTERACTION WITH CALCITE - BATCH
EXPERIMENTS AND SPECTROSCOPIC CHARACTERIZATION
A. Sabău (1),(2) , N. Jordan (3) , C. Lomenech (2) , N. Marmier (2)(*) , V. Brendler (3) ,
A. Barkleit (3) , S. Surblé (4) , N. Toulhoat (5) , Y. Pipon (5) , N. Moncoffre (5) , E. Giffaut (1)
(1) Agence Nationale pour la gestion des Déchets RAdioactifs (ANDRA)
(2) University of Nice - Sophia Antipolis,
Ecosystèmes Côtiers Marins et Réponses aux Stress (ECOMERS)
(3) Helmholtz-Zentrum Dresden-Rossendorf (HZDR)-Institute of Resource Ecology (IRE)
(4) UMR 3299 CEA/CNRS SIS2M/LEEL
(5) Institut de Physique Nucléaire de Lyon CNRS/Université Claude Bernard Lyon
(6) CEA/DEN, Centre de Saclay, 91191 Gif sur Yvette, France
* Email corresponding author: nicolas.marmier@unice.fr
Interactions between cations and natural or synthetic calcite may include incorporation processes, resulting
in the irreversibility of some sorption reactions. Calcite is present in soil and sediment materials, and in
particular in the Callovo-Oxfordian clay samples from of the French underground laboratory of Bure
(France), studied in the context of an underground repository for radioactive waste. Europium has been
chosen to be investigated by TRLFS due to its fluorescent properties and because it can serve as an analogue
for trivalent actinides. Nickel is toxic as a heavy metal as well as in its radioactive form. Few experimental
studies have been made to define its interaction with soil and sediment minerals in general and only a
handful of articles report investigations of Ni interaction with calcite. To investigate these irreversible
processes, we have chosen to work on the Eu-CO 2 -NaCl-CaCO 3 and Ni-CO 2 -NaCl-CaCO 3 systems at pH 8.4,
buffered by calcite under atmospheric conditions.
Our study combines macroscopic batch experiments with spectroscopic investigations (Time Resolved Laser
Fluorescence Spectroscopy - TRLFS and Rutherford Backscattering Spectrometry - RBS) to
comprehensively characterize these systems.
First, appropriate material for sorption experiments were selected based on characterization studies.
Eventually, a calcite powder from SOLVAY (SOCAL U1-R) with a particle size of 0.2 μm for TRLFS
investigations was chosen, mainly due to its large BET specific surface area (i.e. 18.4 m 2 /g). In addition, a
calcite powder from OMYA (BL 200), with a bigger particle size (56 μm) and a lower specific surface area
(0.7 m 2 /g) was used for Rutherford Backscattering Spectrometry (RBS) measurements, due to the specific
requirements of this technique.
For both powders, Diffuse Reflectance Infra-Red Fourier Transform Spectroscopy (DRIFT), X-ray
Diffraction (XRD), Scanning Electron Microscopy (SEM) and elementary analysis confirmed the absence of
polymorphic CaCO 3 compounds (i.e. vaterite and aragonite).
260

In order to get a better understanding of incorporation of cations in the structure of calcite, we compared our
results obtained on powders with studies on millimetric calcite single crystals from Alfa Aesar, performed
under the same experimental conditions as for powders.
The sample preparation consists in open reactor experiments under atmospheric conditions (pCO 2 = 10 -3.5
atm) in 0.1 M NaCl media. The studied concentrations range from 10 -6 to 10 -3 M. The experiments were
carried out for contact times ranging from 4 hours to 6 months for europium and from 4 hours up to 3 months
for nickel. For europium, ICP-AES/ICP-MS analysis of the supernatants showed a strong retention by calcite
whatever the initial concentration, contrary to nickel where the retention is depending on the initial
concentration.
The second step of the work involved efforts to better understand the time-dependence of Eu and Ni sorption
and respective mechanisms.
For each concentration of europium investigated by TRLFS, two species are identified and their fluorescence
lifetime increases as the initial concentration decreases and time goes on, corresponding to a gradual loss of
water molecules surrounding the europium. For higher concentrations, the species identified appear to
correspond to a (co-) surface precipitate and possibly an inner-sphere surface complex with two water
molecules retained in the hydration sphere. For lower concentrations, the longer lifetimes observed for one of
the two species suggest the incorporation of europium in calcite [1,2].
Rutherford Backscattering Spectrometry (RBS) experiments have been carried out using an alpha particle
millibeam at the 4MV Van de Graaff accelerator of IPNL and also on nuclear microprobe of CEA-Saclay.
This technique is well adapted to discriminate sorption processes such as: (i) adsorption or co precipitation at
the mineral surfaces or (ii) incorporation into the mineral structure (through diffusion for instance). The
interpretation of the results shows different sorption behaviors for Ni and Eu. Ni accumulates at the calcite
surface whereas Eu is also incorporated at a greater depth. Eu seems therefore to be incorporated into two
different states in calcite: (i) heterogeneous surface accumulation, which confirms the hypothesis of the
surface precipitate, and (ii) incorporation up to depths greater than 160 nm after 1 month of sorption.
Complementary Scanning Electron Microscopy (SEM) observations of the mineral surfaces at low voltage
have also been carried out, which confirmed the heterogeneities detected by RBS measurements.
[1] Fernandes, M. M.; Schmidt, M.; Stumpf, T.; Walther, C.; Bosbach, D.; Klenze, R.; Fanghänel, Th., Journal of
Colloid and Interface Science (2008), 321(2), 323-331.
[2] Piriou, B; Fedoroff, M.; Jeanjean, J.; Bercis, L., Journal of Colloid and Interface Science (1997), 194, 440-447.
PA5-39
SELECTIVE REMOVAL OF METALS FROM AQUEOUS SOLUTIONS USING SILICA
ATTACHED LIGANDS
James Holt, Steve Christie, Nick Evans and Steve Edmondson
Department of Chemistry, Loughborough University, Loughborough, Leicestershire. LE11 3TU.
In order to facilitate the selective sequestration of important radionuclides and contaminants, a surface
attached ligand, (3-Aminopropyl)triethoxysilane (APTES) has been attached to 2 types of silica with very
different surface areas. ZEOprep 60 HYD Silica gel, 40-63 μm and fumed silica, 0.007 μm were used as the
solid support for APTES attachment. A further 14 ligands attached to silica substrates by PhosphonicS Ltd
have also been tested for their selectivity of sequestration in a solution containing cobalt, nickel, copper,
zinc, cadmium, europium and uranium at a concentration of ca. 20 ppm.
O
Si O Si NH 2
O
Figure 7 - Silica attached APTES
261

Using solid-state NMR (SSNMR), IR, TGA and elemental analysis, it has been demonstrated that the
APTES attachment was successful. Following this process we have also grafted polymers from these
surfaces with the intention of increasing the available surface area and therefore removing even more metal.
A potentially novel approach to its characterisation has been carried out by producing a step by step SSNMR
whereby the organic polymeric chain can be assigned.
Figure 8 – Demonstrative view of the object of this work where the final step shows the product to be used
for selective metal sequestration
Following successful attachment of polymer to the silica surface, our focus has moved to the attachment of
ligands to silica and the testing of their selectivity. By utilising two different silica types, we have been able
to investigate the differences in the amount of ligand we can attach but then also the ability for each of these
to sequester metal. Following successful sequestration of some transition metals, we have extended our
research by using radionuclides including 57 Co, 63 Ni, 109 Cd, 152 Eu and 238 U. Their concentrations ranged from
2.5 ppm up to 160 ppm in some cases. To make the investigation more relevant to real case scenarios, a great
excess of potentially competitive groundwater cations, such as Na + and Ca 2+ were added to further the study.
It has been shown that the metal sequestration is not significantly affected by the addition of these ions.
R d ’s for the sequestration of 63 Ni from deionised water range from 4 x 10 4 to 1.2 x 10 7 ml/g compared to 5.3
x 10 4 to 7.9 x 10 5 ml/g for potentially competitive calcium in solution and 1.2 x 10 5 ml/g to 7.3 x 10 6 ml/g for
potentially competitive sodium sequestration. Isotherms have also been produced across a pH range from
5.01 to 6.80 before addition of the material, to a final pH of 6.90 to 9.49 depending on the original
concentration and competitive ions in solution. Similar R d values or better have been recorded for other
metals including 57 Co, 109 Cd, 152 Eu and 238 U.
Figure 9 – Demonstrative view of how the APTES ligand has been attached to the ZEOprep and fumed silica
surface
After encouraging results on the capabilities of silica attached APTES when in a solution with one metal, this
and each of the silica attached ligands acquired from PhosphonicS Ltd. were tested in an aqueous solution
that contained Co, Ni, Cu, Zn, Cd, Eu and U. This solution was made to a concentration of ca. 20 ppm for
each metal and pH 4.7.
262

Uranium was found to be highly selectively removed by the APTES ligand with copper being the next
selected. These results were obtained by measuring the metal concentration before and after the modified
silica had been added to the aqueous solution using ICP-OES. A noticeable difference between the ZEOprep
and Fumed silica materials was also noticed in terms of quantity of metal removed.
Investigations with the PhosphonicS Ltd. samples have shown a range of selectivity with the molecules
attached to the silica which include a range of sulphur, nitrogen, carbonyl and phosphonate ligands.
With many nitrogen containing ligands, uranium was found to be removed from a concentration of 21.5 to
1.5 ppm over a four week period, whilst in the same solution, copper was only reduced to 8.5 ppm from an
original concentration of 19.8 ppm. The remaining metals in solution did not see a noticeable change. The
other tested ligands have been found, in this seven metal system to be selective for europium and copper
whilst none of the other metals have seen a noticeable change.
PA5-40
EFFECTS OF CEMENT SUPERPLASTICIZERS ON EU SORPTION ONTO KIVETTY GRANITE
S. Holgersson
Chalmers University of Technology, Department of Chemical and Biological Engineering, Nuclear
Chemistry, Kemivägen 4, 41296 Göteborg, Sweden
Superplasticisers (SP) are cement additives used for the conditioning of the cement for increasing its
workability in construction work. In the construction of a deep underground repository for nuclear waste in a
geological formation there are a number of applications where the use of cement cannot be avoided.
Examples of use can be floor plates, tunnel plugs and grout injections against water intrusion during the
construction work. A recent literature review have shown that there is a potential for finding adverse effects
of SP on radionuclide sorption onto host rock [1].
In this work a number of parameters that may affect Eu sorption in the presence of SP were investigated:
• Solid phase: one rock type from the Finnish Kivetty site, graphite-free, 0.045-0.2 mm sieved
fraction.
• SPs from concrete leaching: two types of concrete, one OPC type and one low-alkaline type used for
preparing the leaching solutions.
• SPs: 4 types: Sikament EVO26 (polycarboxylate type), Glenium51 (polycarboxylate type) and
Rheobuild1000 (polynaphtalenesulphonate type) and Mighty150 (polynaphtalenesulphonate type).
• SP concentrations: in experiments with SP additions, 0.1 and 1g/L were used.
• Water phase: the saline Olkiluoto water [2], adjusted to three different pH: 8.5, 9 and 10
Batch sorption experiments were accomplished by preparing batches of 0.5 g of crushed rock in 10 mL
synthetic groundwater phase (S:L=1:20). Samples of the water phase were taken after 1 day, 1 week, 1
month, 3 months and 6 months of contacting time. All experiments were made in a nitrogen-flushed glove
box. The final results at 6 months show for the reference case R d (m 3 /kg)=2.5±0.3, 1.5±0.4 and 1.4±0.8 for
pH 8.5, 9 and 10, respectively. The addition of 0.1g/L EVO26 gave R d (m 3 /kg)=0.13±0.01, 0.048±0.000 and
1.2±0.0, respectively. The addition of 1g/L EVO26 gave R d (m 3 /kg)=0.18±0.01, 0.12±0.03 and 0.076±0.002,
respectively. Thus, a clear effect of EVO26 on Eu sorption onto granite was established and the magnitude of
effect seems to be pH dependent. The addition of 0.1g/L Glenium51 gave R d (m 3 /kg)=1.1±0.1, 0.27±0.01 and
0.18±0.00, respectively. The addition of 1g/L Glenium51 gave R d (m 3 /kg)=0.14±0.01, 0.045±0.003 and
0.016±0.000, respectively. The effect of Glenium51 was larger than for EVO26 and it is also pH dependent.
The addition of 0.1g/L Rheobuild1000 gave R d (m 3 /kg)=2.2±0.8, 2.4±0.7 and 1.5±1.0, respectively. The
addition of 1g/L Rheobuild1000 gave R d (m 3 /kg)=1.8±0.2, 1.5±0.1 and 0.48±0.01, respectively. Thus, an
effect of Rheobuild1000 was seen only for the highest concentration and at pH10. Finally, The addition of
0.1g/L Mighty150 gave R d (m 3 /kg)=3.1±1.1, 0.60±0.03 and 0.87±0.13, respectively. The addition of 1g/L
Mighty150 gave R d (m 3 /kg)=1.6±0.2, 0.84±0.17 and 0.68±0.14, respectively. The effect of Mighty150 seems
263

to be the same for both concentrations: a reduction in of Eu sorption at pH9 and 10. The batch experiments
were repeated with leaching solutions of OPC and low-alkaline cement with the SPs added in the cement
curing at realistic concentrations. However, no effects on Eu sorption with these leaching solutions could be
established. Based on these results, normal usage of SPs in cement curing should not have an effect on Eu
retention of the surrounding host rock. However, the elevated concentrations clearly establish the
Rheobuild1000 (polynaphtalenesulphonate type) as the most inert SPs of the four types tested here and it
may therefore be preferable for use, especially over the polycarboxylate type of SPs, from a safety
perspective.
[1] Hakanen, M. and Ervanne, H.: The Influence of organic cement additives on radionuclide mobility - A literature
survey. Posiva Working Report 06-06. Posiva OY.(2006).
[2] Vuorinen, U. and Snellman, M.: Finnish reference waters for solubility, sorption and diffusion studies. Posiva
Working Report 98-61. Posiva OY. (1998).
PA5-41
URANIUM (VI) SORPTION ONTO ROCK SAMPLES FROM AREAS OF THE PROPOSED HLW
AND SNF REPOSITORY IN RUSSIA (NIZHNEKANSKY MASSIVE)
N.V. Kuzmenkova (1) , V.G. Petrov (2) , I.E. Vlasova (2) , V.A. Petrov (3) , V.V. Poluektov (3) , S.N. Kalmykov (2)
(1)
Vernadsky Institute of Geochemistry and Analytical Chemistry, Russian Academy of Science, Moscow,
Russia
(2) Lomonosov Moscow State University, Department of Chemistry, Moscow, Russia
(3)
Institute of Geology of Ore Deposits, Petrography, Mineralogy and Geochemistry, Russian Academy of
Sciences (IGEM RAS), Moscow, Russia
The accepted in Russia concept for high level wastes (HLW) and spent nuclear fuel (SNF) disposal is based
on their isolation into the deep underground crystalline rock formations. Three areas of the Nizhnekansky
Granite Massive, namely “Kamenny”, “Itatsky” and “Eniseysky”, are supposed as the most perspective
locations for the future HLW and SNF repository. Sorption properties of granites towards different
radionuclides are of great importance for the modeling of radionuclides migration through the granite body
of the repository and, thus, for the Safety Assessment. In this work sorption behavior of uranium (VI) onto
granite samples from aforementioned areas is studied.
Core materials from two perspective areas (“Kamenny” – drilling depth down to 700 m; “Itatsky” – drilling
depth down to 500 m) have been studied in terms of petrographic and mineralogical characterization;
definition of filtration, elastic, petro-physical and strength properties; estimation of hydrothermalmetasomatic
transformation of rocks.
Four undisturbed slices from both areas were collected to perform sorption experiments to reveal mineral
phases responsible for the U(VI) sorption. Following rock samples from different depths were chosen for this
experiment: granodiorite biotite (K-106.8) and granodiorite-tonalite biotite (K-524.3) from “Kamenny” area;
hornblende diorite from “Itatsky” area (I-408.2). Rock samples were crushed and sieved with size less than
1 mm.
Sorption experiments were performed in glove box with inert atmosphere (N 2 ) to exclude possible influence
of carbonates. Sodium perchlorate was used as a background electrolyte (0.01 M). Initial concentration of
uranium was 1·10 -7 mol/L and isotope 233 U (T 1/2 = 1.59·10 5 years) was used for liquid-scintillation counting.
The solid to liquid ratio was 1:4.
Preliminary results show significant differences between used rock samples in sorption rate and pHdependence.
Equilibration time is five days and two weeks in the case of samples K-524.3 and both K-106.8,
I-408.2, respectively. The pH-dependence of sorbed uranium fraction has typical hump-shape for samples K-
524.3 and I-408.2: increase of sorption percentage with increasing pH values to 6, plateau (90-98 % of
uranium sorbed), decrease of sorption percentage with increasing pH values from 8 due to U(VI) hydrolysis.
In contrast, the fraction of sorbed uranium (VI) was around 95-98 % in the pH range from 4 to 10 in the case
of sample K-106.8.
264

Local distribution and preferential sorption of uranium (VI) onto different minerals within the sample was
studied by radiography, SEM-EDX, SIMS, etc. These data accompanying with rock sample composition
allow the development of quantitative model for U(VI) sorption onto investigated rocks.
This work was supported by the Ministry of Education and Science of Russian Federation (contract
14.U02.21.1527).
PA5-42
SORPTION OF SELENIUM OXYANIONS ONTO HEMATITE
N. Jordan 1) , S. Domaschke 2) , H. Foerstendorf 1) , A. C. Scheinost 3) , S. Weiß 1) , K. Heim 1)
1) Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology,
P.O. Box 510119, D-01314 Dresden, Germany
2) Hochschule Zittau/Görlitzhe, University of Applied Sciences, Germany
3)
The Rossendorf Beamline at ESRF, P.O. Box 220, F-38043 Grenoble, France
The 79 Se isotope, which is a long-lived (t 1/2 ~ 3.27 × 10 5 years[1]) and radiotoxic fission product found in
spent nuclear fuels, is of high importance in the context of geological disposal facilities. Safety calculations
assessments have shown it to be one of the most contributing isotopes to the total radioactivity that could be
potentially released to the biosphere. Selenium has a quite complex speciation, with four main oxidation
states, depending on both the pH and the redox potential of the surrounding environment. The concentration,
the bioavailability, the mobility, the distribution and the oxidation state of selenium in the environment are
greatly influenced by the pH, nature of mineral sorbent as well as potential redox reactions at mineral
surfaces. Among the mechanisms which enable selenium retardation and reduces its migration, adsorption
processes onto solid surfaces (iron, alumina, titanium oxides) has been extensively investigated at room
temperature [2-4].
Our study focuses on selenium(VI) and selenium(IV) sorption onto hematite (α-Fe 2 O 3 ), which was so far
not thoroughly characterized yet. By means of EXAFS and ATR FT-IR spectroscopic studies, it was
observed that selenium(VI) forms purely monodentate inner-sphere complexes onto hematite, but the study
was only performed at pD 3.5 [5]. To our knowledge, the only spectroscopic characterization of the binary
selenium(IV)/hematite system concluded the formation of bidentate bridging inner-sphere complexes onto
single hematite crystal using x-ray standing wave (XSW), but the measurements were only performed at pH
4.0 [6]. Hematite was chosen because it is a ubiquitous iron oxide mineral present in the environment. In
addition, it is an iron phase often found in rocks and soils in the vicinity of underground repositories [7]).
At the macroscopic level, the effect of pH and ionic strength was studied by means of batch experiments.
Sorption of both oxyanions was found to decrease with increasing pH. An increase of the ionic strength
(from 0.01 M to 0.1 M) impacted the sorption of selenium(VI), while the selenium(IV) uptake was found to
be not significantly affected. Electrophoretic mobility measurements revealed that selenium(IV) sorption
shifted the isoelectric point (pH IEP ) of hematite to lower pH values, while the pH IEP was not significantly
modified upon selenium(VI) sorption. At the molecular level, in situ ATR FT-IR measurements revealed the
formation of inner-sphere complexes during selenium(IV) sorption onto hematite, while the sorption of
selenium(VI) proceeded via the formation of outer-sphere complexes. Complementary information about the
Se reactivity at the hematite surface is provided by EXAFS spectroscopy.
High level and long-lived radioactive wastes are well-known to increase the temperature at the vicinity of
the waste disposal site. Such a thermal effect raises the question how the retention of selenium is influenced
at elevated temperatures. By means of batch sorption experiments, electrophoretic mobility measurements
and in situ ATR FT-IR spectroscopic studies, information and insights about mechanisms involved at higher
temperatures (from 25 °C to 60 °C) are provided.
[1] G. Jörg, R. Buhnemann, S. Hollas, N. Kivel, K. Kossert, S. Van Winckel, C.L.V. Gostomski, Appl. Radiat.
Isotopes 68 (2010) 2339.
[2] E.J. Elzinga, Y.Z. Tang, J. McDonald, S. DeSisto, R.J. Reeder, J. Colloid Interface Sci. 340 (2009) 153.
265

[3] N. Jordan, A. Ritter, H. Foerstendorf, A.C. Scheinost, S. Weiss, K. Heim, J. Grenzer, A. Mucklich, H. Reuther,
Geochim. Cosmochim. Acta 103 (2013) 63.
[4] A.C. Scheinost, L. Charlet, Environmental Science & Technology 42 (2008) 1984.
[5] D. Peak, D.L. Sparks, Environ. Sci. Technol. 36 (2002) 1460.
[6] J.G. Catalano, Z. Zhang, P. Fenter, M.J. Bedzyk, J. Colloid Interface Sci. 297 (2006) 665.
[7] F. Claret, B.A. Sakharov, V.A. Drits, B. Velde, A. Meunier, L. Griffault, B. Lanson, Clay Clay Min. 52 (2004) 515.
PA5-43
EFFECT OF NATURAL ORGANIC LIGANDS ON PLUTONIUM SORPTION TO
MONTMORILLONITE: OBSERVATIONS ON DIFFERENCES DUE TO ORDER OF ADDITION
Boggs, M. A. 1 Zavarin, M 1 . Powell B.A 2 . Kersting A.B 1 .
1 Glenn T. Seaborg Institute, Lawrence Livermore National Laboratory, 7000 East Avenue,
Livermore, California. 94551 USA
2 Environmental Engineering and Earth Sciences, Clemson University, 342 Computer Ct.
Anderson, South Carolina. 29625 USA
The interactions of plutonium (Pu) and clay is of great importance to understanding currently contamination
sites (Hanford Site, Washington; NNSS, Nevada) and for the long-term storage of radionuclides, as clays are
expected to be used as backfill at the majority of proposed sites. Clays such as montmorillonite have
previously been shown to strongly sorb radionuclides, coupled with their ability to be transported as colloids,
Pu 1 , can lead to an increase in the environmental mobility of radionuclides 2 . Little is currently known about
how organic matter, will affect the sorption characteristics of Pu(IV) in the presence of montmorillonite.
We have studied the effect of Pu in the presence of montmorillonite with three organic ligands: humic acid,
fulvic acid, and deferrioxamine B (DFOB). Our experiments were designed to determine if the order in
which the mineral, metal, and ligand were added altered the sorption characteristics of Pu. The experiments
were carried out three different ways; 1) adding Pu and a ligand complex prior to addition to montmorillonite
(Pu+L), 2) equilibrating the ligand and montmorillonite prior to addition of Pu (L+Mont) and lastly 3)
equilibrating Pu to montmorillonite prior to the addition of a ligand (Pu+Mont). Mineral ligand and metal
concentrations were held constant at 0.1 g/L, 0.8 ppm carbon and 10 10 M respectively, while the pH ranged
from 4 to 10.
Results for systems containing humic acid and DFOB are compared to that of a ligand free system of Pu(IV)
and montmorillonite in Figure 1. A decrease in sorption, over a wide range of pH, was observed for the
complex of humic acid and Pu(IV) compared to the ligand free studies, and follow trends that have been
previously observed with other minerals. The same trend was observed in the Pu+Mont systems. Total
organic carbon analysis shows that there is no measureable sorption of humic acid onto montmorillonite over
the time scale of these studies. The results suggest a relatively straightforward competition reaction between
montmorillonite and humic acid for Pu complexation. Weaker, but similar trends were observed for fulvic
acid. It is expected that fulvic acid, behaving as a relatively weaker complexant, would not be able to
compete as well as humic acid for Pu.
In contrast to the studies performed with humic and fulvic acid, DFOB did not decrease sorption of Pu(IV),
rather an increase in sorption was observed for the Pu-L system. Using total organic carbon analysis we
demonstrated that montmorillonite is able to sorb up to1 mM DFOB/g mineral. Further investigations of the
interaction of montmorillonite and DFOB were carried out using powder XRD and TEM to characterize the
mineral pre and post sorption. No change in the over mineral structure or composition was observed, over a
large concentration range (1 – 1000 ppm carbon), using both TEM and XRD. However, both methods
indicated an increase in interlayer spacing from 10 to 18 Å as the concentration of DFOB was increased.
The current study highlights the need to both understand how organic matter influences radionuclide
interaction in simple binary inorganic interactions, as well as direct organic-inorganic interactions.
Experiments of this nature are crucial to validation long-term storage plans and to aiding in remediation
efforts at contaminated sites.
266

Figure 1. Sorption of Pu(IV) (10^10 M) on montmorillonite (0.1 g/L) for humic acid and DFOB (0.8 ppm C)
1
DFOB
1
Humic acid
Fraction Sorbed
0.5
L+Mont
ligand Free
Fraction Sorbed
0.5
L+Mont
ligand Free
0
Pu+L
4 6 pH 8 10
0
Pu+L
4 6 pH 8 10
1. Zavarin, M.; Powell, B. A.; Bourbin, M.; Zhao, P.; Kersting, A. B., Environmental Science & Technology 2012,
46, 2692-2698.
2. Kersting, A. B.; Efurd, D. W.; Finnegan, D. L.; Rokop, D. J.; Smith, D. K.; Thompson, J. L., Nature 1999, 397, 56-
59.
PA5-44
SORPTION AND SPECIATION OF MOLYBDENUM ON BOREAL FOREST SOIL SAMPLES
Mervi Söderlund and Jukka Lehto
Laboratory of Radiochemistry, Department of Chemistry, University of Helsinki, A. I. Virtasen Aukio 1,
00560 Helsinki, Finland
A KBS-3-type repository for the spent fuel from the Finnish nuclear power reactors in Olkiluoto and Loviisa
is to be built in the bedrock at the Olkiluoto site at the depth of approximately 400 m [1]. The final disposal
plan includes a safety assessment of the spent nuclear fuel, where the potential dose contributing nuclear
waste nuclides for man are specified. As a part of this assessment, the transport of these prioritized
radionuclides from the geosphere to surface environment and their fate within is modelled and evaluated [2].
93 Mo is classified as high priority (I) radionuclide in the long-term safety assessment of spent nuclear fuel [2]
due to its anionic nature and low retention on soil mineral constituents. This study was made to get insight on
the speciation of molybdenum and retention on humus samples and mineral soil samples in aerobic and
anaerobic conditions to ascertain the importance of molybdenum in the safety assessment. Also, the effect of
pH on the sorption and speciation of molybdenum was investigated.
The sorption and speciation of molybdenum was investigated on boreal forest soil samples taken from the
vicinity of Olkiluoto nuclear power plant in Eurajoki, south-western Finland. Soil samples presented humus
layer and poorly developed mineral soil layers. Parallel anaerobic samples were taken from the mineral soil
layers. Sampling extended from the soil surface to the bedrock surface. Soil samples were preserved without
any pre-treatment, e.g. drying or sieving and used as such.
The first objective of the study was to determine the sorption of molybdenum on soil samples as a function
of equilibrium time. Batch sorption test were done for humus samples and aerobic and anaerobic soil
samples. The liquid phase used in the tests was synthetic soil solution simulant, which composition is similar
to the composition of Olkiluoto soil solution [3]. Equilibrium time ranged from 1 week to 12 weeks, after
which the equilibrium concentration of Mo was measured with ICP-MS. Mass distribution coefficient, K d ,
describing the effectiveness of the retention was determined for every sample.
In the beginning of sorption tests, the initial concentration of stable molybdenum to be used was examined in
an isotherm experiment. Based on the results, two concentration levels were selected; 10 -3 M and 10 -6 M.
Previously, it has been reported that molybdenum begins to form polynuclear species such as HMo 2 O 6 + ,
HMo 2 O 7 - , H 2 Mo 7 O 24 4- and Mo 8 O 26 4- when the concentration exceeds 10 mmol Mo/l [4]. The chemisorption
267

of these species does not take place because of the octahedral coordination of Mo [4]. In a previous study by
Lusa et al. (2009) with soil samples of the same region, the highest association of molybdenum was found
with fractions of easily exchangeable ions and bound to Fe and Mn oxides, whereas the concentration of Mo
related to organic matter was smaller. Information concerning the difference in sorption behaviour between
aerobic and anaerobic conditions was not found.
In comparison with the results from Lusa et al. (2009), the retention of molybdenum on initial concentration
of 10 -3 M was higher on humus samples than on mineral soil samples. This is in good agreement with the
observation that organic matter is the most important sorbent for molybdenum in soils [5]. The average K d -
value for humus samples, aerobic soil and mineral soil were 3.45, 0.61 and 1.63 ml/g, respectively. The
retention on humus samples increased with time, whereas no effect on mineral soil samples was seen. The
results from the 10 -6 M experiment series are coming later on.
The second aim was to study the changes in the speciation of molybdenum in the course of test. The
experiment arrangement was similar to the sorption test with an exception of the molybdenum concentration
used (10 -5 M). Molybdenum was added in the molybdate form (MoO 4 2- ) and the equilibrium speciation was
investigated with HPLC. The results of the speciation analysis are on the way. Also, the speciation of Mo in
the experimental system was modelled with PhreeqC and are to be compared with the experimental results.
It has been reported that the sorption of molybdenum decreases with increasing pH mainly because of the
shift in the surface charge of soil particles towards more negative values, and that the maximum sorption
takes place when pH is below 5 [6,7]. This phenomenon was confirmed in an experiment series, where the
sorption of Mo was investigated on concentration level 10 -3 M and 10 -6 M in pH-range 4-10 for humus
samples and aerobic and anaerobic mineral soil samples. On the other hand, the pK a -values for H 2 MoO 4 and
HMoO 4 - are 3.66 and 3.81 [4], which would correspond the increase in the concentration of MoO 4 2- species
with increasing pH. This is to be confirmed in an experiment series with 10 -5 M Mo initial concentration and
pH-range 4-10. The results from the speciation analysis are compared with sorption data to get a clearer
image and predictability of the sorption behaviour and speciation of molybdenum in different circumstances.
[1] Posiva Oy, 2009. Olkiluoto Site Description 2008. Posiva 2009-01, 714 p.
[2] Hjerpe, T., Ikonen, A. T. K, and Broed, R., 2010. Biosphere Assessment Report 2009. Posiva report 2010-03, 186 p.
[3] Lusa, M., Ämmälä, K., Hakanen, M., Lehto, J., and Lahdenperä, A.-M., 2009. Chemical and geotechnical analyses
of soil samples from Olkiluoto for studies on sorption in soils. Posiva Working Report 2009-33, 151 p.
[4] Cruywagen, J. J., and De Wet, H. F., 1988. Equilibrium study of the adsorption of molybdenum (VI) on activated
carbon. Polyhedron 7, 547-556.
[6] Mikkonen, A., and Tummavuori, J., 1993b. Retention of molybdenum(VI) by three Finnish mineral soils. Acta
Agriculturae Scandinavica Section B: Soil and plant Science 43, 206-212.
[7] Goldberg, S., and Forster, H. S., 1998. Factors affecting molybdenum adsorption by soils and minerals. Soil Science
163, 109-114.
PA5-45
Sr-85 and Eu-152 SORPTION ON MX-80 BENTONITE COLLOIDS
P. Hölttä, O. Elo, S. Jortikka, S. Niemiaho, M. Lahtinen and J. Lehto
University of Helsinki, Department of Chemistry, Laboratory of Radiochemistry, P. O. Box 55, FI-00014
University of Helsinki, Finland.
Crystalline rock in Olkiluoto is being considered as a host medium for the repository of highly radioactive
spent nuclear fuel in Finland. The spent uranium fuel will be placed in the final disposal tunnels in copper
iron canisters, which will be surrounded with bentonite clay. Colloids produced from the degraded bentonite
buffer may effect on the migration of radionuclides and colloid-facilitated transport may be significant to the
long-term performance of a spent nuclear fuel repository. The potential relevance of colloids for radionuclide
transport is highly dependent on the colloid formation, the stability of colloids in different chemical
environments and their interaction with radionuclides. In this work, the radionuclide sorption was studied on
powdered bentonite and bentonite colloids as a function of ionic strength. The bentonite used was MX-80
type bentonite powder which consists mainly of montmorillonite. Sr-85 and Eu-152 sorption were
determined in NaCl and CaCl 2 solutions as well as in diluted OLSO reference groundwater which simulates
268

the current saline groundwater in Olkiluoto in oxic conditions. The ionic strengths of the solutions were
adjusted between 0.001 and 0.1 M. All of these experiments were made in ambient conditions. Sorption was
quantified by the determination of the distribution ratio (R d ) of radionuclide activity between solid and liquid
phase.
MX-80 bentonite powder was added to the solution spiked with a tracer. Aliquots were taken and the solid
phase was separated from the colloids containing liquid phase by centrifuging. The activity of the liquid was
measured and the size and zeta potential of particles in the solution was analyzed. The liquid was then
returned to the sample vial and the sample was shaken until the centrifugation and activity, particle size and
zeta potential measurements were repeated after three and after seven days. Finally, after the last
centrifugation, the remaining particles in the liquid were separated by filtering the sample with polycarbonate
filters (pore size 1.2 and 0.05 µm) and the colloids contribution to the sorption was determined.
Radionuclide sorption onto bentonite colloids was studied using colloid dispersion solution made from MX-
80 bentonite clay powder which was mixed with Milli-Q water. The suspension was shaken for one week
and the colloidal fraction was then separated by centrifugation and the concentration of the bentonite colloids
was determined by a gravimetric method after drying the suspension. 4.7 mL aliquots were taken after 2 h, 1,
2 and 7 days and solid colloid fraction was separated by ultracentrifugation (90000 rpm/60 min). The gamma
activity of Sr-85 or Eu–152 was detected using a Wizard gamma counter. From the separated liquid phase,
colloidal particle size distribution was determined applying the photon correlation spectroscopy and zeta
potential applying the dynamic electrophoretic mobility (Malvern Zetasizer Nano ZS). Colloid concentration
was determined using a standard series made from MX-80 bentonite and a derived count rate obtained PCS
measurements as well as using the Al content of montmorillonite analyzed using ICP-MS.
In the experiments with powdered bentonite, the obtained results showed that in low salinity solutions, about
85 % of the tracer was adsorbed on the solid bentonite powder, 5 % was absorbed on particles larger than 1.2
µm and 10 % was adsorbed on the colloidal fraction (50 – 1.2 µm). At the beginning, the tracer was mostly
adsorbed on bentonite powder but during one week, the activity was increasing in the liquid phase as the
colloid concentration was also increasing. Practically all of the activity in the electrolyte solution disappeared
in the filtration, which indicates that the tracers were adsorbed on the colloids in the solution. In more saline
solutions (I > 0.02 M) 100 % of the tracer was adsorbed immediately on the solid bentonite powder.
Practically all of the Eu-152 activity was in the solid phase from the beginning to the end of the experiment
regardless of the electrolyte solution. The particle size was large and the particle concentration low
indicating that no colloids were present. Using strontium and sodium chloride solution, about 70 % the
activity adsorbed immediately on the solid phase. The mean particle size was again large and no activity
disappeared from the liquid phase during filtration. In calcium chloride solution, the activity of Sr-85 was
much lower in the solid phase. In the experiments with bentonite colloids, the obtained results confirmed the
influence of ionic strength and Ca 2+ concentration on the sorption of Sr-85 and Eu-152 onto bentonite
colloids. Nearly all of Sr-85 and Eu-152 was sorbed onto bentonite colloids in 0.001 M NaCl, CaCl 2 and
OLSO solution. The distribution coefficient (R d ) decreased when the ionic strength increased. The results are
given and the importance of bentonite colloids to the migration of radionuclides is discussed.
PA5-46
Eu AND Cm SORPTION ONTO UNPURIFIED ILLITE: BATCH-TYPE EXPERIMENTS AND
TIME RESOLVED LASER FLUORESCENCE SPECTROSCOPY (TRLFS)
T. Kupcik 1) , R. Marsac 1) , M. Hedde 1) , T. Rabung 1) , T. Schäfer 1) , M. Marques Fernandes 2) ,
B. Baeyens 2) , H. Geckeis 1)
1) KIT, Institute for Nuclear Waste Disposal (INE), P.O. Box 3640, 76021 Karlsruhe, Germany
2) Paul Scherrer Institut, Laboratory for Waste Management, 5232 Villigen PSI, Switzerland
Clay mineral dominated geological formations (e.g. argillaceous rocks) are under investigations as potential
host rocks for high-level radioactive waste repositories. In recent years a vast amount of work has been
devoted to characterise the sorption properties of radionuclides onto clay minerals [1-4], but so far most of
the work has been performed on purified material. However, accessory mineral phases initially present (e.g.
269

calcite, Fe- or Ti-phases) are also known to be effective sorbents for actinides [5,6] and can influence
sorption properties especially at trace metal ion concentrations [7]. In the present study, the sorption
behaviour of trivalent lanthanides/actinides onto a natural illite is investigated by batch and time resolved
laser fluorescence spectroscopy (TRLFS) experiments.
The “as received” illite material obtained from the region of Le Puy en Velay (France) is mainly composed
of illite and illite/smectite mixed layers with small contributions of kaolinite, feldspar, quartz and calcite.
The

Fig. 1: pH sorption edges at trace Eu concentrations (in 0.1 and 0.5M NaCl/NaClO 4 ) (a) and Eu sorption
isotherms at pH 5.5 (in 0.1M NaCl/NaClO 4 ) (b) for a natural illite du Puy and two purified Na-illite du Puy
batches with different size fractions (

idging” ternary surface complexes that are adsorbed on mordenite surface, while reduction is attributed to
the formation of soluble Ni(II)-HA/FA complexes in solution that compete with uptake processes.
EXAFS analysis from the second shell suggested that three phenomena occurred at the diatomite/water
interface: (1) outer-sphere and/or inner-sphere complexation; (2) dissolution of Si which is the rate limiting
step during Ni uptake; and (3) extensive growth of surface (co)precipitates. Under acidic conditions, outersphere
complexation is the main mechanism controlling Ni uptake. Surface loading increases with
temperature increasing, and surface coprecipitates become the dominant mechanism at elevated temperature.
1. Shitong Yang, Guodong Sheng, Xiaoli Tan, Jun Hu, Jinzhou Du, Gilles Montavon, Xiangke Wang, 2011.
Determination of Ni(II) Sorption Mechanisms on Mordenite Surfaces: A Combined Macroscopic and Microscopic
Approach. Geochimica Cosmochimica Acta 75, 6520-6534.
2. Guodong Sheng, Shitong Yang, Jiang Sheng, Jun Hu, Xiaoli Tan, Xiangke Wang, 2011. Macroscopic and
Microscopic Investigation of Ni(II) Sequestration on Diatomite by Batch, XPS and EXAFS Techniques.
Environmental Science & Technology. 45, 7718-7726
PA5-48
SURFACE MODIFIED MINERALS FOR RADIONUCLIDE SEQUESTRATION
H. Gillings 1 *, S.E. Dann 1 , D. Read 1
[1] Department of Chemistry, Loughborough University, Loughborough, Leics. LE11 3TU
With the growing requirement for alternatives to fossil fuels, the need for development of the UK’s nuclear
capacity is apparent. There is a substantial legacy of waste generated by the nuclear industry over the last 60
years, and new developments within the industry will continue to add to this inventory. Therefore, the safe
processing and storage of these wastes is an area of great importance in gaining acceptance for new build.
Ion exchange is a key procedure used in removing radionuclides from solution. The Sellafield Ion Exchange
Effluent Plant (SIXEP) is an example, allowing the removal of radionuclides, principally caesium and
strontium, from waste effluent streams using clinoptilolite. However, ion exchange is not suitable for
actinide ions, such as those of uranium and plutonium, and consequently, there is a need for additional
selective materials that can be easily implemented into existing treatment systems; this work considers the
use of modified zeolites and clays for this purpose.
Natural and synthetic zeolites and clay minerals have been reacted with organic ligands to modify their
surfaces for the removal of higher oxidation state actinide ions, such as those of uranium. (3-
aminopropyl)trimethoxysilane, N-[3-(trimethoxysilyl)propyl] ethylenediamine and N-[(3-
trimethoxysilyl)propyl]ethylenediamine triacetic acid (trisodium salt)) have been reacted with two
clinoptilolite samples and the zeolite ZSM-5. The influence of a polar versus non-polar solvent has been
investigated on one of the clinoptilolite samples. The reacted materials have been characterised using powder
XRD, FTIR, CHN and SSNMR to determine uptake of the ligand. All materials show an increase of carbon
concentration, though the extent of uptake differs. SSNMR of the clinoptilolite shows differences between
the reacted and unreacted samples with regards to the silicon and carbon spectra.
The capacity and kinetics of unreacted ZSM-5 and the two clinoptilolite samples for uptake of strontium and
caesium have been studied using static batch sorption methods. The clinoptilolite samples show the highest
uptake in the case of both strontium and caesium with capacities of 2.74x10 -3 g g -1 and 1.7 x10 -3 g g -1
respectively. The kinetic studies suggest that caesium uptake occurs noticeably faster than for strontium in
all cases. The capacity of unreacted ZSM-5 and the two clinoptilolite samples has also been studied for
uranium and thorium; preliminary results are described.
272

PA5-49
Radionuclide Interactions with the Fe 1-x O Surface under GDF Conditions
O. Preedy 1 , A. Van Veelen 2 , J. Qi 3 , M.P. Ryan 3 , R.A. Wogelius 2 , K. Morris 2 , G.T.W. Law 2 , N.A Burton 2 , F.
Mosselmans 4 and N.D.M. Evans 1
[1] Loughborough University, [2] University of Manchester [3] Imperial College London [4] Diamond light
source
Over the last 60 years the UK has produced an extensive and complex legacy of radioactive wastes. In light
of the future development of the UK nuclear industry there is a need for this hazard to be addressed. In 2008
the concept of deep geological disposal of this waste legacy was adopted by the government as official
policy, and the Nuclear decommissioning authority (NDA) was set up to action this. It is necessary to
understand the complex processes associated with deep geological disposal in order to develop a robust
safety case.
This project is part of the AMASS
consortium an EPSRC NDA
funded consortium “Atomic and
macro-scale studies of surface
processes. which aims to address
fundamental questions about the
structure, chemistry, and
morphology of key
material/groundwater interfaces
involved in deep geological
disposal, by working on the
premise that mineral surface
characteristics control several
critical processes involved in the
retardation of radionuclide release
such as sorption, complexation and
incorporation, depicted in the
figure above.
Iron oxides, resulting from the corrosion of the extensive use of steel used throughout the facility, are one of
the key surfaces of interest. Over the lengthy timescales associated with geological disposal the facility will
become anoxic and reducing, and iron(II) species will be stabilised; hence this study focuses on wϋstite (Fe 1-
xO) as a starting material.
Batch experiments using pure phase powders have been employed to investigate how the wϋstite surface
alters with exposure to conditions relevant to a GDF, as well as the effect that pH evolution has on sorption
of the Uranyl (UO 2 [2+] ) and pertechnetate (TcO 4 [-] ) ions. The surface co-ordination and speciation was
determined by ex-situ EXAFS measurements. All experiments showed that the bulk material remained
largely unaltered over the time frame of these experiments with evidence that the oxidation kinetics are
relatively slow. Surface morphology however is significantly altered with a large degree of roughening
observed. In the case of both the uranyl and pertechnetate sorption experiments it was determined that both
ions show affinity for the surface, with the XANES data showing evidence of reduction of the Uranium and
Technetium ions.
273

PA7 EXPERIMENTAL METHODS
PA7-1
PA7-2
PA7-3
PA7-4
PA7-5
PA7-6
PA7-7
PA7-8
PA7-9
PA7-10
PA7-11
PA7-12
PA7-12
DEVELOPMENT OF DIFFUSIVE GRADIENTS IN THIN-FILMS TECHNIQUE FOR THE
DETERMINATION OF TECHNETIUM-99 IN FRESHWATERS
J.J. Surman, J.M. Pates, H. Zhang (UK)
THE ENHANCEMENT OF ANALYTICAL METHOD FOR IODINE-129 DETERMINATION
IN LOW-LEVEL RADIOACTIVE WASTE
Yi-Kong Hsieh, TsingHai Wang, Li-Wei Jian, Wei-Han Chen, Tsuey-Lin Tsai, Chu-Fang Wang
(Taiwan)
SELECTIVE CHEMILUMINESCENCE SPECROSCOPY OF ACTINIDES AND
LANTHANIDES IN SOLUTIONS
I.N. Izosimov, N.G. Firsin, N.G. Gorshkov, V.A. Mikhalev, S.N. Nekhoroshkov (Russia)
SPECTROPHOTOMETRIC DETERMINATION OF MICROAMOUNTS OF THORIUM WITH
THORIN IN THE PRESENCE OF CETYLPRIDINIUM CHLORIDE AS SURFACTANT IN
PERCHLORIC ACID
Muhammad Haleem Khan, Syed Manzoor Hussain Bukhari, Akbar Ali (Pakistan)
APPLICATION OF ACCELERATOR MASS SPECTROMETRY TO MIGRATION STUDIES OF
ACTINIDES AT A LEGACY WASTE DISPOSAL SITE
T.E. Payne, J.J. Harrison, D.P. Child, K.L. Wilsher, S. Thiruvoth, M.A.C. Hotchkis, A. Ikeda Ohno, M. P.
Johansen (Australia)
ICP-MS MEASUREMENT OF SAMPLES WITH HIGH SALINITY - SAMPLE CLEAN UP OR
TRANSIENT MEASUREMENT
C. Hein, J.M. Sander, R. Kautenburger, H.P. Beck, G. Kickelbick (Germany)
BATCH EXPERIMENTS MADE EASY - AUTOMATION OF ICP-MS BATCH EXPERIMENTS
J.M. Sander , C. Hein , R. Kautenburger , H.P. Beck , G. Kickelbick (Germany)
ELECTROCHEMICAL ASSESSMENT OF REDOX PROPERTIES OF ARGILLACEOUS
SEDIMENTS USING ORGANIC ELECTRON TRANSFER MEDIATORS
A.L. Hoving, T. Behrends, M. Sander, N. Maes, C. Bruggeman (Netherlands, Switzerland, Belgium)
THE USE OF DIAMOND LIGHT SOURCE TO STUDY RADIONUCLIDES BY X-RAY
ABSORPTION SPECTROSCOPY
F. Mosselmans, G. Law, K. Morris, S. Shaw, S. Parry (UK)
SPECIATION OF 79 SE AT ULTRATRACE LEVELS USING AN AUTOMATED
CHROMATOGRAPHIC SYSTEM COUPLED TO HIGH RESOLUTION ICPMS. APPLICATION
TO NUCLEAR SPENT FUEL CORROSION STUDIES
L. Aldave de las Heras, M. Sandow, D. Serrano-Purroy, R. Sureda Pastor, S. Van Winckel, J.P. Glatz
(EU, Spain)
THE COMMISSIONING AND VALIDATION OF LOW-FLOW SAMPLING TECHNIQUE
K. Farrow, R. Stoate, L. Vivian, L. Penrose, R. Hackett, J. Rice (UK)
DEVELOPING A ROBUST ANALYTICAL METHOD FOR RADIOANALYTICAL
SEPARATION OF AMERICIUM-241
K. Farrow, G. Murphy, T. Cartledge (UK)
ANALYSIS OF URANIUM AND PLUTONIUM USING ACTINIDE RESIN
A.R. King, D. Knight, A. Fairhead (UK)
274

PA7-1
DEVELOPMENT OF DIFFUSIVE GRADIENTS IN THIN-FILMS TECHNIQUE FOR THE
DETERMINATION OF TECHNETIUM-99 IN FRESHWATERS
Surman, J.J, Pates, J.M. and Zhang, H.
Lancaster Environment Centre, Lancaster University, Lancaster, LA1 4YQ
Phone: 01524 593866, e-mail: j.surman@lancaster.ac.uk
Technetium-99 is one of the dominant components in medium and high-level waste and is important in the
calculation of long-term collective dose [1]. Releases to the environment at nuclear sites worldwide have
resulted in the contamination of environmental waters, particularly groundwater [2]. As technetium-99 forms
the highly mobile pertechnetate anion in oxidising waters [3], it has the capacity to migrate away from the initial
source and as such requires monitoring. Measurement of technetium-99 typically involves grab sampling,
followed by subsequent laboratory separation and measurement [4]. To date the use of technetium-specific
passive samplers has not been pursued for use in terrestrial waters.
The diffusive gradients in thin-films (DGT) technique is an established passive sampling method for measuring
dissolved trace metal concentration and speciation in water, in which the analyte of interest diffuses through a
thin diffusive gel and selectively binds to a resin [5]. DGT devices are deployed in situ meaning that possible
post-sampling changes to speciation and analyte losses to container walls are avoided. The use of a highly
specific technetium binding layer would avoid the collection of large (up to 20 litre) samples and complex
separation chemistry. This has the added advantage of avoiding issues with yield tracers as each available
technetium isotope has its own drawbacks of high costs, impurity issues, and/or a short half-life. Additionally,
as DGT devices bind and accumulate the analyte of interest over the period of time they are deployed, DGT
represents a method of obtaining time-integrated data. For borehole monitoring, a further benefit is that no waste
purge water is created and the amount of sampling equipment needed is reduced.
Currently technetium-DGT has been developed for use in seawater, using TEVA resin embedded in
polyacrylamide gel for the binding layer [6]. Seawater represents a relatively constrained range of solute
concentrations, thus technetium binding characteristics in a range of environmentally realistic anion
concentrations have been investigated. The response of technetium-DGT was found to be independent of the
following range of anion concentrations: 75-300 mg l -1 HCO 3 - and 20-2700 mg l -1 SO 4 2- . As nitrate is a common
co-contaminant at nuclear legacy sites, a range of realistic contamination concentrations was investigated.
Technetium binding was found to be unaffected in solutions containing 25-150 mg l -1 NO 3 - . These results
display the potential of DGT as a simple and robust method for the measurement of technetium-99 in terrestrial
waters for field experiments by researchers and for monitoring schemes by the nuclear industry.
[1] Harvey, B., Williams, K., Lovett, M. & Ibbett, R. 1992. Journal of Radioanalytical and Nuclear Chemistry, 158, 417-
436.
[2] Hu, Q-H., Weng, J-Q. & Wang, J-S. 2010. Journal of Environmental Radioactivity, 101, 426-437.
[3] Icenhower, J. P., Qafoku, N. P., Zachara, J. M. & Martin, W. J. 2010. American Journal of Science, 310, 721-752.
[4] Shi, K., Hou, X., Roos, P., & Wu, W. 2012. Analytica Chimica Acta, 709, 1-20.
[5] Zhang, H. & Davison, W. 1995. Analytical Chemistry, 67, 3391-3400.
[6] French, M.A., Zhang, H., Pates, J.M., Bryans, S.E. & Wilson, R.C. 2005. Analytical Chemistry, 77, 135-139.
275

PA7-2
THE ENHANCEMENT OF ANALYTICAL METHOD FOR IODINE-129 DETERMINATION IN
LOW-LEVEL RADIOACTIVE WASTE
Yi-Kong Hsieh 1) , TsingHai Wang 1) , Li-Wei Jian 1) , Wei-Han Chen 1) , Tsuey-Lin Tsai 2) , Chu-Fang Wang 1)*
1) Biomedical Engineering and Environmental Sciences, National Tsing Hua University, Hsinchu, 300,
Taiwan
2)
Chemical Analysis Division, Institute of Nuclear Energy Research, Atomic Energy Council, Taoyuan,
320, Taiwan
Iodine-129 is of special interest nuclide in radioactive waste due to its long live time (15.7 ma) and its
potential radiological contaminant. Pyrohydrolysis that is operated around 1000 o C was the most common
pre-treatment method used for the determination of iodine (I) in the solid samples. However, additional
attentions have to be paid during the pyrohydrolysis to avoid any possible leaking along with the gas flow.
Alternatively, alkaline digestion seems to be a more friendly method for iodine sample pre-treatment because
of its convenience. In this study, we conducted alkaline digestion with three different formulas at 190 o C for
3 hours and compared their performances to the one conducted with pyrohydrolysis. Forty mg of solid
sample was tested with 12 mL of mixture solution with following formula 1:1:1 (v/v):
Formula A: tetramethyl ammonium hydroxide (TMAH) + H 2 O 2 and Triton X-100
Formula B: ammonium hydroxide (NH 4 OH) + H 2 O 2 and Triton X-100
Formula C: TMAH, NH 4 OH, and H 2 O 2
To examine the validity of our proposed analytical method, I-129 spiked reference cement SRM 2709 was
alkaline digested followed by ICP-MS quantification. With triplicate analyses, the recoveries of these three
alkaline digestion formulas are 91 +/- 3 % for Formula A, 76 +/- 5 % for Formula B, and 83 +/- 3 % for
Formula C. In comparison with Formula A to Formula B, it is clear that the usage of TMAH can greatly
increase the iodine-129 recovery. This is likely that TMAH can serve as both an alkaline source and an
oxidizer. On the other hand, the non-ionic surfactant Triton X-100 seems to be able to stabilize iodide as the
recovery of Formula A is higher than Formula C. We further applied our analytical method employing
Formula A to quantitatively analyze eight real low level radioactive waste samples from Lan-Yu temperate
storage site, Taiwan. Notably, determined iodine-129 concentrations in these samples based on our proposed
method were approximately one order of magnitude lower than those determined from liquid scintillation
counter. This clearly reflects the higher detection limit of ICP-MS than that of liquid scintillation counter.
This exciting result strongly suggests the current used scaling factor derived from liquid scintillation counter
could be appropriately revised. Such revision could help to reclassify those existing radioactive wastes and
therefore reasonably reduce the expense of radioactive waste management.
* corresponding author: Chu-Fang Wang, Professor, cfwang@mx.nthu.edu.tw
[1] S.K. Sahoo et al., J. Radiation Res. 50, 325, (2009)
[2] J.L. Mas, K. Tagami, S. Uchida, Anal. Chim. Acta, 509, 83, (2004)
[3] K.S. Leonard, D. McCubbin, M.B. Lovett, Sci. Total Environ. 175, 9, (1995)
[4] J.H. Chao et al., Appl. Radiation Isotopes, 51, 137, (1999)
[5] M. Mesko et al., Anal. Bioanal. Chem. 398 1125 (2010)
[6] P.P. Toribio, J.M. Campos-Martin, J.L.G. Fierro, Appl. Catalysis A: General, 294, 290, (2005)
276

PA7-3
SELECTIVE CHEMILUMINESCENCE SPECROSCOPY OF ACTINIDES AND LANTHANIDES
IN SOLUTIONS
I.N.Izosimov 1* , N.G.Firsin 2 , N.G.Gorshkov 2 , V.A.Mikhalev 2 , S.N.Nekhoroshkov 2
1 Joint Institute for Nuclear Research, 141980 Dubna, Russia,
2 Khlopin Radium Institute, 194021 St. Petersburg, Russia
* e-mail: izig@mail.ru
Of lanthanide or actinide properties analysis interest was to use advantages of luminescence procedure for
detection of lanthanides and actinides having no self-luminescence, as an example, by initiation of
luminescence of some agents through excitation of lanthanide or actinide element to be detected. An effort
was made on initiation of luminol chemiluminescence through excitation of lanthanide and actinide ions with
laser radiation [1,2]. The use of this type of chemiluminescence for detection of actinides is possible only in
the case of selective excitation of chemiluminescence [2-4].
In this work the details of luminol chemiluminescence initiation through excitation of Sm(III), U(IV) and
Pu(IV) ions with laser radiation are presented. Data on luminol chemiluminescence in solutions containing
Sm(III), U(IV) and Pu(IV) are discussed. Chemiluminescence was induced by two-quanta excitation of
lanthanide or actinide ions in the range of 4f or 5f electron transitions by the scheme two steps-one color, i.e.
in irradiation of actinide-containing solution by one laser and by the scheme two steps-two colors, when a
solution is irradiated by two lasers operating at different wavelengths. A multi-step scheme of
chemiluminescence excitation makes this procedure not only highly sensitive but also highly selective
procedure of detection of substances. Appropriate selectivity was reached in our experiments when
chemiluminescence was initiated by transitions within 4f or 5f electron shell of lanthanide or actinide ions,
which correspond to visible spectral range. Since the energy of one-quantum excitation in visible range is
insufficient for initiation of luminol chemiluminescence, we selectively excited lanthanide or actinide ion by
multi-quantum absorption of visible light. This fact allows using highly sensitive chemiluminescence
procedure for selective detection of various valence lanthanide or actinide species in solutions based on
individual features of their absorption spectra.
The experiments were performed on an installation involving a pulse nitrogen laser OBB 1010 with a pulse
length of 1 ns and a pulse power of approximately 1.4 MW per a pulse and two tunable dye lasers OBB 1011
and OBB 1012 with a pulse length of 1 ns and 800 ps respectively. The pulse power 300 kW was reached for
dye lasers. A delay time for luminescence registration was 2 µs. The spectra of chemiluminescence initiation
as a result of excitation of Sm 3+ ions with dye laser by using two steps-one color scheme (two photons
absorbed during one laser pulse) is shown in fig.1. There is no complete similarity between the spectrum of
chemiluminescence excitation and Sm 3+ ions absorption spectrum. This experimental fact connected with the
difference in the selection rule for single-quantum and multi-quantum absorption.
The spectrum of chemiluminescence excitation obtained in tuning of generation wavelength of the second
laser (two steps-two colors scheme) is similar to the absorption spectrum of uranium (Fig. 2). The presence
of absorption band of U(IV) in the range of retuning of the second laser results in appearance of a peak of
luminol chemiluminescence. This fact undeniably confirms the selective mechanism of chemiluminescence
excitation. Initiation of chemiluminescence as a result of excitation of Pu(IV) with two dye lasers was
demonstrated for a solution containing CsF, luminol, and Pu(IV). A choice of solution composition was
made based on an attempt to provide favorable conditions for observation of luminol chemiluminescence and
to avoid formation of colloidal species of hydrolyzed Pu(IV). The spectrum of chemiluminescence excitation
in both two steps-one color (Fig.3) and two steps-two colors (Fig.4) schemes correlated with absorption
spectrum of Pu(IV). In both schemes we realized selective excitation of chemiluminescence and this
selectivity is caused by the features of absorption spectra of Pu(IV) solutions.
277

Fig. 1. Two-steps one-color excitation of
chemiluminescence. Excitation spectrum (a)
of luminol chemiluminescence with dye laser
in the range of absorption bands of Sm 3+ .
Chemiluminescence is detected at the
wavelength of 460 nm. Absorption spectrum
of Sm 3+ is shown below.
Fig. 2. Two-steps two-colors excitation of
chemiluminescence in solution luminol + U(IV) + HCl.
1.Chemiluminescence intensity dependence on the
wavelength of laser radiation at the first excitation step.
The wavelength of laser radiation at the second step was
fixed at 500 nm.
2.Absorption spectrum of U(IV)+HCl solution.
Fig. 3. Spectrum of chemiluminescence
excitation by the scheme two steps-one color
in solution containing CsF, luminol, and
Pu(IV) (2). Absorption spectrum of Pu(IV)
(1).
Fig. 4. Spectrum of chemiluminescence excitation in CsF
+ luminol + Pu(IV) solution by the scheme two steps-two
colors (2). Wavelength of laser radiation at the first
excitation step was varied (2), the wavelength of laser
radiation at the second step was fixed at 490 nm.
Absorption spectrum of Pu(IV) (1).
[1]. Izosimov I.N., Gorshkov N.G., Mashirov L.G. et al., Proc. Int. Conf. Actinides 2005, Manchester, UK, 2005, p.779.
[2]. Izosimov I.N., Phys. Part. Nucl. 38, 177 (2007).
[3]. Izosimov I.N., Firsin N.G., Gorshkov N.G. et al., Preprint of the Joint Institute for Nuclear Research, E6-2012-62,
Dubna, 2012, 11 P.
[4]. Gorshkov N.G., Izosimov I.N., Mikhalev V.A. et al., Radiochemistry. 6, 525 (2012).
278

PA7-4
SPECTROPHOTOMETRIC DETERMINATION OF MICROAMOUNTS OF THORIUM WITH
THORIN IN THE PRESENCE OF CETYLPRIDINIUM CHLORIDE AS SURFACTANT IN
PERCHLORIC ACID
Muhammad Haleem Khan* 1 , Syed Manzoor Hussain Bukhari*, Akbar Ali**
*Department of Chemistry, University of Azad Jammu & Kashmir, Muzaffarabad, Pakistan
**Chemistry Division, Pakistan Institute of Nuclear Science & Technology, Nilore, Islamabad, Pakistan
Thorium is surprisingly abundant in Earth's crust, being almost as abundant as lead 1 . Among actinides,
thorium is important due to its convertibility into fissile material and as a breeder reactor fuel 2 . People will
always exposed to small amounts of thorium through air, food and water because it is found nearly
everywhere on earth. Thorium and its compounds are seriously hazardous and cause various types of
diseases including lungs, pancreas and bone cancers 3 . Thorium and rare earths often coexist in minerals due
to similar behavior and their determination is a tedious process 4 . Advanced techniques require costly
equipment and most of the analytical laboratories in the developing countries are not equipped with these
instruments. Moreover, there is no γ-emitting radiotracer of thorium and proper thorium lamp for atomic
absorption spectrophotometry 5 .
Spectrophotometric methods using suitable chromogenic reagents are reported to be more reliable
and precise, but most of these were found to have some limitations in their sensitivity and selectivity. The
improved spectrophotometric determination methods for important metals having high sensitivity and
selectivity by using surfactants provides an inexpensive alternative to modern analytical techniques 6 .
Surfactants are currently used in the spectrophotometric determination of metal ions because of their unique
properties in controlling solubility, reactivity, sensitivity, and selectivity.
Keeping in view, the importance of the thorium metal, a simple, sensitive and efficient
spectrophotometric method is developed for its rapid determination using thorin as chromogenic reagent in
the presence of cetylpridinium chloride (CPC) in perchloric acid. The reaction between thorin and thorium
was found instantaneous in 3 mol L -1 HClO 4 . Cetylpridinium chloride was used for increasing the sensitivity
and selectivity of the complex. The absorption maximum of thorium-thorin complex was observed at 544
nm, whereas addition of CPC cause shifting of absorption band towards longer wavelength(581 nm). The
absorbance was studied for 168 hours in the presence of surfactant and found stable. The method allows the
determination of thorium in the range of 1-35μg mL -1 with a molar absorptivity, 2.95 ×10 4 L mol -1 cm -1 at
25±5 0 C. Sandell’s sensitivity of the complex was calculated to be 7.8 ng cm -2 at λ max 581nm. A significant
(~ 2 fold) enhancement in the sensitivity was achieved in the presence of surfactant. Relative Standard
Deviation was reduced from 4.25 to 2.5 in the presence of surfactant. Among various counter ions studied,
Be 2+ , Fe 3+ , Zn 2+ , Mn 2+ , Zr 4+ ,U 6+ , and tartrate, do not interfere, whereas, phosphate interfere beyond the
tolerance limit. The interfering effect of Ni 2+ and sulphate has been minimized in the presence of CPC in
HClO 4 . The validity of the proposed method was tested by determining thorium in environmental water
samples and Certified Reference Materials(NBL-79- A, Granite and Lujavrite) and the results agreed very
closely with the reported values. The proposed method is easy and applicable for thorium determination in
environmental and ores samples.
The main objective of the proposed method is to enhance the colour intensity and determination
range of the metal complex with chelating dye in the presence of surfactant. This will help in determination
of low prevailing precious and environmentally toxic metals in ordinary labs which provides a base for
"green Chemistry".
1. S. Prakash “Advanced Chemistry of Rare elements” (1960) P-280.
2. “Thorium dioxide properties and nuclear applications”, by J. Bell & R. Berman, USDOE, Washington, D.C. (1984)
3. Y. Chen, Z. Li, Z. Zhu, J. Pan, Analyst 124 (1999) 1839.
4. E. A. Aleksandrova, A. K. Charykov, Radiokhimiya, 31 (1989) 96.
5. W. C. Johnson, Jr., J. Milton, H. Campbell, Anal. Chem., 37 (1965)1440.
6. M.A.H. Hafez, I.M.M. Kenawy, M.A.M. Ramadan , Anal. Lett. 27 (1994) 1383.
7. V.N. Tikhonov, Zh. Anal. Khim. 32 (1977) 1435.
279

Table 3: Determination of Thorium with Thorin in the absence as well as in the presence CPC as Surfactant
Parameter
Comparison of Data
Existing Improved
In the absence of CPC In the presence of CPC
Wave length(λ max ) (nm) 544 581
Reaction medium Perchloric acid (3.0 mol L -1 )
Stability (hour) 36 168
Temperature ( o C ) Room temperature (preferably 20 ± 5)
Beer’s law range (µg mL -1 ) 1.0 - 16.0 1.0 - 35.0
Molar extinction coefficient
(L mol -1 cm -1 )
1.94 × 10 4 2.95 × 10 4
Sandell’s sensitivity (ng cm -2 ) 11.9 7.8
Regression equation y = 0.0838 x +0.0062 y = 0.1272 x +0.0013
Correlation coefficient 0.9993 0.9994
Relative standard deviation (% )
[Th 4+ ] = 5.0 µg mL -1 , n = 5
4.25 2.52
Comparison Figure
Absorption spectra of
(A) Th 4+ -thorin complex; (B) Th 4+ -thorin -CPC complex system.
280

PA7-5
APPLICATION OF ACCELERATOR MASS SPECTROMETRY TO MIGRATION STUDIES OF
ACTINIDES AT A LEGACY WASTE DISPOSAL SITE
T.E. Payne 1)* , J.J. Harrison 1) , D.P. Child 1) , K.L. Wilsher 1) , S. Thiruvoth 1) , M.A.C. Hotchkis 1) ,
A. Ikeda-Ohno 1,2) , and M. P. Johansen 1)
1)
Australian Nuclear Science and Technology Organisation, Kirrawee, Australia
2)
School of Civil and Environmental Engineering, University of NSW, Australia
The Little Forest Burial Ground (LFBG), located on the urban fringes of Sydney, Australia, was used by the
Australian Atomic Energy Commission (AAEC) to dispose of low level radioactive waste in shallow
trenches during the 1960s. According to operational records, various radionuclides were disposed, including
plutonium and uranium. The uranium waste included the man-made isotope 233 U and uranium enriched in
235 U. The total amount of plutonium disposed was a few grams, with an unknown and possibly variable
isotopic composition (although the isotope 239 Pu was most commonly mentioned in available disposal
records). The presence of these isotopes in the waste is attributable to research into nuclear power reactors,
which was being undertaken by the AAEC at the time. This paper outlines the application of Accelerator
Mass Spectrometry (AMS) to detect and quantify low levels of actinides in groundwater, soil, vegetation,
and other environmental samples from the LFBG.
Historical records indicate that there has been mobility of radionuclides from the trenches at the site since the
cessation of disposal, particularly tritium which is detectable in groundwater some distance from the disposal
area [1]. Other radionuclides have exhibited lower mobility, but 60 Co has been reported in vegetation (acacia
species) samples taken from the site [2] and there has been some surface expression of radionuclides
including alpha-emitters [2,3]. Preliminary thermochemical modelling using a representative LFBG
groundwater composition suggests that the dominant chemical form of Pu may be soluble Pu IV species.
Uranium is likely to be present as U VI , predominantly forming a monocarbonate species (UO 2 CO 3(aq) ) with
smaller proportions of hydroxide, sulphate and silicate species. The role of organic complexation of actinides
is being assessed. Together with additional chemical speciation information, accurate measurements of the
distributions of both U and Pu isotopes will provide input for models of actinide mobility at the site.
Typically alpha-spectrometry is used to assess samples for unusual uranium isotopic composition and for the
presence of plutonium. The primary region of interest of 233 U in alpha spectra is a doublet peak (emission
energies of 4.784 MeV (13%) and 4.824 MeV (84%)). These peaks will overlap with the primary peaks of
234 U, which is ubiquitous in environmental samples (Figure 1). Similarly the 239 Pu and 240 Pu cannot be
resolved by alpha spectrometry, and are therefore typically recorded as the sum of the activity of 239 Pu and
240 Pu.
234
U
233
U
Figure 1. Alpha spectrum of U showing major peaks (from left) of 238 U (doublet), 235 U (minor), 234 U
(doublet), 233 U (partly obscured) and 232 U (doublet), the 232 U being a radiochemical tracer for alpha
spectroscopy. In this sample 233 U is clearly present, but not readily quantified from this spectrum.
281

Figure 2. Energy spectra of uranium isotopes measured in the AMS. Lower energy peaks correspond to
other ions of lower mass and charge state. ‘P’ denotes a pulser peak. Masses 233, 234, 235 and 236 were
injected sequentially with time periods of 3.5, 1.0, 0.4 and 1.0s respectively. In this case 236 U is the chemical
yield tracer.
The use of Accelerator Mass Spectrometry (AMS) enables actinide isotopes to be quantified in these types of
environmental samples [4]. In AMS, a Tandem accelerator system is used to achieve high energy ion
analysis, which enhances the sensitivity relative to that achievable with conventional mass spectrometers.
The isotopes of interest are injected sequentially and the ions counted in an ionisation detector following
acceleration and mass and energy analysis. An example of energy spectra of the U ions is shown in Figure 2.
The 240 Pu/ 239 Pu isotopic ratio in vegetation (trees of syncarpia species) from the trenched area, measured by
AMS, is in the range ~0.07 to 0.10. This is significantly different to the isotope ratio (~0.18) previously
measured for a soil sample from Melbourne, Australia, reported as part of a global fallout study [5]. The
average of six measurements taken in the 30-53 o S latitudes was 0.185 [5]. Therefore, the inventory of this
actinide in the LFBG trees appears to derive a contribution from a source other than fallout, likely the waste
disposed in the trenches. Similarly, analysis of uranium isotopes in vegetation using AMS has shown that
some samples from the trenched area have unusual amounts of 233 U, with 233 U/ 234 U activity ratios as high as
1.9 (±0.1). Measurements of other environmental samples are in progress.
The AMS measurements are particularly important for discriminating between actinides derived from
various sources. The results clearly indicate the presence of anomalous actinide isotope ratios in some
environmental samples from the vicinity of the LFBG trenches, confirming a local source of uranium
(containing a significant component of 233 U) and exhibiting a plutonium isotopic signature which cannot be
explained by fallout.
[1] Hughes, C.E., Cendon, D.I., Harrison, J.J., Hankin, S., Johansen, M.P. Payne, T.E., Vine, M., Collins, R.N.,
Hoffmann, E., Loosz, T., M (2011). Movement of a tritium plume in shallow groundwater at a legacy low level
radioactive waste disposal site in eastern Australia. Journal of Environmental Radioactivity, 102: 943-952.
[2] AAEC (1985). The Little Forest Burial Ground – An Information Paper. AAEC report DR-19. Australian Atomic
Energy Commission, Australia.
[3] Payne ,T.E. (2012). Background Report on the Little Forest Burial Ground Legacy Waste Site, ANSTO Report E-
780. Australian Nuclear Science and Technology Organisation, Australia.
[4] Hotchkis, M.A.C., Child, D.P., Zorko, B. (2010). Actinides AMS for nuclear safeguards and related applications.
Nucl. Instr. Meth. B, 268: 1257-1260.
[5] Kelley, J.M., Bond, L.A., Beasley, T.M. (1999). Global distribution of Pu isotopes and 237 Np. The Science of the
Total Environment, 237/238: 483-500.
282

PA7-6
ICP-MS MEASUREMENT OF SAMPLES WITH HIGH SALINITY – SAMPLE CLEAN UP OR
TRANSIENT MEASUREMENT
Ch. Hein, J. M. Sander, R. Kautenburger, H. P. Beck, G. Kickelbick
Institute of Inorganic Solid State Chemistry, Saarland University, Campus Dudweiler, Am Martk Zeile 5,
66125 Saarbrücken-Dudweiler - Germany
The realisation of a high level nuclear waste (HLW) disposal in deep and stable geological formations is a
very important task for the next years or even generations. In Germany argillaceous rock as host rock for
such a disposal is one of the possibilities. For investigation of safe disposals for such high active and heat
developing wastes sorption experiments of metal ions onto the clay under natural conditions (pH, organic
substances and temperature) are necessary. With these experiments relevant geochemical parameters like K d -
values which are used in geochemical modelling experiments can be determined [1]. As in former studies [2,
3], Opalinus clay with its pore water was in the main focus of attention. In Germany not only Opalinus clay
but also other types of clay rocks are potential host rocks for a HLW disposal site. In northern Germany
another kind of clay is available and the pore water of the clay shows a very high salinity up to 5 M. The
main component of this pore water is sodium chloride with magnesium or calcium as additional cations. For
high sensitive measurement of metal ions mass spectrometry with inductively coupled plasma (ICP-MS) is
the method of choice. Samples with high salinity cannot be measured with ICP-MS due to contamination of
the cones and ion lens system in the ICP-MS. Due to the high metal sorption capacity of the clay only very
low metal concentrations have to be measured in the supernatant after the experiments. Therefore, a high
dilution of the samples before the measurement is not possible and the high matrix introduction system
(HMI) of Agilent is only capable of handling diluted salt contents of up to 6%. As solution for this problem
different possibilities are available. At first a sample clean-up and at second a transient measurement, which
allow a sample injection into the ICP-MS for only a few seconds.
A sample clean-up needs additional working steps with ion exchanging materials during solid phase
extraction. With this method a new source of errors and expenditure of time arises. In comparison, the time
resolved transient ICP-MS measurement is a good alternative. With a modification of the ICP-MS system
additional tools like a flow injection system are not necessary. The goal of the modification is to obtain a
time resolved signal for the analyte that can be integrated. This peak area should be directly proportional to
the analyte concentration. The used autosampler ASX-500 (Cetac Company, USA) was externally controlled
with the program Hyperterminal which is included in the Windows® (Microsoft Corporation, Redmond,
USA) operating system. Additionally, the modified setup allows the sample to be diluted with HNO 3 added
as make up flow via the peristaltic pump of the ICP-MS system.
Figure1. Modification of the ICP-MS system for transient measurement
283

To check the proof of principle for the transient method, some tests were performed. In the first step, a
calibration for the analytes uranium and europium (0.1, 0.5, 1, 5, 10, 50, 100 µg L -1 ) were made, which show
a good linear regression. In association to this measurements synthetic samples with a high salt content (1
and 3 M NaCl) and known concentrations of europium and uranium (0.8, 3, 8, 15, 30,
80 µg L -1 ) were prepared and analysed. With the transient measurement the determined metal concentrations
have a good agreement with the added metal concentration. For comparison of the transient method with the
standard ICP-MS method, sorption experiments with europium and uranium (0.04 mmol L -1 ) onto Opalinus
clay (4 g L -1 ) in 10 mM sodium perchlorate dissolution were performed. The samples were measured with
both methods and the results show only small differences when compared to each other. The application of
the transient method to real samples with 1 or 3 M ionic strength provides reproducible and congruent
results. The preliminary experiments in the ternary system Opalinus clay - metal - high ionic strength show
that with an increasing ionic strength the sorption of the metal ions onto the clay decrease strongly.
[1] Tertre, E., Hofmann, A. and G. Berger (2008). "Rare earth element sorption by basaltic rock: Experimental data and
modelling results using the “Generalised Composite approach”." Geochim. Cosmochim. Acta. 72: 1043-1056
[2] Kautenburger, R. (2011). "Batch is bad? Leaching of Opalinus clay samples and ICP-MS determination of extracted
elements." J. Anal. At. Spectrom. 26: 2089-2092
[3] Fröhlich, D., Amayri, S., Drebert, J. and T. Reich (2012). "Influence of temperature and background electrolyte on
the sorption of neptunium(V) on Opalinus Clay." Appl. Clay Sc. 69: 43-49
PA7-7
BATCH EXPERIMENTS MADE EASY – AUTOMATION OF ICP-MS BATCH EXPERIMENTS
J. M. Sander (1) , C. Hein (1) , R. Kautenburger (1) , H. P. Beck (2) , G. Kickelbick (1)
(1)
Inorganic Solid State Chemistry, Saarland University, Campus Dudweiler ;
(2)
Inorganic and Analytical Chemistry and Radiochemistry, Saarland University, Campus Dudweiler, D-
66125 Saarbrücken - Germany
According to a decision of the Council of the European Union (EU), all EU member states that operate
nuclear power reactors are to establish a national framework including arrangements on the final disposal of
their spent fuel and high level radioactive waste (HLW) by no later than August, 23 rd 2015 [1]. Today, a
broad and international consensus has been reached about the final disposal of high-level nuclear waste
(HLW) in deep and stable geological formations. A multi-barrier concept consisting of a technical,
geotechnical and geological barrier has to ensure an adequate protection of the environment from the stored
toxic and radioactive materials over a long period of time. Among the possible host rock materials, such as
salt, granite, and clay, the latter is a very promising candidate due to its favourable properties: A very low
hydraulic conductivity, a high sorption capability, an efficient filtration ability for interacting molecules,
small solubility of the clay components in water, and a high chemical buffer capacity. In order to guarantee
the sufficient retardation of radionuclides along their path of migration in the case of an intrusion of water
caused by an incident, the interactions with the host rock material leading to sorption and mobilisation
processes have to be examined.
A widespread method to analyse the behaviour of different reference clay minerals and naturally occurring
clays with respect to their suitability for a possible HLW disposal site is based on so called batch
experiments [2]. For this method, a set of small amounts of homogenised clay is equilibrated under the
aqueous conditions of interest and pH has to be adjusted iteratively due to the buffering of the clay sample
constituents. Once the sorption equilibrium has been reached, the extent of metal ion sorption is determined
by analysing the remaining amount of metal ions in the supernatant using inductively coupled plasma mass
spectrometry (ICP-MS). The sorption is dependent on a number of parameters, such as pH value, metal ion
concentration, salinity, and temperature. Furthermore, the presence or rather the concentration of naturally
occurring organic (e. g. lactate, formate or propionate [3] and humic substances) and inorganic ligands (e.g.
borate) can be varied.
This batch technique allows the precise adjustment of the different parameters listed above applying the
described and highly standardised batch experiments thus enabling researchers to produce a consistent and
detailed data record. However, these experiments are performed by hand and therefore rather time
284

consuming, especially with regard to the iterative nature of pH adjustment steps and subsequent equilibration
time. In order to facilitate the exploration of the widespread parameter space and with a possible personnel
layoff in mind, automated batch experiments are being developed in our group.
This includes the automation of the whole batch cycle starting with the weigh out of the clay at the beginning
followed by the addition of (saline) matrix, metal solutions, ligands etc. and conditioning steps including
thermostatisation of the samples as well as the iterative adjustment of the pH value and the final sample
preparation for ICP-MS analysis. Our instrumentation is based on a commercial platform that was modified
to suit the needs of the above described experiments. Optionally, anaerobic conditions during all
experimental steps can be realised. Obtained results can be compared with respect to sample throughput per
time, reproducibility of repeated measurements, congruence with manually performed batches, work
simplification or personnel commitment. Exemplary results will be presented on the poster.
This work is supported by the German Federal Ministry of Economics and Technology (BMWi) on the basis
of a decision by the German Bundestag. In addition, we would like to thank the Project Management Agency
Karlsruhe in Karlsruhe Technology Institute (PTKA-WTE) for funding (projects: 02E9683, 02E10196 and
02E10991) and our project partners for the kind collaboration.
[1] M. Sawicki, Official Journal of the European Union (2011) 199, 48–56.
[2] R. Kautenburger (2011). "Batch is bad? Leaching of Opalinus clay samples and ICP-MS determination of extracted
elements. " J. Anal. At. Spectrom. 26: 2089-2092.
[3] A. Courdouan, I. Christl, S. Meylan, P. Wersin, R. Kretzschmar (2007). "Characterization of dissolved organic
matter in anoxic rock extracts and in situ pore water of the Opalinus Clay." Appl. Geochem. 22: 2926-2939.
PA7-8
Electrochemical assessment of redox properties of argillaceous sediments using organic electron
transfer mediators.
A.L. Hoving a , T. Behrends a , M. Sander b , N. Maes c , C. Bruggeman c
a Department of Earth Sciences, Faculty of Geosciences, Utrecht University, Budapestlaan 4, 3584 CD
Utrecht, Netherlands
b Institute of Biogeochemistry and Pollutant Dynamics (IBP), ETH Zurich, Zurich, Switzerland
c SCK●CEN, Expert Group Waste Disposal, Boeretang 200, B-2400 Mol, Belgium
Argillaceous formations have been proposed as possible host rocks for long-term disposal of nuclear waste.
Since the mobility of some of the long-lived radionuclides, e.g. uranium and selenium, is dependent on their
redox state, knowledge of the redox properties of the solids in the formations is of pivotal importance to
predict long-term behavior of these redox sensitive radionuclides in the far field.
In a first instance, a redox potential has to be assigned to the system in order to assess the equilibrium redox
speciation of the radionuclides. However, conventional methods using redox electrodes to determine redox
properties and potentials of suspensions often lead to irreproducible and inaccurate measurements due to the
slow or absent electron transfer between the electrode and solid particles. In this study, we tested a novel
method using organic electron transfer mediators to facilitate electron transfer between the electrode and the
solids (Aeschbacher et al., 2010; Gorski et al., 2012) for determining the redox potential (E H ) and redox
properties (Electron donating/ accepting capacities, Q EDC /Q EAC ) of suspensions containing argillaceous
sediments. The goal is to obtain more reliable measurements of redox properties of argillaceous sediments
for predicting the behaviour of redox sensitive radionuclides in clay formations.
Here, we investigated the redox properties of Boom Clay samples collected from different locations in the
Netherlands. For the E H measurements, the experimental setup consisted of a two electrode electrochemical
cell, containing very small amounts of several electron transfer mediators covering a range of redox
potentials. The Q EDC and Q EAC were determined using a three electrode electrochemical cell with an electron
transfer mediator, one active around oxidizing conditions and one around reducing conditions, respectively.
The electrochemically determined redox properties will be discussed in view of the Fe-speciation in the
285

materials in order to assess whether the Fe(II)/Fe(III) redox couple is dominating the redox properties. In
particular, the measured redox potential will be compared with redox potentials calculated for various redox
couples based on the composition of the pore waters and solids in the clay formations. Finally, the measured
redox potentials will be used to evaluate the redox speciation of U and Se in Boom Clay based on
equilibrium thermodynamics.
Aeschbacher, M.; Sander, M.; Schwarzenbach, R. P. Novel electrochemical approach to assess the redox properties of
humic substances. Environ. Sci. Technol. 2010, 44, 87–93.
Gorski, C. A.; Aeschbacher, M.; Soltermann, D.; Voegelin, A.; Baeyens, B.; Marques, M.; Hofstetter, T. B.; Sander, M.
Redox properties of structural Fe in smectites: 1. Electrochemical quantification of electron donating and accepting
capacities. Environ. Sci. Technol. 2012, 46 (17), 9360-9368
PA7-9
THE USE OF DIAMOND LIGHT SOURCE TO STUDY RADIONUCLIDES BY X-RAY
ABSORPTION SPECTROSCOPY.
F. Mosselmans 1) , G. Law 2) , K. Morris 3) , S. Shaw 3) , S. Parry 1)
1) Diamond Light Source, Harwell Campus, Didcot OX11 0DE, United Kingdom
2) School of Chemistry, University of Manchester, Manchester M13 9PL, United Kingdom
3) School of Earth, Atmospheric and Environmental Sciences, University of Manchester, Manchester M13
9PL, United Kingdom
X-ray absorption spectroscopy (XAS) is a well-established technique for the study of radionuclides in
complex environments [1]. The Diamond Light Source, which opened in 2007, is the United Kingdom's
dedicated synchrotron light facility.
X-ray absorption spectroscopy involves scanning the energy of an incident X-ray beam over the absorption
edge of the element of interest (where the X-ray has sufficient energy to cause the ejection of a core-hole
electron). The resultant X-ray absorption spectrum contains information about the local electronic structure
and site geometry of a particular element. Modern synchrotron sources enable the speciation of elements in
low concentrations in small amounts of material to be examined, which reduces any radiation hazard and
comes closer to real environmental situations. For example the new I20 LOLA beamline has collected
EXAFS spectra from sub ppm metal-bearing aqueous solutions, while the microfocus spectroscopy beamline
has been used to probe the oxidation state of uranium in particles of a few tens of microns in size deposited
around a uranium processing site [2]. Diamond has built a small radiochemistry laboratory, where samples
are stored, when not being measured on the beamline and where the manipulation of uranium and technetium
bearing samples is permitted.
In the last two years, the core EXAFS beamline B18 has been used extensively by members of the BIGRAD
consortium to perform XAS experiments on samples to aid the understanding of the long-term behaviour of
radionuclides stored in a deep geological disposal facility (GDF). These experiments have looked at how
uranium and technetium behave with relevant minerals in the very alkaline fluids (pH 10 -13) likely to be
found on the border of the GDF with the host rock.
We have looked at how uranium and technetium react with iron oxide phases such as ferrihydrite and
magnetite, looking at the differences in behaviour if the radionuclide is present during the formation of the
iron oxide phase or added to the iron phase in an adsorption experiment [3]. This affects how the uranium is
partitioned in the iron oxide phase, either being largely incorporated into the host mineral lattice or at least
partially adsorbed on the surface of that phase. Further studies have looked at the effect of carbonate on some
of these processes. We have also explored the extent of microbiologically-mediated U(VI) reduction under
GDF relevant alkaline conditions [4].
Moreover we have studied the speciation of uranium when reacted with the mineral phases orthoclase,
quartz, and biotite, and with these same phases after they have been altered biogenically in very alkaline
fluids [5]. We have also studied the uranium speciation in the solid phase after reaction with Sherwood
286

sandstone, a model mineralogical system used in BIGRAD, which has been exposed to very alkaline fluids
for some time and monitored the evolution of the uranium speciation.
[1] X. Tan, et al., Molecules 15, 8431 (2010)
[2] N. E. Lloyd, et al., Miner. Mag. 73, 495 (2009).
[3] T. Marshall, et al. Miner. Mag. 76, 2071 (2012)
[4] A. Williamson, et al., Miner. Mag. 76, 2548 (2012)
[5] D. R. Brookshaw, et al., Miner. Mag. 76, 1519 (2012)
PA7-10
SPECIATION OF 79 SE AT ULTRATRACE LEVELS USING AN AUTOMATED
CHROMATOGRAPHIC SYSTEM COUPLED TO HIGH RESOLUTION ICPMS. APPLICATION
TO NUCLEAR SPENT FUEL CORROSION STUDIES.
L. Aldave de las Heras 1)* , M. Sandow 1) , D. Serrano-Purroy 1) , R. Sureda Pastor 2) , S. Van Winckel 1) , J.P.
Glatz 1)
1)
European Commission, DG Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany
2) Department of Chemical Engineering, Universitat Politècnica de Catalunya, Spain
Assessment of the safety of geological disposal of spent nuclear fuel requires substantial information
regarding the mechanisms and rates of release of the various radionuclides present in the fuel. Previous
studies have shown that release of radionuclides from spent fuel pellets is controlled by two processes – the
slow dissolution rate of the UO 2 grains and the rapid release of some elements, such as Cs, I and Cl 1,2 . In the
long-term safety assessment of spent fuel disposal, the rapid release is often referred to as the instant release
fraction or IRF and is considered to include (1) the release from the fuel/sheath gap which occurs in the first
weeks to months of contact with groundwater and (2) release of material segregated at grain boundaries. The
IRF is of particular interest in safety assessments, because some of the preferentially released radionuclides
(e.g. 129 I, 36 Cl 135 Cs, 79 Se, 14 C, 99 Tc and 126 Sn) are both long-lived and geochemically mobile 3 . In present
safety analysis, these radionuclides could have a significant contribution to the dose to man. The basis for the
calculated dose contribution is a simplified description of the release function. Assessment studies conducted
up to now lead to the conclusion, that a realistic release function results in lower peak doses rates, and thus
contributes to acceptance of nuclear waste disposal 4 . 79 Se is a β emitter produced by the fission of 235 U and is
one of the radionuclides of interest in the management of nuclear waste repositories because of its long halflife,
recently determined as 327000 years 5 , and its potential migration ability from the repository to the
environment. Individual selenium species differ greatly in their behaviour. Speciation analysis should
provide a more realistic and true status of this nuclide in waste matrices and its transport in the environment
using more accurate estimation and modelling scenarios.
Pure ß-emitting nuclides pose a challenging task for reliable, quantitative determination due to vulnerable
radiometric and mass spectrometric methodologies, both requiring chemical purification in advance for the
removal of interfering activity and isobars. Inductively coupled plasma mass spectrometry (ICPMS) has been
used increasingly for the determination of long-lived radionuclides. This technique combines several
advantages over classical radiochemical techniques such as high sensitivity and selectivity, low detection
limits, and the possibility to couple ICPMS on line to different separation techniques. Due to isobaric
interferences on several of its most abundant isotopes and its high ionisation potential (9.75 eV), Se is one of
the most difficult elements to quantify by ICPMS. Nevertheless, the introduction of ICPMS equipped with
collision/reaction cells overcomes problems linked with the occurrence of polyatomic interferences,
especially the elimination of the spectral background due to argon species. Methods for selenium speciation
analysis employ typically chromatography coupled with ICPMS 6,7 .
According to solubility data available for selenium 8 , Se(0) and Se(-II) (selenide) are moderately soluble (~10 -
8 M), but the Se solubility increases with the oxidation state (Se(VI)>Se(IV)). If Se is present as metal in the
fuel, the oxidation is probably much slower than when it is present initially already as selenite, under air
atmosphere or even more in anoxic conditions. So, the amount of leached Se will depend on the redox
287

Se (IV) y = 8253.5x
R 2 = 0.9992
250000
200000
150000
100000
50000
0
0 5 10 15 20 25 30
Concentration ng/g
conditions, and – under anoxic conditions – also on the pH. There is a possibility that the Se leaching will be
controlled by the solubility of Se(0) or selenide in anoxic conditions. In oxic conditions, this is less likely,
and it may have faster leaching as selenate, not limited by solubility (see figure 1) 9 .
The aim of the present study was the development of a method for the determination of 79 Se species at trace
levels by high resolution ICP-MS coupled to an automated chromatographic system (Figure 2).
FAST valve
250 µl loop
Chilled cyclonic
Spray chamber
PC3
ICPMS
Vacuum
Element 2
SC-DX Autosampler
Waste
Se 4/6 Speciation column
Eluent
(Phosphate buffer
pH 8)
1
2
Waste
ESI Micro Peripump
©Laura Aldave
Figure 1: Equilibrium diagram potential
Figure 2: System diagram for the determination of Se species
pH of the system selenium – water at 25 °C.
Parameters of the chromatographic method developed using analytical columns packed with a specific
stationary phase that selectively separate Se species, the choice of the eluent, and the impact of matrix
components in the separation are optimised in order to enhance sensitivity, improve detection limits and
preserve species presents in the sample. The method was applied to diluted nuclear spent fuel leachates
analogues consisting in a buffer solution (NaCl 19mM, NaHCO 3 1mM pH 7.4) with uranium concentration
varying from 10 -7 to 10 -5 M with individual Se species in different concentrations. In Figure 3 the elution Se
species (Se(IV) and (Se(VI)) profiles obtained as well as the Se(IV) calibration curve are shown. A fit for
purpose curve is obtained which does not introduce an extra uncertainty component showing that the method
has potential for the determination of Se species in nuclear spent fuel leachates. The detection limit is 12 pg
g -1 , representing an absolute amount of 3 pg. The repeatability, on the basis of three repetitions, is between
0.2 and 1 % in this concentration interval.
30000
Intensity (cps) Se-82 MR
25000
20000
15000
10000
Se(IV)
Peak area ( Se-82)
Se(VI)
5000
0
0 100 200 300 400 500 600 700 800
Time (sec)
Figure 3: Se species elution chromatograms and calibration curve.
1. Johnson., L.H. and J.C. Tait., SKB Technical Report 97-18, Sweden, 1997.
2. Johnson, L., et al., Journal of Nuclear Materials, 2012. 420: p. 54-62.
3. Nagra, Project Opalinus Clay. Nagra NTB 02-05, Nagra, Wettingen, Switzerland, 2002.
4. FIRST-Nuclides Euratom Project, http://www.firstnuclides.eu/.
5. Jörg, G., et al., Applied Radiation and Isotopes, 2010. 68(12): p. 2339-2351.
6. Darrouzès, J., et al., Journal of Analytical Atomic Spectrometry, 2005. 20(2): p. 88-94.
7. Montes-Bayón., et al., J. Chromatography A, 2003. 1000: p. 457-476.
8. Lemmens, K.,. SCK report R-3344, 1999.
9. Pourbaix, M., Gauthier - Villars, Paris, 1963: p. 554-559.
288

PA7-11
THE COMMISSIONING AND VALIDATION OF LOW-FLOW SAMPLING TECHNIQUE
Rhiannon Stoate; Laura Vivian; Tim Cartledge; Lucy Penrose; Simon Waldren
AWE, Aldermaston, Reading, Berkshire, RG7 4PR, United Kingdom
The groundwater monitoring programme at AWE monitors for specific radionuclides within the local
environment. The programme currently utilises numerous collection methods for groundwaters that are
dependent on sampling location and the specific history of the borehole unit; current techniques for sample
collection include manual pumps, bailers, submersible and peristaltic pumps. Collections involving the
pumping of water have the potential to disturb sediments at the sampling site and as a result the turbidity of
samples from these particular sites is inconsistent. One technique that is applied is that stagnant water is
removed via the associated purging process that is carried out prior to sampling, however, the execution of
this purging process 24 hours prior to collection raises questions as to how representative the sample is at the
time of collection. There is a requirement to improve current ground water monitoring techniques with the
use of a new method; previous studies have established that low-flow sample collection is the best available
technique (BAT) and fit for purpose regarding the groundwater monitoring programme at AWE.
The use of low-flow sampling was trialled at numerous locations to determine the applicability of
introducing this new method within the programme. The full utilisation of this method has resulted in
improvements to the process; the volume of water samples extracted from boreholes and thus the volume of
waste water for disposal has been reduced, as has the turbidity of these samples. The low-flow method also
results in a more representative sample being collected; the water samples collected through low-flow cells
were subjected to continuous monitoring until parameter conditions prior to sample collection had stabilised;
this ensured fresh water in the screen was collected. The monitoring process is conducted using a series of
probes; variables monitored as part of this process include: pH, conductivity, redox potential and dissolved
oxygen. Water levels were also monitored at the individual sampling sites to ensure that draw down was
minimised. Upon evaluation of the final trials in February 2013 low-flow was deemed as being the BAT and
was successfully implemented at all sampling locations at AWE site A for the April 2013 regulatory
collection; average stabilisation times for each parameter and maintenance and calibration procedures have
been established as pre-requisites for the operating procedure for this method.
PA7-12
DEVELOPING A ROBUST ANALYTICAL METHOD FOR RADIOANALYTICAL SEPARATION
OF AMERICIUM-241.
Kirsty Farrow; Gemma Murphy; Tim Cartledge; Simon Waldren
AWE, Aldermaston, Reading, Berkshire, RG7 4PR, United Kingdom
The implementation of the Environmental Permitting Regulations (EPR2010) sees the requirement for
improvements to methodology used to determine radionuclide composition of bulk waste materials at AWE.
EPR2010 provides detailed guidelines on radionuclide concentrations, below which a material is deemed to
be out of scope; if a sample has concentrations below the outlined limits, there is no need for sentencing as
radioactive waste. The principal change leading to this work is reductions in solid exemption and exclusion
rates. Current methodology is to be reviewed due to the recent changes to this legislation; defined
concentrations for actinides are significantly lower than the previously accepted values. As a result, previous
preferential radio-analytical techniques are no longer applicable as they can no longer provide the required
sensitivity.
A review of current methodology, with an aim to identify and implement the best available technique (BAT)
for analysis of the key radionuclides of interest plutonium, uranium and radium in order to meet the revised
limits is required. Studies will include a review of the current techniques, evaluation of current techniques
against required measurement levels arising from the changes in legislation and an investigation of
289

alternative techniques for the analysis of the key radionuclides in environmental samples. It is hoped that as a
result of this research, a BAT for the analysis of 241 Am, specifically in environmental and bulk waste
clearance samples can be developed and implemented as part of the bulk waste clearance programme at
AWE. The development of a method which results in a rapid and effective extraction of 241 Am, as well as a
precise and reliable measurement technique, will allow essential monitoring work to continue in line with the
EPR2010.
PA7-13
ANALYSIS OF URANIUM AND PLUTONIUM USING ACTINIDE RESIN
A R King, D Knight and A Fairhead
AWE, Aldermaston, Reading, Berkshire, RG7 4PR, United Kingdom
Analysis of environmental samples for uranium and plutonium has traditionally used radiochemical
separation using anion exchange resins. Eichrom’s ‘Actinide Resin’, which selectively pre-concentrates
uranium, plutonium, americium, thorium and neptunium from neutral and slightly acidic aqueous solutions,
is an alternative. Its selectivity for actinides provides an advantage over traditional anion exchange
separation for concentrating actinide isotopes. Actinide Resin was trialled at AWE to identify a method for
measuring uranium and plutonium isotopic ratios using TIMS (Thermal Ionisation Mass Spectrometry), and
using Actinide Resin to pre-concentrate the actinides of interest. A method has recently been developed for
the successful pre-concentration, recovery, separation and measurement of uranium and plutonium from the
resin.
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PA8
PA8-1
PA8-2
COMPUTATIONAL CHEMISTRY
QUANTUM CHEMICAL INVESTIGATION OF THE SORPTION OF SELENITE ON THE
CALCITE (104) SURFACE AND INCORPORATION INTO THE BULK PHASE
R. Polly, F. Heberling, B. Schimmelpfennig, H. Geckeis (Germany)
MOLECULAR DYNAMICS SIMULATIONS OF URANYL SPECIES
ADSORPTION/DIFFUSION IN MONTMORILLONITE PORES
Xiaoyu Liu, Zhong Zheng, Na Zhang, Chunli Liu (China)
PA8-1
QUANTUM CHEMICAL INVESTIGATION OF THE SORPTION OF SELENITE ON THE
CALCITE (104) SURFACE AND INCORPORATION INTO THE BULK PHASE
R. Polly 1)* , F. Heberling 1) , B. Schimmelpfennig 1) , H. Geckeis 1)
1) Karlsruher Institut für Technologie (KIT), Campus Nord, Institut für Nukleare Entsorgung (INE), Postfach
3640, 76021 Karlsruhe, Germany
Sorption reactions with mineral phases may have an essential impact on the mobility and bioavailability of
the oxidized selenium species in soils and sediments. Both incorporation and surface sorption are important
retardation mechanisms of the transport of radionuclides in ground water.
In the context of nuclear waste disposal, the radioisotope 79 Se is of special concern due to its long half-life
(1.1∙10 6 a) and the expected potentially high mobility. Selenite (SeO 3 2- ) and selenate (SeO 4 2- ) as anions are
retarded only weakly by common mineral surfaces. Therefore, 79 Se has been identified as a potentially dose
dominating fission product in long term safety assessment calculations for nuclear waste repositories (e.g.
[1]). Calcite is the most common polymorph of calcium carbonate. In the geological environment of potential
nuclear waste disposal sites calcite is e.g. present as a mineral constituent in clay formations (up to 20 % in
some cases), as fracture filling material in granitic rocks, or as a corrosion product of concrete based
materials in the technical barrier. Recent studies showed that selenite can be incorporated into calcite [2].
We studied the incorporation of selenite into the bulk phase as well as into the first surface layer of calcite in
the presence and absence of hydration water. Density functional theory (DFT) calculations are carried out
with periodic boundary conditions and plane-wave basis sets as implemented in the Vienna ab initio
simulation package (VASP). Electron exchange and correlation are described using the Perdew-Burke-
Ernzerhof (PBE) functional. The ion cores are dealt with by the projector augmented wave (PAW) method.
We employed an energy cut-off of 400 eV for the kinetic energy of the plane waves for all bulk and surface
calculations. For these calculations we employed monoclinic and hexagonal unit cells.
291

For the study of the incorporation of selenite into bulk calcite a 2x2x1 supercell based on the hexagonal unit
cell of calcite was used. This was necessary in order to retain the three-fold symmetry axis. SeO 2- 3 is not
planar as CO 2- 2-
3 . Therefore the incorporation of SeO 3 into calcite leads strain within the bulk phase of
calcite. Compared to the gas phase (r gas,Se-O =176 pm) the Se-O bond distance is shorter in the bulk phase with
r bulk,Se-O =173 pm.
The hydrated calcite (104) surface and the incorporation of SeO 2- 3 into the first surface layer of calcite
2-
replacing one CO 3 was studied with a 2x2x1 supercell based on a monoclinic surface unit cell [3]
resembling the calcite (104) termination and containing five layers of CaCO 3 . A vacuum gap of 1500 pm was
introduced above the surface so that the slabs are well isolated from their periodic images. Here, of course,
the spatial limitations do not hold as in the case above and the Se can protrude from the surface into vacuum
or the water phase . The three calculated Se-O bond lengths are r bulk,Se-O,1 =170 pm, r bulk,Se-O,2,3 =175 pm,
slightly longer as in the bulk phase. Additionally the symmetry of the SeO 2- 3 is reduced from C 3v to C s .
Besides the structural characterization of the selenite incorporation species in calcite, the main focus of this
study is the quantification of the selenite incorporation into calcite at standard conditions, and the
phenomenological thermodynamic description of selenite doped calcite, and of selenite coprecipitation with
calcite. Our DFT calculations confirm thermodynamic and theoretical calculations based on the single defect
method [4], that selenite incorporation into the bulk calcite structure is energetically and structurally
unfavorable. Insertion into the first surface monolayer of the calcite/aqueous solution interface can however
represent a relevant retention mechanism for SeO 3 2- .
[1] Ondraf/Niras, Technical Overview of the SAFIR 2 Report, Safety Assessment and Feasibility Interim Report 2; 2001.
[2] G. Aurelio, A. Fernandez-Martinez, G. J. Cuello, G. Roman-Ross, I. Alliot, L. Charlet, Structural study of
selenium(IV) substitutions in calcite. Chem. Geol. 2010, 270, (1-4), 249-256.
[3] F. Heberling, T. P. Trainor, J. Lützenkirchen, P. Eng, M.A. Denecke, D. Bosbach, Structure and reactivity of the
calcite-water interface. Journal of Colloid and Interface Science 2011, 354, (2), 843-857.
[4] F. Heberling, V. Vinograd, R. Polly, S. Heck, J. Rothe, J. Gale, D. Bosbach, H. Geckeis, B. Winkler, in preparation
PA8-2
MOLECULAR DYNAMICS SIMULATIONS OF URANYL SPECIES
ADSORPTION/DIFFUSION IN MONTMORILLONITE PORES
Xiaoyu Liu 1 , Zhong Zheng 1 ,Na Zhang 1 , Chunli Liu 1,*
1. Beijing National Laboratory for Molecular Science, Radiochemistry & Radiation Chemistry
Key Laboratory for Fundamental Science, College of Chemistry and Molecular Engineering,
Peking University, Beijing 100871, P.R. China
Molecular dynamics (MD) simulations of uranyl species adsorption/diffusion in
montmorillonite pores were carried out. 0, 0.05 and 0.10 M NaCl were introduced into the
simulation cell to investigate the influence of internal electrolyte concentration on the uptake
amount, species distribution and electrical double-layer (EDL) structures. Five of six cases
revealed that the β- and d-planes of the adsorbates are located 4.3-4.6 Å and 5.5-5.8 Å from the
clay planes, respectively. However, based on the split carbonate peaks, the remaining case formed
292

more uranyl-(bi)carbonate complexes than the other five, as shown by a peak or a minimum
shifted slightly farther away from the clay plane. For this outlying case, the charge density profile
confirmed the charge inversion of carbonates near like-charged surfaces. Collectively, the
simulations revealed the subtle influence with the internal NaCl concentration (0-0.05 M) on the
EDL structures, uptake amount and species distribution. In particular, a threshold concentration
(0.10 M NaCl in this study) for charge inversion within the β-plane may exist. Under this
condition, a pronounced change in the EDL structure occurs, which in turn causes a dramatic
alteration in the uranyl species adsorption relative to lower electrolyte concentrations.
Based on the species-based diffusion concept [1] , molecular dynamics simulations were
carried out to investigate the diffusive behavior of aqueous uranium species in montmorillonite
pores. Three uranium species (UO 2+ 2 , UO 2 CO 3 , UO 2 (CO 3 ) 2- 2 ) were confirmed in both the adsorbed
4-
and diffuse layers. UO 2 (CO 3 ) 3 was neglected in the subsequent analysis due to its scare
occurrence. The species-based diffusion coefficients in montmorillonite pores were then calculated,
and compared with the water mobility and their diffusivity in aqueous solution/feldspar nanosized
fractures. Three factors were considered that affected the diffusive behavior of the uranium species:
the mobility of water, the self-diffusion coefficient of the aqueous species, and the electrostatic
forces between the negatively charged surface and charged molecules. The mobility of U species in
the adsorbed layer decreased in the following sequence: UO 2+ 2 >UO 2 CO 3 >UO 2 (CO 3 ) 2- 2 . In the
2-
diffuse layer, we obtained the highest diffusion coefficient for UO 2 (CO 3 ) 2 with the value of
5.5×10 -10 m 2 s -1 , which was faster than UO 2+ 2 . For these two charged species, the influence of
electrostatic forces on the diffusion of solutes in the diffuse layer is overwhelming, whereas the
influence of self-diffusion and water mobility are minor. Our study demonstrated that the negatively
charged uranyl carbonate complex must be addressed in the safety assessment of potential
radioactive waste disposal systems.
[1] S. Kerisit, C. Liu, Molecular simulation of the diffusion of uranyl carbonate species in aqueous solution, Geochim.
Cosmochim. Acta 74 (2010) 4937-4952.
PB1
PB1-1
PB1-2
PB1-3
PB1-4
SORPTION/DESORPTION PHENOMENA IN
DYNAMIC SYSTEMS
AGING EFFECT OF SORPTION OF 32-YEAR-OLD PLUTONIUM COMPLEXES ON
SAVANNAH RIVER SITE SEDIMENTS
H.P. Emerson, B. A. Powell (USA)
MIGRATION OF URANIUM THROUGH SANDSTONE IN THE ALKALINE
DISTURBED PLUME FROM A CEMENTITIOUS REPOSITORY
M. Felipe-Sotelo, A.E Milodowski , N. Bryan, N.D.M. Evans (UK)
ANALYSING METAL SPECIATION AND MOBILITY IN CLAY - FROM ICP-MS
BATCH EXPERIMENTS TO A NEW APPROACH OF MINIATURISED CLAY
COLUMN EXPERIMENTS (MCCE) USING LC-ICP-MS
R. Kautenburger, C. Hein, J.M. Sander, H.P. Beck, G. Kickelbick (Germany)
STUDY OF SORPTION AND DESORPTION BEHAVIOUR OF RADIONUCLIDES IN
COLUMN FILLED WITH CRUSHED GRANITE
K. Videnská, Š. Palágyi, K. Štamberg, V. Havlová, H. Vodičková (Czech Republic)
293

PB1-5
STUDY ON THE INTERACTION BETWEEN HUMIC ACIDS AND GRANITE FROM
BEISHAN AREA, CHINA
Chunli Wang, Chun Li, Chunli Liu (China)
PB1-6
PB1-7
DETERMINATION OF MIGRATION PARAMETERS OF CRYSTALLINE ROCKS:
APPLICATION OF ELECTRO-MIGRATION METHOD ON SAMPLES WITH
DIFFERENT LENGTHS
P. Vecerník, V. Havlová (Czech Republic)
THE INFLUENCE OF HUMIC ACID ON THE MIGRATION OF CAESIUM, NICKEL
AND EUROPIUM CATIONS THROUGH QUARTZ SAND
S. L. Jain, N. Evans, N. Bryan (UK)
PB1-1
AGING EFFECT OF SORPTION OF 32-YEAR-OLD PLUTONIUM COMPLEXES ON
SAVANNAH RIVER SITE SEDIMENTS
Hilary P. Emerson and Brian A. Powell
Department of Environmental Engineering and Earth Science, Clemson University, 342 Computer
Court, Anderson, SC 29625
It has been previously noted that the reversibility of actinide sorption decreases over time even though the
specific mechanisms responsible for this process are unknown [1-3]. Further studies have suggested that
groundwater constituents may play a role in mobilizing actinide elements, especially natural organic matter
[4,5]. In this study sorption/desorption experiments and selective iron extractions were used to investigate
the desorption processes and aging effects of plutonium sorption in the presence of a subsurface clayey
sediment from an 11 year lysimeter experiment completed at the Savannah River Site 21 years ago (soils
have now aged in the presence of 239/240 Pu for a total of 32 years) [6].
Desorption experiments were completed for 7 and 28 days on source materials amended with ≈6 µg/g of
239/240 Pu at a soil concentration of 47.9±0.4 g soil /L and pH of 6.4±0.4. A variety of inorganic and organic
ligands were added to the solutions including:
- NaCl, CaCl 2 , Na 2 PO 4 , or NaF from 0.1 to 10 mM
- Citric Acid, Suwanee River Fulvic Acid, or DFOB from 0.05 to 5 mg carbon /L
- Or H 2 O 2 from 0.01 to 1% (3 to 300 mM) or NH 2 OH . HCl at 1 mM
The results showed that desorption of 239/240 Pu was independent of the chosen ionic strengths and ligands
with an average desorption K d of 51000±40000 mL/g or mobilization of 0.08±0.13% of the total 239/240 Pu
activity. Previous sorption K d measurements done by the Savannah River Site at 20 mM NaClO 4 ranged
from 9000 to 18000 mL/g for Pu(IV) and Pu(V) sorption over 33 days for similar pH values [7]. Despite the
variability of the K d values measured in this work, it is remarkable that they are within an order of magnitude
of each other with the diverse ligands under investigation, that the desorption of the ‘aged’ plutonium
complexes is so minor, and that the 7 and 28 day desorption periods did not produce significantly different
results.
To examine the potential aging effect, 242 Pu was used as a “fresh“ isotopic tracer. Selective extractions for
total free iron and amorphous iron oxides [8] were completed on source sediment suspensions of 70±1.4
g soil /L following a 3 day dual isotope sorption ( 242 Pu) and desorption ( 239/240 Pu) experiment. The 239/240 Pu
contaminated soil suspensions were reacted with 0.063±0.006 µg/g of 242 Pu for 3 days in the presence of ≈10
294

mM NaCl, 10 mM CaCl 2 ,10 mM Na 2 PO 4 , 10 mM NaF, 5 mg C/L Citric Acid, 5 mg C/L Suwanee River
Fulvic Acid, 5 mg C/L DFOB, 1% H 2 O 2 , or 1 mM NH 2 OH . HCl prior to selective extraction. In this manner,
desorption of 239/240 Pu and sorption of 242 Pu in the presence of these ionic strengths or ligands were analyzed
in addition to their association with the amorphous and crystalline iron fractions in the sediments. After the
three day sorption/desorption period, 97.9±2.1% of the 242 Pu and 99.95±0.05%of the 239/240 Pu was sorbed.
Solid phase total free iron and amorphous iron oxide extractions showed a significant fraction of plutonium
associated with iron phases with 22±2% of the 239/240 Pu (14±1% with the amorphous iron oxides and 8±3%
with the crystalline iron oxides) and 91±14% of the 242 Pu being associated with the total free iron fraction of
the soil (shown in Figure 1 below). It is important to note that the fractions of crystalline iron oxide were
calculated by difference because the extractions were not performed sequentially but on duplicate samples.
The average removal of 242 Pu by the amorphous iron oxide extraction was 81%. It is likely that the 242 Pu and
239/240 Pu are being mobilized by the specific conditions of the iron extraction solution rather than being
associated with those specific iron fractions. However, these data clearly show that the 3-day-old plutoniumsurface
complex is not nearly as strong as that after 32 years of aging.
Figure 1: Fraction of (a) 239/240 Pu after 32 years of aging and (b) 242 Pu after 3 days of aging associated with
the amorphous (red) and crystalline (blue) iron fractions on a Savannah River Site sediment
[1] Tinnacher, R.M., Zavarin, M., Powell, B.A., Kersting, A.B. (2011) ”Kinetics of neptunium (V) sorption and
desorption on goethite: An experimental and modeling study.“ Geochimica et Cosmochimica Acta, 75, 6584-6599.
[2] Missana, T., Garcia-Gutierrez, M., and Alonso, U. (2004) ”Kinetics and irreversibility of cesium and uranium
sorption onto bentonite colloids in a deep granitic environment.“ Applied Clay Science, 26, 137-150.
[3] Kaplan, D.I., Powell, B.A., Demirkanli, D.I., Fjeld, R.A., Molz, F.J., Serkiz, S.M., and Coates, J.T. (2004)
“Influence of oxidation states on plutonium mobility during long-term transport through an unsaturated subsurface
environment.” Environmental Science and Technology, 38, 5053-5058.
[4] McCarthy, J.F., Sanford, W.E., Stafford, P.L. (1998) “Lanthanide field tracers demonstrate enhanced transport of
transuranic radionuclides by natural organic matter.” Environmental Science and Technology, 32 (24), 3901-3906.
[5]McCarthy, J.F., Czerwinski, K.R., Sanford, W.E., Jardine, P.M., and Marsh, J.D. (1998) “Mobilization of transuranic
radionuclides from disposal trenches by natural organic matter.” Journal of Contaminant Hydrology, 30, 49-77.
[6] Kaplan, D.I., Demirkanli, D.I., Gumapas, L., Powell, B.A., Fjeld, R.A., Molz, F.J., and Serkiz, S.M. (2006) “Elevenyear
field study of Pu migration from Pu III, IV, and VI sources.” Environmental Science and Technology, 40 (2), 443-
448.
[7] Powell, B.A., Fjeld, R.A., Coates, J.T., Kaplan, D.I., and Serkiz, S.M. (2003) “Plutonium Oxidation State
Geochemistry in the SRS Subsurface Environment (U).” Savannah River Site Report# WSRC-TR-2003-00035.
[8] Sparks, D.L., Page, A.L, Helmke, P.A., Loeppert, R.H., Soltanpour, P.N., Tabatabai, M.A., Johnston, C.T., and
Sumner, M.E. (1996) “Methods of Soil Analysis, Part 3: Chemical Methods.” Soil Society of America, Madison, WI.
295

PB1-2
MIGRATION OF URANIUM THROUGH SANDSTONE IN THE ALKALINE DISTURBED
PLUME FROM A CEMENTITIOUS REPOSITORY
M. Felipe-Sotelo (1) , A.E Milodowski (2) , N. Bryan (3) , N.D.M. Evans (1)
(1) Chemistry Department, Loughborough University, Loughborough, LE11 3TU, UK
(2) British Geological Survey, Keyworth, Nottingham, NG12 5GG, UK
(3) Centre for Radiochemistry Research, School of Chemistry, The University of Manchester, Manchester,
M13 9PL, UK
Some of the concepts for the disposal of radioactive waste are based on a multibarrier system, designed to
mitigate the migration of radionuclides from the geological disposal facility (GDF) into the
geosphere/biosphere. The present UK concept for the geological disposal of intermediate (ILW) and low
level waste (LLW) makes use of cement and concrete for the encapsulation of waste in stainless steel drums,
the backfilling of vaults and the construction of the engineered facilities. While the cement will buffer the
porewater in the wasteform and the near-field repository environment to high pH, thereby contributing to the
retardation of many radionuclides by precipitation, this could potentially create a hyperalkaline plume
moving from the repository into the host rock. The alkaline cement leachate can react with the host rock,
promoting dissolution of some mineral phases and precipitation of new ones. Previous work on the effect of
the alkaline fluids on clays and crystalline rocks have shown that, in general terms, the alkaline solution
caused dissolution of aluminosilicates and precipitation of secondary phases as a result of its interaction with
the surrounding rocks; young cement leachates (mainly K/NaOH) generally cause the formation of zeolites
and less common feldspars, while Ca(OH) 2 solutions representing matured cement leachates cause
precipitation of CSH gels [1]. However, there is no agreement on the potential effect on the migration of
radionuclides, and while some authors have suggested that the formation of secondary phases could be
beneficial for the performance of a GDF [2], others have concluded that the restricted access to matrix
porosity as a result of the precipitation of secondary phases may potentially reduce the sorbing capacity of
the rock and impair the chemical and physical retardation of radionuclides [3].
The aim of this work is the assessment of the transport of uranium(VI) under advective conditions through
intact cores of sandstone in an alkaline environment, and the effect of the changes induced in the rock by the
reaction with cement leachates on the sorption of uranium. Intact cores of sandstone (100 mm length and 50
mm Ø), were pumped continually with a young cement leachate (YCL), spiked with uranium at a
concentration of 10 -7 mol L -1 , added as UO 2 2+ . The YCL is a K-Na-Ca-OH synthetic fluid of pH 13.1 and Eh
-170 mV (against Ag/AgCl electrode). Two scenarios were studied: (i) the sandstone had not been pretreated
and, (ii) the core had been saturated with humic acid (HA) before the injection of the YCL spiked
with U. Samples at the outlet were collected and the concentrations of U and HA were determined by ICP-
MS and UV-vis, respectively. The concentration of other constituents of the rock such as Al and Si, and also
elements originally present in the YCL such as Ca, were was also determined in the leachate, in order to
monitor mineralogical changes in the sandstone, caused either by dissolution or precipitation. At the end of
the experiments, the sandstone cores were cut axially and the migration pathways of U were observed by
autoradiography.
In the absence of HA, the sandstone showed a significant capacity to sorb U, and breakthrough was not
observed until after 280 pore volumes (≈ 6.5 L) of YCL, after which the concentration of U in solution
increased linearly (Figure 1). No differences were observed for the concentration of U in solution for filtered
(30 kDa membrane filters) and unfiltered samples, which would suggest the absence of colloids, or that any
present had an insignificant effect the transport of U. The analyses of the leachate showed significant
dissolution of Al and Si, with a clear maximum between 10 and 75 pore volumes, after which they reached
steady state. The same results were obtained for filtered samples, confirming the absence of colloidal
particles in the eluates. On the other hand, the concentration of Ca in the eluate decreased steadily over the
duration of the experiment (435 days), in comparison with the original concentration in the YCL. Preliminary
observations with scanning electron microscope indicate precipitation of Ca-rich phase with morphology
similar to zeolites, as well as formation of secondary Ca-K-Al-silicate phases.
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For the sandstone core saturated with HA, the injection of YCL caused the rapid leaching of the accumulated
organic matter, and the concentration of HA reached values around 3000 ppm in the first few pore volumes
of YCL through the core (about 50-200 mL). During this stage no sorption of U was observed, but as the
concentration of HA in the leachate decreased, the retention of U on the rock reached similar levels to those
observed in the previous case, even when the concentration of HA in solution was between 5 and 0.5 ppm.
Figure 1. Elution profile of uranium through an intact core of sandstone
in YCL in the absence of humic acid (unfiltered samples)
Observation of the migration pathways of U through the cores by autoradiography (Figure 2) showed clear
enrichment of U on the inlet end of the core (bottom of Figure 2), and also diffuse but significant U uptake
on the finer clay- and Fe 2 O 3 -rich laminae adjacent to the coarser and more porous and permeable sand layers.
Modelling results suggested that retention of U occurred by precipitation of CaUO 4 and the Eh conditions in
solution did not seem to promote reduction of UO 2 2+ to U 4+ . However additional spectroscopic experiments
are required to determine whether the clay mineral present in the sandstone might cause reduction of
oxidation state of U.
Figure 2. Photograph of sandstone core at end of experiment and digital laser-photostimulated storage
phosphor imaging autoradiograph of the migration pathways of uranium
Acknowledgments: The authors would like to thank NERC for the funding of the BIGRAD consortium.
[1] I. Devol-Brown, E. Tinseau, D. Bartier, A. Mifsud and D. Stammose (2007). “Interaction of Tournemire argillite
(Aveuron, France) with Hyperalkaline fluids: batch experiments performed with powdered and/or compact materials”.
Physics and Chemistry of the Earth. 32: 320-333.
[2] J.M. Soler and U.K. Mader (2005). “Interaction between hyperalkaline fluids and rocks hosting repositories for
radioactive waste: reactive transport simulations”. Nuclear Science and Engineering. 151: 128-133.
[3] C.I. Steefel and P.C. Lichter (1994). “Diffusion and reaction in rock matrix bordering a hyperalkaline fluid-filled
fracture”. Geochimica et Cosmochimica Acta. 58: 3595-3612.
297

PB1-3
ANALYSING METAL SPECIATION AND MOBILITY IN CLAY - FROM ICP-MS BATCH
EXPERIMENTS TO A NEW APPROACH OF MINIATURISED CLAY COLUMN EXPERIMENTS
(MCCE) USING LC-ICP-MS
R. Kautenburger (1) , Ch. Hein (1) , J. M. Sander (1) , H. P. Beck (2) , G. Kickelbick (1)
(1)
Inorganic Solid State Chemistry, Saarland University, Campus Dudweiler;
(2)
Inorganic and Analytical Chemistry and Radiochemistry, Saarland University, Campus Dudweiler, D-
66125 Saarbrücken - Germany
Nowadays, there is a broad scientific consensus on the technical merits of the disposal of high-level nuclear
waste (HLW) in deep and stable geological clay formations. Particularly, clay formations with a high
sorption capacity for metal ions are considered as one part of the natural barrier of a future high-level nuclear
waste disposal protecting mankind and nature from the possible impact of radionuclide contamination. A
wide set of geochemical parameters can influence the migration behaviour of radionuclides originated from a
leakage in a waste disposal for example competing ions released from the clay by infiltration of percolating
water, natural organic matter (NOM) as complex forming ligands, changes in temperature or pH-milieu of
the aquifer.
To study the behaviour of the released waste compounds, most sorption laboratory experiments were carried
out with the batch technique: A certain amount of the clay is homogenised and added to a high volume of an
aqueous solution like water or clay porewater including the appropriate sample components. Using this
technique a lot of different parameters can be tested in a short time but the obtained results are far apart from
natural conditions [1]. In contrast to the batch method, laboratory clay diffusion or in-situ diffusion
experiments can reveal results that are very close to nature. However, they have their great disadvantage in
an extremely long experimental time period (months or even years for higher valent metal ions). As a
consequence, only a very limited number of experiments can be performed.
In this study trivalent europium (homologue of americium) and uranium (VI) were used and their sorption
and desorption behaviour onto Opalinus clay were studied [2]. NOM can play an important role in the
immobilisation or mobilisation of metal ions due to complexation and colloid formation. This complexation
could interfere with the sorption of metal ions onto clay. In addition to humic acid (HA), we used other
naturally occurring organics in Opalinus clay like lactate, formate or propionate [3]. Therefore, we
investigated the complexation behaviour of the metal ions with NOM as well as the influence of present
NOM on the metal sorption onto clay [4].
Iodide
Salicylate
Iodide
Salicylate
Figure 1. Proof of principle for the MCCE: Influence of Eu(III) on the retention time of salicylate. Injection
of 2 µL of iodide (0.1 mM) and salicylate (0.1 mM) in the absence (left hand panel) or presence (right hand
panel) of europium (0.3 mM). LC-separation on a 120 mg kaolinite clay column (10 x 3.5 mm, L x ID) at a
LC-flow rate of 40 µl min -1 ultra-pure water (Milli-Q)
298

As methods, we used capillary electrophoresis hyphenated with inductively coupled plasma mass
spectrometry (CE-ICP-MS) to study the complexation behaviour of Eu(III) and U(VI) with HA. The
influences of metal concentration as well as the presence of competing cations from clay dissolution [1] as
well as cations from clay porewater on the complexation behaviour was analysed [5]. For the
sorption/desorption behaviour common batch experiments with mineral suspensions are performed, and in
comparison a new developed setup of miniaturised clay column experiments (MCCE) with compressed clay
was used to study the influence of NOM on the metal mobility in compact clay like Kaolinite or Opalinus
clay.
By the use of MCCE, mobility experiments with trivalent metal ions like europium in compacted clay can be
performed within only one day or for kaolinite just in a few minutes as shown in Figure1. As consequence,
an online coupling of the MCCE with ICP-MS should be possible. This hyphenation leads to quantitative
information on the elemental composition of the eluent directly after determination of the UV/Vis-active
compounds in the diode array detector of the LC.
We would like to thank the German Federal Ministry of Economics and Technology (BMWi), represented
by the Project Management Agency Karlsruhe (PTKA-WTE) for funding (projects: 02E9683, 02E10196 and
02E10991) and our project partners for the kind collaboration.
[1] R. Kautenburger (2011). "Batch is bad? Leaching of Opalinus clay samples and ICP-MS determination of extracted
elements." J. Anal. At. Spectrom. 26: 2089-2092
[2] R. Kautenburger, H.P. Beck (2010). "Influence of geochemical parameters on the sorption and desorption behaviour
of europium and gadolinium onto kaolinite." J. Environ. Monit. 12: 1295-1301
[3] A. Courdouan, I. Christl, S. Meylan, P. Wersin, R. Kretzschmar (2007). "Characterization of dissolved organic
matter in anoxic rock extracts and in situ pore water of the Opalinus Clay." Appl. Geochem. 22: 2926-2939
[4] C. Möser, R. Kautenburger, H.P. Beck (2012). "Complexation of Europium and Uranium by Humic Acids Analysed
by Capillary Electrophoresis - Inductively Coupled Plasma Mass Spectrometry." Electrophoresis 32: 1482-1487
[5] R. Kautenburger (2009). "Influence of metal concentration and the presence of competing cations on lanthanide
speciation with humic acid analysed by CE-ICP-MS." J. Anal. At. Spectrom. 24: 934-938
PB1-4
STUDY OF SORPTION AND DESORPTION BEHAVIOUR OF RADIONUCLIDES
IN COLUMN FILLED WITH CRUSHED GRANITE
K. Videnská (1),(2) , Š. Palágyi (1) , K. Štamberg (3) , V. Havlová (1) , H. Vodičková (1)
(1) Fuel CycleChemistry Department, ÚJV Řež, a.s., 250 68 Husinec-Řež, Czech Republic
(2) Department of Analytical Chemistry, Institute of Chemical Technology, Prague 6, 166 28 Prague, Czech
Republic
(3) Department of Nuclear Chemistry, Faculty of Nuclear Sciences and Physical Engineering, Czech
Technical University, Prague 1, 115 19 Prague, Czech Republic
The sorption and desorption of radionuclides were studied in columns of crushed granite and fracture infill of
various grain size. The aim of experiments was to study and compare tracer behaviour passing through
column filled with crushed granite and to quantify the effect of grain size and granite composition on
retardation of studied tracers.
Migration of radionuclides was investigated under dynamic conditions in column arrangement because the
dynamic continual transport experiments have many advantages in comparison with the static batch
operation; e.g. the dynamic arrangement and the aqueous volume to mass ratio are closer to conditions
existing in environmental; change of material quality is insignificant; column experiments allow
determination of transport parameters such as the sorption and desorption retardation coefficients (R),
sorption and desorption distribution coefficients (K d ), Peclet number and hydrodynamic dispersion
coefficient under dynamic conditions [1]. Those are determined by fitting the experimental breakthrough
curves using regression procedures based on the erfc-function, assuming a non-linear reversible equilibrium
sorption/desorption isotherm [2].
299

The radioactive ( 3 H, 36 Cl - , 85 Sr 2+ , 137 Cs + ) and stable (SeO 3 2- , SeO 4 2- ) tracers and two type of crystalline rock
materials were investigated (granite, coded as PDM1-1 and fracture filling materials, coded as PDM1-2).
Grain size of used crystalline rock was in the range of 0.063 – 1.25 mm. Sorption was studied using synthetic
granitic water (SGW) spiked radionuclides, desorption was studied using pure SGW.
The breakthrough curves showed that behaviour of 3 H and 36 Cl is different in comparison with other tracers,
both displaying behaviour of conservative tracers without any sorption. For this reason tritium is suitable for
determination of transport parameters. The anionic exclusion was observed in case of 36 Cl in granite PDM 1-
1. The selenate (Se(VI)) showed resembled 36 Cl and 3 H pattern as conservative non-sorbing tracer. On the
other hand selenite (Se(IV)) showed completely different sorption behaviour in comparison with selenate.
The significant sorption of selenite was observed on the both type of granitic rock. The values of sorption R
and K d showed influence of granitic size and granite composition on selenite sorption. The sorption of 85 Sr is
significantly higher than sorption of 3 H and 36 Cl, nevertheless the values of K d and R showed differences on
different granite material. The presence of fracture infill minerals (chlorite) significantly increased sorption
of 85 Sr in comparison with pure granite. The desorption transport parameters revealed similar patterns as
sorption parameters in case of both granitic materials which reflects the reversibility of sorption process. The
reversibility of strontium sorption was observed in case of both studied materials, 95% of sorbed strontium
was desorbed. The most probable sorption mechanism was ion exchange.
The experimental and theoretical sorption breakthrough curves are displayed in Figure 1.
A rel, C rel
n PV
Figure 1. Theoretical sorption breakthrough curves (solid line, resp. dashed line) and experimental sorption
breakthrough curves of transport: 3 H + (○, ●), Se IV (∆, ▲), Se VI (□, ■) a 85 Sr 2+ (◊, ♦) in crushed granite (blank
symbols) and in fracture infill materials (dark symbols) with grain size 0.125 – 0.63 mm in synthetic granitic
water.
[1] Š. Palágyi, K. Štamberg, Radiochim. Acta 98, 259 (2010).
[2] Š. Palágyi, K. Štamberg, H. Vodičková, J. Radioanal. Nucl. Chem. 283, 629 (2010).
300

PB1-5
Study on the interaction between humic acids and granite from Beishan Area, China
Chunli Wang, Chun Li, Chunli Liu *
Beijing National Laboratory for Molecular Sciences, Radiochemistry & Radiation Chemistry Key
Laboratory for Fundamental Science, College of Chemistry and Molecular Engineering, Peking
University, Beijing 100871,China
Author for correspondence liucl@pku.edu.cn
Being a kind of widely spread natural organic materials, humic acids can interact with various metal
ions and minerals, thus affect their speciation, translation and other chemical behaviours. It is necessary to
study the effect of humic acids on the adsorption and migration behavior of key nuclides in corresponding
minerals. Meanwhile, Beishan granite in Gansu province has been pre-selected as one of the most potential
host rock for HLW repository of China. The objective of our study was to investigate the interaction between
humic acids and granite from Beishan Area, China.
The sorption of humic acids on granite was investigated as a function of solid/liquid ratio, pH, ionic
strength and temperature. The sorption ratio increased with the increase of solid/liquid ratio. The sorption
ratio decreased while the pH increased, and it reached 90% when pH < 6. When ionic strength was lower
than 0.1 mol/L, the sorption decreased with the increase of ionic strength, due to the competition effect from
background electrolyte. On the other hand, when ionic strength reached 0.5 mol/L, it exhibited stronger
sorption due to the deposition of humic acid colloid. Furthermore, it indicated that the sorption of humic
acids on Beishan granite could be described as Freundlich isotherm, and the process could be described by
Pseudo-second-order kinetic model. Thermodynamic calculations showed that the sorption of humic acids on
Beishan granite were endothermic and spontaneous. This study suggested that the interaction between humic
acids and granite surface was outer-complexation at higher pH.
Figure 1 The influence of pH(a) and temperature(b) on the interaction between humic acid and granite
from Beishan Area, China.
[1] M. Samadfam, T. Jintoku, S. Sato. Radiochimica Acta, 88:717-721(2000).
[2] M. Salman , B. El-Eswed, F. Khalili. J.Applied Clay Science, 38:51-56(2007).
[3] Tan, X. L.; Fang, M.; Li, J. X.; Wang, X. K. J Hazard Mater, 168 (1), 458-465(2009).
301

PB1-6
DETERMINATION OF MIGRATION PARAMETERS OF CRYSTALLINE ROCKS:
APPLICATION OF ELECTROMIGRATION METHOD ON SAMPLES WITH DIFFERENT
LENGTHS
Petr Večerník, Václava Havlová
ÚJV Řež, a. s., Hlavní 130, 250 68, Husinec- Řež, Czech Republic
Diffusion of radionuclides from fractures into adjacent altered or fresh rock is considered to be one of the
important retardation processes for safety assessment of radioactive waste in a deep geological repository.
Diffusion within crystalline rocks with low porosity (< 1 %) is a slow process that results in the need to use
samples of limited length in the laboratory in order to receive results in reasonable time scale. The usage of
short samples brings some contradictions: even though the experiments can deliver the results within a
month, samples in fact do not represent real rock conditions, namely due to possible interconnections of
pores that would not be connected in real rock massive. This causes increased pore connectivity and possible
overestimation of laboratory diffusion results. Moreover, it is known from the literature that the disturbed
zone on the sample surface, produced by sawing and drilling, can extent up to 15 mm from the edge [1],
giving rise or increased tracer diffusivity through the sample. It was reported in the literature, that formation
factor F f decreases with increasing sample length [2] as non-realistic connectivity vanishes with less
disturbed samples.
Electromigration methods enable speeding up the laboratory diffusion experiments to gain rock migration
parameters (formation factor F f and effective diffusion coefficient D e ) in relatively short times (hours) in
comparison with through diffusion experiments (months or even years). Diffusion coefficient calculation is
based on the Einstein relation, describing the relation between the diffusivity and ionic mobility of ionic
solutes. Formation factor can be evaluated as a ratio of effective diffusion coefficient (D e ) and specie
diffusion coefficient in free water (D w ). Apparent formation factor is calculated as a ratio of the pore water
resistivity and rock resistivity.
As electromigration is the main process of solute transport, the studied tracer should be ionic one, therefore
the nonradioactive iodide anion I – is used. Methodology was based on through electromigration method
described by Löfgren [3, 4] and was applied and further developed in ÚJV Řež (former name - Nuclear
Research Institute Řež) laboratories [5]. Rock sample is placed between two compartments, the input one
which is holding an electrolyte with high tracer concentration and the output one holding an electrolyte
initially free of the tracer. The potential gradient over the sample could be achieved by placing an electrode
in each electrolyte and connecting the electrodes to a direct current power supply. The first generation of
electromigration cell enables used 1 cm thick samples only. The second generation of electromigration cell
was designed to allow using samples of different diameters and lengths. Nowadays rock samples of 4 – 5 cm
in diameter and 1 – 10 cm long of drilled core can be placed into the experimental cell.
The aim of the presented study was to test the use of the apparatus for longer samples of different rock types
in order to obtain migration data for less disturbed rock; furthermore to study the dependence of migration
parameters on sample length and type. The apparatus and method was successfully applied on rock samples
of different lengths from Czech Republic (Melechov granite) and Sweden (Äspö diorite). In Figure 1 there
are presented the breakthrough curves of iodide anion in through-electromigration experiments with different
samples length of Melechov granite. Migration parameters, gained by both resistivity and throughelectromigration
methods, revealed a good consistency for samples with different lengths. Determination of
apparent diffusion coefficient F f a using resistivity measurements enables laboratory support to resistivity
measurements in-situ.
302

4,5E-05
4,0E-05
1 cm 3 cm 5 cm
I - concentration (mol/l)
3,5E-05
3,0E-05
2,5E-05
2,0E-05
1,5E-05
1,0E-05
5,0E-06
0,0E+00
0 250 500 750 1000 1250 1500 1750 2000
time (min)
Figure 1: Breakthrough curves of iodide anion in through electromigration experiments with different
samples length of Melechov granite.
[1] Möri A. (2009). In situ matrix diffusion in crystalline rocks – An experimental approach. PhD thesis. University
Bern.
[2] Valkiainen M., Alto H., Lehikonen H., Uusheimo K. (1996). The effect of thickness in through-diffusion
experiments. Final Report. VTT Research Notes 1788, VTT Vuorimiehentie 5, Finland.
[3] Löfgren M. (2004). Diffusive properties of granitic rock as measured by in-situ electrical methods, Doctoral thesis
(2004), Royal Institute of Technology Stockholm.
[4] Löfgren M., Neretnieks I. (2006). Through electromigration: A new method of obtaining formation factors and
investigating connectivity. J. of Contam. Hydrol. 87, 237-252.
[5] Löfgren M., Večerník P., Havlová V. (2009). Studying the influence of pore water electrical conductivity on the
formation factor, as estimated based on electrical methods. SKB R-09-57, Svensk Kärnbränslehantering AB,
Stockholm, Sweden.
PB1-7
THE INFLUENCE OF HUMIC ACID ON THE MIGRATION OF CAESIUM, NICKEL AND
EUROPIUM CATIONS THROUGH QUARTZ SAND
S. L. Jain, a, 1 N. Evans, a * and N. Bryan b
a Department of Chemistry, Loughborough University, Loughborough, LE11 3TU
b Department of Chemistry, Manchester University, Manchester M13 9PL
1 Present address: School of Chemistry, University of St Andrews, Fife, KY16 9ST
The behaviour of anthropogenically produced radionuclide cations in the geosphere is an important, and ever
growing, area of study. The movement of cations with respect to silicate minerals is of particular importance,
as these minerals and rocks constitute a large proportion of the geosphere. While silicates are generally
poorly sorbing of radionuclides, complexation by humic acid (HA), a naturally occurring colloidal substance
of ill-defined chemical structure which is ubiquitous in soils and groundwaters, may change sorption and
transport behaviour. Sorption of HA-radionuclide species to a silicate surface could immobilise it, or HA
complexation may enhance radionuclide transport, or leave transport properties unaffected. Ionic strength
and pH may also affect radionuclide migration through sands when HA is present.
We have previously reported results of our sorption and desorption studies of radionuclides on quartz sand
and on Hollington sandstone. This poster describes the findings of our radiometric investigations, using
303

column and batch sorption studies, of the migration properties of caesium, nickel and europium ions with
respect to sand, at metal ion concentrations ranging from 1 × 10 -2 - 1 × 10 -12 mol dm -3 . We have also
monitored the impact of HA on sand, and investigated the sand-metal-HA ternary systems, using HA
concentrations ranging from 0-200 ppm, at fixed ionic strength and pH, and at ambient temperature.
Sand has a very low surface area (approximately 0.05 m 2 g -1 ), however can sorb HA well. Sand sorbs up to 25
cm -3 g -1 of CsNO 3 , and the presence of Cs + does not affect HA sorption onto sand. With a 1:1 solid to liquid
ratio for sorption studies, very low R d values are obtained for Cs + . For Ni 2+ , the amount of HA sorption is
dependent upon both the HA and Ni 2+ concentrations. At HA concentrations below 50 ppm, the amount of
HA bound to the sand is reduced in the presence of Ni 2+ , suggesting competition of Ni 2+ for sites on the sand.
However, at higher HA concentrations (70 – 100 ppm), increasing Ni 2+ concentration enhances the amount of
HA which binds to the sand, perhaps due to the formation of a Ni-HA-sand ternary complex.
Our column experiments have shown that both Cs + and Ni 2+ interact with sand. For Ni 2+ solutions, this is
thought to be due to precipitation on the sand as the hydroxide, rather than binding of the ions onto the
silicate surface, as is the case with Cs + . Fixing the pH to 6 allows Ni 2+ to travel through the column.
Introducing a high HA concentration retards Ni 2+ movement through the column, as HA bound to the silicate
surface complexes the Ni 2+ ions to form a ternary Ni-HA-sand species. By contrast, the presence of HA
increases the mobility of Eu 3+ through a silica column.
1. S.L. Jain, N. Bryan, N. D. M. Evans, The influence of humic acid on the migration of +1 and +2 cations through
quartz sand. Proceedings of the second Annual Conference on Decommissioning, Immobilisation and Management of
Nuclear Waste for Disposal, 15-16 December 2010, Manchester, UK.
2. S. L. Jain, N. D. M. Evans, N. Bryan, Sorption and Desorption of Radionuclides, Cs-127 and Ni-63, to Hollington
sandstone under far field conditions, Proceedings of the third Annual Conference on Decommissioning, Immobilisation
and Management of Nuclear Waste for Disposal, 14-15 December 2011, Coventry, UK.
PB3
PB3-1
PB3-2
PB3-3
PB3-4
PB3-5
COLLOID MIGRATION
HYDRO-GEOCHEMICAL EFFECTS ON COLLOID MIGRATION IN THE
ARTIFICIAL, SINGLE-FRACTURED GRANITIC ROCK
S. Lee, J.-W. Kim, M.-H. Baik, J. Jeong (Korea)
COLLOIDAL TRANSPORT IN A PODZOLIC SOIL CONTAMINATED BY
DEPLETED URANIUM: SITE INVESTIGATION, LABORATORY STUDIES AND
MODELING
S. Harguindeguy, P. Crançon, G. Lespes, L. De Windt,
F. Pointurier, M. Potin Gautier (France)
THRESHOLD OF BENTONITE COLLOID EROSION UNDER STATIC AND FLOW
CONDITIONS: RELEVANCE OF SCENARIO EVOLUTION
U. Alonso, T. Missana, M. García-Gutiérrez, N. Albarran,
M. Mingarro (Spain)
THE EFFECT OF COLLOIDS ON RADIONUCLIDE TRANSPORT IN COLUMN
EXPERIMENTS
S. Niemiaho, P. Hölttä, M. Voutilainen, J. Lehto (Finland)
THE EFFECT OF COLLOIDS ON RADIONUCLIDE TRANSPORT IN COLUMN
EXPERIMENTS
S. Niemiaho, P. Hölttä, M. Voutilainen, J. Lehto (Finland)
304

PB3-1
CESIUM MIGRATION IN GRANITE FRACTURES IN THE PRESENCE OF BENTONITE
COLLOIDS: COMPARISON BETWEEN THE GRIMSEL AND THE ÄSPÖ CASE.
T. Missana, M. Garcia-Gutierrez, U. Alonso, N. Albarran, M. Mingarro.
CIEMAT, Department of Environment, Avenida Complutense, 40
28040 Madrid (SPAIN)
Compacted bentonite is considered an adequate engineered barrier in high level radioactive waste (HLRW)
repositories, because it is expected to prevent the water influx to the waste and to delay radionuclide
transport towards the geosphere, providing high sorption capability for most radionuclides.
The possibility of erosion of the bentonite surface in contact with groundwater and the generation of
bentonite colloids is currently the subject of many researches (for example, EC BELBaR [1] and CFM [2]
projects). In fact, to establish under which conditions the presence of colloids can favour radionuclide
migration is a critical task for the performance assessment (PA) of HLRW repositories.
Colloid-driven contaminant transport depends on several chemical and hydrodynamic factors and on the
colloid-contaminant bond strength. The chemistry of the groundwater/clay system plays an important role on
the erosion of the bentonite, on the stability and mobility of generated colloids and also on the adsorption of
the contaminant at the colloid surface. On the other hand, water flow and flow paths are decisive on colloid
transport.
In this study, cesium transport in two granite fractures and in the presence of bentonite colloids will be
analysed comparing two cases, very different from the chemistry point of view: the “Grimsel” and the
“Äspö” case. These specific cases are of interest because: a) Grimsel (Switzerland) and Äspö (Sweden) are
underground laboratories, where investigations on radionuclide migration in crystalline rocks have been
performed for years; b) the chemistry of the groundwater of these two sites is very different, providing a very
different environment for colloids stability and mobility.
In general, bentonite colloids are quite stable in low ionic strength (< 1·10 -3 M) and alkaline waters (Grimsel
case), but became unstable when the ionic strength of the water increase over 1·10 -2 M or in the presence of a
significant amount of divalent cations (Äspö case).
Prior to transport studies, the adsorption of cesium in Grimsel and Äspö granite and in bentonite colloids was
analysed under a wide range of experimental conditions. As the main mechanism for cesium sorption is ionic
exchange, higher adsorption was always observed under low ionic strength conditions of the Grimsel case.
Transport experiments were carried out using water flow velocity from 1·10 -6 to 1·10 -5 m/s in two artificial
fractures (of Grimsel and Äspo granite, respectively) injecting 100 ppm of bentonite colloids in which
cesium ([Cs]=1·10 -7 M) was previously adsorbed (80 % in the Grimsel case and 20 % in the Äspö case)
suspended in synthetic waters representative of the two different cases.
The eluted water was analysed to measure cesium activity and the presence of bentonite colloids by photon
correlation spectrometry (PCS) in the same samples and to obtain simultaneously colloid and cesium
breakthrough curves. The breakthrough curve of cesium, without the colloids, presented a single peak with a
retardation factor (R f ) of ~200 in the Grimsel case and of ~170 in the Äspo case.
In the presence of bentonite colloids, in the Grimsel case, the breakthrough curve clearly showed a small
cesium peak in a position very similar to that of the conservative tracer (HTO), with an R f of 0.8, coincident
with the colloid breakthrough peak measured by PCS. However, the quantity of cesium recovered after this
peak was only the 0.15 % of the injected.
Since the colloid recovery was quite high (80%), this clearly demonstrated that cesium desorbed from the
colloid during the transport in the column. Other two peaks were observed in the breakthrough curve with R f
305

of 200 and 370 approximately. In the Äspö case, no elution of colloids was seen at all, which was no
surprising considering that they aggregate fast under the Äspö water conditions, but the “split” of the main
cesium peak in two peaks with R f of approximately 170 and 270 was also clearly observed. The appearance
of new peaks with retardation factors higher than those showed in the same system without colloids, is most
probably due to the modification of the sorption properties of the fracture surface caused by bentonite colloid
deposition [3].
It is important to remark that in similar experiments with other radionuclides (Sr [3], Eu[4] and U[5]) and
bentonite colloids, desorption was also clearly observed. Desorption from the colloids is a very important
factor limiting the role of colloids in radionuclide transport.
As a main conclusion, in both the Grimsel and Äspö cases, the presence of bentonite colloids did not
facilitate cesium transport in the medium, rather produced a slight increase in the overall cesium retardation
in the fracture.
Acknowledgments
This work has been partially supported by the EC within the BELBaR Project (Fp7 Fission, nº 295487)
[1] www.skb.es
[2] www.grimsel.com
[3] Albarran N., Missana T., Garcia-Gutierrez M., Alonso U., Mingarro M, Journal of Contaminant Hydrology 122
(2011) 76-85.
[4] Missana T., Alonso U., Garcia-Gutierrez M., Mingarro M, Applied Geochemistry 23 (2008) 1484-1497.
[5] Albarran N., Ph.D. Thesis, (2010) Universidad Autónoma de Madrid.
PB3-2
HYDRO-GEOCHEMICAL EFFECTS ON COLLOID MIGRATION IN THE ARTIFICIAL,
SINGLE-FRACTURED GRANITIC ROCK
Sanghwa Lee (1) , Jung-Woo Kim (2) , Min-Hoon Baik (2) , Jongtae Jeong (2)
(1) Kyung Hee University, 1732 Dukyoungdaero, Giheung-gu, Yongin, Gyeonggi-do, Korea
(2) Korea Atomic Energy Research Institute, 989-111 Daedeokdaero, Yuseong-gu, Daejeon, Korea
shluvu@khu.ac.kr
The granitic rock has been proposed as a common host rock for deep geological radioactive waste repository
in many countries [1, 2] including Korea [3]. Colloids generally show different behaviors for every
dissimilar experimental setup. Even same type of granites can bring out different results depending on their
origins. Since recent studies have investigated the theory supporting the fact that colloids can facilitate the
migration of contaminants in fractured media [4], investigating the fate and transport of colloids in the
fractured rock is important for safety assessment of geological radioactive waste disposal system. In the
present study, the laboratory scale experiments of mono-dispersed microspheres and bentonite colloids
transport were conducted to observe hydro-geochemical effects on the colloid transport in artificial, singlefractured
granitic rock.
Fig. 1 shows a picture and a schematic diagram of the experimental apparatus. Each rectangular rock block
was embedded in two acryl plates with a gasket between the acryl plate and the rock to prevent leaking of the
solution. The width and length of the fracture are 2 cm and 19 cm, respectively, and the hydraulic aperture
can be modulated. Preliminary to colloid transport test, a conservative tracer test was performed to get
hydraulic properties of the fracture. As a conservative tracer, bromide was chosen. From the tracer test, a
dispersion coefficient (= 0.024 cm 2 /s) and a linear velocity (= 1.448 cm/min) were calibrated by global
optimization using GlobalSearch toolbox included in MATLAB [5]. Using calibrated parameters, the
hydraulic aperture and the pore volume of the column were calculated. The governing equation used for the
calibration was the simple 1-D advection-dispersion equation [6]. After characterizing the fracture, several
colloid transport tests were performed on different hydro-geochemical conditions listed in Table 1. For the
first step, the effect of colloid size on their transport behavior in the fracture was investigated using four
306

different sizes of fluorescent microsphere (0.044, 0.1, 0.34 and 1.0 µm). The concentration of microspheres
in the effluent was measured by Hitachi F-7000 Fluorescence Spectrophotometer.
Fig. 2(a) illustrates the bromide break through curve (BTC) with error bars calculated from three times
repetition of experiments. The existence of dead volumes around the inlet and the outlet ports presumably
resulted in the tail on the bromide BTC. Fig. 2(b) illustrates BTCs of colloids for each size. The mass
balance was also checked for the colloid recovery at the end of the experiment. From the results, different
degrees of filtration for each size of colloids were observed. The results presumably imply that different kind
of filtration mechanisms are dominant or the different weighting factors of the filtration are applied
depending on the colloid size. Elimelech et al. [7] suggested four mechanisms for the colloid deposition:
straining, Brownian motion (diffusion), interception and sedimentation. Straining is a physical filtration
related to the relative particle size on the fracture aperture. Since even the largest microspheres used in this
study were about two orders smaller than fracture aperture, straining is not likely to play a significant role in
colloid deposition. Colloid interception occurs as colloids moving along a streamline come into contact with
the fracture wall. Degree of interception is highly dependent on the velocity of the colloids toward the
fracture surface. Sedimentation is negligible when the colloids are sufficiently buoyant and the flow direction
is vertical. In contrast to sedimentation, Brownian motion acts in all directions and is significant for colloids
smaller than 1.0 µm. For now, therefore, we assume that Brownian motion is dominating filtration process
which changes the probability of collision of colloid with fracture wall. However, the reliable discussions
should be made by further studies.
The final aim of this study is to investigate the effects of hydro-geochemical conditions on the fate and
transport behavior of colloids in the fractured matrix. Further, the most significant factor in the colloid
transport in the fractured granitic rock will be revealed. This experimental study is expected to nicely support
the relative modeling works and safety assessment of geological radioactive waste disposal system in the
future.
Acryl
Plate
20 cm
Outlet
Rock
Matrix
Fraction
Collector
Gasket
Syringe
Pump
Inlet
2 cm
Figure 1. A picture (left) and a schematic diagram (right) of the experimental apparatus
307

(a)
Relative Concentraion, C/C_0
1.2
1
0.8
0.6
0.4
0.2
0
The effect of dead volumes at in
let and outlet ports.
0 5 10 15 20
Time (min)
(b)
Relative Concentration, C/C_0
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0
0.044
0.1
0.34
1
0 20 40
Time (min)
60 80 100
Figure 2. Break Through Curve of (a) conservative bromide tracer and (b) four different size microspheres
(0.044, 0.1, 0.34 and 1.0 µm)
Table 1. Different hydro-geochemical parameters considered in the colloid transport experiment and the
experimental conditions for each parameter
Parameter
Condition
Colloid type
Latex, Bentonite
Colloid size (Latex) 0.044, 0.1, 0.34, 1.0 (um)
Flow rate
0.1, 0.5, 1.0 (ml/min)
Fracture aperture 100 to 600 (um)
Fracture roughness 2 types of roughness
Column gradient 0 ° ~ 180 °
pH 7 to 9
Ionic strength
2 ~ 10 (mM)
308

[1] Wang, T.-H.; Li, M.-H.; Teng, S.-P., 2009. Bridging the gap between batch and column experiments: A case study
of Cs adsorption on granite, Journal of Hazardous Materials 161: 409-415.
[2] Yamaguchi, T.; Nakayama, S.; Vandergraaf, T. T.; Drew, J. D.; Vilks P., 2008. Radionuclide and Colloid Migration
Experiments in Quarried Block of Granite under In-Situ Conditions at Depth of 240 m, Journal of Power and Energy
Systems 2(1): 186-197
[3] Baik M.-H.; Lee, J.-K.; Choi J.-W., 2009. Research Status on the Radionuclide and Colloid Migration in
Underground Research Facilities, J. of the Korean Radioactive Waste Society 7(4): 243-253
[4] Delos, A.; Walther, C.; Schäfer, T.; Buchner, S., 2008, Size dispersion and colloid mediated radionuclide transport
in a synthetic porous media, Journal of Colloid and Interface Science 324: 212-215
[5] MATLAB Primer R2013a, MathWorks, 2013
[6] Bedient, B. P.; Rifai, S. H.; Newell, J. C., 1999, Ground Water contamination Transport and Remediation – second
edition, Prentice-Hall
[7] Elimelech, M.; Gregory, J.; Jia, X.; Williams, R. A., 1995, Particle Deposition and Aggregation, Measuring,
Modeling, and Simulation. Elsevier, New York
PB3-3
COLLOIDAL TRANSPORT OF DEPLETED URANIUM IN A PODZOLIC SOIL: SITE
INVESTIGATION, LABORATORY STUDIES AND MODELING
S. Harguindeguy (1,2) , P. Crançon (2) , L. De Windt (3) , F. Pointurier (2) , M. Potin Gautier (1) , G. Lespes (1)
(1) Laboratoire de Chimie Analytique Bio-Inorganique et Environnement, IPREM-UMR 5254, Pau (France)
( 2) CEA, DAM, DIF, F-91297 Arpajon (France)
( 3) Ecole des Mines de Paris/Mines-ParisTech, Centre de Géosciences, F-77300 Fontainebleau (France)
The site considered in this study is located in Landes Gascony (France). The sandy podzolic soil in was
contaminated some decades ago by depleted uranium. Numerous years after the surface contamination, a
field monitoring revealed a 20 cm deep uranium migration in the soil containing 0.6wt.% of organic matter,
with a significant colloidal implication.
According to this history, the uranium speciation and colloidal mobility were investigated in the soils and
waters of this site. For that, contaminated soil samples were leached in batch tests using synthetic rainwater
at equilibrium with atmospheric CO 2 and O 2 at a Liquid/Solid ratio of 10. The soil samples were also
subjected to column experiments of 20-cm length considering several types of water chemistry: rainwater,
saline water and deionized water. Complementary, the subsurface water sampling was made in both
piezometers and the drainage systems.
Several analytical methods were consistently applied to both the groundwater samples and the laboratory
leachates. The total uranium concentrations and isotopic compositions were determined by ICP-MS after
ultrafiltration and uranium purification. The size and nature of colloids, as well as their association with
uranium, were reached using Asymmetric Flow Field-Flow Fractionation (As-Fl-FFF) coupled to multidetection
by ultraviolet (UV), multi angle laser light scattering (MALLS), dynamic light scattering (DLS)
and atomic mass spectrometry (ICP-MS). The uranium speciation in soil leachates as well as the chemical
interactions between U(VI) species and the reactive sites of the soil material were evaluated using the
geochemical code CHESS. In a second stage, the batch model was implemented in the reactive transport
HYTEC to simulate the uranium colloidal mobility in the column experiments.
The colloid sizes range from 10 to 200 nm. Two types of colloids co-exist: a small-sized organic fraction
ranging from 10 to 50 nm, and an inorganic population above 50 nm. Aluminum, iron and uranium are
mainly associated with the colloidal organic matter. In the batch tests, 10% of the total amount of uranium is
leached and the anthropogenic uranium is more mobile than the natural uranium in agreement with the field
data. Indeed, 25% of the total uranium in soil is depleted uranium, while in the leachates, 80% of uranium is
from anthropogenic origin. In the batch leachates, 80% of uranium is under colloidal forms and 20% under
dissolved species (< 10 kDa). In the column experiments, the organic colloids are more easily leached during
the low ionic strength events. The analysis of the column leachates indicates a strong correlation between
uranium concentration and total organic carbon (TOC). The uranium isotopic ratio of leachates does not
change neither with time nor the chemical composition of the leachate. On the contrary, the uranium
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concentration, colloidal/dissolved ratio and TOC asymptotically decrease with time. Complementary, the
subsurface water sampling clearly shows that the organic colloidal phases played a major role in the uranium
transport, with a dominant contribution of small-sized colloids (typically 10 – 50 nm). The opposite situation
takes place in the water sampled in the piezometers where 90% of uranium is found to be associated with the
dissolved fraction (< 10 kDa).
Simulation of the laboratory tests provides a mechanistic simple description of the major processes involved
in the speciation of U(VI) among the different phases. Modeling indicates that the functional groups of
humic matter bound to the sand aggregates are far from being saturated with respect to the aqueous
concentrations in the soil pore-water. U(VI) ions are strongly adsorbed onto the organic phase despite the
relatively low equilibrium pH of the soil in the batch tests (pH 4 – 5). In leachates, the presence of mobile
organic ligands competing with surface bound organics for U(VI) complexation must be included in the
speciation modeling. Without this mobile organic phase presumably fulvic acids, the model overestimates
the actual bound fraction, thus underestimating the actual amount of mobile U(VI).
PB3-4
THRESHOLD OF BENTONITE COLLOID EROSION UNDER STATIC AND FLOW
CONDITIONS: RELEVANCE OF SCENARIO EVOLUTION
U. Alonso 1) , T. Missana 1) , M. García-Gutiérrez 1) , N. Albarran 1) , M. Mingarro 1)
1) CIEMAT, Avda. Complutense 40, 28040 Madrid, Spain
The quantification of colloid erosion from compacted bentonite barrier, in a high-level waste repository, is
considered especially relevant because the eroded colloids may be stable and mobile, affecting radionuclide
transport. Additionally, if particle erosion were continue the barrier integrity will be compromised.
At present, many experimental and modelling efforts are being devoted to elucidate the mechanisms, and the
main chemical and physical factors affecting the formation of the bentonite gel, as forerunner of colloid
formation. The possibility that a colloid erosion threshold existed, under given chemical and flow conditions,
is also an issue in question.
Colloid erosion experiments were mainly carried out with natural raw FEBEX bentonite, a Na-Mg bentonite
with 92% smectite content and a 30 % of exchangeable Na.
Colloid erosion results, obtained under static conditions with different bentonite and different electrolytes,
showed that an initial fairly linear erosion rate exists but the system, if conditions are not changed, tends to
achieve equilibrium.
The maximum generation masses were both dependent on the clay properties (main exchangeable cation,
compaction density and bentonite type), on the water conditions (chemical composition, Ca content and ionic
strength) and on the extrusion path area.
Experiments carried out under constant flow velocities suggested a continuous linear erosion rate, within
experiments that lasted 200 days. These rates were dependant on initial electrolyte, higher rates measured in
Ca-free electrolytes. Constant and continuous bentonite erosion would imply that the barrier integrity could
be compromised and therefore, to check whether the erosion remained constant at longer experimental times
was considered necessary [1].
Longer times (3 years) erosion experiments clearly showed that erosion slows down and even stops in certain
cases. Thus, also under dynamic conditions, erosion seems to achieve a maximum possible value, depending
on the conditions of the experiments, as observed in static tests. The relevance of ionic exchange effects on
bentonite erosion rates at longer times is again pointed out.
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Results suggested that there is a maximum threshold of colloid generation of colloids, for given initial
conditions, dependent on the characteristics of the gel layer initially formed and on the feasibility of its
transport.
Nevertheless, the possibility that physico-chemical conditions change over time cannot be ruled out
considering the lifetimes of a repository. Thus to generation cells in which that had already reached
equilibrium, colloid generation already stopped or reached equilibrium, successive changes in water
chemistry, or flow rates (water velocities varied from 1.5·10 -8 m/s to 3.5·10 -6 m/s), were applied.
Significant changes in colloid erosion rates were not observed, in any case, under successive flow changes,
and the experimental conditions analyzed. The chemistry of the water/clay system seems to be much more
relevant on colloid erosion rates than flow velocity.
The research leading to these results has received funding from EU Seventh Framework Programme
(FP7/2007-2011) under the grant agreement Nº 295487 (BELBAR, Bentonite Erosion: effects on the Long
term performance of the engineered Barrier and Radionuclide Transport) and by the Spanish Government
under the project NANOBAG (CTM2011-2797).
[1] T. Missana, et al. Phys. Chem. of Earth 36 (17-18), 1607-1615 (2011).
PB3-5
THE EFFECT OF COLLOIDS ON RADIONUCLIDE TRANSPORT IN COLUMN EXPERIMENTS
S. Niemiaho, P. Hölttä, M. Voutilainen and J. Lehto
University of Helsinki, Department of Chemistry, Laboratory of Radiochemistry, P.O. Box 55,
FIN-00014 University of Helsinki, Finland.
Crystalline rock in Olkiluoto is being considered as a host medium for the final disposal of highly radioactive
spent nuclear fuel in Finland. The spent uranium fuel will be placed in final disposal tunnels in copper iron
canisters, and tunnel void space will be filled with bentonite clay. Colloids produced from the degraded
bentonite buffer may affect the migration of radionuclides through colloid-facilitated transport, which could
be of significance to the long-term performance of a spent nuclear fuel repository. The potential relevance of
colloids for radionuclide transport is highly dependent on the stability of colloids in different chemical
environments and the interaction of radionuclides with these mobile solid phases. In this work, radionuclide
and colloid interaction with natural rock was studied in fracture column and crushed rock column
experiments.
The fracture column was made from Olkiluoto tonalite; a drill core was artificially opened along the natural
fracture containing an altered zone and filling minerals. The final fracture width was about 3.5 cm, the
column length 6.8 cm and the fracture aperture 100 μm. The crushed rock columns were made of Kuru Grey
granite and strongly altered tonalite from the Syyry area in Sievi. The rock matrix of Kuru gray granite is
intact, fine-grained, non-foliated and equigranular. The total porosity is 0.2 % and the density is 2660 kg m -3 .
The rock matrix of the Syyry column was strongly and homogeneously altered containing visible mm-scale
pores. The total porosity is 2-8 % and the density is 2400 kg m -3 . The crushed rock column diameter was
1.5 cm and length 15 cm or 30 cm. In the flow experiments water was pumped into the columns at different
flow rates of 5–25 µl min -1 using a peristaltic pump to control the water flow rate. A short tracer pulse
(20 µl) was injected into the water flow using an injection loop and the tracer out-flow was collected using a
fraction collector.
At the beginning, the hydraulic properties in the columns were determined using non-sorbing tracers, Cl-36
and I-125 without colloids. In the low salinity Allard reference ground water, the colloids are assumed to be
stable and mobile. The transport of Sr-85 and Eu-152 in the columns was studied both in the presence and
absence of bentonite colloids. The colloid dispersion solution was made from MX-80 bentonite clay powder
which was mixed with Milli-Q water. The suspension was shaken for one week and the colloidal fraction
was then separated by centrifugation and the concentration of the bentonite colloids was determined by a
gravimetric method after drying the suspension. The beta activity of Cl-36 was detected using liquid
311

scintillation counting (Perkin Elmer Tri-Carb) and the gamma activity of I-125, Sr-85 and Eu–152 was
detected using a Wizard gamma counter. The colloidal particle size distribution and zeta potential were
determined by applying photon correlation spectroscopy (PCS) and dynamic electrophoretic mobility,
respectively (Malvern Zetasizer Nano ZS). The colloid concentration was determined by two different
techniques; by comparing the derived count rate obtained in PCS measurements of the colloid dispersions
with the count rate of a standard series of MX-80 bentonite with a known colloid concentration as well as by
using the Al content of montmorillonite analyzed by ICP-MS.
The transport of tracers through the columns was interpreted using a numerical compartment model modified
to calculate the advection and hydrodynamic dispersion in the columns. Based on the retention time
distribution without matrix diffusion in the column, the effect of matrix diffusion was calculated for each
short time interval by means of an analytic expression of the matrix diffusion. The effects of matrix diffusion
were observed in the elution curves of I-125 in the Olkiluoto fracture column and altered Syyry tonalite
column. Particularly Eu-152 as well as Sr-85 were strongly retarded in the Olkiluoto fracture column and
altered Syyry tonalite column in the absence of colloids. In the presence of bentonite colloids, however, a
slow elution of Eu-152 and Sr-85 was observed. The results show an effect of the water flow rate on the
colloid recovery from the rock columns, indicating colloid filtration in the fracture or at the crushed rock
surfaces. Preliminary results are presented and the effect of colloids on radionuclide migration in Olkiluoto
conditions is discussed.
312

PB5 FIELD AND LARGE-SCALE EXPERIMENTS
PB5-1
PB5-2
PB5-3
PB5-4
PB5-5
PB5-6
PB5-7
AN OVERVIEW OF THE LONG-TERM DIFFUSION TEST, GRIMSE TEST SITE,
SWIZTERLAND
A. Martin, M. Siitari-Kauppi, V. Havlová, Y. Tachi, J. Miksova (Switzerland, Finland, Czech
Republic, Japan)
IN-SITU MIGRATION EXPERIMENTS AT THE ROCK FRACTURES OF KURT (KAERI
UNDERGROUND RESEARCH TUNNEL)
J.-W. Kim, J.-K. Lee, M.-H. Baik, J. Jeong (Korea)
THE LATEST RESULTS ON COLLOID ASSOCIATED RADIONUCLIDE MOBILITY
FROM THE CFM PROJECT, GRIMSEL (SWITZERLAND)
T. Schäfer, I. Blechschmidt, M. Bouby, S. Büchner, J. Brendlé, G. Darbha, H. Geckeis, T.
Kupcik, R. Götz, W. Hauser, S. Heck, F. Huber, M. Lagos, A. Martin (Germany, Switzerland,
France)
QUANTIFYING 14 CH 4 MIGRATION AND FATE FOLLOWING SUB-SURFACE
RELEASE TO AGRICULTURAL SOIL
G. Shaw, B. Atkinson, W. Meredith, C. Snape, A. Hoch, D. Lever (UK)
REPRO - THE IN SITU RADIONUCLIDE MIGRATION EXPERIMENT AT ONKALO
UNDERGROUND FACILITY
A. Poteri, K. Helariutta, J. Ikonen, M. Voutilainen , M. Siitari-Kauppi, P. Andersson, J.
Byegård, K Nilsson, M. Skålberg, J. Kuva, P. Kekäläinen, J. Timonen, P. Pitkänen, K.
Kemppainen, J. Liimatainen, I. Aaltonen, L. Koskinen (Finland, Sweden)
CONTRASTING ACTIVITIES OF FALLOUT RADIONUCLIDES BETWEEN TWO
TYPES OF ARCTIC SOILS
E. Łokas, P. Wachniew , M. Gąsiorek (Poland)
AN IN-SITU EXPERIMENT TO STUDY SORPTION AND DIFFUSION OF SEVERAL
RADIONUCLIDES IN UNDISTURBED GRANITIC ROCK AT NATURAL CONDITIONS
IN THE ÄSPÖ HARD ROCK LABORATORY, SWEDEN. LTDE-SD (LONG TERM
SORPTION DIFFUSION EXPERIMENT)
J. Byegård, E. Gustafsson, S. Höglund, M. Skålberg, H. Widestrand (Sweden)
PB5-1
AN OVERVIEW OF THE LONG-TERM DIFFUSION TEST, GRIMSEL TEST SITE,
SWIZTERLAND
A. Martin (1) , M. Siitari-Kauppi (2) , V. Havlová (3) , Y. Tachi (4) , J. Miksova (5)
(1) Nagra, Hardstrasse 73, 5430 Wettingen, Switzerland
(2) Laboratory of Radiochemistry, University of Helsinki, A.I. Virtasen aukio 1, 00014 Helsinki, Finland
(3)
Dept. of Fuel Cycle Chemistry, UJV Rez, a.s., Hlavni 130, 250 68 Řež, Czech Republic
(4) Japan Atomic Energy Agency, Tokai, Ibaraki, Japan
(5) Radioactive Waste Repository Authority (RAWRA), Dlažděná 6, 110 00 Praha 1, Czech Republic
Matrix diffusion is very important in the context of a radioactive waste repository in that it provides a
mechanism for greatly enlarging the volume of rock that is accessible from a flow path and thus increases the
global pore volume which can retard radionuclides by diffusion and sorption. This is of special relevance
when considering the calculated contribution to dose from weakly- and non-sorbing radionuclides such as I-
129 and C-14.
Diffusion coefficients used in performance assessment calculations have mainly been derived from rock
samples that are de-stressed once removed from the host rock. This can lead to an over-estimation of matrix
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diffusion from lab experiments primarily due to sample preparation in comparison with real conditions in
crystalline rocks. Despite this, there have been very few long-term field-scale experiments to realistically
evaluate matrix diffusion of radionuclides in fractured rock with minimal disturbance to in-situ conditions.
These questions have formed the basis for the design of the Long Term Diffusion (LTD) project*. The LTD
project is a series of experiments which aims to obtain quantitative information on matrix diffusion under insitu
conditions and is an integral part of Phase IV at the Grimsel Test Site (www.grimsel.com). The project
was started in 2004 and is divided into two phases.
As part of Phase 1 (2004-2008), initial in-situ tests involved the characterisation of pore space geometry
(including determination of in-situ porosity for comparison with laboratory-derived data) using the C-14
labeled poly-methylmethacrylate (C-14 PMMA) technique and the NHC-9 chemical porosimetry technique.
With both techniques, in-situ derived porosity values were found to be 10-20% lower than porosity values
derived from measurements of rock samples in the laboratory. This difference was interpreted to be as a
result of de-stressing and mechanical artefacts of the rock samples. Next an in-situ diffusion experiment was
carried out where radionuclides (H-3, I-131, Na-22, Cs-134) and stable isotopes (Na, I) were allowed to
diffuse continuously into undisturbed rock matrix for 780 days with subsequent geochemical analysis of
matrix samples combined with predictive and post mortem modelling. After stopping the radionuclide
circulation, the diffusion interval was injected with resin, overcored and subsampled in order to measure the
radionuclide profiles in the rock. The main results can be summarized as follows: i) Based on post-mortem
modelling, effective diffusion coefficients of H-3 (HTO) were estimated to be lower than the lowest values
derived from rock samples analysed in the laboratory. However, it should be pointed out that there was
significant loss of HTO during the overcoring process. ii) Cs-134 penetration ranged from one to two
centimetres from the borehole wall depending whether the diffusion was perpendicular or parallel to the
foliation and was found to have predominantly sorbed onto mica. iii) Na-22 was found to have diffused up to
10 cm from the borehole wall. iv) Diffusion and sorption coefficients of Na-22 and Cs-134 derived from the
in-situ test were found to be similar to diffusion coefficients derived from rock samples.
As part of the ongoing Phase 2 (2009 – up to now) of the LTD project, a second in-situ experiment was setup
consisting of a borehole for circulation of the radionuclides, and an observation borehole. Several criteria
were used in the selection of radionuclides including duration of half-lives, safety aspects, feasibility,
comparison with results from the first in-situ experiment as well as the study of more safety relevant species
namely Cl-36 and Se. Based on this criteria, the LTD partners selected H-3, Cl-36, Na-22, Ba-133 and Cs-
134 as well as stable Se for use in the second long-term in-situ diffusion test. Injection of the radionuclide
cocktail is planned in 2013 and will be circulated for two to three years before overcoring, extraction and
analysis of the rock volume is performed.
It is envisaged that by bringing together the results from all work-packages in both phases, realistic values
for diffusion rates for in-situ matrix diffusion will be determined and improvements in confidence made to
existing performance assessment calculations.
*LTD experimental partners are Laboratory of Radiochemistry, University of Helsinki, Finland, Japan
Atomic Energy Agency (JAEA), Japan, Nuclear Research Institute (NRI) and the Radioactive Waste
Repository Authority (RAWRA), Czech Republic and NAGRA, Switzerland.
314

PB5-2
IN-SITU MIGRATION EXPERIMENTS AT THE ROCK FRACTURES OF KURT (KAERI
UNDERGROUND RESEARCH TUNNEL)
Jung-Woo Kim, Jae-Kwang Lee, Min-Hoon Baik, Jongtae Jeong
Radioactive Waste Disposal Research Division, Korea Atomic Energy Research Institute, 989-111
Daedeokdaero, Yuseong-Gu, Daejeon 305-353, Rep. of Korea
Radionuclide migration and retardation in a fractured rock has been considered as one of the most important
processes in the geological disposal of radioactive wastes. Many in-situ migration experiments have been
carried out to understand and predict the radionuclide migration and retardation processes in various
geological environments [1-4]. In this study, in-situ solute and colloid migration experiments were carried
out at KURT (KAERI Underground Research Tunnel), located within the research area of Korea Atomic
Energy Research Institute (KAERI), Daejeon, Korea.
For the in-situ migration experiments, we developed an experimental system which is mainly composed of
three main parts such as injection, extraction, and data treatment parts. The injection rate and extraction rate
were 1.3 L/hr and 24.6 L/hr, respectively. For the selection of a water-conducting fracture for solute
migration experiments, boreholes were drilled and fractures in the drilled boreholes were investigated by
water pressure tests and borehole image analyses using BIPS (Borehole Image Processing System). It was
observed from the water pressure tests that there was a water-conducting fracture in YH3 borehole, and the
hydraulic conductivity (K) was measured as 4.38×10 -6 cm/s. In addition, the borehole image analyses for the
YH3, YH3-1, and YH3-2 boreholes showed that a fracture of the YH3-1 borehole and a fracture of the YH3-
2 borehole were connected each other.
In the selected rock fractures, the migration of solutes and colloids in the fractured rock were investigated
using non-sorbing tracer (uranine) and latex colloid (size = 1.0 um), respectively. The concentrations of
uranine and latex colloid were 1,000 mg/L and 500 mg/L, respectively, and pH and IS of the groundwater
were about 8.7 and about 5 mM, respectively. From the in-situ experiments, 95% of uranine and 32% of
latex colloids were recovered at the extraction well. Although uranine was assumed to be non-reactive, about
5% of uranine was lost in the experiment. The larger loss of latex colloids was presumably contributed by the
interaction with rock fracture surfaces.
Based on the in-situ experimental results, a numerical model, so called fracture network model, was
developed to analyze the transport mechanisms of solutes and colloids in the rock fracture. Firstly, fractures
which were assumed to be flat planes connecting the two boreholes were surveyed based on the dip angle
and azimuth of the fractures measured from BIPS data at the locations of packers in both boreholes, and then
it was found that the two boreholes were connected by 3 intersecting fractures as shown in Fig. 1. Although
there was the shortcut for the groundwater flow through the 3 fractures, the horizontal flow path line was
selected in the modeling work. In the fracture network model, it was assumed that 1) all groundwater flows
in the fracture are only attributed to the extraction flow rate because of the large extraction rate, 2) the solute
transport follows 1-D advection dispersion equation in each fracture, 3) the solute can be lost owing to the
partial flow in the fracture intersection, 4) the solute can be diluted owing to the groundwater addition in the
fracture intersection, and 5) colloids can be filtered in the fracture unlike solutes. Colloid acceleration was
not considered in the model.
Using the fracture network model, dispersivities of uranine (= 0.362 m) and colloid (= 0.149 m), colloid
filtration coefficient (= 0.181 /m), and flow rate contribution fractions of each fracture (f 2 =1/1.2 and f 3 =1/4.7)
were calibrated. Generally, the flow rate contribution fraction at the fracture intersection is expected as 1/3.
f 2 , which is larger than 1/3, implies that the groundwater flow at the fracture 2 is mainly attributed by the
groundwater flow at the fracture 1. On the other hand, f 3 , which is smaller than 1/3, implies that the
groundwater flow at the fracture 3 is mainly attributed by the groundwater flow at the other than fracture 2.
Based on the calibrated parameters, the optimized modeling results were well correlated with the respective
experimental results (Figs. 2 and 3). The high accuracy of colloid transport model without considering
315

colloid acceleration implies that the colloid filtration is more significant factor rather than colloid
acceleration in the horizontal transport of 1.0 um sized colloids.
For the in-situ migration experiments in KURT considering the fracture network, a new modeling approach
was suggested and the modeling results were reasonably agreed with the experimental results.
Figure 10. Three intersecting fractures and potential path line between two boreholes.
Figure 11. Model verification results for the uranine transport.
Figure 12. Model verification results for the colloid transport.
[1] M. Mazurek, A. Jakob, P. Bossart, J. Contam. Hydrol. 61, 157 (2003).
[2] A. Möri, W. R. Alexander, H. Geckeis, W. Hauser, T. Schäfer, J. Eikenberg, Th. Fierz, C. Degueldre, T. Missana,
Colloids Surf. A 217, 33 (2003).
[3] K. S. Navakowski, G. Bickerton and P. Lapcevic, J. Contam. Hydrol. 73, 227 (2004).
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[4] P. Anderson, J. Byergård, E. -L. Tullborg, Th. Doe, J. Hermanson, A. Winberg, J. Contam. Hydrol. 70, 271 (2004).
Acknowledgement
This study was supported by Nuclear R&D program of Ministry of Education, Science, and Technology
(MEST), Korea.
PB5-3
THE LATEST RESULTS ON COLLOID ASSOCIATED RADIONUCLIDE MOBILITY FROM
THE CFM PROJECT, GRIMSEL (SWITZERLAND)
T. Schäfer 1* , I. Blechschmidt 2 , M. Bouby 1 , S. Büchner 1 , J. Brendlé 3 , G. Darbha 1 , H. Geckeis 1 ,
T. Kupcik 1 , R. Götz 1 , W. Hauser 1 , S. Heck 1 , F. Huber 1 , M. Lagos 1 , A. Martin 2
1
Karlsruhe Institute of Technology (KIT), Institute for Nuclear Waste Disposal (INE), Karlsruhe, Germany;
2 NAGRA Natl. Cooperat. Disposal Radioact. Waste, Wettingen, Switzerland; 3 Lab. Mat. Mineraux, UMR
CNRS, Univ. Haute Alsace, Mulhouse, France *thorsten.schaefer@kit.edu
The influence of colloidal/nano-scale phases on the radionuclide (RNs) solubility and migration behavior is
still one of the uncertainties in repository safety assessment [1]. In our work, we aim 1) to identify the
presence and the formation of relevant colloids in repository specific areas, 2) to determine their stability as a
function of geochemical parameters, 3) to elucidate the thermodynamics and kinetics of the colloid
interaction with radionuclides, 4) to perform laboratory and field experiments to quantify the colloid mobility
and their interaction with surfaces. The final goal is to state on the relevance of the nanoparticles (NPs) /
colloids for the radionuclide migration under natural geochemical conditions. In this contribution we report
on the progress concerning the colloid migration under near natural hydraulic conditions at the Grimsel Test
Site.
Within the Colloid Formation and Migration (CFM) project at the Grimsel Test Site (GTS Switzerland) [2] a
huge geo-technical effort was taken to isolate hydraulically a shear-zone from the artificially introduced
hydraulic gradient due to the tunnel construction. The construction is a combination of polymer resin
impregnation of the tunnel surface and a steel torus to seal the tunnel surface. The steel tube with
reinforcement rings is sealed at both ends with rubber based hydraulic “donut” packers and the annulus
between resin and steel ring is filled and pressurized with water to counteract the hydrostatic pressure of the
water conducting feature giving mechanical support to the resin. Natural outflow points of the MI shear zone
were localized prior to the construction and sealed by surface packers. This design gives the opportunity to
regulate outflow and thereby adjust the flow velocity in the fracture.
After optimization of the experimental setup and injection procedure through a number of conservative tracer
tests using fluorescence dyes so-called “homologue” tracer tests were performed by injecting a suspension of
Febex bentonite colloids containing adsorbed Eu, Tb, Hf in addition to the conservative tracer. A license was
granted in January 2012 by the Swiss regulator (BAG) to perform the first radionuclide tracer test under
these low-flow conditions. The injection cocktail of 2.25L volume was prepared at INE and transferred to the
GTS. A total colloid concentration of 101.4 ± 2.5 mg/L montmorillonite clay colloids were used, whereas 8.9
± 0.4mg/L were present as synthetic montmorillonite with structural incorporated Ni. For details on the
structural characterization of the Ni-montmorillonite phyllosilicate, see [3]. Beside the colloids and the
conservative tracer Amino-G (1646 ± 8ppb) the radioisotopes 22 Na, 133 Ba, 137 Cs, 232 Th, 237 Np, 242 Pu and
243 Am were injected. The trivalent and tetravalent actinides were quantitatively associated with the colloids
present as well as a part of the Cs, whereas Np(V) and Na are not bentonite colloid bond.
The injection of the radionuclide bentonite colloid cocktail was performed through borehole CFM 06.002-i2
into the MI shear zone and the water extracted under a constant flow rate of approx. 25mL/min from the
“Pinkel” surface packer (dipole distance 6.08m). The tracer cocktail was recirculated in the injection loop to
monitor the fluorescence decrease of the conservative tracer (Amino-G) and therefore providing an injection
function for consequent transport modeling. For on-site colloid analysis a mobile Laser-Induced Breakdown
Detection (LIBD) system similar to the one used in the CRR experiments [4] was transferred to Grimsel and
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installed in-line at the “Pinkel” outlet to directly monitor the mobile colloid fraction throughout the
experiment.
The conservative tracer Amino-G was recovered quantitatively both by on site inline and off-site
fluorescence detection. Analysis of the weakly sorbing tracers by γ-spectrometry performed by PSI-LES and
INE showed very good conformity and revealed recoveries for 22 Na, 137 Cs and 133 Ba of 64%, 10% and 1%,
respectively. The clay colloid recovery (integration until 11.4 days) determined by LIBD was quantified to
be ~66%, whereas HR-ICP-MS provided 33-38% by analyzing Al and Ni as structural components of the
clay particles. . The Al/Ni ratio taken as an indicator for the injected clay colloids did not vary during the
experiment (injection cocktail: Al/Ni ratio = 21.6 ± 0.5; effluent samples: Al/Ni ratio = 20.4 ± 0.5) and thus a
mobilization of Al containing colloid phases from additional sources in the fracture can be excluded.
Reasons for the divergent colloid recovery data are currently investigated.
Based on a number of colloid migration experiments performed in the MI shear zone a colloid
attachment/filtration rate can be estimated, which is lower than the value of 2∙10 -2 estimated by filtration
theory in [5]. All injected radionuclides including the strongly sorbing tri- and tetravalent actinides could be
detected in the effluent. For the initial quantitatively colloid associated actinides Am(III) and Pu(IV) a
recovery of at least 16% and 28-32%, respectively, could be determined. Np recovery is significantly
reduced to ~4 % in comparison to the CRR experiment of 82 ± 4 % [4], which hints to a kinetic controlled
Np(V) reduction observed through the increased residence time in Run 12-02 in line with the measured
pH/Eh conditions. The data of Run 12-02 obtained so far clearly show the mobility of bentonite derived
montmorillonite colloids under near-natural flow conditions in the MI shear zone of the Grimsel Test Site
[6]. The experimental data will be discussed in detail in the presentation.
Acknowledgement
The research leading to these results has received funding from the German Federal Ministry of Economics
and Technology (BMWi) under project KOLLORADO-2 (02E10679).
[1] T. Schäfer, et al. Appl. Geochem., 27 (2012) 390-403.
[2] H. Geckeis, et al., Actinide - Nanoparticle interaction: generation, stability and mobility, in: S.N. Kalmykov, M.A.
Denecke (Eds.) Actinide Nanoparticle Research, Springer- Verlag, Berlin, Heidelberg, 2011, pp. 1-33.
[3] Reinholdt, et al., Nanomaterials, 3 (2013) 48-69.
[4] H. Geckeis, et al., Radiochim. Acta, 92 (2004) 765-774.
[5] T. Schäfer & U. Noseck, Colloid/ Nanoparticle formation and mobility in the context of deep geological nuclear
waste disposal (Project KOLLORADO; Final report), FZKA report 7515, 2010.
[6] www.grimsel.com
PB5-4
QUANTIFYING 14 CH 4 MIGRATION AND FATE FOLLOWING SUB-SURFACE RELEASE TO
AGRICULTURAL SOIL
G. Shaw a , B. Atkinson a , W. Meredith b , C. Snape b , A. Hoch c and D. Lever c
a School of Biosciences , Agricultural and Environmental Science, Sutton Bonington Campus, University of
Nottingham, Sutton Bonington, LE12 5RD, UK. b School of Chemical and Environmental Engineering,
University Park, University of Nottingham, Nottingham, NG7 2RD, UK. c AMEC, B150 Thomson Ave,
Harwell Oxford, Didcot, Oxfordshire, OX11 0QB, UK.
Carbon-14 (half-life 5730 years) is expected to be released from a geological disposal facility (GDF) over a
time scale of several thousand years. Both 14 CO 2 and 14 CH 4 will be generated from waste materials within a
GDF but 14 CH 4 is likely to be the dominant carbon-14 species transported in the gas phase, ultimately
reaching the biosphere at low activity concentrations. Despite much work in recent years on the behaviour of
both CH 4 and CO 2 in soils, considerable uncertainty remains over the potential rate of transport of 14 C as
14 CH 4 in the soil, its oxidation to 14 CO 2 and subsequent absorption by plants, and its ultimate storage in soil
organic matter.
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Over the past three years the transport and retention of 13 CH 4 and 14 CH 4 in agricultural soil has been
investigated in laboratory [1] and field experiments [2] at the University of Nottingham, on behalf of the
Nuclear Decommissioning Authority Radioactive Waste Management Directorate. The experiments have
focused on the behaviour and fate of 13 CH 4 (a non-radioactive surrogate for 14 CH 4 , traceable using GC-
IRMS) injected into sub-surface soil, its subsequent transport through the soil gas phase and its oxidation to
13 CO 2 under laboratory and field conditions.
Transport in the gas phase within soils is dominated by diffusion, which results in rapid transport of labelled
CH 4 . Typically, 13 CH 4 injected 50 cm below the soil surface had completely diffused from the soil surface
within 24 to 48 hours. A number of ‘standard models’ have been published to calculate the effective diffusion
coefficient of gases in soils in relation to total soil porosity and water content. The observed rates of diffusion
in our field and laboratory experiments are generally greater than those predicted by these standard models.
Bacterially mediated oxidation of 14 CH 4 to 14 CO 2 (methanotrophy) in soils is an important determinant of the
ultimate radiological impact of 14 C migration from a GDF since 14 CO 2 outgassing from soils has the potential
to be fixed by photosynthesising crop plants. In the absence of better data, NDA currently assumes 100% of
carbon-14 in methane arriving at the surface is incorporated in plant tissues. To address this conservative
assumption, rates of oxidation of CH 4 to CO 2 in laboratory and field experiments were quantified using both
ambient atmospheric CH 4 and injected pulses of 13 CH 4 . For the former, a reduction in atmospheric CH 4
concentration is observed when gas samples are taken from non-saturated soils at increasing soil depth. This
results from ‘consumption’ of CH 4 by methanotrophic bacteria within the soil which provides a net sink of
atmospheric methane. After injecting 13 CH 4 into soils, the degree and rate of oxidation is indicated by shifts
in the δ 13 C value of CO 2 in the soil gas. Oxidation rates observed using both of these methods were variable
but usually low, hence a large proportion of 13 CH 4 injected into subsoil is transported to the soil surface
where it is lost to the free atmosphere.
Experimental data are being interpreted using physically-based models as part of the development of an
assessment model to be applied in a safety case for deep geological disposal of radioactive waste in the UK.
[1] ATKINSON, B., W. MEREDITH, C. SNAPE, M. STEVEN and G. SHAW (2011) Experimental and Modelling
Studies of Carbon-14 Behaviour in the Biosphere: Diffusion and oxidation of isotopically labelled methane ( 13 CH 4 ) in
laboratory soil column experiments. Report prepared for NDA Radioactive Waste Management Directorate.
[2] ATKINSON, G. SHAW, B., W. MEREDITH, C. SNAPE, M. STEVEN and A. HOCH (2012) Uptake of Carbon-14
in the Biosphere - Data from First Field Experiment, and Modelling of Initial Laboratory Experiments. Report prepared
for NDA Radioactive Waste Management Directorate.
PB5-5
REPRO – THE IN-SITU RADIONUCLIDE MIGRATION EXPERIMENT AT ONKALO
UNDERGROUND FACILITY
A. Poteri (1) , K. Helariutta (2) , J. Ikonen (2) , M. Voutilainen (2) , M. Siitari-Kauppi (2) , P. Andersson (3) , J. Byegård
(3) , K. Nilsson (3) , M. Skålberg (3) , J. Kuva (4) , P. Kekäläinen (4) , J. Timonen (4) , P. Pitkänen (5) , K. Kemppainen (5) ,
J. Liimatainen (5) , I. Aaltonen (5) , L. Koskinen (5)
(1) VTT, P.O.B. 1000, 02044 VTT, Finland
(2) Department of Chemistry, University of Helsinki, P.O. Box 55, 00014 University of Helsinki, Finland
(3) Geosigma AB, P.O. Box 894, 75108 Uppsala, Sweden
(4) Department of Physics, University of Jyväskylä, P.O. Box 35, 40014 University of Jyväskylä, Finland
(5) Posiva Oy, Olkiluoto, 27160 Eurajoki, Finland
Spent nuclear fuel from nuclear power plants, owned by TVO (Teollisuuden Voima Oy) and Fortum, is
planned to be disposed to a repository at a depth of more than 400 meters in the bedrock of Olkiluoto
(Eurajoki, Finland). The repository system of multiple release barriers: the nuclear fuel, copper canister with
the cast iron insert and bentonite buffer around the canister may fail during passage of the millennia.
Therefore, safe disposal of spent nuclear fuel requires information on the radionuclide transport and retention
properties within the porous and water-containing rock matrix along the water conducting flow paths.
319

Due to sorption and matrix diffusion radionuclides migrate at a lower speed than the velocity of the flowing
water. The rock matrix retention properties are determined by porosity of the rock matrix and solute
diffusivity in the rock matrix. In the performance assessment conditions most of the retention takes place in
the vicinity of the deposition holes. In-situ diffusion experiments provide data on the rock matrix retention
properties that is relevant for the conditions in bedrock close to the repository.
The first in-situ experiment using radioactive tracers have been performed within ONKALO, the
underground rock characterization facility in Olkiluoto, during the spring and summer of 2012. The
performed experiment is a part of REPRO project (experiments to investigate rock matrix REtention
PROperties). The REPRO project is realized as collaboration between universities, VTT and private sector
companies.
An artificial flow channel was built along a drill hole at depth of 420 m to study the matrix diffusion of
selected radionuclides. Water was pumped through the flow channel with a constant flow rate of 20 µl/min
and a mixture of HTO, 22 Na, 36 Cl and 125 I was injected as a short tracer pulse to the flowing water. The
passage of the tracers through the flow channel was measured both with an online gamma detector and by
taking laboratory samples from the water discharging from the test section. The experiment was completed
in about 5 months when the injected tracers were practically completely collected from the flow channel.
Concentrations of the tracer radionuclides as a function of time in the outflow from the test section, i.e.
breakthrough curves, were obtained. Diffusion and distribution coefficients of the rock matrix were
determined from the breakthrough curves for the used tracers by assuming advection, matrix diffusion and
sorption as the transport and retention processes (see Table 1). Porosities from laboratory experiments were
used in the analysis. These results provide data on matrix diffusion and sorption of radionuclides under insitu
conditions in the repository depth and are utilized to demonstrate that assumptions applied in the safety
case are in line with the site evidence.
Table 1. Rock matrix porosities, diffusion coefficients and distribution coefficients determined from
measured breakthrough curves of HTO, 22 Na, 36 Cl and 125 I
Element ɛ [%] D p [m 2 /s] K d [m 3 /kg]
HTO 0.7 7 × 10 -12 -
22 Na 0.7 7 × 10 -12 (0.5 – 5) × 10 -5
36 Cl 0.3 1 × 10 -12 -
125 I 0.7 7 × 10 -12 (0.5 – 5) × 10 -5
PB5-6
CONTRASTING ACTIVITIES OF FALLOUT RADIONUCLIDES BETWEEN TWO TYPES OF
ARCTIC SOILS
E. Łokas (1) , P. Wachniew (2) , M. Gąsiorek (3)
1) Department of Nuclear Physical Chemistry, Institute of Nuclear Physics (PAS), Radzikowskiego 152,
31-342 Krakow, Poland
2) Department of Applied Nuclear Physics, AGH – University of Science and Technology, Mickiewicza 30, 30-
059 Krakow, Poland
3)
Department of Soil Science and Soil Protection, University of Agriculture, Mickiewicza 21, 31-120
Krakow, Poland
Radioactive isotopes released by testing of nuclear weapons in the atmosphere, which peaked in the 1960’s
are still present in the environment, especially in soils that play important role in accumulation of airborne
radionuclides. This fact provides a rationale for a better understanding of the inventories, behaviour and
transfers of the fallout radionuclides in the Arctic environment in conditions of current climatic change.
Thawing, increased precipitation and elevated summer temperatures will enhance transport of radionuclides
into deeper layers of soils and across the landscape. Retreating glaciers uncover areas that are sites of intense
320

iogeochemical and geomorphic activity. Information on levels and behaviour of radionuclides in these
dynamic parts of the Arctic environment is limited. The aim of this study was to investigate activity
concentrations and inventories of the fallout radionuclides 137 Cs, 238 Pu, 239+240 Pu and 241 Am in soil profiles
from the proglacial zone of the Werenskiold glacier and from the nearby tundra in the southern part of the
Wedel Jarlsberg Land (Svalbard). Anomalously high activities of these fallout radionuclides were found in
the proglacial zone where up to 15 cm thick vertical sequences of fine-grained deposits were collected. These
deposits consist of material washed-out into the small depressions of the bottom moraine uncovered by the
retreating glacier in the last few tens of years. Detailed mineralogical, granulometric and chemical
investigations reveal vertical homogeneity of properties of these initial soils. Despite this lack of
differentiation in physical and chemical characteristics, the depth profiles of radionuclide activities show
pronounced peaks between depths of 2.5 to 3 cm for 137 Cs, 238 Pu, 239+240 Pu and 241 Am with the highest
activities of 3000 Bq/kg, 1 Bq/kg, 20 Bq/kg and 14 Bq/kg, respectively. Depth distribution of the
radionuclides is thus not related to soil properties and might reflect past variations in radionuclide fluxes to
soils. Total inventories of 137 Cs, 238 Pu, 239+240 Pu and 241 Am reach very high values of 100000 Bq/m 2 , 50
Bq/m 2 , 900 Bq/m 2 and 600 Bq/m 2 , respectively. A primary source of the highly active material found in the
studied soils iscryoconite holes that develop on glacier surface and temporarily accumulate airborne material
deposited on glacier surface. Melting of the glacier leads to release of this material and its downstream
transport with melt waters until it reaches deposition sites in the proglacial zone. Development of plant cover
on these areas will lead to incorporation of the accumulated radionuclides into the food chain of the fragile
Arctic ecosystem. Contrastingly, activities of these radionuclides in soil profiles collected in the nearby
tundra showed considerably lower values (for 137 Cs 120 Bq/kg, for 238 Pu 0.11 Bq/kg, for 239+240 Pu 2.3 Bq/kg
and for 241 Am 0.9 Bq/kg). Furthermore, the radionuclides concentrated in the uppermost soil layers showed
the ability of organic-rich material of topsoil to retain airborne radioactive contamination.
Activity ratios of 238 Pu/ 239,240 Pu and 241 Am/ 239+240 Pu were investigated in order to identify primary sources of
these radionuclides. Generally the average 238 Pu/ 239+240 Pu activity ratios were similar to the global fallout
ratios for Svalbard of 0.025. 241 Am/ 239+240 Pu activity ratios for the soil exceeded in some cases the global
fallout ratio of 0.37 due to the ingrowth of Am from the decay of 241 Pu.
Pecularities of the Arctic environment may give rise to unusual patterns of radionuclide distribution. Yet,
activity concentrations and inventories in different compartments and regions of the Arctic environment are
poorly investigated. Such information is important for assessment of radioecological situation of the Arctic
but can be also used to estimate time scales of the geomorphic and pedogenic processes.
This study was supported by the Foundation for Polish Science PARENT-BRIDGE Programme co-financed
by the EU European Regional Development Fund.
PB5-7
AN IN-SITU EXPERIMENT TO STUDY SORPTION AND DIFFUSION OF SEVERAL
RADIONUCLIDES IN UNDISTURBED GRANITIC ROCK AT NATURAL CONDITIONS IN THE
ÄSPÖ HARD ROCK LABORATORY, SWEDEN. LTDE-SD (LONG TERM SORPTION
DIFFUSION EXPERIMENT)
Byegård, Johan; Gustafsson, Erik; Höglund, Sanne; Skålberg, Mats; Widestrand, Henrik*
GEOSIGMA AB, Stora Badhusgatan 18-20, SE-411 21 GÖTEBORG, Sweden
*) Mechanical and Process Engineering/Nuclear Power, Vattenfall Research and Development AB, BU
Engineering/BD Asset Development, P.O. Box 475, SE-401 27 Göteborg
An experiment has been performed at a depth of 410 m in the Äspö Hard Rock Laboratory in Sweden in
which studies of radionuclide retention processes (sorption and diffusion) has been studied. The work was
mainly focused on the aim to study radionuclide retention under natural conditions (e.g., pressure, rock stress
conditions, groundwater chemistry, mineralogy, etc.) and to confirm the applicability of the laboratory
derived retention parameters (porosity, sorption coefficients and diffusivities). The experimental set-up
consisted of a borehole into the rock in which a natural fracture surface was isolated and natural groundwater
was allowed to circulate in a closed loop connecting the isolated borehole section to a control unit (e.g.,
pump, pressure regulator, pH-electrode, Eh-electrode, on-line detector, sampling and injection devices).
321

Radionuclides were injected and the sorption on the fracture surface, combined with the diffusion in to the
inner pores of the rock, was studied by following the concentration decrease of the radionuclides. Besides the
natural fracture surface, a small size borehole was drilled in the centre of the fracture surface and a small part
of this was isolated and exposed to the circulated groundwater; this in order obtain information of the direct
interaction of the radionuclides with non-altered oppositely to the natural fracture surface where altered rock
was expected to cover the surface.
After half a year of contact of the tracer solution to the rock, the circulation experimental phase was finalized
and the test section was overcored; this in order to isolate small drill core samples which should be used to
measure penetration profiles for the different tracers. However, the present work will focus on the
experimental aspects of the first circulation phase of the experiment; the results and evaluation of the
measured penetration profiles will be dealt with in a future paper.
In the experiment, 23 different radionuclides were injected where the majority comprised of elements which
are present with long-lived radioisotopes in the spent nuclear fuel. Radioisotopes with decay involving γ-
emission were preferred since this enabled multi-tracer measurement using γ-spectrometry with High Purity
Germanium (HPGe) detectors. These measurements were performed both for sample measurements and for
on-line measurements. However, the strong variation of sorption behaviour for the different radionuclides
made it necessary to develop a separation method to enable quantification of radionuclides, otherwise
strongly interfered by the Compton interaction from the radionuclides present in high concentrations.
For the radioelements that were used as tracers, speciation calculations was performed in order to predict
what chemical species of each element was expected to persist in the groundwater. These results are
presented and are discussed in relation to results from the experiments in which the sampled groundwater
was passed through 1) anion exchanger 2) cation exchanger and 3) 20 nm filter. Discussions will also be
made concerning some of the limitations performing a multielement radionuclide experiment, e.g., the
possibilities of preventing adsorption of strongly hydrolysed elements (tri and/or tetra-valent elements) on
the experimental equipment.
Results will be given for the concentration decrease in the water phase a function of the experimental time;
this will be compared to predictions where diffusion and adsorption are calculated using a linear adsorption
matrix diffusion model. There will also be a short introductory to the results of the measured penetration
profiles showing the magnitude and the heterogeneity of the diffusion processes in to the pores of the rock.
322

PC2 COUPLING CHEMISTRY AND TRANSPORT
PC2-1
PC2-2
PC2-3
PC2-4
PC2-5
PC2-6
RADIONUCLIDES TRANSPORT OF PYRO-PROCESSED WASTE IN KOREAN
REFERENCE DISPOSAL SYSTEM
C.H. Kang, J.T. Jeong (Korea)
NUMERICAL MODELING OF IRON-CORROSION AND INTERACTION WITH
BENTONITE IN CLAY FORMATIONS
C. Hansmeier, G. Bracke, B. Reichert (Germany)
IMPACT OF POROSITY CLOGGING ON DIFFUSION RATE: EXPERIMENTS VERSUS
MODELING
I. Fatnassi, S. Savoye, P. Arnoux, P. Gouze, O. Bildstein, V. Detilleux, C. Wittebroodt
(France, Belgium)
COUPLED THCM MODELS OF HEATING AND HYDRATION EXPERIMENTS
PERFORMED ON SAMPLES OF COMPACTED FEBEX BENTONITE IN CONTACT
WITH CONCRETE AND CARBON STEEL
J. Samper, A. Mon, L. Montenegro, J. Cuevas, R. Fernández, M.J. Turrero, E. Torres, A.
Naves (Spain)
QUANTIFYING PLUTONIUM SORPTION AND DESORPTION RATES FROM
MINERAL SURFACES: A NUMERICAL APPROACH TO MODELING BATCH AND
FLOW CELL EXPERIMENTAL DATA
M. Zavarin, B.A. Powell, J. Begg, A.B. Kersting (USA)
ORCHESTRA: A FRAMEWORK FOR INTEGRATING DETAILED REACTIVE
TRANSPORT PROCESSES OF RADIONUCLIDES INTO PERFORMANCE
ASSESSMENT MODELS
J.C.L. Meeussen, E. Rosca-Bocancea J. Grupa, T.J. Schröder (Netherlands)
PC2-1
RADIONUCLIDES TRANSPORT OF PYRO-PROCESSED WASTE IN KOREAN REFERENCE
DISPOSAL SYSTEM
C. H. Kang, J. T. Jeong
Korea Atomic Energy Research Institute, 989-111 Daedeok-daero, Yuseong-Gu,
Daejeon 305-353, Korea
Korea Atomic Energy Research Institute (KAERI) has been developing the KIEP-21 (Korean, Innovative,
Environmentally Friendly, and Proliferation Resistant System for the 21st Century). It is an advanced nuclear
fuel cycle option with a pyro-processing and a SFR. Pyro-processing considered in KAERI adopts three
components; electrolytic reduction, electro-winning and electro-refining, and produces several different
waste streams [1]. Currently two waste streams, metal wastes and ceramic high level wastes are considered
to be disposed. The metal waste contains cladding hulls and insoluble noble metal fission products. The
ceramic high level waste contains a considerable amount of rare earth elements, a small amount of TRU and
Sr because most uranium and TRU are recovered through pyro-processing. To accommodate these wastes,
the advanced Korean reference disposal system (A-KRS) is being developed by KAERI. The current A-KRS
concept is two tier repository. The shallow floor at 200 meters is designed to host ILW by construction of a
silos or tunnels, while the deeper one at 500 meters will take care of HLW by construction of tunnels.
KAERI has also developed a safety assessment program for the disposal of waste streams from pyro-process
by utilizing GoldSim [2].
In this paper a systematic approach to assess the post closure safety of a potential repository for pyro
processed wastes is presented. To evaluate pyro-process advantage over the direct disposal of Spent Nuclear
323

Fuel, the mass balance assuming 4.5 wt% U-235, 45,000 MWD/MTU, 5 years cooling through the entire
pyro-process has been estimated [3, 4]. Probabilistic safety assessment and sensitivity analysis for the input
parameters have been performed to classify important radionuclides and input parameters. Results show that
the important nuclides are Sn-126, C-14, Nb-94, Zr-93, Pu-239, Pu-242, Pa-231, Ra-226, etc. Also the
results of the sensitivity analysis show waste form release mechanism, dissolution rate, hydraulic
conductivity of fractured rock, sorption coefficient, etc. are important parameters in safety assessment.
Figure 1 shows the peak dose’s sensitivity to the waste form dissolution rate. In this figure the peak dose of
fission product is very sensitive to the waste dissolution rate while that of actinide is not sensitive because
the solubility of actinides is low enough and most actinides remain as precipitates in the engineered barrier
system.
This work suggests the importance of developing a waste form of low dissolution rate and more systematic
study for hydraulic conductivity, sorption, etc. in the waste disposal project.
1.0E-03
1.0E-04
Max_Dose [mSv/yr
1.0E-05
1.0E-06
C14
Nb94
Pa231
Ra226
Sn126
Zr93
Pu239
1.0E-07
1.0E-08
1.0E-07 1.0E-06 1.0E-05 1.0E-04
Waste Dis. Rate [1/yr]
Figure 1. Peak Dose’s Sensitivity to Waste Dissolution rate
[1] Yoo, J. H., C. S. Seo, E. H. Kim and H. S. Lee, “A Conceptual Study of Pyroprocessing for Recovering Actinides
from Spent Oxide Fuel,” Nuclear Engineering and Technology 40(7) 581-592 (2008).
[2] User’s Guide, GoldSim Contaminant Transport Module, GoldSim Technology Group (2006).
[3] Kang, C. H. et al., “A Post Closure Safety Assessment for Radioactive Waste from Advanced Nuclear Fuel Cycle,”
KAERI/TR-3989/2010, KAERI (2010).
[4] Hwang, Y. and C. H. Kang, “The development of safety assessment approach and its implication on the advanced
nuclear fuel cycles,” Nuclear Engineering and Technology 42(1) 37-46 (2010).
PC2-2
NUMERICAL MODELING OF IRON-CORROSION AND INTERACTION WITH BENTONITE IN
CLAY FORMATIONS
C. Hansmeier 1) , G. Bracke 1) , B. Reichert 2)
1)
Department of Final Disposal, GRS mbH, Schwertnergasse 1, 50667 Cologne, Germany
2)
Steinmann-Institute, Department Hydrogeology, Nussallee 8, 53115 Bonn, Germany
In deep geological repositories, bentonite is preferred as a possible Engineered Barrier (EB), to isolate the
High Level Waste (HLW) from the biosphere. Over time, the iron canister of the HLW corrodes and
interactions between the EB and the corrosion products of the iron canister modify the properties of the
bentonite. Besides, formation water in the host rock (Opalinus clay) affects the bentonite, too. Therefore this
study focuses on (1) iron corrosion by dissolution of iron and precipitation of iron corrosion products, (2) the
influence of corrosion products on the chemical and physical properties of the bentonite and (3) the influence
of the formation water on the bentonite. The aim is to calculate changes in porewater chemistry, mineral
324

dissolution and precipitation as well as its effects on permeability and porosity changes to assess the long
term behavior. Cation exchange and surface complexation are not considered in this model.
For the calculations the code TOUGHREACT 1.2 [1] is used with PetraSim 5 [2] as user interface.
TOUGHREACT adds reactive geochemistry to the multi-phase flow code TOUGH2 [3]. The modeled
system contains of an iron canister, bentonite barrier and Opalinus clay. The system is modeled as a 1D
model for a timescale of 10.000 years. Due to low permeabilities of both bentonite and clay the mass
transport entirely takes place by diffusion. In a first step, a constant temperature of 25 °C and a completely
water-saturated regime is assumed.
After the closure of the repository the oxygen will be consumed rapidly and anaerobic conditions will
prevail. The iron corrosion of the canister is represented by a cell where Fe(s) is added to the bentonite to
simulate the contact to the iron canister. The corrosion is simulated by dissolution of the iron.
Bentonites are used for engineered barrier systems in high level waste repositories because of their swelling
ability and therefore to retain radionuclides from the biosphere. The MX-80 bentonite consists primarily of
montmorillonite (~70%). Therefore the physically and chemically behavior of the bentonite is predominately
controlled by the transformation of the montmorillonite. The initial composition of the MX-80 porewater and
the composition of the Opalinus clay porewater were taken from Nagra [4] and initial mineral compositions
of the MX-80 bentonite from PSI [5]. Dissolution and precipitation of minerals are mainly kinetically
controlled.
The corrosion products at the cell which represents the iron canister (1) as well as in the bentonite next to the
canister (2) are mainly magnetite and smaller amounts of siderite, goethite as well as some other Fe-rich
minerals. The porosity decreases due to mineral precipitation (e.g. magnetite) which affects the diffusion
process. The corrosion process raises the pH from ~7 to higher values of ~12.
At the contact to the Opalinus clay (3) the bentonite minerals transform under influence of the diffusive
formation water of the host rock. Mainly the montmorillonite is affected by transformation to illite,
vermiculite and other clay minerals. Changes in porosity are as well referable to mineral dissolution and
precipitation.
[1] Xu, T., Sonnenthal, E., Spycher, N., Pruess, K.: TOUGHREACT User’s Guide: A Simulation Program for
Non-isothermal Multiphase reactive Geochemical Transport in Variably Saturated Geologic Media, 2006,
Lawrence Berkeley National Laboratory, University of California: Berkeley, California, USA.
[2] Thunderhead Engineering: PetraSim 5, User Manual, 2005, Manhattan, Kansas, USA.
[3] Pruess, K.: TOUGH2: A general purpose numerical simulator for multiphase fluid flow, 1990, Lawrence
Berkeley National Laboratory, University of California: Berkeley, California, USA.
[4] Nagra: Technical Report 02-05, Project Opalinus Clay, Safety Report, Demonstration of disposal feasibility for
spent fuel, vitrified high-level waste and long-lived intermediate-level waste (Entsorgungsnachweis), 2002,
Wettingen, Switzerland.
[5] Bradbury M. H. & Baeyens, B.: Porewater chemistry in compacted re-saturated MX-80 bentonite: physicochemical
characterization and geochemical modeling, 2002, PSI Bericht Nr. 02-10.
325

PC2-3
IMPACT OF POROSITY CLOGGING ON DIFFUSION RATE: EXPERIMENTS VERSUS
MODELING
I. Fatnassi (1,2) , S. Savoye (1) , P. Arnoux (1) , P. Gouze (2) , O. Bildstein (3) , V. Detilleux (4) , C. Wittebroodt (5)
(1) CEA, DEN, DPC, Laboratory of Radionuclides Migration Measurements and Modeling, F-91191 Gifsur-Yvette,
France. (ikram.fatnassi@cea.fr; sebastien.savoye@cea.fr; patrick.arnoux@cea.fr )
(2) Géosciences Montpellier, UMR 5243, Université de Montpellier 2, F-34095 Montpellier Cedex 5,
France. (philippe.gouze@univ-montp2.fr )
(3) CEA, DEN, DTN, Laboratory for modeling the transfers in the Environment, F-13108 Saint Paul lez
Durance, France. (olivier.bildstein@cea.fr )
(4) Bel V, 148 Walcourtstraat, B-1070 Brussels, Belgium. (valery.detilleux@belv.be)
(5) IRSN, PRP-DGE, SRTG, LETIS F-92260 Fontenay-aux-Roses, France. (charles.wittebroodt@irsn.fr )
Disposal in deep geological clay formations is one of the options chosen by several countries for managing
the fate of high level/intermediate level and long-lived nuclear wastes. The long-term evolution of these deep
repositories should be, for a major part, governed by geochemical processes that could irreversibly modify
the containment properties of the materials used in the multi-barrier system, and especially at their interfaces.
However, addressing the feedback of porosity changes in the long-term simulations coupling chemistry and
transport is still an issue, notably because of the lack of quantitative experiments used to calibrate the current
numerical models [1]. Up to now, even though diffusion is expected to be the main transport process taking
place in the context of the nuclear waste geological disposal, only so-called quantitative experiments carried
out under advective conditions are described in literature for investigating the clogging effect [2, 3].
Tartakovsky et al. [2] and Katz et al. [3] used quasi-two dimensional flow cells packed with quartz sand and
evidenced precipitation of calcium carbonate that clearly impacts the transport properties of their cells.
The present work aims at investigating the capability for reproducing clogging experiments of two
chemistry-transport codes (HYTEC and CRUNCH) that use distinct formulation for taking into account
diffusion change with porosity. The experimental design consists of a modified through-diffusion cell, in
which the cell body has a quasi-two-dimensional shape and is made of plexiglass® for visualizing the
clogging, if any. Different porous media (including sand) have been tested and their initial containment
properties have been determined by means of a through-diffusion method with deuterium-enriched solutions.
Afterwards, a solution containing Na 2 C 2 O 4 has been injected in the outlet reservoir, while another solution
containing CaCl 2 and deuterium has been put into the inlet reservoir for studying the impact of the porosity
clogging by the calcium oxalate precipitation on the diffusive behaviour of the deuterium. First experimental
results clearly show a decrease of the diffusive flux of deuterium in the outlet reservoir related to the porosity
clogging. Comparison between experiments and model predictions is presented and discussed, especially
regarding some key parameters such as kinetics parameterization, pore size distribution, space and time
discretization, for instance.
[1] Kosakowski, G., et al. 2009. Evolution of generic clay/cement interface: first reactive transport calculations utilizing
a Gibbs energy minimization based approach for geochemical calculation. J. Environ. Sci. Sustain. Soc. 3, 41-49.
[2] Tartakovsky, A. M., et al., 2008. Mixing-induced precipitation: Experimental study and multiscale numerical
analysis. Water Resources Research 44, W06S04.
[3] Katz, G.E., et al., 2011. Experimental and modeling investigation of multicomponent reactive transport in porous
media. J. Contam. Hydrol. 120-121, 27-44.
326

PC2-4
COUPLED THCM MODELS OF HEATING AND HYDRATION EXPERIMENTS PERFORMED
ON SAMPLES OF COMPACTED FEBEX BENTONITE IN CONTACT WITH CONCRETE AND
CARBON STEEL
J. Samper (1) , A. Mon (1) , L. Montenegro (1) , J. Cuevas (3) , R. Fernández (3) , M.J. Turrero (2) , E. Torres (2) & A.
Naves (1)
(1)
University of La Coruña, 15071 La Coruña, Spain, jsamper@udc.es
(2)
Centro de Investigaciones Energéticas, Medio Ambientales y Tecnológicas, Av. Complutense 40,
28040 Madrid, Spain
(3)
Universidad Autónoma de Madrid. Facultad de Ciencias, 28049 Madrid, Spain
An Engineered Barrier System (EBS) is foreseen for backfilling and sealing of a radioactive waste
repository. Quantifying the time evolution of the EBS requires the use of coupled thermal (T), hydrodynamic
(H), chemical (C) and mechanical (M) models. Such models are also required to interpret laboratory
experiments of canister corrosion, corrosion-bentonite interactions and concrete-bentonite interactions. Here
we present coupled THCM models of heating and hydration experiments performed at the Ciemat facilities
to: (1) Study the interactions of concrete and bentonite and bentonite and iron under repository conditions
and (2) Analyse how such interactions could affect the bentonite properties. These models were solved with
the INVERSE-FADES-CORE code [1]. Experiments were performed with different sample lengths (25 mm
for 2-I experiments and small corrosion cells and 100 mm for the HB and FB experiments [2]) and
temperatures (25º, 50º, 60º and 100ºC). The small corrosion cells include a 21 mm thick layer of compacted
bentonite and a 4 mm thick layer of Fe powder. The FB3 experiment includes a 87 mm layer of bentonite
and a 13 mm layer of Fe powder in contact with the heater. The so-called double interface experiments, 2-I
experiments, include a 3 mm thick layer of cement mortar which is in contact with the hydration system, a 18
mm thick layer of bentonite and a 2 mm layer of powdered magnetite. The HB experiments, on the other
hand, include a 30 mm layer of concrete which is in contact with the hydration system and a 71.5 mm thick
layer of bentonite. The corrosion cells were hydrated with Grimsel granitic water. The concrete-bentonite
cells were hydrated with a Reference Argillaceous Formation water. Bentonite has an initial porosity of 0.40
and a gravimetric water content of 14 % which corresponds to a saturation degree of 56.8 % and an
approximate suction of 1.29·10 8 Pa. The mortar of the 2I cells has a porosity of 0.3. The concrete of the HB
cell has a porosity of 0.125 and an initial gravimetric water content of 2.6 %.The initial gas pressure is equal
to the atmospheric pressure.
The experiments were modeled with a THCM model using 1D finite element grids. A uniform liquid
pressure was adopted at the injection boundary. Hydrodynamic, thermal and solute transport parameters were
taken from those calibrated previously from previous heating and hydration experiments [3]. The hydration
boundary was modeled with a zero-vertical displacement, a Cauchy condition for the energy equation and a
Neuman condition for solute transport with the solute flux equal to the product of the water flux times the
solute concentration of the inflow water. The effective diffusion coefficient is equal to 2·10 -10 m 2 /s and is the
same for the all the dissolved species except for Cl - which has a diffusion coefficient of 9·10 -11 m 2 /s. The
initial chemical composition of the bentonite was taken from Fernández et al. (2001) [4]. The chemical
system is defined in terms of the following primary species: H 2 O, H + , Ca 2+ , Mg 2+ , Na + , K + , Cl - , SO 2- 4 , HCO - 3 ,
O 2 (aq), Al 3+ and SiO 2 (aq). The model accounts for homogeneous reactions such as acid-base, aqueous
complexation and redox reactions and heterogeneous reactions such as mineral dissolution/precipitation,
cation exchange and surface complexation. The following mineral phases were considered for the concretebentonite
experiments: Calcite, gypsum/anhydrite, cristobalite, and portlandite, brucite, ettringite, sepiolite,
anorthite, C0.8SH and C1.8SH. Calcite, gypsum/anhydrite, cristobalite, magnetite, siderite, goethite, and
Fe(OH) 2 (s) were considered for the bentonite-canister experiments. Cation exchange of Ca 2+, Mg 2+ , Na + , K +
and Fe 2+ was modelled with the Gaines-Thomas convention. Surface complexation reactions take place at the
following types of protolysis sites: S S OH, S W1 OH and S W2 OH. Canister corrosion is modeled with a constant
corrosion rate. Magnetite precipitation is controlled by a kinetic law depending on the mineral saturation
degree. Several hypotheses were tested with the numerical model to explain the observed mineral phases.
327

The THM models were calibrated with the water content and dry density data measured at the end of the
experiments and temperature and relative humidity data measured during the experiment. The parameters of
the reactive transport model were calibrated with aqueous extract data measured at the end of the experiment.
The kinetic parameters of magnetite precipitation were calibrated by using Fe weight content data which are
representative of the precipitation of Fe(OH) 2 (s). For the most part, model results agree well with
experimental data. The results of the bentonite-canister interface models indicate that: 1) The main properties
of the bentonite remain unaltered; 2) There is a sequence of corrosion products, Fe(OH) 2 (s) and magnetite
being the end members [5]; 3) Fe 2+ is sorbed by surface complexation; 4) Fe 2+ cation exchange is less
relevant than Fe 2+ sorption; and 5) Corrosion products penetrate a few mm into the bentonite. The concretebentonite
experiments were modelled with a THMC code (INVERSE-FADES-CORE) and a THC reactive
transport code (CrunchFlow, [6]). The latter model was targeted to identify the relevant secondary minerals
and analyze the alterations observed at a millimetric scale at the concrete-bentonite interface. This model
was then used for long-term chemical and mineralogical predictions of the bentonite-concrete interface.
[1] Zheng, L. & J. Samper (2008). Coupled THMC model of FEBEX mock-up test, Physics and Chemistry of the Earth,
Physics and Chemistry of the Earth, Vol. 33, S486–S498.
[2] Turrero, M.J., M.V. Villar, E. Torres, A. Escribano, J. Cuevas, R. Fernández, A.I. Ruiz, R. Vigil de la Villa, I. de
Soto (2011). Laboratory tests at the interfaces: First results on the dismantling of tests FB3 and HB4. Deliverable
D2.3-3-1. PEBS Project.
[3] Zheng L, J Samper, L Montenegro, AM Fernández (2010). A coupled THMC model of a heating and hydration
laboratory experiment in unsaturated compacted FEBEX bentonite, J of Hydrol, 386 (3-4): 80-94.
[4] Fernández, A., J. Cuevas, & P. Rivas (2001). Pore water chemistry of the FEBEX bentonite. Mater. Res. Soc. Symp.
Proc. 663, 573–588.
[5] Torres, E., J. Peña, M.J. Turrero, P.L. Martín, A. Escribano, M.V. Villar (2008). Different stages on the corrosion
phenomena during the lifetime of a Deep Geological Repository. Eurocorr 2008. Edinburgh, UK, 7-11.
[6] Steefel, C. (2009). CrunchFlow. Software for Modeling Multicomponent Reactive Flow and Transport. User's
manual. Earth Sciences Division. Berkeley, CA.
Acknowledgements: The work presented here was performed within the framework of the PEBS Project (Long-term
Performance of the Engineered Barrier System, EBS) which aims at evaluating the sealing and barrier performance of
the EBS with time, through the development of a comprehensive approach involving experiments, models and the
consideration of the potential impacts on long-term safety functions. PEBS has received funding from the European
Atomic Energy Community’s Seventh Framework Programme (FP7/2007-2011) under grant agreement 232598. This
work was partly funded by ENRESA (Spain) and the Spanish Ministry of Economy and Competitiveness (CICYT
Project CGL2012-36560). We acknowledge the contribution of Juan Carlos Mayor from ENRESA.
PC2-5
QUANTIFYING PLUTONIUM SORPTION AND DESORPTION RATES FROM MINERAL
SURFACES: A NUMERICAL APPROACH TO MODELING BATCH AND FLOW CELL
EXPERIMENTAL DATA
Mavrik Zavarin 1 , Brian A. Powell 2 , James Begg 1 , Annie B. Kersting 1
1. Glenn T. Seaborg Institute, Physical & Life Sciences, Lawrence Livermore National Laboratory, POS Box
808, L-231, Livermore, CA 94550
2. Environmental Engineering and Earth Sciences, Clemson University, 342 Computer Court, Anderson,
South Carolina 29625, United States
It is generally thought that due to its low solubility and high sorptivity, Pu migration in the environment
occurs only when facilitated by transport on particulate matter (i.e., colloidal particles). The ability of
colloids to facilitate Pu transport will be a function of Pu sorption to mineral colloids, colloid filtration rates,
and Pu desorption rates. Experiments to obtain kinetic data for Pu interaction with colloid particles have
identified both rate-limited desorption and redox speciation effects (e.g., [1, 2]). However, most transport
models, which are designed to predict the environmental mobility of Pu, employ only empirical (net)
desorption rates or ignore desorption processes altogether (e.g., [3]). Recent experimental data from Powell
and others [4, 5] suggest that these simplified models cannot adequately predict Pu transport because they do
not capture the coupled processes controlling Pu sorption and desorption.
328

The focus of our effort is to quantify Pu – mineral colloid sorption/desorption rates using a combination of
batch/flow cell experimental techniques and numerical modeling approaches. A simple yet flexible model
was developed to simulate the known Pu redox transformations occurring on the surface of minerals and
quantify the various adsorption, desorption, and redox transformation rates. The model was based on
previous efforts focused on modeling Np(V) sorption behavior on goethite [6]. The original Np(V) model of
Tinnacher et al. [11] was expanded into the following conceptual model:
Each reaction in the model was represented by a formulation with multiple optional parameters. For
example, the mass balance for Pu(V) had the following form
dPu(V) aq
dt
where
= −k 1 × Pu(V) n′ aq × 1 − Q φ
K + F V (C in − C out )
Q = Pu(V) s1
n"
Pu(V) n′ and K = k 1 V
.
aq
k −1 m
F, V, and m are the water flux, flow cell volume , and solid mass, respectively, C in and C out are the inflow
and outflow Pu(V) concentrations, and n’ and n” are the Freundlich and empirical exponents. The ψ term
was added to account for observed differences in sorption and desorption kinetics. The ψ term is the only
parameter that was allowed to change depending on whether the overall reaction is a forward or reverse
reaction. Thus, while the same equilibrium may be reached from both the adsorption and desorption
directions, the rate of approach to equilibrium may differ substantially. The code was written in visual Basic
and coupled to the PEST minimization routine [7].
Using this approach, we are fitting a number of Pu batch and flow cell experiments performed in our
laboratory. To date, the focus has been on the minerals goethite and montmorillonite. In the case of Pu(V)
interactions with goethite, we were able to fit our pH 8 adsorption and desorption flow cell data, including
the transient stop flow events, using this model (Figure 1). However, the fast adsorption in the first few pore
volumes of the experiment could not be accounted for. Importantly, the model fit was found not to be
sensitive to most of the optional model parameters. Instead, a good fit could be achieved while ignoring all
Freundlich and empirical exponents except in the case of Pu(V) adsorption. However a kinetic hysteresis
effect, parameterized as ψ, was essential in fitting Pu(IV) adsorption and surface mediated reduction
reactions. This led to a numerical model not unlike the earlier Np(V) model [11] and consistent with batch
equilibrium Kds, surface mediated reduction rates reported in the literature, and measured aqueous and
sorbed Pu oxidation states.
329

Figure 1. Flow cell Pu(V) adsorption and desorption breakthrough data and model fit.
Prepared by LLNL under Contract DE-AC52-07NA27344.
[1] N. Lu, et al., Radiochim. Acta 83, 167 (1998).
[2] W. Runde, et al., Appl. Geoch. 17, 837 (2002).
[3] D.A. Pickett, Nucl. Sci. Eng. 151, 114 (2005).
[4] D.I. Kaplan, et al. Env. Sci. Tech. 38, 5053 (2004).
[5] B.A. Powell, et al., Env. Sci. Tech. 39, 2107 (2005).
[6] R.M. Tinnacher, et al., Geochim. Cosmochim. Acta 75, 6584 (2011).
[7] J. Doherty, Watermark Numerical Computing (2003).
PC2-6
ORCHESTRA: A FRAMEWORK FOR INTEGRATING DETAILED REACTIVE TRANSPORT
PROCESSES OF RADIONUCLIDES INTO PERFORMANCE ASSESSMENT MODELS
Johannes C.L. Meeussen * , Ecaterina Rosca-Bocancea, Jacques Grupa and Thomas J. Schröder
1) Nuclear Research and Consultancy Group(NRG) Westerduinweg 3 Petten, 1755 ZG, The Netherlands.
The estimation of long term migration rates of radionuclides in soil, host rock as well as cementitious waste
and concrete barriers plays a central role in the evaluation and design of geological radioactive waste
disposal. Migration rates of radionuclides in such reactive porous materials are determined by the
distribution of ions over the mobile aqueous phase and the immobile solid phase. At a small scale this
distribution is determined by precipitation of minerals, adsorption of ions to mineral (clay, Al- and Feoxides),
organic surfaces, and adsorption to colloidal particles. In most cases the multicomponent nature of
such interactions results in a strong dependence of single element behaviour on actual chemical conditions.
This complicates the translation of short-term experimental results to (very) long-term field conditions.
Ideally, a long term migration model should contain a description of the main chemical interaction processes
for the macro-elements as well as the trace elements[1].
Over recent years significant progress is made in the development of insight into interactions of ions with the
most relevant reactive phases / surfaces in soil, e.g. the different oxide and organic matter surfaces, dissolved
organic carbon, including their mutual chemical interactions in a consistent thermodynamic way. The
combined CD-MUSIC models for oxides, the NICA-DONNAN model for organic matter, and the LCD
model are a clear demonstration of this[2].
330

For PA (performance assessment) purposes, however, it is generally not practical to compose a model that
includes complete detailed process models. This may lead to long calculation times and huge input parameter
demand. For such purposes a model should be as detailed as necessary but as simple as possible. Proper
evaluation of the effect of model simplifications, however, can only be done by comparing the detailed with
simplified models.
In this work we demonstrate the application of the ORCHESTRA model framework[3] that not only can
accommodate detailed state-of-the art radio-nuclide chemical interaction and transport models, but is also
flexible enough to act as a integrating framework for implementing PA models. This is demonstrated by
discussing the PA model as implemented within the context of the Dutch OPERA research programme[4].
The following properties of ORCHESTRA make it particularly suitable as integrating framework:
- User definable (programmable) modules
- Predefined and tested modules for complex aqueous, surface and cement chemistry
- Predefined and tested modules for (gas) diffusion and convection
- User definable interfaces and links between different sub-models
- Possibility to interchange detailed sub-models with simplified alternatives
- Numerically efficient
- Platform independent (runs on windows, linux, OSX)
[1] J.J. Dijkstra, et al., Environ. Sci. Technol. 43 , 6196-6201 (2009).
[2] Weng et al., Environ. Sci. Technol. 42: 8747-8752 (2008)
[3] J.C.L. Meeussen, Environmental Science & Technology, 37, 1175-1182 (2003).
[4] OPERA, Dutch research programme on radioactive waste disposal. 2012-2016.
http://www.covra.nl/infocentrum/opera
PC5
PC5-1
PC5-2
PC5-3
PC5-4
PC5-5
SAFETY ASSESSMENT AND REPOSITORY
CONCEPTS
DECOMPOSITION OF U(VI)-ARSENAZO III COMPLEX BY FENTON REACTION
INDUCED BY GAMMA IRRADIATION
Z. Zheng, M.L. Kang, C.L. Liu, T. Chen, B. Grambow, L. Duro, T. Suzuki-Muresan (China,
France, Spain)
THE BIGRAD CONSORTIUM - SPECTROSCOPIC STUDIES OF THE TRANSFORMATION
OF MINERALS IN SANDSTONE UNDER HYPERALKALINE CONDITIONS
A.F. Seliman, S.A. Banwart, M.E. Romero-Gonzales (UK)
RADIATION INDUCED CORROSION OF COPPER IN DEEP GEOLOGICAL
REPOSITORIES FOR SPENT NUCLEAR FUEL
Å. Björkbacka, S. Hosseinpour, M.C. Johnson, C. Leygraf, M. Jonsson (Sweden)
GEOCHEMICAL MODEL OF CRUSHED GRANITE DISSOLUTION AT 70°C
I. Brusky, J. Sembera (Czech Republic)
INVENTORY OF A DEEP GEOLOGICAL RADIOACTIVE WASTE REPOSITORY IN A
SALT FORMATION
H. Seher, H. Fischer, J. Larue (Germany)
331

PC5-1
DECOMPOSITION OF U(VI)-ARSENAZO III COMPLEX BY FENTON REACTION
INDUCED BY GAMMA IRRADIATION
Z. Zheng 1 , M.L. Kang 1 , C.L. Liu 1, *, T. Chen 1,2 , Bernd Grambow 3 , Lara Duro 4 , Tomo Suzuki-Muresan 3
1
Beijing National Laboratory for Molecular Science, Radiochemistry & Radiation Chemistry Key
Laboratory for Fundamental Science, College of Chemistry and Molecular Engineering, Peking University,
Beijing 100871, China
2
North China Electric Power University, Beijing 102206, China
3 SUBATECH, Unité Mixte de Recherche 6457, Ecole des Mines de Nantes, CNRS/IN2P3, Université de
Nantes, 4 rue Alfred Kastler, BP 20722, 44307 Nantes cedex 03, France
4 AMPHOS21,P. Garcia Ifaria49-51,Barcelona, E-08019, Spain
*Corresponding author liucl@pku.edu.cn
Water radiolysis can produce hydrogen peroxide.[1] The results in this study demonstrate that
irradiation of U(VI) and iron (Fe 2+ or Fe 3+ ) mixed solution or introduction of iron into a gamma irradiated
U(VI) solution can give rise to Fenton reaction[2], which can destruct the subsequently added Arsenazo III,
leading to the sharp decrease of the UV-vis absorbance of U(VI)-Arsenazo III complex. [3] The percentage
of absorbance decrease is iron content- and dose-dependent. The adding sequence of the iron (before or after
the irradiation process) and the type of iron (ferrous or ferric) [4] both has apparent impact on the
decomposition of U(VI)-Arsenazo III complex. Furthermore, the Fenton reagent or H 2 O 2 induced by gamma
irradiation is fairly stable. Even if the irradiated solution has been aged for one hundred days in the air, there
is no discernible change of its oxidation ability on the subsequently added Arsenazo III. Through the FT-MS,
destroying of the U(VI)-Arsenazo III structure by irradiation and Fenton process was observed. This study
advances some matters needing attention on the application of Arsenazo-III in radioanalytical chemistry.
Absorbance
0.6
0.5
0.4
0.3
0.2
irradiated with Fe 2+
Concentration (mmol/L)
1.0
0.8
0.6
0.4
0.2
Fe 2+
Total Fe
Fe 3+
irradiated without Fe 2+ 0 200 400 600 800 1000
0.1
0.0
0 200 400 600 800 1000 1200 1400 1600 1800
Dose, Gy
Dose, Gy
Fig.1 Dose dependence of the absorbance (655 nm) of U(VI)-Arsenazo III complex (left panel) and the Fe 2+
content in the solution (right panel). During the irradiation process, the dose rate was kept at 30 Gy/min.
[1] I.G. Draganic, Nenadovi. Mt, Z.D. Draganic, RADIOLYSIS OF HCOOH + O2 AT PH 1.3-13 AND YIELDS OF
PRIMARY PRODUCTS IN GAMMA RADIOLYSIS OF WATER, Journal of Physical Chemistry, 73 (1969) 2564-&.
[2] E. Khan, W. Wirojanagud, N. Sermsai, Effects of iron type in Fenton reaction on mineralization and
biodegradability enhancement of hazardous organic compounds, Journal of Hazardous Materials., 161 (2009) 1024-
1034.
[3] S.B. Savvin, The use of Arsenazo III for the photometric determination of Th, U, Zr, Pa, Sc, and rare-earth elements,
Industrial Laboratory, 29 (1963) 113-121.
[4] E. Viollier, P.W. Inglett, K. Hunter, A.N. Roychoudhury, P. Van Cappellen, The ferrozine method revisited:
Fe(II)/Fe(III) determination in natural waters, Applied Geochemistry, 15 (2000) 785-790.
PC5-2
332

SPECTROSCOPIC STUDIES OF THE TRANSFORMATION OF MINERALS IN SANDSTONE
UNDER HYPERALKALINE CONDITIONS
A.F. Seliman (1) , S.A. Banwart (1) , M.E. Romero-González (1)
(1) Cell-Mineral Research Centre, Kroto Research Institute. The University of Sheffield. Sheffield S3 7HQ –
United Kingdom
The geological disposal of radioactive waste requires a detailed understanding of the critical processes
affecting geochemical transformations at laboratory and field scale. The interfaces between storage
containers, repository and geological substrata can be viewed as a series of continuous barriers with varying
porosity and chemical composition that governs flow and physical-chemical transformation of radionuclides
during their long term transport. These barriers will be transformed over time when interacting with fluids
originating both from inside the facility and the surrounding natural groundwater. The pH of the solution is
predicted to reach values of 13 in the first instance and equilibrate at pH 10 in the long term. In the UK,
sandstone has been proposed as a candidate geological environment for disposal of radioactive material. The
aim of this work is to investigate the transformation of minerals present in sandstone under hyperalkaline
conditions representative of geological disposal facilities. This approach provided spatial and temporal
understanding of dissolution reactions and potential pathways for transport of radionuclides in the context of
reactive transport modelling. The experimental set up consist of a 2D flow cell packed with Sherwood
sandstone that is in contact with young cement leachate (YCL) at pH 13. The flow rate was kept at 0.7 mL h-
1 for approximately 37 days. The flow cell was equilibrated with synthetic ground water before being in
contact with YCL solution at pH 13. Samples were collected for analysis of pH, Na + , K + , Ca +2 , SiO 2 and Al +3
as main indicators of mineral changes in the sandstone over time. A conservative tracer (Br - ) was also
quantified. At the end of the experiments, the flow cell were destructively sampled and analysed using XRD,
XPS, SEM and TEM in order to establish changes in mineralogy and formation of secondary minerals. The
obtained breakthrough curves for pH showed that equilibrium to pH 13 (input value) is reached after 2 pore
volumes and maintained throughout the duration of the experiment. The breakthrough curve for Na + and Br -
showed a conservative behaviour during the experiment. Dissolution of K + was observed but not to a great
extent. The dissolution of SiO 2 and Al 3+ was significant (Figure 1) and the breakthrough coincides with the
stabilisation of Ca 2+ in the output solution as evident from its breakthrough curve.
Fig 1. Breakthrough curves for SiO 2 and Al 3+ from the dissolution of sandstone in a 2D flow cell experiment.
The spectroscopic and microscopic analyses revealed the formation of Ca-silicates. This result corresponds
with the dissolution and precipitation profiles obtained from the aqueous phase analyses. The formation of
Na-Al-silicates was also evident from the XRD results indicating the potential precipitation of other
compounds due to the high alkaline experimental conditions. The dissolution and formation of secondary
minerals occurred during a short period of time, this will have an impact on change of porosity and
permeability of sandstone materials due to the formation of Ca-silicates. The results presented here are a first
step towards modelling of dissolution and precipitation of minerals in sandstone, reactive transport
modelling and studies of transport of radionuclides under hyperalkaline conditions are needed to fully
understand the implications of sandstone transformations around a geological disposal facility.
333

PC5-3
RADIATION INDUCED CORROSION OF COPPER IN DEEP GEOLOGICAL REPOSITORIES
FOR SPENT NUCLEAR FUEL
Åsa Björkbacka 1 , Saman Hosseinpour 2 , Magnus C. Johnson 2 , Christofer Leygraf 2 , Mats Jonsson 1
1 Division of Applied Physical Chemistry, KTH Royal Institute of Technology, School of Chemical Science
and Engineering, SE-100 44 Stockholm, Sweden.
Fax: 0046(0)8 790 87 72; Tel: 0046(0)8 790 91 23; E-mail: matsj@kth.se
2 Division of Surface and Corrosion Science, KTH Royal Institute of Technology, School of Chemical
Science and Engineering, SE-100 44 Stockholm, Sweden.
Fax: 0046(0)8 820 82 84; Tel: 0046(0)8 790 64 68; E-mail: chrisl@kth.se
One of the most developed concepts of long term storage of spent nuclear fuel is the Swedish KBS-3
method. 1 The spent nuclear fuel will be encapsulated in copper canisters with a cast iron insert, then placed
one by one in tunnels 500 meters below ground and embedded in bentonite clay. 1 Water in the bentonite in
close proximity to the outer canister surface will undergo water radiolysis where two oxidants, hydrogen
peroxide (H 2 O 2 ) and the hydroxyl radical (HO∙), 2 have significantly higher standard reduction potentials (E ◦ )
than copper 3-5 and are therefore thermodynamically capable of initiating corrosion of the canisters.
Despite the importance of the process of radiation induced corrosion of copper relatively few studies have
been reported. In this work the effect of gamma irradiation, total gamma dose and dose rate on radiation
induced corrosion of copper in anoxic pure water have been studied experimentally. Copper samples were
submerged in water and exposed to a series of total doses using dose rates of 80, 370 and 770 Gy·h -1 .
Unirradiated samples were used as reference samples throughout. The copper surfaces were examined
qualitatively using infrared absorbance spectrometry (IRAS), scanning electron microscopy with energy
dispersive spectroscopy (SEM-EDS) and x-ray photoelectron spectroscopy (XPS) and quantitatively using
cathodic reduction and atomic force microscopy (AFM). The dissolution of copper after irradiation was
measured using ICP-AES. The experiments show that the irradiated samples are more corroded than
corresponding reference samples. The dissolution as well as the oxide layer thickness increases after
irradiations. Surface examinations also reveal local corrosion features. Based on numerical simulations the
influence of aqueous radiation chemistry on the corrosion process was evaluated. Interestingly, the
evaluation indicates that aqueous radiation chemistry is not the only process driving the corrosion of copper
in these systems.
1. Long-term safety for the final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site project. TR-
11-01, Swedish Nuclear Fuel and Waste Management Co., Stockholm, 2011.
2. G. R. Choppin, J. O. Liljenzin and J. Rydberg, Radiochemistry and nuclear chemistry, 2nd edn., Butterworth-
Heinemann, Oxford, U.K., 1995.
3. F. A. Cotton and G. Wilkinson, Advanced inorganic chemistry, 3rd edn., Wiley Interscience, New York, 1972.
4. W. H. Koppenol and J. F. Liebman, Journal of Physical Chemistry, 1984, 88, 99-101.
5. R. R. Conry, in Encyclopedia of Inorganic Chemistry, ed. R. B. King, Wiley, New York, Edition edn., 2005, pp.
940-958.
334

PC5-4
GEOCHEMICAL MODEL OF CRUSHED GRANITE DISSOLUTION AT 70°C
I. Brusky (1) , J. Sembera (2)
(1) Institute for Nanomaterials, Advanced Technology and Innovation
Technical University of Liberec, Studentska 1402/2 Liberec 461 17 – Czech Republic
(2)
Faculty of Mechatronics, Informatics and Interdisciplinary Studies,
Technical University of Liberec, Studentska 1402/2 Liberec 461 17 – Czech Republic
Granitic rock has been proposed as a potential host rock for deep radioactive waste repository in the Czech
Republic. For modelling of behaviour of radionuclides in the water solution need to know the composition of
the water solution. Our aim is to model the water solution in granite evolving in time. Experiments of
dissolution of crushed granite in deionized water were done. We simulate chemical evolution of water
solution which is in contact with crushed granite and with atmosphere. We are focusing on few governing
chemical interactions between water and granite.
Experiments were done as follows. In one reactor, there are 60 g of crushed granite and 600 ml of water.
They interact during a certain time interval. The grain size of the crushed granite is between 0.125 and 0.25
mm. Deionized water was used for experiments. Water stays in contact with atmosphere. The reactor
temperature is 70 °C. Chemical experiments and analyses are well described in the technical report [1].
This experiment was done five times. Each experiment took another time. Times were the following: one
day, five days, ten days, twenty days and forty days. Results of these experiments are interpreted as a time
development of one experiment. This interpretation introduces uncertainty in time development result since
each experiment was done with another 60 gram sample of the same fraction.
The chemical analysis of crushed granite was done. Reaction surface of crushed granite was analyzed by
B.E.T. method. We have to suppose that all minerals have the same reaction surface 510 m 2 /kg. For each
experiment, pH and concentrations of Si, Al, Ca, Na, K, and Cl were measured.
We created a model in The Geochemist’s Workbench 8.0.12 (GWB) [2] using the thermodynamical database
(TDB) named thermo.dat [3]. This software automatically considers all possible equilibrium chemical
reactions that can appear in the system if the used TDB includes the relevant data.
The model does not begin in the zero time but in the time of one day because during the first day of
experiment very small pieces of minerals from crushing dissolved to deionized water very rapidly. This
phenomenon is hard to describe and harder to simulate. The initial water composition used in the model is
the result of chemical analysis of the water taken after one day experiment. The first step in modeling was to
solve the equilibrium of the water with atmosphere (especially CO 2 ) under temperature 70 °C. We had to
solve this equilibrium for all water samples to be able to compare pH.
In the model, there are three kinetic reactions. They are dissolutions of albite, anorthite, and quartz. We
thought about dissolution of orthoclase, too. We evaluated chemical analyses and it showed that all water
solutions are oversaturated by orthoclase. And our past work shows this phenomenon, too [4].
The reaction rate of the r-th reaction [mol/l/s] is given by the following equation:
( r)
( r) ( r) ( r) ( r)
⎛ Q ⎞
r = m S k ⎜1−
( r)
⎟
⎝ K ⎠
where m (r) means the weight of the dissolved mineral, S (r) [m 2 /kg] is its specific surface area, k (r) [mol/l/m 2 /s]
( r )
Q
( r )
stands for the kinetic constant and is the saturation ratio. Here Q is the activity product of the r-th
( r )
K
reaction and K (r) is the equilibrium constant of the r-th reaction.
335

Precipitated minerals are limited only on Beidelite-K ( K0.33Al2.33Si3.67O 10(OH) 2
). Kinetic constants are
unknown in the model, they are the calibrated parameters. We compare them with values form scientific
literature, see Table 1.
Table 1. Kinetic constants from literature and calibrated in model, temperature 70°C
Kinetic constant [mol / m 2 / s]
Mineral Researched Model
Albite 7.4981E-12 [5] 4.00E-12
Anorthite 1.9467E-09 [5] 1.50E-11
Quartz 5.3269E-12 [5] 1.00E-11
The model outputs have good agreement with chemical analyses, see Figure 1. Calibrated kinetic constants
are compared with values from scientific literature Table 1.
Figure 1. Evolution of pH, aluminium and dissolve silicon dioxide. Model results are represented by a line
and measurements are pictured as circles.
The article was realized under the project TA01021331 “Simulation tools for the evolution prediction of
THC processes and their influence on the migration of radionuclides in geosphere”.
[1] P. Martinec, et. al (2011) “Supplement number 2 – Granite from Liberec, quarry Ruprechtice.” [in Czech], Project
SÚRAO-D 2011/ÚGN
[2] C. M. Bethke (1996). “Geochemical Reaction Modeling Oxford” University Press
[3] http://www.gwb.com/thermo.htm
[4] I. Brusky and J. Sembera (2012) “Using a geochemical model for interpretation of a laboratory experiment”
Proceedings of the 5 th WSEAS International Conference on Environment and Geological Science and Engineering,
Vienna, ISBN 978-1-61804135-7
[5] H. Yasuhara et. al (2011) “Temporal alteration of fracture permeability in granite under hydrothermal conditions
and its interpretation by coupled chemo-mechanical model” Applied Geochemistry, Vol. 26, Issue 12, pp. 2074–2088
336

PC5-5
Inventory of a deep geological radioactive waste repository in a salt formation
H. Seher, H. Fischer, J. Larue
Department of Final Disposal, GRS mbH, Schwertnergasse 1, 50667 Cologne, Germany
Rock salt is one of the possible host rock formations for the disposal of high-level radioactive waste in
Germany. The Preliminary Safety Analysis of the Gorleben Site (Vorläufige Sicherheitsanalyse Gorleben,
VSG) evaluated the long-term safety of a hypothetical repository in the salt dome of Gorleben, Germany.
Within the project the radionuclide release and transport is numerically modelled using the code MARNIE
[1] for the transport in the liquid phase and the code TOUGH2 [2] for the transport both in the liquid and the
gas phase.
During the project, several alternative repository concepts were developed [3, 4]. The so called AB1 concept
foresees the emplacement of high active heat-generating waste in 12 emplacement fields which are located
east of the shafts and 3 emplacement fields for negligible heat generating waste located west of the shafts.
All emplacement fields were positioned between two main drifts. The sequence of waste emplacement
(sequence of loading) in the different emplacement field as described in [3] was only considered in the
calculations with MARNIE.
For the numerical modelling of radionuclide transport an inventory of relevant radionuclides was compiled
for fission and activation products as wells as for radionuclides from the Thorium, Neptunium, Uranium and
Actinium decay chains [5]. Waste forms were spent nuclear fuel of research reactors and power plants (PWR,
BWR, WWER), cladding materials, reprocessing waste (CSD-B, CSD-C and CSD-V), graphite based waste,
other mixed waste and uranium tails. For each waste form a specific container were foreseen (e.g. spent fuel:
POLLUX ® -10, CSD-C: POLLUX ® -9, cladding materials: cast iron container type II).
For all waste forms a typical nuclide specific inventory was compiled in [6]. The inventory, e.g. for spent
fuel, was classified according to decay storage age-groups, each spanning 5 years, by [6]. The inventory was
quantified in Bq per tons of heavy metal (t HM ) of spent nuclear fuel. The POLLUX ® -10-container is used for
PWR, BWR and WWER spent fuel. However, the inventory in Bq for a typical container was derived for
PWR spent fuel which overestimates the inventory for BWR and WWER spent fuel.
In order to determine the emplaced activity for a given activity by [6] in [Bq∙t HM -1 ] the total mass of heavy
metal in the fuel per POLLUX ® -10-container has to be estimated. The mean load of heavy metal in the spent
fuel of one POLLUX ® -10-container was assumed to be 5.288 t HM , which sums up to a total mass of
11210 t HM for all 2120 POLLUX ® -10-containers in the repository.
The MARNIE calculation started at the time when the first emplacement field is loaded and not after closure
of the final repository. The remaining emplacement fields were filled successively according to the loading
sequence within 40 years [3]. The inventory, radioactive decay and compaction of salt grit were therefore
specific for each emplacement field.
The integration of the loading sequence in the calculations improves the activity distribution at the time of
closure. The calculated activity inventory was used as input for subsequent calculations with TOUGH2. The
total activity in the repository was 4.46∙10 19 Bq at the time of closure.
[1] Martens, K.-H., et al., Beschreibung des Rechenprogrammes MARNIE, GRS-A-3027, 2002.
[2] Pruess, K., TOUGH2: A general purpose numerical simulator for multiphase fluid flow, Report LBL-29400, 1990.
[3] Bollingerfehr, W., et al., Endlagerkonzepte, GRS-272, 2011.
[4] Bollingerfehr, W., et al., Endlagerauslegung und -optimierung, GRS-281, 2012.
[5] Larue, J., et al., Radiologische Konsequenzenanalyse, GRS-289, 2013
[6] Peiffer, F., et al., Abfallspezifikation und Mengengerüst. Basis Ausstieg aus der Kernenergienutzung (Juli 2011),
GRS-278, 2011
337

PFS
PFS-1
PFS-2
PFS-3
ENVIRONMENTAL BEHAVIOUR OF RADIO-
NUCLIDES AFTER THE FUKUSHIMA ACCIDENT
FIXATION AND DYNAMICS OF RADIOACTIVE CESIUM IN FUKUSHIMA SOILS
T. Saito , H. Makino , S. Tanaka (Japan)
QUANTIFICATION OF THE NATURAL REDISTRIBUTION OF RADIOCAESIUM IN
THE AFTERMATH OF THE FUKUSHIMA DAI-ICHI NUCLEAR ACCIDENT
H. Sato, K. Iijima, T. Niizato, S. Nakayama, S.M.L. Hardie,
I.G. McKinley (Japan, Switzerland)
DEVELOPMENT OF A BOX-MODELLING APPROACH TO INTEGRATE
CONTAMINATED SITE UNDERSTANDING IN FUKUSHIMA
S.M.L. Hardie, I.G. McKinley, T.M. Beattie, E.M. Scourse, L. Klein (Switzerland, UK)
PFS-1
FIXATION AND ITS DYNAMCIS OF RADIOACTIVE CESIUM IN FUKUSHIMA SOILS
T. Saito (1) , H. Makino (2) , S. Tanaka (2)
(1) Nuclear Professional School, School of Engineering, The University of Tokyo, 2-22 Shirakata Shirane,
Tokai-mura, Ibaraki, 319-1188, JAPAN
(2) Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo,
7-3-1 Hongo, Bunkyo-ku, Tokyo, 113-8656, JAPAN
Decontamination and radiation protection of contaminated environments by radionuclides released from the
accident of the Fukushima Daiichi nuclear plant (F1 NPP) are on-going challenges for the recovery of local
societies from the accident. Radioactive cesium, 137 Cs (T 1/2 = 30.10 a) and 134 Cs (T 1/2 = 2.07 a), are major
contaminants in large areas around the F1 NPP. Understanding the speciation and migration of radioactive
Cs in soils is essential for the challenges. The objective of this study is to reveal chemodynamics of 137 Cs in
soils through sequential extraction and size fractionation, supplemented by chemical and mineralogical
analyses of soils. The distributions of 137 Cs over the different fractions were compared with those of 133 Cs,
the stable isotope of cesium, which had existed in soils before the accident; the variation of the isotopic ratio
enables us to assess long-term dynamics of the fixation process of Cs in the soils, as the both isotope
chemically behave in the same way and the only plausible explanation for such a variation is the different
residence time of the isotopes in the soils. Eleven soils from different locations were tested for better
statistics of the outcomes.
Soil samples were collected on April 20th, 2011, in the eastern area of Fukushima Prefecture. The details of
the sampling and inventories of 137 Cs and other radionuclides are found in the literature [1]. Sequential
extractions of 137 Cs and 133 Cs, based on Hou et al. [2], were performed on the finely ground and 2-mm sieved
soil samples, from which the six fractions were obtained: (i) water soluble fraction, (ii) ion exchangeable
fraction, (iii) amorphous iron (hydr)oxide fraction, (iv) organic fraction, (v) acid digestible fraction, and (vi)
residue. Sub-samples after the decomposition of organic materials by H 2 O 2 were size-fractionated by wetsieving
and sedimentation to the four mineral fractions: (M-i) coarse sand (CS), (M-ii) fine sand (FS), (M-iii)
silt, and (M-iv) clay. The amounts of 137 Cs in the chemical and mineral fractions were determined by γ-ray
counting after capturing 137 Cs to Cu 2+ -substituted potassium ferrocyanide supporting resins [3] and those of
133 Cs in the chemical fractions (i ~ vi) by ICP-MS.
Most of 137 Cs were found in in the acid digestible fraction and residue, as is shown in Figure 1. Among the
mineral fractions, 137 Cs tended to be more and more associated with smaller fractions, silt and clay (Figure
2), where illite and mica family were abundant. The K d values of 137 Cs calculated for these fractions were
similar to those reported for illite in laboratory experiments [4]. Thus, it was concluded that micaceous
minerals rich in silt and clay fractions were responsible for the fixation of 137 Cs in the soils. The isotope
338

atios of 137 Cs/ 133 Cs were found to be smaller for the fractions (vi) to (vi), suggesting the presence of
exchange processes with slow kinetics between these fractions and more labile fractions (the fractions (i) ~
(iii)). Model calculations with multi-site ion exchange models further confirmed the presence of at least three
sites for 137 Cs in the soils: labile, less-labile, and inert sites. The labile sites could correspond to regular
exchange sites; meanwhile, less-labile and inert sites could be so-called frayed edge sites with different
exchangeability by competing ions like NH 4 + . Correlation analyses of the 137 Cs and 137 Cs/ 133 Cs distributions
with physicochemical properties of the soils will be demonstrated to clarify the chemodynamics of
radioactive cesium in soils.
Figure. 1. Box-plots of the distribution of 137 Cs over the different chemical fractions.
Figure. 2. Box-plots of the distribution of 137 Cs over the different mineral fractions.
[1] T. Fujiwara, et al., Radioact. 113, 37 (2012).
[2] X. Hou, et al., Sci. Total Environ., 308, 97 (2003).
[3] K. Tanihara, J. Radioanal. Nucl. Chem., 201, 509 (1995).
[4] N. J. Comans, et al., Geochim. Cosmochim. Acta 55, 433 (1991).
339

PFS-2
QUANTIFICATION OF THE NATURAL REDISTRIBUTION OF RADIOCAESIUM IN THE
AFTERMATH OF THE FUKUSHIMA DAI-ICHI NUCLEAR ACCIDENT
H. Sato 1 , K. Iijima 1 , T. Niizato 1 , S. Nakayama 1 , S. M. L. Hardie 2 , and I. G. McKinley 2
1 Japan Atomic Energy Agency, Fukushima, Japan
2 MCM Consulting, Baden-Dättwil, Switzerland
The accident at the Fukushima Dai-ichi Nuclear Power Plant (1F-NPP) in March 2011, led to the release of
volatile radionuclides which were deposited on the surrounding environment (soils, forests, residential land,
etc.) within the Fukushima Prefecture. The inventory of radionuclides released by the beginning of April
2011 was between 1.2-1.5 x 10 16 Bq for both 137 Cs and 134 Cs and 1.5 x 10 17 Bq for and 131 I [1]. After the
decay of short-lived 131 I, radiocaesium is now the main contributor to radiological dose and therefore
understanding its transport behaviour is essential to assess potential doses in the future – both in areas being
remediated and in those where only natural processes change distribution of dose rates.
In November 2012, JAEA began a research project entitled the “Long-Term Assessment of Transport of
Radioactive Contaminant in the Environment of Fukushima” (F-TRACE). This project has 3 objectives; (1)
to elucidate the transport behaviour of radiocaesium from contaminated mountain forests, through the
biosphere to lacustrine and coastal sinks, (2) to develop an appropriate dose evaluation system, and (3) to
implement a comprehensive system for predicting radiocaesium transport and the impact of various countermeasures
to limit resultant doses. The project involves an integrated assessment of all physical and chemical
processes contributing to the mobilisation/immobilisation of radiocaesium within the Fukushima evacuated
zone and the wider prefecture, which requires coordination of field investigations, a laboratory programme
and modelling studies that cover a wide range of technical disciplines.
The research programme includes 9 study topics; (1) forest ecosystems, (2) the fluvial environment (3) the
lacustrine environment, (4) the estuarine environment, (5) laboratory investigations of Cs transport, (6)
model analysis (simulation of Cs transport), (7) lichen biomonitors, (8) development of techniques to restrict
Cs transport and (9) follow-up monitoring. In evaluation of Cs transport from forest sources to estuary sinks,
information and data obtained from research areas (1) to (5) are directly used as input data for model
analyses. The lichen studies are carried out to assess application as a biomonitor of the initial distribution of
radiocaesium fallout and radiocaesium dynamics in the surface environment. The development of techniques
to control Cs transport focuses on interception of radiocaesium adhering to colloids and suspended solids in
fluvial systems. Finally, follow-up monitoring is performed to track changes in dose rate so that these can be
compared with model predictions and the causal mechanisms elucidated for any divergences observed.
Figure 1 shows the field investigation areas used to develop the radiocaesium transport model. Forest
ecosystem studies have been performed within 2 areas (Kawamata Town and Kawauchi Village), involving
analysis of geomorphology/topography and soil and vegetation analyses. Additionally, monitoring of surface
runoff and downward movement of soil (bioturbation) in field observation plots is ongoing.
Five rivers were studied and characterised in terms of cross-section shape, flow rate, turbidity, suspended
solids and sediments present as a function of time. Investigations have also been carried out at Ogi dam
(Kawauchi Village), including sounding and acoustic profiling for 3 dimensional determination of
bathymetry. Sampling of water and bottom sediments (using sediment corers) has been carried out to obtain
concentration profiles of radiocaesium, used to analyse sedimentation rate as a function of time and identify
any cases of re-suspension (particularly associated with extreme rainfall associated with typhoons).
The estuaries of these 5 rivers are also studied to determine the behaviour of soil particle transport as a
function of salinity and resultant radiocaesium depth profiles in sediment. An important factor here is
assessing geographical change of the river mouth as, from investigations to date, it has been seen that river
mouth-bars are reshaped as part of the recovery from changes caused by the massive Tohoku earthquake and
tsunami.
340

Supporting laboratory studies have focused on mechanisms of Cs uptake by solids and have included
mineralogical analyses, determination of particle size distribution and sorption/desorption isotherms on
representative clay, silt, sand and gravel samples.
Numerical simulation of Cs transport from forest to estuary will be carried out using two or more models,
wherever possible based on existing codes. Initially, Cs removal from forest (land erosion) will be analysed
using USLE (Universal Soil Loss Equation) while Cs transport in river networks will be analysed using
TODAM, which was developed by Pacific Northwest National Laboratory (PNNL). Cs transport in estuaries
can be analysed using FLESCOT, ROMS and Nays2D. All field and laboratory studies are planned to
provide input data and refine the assumptions forming the basis of these models, allowing them to be tailored
to the particular boundary conditions of this region of Japan.
This paper will provide an overview of the F-TRACE Project, progress to date and the programme of future
work to develop and test models and databases. Because of great interest by both the international scientific
community and the general public, it is important that work is carried out with strict quality control and is
well communicated to all stakeholders, therefore these aspects will also be discussed.
Sakashita,
Kawamata Town
◇
Matsukiyama,
Namie Town
◇ ◇
Odaka Riv.
◇ ◇
◇
Ottozawa,
Okuma Town
Okuma Town
office
Yonomori Park◇
◇ ◇
◇
Ukedo Riv.
Maeda Riv.
Kuma Riv.
Air dose rate at 1m
height (µSv/h)
[31 May 2012]
Tomioka Riv.
Kainosaka, Ogi dam
Kawauchi Vil.
(1) Water flow transport scenario
● Forest investigation
〇 River & estuary investigation
◎ Dam & reservoir investigation
(2) Other transport scenario
△ Follow-up monitoring
(3) Constraint of transport
□ Prototype test for constraint of transport
◇ Lichen investigation
Figure 1. Field areas studied for the development of a radiocaesium transport model
[1] Ohara, T., Morino, Y. and Tanaka, A., J. Natl. Inst. Public Health, 60 (4) (2011).
◇
341

PFS-3
DEVELOPMENT OF A BOX-MODELLING APPROACH TO INTEGRATE CONTAMINATED
SITE UNDERSTANDING IN FUKUSHIMA
S.M.L. Hardie (1) , I.G. McKinley (1) , T.M. Beattie (2) , E.M. Scourse (2) , and L. Klein (1)
(1) MCM, Baden-Dättwil, Switzerland
(2) MCM, Bristol, UK
Integrating understanding of all processes influencing the mobility, biological uptake and subsequent health
risk of radionuclides in a large site contaminated by either accidents or past waste disposal activities is a
major challenge. In the past, predominantly due to the limitations of available computing power, great
simplifications were required in order to reduce this to something that could be simulated in a model
(McKinley et al., 1981). Typically, soils & sediments were treated as equivalent porous media, radionuclides
were considered to travel in true solution constrained by the solubility of pure crystalline phases, all uptake
processes were integrated as a K d , biological uptake was integrated within a transfer or “dose conversion”
factor and complications from microbial activity, colloids, gas phases, etc. simply ignored.
Although both system understanding and, in particular, the power of computers has increased dramatically in
the last 2 decades, this has been little reflected in the approach to site modelling. This paper reviews the
approaches currently used and re-assesses the potential of box models which, although computationally
inefficient, provide a flexible framework for integrating empirical system observations with more
fundamental system quantification. This is illustrated for the case of off-site contamination resulting from
releases from Fukushima Daiichi. Although the only significant radionuclides are isotopes of radiocaesium,
the complexity and dynamic nature of this location provides major challenges for both model development
and testing (Hardie & McKinley, In Press). Nevertheless, this appears to lie within the current state of the art
and is currently being discussed with Japanese organisations involved with Fukushima remediation. If
successful implementation can be demonstrated, the approach may be worth considering for other
contaminated legacy sites.
McKinley, I.G., Baxter, M.S., Jack, W. (1981). A simple model of radiocaesium transport from Windscale to the Clyde
Sea Area. Estuarine, Coastal and Shelf Science, 13, 59-67.
Hardie, S.M.L. & McKinley, I.G. (In Press). Fukushima remediation: status and overview of future plans. Journal of
Environmental Radioactivity.
342

PUK SPECIAL UK SESSION
PUK-1
PUK-2
PUK-3
PUK-4
PUK-5
PUK-6
PUK-7
PUK-8
PUK-9
PUK-10
SYNTHESIS OF SUPERPLASTICISERS TAILORED FOR APPLICATIONS IN
NUCLEAR DECOMMISSIONING AND STORAGE
M. Isaacs, S. Edmondson, S. Christie, D. Read (UK)
CARBONATION OF REPOSITORY CEMENT:
IMPACT OF CO 2 ON CEMENT MINERALOGY, WATER CHEMISTRY AND
PERMEABILITY
G. Purser, C. Rochelle, A. Milodowski, D. Noy, J. Harrington D. Wagner, A. Butcher (UK)
THE ROLE OF MONITORED NATURAL ATTENUATION IN MANAGEMENT OF
RADIOACTIVELY CONTAMINATED GROUND FROM THE SELLAFIELD MAGNOX
SWARF STORAGE SILO
J. Graham (UK)
A STATISTICAL APPROACH TO INVESTIGATING ENHANCEMENT OF
POLONIUM-210 IN THE EASTERN IRISH SEA ARISING FROM DISCHARGES
FROM A FORMER PHOSPHATE PROCESSING PLANT
A. Dewar, W. Camplin, J. Barry, P. Kennedy (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM
K. Morris, N.D. Bryan, N.D.M Evans, F.R. Livens, J.R. Lloyd, J.F.W. Mosselmans, A.
Milodowski, S. Shaw, J.S. Small, S. Thornton (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM: MODELLING AND SYNTHESIS
J.S. Small, L. Abrahamsen, X. Chen, H Steele, S. F. Thornton (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM: WORK PACKAGE 2 – RADIONUCLIDE FORM,
REACTION, AND TRANSPORT
N.D. Bryan, K. Morris, G.T.W. Law, N.D.M Evans, F.R. Livens, J.R. Lloyd, J.F.W.
Mosselmans, S. Shaw, K. Smith, A. Stockdale, P. Bots, M. Felipe-Sotelo (UK)
MICROBIOLOGICAL CROSS-CUTTING RESEARCH; THE NERC
BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT BIGRAD
CONSORTIUM
J.R. Lloyd, A. Williamson, C. Boothman, A. Rizoulis, N. Bassil, S. Smith, K. Morris, S. Shaw,
G.T.W. Law, A. Milodowski, J.M. West, J.S. Small, J.F.W. Mosselmans (UK)
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT
BIGRAD CONSORTIUM: WORK PACKAGE 1 – GEOSPHERE EVOLUTION
S. Shaw, A.E. Milodowski, E.B.A. Moyce C. Rochelle, G.T.W. Law, J.F.W. Mosselmans, P.
Bots, T.A.M. Marshall, M.E. Romero-Gonzalez, A. Seliman, K. Morris (UK)
THE BIGRAD CONSORTIUM – RADIONUCLIDE SORPTION AT HIGH pH IN
CALCITE SYSTEMS
Kurt Smith, Nick Bryan, Katherine Morris
343

PUK-1
SYNTHESIS OF SUPERPLASTICISERS TAILORED FOR APPLICATIONS IN NUCLEAR
DECOMMISSIONING AND STORAGE
M. Isaacs ab , S. Edmondson a , S. Christie b , D. Read b
a) Department of Materials and b) Department of Chemistry, Loughborough University, Leicestershire,
UK, LE11 3TU
Safe handling and storage of nuclear waste is an important consideration within the nuclear industry in order
to minimise exposure to workers and release to the environment. In the UK, low to intermediate level nuclear
waste is encapsulated in cement via which the waste in question is used as the aggregate so that it can be
stored and handled in solid form. However, the presence of commercial superplasticisers has been shown to
increase the mobility of radionuclides in cement and increase bleed. In this work attempts are being made to
develop a viable superplasticiser that confers the necessary rheology and also fits the needs of the nuclear
industry.
It is well known that the presence of organics can increase the solubility and mobility of radionuclides in
cementitious systems and, in the case of older types of superplasticisers, such as sulfonated melamine
formaldehyde condensates or modified lignosulfonates, metal lability is increased to an unacceptable level.
More recent developments in superplasticiser synthesis have led to the use of polycarboxylate comb
polymers as they have been shown to be the most effective at improving workability. ADVA Cast 551, a
commercial superplasticiser, was the subject of previous research into the effects of such polymers on the
mobility of radionuclides. It was shown that ADVA Cast 551 increased the solubility of U, Ni and Eu and
hence would not be suitable for use in an encapsulation grout.
An initial objective of the project was to synthesise a polycarboxylate superplasticiser from its monomer
constituents and fully characterise the polymer, the aim being to determine whether it is the superplasticiser
or impurities and additives in the commercial product that causes the increase in mobility. A number of
superplasticisers has been synthesised using a protocol derived from the literature based on an aqueous
polymerisation of poly(ethylene glycol) methyl ether methacrylate, sodium methacrylate and methallyl
sulfonate, which acts as a chain transfer agent. These polymers have been characterised using NMR and IR
spectroscopy and gel permeation chromatography.
The effect of the new superplasticisers on the rheology of EN 197-1 CEM I Ordinary Portland Cement
(OPC) pastes has been tested using 'mini-slump' as well as Colcrete flow channel tests in order to determine
their performance with regards to the increase in workability and reduction in free water content. The effect
of the synthesised superplasticisers on the solubility of metal salts is being investigated by oversaturation
solubility measurements. The results will be compared to those obtained from tests using ADVA Cast 551.
Novel comb-polymers resembling polycarboxylate superplasticisers have been synthesised to assess whether
methallyl sulfonate can be replaced with molecules such as diethyl(2-methylallyl) phosphonate in the hope
that the resulting polymer may still act as a satisfactory superplasticiser but coordinate less strongly with
radionuclides and thus not increase their mobility in cement paste.
PUK-2
CARBONATION OF REPOSITORY CEMENT:
IMPACT OF CO 2 ON CEMENT MINERALOGY, WATER CHEMISTRY AND PERMEABILITY
G. Purser, C. Rochelle, A. Milodowski, D. Noy, J. Harrington D. Wagner, A. Butcher
British Geological Survey, Keyworth, NG12 5GG, UK
In the current UK reference concept for low/intermediate radioactive wastes, large quantities of cementitious
materials will be used for repository construction and buffer/backfill. Degradation of organic material within
the waste will produce CO 2. This will lead to cement carbonation, potentially reducing the capacity of the
344

uffer/backfill cement to maintain highly alkaline conditions, and as a consequence increase the likelihood of
radionuclide migration. The movement of CO 2 will be controlled by a complex interplay between transport
processes and chemical reactions, and it is unclear whether the overall changes due to carbonation will be
beneficial or deleterious to long-term radionuclide immobilisation. In this work we quantified changes in
cement mineralogy, structure, porosity/permeability, and the composition of coexisting aqueous solutions.
Laboratory investigations were undertaken using Nirex reference vault backfill (NRVB). Diffusive processes
were studied using ‘Batch experiments’ and advective ones using ‘flow experiments’.
- Thirty-two ‘batch experiments’ were assembled using cement cores (25 mm diameter x 50 mm long).
These had durations of between 10-365 days, and were conducted under a range of potential in-situ
conditions; 20°C or 40°C, 4 MPa or 8 MPa, with or without ‘young’ (Na/K/Ca-rich) or ‘evolved’ (Carich)
cement porewaters, with or without CO 2 .
- Four ‘flow experiments’ were performed in which the cement cores (49 mm diameter x 49 mm long)
were isotropically confined. Experiments were conducted at 40°C and 4 MPa (two tests with gaseous
CO 2 and one with dissolved CO 2 ) and at 40°C and 8 MPa (supercritical CO 2 ).
In the batch experiments, carbonation was associated with an increase in weight of over 8% but the cement
cores did not change in size. Carbonation reactions progressively moved into the cement creating four
distinct reaction zones/fronts.
• Zone 1 (innermost zone) - minor carbonation with minimal apparent volume change
• Zone 2 - significant carbonation and localised shrinkage expressed as small fractures. Breakdown of
portlandite (Ca(OH) 2 ) and calcium silicate hydrate (CSH) phases was actively ongoing in this zone.
• Zone 3 - complete carbonation and alterations of the cement matrix. A 3D ‘chicken wire’ meshwork
of interconnected, higher-density, carbonate-filled micro-fractures separated silica-rich areas having
lower-density and high porosity, and concentric ‘relic’ reaction fronts.
• Zone 4 - dissolution of initially-formed carbonate minerals in the outer parts of the sample by the
surrounding, slightly-acidic water.
Casting induced heterogeneities occurred within a few samples of cement (coarser layers were dominated by
portlandite and finer layers dominated by CSH) The coarser layers underwent relatively more carbonation. A
Secondary Cl-rich, calcium chloroaluminate phase formed in most experiments, though greater quantities
were formed in the CO 2 experiments. It occurred at the boundary of Zones 2 and 3 in a region just in advance
of the main carbonation front. The formation of Cl-rich phases within a repository could benefit its safety
functions, as it might help to immobilise/retard 36 Cl if it were to leach from the waste. Aqueous fluids in
contact with the cement remained alkaline in the CO2-fee experiments. With the addition of CO 2 the pH
changed to a more neutral pH 7. A lowering of the pH may facilitate metal corrosion and migration of certain
radionuclides.
Flow experiments revealed decreases in overall sample permeability as a consequence of porosity reduction
due to conversion of portlandite and CSH to secondary carbonate minerals and silica gel. Injection of freephase
(gaseous and supercritical) CO 2 halved permeability, whereas use of dissolved CO 2 reduced hydraulic
permeability by about 3 orders of magnitude.
Detailed petrographic observations of a partly-reacted cement sample revealed a series of reaction zones as
per the batch experiments. These observations, coupled with micropermeameter data, show the greatest
reductions in porosity and permeability in a very narrow zone at the leading edge of the visible alteration
front. However, it was found that cement permeability was not completely sealed, allowing potential for the
cement to vent gas, preventing the build-up of excess gas pressure in and around the waste canisters. Overall
reduction in permeability could be beneficial within a repository setting, as it could lower the potential for
radionuclide migration. Reaction processes observed in these NRVB samples bear many similarities to
carbonation of borehole cements and carbonation of naturally occurring CSH phases.
Acknowledgements
The research leading to these results received funding from the European Atomic Energy Community’s
Seventh Framework Programme (FP7/2007-2011) under Grant Agreement no. 230357, the FORGE project.
345

The Nuclear Decommissioning Authority (NDA) is also thanked for helping support this research. This
abstract is published with the permission of the Executive Director of the British Geological Survey, NERC.
PUK-3
THE ROLE OF MONITORED NATURAL ATTENUATION IN MANAGEMENT OF
RADIOACTIVELY CONTAMINATED GROUND FROM THE SELLAFIELD MAGNOX SWARF
STORAGE SILO
James Graham
Environmental Services Team, Waste Management & Decommissioning, National Nuclear Laboratory, 5th
Floor, Chadwick House, Warrington Road, Birchwood Park, WA3 6AE, UK
Sellafield’s Separation Area contains a high number of Legacy plants associated with radioactive waste
reprocessing. Radioactive Liquor is known to have leaked to ground from a number of these plants including
The Magnox Swarf Storage Silos (MSSS). The MSSS was constructed in the 1960s to 1980s to hold
irradiated decanning wastes (Magnox swarf and miscellaneous intermediate level waste). The MSSS
received waste until 2000 but is now a priority for waste retrievals and decommissioning. Following these
activities the remediation of the ground beneath the building foundations can be undertaken. This final
activity is planned for the year 2045. While waste retrievals operations have been designed to avoid further
liquor leakage, this cannot be entirely ruled out. As such interim measures are needed to manage the
historical ground contamination and also any future ground contamination if additional leakage should occur.
Natural attenuation refers to the combination of physical, chemical and biological processes that act, without
human intervention, to decrease contaminant concentrations, flux and toxicity, and thereby reduce the risks
posed by contamination. Monitored natural attenuation (MNA) is therefore a method that takes advantage of
natural processes of attenuation. A key benefit of MNA in relation to the highly developed industrial area
around the MSSS is that it requires only minimal ground intrusion.
The potential application of MNA as an interim measure to manage historical and possible future ground
contamination from the MSSS is currently under investigation. Building upon the current local and site
conceptual models work has been undertaken to:
‣ Identify the key contaminants of concern (COCs) through review of the silo inventory, soil and
groundwater data.
‣ Delineate the extent and development (contracting, stable, expanding) of COC groundwater plumes
in order to assess the effectiveness of natural attenuation processes.
‣ Improve the current understanding of the geochemistry of the identified COCs and their behaviour
under the prevailing Sellafield site geochemistry.
‣ Consider the likely impact of future leaks on the mobility of existing ground contamination.
The work undertaken is allowing more accurate and less conservative parameterisation of predictive models
of MSSS contaminant transport to be made and therefore more accurate predictions of the likely success of
MNA in managing groundwater radionuclide activities and risk. In turn these studies will help inform the site
of the requirement for any additional intrusive mitigation measures.
346

PUK-4
A STATISTICAL APPROACH TO INVESTIGATING ENHANCEMENT OF POLONIUM-210 IN
THE EASTERN IRISH SEA ARISING FROM DISCHARGES FROM A FORMER PHOSPHATE
PROCESSING PLANT
Alastair Dewar a , William Camplin a , Jon Barry a and Paul Kennedy c
a Centre for Environment, Fisheries and Aquaculture Science, Lowestoft, United Kingdom, NR33 0HT
b Food Standards Agency, London, United Kingdom, WC2B 6NH
Since the cessation of phosphoric acid production (in 1992) and subsequent closure and decommissioning
(2004) of the Rhodia Consumer Specialties Limited plant in Whitehaven, the concentration levels of
polonium-210 ( 210 Po) in local marine materials have declined towards a level more typical of natural
background. However, enhanced concentrations of 210 Po and lead-210 ( 210 Pb), due to this historic industrial
activity (plant discharges and ingrowth of 210 Po from 210 Pb), have been observed in fish and shellfish samples
collected from this area over the last 20 years. The results of this monitoring, and assessments of the dose
from these radionuclides, to high-rate aquatic food consumers are published annually in the Radioactivity in
Food and the Environment (RIFE) report series. The RIFE assessment uses a simple approach to determine
whether and by how much activity is enhanced above the normal background.
As a potential tool to improve the assessment of enhanced concentrations of 210 Po in routine dose
assessments, a formal statistical test, where the null hypothesis is that the Whitehaven area is contaminated
with 210 Po, was applied to sample data. This statistical, modified “green”, test has been used in assessments
of chemicals by the OSPAR commission. It involves comparison of the reported environmental
concentrations of 210 Po in a given aquatic species against its corresponding Background Assessment
Concentration (BAC), which is based upon environmental samples collected from regions assumed to be not
enhanced by industrial sources of 210 Po, over the period for which regular monitoring data is available
(1990–2010). Unlike RIFE, these BAC values take account of the variability of the natural background level.
As an example, for 2010 data, crab, lobster, mussels and winkles passed the modified “green” test (i.e. the
null hypothesis is rejected) and as such are deemed not to be enhanced. Since the cessation of phosphoric
acid production in 1992, the modified “green” test pass rate for crustaceans is ~42% and ~60% for molluscs.
Results of dose calculations are made (i) using the RIFE approach and (ii) with the application of the
modified “green” test, where samples passing the modified “green” test are assumed to have background
levels and hence zero enhancement of 210 Po. Applying the modified “green” test reduces the dose on average
by 40% over the period of this study (1990-2010).
PUK-5
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT BIGRAD
CONSORTIUM
1 K.Morris, 2 N.D. Bryan, 3 N.D.M Evans, 1,2 F.R. Livens, 1 J.R. Lloyd, 3 J.F.W. Mosselmans, 4 A.Milodowski, 1 S.
Shaw, 5 J.S. Small, 6 S. Thornton
1 School of Earth, Atmospheric and Environmental Sciences. The University of Manchester, Manchester, M13
9PL. UK.
2 School of Chemistry. The University of Manchester, Manchester, M13 9PL. UK.
3 School of Chemistry, Loughborough University, Loughborough, LE11 3TU, UK.
3 Diamond Light Source, Didcot, Oxfordshire, OX11 0DE, UK.
4 British Geological Survey, Keyworth, Nottingham, NG12 5GG.
5 National Nuclear Laboratory, Risley, Warrington, WA3 6AE.
6 Kroto Research Institute, The University of Sheffield, Sheffield, S3 7HQ.
The NERC BIogeochemical Gradients and RADionuclide Transport, BIGRAD, Consortium will run over
the period from 2010 – 2014. The work of the BIGRAD consortium is focussed on understanding the
chemically disturbed zone of an intermediate level radioactive waste geological disposal facility. Its
objectives are to: 1) gain a holistic understanding of biogeochemical processes and their controls on
347

adionuclide behaviour in the chemically disturbed zone; and 2) to develop a predictive modelling capability,
firmly rooted in scientific advances and experimental results, to describe radionuclide mobility in the CDZ.
Here we will present the research strategy for the consortium and introduce the breadth of the work
performed in the areas of (i) geosphere evolution, (ii) radionuclide form, reaction and transport, (iii)
synthesis and application to performance assessment (iv) biogeochemical processes in the chemically
disturbed zone and (v) predictive modelling. We will also highlight key results that will be discussed in more
detail in several presentations at the Migration 2013 conference by BIGRAD members. In addition, we will
highlight the wider impacts from the work of the BIGRAD consortium to date.
PUK-6
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT BIGRAD
CONSORTIUM: MODELLING AND SYNTHESIS
1 J.S. Small, 1 L. Abrahamsen, 2 X. Chen, 1 H Steele, 2 S. F. Thornton.
1 National Nuclear Laboratory, Risley, Warrington, WA3 6AE.
2 Kroto Research Institute, The University of Sheffield, Sheffield, S3 7HQ.
The NERC BIogeochemical Gradients and RADionuclide Transport, BIGRAD, Consortium will run over
the period from 2010 – 2014. The work of the BIGRAD consortium is focussed on understanding the
chemically disturbed zone (CDZ) of an intermediate level radioactive waste geological disposal facility
(GDF). Geochemical-based modelling provides an important cross-cutting tool to interpret experimental data
obtained from mineralogical, biogeochemical and radiochemical experiments undertaken within BIGRAD.
Models have been developed with PHREEQC to interpret the interaction between hyper alkaline cement
leachate and silicates in representative host rocks in long-term (15 year) batch experiments and shorter term
column experiments, to evaluate the pH evolution of the CDZ. These models provide insight into the
geochemical, mineralogical and physical changes that will occur in the host rock matrix as the chemistry of
the cement leachate released from a GDF evolves over time. In turn the conceptual process model developed
from this analysis can identify the relevant controls on radionuclide migration through the CDZ and
geosphere for GDF design and risk management. The thermodynamics of microbially-mediated reduction
processes at the pH of the alkaline CDZ have been evaluated and biogeochemical models have examined the
microbial growth kinetics of denitrification, iron reduction and the associated reduction of radionuclides that
have been demonstrated by BIGRAD research to occur up to pH 11.
Models of BIGRAD experiments and wider research findings, such as relating to radionuclide sorption and
complexation, provide a conceptual basis of upscaled models that apply BIGRAD research to the geological
disposal safety case. This synthesis of information utilises wider international collaborations at rock
laboratories and expertise of implementers and regulators through the BIGRAD Advisory Group to define
scenarios for analysis. Scenario models will examine the pH and Eh evolution of the CDZ for both advection
and diffusion-dominated transport processes. These consider the temporal evolution of the GDF and the
timing of releases of reactive biogeochemical species to evaluate the effect on redox sensitive radionuclides
(Tc, U, Pu, Np) examined by BIGRAD.
348

PUK-7
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT BIGRAD
CONSORTIUM: WORK PACKAGE 2 – RADIONUCLIDE FORM, REACTION, AND TRANSPORT.
1 N.D. Bryan, 2 K. Morris, 1 G.T.W. Law, 3 N.D.M Evans, 1,2 F.R. Livens, 1 J.R. Lloyd, 4 J.F.W. Mosselmans, 1 S.
Shaw, 1,2 K. Smith, 2 A. Stockdale, 2 P. Bots, 3 M. Felipe-Sotelo
1 School of Chemistry. The University of Manchester, Manchester, M13 9PL. UK.
2 School of Earth, Atmospheric and Environmental Sciences. The University of Manchester, Manchester, M13
9PL. UK.
3 School of Chemistry, Loughborough University, Loughborough, LE11 3TU, UK.
4 Diamond Light Source, Didcot, Oxfordshire, OX11 0DE, UK.
The NERC BIogeochemical Gradients and RADionuclide Transport, BIGRAD, Consortium (2010 – 2014)
is studying the chemically disturbed zone (CDZ) of a cementitious intermediate level radioactive waste
geological disposal facility. Work Package 2 is focussed on the influence of the conditions of the CDZ on
radionuclide behaviour and transport. Here we present some examples of the research.
The migration of uranium through sandstone in an alkaline environment has been studied: the aim of the
work is to assess the transport of uranium(VI) under advective conditions through intact cores of sandstone,
and the effect of the changes induced in the rock by the reaction with cement leachates on the sorption of
uranium. In the absence of HA, the sandstone showed a significant capacity to sorb U. No differences were
observed for the concentration of U in solution for filtered (30 kDa membrane filters) and unfiltered samples,
which would suggest the absence of colloids, or that any present had an insignificant effect the transport of
U. Observation of the migration pathways of U through the cores by autoradiography showed clear
enrichment of U on the inlet end of the core, and also diffuse, but significant, U uptake on the finer clay- and
Fe 2 O 3 -rich laminae adjacent to the coarser and more porous and permeable sand layers.
The behaviour of U(VI) in calcite systems under CDZ conditions has also been studied. At pH 10.5,
luminescence spectroscopy has been used to identify a liebigite like Ca 2 UO 2 (CO 3 ) 3 surface complex at the
lowest loadings. At higher loadings, the calcite surface encourages precipitation of a U(VI) containing phase
which TEM suggests is related to calcium uranate. At higher pH, U(VI) shows no interaction with the calcite
solids, which SAXS analysis suggests is because U(VI) exists in a colloidal form.
The influence of dissolved organic matter on radionuclide mobility at high pH has been studied. Laboratory
experiments investigated binding of Np(V) and U(VI) to dissolved organic matter (DOM) in simple
solutions. Neptunyl binding to DOM shows a maximum over the pH range expected within an evolving
repository and uranyl exhibits decreasing binding with pH, however, the majority of metal in solution is
present as organic complexes under the lower pH conditions investigated (10-10.5). Using these data and
other datasets from published literature, we have updated the WHAM/Model VII chemical speciation code.
These updates now allow application of the model for more complex mixtures that are representative of the
conditions expected in and around a cementitious repository. Calculations for three simulated interstitial
waters (representing different cement degradation phases) suggest U(VI) and Np(V) are not likely to be
significantly bound to DOM under these conditions.
Iron oxides are ubiquitous throughout the environment and will be present in a repository. Therefore, we
have studied the interaction of radionuclides with iron oxide surfaces. During all experiments performed on
the effect of iron oxides on the mobility of neptunium, > 95% of the added Np was removed from solution.
Furthermore, Np XANES and EXAFS revealed that during the transformation to magnetite, neptunium was
reduced to Np(IV). Additionally, the EXAFS spectra from Np adsorbed to the iron oxide surfaces are
distinctly different to those from Np associated with the crystalline end products of the experiments,
suggesting that Np may have become incorporated into the structure.
349

PUK-8
MICROBIOLOGICAL CROSS-CUTTING RESEARCH; THE NERC BIOGEOCHEMICAL GRADIENTS
AND RADIONUCLIDE TRANSPORT BIGRAD CONSORTIUM
1 J.R. Lloyd, A. Williamson, C. Boothman, A. Rizoulis, N. Bassil, S. Smith, 1 K.Morris, 1 S. Shaw, 2 G.T.W.
Law, 3 A.Milodowski, 3 J.M. West, 4 J.S. Small, 5 J.F.W. Mosselmans.
1 School of Earth, Atmospheric and Environmental Sciences. The University of Manchester, Manchester, M13
9PL. UK.
2 School of Chemistry. The University of Manchester, Manchester, M13 9PL. UK.
3 British Geological Survey, Keyworth, Nottingham, NG12 5GG.
4 National Nuclear Laboratory, Risley, Warrington, WA3 6AE.
5 Diamond Light Source, Didcot, Oxfordshire, OX11 0DE, UK.
The NERC BIogeochemical Gradients and RADionuclide Transport, BIGRAD, Consortium will run over
the period from 2010 – 2014. The work of the BIGRAD consortium is focussed on understanding the
chemically disturbed zone of an intermediate level radioactive waste geological disposal facility. Its
objectives are to: 1) gain a holistic understanding of biogeochemical processes and their controls on
radionuclide behaviour in the chemically disturbed zone; and 2) to develop a predictive modelling capability,
firmly rooted in scientific advances and experimental results, to describe radionuclide mobility in the CDZ.
Here we will present work under a microbiological cross-cutting theme. Several projects within the BIGRAD
consortium focus on the impact of microbial metabolism in the chemically disturbed zone, and include work
on microbial diversity and ecology of high pH analogues sites, mineral biotransformations, radionuclide
redox biogeochemistry, ISA biodegradation at high pH, microbial gas metabolism and biofilm formation.
The multidisciplinary approaches being utilised in this project will be discussed including classical
microbiology, genomic and postgenomic tools, spectroscopy, radiochemistry, modelling, electron
microscopy and other imaging techniques. The key impacts of these microbial processes on radionuclide
migration in the CDZ will also be presented.
PUK-9
THE NERC BIOGEOCHEMICAL GRADIENTS AND RADIONUCLIDE TRANSPORT BIGRAD
CONSORTIUM: WORK PACKAGE 1 – GEOSPHERE EVOLUTION.
1 S. Shaw, 2 A. E. Milodowski, 3 E.B.A. Moyce 2 C. Rochelle, 4 G.T.W. Law, 5 J.F.W. Mosselmans, 1 P. Bots,
1 T.A.M. Marshall, 6 M.E. Romero-Gonzalez, 6 A. Seliman and 1 K. Morris
1 School of Earth, Atmospheric and Environmental Sciences. The University of Manchester, Manchester, M13
9PL, UK.
2 British Geological Survey, Keyworth, Nottingham, NG12 5GG, UK.
3 School of Earth and Environmental Sciences, The University of Leeds, Leeds, LS2 9JT, UK.
4 School of Chemistry, The University of Manchester, Manchester, M13 9PL, UK.
5 Diamond Light Source, Didcot, Oxfordshire, OX11 0DE, UK.
6 Kroto Research Institute, The University of Sheffield, Sheffield, S3 7HQ, UK.
The NERC BIogeochemical Gradients and RADionuclide Transport, BIGRAD, Consortium will run
over the period from 2010 – 2014. The work of the BIGRAD consortium is focussed on understanding the
chemically disturbed zone (CDZ) surrounding an intermediate level radioactive waste geological disposal
facility (GDF). The objectives of Work Package 1 (WP 1) are focussed on identifying the products and
phases arising from the reaction of high pH cement fluids with the host rock; and 2) to quantify the effects of
reaction between the host rock, CDZ relevant fluids and biogeochemical processes on porosity and
permeability evolution.
The biogeochemical environment within the CDZ will evolve significantly with time as the various
cement pore fluids react with the host rock surrounding the GDF. Little is known of the chemical and
350

physical properties of the secondary phases that will form, or how these phases will affect the physical
properties of the CDZ (e.g. porosity and permeability). Here we will present work aimed at quantifying the
kinetics and mechanisms of the biogeochemical alteration processes within the CDZ, using a range of
advanced experimental techniques to mimic conditions inside the CDZ over a range of scales. A combination
of batch and column experiments have been conducted on a selection of rocks (e.g. sandstone) and pure
minerals (e.g. feldspar and biotite), at timescales from months to >15 years. Analyses of the experimental
products have been used to determine the complex pathways of rock and mineral alteration, which vary
significantly with time, rock/mineral composition and fluid chemistry. In addition, the column experiments
have shown that the formation of secondary phases can have a significant effect on permeability. The key
impacts of these geochemical and mineralogical processes on radionuclide migration in the CDZ will also be
presented.
PUK-9
THE BIGRAD CONSORTIUM – RADIONUCLIDE SORPTION AT HIGH PH IN CALCITE
SYSTEMS
Kurt Smith 1,2 , Nick Bryan 2,1 , Katherine Morris 1
1 Research Centre for Radwaste and Decommissioning and SEAES, University of Manchester, Williamson
Building, Oxford Road, Manchester, M13 9PL
2 Centre for Radiochemistry Research and School of Chemistry, The University of Manchester, Oxford Road,
Manchester, M13 9PL, United Kingdom.
Cement is ubiquitous in geological disposal facilities and carbonation will, over geological time,
convert most of the portlandite components of cement into calcite. In addition, any cementitious backfill
material of the type proposed in the UK programme contains a large calcite component. Thus, calcite is
likely to be an important mineral phase in any Geological Disposal Facility (GDF) and may play an
important role in the sequestration of radionuclides. Calcite has already been shown to be good at
sequestering radionuclides including U(VI) under ambient conditions[1]. Understanding the sorption
behaviour of UO 2
2+
in high pH conditions representative of a GDF is the focus of this work using model
Young Cement Leachate (YCL, pH 13.3) and Old Cement Leachate (OCL, pH 10.5) and apply modelling,
batch and EXAFS techniques to define the mechanisms of the radionuclide sorption in these systems.
Geochemical modelling showed that at 0.42 µM U(VI) both calcite equilibrated YCL and OCL are
supersaturated with respect to uranium. The phases predicted to be supersaturated vary: in OCL it is
primarily becquerelite (Ca(UO 2 ) 6 O 4 (OH) 6·8(H 2 O)) and schoepite ((UO 2 ) 8 O(OH) 12 ∙12(H 2 O)) that are
predicted to precipitate; in YCL it is clarkeite ((Na,Ca) 2 (UO 2 ) 2 (O,OH) 3 ) and calcium uranate (CaUO 4 ).
However, experimentally 42.0 µM and 4.20 µM U(VI) concentrations are stable in solution (

Ca scatterers. ESEM in backscattering mode was used to identify select calcite crystals that were coated in
electron dense uranium (see Figure 1) and supporting our interpretation that U(VI)-precipitate was
dominating these systems. Overall these data suggest that calcite is reactive to uranium across the
hyperalkaline conditions expected in an intermediate level waste GDF.
[1] E. J. Elzinga, Geochemica Acta. 68, 2437 (2003) .
352

SESSION 11
B4: EFFECTS OF BIOLOGICAL AND
ORGANIC MATERIALS
MICROBIAL METABOLISM; CRITICAL CONTROLS ON RADIONUCLIDE MIGRATION IN
NATURAL AND ENGINEERED ENVIRONMENTS
J.R. Lloyd, A. Rizoulis, N. Bassil, L. Newsome, C.L. Thorpe, V.E. Evans, A.R. Brown, R. Kimber, C.
Boothman, G.T.W. Law, F.R. Livens, K. Morris (INVITED) (UK)
AQUEOUS STABLE 127 I AND RADIOACTIVE 129 I SPECIATION AND UPTAKE BY
SUBSURFACE SEDIMENTS IN AN ARID ENVIRONMENT
D. I. Kaplan, C. Xu, S. Zhang, H.-S. Chang, H.-P. Li, Y.-F. Ho, K.A. Schwehr, C.M. Yeager, D.
Wellman, P.H. Santschi (USA)
EXPERIMENTS TO ASSESS THE MOBILITY OF NICKEL ISOTOPES THROUGH A
CEMENTITIOUS BACKFILL IN THE PRESENCE OF CELLULOSE DEGRADATION
PRODUCTS, USING 63 NI AS A TRACER
J. Hinchliff, M Felipe-Sotelo, D. Read, N.D.M. Evans, S. J. Williams, D. Drury, A.E. Milodowski
(UK)
BIOSORORPTION OF ACTINIDES TOWARDS HALOPHILIC MICROORGANISMS
D.T. Reed, J.S. Swanson, K. Simmons, J.F. Lucchini, D. Cleveland, M.K. Richman (USA)
B4-1
B4-2
B4-3
B4-4
B4-1
MICROBIAL METABOLISM; CRITICAL CONTROLS ON RADIONUCLIDE MIGRATION IN
NATURAL AND ENGINEERED ENVIRONMENTS
J.R. LLOYD, A. RIZOULIS, , N. BASSIL, L. NEWSOME, C.L. THORPE, V.E. EVANS, A.R. BROWN, R. KIMBER, C.
BOOTHMAN, G.T.W. LAW, F.R. LIVENS, K. MORRIS
SEAES, WRC and RCRD, University of Manchester, M13 9PL, UK
Microbial metabolism can have a controlling influence on the solubility of actinides and fission products in
engineered and natural environments. In the “far field” surrounding a nuclear repository, microorganisms can
immobilise redox active radionuclides via respiratory processes that either change directly the oxidation state
of the element, or produce new biogenic phases for enhanced sorption. In the “near field” of the repository,
the direct and indirect impacts of microbial metabolism are less well characterised but have the potential to
have a significant impact on wasteform evolution and radionuclide mobility, and must be incorporated into
the safety case of the repository.
Recent work on the redox cycling of U, Np, Pu and Tc will be discussed, including both reduction and
oxidation reactions and their impact on soluble and insoluble radionuclide inventories. The roles of proteins,
secreted electron shuttles and other microbial products will be discussed alongside additional controls
coupled to bulk element cycles e.g. the production of new mineral phases or significant changes in the
geochemical environment such as pH. Studies from a range of contrasting natural and engineered systems
will highlight how microbial communities can respond to the radioactive inventory and the extreme
(radio)chemistry of some disposed wasteforms, and ultimately control the biogeochemical fate of key
radioactive elements, including studies on microbial gas metabolism, metal chelate (ISA) biodegradation and
radionuclide biotransformations mediated via direct and indirect interactions with microbial systems.
[1] Williamson, A.J., Morris, K., Shaw , S., Byrne, J.M., Boothman, C. and Lloyd, J.R. (2013) Microbial reduction of
Fe(III) under alkaline conditions relevant to geological disposal. Applied and Environmental Microbiology. DOI:
10.1128/AEM.03063-12
[2] Brookshaw, D.R., Pattrick, R.A.D, Lloyd, J.R. and Vaughan, D.J. (2012) Microbial effects on mineral-radionuclide
interactions and radionuclide solid-phase capture processes. Mineralogical Magazine 76 777–806
[3] Kimber, R.L., Boothman, C., Purdie, P., Livens, F.R. and Lloyd, J.R. (2012). Biogeochemical behaviour of
plutonium during anoxic biostimulation of contaminated sediments. Mineralogical Magazine 76 567-578
353

[4] Rizoulis, A., Steele, H.M., Morris, K. and Lloyd, J.R. (2012) The potential impact of anaerobic microbial
metabolism during the geological disposal of intermediate-level waste. Mineralogical Magazine 76 397–406
[5] Thorpe C.L., Lloyd J.R., Law G.T.W., Burke, I.T., Shaw, S, Bryan, N.D. and Morris, K. (2012) Strontium sorption
and precipitation behaviour during bioreduction in nitrate impacted sediments Chemical Geology 306 114-122 DOI:
10.1016/j.chemgeo.2012.03.001
[6] Williams, K.H., Bargar, J.R., Lloyd, J.R. and Lovley, D.R. (2012) Bioremediation of uranium-contaminated
groundwater: a systems approach to subsurface biogeochemistry. Current Opinion in Biotechnology 24 1–9 DOI:
10.1016/j.copbio.2012.10.008
[7] Lloyd, J.R. and Gadd, G.M. (2011). The geomicrobiology of radionuclides. Geomicrobiol. J. 28:5-6, 383-386
[8] Law GTW, Geissler A, Lloyd JR, Livens, FR, Boothman, C, Begg, JDC, Denecke, MA, Rothe, J., Dardenne, K.,
Burke, IT, Charnock, JM and Morris, K (2010) Geomicrobiological Redox Cycling of the Transuranic Element
Neptunium. Environmental Science and Technology 44 8924-8929 DOI: 10.1021/es101911v
[9] Lear, G., McBeth, J.M., Boothman, C., Gunning, D., Ellis, B.L., Lawson, R.S., Morris, K., Burke, I.T., Bryan, N.D.,
Brown, A.P., Livens, F.R. & Lloyd, J.R. (2010). Probing the Biogeochemical Behavior of Technetium Using a Novel
Nuclear Imaging Approach. Environmental Science and Technology 44 156–162 doi 10.1021/es802885r
B4-2
AQUEOUS STABLE 127 I AND RADIOACTIVE 129 I SPECIATION AND UPTAKE BY
SUBSURFACE SEDIMENTS IN AN ARID ENVIRONMENT
Daniel I. Kaplan 1 , Chen Xu 2 , Saijin Zhang 2 , Hyun-shik Chang 3 , Hsiu-Ping Li 2 , Yi-Fang Ho 2 , Kathleen A.
Schwehr 2 , Chris M. Yeager 4 , Dawn Wellman 5 , Peter H. Santschi 2
1 Savannah River National Laboratory, Aiken, South Carolina, United States, daniel.kaplan@srnl.doe.gov;
2 Department of Marine Sciences, Texas A&M University, Galveston, Texas, United States
3 Savannah River Ecology Laboratory, Aiken, South Carolina, United States
4 Los Alamos National Laboratory, Los Alamos, New Mexico, United States
5 Pacific Northwest National Laboratory, Richland, Washington, United States
The worldwide inventory of 129 I continues to increase at a staggering rate: 1 Ci 129 I is created as a by-product
for every gigawatt of electricity produced by nuclear power. This results in 13,000 metric tons of 129 I per
year in spent nuclear fuel. As nuclear energy continues to be created, more 129 I waste will be produced,
increasing the chance of 129 I entering the environment. 129 I in the environment often poses unusually high
human risk because it has a perceived high mobility in the subsurface environment, a long half-life (16
million years), and an extremely high toxicity, where 90% of the body’s iodine accumulates in the thyroid.
Remediation of existing 129 I contaminated plumes are of immediate concern. For example, remediation
strategies are needed for >50 km 2 of 129 I contaminated groundwater within the Hanford Site, Washington.
Field studies were conducted in the alkaline (pH 7 to 9) and semi-arid environment of the Hanford Site,
Washington. Stable 127 I and radioactive 129 I speciation, pH, and dissolved organic carbon of groundwater
samples collected from seven wells located in the 200-West Area were investigated. The most striking
finding was that iodate (IO 3 - ) was the most abundant dissolved species. Unexpectedly, iodide (I - ), which was
the expected dominant groundwater species based on traditional assumptions associated with thermodynamic
calculations, only accounted for 1-2% of the total iodine concentration. It is likely that the relatively high pH,
low sediment organic matter (SOM) concentrations slowed down or even inhibited the reduction of iodate, as
SOM abiotically can reduce iodate to iodide. Low concentrations of groundwater organo-iodine, on average
about 7% of the total iodine, were detected in most samples, with the highest concentrations measured near
the radionuclides sources (surface cribs) and lower concentrations were detected in the far-field.
Batch sorption studies were conducted with vadose zone and surface aquifer sediments amended with 125 I - or
125 IO 3 - at ambient 129 I concentrations. Iodate uptake (K d values) was on average 89% greater than iodide
uptake. Sediment organic matter, even at very low concentrations (

dominant subsurface species. Iodine and SOM can form essentially irreversible covalent bonds, thereby
providing a yet unstudied 129 I retardation reaction at the Hanford Site.
Finally, tests were conducted to determine whether it was possible to remove 129 I from Hanford groundwater
(Hanford has some >50 km 2 of 129 I plumes) through coprecipitation with calcite. As much as 40% of the
dissolved iodine at ambient concentrations were removed through coprecipitation in calcite. Iodate was the
main species incorporated into the calcite and this incorporation process was impeded by elevating the pH
and decreasing ionic strength in the groundwater. Iodate, the dominant iodine species in the groundwater,
and carbonate have identical ionic radii, readily permitting iodate substitution into the calcite structure.
Although the 129 I plumes at the Hanford Site appear to be generally moving rapidly and at rates only slightly
slower than tritium, this study provides insight into new geochemical processes, suggesting the presence of a
second SOM-bound fraction that may be moving extremely slowly.
B4-3
EXPERIMENTS TO ASSESS THE MOBILITY OF NICKEL ISOTOPES THROUGH A
CEMENTITIOUS BACKFILL IN THE PRESENCE OF CELLULOSE DEGRADATION
PRODUCTS, USING 63 NI AS A TRACER
J. Hinchliff 1 , M Felipe-Sotelo 1 , D. Read 1 , N.D.M. Evans 1 , S. J. Williams 2 , D. Drury 3 , A.E. Milodowski 4
1. Loughborough University, UK. 2. NDA RWMD, Harwell, Oxford UK 3. Golder Associates (UK) Ltd, 4.
British Geological Survey, Keyworth, Nottingham, UK.
Many concepts for the geological disposal of intermediate level waste (ILW) and low level radioactive waste
(LLW) include backfill materials based on admixtures of Ordinary Portland Cement (OPC). These backfill
materials will generate high pH conditions and further, the eventual corrosion of the metal canisters used for
disposal will promote a low E h environment. Generally, it has been assumed that these safety functions of the
backfill will reduce the solubility of many radionuclides and retard their migration as a result of sorption and
incorporation within the near field of a Geological Disposal Facility (GDF).
The radioisotope 59 Ni is produced by activation of nickel in stainless steel reactor parts and may be a
significant component of some disposed waste. 59 Ni is a low energy beta gamma emitter (4.6 and 2.4 keV
respectively) and its half-life of ~76000 years means that escape and migration from the GDF could be of
long-term significance. The tracer used these experiments is 63 Ni which may also be present in waste.
However, its half-life of ~100 years means that escape and migration would be of importance only if it
occurs within the operational phase or first few hundred years, post closure. Cellulosic materials will also be
disposed and previous studies have shown that cellulose degradation products (CDP) produced at high pH
can enhance the migration of metal ions.
Nickel solubility in the cement backfill (in this case NRVB: Nirex Reference Vault Backfill) liquor at pH >
12.5, in the absence of CDP was found in the range 2.0 x 10 -8 – 2.4 x 10 -7 mol dm -3 . However, in the
presence of CDP nickel solubility was observed to increase by over two orders of magnitude. These
observations infer that mobility and consequently migration rates should also increase in the presence of
CDP.
Radial diffusion and advection experiments have been undertaken to compare the mobility of nickel in the
presence and absence of CDP. The CDP solution was prepared from a mixture of tissue paper and the
components of NRVB sealed in a steel vessel and maintained at 80⁰C for 30 days. The resulting solutions
have nickel concentrations in the range 2.3 x 10 -6 – 1.9 x 10 -5 mol dm -3 .
The radial diffusion experimental technique uses small (40 mm diameter x 40 mm height) pre-cast cylinders
of NRVB. An appropriate concentration of the tracer, in this case ~30 kBq 63 Ni, was introduced into a cavity
in the centre of the cylinder, which was then sealed and submerged in a solution previously equilibrated with
the solid matrix. The increase in concentration of the tracer as it migrated into the external solution was then
determined at defined time intervals. The experiments were carried out in a nitrogen filled glovebox. The
radial advection technique uses a similar approach. The equilibrated solution is pushed, using nitrogen
355

pressure into a drilled central core in the centre of the block and collected as it exits from the outer surface.
The experiment is contained in a purpose designed cell to facilitate addition of the tracer, collection of the
exiting solution and to prevent ingress of atmospheric carbon dioxide
The results of the diffusion and advection experiments indicate that nickel mobility is significantly increased
in the presence of CDP. Initial breakthrough in the CDP diffusion experiments occurred in less than one
month and a steady rise in tracer concentration has been observed for one year. In the CDP advection
experiments nickel moves rapidly through the backfill material in a manner similar to that observed for
readily soluble species.
In the absence of CDP, initial modelling suggests that nickel mobility is limited by its solubility. The
relevant diffusion experiment has been running for a year and no breakthrough over the 20 mm flow path has
been observed.
Autoradiography has been used to visualise the flow through the backfill material and identify any residual
tracer in the central core that may have affected the mass balance.
More detailed investigations continue as there are potentially important variations in ionic concentration
between the CDP solution and the backfill liquor which may also be factors in the increased mobility of
nickel. Finally, it is recognised that the CDP solution is a complex mixture of organic chemicals; it would
therefore be desirable to run a series of experiments to identify which one or which combination of these
organic substances enhance migration most significantly.
The authors wish to thank NDA RWMD for funding this research.
E. Wieland, J. Tits, J.P. Dobler and P. Spieler. The effect of α-isosaccharinic acid on the stability of Th(IV) uptake by
hardened cement paste. Radiochimica Acta 90, 683-688 (2002).
E. Wieland, J. Tits, A. Ulrich and M. H. Bradbury. Experimental evidence for solubility limitation of the aqueous Ni(II)
concentration and isotopic exchange in cementitious systems. Radiochimica Acta 94, 29-36 (2006).
J. Hinchliff, M. Felipe-Sotelo, D. Drury, N. D. M. Evans, D. Read. Methods to Assess Radioisotope Migration in
Cementitious Media using Radial Diffusion and Advection. Proceedings, NUWCEM 1st International Symposium on
Cement-based Materials for Nuclear Wastes 11-14 October 2011 Avignon, France
Francis, A J, Cather, R, and Crossland, I G, Development of the Nirex Reference Vault Backfill; report on current status
in 1994, Nirex Science Report S/97/014, UK Nirex Ltd., Harwell, UK, 1997.
B4-4
BIOSORORPTION OF ACTINIDES TOWARDS HALOPHILIC MICROORGANISMS
D. T. Reed, J. S. Swanson, K. Simmons, J. F. Lucchini, D. Cleveland and M. K. Richmann
Earth and Environmental Sciences Division,
Los Alamos National Laboratory, Carlsbad NM, 88220, USA
Microbial processes have been shown to influence the long-term migration of actinides in a wide variety of
subsurface environments [1]. These studies however tend to focus on the near-surface contaminant problems
in low ionic-strength groundwater that is present at many DOE sites. The ongoing recertification of the
WIPP TRU repository and the broader consideration of salt repository concepts for the permanent disposal of
HLW/SF fuel has focused attention on the effects of the halophilic microorganisms that are indigenous to
Salt and high ionic-strength brine systems.
A key and important actinide microbial interaction is the bioassociation of actinide aqueous colloidal and
dissolved species. This interaction is a potential contributor to the source term of “mobile” actinide
concentration when physical transport due to intrusion scenarios is the predominant release pathway. If
transport pathways are present, bioassociation can lead to reduced mobilization (e.g., if the microbe itself is
not mobile) or enhanced mobilization if the microbe or the degradation “fragments” are mobile. When,
migration pathways exist, biocolloidal species may contribute to enhanced colloidal subsurface transport.
We are investigating the bioassociation of actinides toward halophilic microorganisms to extend what has
been observed in soil bacteria [2-3] to the microorganisms that are typically found in high ionic-strength
356

ine systems. A number of isolates have been identified from brine systems and both halophilic archaea and
bacteria have been identified [4]. The biosorption of actinides towards these microorganisms is being
investigated using redox-invariant analogs (e.g., thorium and neodymium) or actinides that can be stabilized
in specific oxidation states (e.g., Am 3+ or NpO 2 + ) since the biosorption observed is specific to the oxidation
state of the actinide and the its aqueous speciation.
We have extensively investigated [5] the biosorption of Np(V) towards Chromohalobacter sp., (a halophilic
bacteria). Np(V) is somewhat of a model system in that it does not hydrolyze until pH ~10 and reversible
sorption that could be modeled using surface complexation models was observed over a broad pH range (See
Figure 1). Halophilic bacteria, based on titration experiments, have very similar surface complexation
behavior as soil bacteria. Although some Pitzer data exists for neptunium [6, 7], sorption models that
incorporate these data are not yet available. This study has now been extended in two ways. First, the Np(V)
bioassociation with Halobacterium noricense (an archaea) led to significantly lower sorption that is believed
to be the result of significant differences in the cellular surface (this is currently under investigation).
Second, the bioassociation of Nd 3+ and Th 4+ , which are used as redox-invariant analogs for An(III) and
An(IV) species, with halophilic bacteria and archaea was investigated. The interpretation of these results
become somewhat difficult at higher pH due to the coupling of precipitation and sorptive processes as well
as the prevalence of colloidal species that interact differently than dissolved species. In the thorium system,
preferential bioassociation of colloidal material was noted and there is overall significantly less sorption due
to the strongly hydrolyzed species that predominate. For neodymium, where colloidal formation was less
prevalent, biosorption was significantly higher. These data, although not yet complete, lead to a much more
thorough understanding of how the association of actinides with halophilic microorganisms may contribute
to the actinide source term. The overall impact of biosorption on the long-term migration of actinides needs
to be considered in the overall safety assessment where high ionic-strength brines are found.
Figure 1. Experimental data for neptunium (V)
adsorption onto Chromohalobacter sp. as a
function of pH in 2 (open circles) and 4 (open
triangles) M NaClO 4 . Adsorption experiments were
performed with 5 x 10 -6 M total neptunium (V) and
5 g/L (wet weight) bacteria [3]. Solid curves
represent best-fit calculated surface complexation
models. Solid diamonds, squares, triangles, and
circles represent the results of desorption
experiments performed with 5 x 10 -6 M total
neptunium (V) and 5 g/L (wet weight) bacteria in 2
M NaClO 4 .
[1] D.T. Reed, R. Deo, and B.E. Rittmann. (2010). “Subsurface Interactions of Actinide Species and Microorganisms,”
Chapter 33 in The Chemistry of the Actinide and Transactinide Elements, 3 rd edition, L.R. Morss, N.M. Edelstein
and J. Fuger eds., Springer Press, Netherlands.
[2] R.P. Deo, W. Songkasiri, B.E. Rittmann, and D.T. Reed (2010). “Surface Complexation of Neptunium (V) onto
Whole Cells and Cell Components of Shewanella alga: Modeling and Experimental Study,” Env. Sci. and Tech.,
vol. 44: 4930-4935.
[3] W. Songkasiri, D.T. Reed and B.E. Rittmann (2002). “Bio-sorption of Neptunium (V) by Pseudomonas
fluorescens,” Radiochimica Acta, vol. 90: 785-789.
[4] J. S. Swanson, D. T. Reed, D. A. Ams, D. Norden, and K. A. Simmons (2012). “Status Report on the Microbial
Characterization of Halite and Groundwater Samples from the WIPP,” Los Alamos National Laboratory Report,
LA-UR-12-22824.
[5] D. A. Ams, J. S. Swanson, J.E. S. Szymanowski, J. B. Fein, M. Richmann, and D. T. Reed (2013). “The Effect of
High Ionic Strength on Neptunium (V) Adsorption to a Halophilic Bacterium,” Geochimica et Cosmochimica
Acta, vol. 110: 45-57.
[6] V. Neck, Th. Fanghänel, G. Rudolph and J.I. Kim (1995). “Thermodynamics of Neptunium(V) in Concentrated
Salt Solutions: Chloride Complexation and Ion Interaction (Pitzer) Parameters for the NpO 2 + Ion,” Radiochimica
Acta, vol. 69: 39-47.
[7] Th. Fanghänel, V. Neck, and J.I. Kim (1995). “Thermodynamics of Neptunium(V) in Concentrated Salt Solutions:
II. Ion Interaction (Pitzer) Parameters for Np(V) Hydrolysis Species and Carbonate Complexes,” Radiochimica
Acta, vol. 69: 169-176.
357
% adsorption
100
90
80
70
60
50
40
30
20
10
0
2 3 4 5 6 7 8 9 10 11
pC H+

SESSION 12
A3: COMPLEXATION WITH INORGANIC AND
ORGANIC LIGANDS
OXYGEN EXCHANGE BETWEEN URANYL(VI) AND WATER: BINUCLEAR SCENARIOS
IN ACID AND IN BASE
S. Tsushima, A. Rossberg, H. Moll (Germany)
THE BIGRAD CONSORTIUM - RADIONUCLIDE INTERACTIONS WITH NATURAL
ORGANIC MATTER AT HIGH pH
A. Stockdale, N.D. Bryan, S. Lofts, E. Tipping (UK)
A COMBINED EXAFS AND TRLFS SPECTROSCOPIC STUDY TO DETERMINE THE
THERMODYNAMIC AND STRUCTURAL PROPERTIES OF TRIVALENT ACTINIDE
COMPLEXES WITH ORGANIC AND INORGANIC LIGANDS AT ELEVATED
TEMPERATURES
D.R. Fröhlich, A. Skerencak-Frech, P.J. Panak (Germany)
A3-4
A3-5
A3-6
A3-4
OXYGEN EXCHANGE BETWEEN URANYL(VI) AND WATER:
BINUCLEAR SCENARIOS IN ACID AND IN BASE
S. Tsushima (1) , A. Rossberg (1) , H. Moll (1)
(1) HZDR, Institute of Resource Ecology, P.O. Box 51 01 19, Dresden, D–01314, Germany
The mechanism of exchange between oxygen in UO 2 2+ and that in solvent water has been disputed over last
50 years. It is well–known that the rate of “yl”–oxygen exchange depends heavily on the pH, and that there is
virtually no exchange at low pH. With increase of the pH (pH > 2) the exchange becomes appreciable, and
under highly alkaline solution there is a rapid oxygen exchange. These observations led to an idea that there
are at least two different exchange mechanisms; one dominating under weakly acidic to neutral pH and
another mechanism at very high pH. Szabó and Grenthe [1,2] used NMR spectroscopy to identify the species
involved in the “yl”–oxygen exchange and they suggested two binuclear complexes as key species;
(UO 2 ) 2 (OH) 2 2+ at low pH and [(UO 2 (OH) 4 2- )(UO 2 (OH) 5 3- )] at high pH. How the oxygen exchange take place
explicitly in these complexes, however, remains unidentified because the lifetimes of the intermediate
species and the transition states of the oxygen exchange are too short to be detected spectroscopically. That
is to say, we know they (dimer complexes) did it but we do not know how they managed to do it. Our
attempt here is to identify the “yl”–oxygen exchange pathways in these complexes using quantum chemical
method thereby proving that oxygen exchange through these complexes are indeed possible, thereby
bringing end to the long–disputed arguments over the “yl”–oxygen exchange mechanisms.
First, we studied the “yl”–oxygen exchange pathway via (UO 2 ) 2 (OH) 2+ 2 [3]. We used hybrid density
functional theory (DFT) with Becke’s three–parameter hybrid functional and Lee–Yang–Parr’s gradient–
corrected correlation functional (B3LYP) employing conductor–like polarizable continuum model (CPCM)
using Gaussian 09 program (Gaussian Inc.). The small core effective core potential and corresponding basis
set was used on uranium and oxygen. Direct proton transfer from the hydroxo bridge or from the
coordinating water to the “yl”–oxygen in (UO 2 ) 2 (OH) 2 (H 2 O) 2+
6 can be ruled out because we found
exceedingly high activation barrier (~170 kJ mol –1 ) through these mechanisms. The exchange mechanism in
(UO 2 ) 2 (OH) 2 (H 2 O) 2+ 6 can be described by a multi–step proton transfer pathway that involves the formation
of an oxo bridge between the two uranyl(VI) centres (U–O yl –U bridge). The activation enthalpy of the
reaction obtained at the B3LYP level is 94.7 kJ mol –1 and is somewhat larger than the experimental value of
80 ± 14 kJ mol –1 . However, the discrepancy is at the acceptable level.
Second, we tried to identify the oxygen exchange pathway through [(UO 2 (OH) 4 2- )(UO 2 (OH) 5 3- )]. For
this attempt, we first studied the speciation of uranium(VI) in highly alkaline solution by quantum chemical
358

calculations as well as by X–ray absorption spectroscopy (XAS). Although various previous studies assumed
that hydrolysis of UO 2 (OH) 4 2– produces UO 2 (OH) 5 3– , our B3LYP calculations together with previous theory
work by others [4,5] suggest that hydrolysis of UO 2 (OH) 4 2– yields UO 3 (OH) 3 3– , as shown in Figure 1. At all
levels of theory we used, we found UO 3 (OH) 3 3– is more stable than UO 2 (OH) 5 3– . We studied this point
further by XAS using the Rossendorf Beamline (ROBL) in ESRF, Grenoble, France, and we found evidence
of the existence of new species UO 3 (OH) 3 3– in the XANES spectra. The sample which contained further
hydrolyzed species showed shift of the uranium L III absorption edge compared to the sample containing only
UO 2 (OH) 4 2– . Similar energy shift was observed in Pa(V) when the species changed from spherical Pa 5+ to
mono–oxo PaO 3+ [6]. Therefore the species beyond UO 2 (OH) 4 2– is better assigned to UO 3 (OH) 3 3– rather than
UO 2 (OH) 5 3– . This assumption was further corroborated by UV-vis absorption as well as time–resolved laser
fluorescence spectroscopy (TRLFS). The complex described as [(UO 2 (OH) 4 2- )(UO 2 (OH) 5 3- )] by Szabó and
Grenthe should therefore be better written as [(UO 2 (OH) 4 2- )(UO 3 (OH) 3 3- )]. We then studied the “yl”–oxygen
exchange pathway in [(UO 2 (OH) 4 2- )(UO 3 (OH) 3 3- )] and found a realistic pathway which has the activation
Gibbs energy of 56.3 kJ mol –1 at the B3LYP level, which is again in good agreement with the experimental
value of 60.8 ± 2.4 kJ mol –1 obtained by Szabó and Grenthe using NMR magnetization transfer technique [2].
2+
Our calculations confirm the “yl”–oxygen exchange mechanisms through (UO 2 ) 2 (OH) 2 and
[(UO 2 (OH) 2- 4 )(UO 2 (OH) 3- 5 )], and underscores the role of binuclear species. The formation of U–O yl –U
bridge seems to play a key role in facilitating intramolecular proton shuttling among the oxygen atoms
thereby contributing to faster “yl”–oxygen exchange.
Figure 1. Formation of 3 is energetically preferred over 2. At all levels of theory we used (B3LYP,
BP86, M06-2X and MP2), both the reaction energy and the activation energy are much lower for 1 →3 than
for 1→2.
[1] Szabó, Z. and Grenthe, I., Inorg. Chem. 2007, 46, 9372–9378.
[2] Szabó, Z. and Grenthe, I., Inorg. Chem. 2010, 49, 4928–4933.
[3] Tsushima, S., Inorg. Chem. 2012, 51, 1434–1439.
[4] Shamov, G. A. and Schreckenbach, G., J. Am. Chem. Soc. 2008, 130, 13735–13744.
[5] Bühl, M. and Schreckenbach, G., Inorg. Chem. 2010, 49, 3821–3827.
[6] Le Naour, C. et al., Inorg. Chem. 2005, 44, 9542–9546.
A3-5
THE BIGRAD CONSORTIUM - RADIONUCLIDE INTERACTIONS WITH NATURAL ORGANIC
MATTER AT HIGH PH
A. Stockdale (1) , N.D. Bryan (1) , S. Lofts (2) , E. Tipping (2)
(1) Centre for Radiochemistry Research, School of Chemistry, University of Manchester, Oxford Road,
Manchester, M13 9PL, UK
(1) Centre for Ecology and Hydrology, Lancaster Environment Centre, Lancaster, LA1 4AP, UK
Few studies to date have sought to quantify the effects of humic substances on radionuclide behaviour under
conditions relevant to cementitious radwaste disposal (i.e. strongly alkali conditions with pH up to ~13).
Using experimental (equilibrium dialysis) and modelling (WHAM / Humic Ion Binding Model VII [1]): we
explore this theme.
Dialysis has been employed to study the equilibrium behaviour of several radionuclides important to long
term (>10 5 years) radwaste geological disposal. The results (Figure 1) show that high fractions of Th(IV) are
359

present as natural organic matter (NOM) complexes at all pH values. Binding of neptunyl to NOM shows a
maximum over the pH range expected within an evolving repository. Uranyl exhibits decreasing binding
with increasing pH, however, the majority of metal in solution is present as organic complexes under the
lower pH conditions investigated (10-10.5).
In addition to laboratory analysis, we have updated WHAM/Model VII to include new humic binding
constants for several potentially important radionuclide cations (Pd 2+ , Sn 2+ , U 4+ , Np(VI)O 2 2+ , Np(V)O 2 + , Pu 4+
and Pu(VI)O 2 2+ , [2]). These were mostly obtained using linear free energy relationships, as described by
Carbonaro and Di Toro [3]. Experimental data from several studies was used in the determination of
constants for pentvalent neptunium, and we have updated the uranyl(VI) values with new experimental
datasets. These new data, combined with data for inorganic complexation can be used to make predictions of
organic speciation over the pH range we are considering. There remains a large uncertainty associated with
the importance of colloidal species for repository safety cases [4], and this work goes some way to
improving our knowledge in this area. We also present a previously unexploited method for obtaining kinetic
data for these high pH systems, using diffusion cell apparatus.
1.0
Fraction of element bound to NOM
0.8
0.6
0.4
0.2
Uranyl (VI)
Neptunyl (V)
Th (IV)
0.0
10 11 12 13
pH
Figure 1. Complexation of U(VI)O 2 2+ , Np(V)O 2 + and Th 4+ with NOM (humic acid) at high pH.
[1] E. Tipping, S. Lofts, J. Sonke, Env. Chem. 2011, 8, 225-235.
[2] A. Stockdale, N.D. Bryan, S. Lofts, J. Environ. Monitor., 13, 2946-2950.
[3] R.F. Carbonaro and D.M. Di Toro, Geochim. Cosmochim. Acta, 2007, 71, 3958-3968.
[4] P. A. Smith, J. Guimerà, G. Kosakowski, A. Pudewills and M. Ibaraki, Eds., The CRR Final Project Report
Series III: Results of the Supporting Modelling Programme. NAGRA Technical Report 03-03, 2003.
A3-6
A COMBINED EXAFS AND TRLFS SPECTROSCOPIC STUDY TO DETERMINE THE
THERMODYNAMIC AND STRUCTURAL PROPERTIES OF TRIVALENT ACTINIDE
COMPLEXES WITH ORGANIC AND INORGANIC LIGANDS AT ELEVATED
TEMPERATURES
D. R. Fröhlich 1)* , A. Skerencak-Frech 2) , P. J. Panak 1),2)
1) Physikalisch-Chemisches Institut, Ruprecht-Karls Universität Heidelberg, Im Neuenheimer
Feld 253, 69120 Heidelberg, Germany
2) Institut für Nukleare Entsorgung, Karlsruher Institut für Technologie, Campus Nord,
P.O. Box 3640, 76021 Karlsruhe, Germany
Concerning the long-term storage of high-level nuclear waste in deep geological formations, different
possible host rock formations (salt, clay, crystalline formations) are investigated throughout Europe [1]. Due
to their long half-lives, the different transuranium elements (Np, Pu, Am, Cm) will determine the long-term
radiotoxicity of the waste materials. Therefore, a well-funded knowledge of the geochemical behavior is
indispensable with respect to the safety assessment of a potential storage site. As a result of the radioactive
360

decay, the near-field of the repository will heat up significantly (T max = 100 °C (clay) or 200 °C (salt) [2]).
This strong increase in temperature will lead to a significant change in the chemical properties of the stored
actinides including their complexation properties with organic and inorganic ligands present in the
repository.
Aim of the present work was to determine the thermodynamic constants of the complex formation of
trivalent actinides with different relevant organic (lactate) and inorganic (chloride) ligands as well as the
structural parameters of the related complexes at increased temperatures by combining different
spectroscopic techniques. Time resolved laser fluorescence spectroscopy (TRLFS) has been used to
determine the aqueous speciation of Cm(III) in the presence of the above mentioned ligands as a function of
ionic strength and ligand concentration at variable temperature. The complex formation constants (log K 0 ) as
well as the related standard reaction enthalpies and entropies (Δ r H 0 m, Δ r S 0 m) were determined according to
the specific ion interaction theory (SIT) and the Van’t Hoff equation, respectively. These results were
complemented by Am L III -edge EXAFS measurements using a newly developed high-temperature EXAFS
cell to determine the structural parameters of the corresponding Am(III) complexes as a function of
temperature.
χ(k)*k 3
25
20
15
10
5
0
25 °C
90 °C
200 °C
Experiment
Fit
-5
2 3 4 5 6 7 8 9 10
k / Å -1
R + ∆ / Å
Figure 13: k 3 -weighted Am L III -edge EXAFS spectra together with the related Fourier Transforms as a
function of temperature ([Am(III)] = 1 mM, [NaCl] = 3 M, pH ≈ 1).
361
FT
5
4
3
2
1
0
0 1 2 3 4
Chloride was chosen as it will occur at high molar concentrations in the case of water access to a repository
in rock salt. From room temperature studies it is known that chloride is a relative weak ligand [3,4].
Nevertheless, the present study clearly shows that in the temperature range between 20 and 200 °C different
Cm(III) chloride complexes are successively formed to a considerable extent, i.e. [CmCl n ] (3-n)+ (n = 1-4).
These were identified by TRLFS. EXAFS measurements of 1 mM Am(III) in 3 M NaCl solution (spectra
shown in Figure 1) showed that Am(III) is coordinated by 2-3 chloride ligands at T = 200 °C, whereas no
chloride complexes were observed at 20 and 100 °C, respectively.
In contrast to rock salt, clay formations contain certain amounts of organic matter which are also present in
the pore waters. Studies on the characterization of dissolved organic matter (DOM) in the pore waters of
different natural clay rocks (i.e., Opalinus Clay (OPA), Callovo Oxfordian (COX)) showed that low
molecular weight organic compounds (e.g., acetate, propionate, lactate) make up a high fraction of the total
DOM (OPA: 36%, COX: 88%) [5,6]. Investigations of the Cm(III)/lactate interaction by TRLFS at variable
temperatures showed that the formation of Cm(III) lactate species is an exothermic process, which is in
contrast to the complexation with simple carboxylic ligands (e.g., propionate [7]). This phenomenon was
investigated further by Am L III -edge EXAFS measurements at T = 20-90 °C which clearly showed that
Am(III) is coordinated side-on through the carboxyl and hydroxyl functions of lactate which was determined
as origin of the exothermic complexation behavior.
The present study provides new insights into the complex formation properties of actinides at elevated
temperatures. The newly determined thermodynamic data is a valuable contribution to the thermodynamic

database which is necessary for reliable long-term predictions of actinide geochemistry in a nuclear waste
repository. Furthermore, high-temperature EXAFS measurements provide molecular-level information on
the structure of the complexes formed.
*Author for correspondence: daniel.froehlich@partner.kit.edu.
[1] W. Kickmaier, et al., Nucl. Eng. Des. 176, 75 (1997).
[2] T. Brasser, et al., GRS report, GRS-247 (2008).
[3] P. G. Allen, et al., Inorg. Chem. 39, 595 (2000).
[4] T. Fanghänel, et al., J. Alloys. Comp. 225, 308 (1995).
[5] A. Courdouan, et al., Appl. Geochem. 22, 1537 (2007).
[6] A. Courdouan, et al., Appl. Geochem. 22, 2926 (2007).
[7] A. Skerencak, et al., J. Sol. Chem., in press.
362

SESSION 13
A4: REDOX REACTIONS AND RADIOLYSIS
IMPACT OF WATER RADIOLYSIS ON URANIUM DIOXIDE CORROSION
A. Traboulsi, J. Vandenborre, G. Blain, J. Barbet, M. Fattahi (France)
REDOX REACTION OF Se(IV)/Se(VI) WITH NATURAL PYRRHOTITE AND IRON
SELENIDES
M. L Kang, B. Ma, F. Bardelli, F. R Chen, C. L Liu, L. Charlet, J.L Xie (China, France)
INTERFACIAL RADIATION CHEMISTRY IN GEOLOGICAL REPOSITORIES FOR SPENT
NUCLEAR FUEL
M. Jonsson (Sweden)
IMPACT OF Fe MINERALS ON THE SAFE DISPOSAL OF NUCLEAR WASTE
C. Pearce, K. Rosso, J. Liu, S. Luksic, M. Schweiger, J. McCloy, A. Felmy (USA)
A4-1
A4-2
A4-3
A4-4
A4-1
IMPACT OF WATER RADIOLYSIS ON URANIUM DIOXIDE CORROSION
A. Traboulsi (1) , J. Vandenborre (1) , G. Blain (1) , J. Barbet (2) ,M. Fattahi (1)
(1) Laboratoire SUBATECH – 4, rue Alfred Kastler – La Chantrerie BP 20722, 44307 Nantes cedex 3 –
France
(2) Cyclotron ARRONAX – 1, rue Arronax – CS 10112, 44817 Saint Herblain cedex – France
Management of nuclear waste such as spent fuel resulting from nuclear industry is one of the major political,
social and scientific concerns of our society [1]. Among various management possibilities, the concept of
direct disposal in deep geological vaults (500 to 1000 m of depth) is being seriously considered by several
countries [2]. The spent nuclear fuel will be stored within canisters (made from steel or copper) and its
disposal is planned to be made in the presence of a series of natural and artificial barriers with the aim of
isolating the waste from the outside environment [3]. However, it is admitted that groundwater will, within
geological timescales (≥ 1000 years) [4] successfully percolate the different barriers and arrive at the nuclear
waste level. Considering the scenario of canister failure, groundwater will then be in direct contact with the
spent fuel whose radioactivity will be limited at these timescales to alpha irradiation [2, 4]. Water in contact
with the waste will be then irradiated and various oxidants species (e.g. OH ∙ , H 2 O 2 , O 2 ...) will be formed near
the spent fuel surface (about 30-40 µm). This will lead to the oxidation of the spent fuel matrix, which is
principally formed of UO 2 [2, 3] (in its reduced form U(IV)), and the formation of oxidized uranium species
(U(VI)). These latter species are more soluble in water and may consequently lead to the migration of
radionuclides dissolved in the spent fuel matrix from the repository to the biosphere [3]. However, it is
known that the oxidation process may be limited by the presence of H 2, which is produced by water
radiolysis, due to its reducing properties [1]. Therefore, the knowledge of the chemistry of the solid-liquid
interface which controls the radionuclides release under the conditions encountered in deep geological
repositories is of great importance in performing assessment and safety studies.
In this work, the oxidation of UO 2 by He 2+ irradiation of water at the solid/solution interface is investigated
as a function of the dose under open and closed atmospheres. Irradiation is realized by the He 2+ beam of the
ARRONAX cyclotron with an energy of 66.5 MeV and a dose rate of 4.37 kGy/min. The aim of this
investigation is to determine the effect of the atmosphere and the dose on the UO 2 oxidation in order to
couple for the first time (1) characterization of the secondary oxidized phases, (2) quantification of H 2 O 2 and
H 2 produced by water radiolysis and (3) determination of the quantity of uranium released into the solution.
The kinetics of the solid surface oxidation are followed by the Raman spectroscopy. H 2 O 2 and H 2 are
measured respectively by UV-VIS spectrophotometry and micro Gas Chromatography (µ-GC). Inductively
Coupled Plasma Mass Spectrometry (ICP-MS) is used to quantify the soluble uranium species released into
the solution. Our results show that He 2+ irradiation of water induced oxidation of the UO 2 surface which
depended on the atmosphere and the dose. We present below the results obtained for two samples irradiated
363

at a dose of 8.73 kGy under the two atmospheres mentioned above. Figure 1 shows the Raman spectra of
unirradiated UO 2 (a) and a sample irradiated under open atmosphere (b, c, d). The spectra of the irradiated
sample are obtained 0.3, 42 and 169 hours after contact with the irradiated water. When comparing the two
spectra, we can see two identical Raman signals at 445 and 560 cm -1 . The first characterizes the U-O stretch
in the fluorite structure of UO 2 and the second indicates the presence of defects in the matrix [5]. After
irradiation, two Raman signals appear at 820 and 865 cm -1 . According to the literature [6], these vibration
bands indicate the presence of studtite (UO 4 4(H 2 O)). We have verified that this is not formed by oxidation
of the UO 2 surface by O 2 present in the atmosphere but by H 2 O 2 produced by water radiolysis. For the
irradiated sample, G(H 2 O 2 ) formed by water radiolysis is 3.00 μmol/J and the quantity of uranium species
released in the irradiated solution is 11.5 10 -7 mol/L. For the sample irradiated under closed atmosphere,
oxidation is much slower due to the reducing effect of H 2 produced by water radiolysis. In these conditions,
G(H 2 O 2 ) and G(H 2 ) are respectively 3.00 and 0.02 μmol/J. The concentration of Uranium species in the
solution is 9.3 10 -7 mol/L. Yields of H 2 obtained in this work are lower than those obtained by radiolysis of
pure water in similar experimental conditions [7]. Moreover, the concentrations of Uranium released into the
solution are comparable to those published in the literature [8]. H 2 O 2 production is similar under open and
closed atmosphere which allows us to conclude that in a closed atmosphere, H 2 produced by water radiolysis
reacts as a reducing agent and limits the oxidation process of UO 2 .
In conclusion, in this work we shine some light on the radiolytic corrosion of UO 2 by identification of (1) the
secondary phase formation and its kinetics, (2) the role of H 2 as a reducing agent, (3) the oxidative role of
H 2 O 2 and (4) the quantity of U species released. Furthermore, detailed mechanisms of UO 2
corrosion/oxidation will be proposed, taking into account the phenomena of water radiolysis.
(a) : unirradiated UO 2
(b) : 0.3 hours
(c) : 42 hours
(d) : 169 hours
(d)
(c)
(b)
(a)
Formation of Studtite
by oxidation of UO 2
Raman signals
of UO 2
Figure1 : Raman spectra of unirradiated UO 2 (a) and a sample irradiated at 8.73 kGy under open atmosphere
((b), (c) and (d)). The spectra (b), (c) and (d) are obtained 0.3, 42 and 169 hours after contact with the
irradiated solution.
Acknowledgements
The authors thank the National French Agency (ANR) for its financial support in the CISSRAD ALPHA
project. The research which led to these results has received funding from the European Atomic Energy
Community (Euratom) – Seventh Framework Programme FP7/2007-2011, under the grant agreement n°
295722 (FIRST-Nuclides project).
[1] P. Carbol, P. Fors, T. Gouder, K. Spahiu, Hydrogen suppresses UO 2 corrosion, Geochimica et Cosmochimica Acta,
73 (2009) 4366-4375.
[2] S. Sunder, D.W. Shoesmith, H. Christensen, N.H. Miller, Oxidation of UO 2 fuel by the products of gamma
radiolysis of water, Journal of Nuclear Materials, 190 (1992) 78-86.
364

[3] S. Röllin, K. Spahiu, U.B. Eklund, Determination of dissolution rates of spent fuel in carbonate solutions under
different redox conditions with a flow-through experiment, Journal of Nuclear Materials, 297 (2001) 231-243.
[4] M. Amme, B. Renker, B. Schmid, M.P. Feth, H. Bertagnolli, W. Döbelin, Raman microspectrometric identification
of corrosion products formed on UO 2 nuclear fuel during leaching experiments, Journal of Nuclear Materials, 306
(2002) 202-212.
[5] H. He, D. Shoesmith, Raman spectroscopic studies of defect structures and phase transition in hyper-stoichiometric
UO 2+x , Physical Chemistry Chemical Physics, 12 (2010) 8108-8117.
[6] A. Canizarès, G. Guimbretière, Y.A. Tobon, N. Raimboux, R. Omnée, M. Perdicakis, B. Muzeau, E. Leoni, M.S.
Alam, E. Mendes, D. Simon, G. Matzen, C. Corbel, M.F. Barthe, P. Simon, In situ Raman monitoring of materials
under irradiation: study of uranium dioxide alteration by water radiolysis, Journal of Raman Spectroscopy, 43 (2012)
1492-1497.
[7] F. Crumière, J. Vandenborre, R. Essehli, G. Blain, J. Barbet, M. Fattahi, LET effects on the hydrogen production
induced by the radiolysis of pure water, Radiation Physics and Chemistry, 82 (2013) 74-79.
[8] J. De Pablo, I. Casas, F. Clarens, F. El Aamrani, M. Rovira, The effect of hydrogen peroxide concentration on the
oxidative dissolution of unirradiated uranium dioxide, Materials Research Society, 663 (2001).
A4-2
REDOX REACTION OF SE(IV)/SE(VI) WITH NATURAL PYRRHOTITE AND IRON
SELENIDES
M. L Kang (1), (2) , B. Ma (1) , F. Bardelli (3) , F. R Chen (2) , C. L Liu (1), * , L. Charlet (3) , J.L Xie (1)
(1 )
Beijing National Laboratory for Molecular Sciences, Radiochemistry and Radiation Chemistry Key
Laboratory for Fundamental Science, College of Chemistry and Molecular Engineering, Peking University,
Beijing, 100871, China.
(2) Key Laboratory of Mineralogy and Metallogeny, Guangzhou Institute of Geochemistry, Chinese Academy
of Sciences, Guangzhou, 510640, China.
(3)
Environmental Geochemistry Group, ISTerre, University of Grenoble I, 38041 Grenoble, France.
Author for correspondence liucl@pku.edu.cn
The radioactive isotope 79 Se, with a half-life of 3.77 × 10 5 years, is presently considered as the key mobile
fission product in the disposal of spent fuel and high-level radioactive waste. Due to its high solubility and
mobility at high oxidation states (+4 and +6), reductive precipitation is considered as the most effective way
to immobilize 79 Se [1].
Pyrrhotite (Fe 1-x S, 0

elatively fast kinetics, while the reaction between Se(VI) and FeSe/FeSe 2 only occurs under limited
conditions (i.e. in the presence of high ferrous content and higher pH) with much slower kineticsTherefore,
reduction of Se(IV) by Fe(II)-bearing minerals, in particular by natural occurring minerals (e.g., pyrite,
pyrrhotite, siderite, etc.), are envisioned to produce Se(0) at the early stage of experiments, rather than FeSe
or FeSe 2 . Due to the formation of bulk Se(0) which has low solubility, the reaction is expected to maintain
redox disequilibrium in laboratory time-scale.
Furthermore, we also investigated the reactivity of FeSe 2 towards Fe 3+ under acidic conditions. The
results indicate that the reaction rate can be described as a second order reaction, with rate constants of 0.49,
0.85, 1.84, and 3.29 L.mol -1 .S -1 at pH 1.6, 1.8, 2.2, and 2.5, respectively. Aqueous Fe 2+ in a heterogeneous
system can reduce Se(IV) to Se(0) with relatively fast kinetics [11, 12], thus the strong reactivity of FeSe 2
towards Fe 3+ implies that ferric iron may play a significant role in FeSe 2 oxidation.
The findings in this study give insight into possible controls on the Se redox process, and can explain
some contrasting observations reported in the literature.
[1] Chen, F. R.; Burns, P. C.; Ewing, R. C. (1999). " 79 Se: geochemical and crystallo-chemical retardation
mechanisms." Journal of Nuclear Materials 275(1): 81-94
[2] Janzen, M. P.; Nicholson, R. V.; Scharer, J. M. (2000). "Pyrrhotite reaction kinetics: Reaction rates for oxidation
by oxygen, ferric iron, and for nonoxidative dissolution." Geochim Cosmochim Ac 64(9): 1511-1522
[3] Diener, A.; Neumann, T.; Kramar, U.; Schild, D. (2012). "Structure of selenium incorporated in pyrite and
mackinawite as determined by XAFS analyses." Journal of Contaminant Hydrology 133: 30-39
[4] Scheinost, A. C.; Kirsch, R.; Banerjee, D.; Fernandez-Martinez, A.; Zaenker, H.; Funke, H.; Charlet, L. (2008).
"X-ray absorption and photoelectron spectroscopy investigation of selenite reduction by Fe-II-bearing minerals."
Journal of Contaminant Hydrology 102(3-4): 228-245
[5] Scheinost, A. C.; Charlet, L. (2008). "Selenite reduction by mackinawite, magnetite and siderite: XAS
characterization of nanosized redox products." Environmental Science & Technology 42(6): 1984-1989
[6] Breynaert, E.; Bruggeman, C.; Maes, A. (2008). "XANES-EXAFS analysis of Se solid-phase reaction products
formed upon contacting Se(IV) with FeS 2 and FeS." Environmental Science & Technology 42(10): 3595-3601
[7] Breynaert, E.; Scheinost, A. C.; Dom, D.; Rossberg, A.; Vancluysen, J.; Gobechiya, E.; Kirschhock, C. E. A.;
Maes, A. (2010). "Reduction of Se(IV) in Boom Clay: XAS Solid Phase Speciation." Environmental Science &
Technology 44(17): 6649-6655
[8] Kang, M. L.; Chen, F. R.; Wu, S. J.; Yang, Y. Q.; Bruggeman, C.; Charlet, L. (2011). "Effect of pH on aqueous
Se(IV) reduction by pyrite. " Environmental Science & Technology 45(7): 2704-2710
[9] Curti, E.; Aimoz, L.; Kitamura, A. (2013). "Selenium uptake onto natural pyrite." Journal of Radioanalytical and
Nuclear Chemistry 295(3): 1655-1665
[10] Charlet, L.; Kang, M. L.; Bardelli, F.; Kirsch, R.; Géhin, A.; Grenèche, J. M.; Chen, F. (2012). "Nanocomposite
pyrite-greigite reactivity toward Se(IV)/Se(VI)." Environmental Science & Technology 46(9): 4869-4876
[11] Chakrabrty, S.; Bardelli, F.; Charlet, L. (2010). "Reactivities of Fe(II) on Calcite: Selenium Reduction."
Environmental Science & Technology 44(4): 1288-1294
[12] Charlet, L.; Scheinost, A. C.; Tournassat, C.; Greneche, J. M.; Gehin, A.; Fernandez-Martinez, A.; Coudert, S.;
Tisserand, D.; Brendle, J. (2007). "Electron transfer at the mineral/water interface: Selenium reduction by ferrous
iron sorbed on clay." Geochimica Et Cosmochimica Acta 71(23): 5731-5749
A4-3
INTERFACIAL RADIATION CHEMISTRY IN GEOLOGICAL REPOSITORIES FOR SPENT
NUCLEAR FUEL
Mats Jonsson
School of Chemical Science and Engineering, Applied Physical Chemistry, KTH Royal Institute of
Technology, SE-100 44 Stockholm, Sweden
Ionizing radiation is a key feature in the safety assessment of a geological repository for spent nuclear fuel.
The very presence of ionizing radiation can influence numerous material properties and processes and
thereby also the integrity of the barriers. In an intact repository of KBS-3 type the copper canister and the
cast iron insert are exposed to neutron and gamma radiation. The interface between the copper canister and
the compacted bentonite clay is also subject to neutron and gamma irradiation which will cause radiolysis of
366

any water present. In case of groundwater intrusion in a breached canister, radiolysis of water will produce
oxidants capable of inducing oxidative dissolution of the spent nuclear fuel matrix and subsequent release of
radionuclides into the groundwater.
While the radiation chemistry of water is a well-known subject, the radiation chemistry of interfacial systems
is still, to a large extent, unknown territory. 1 In this paper the interfacial radiation chemistry of interfaces and
radiation conditions of relevance in a deep geological repository for spent nuclear fuel is discussed.
It is evident that radiation induced redox processes such as oxidative dissolution of the fuel matrix 2 and
radiation induced corrosion of the copper canister 3 can be detrimental to the long term safety of the
repository. However, aqueous radiolysis products can also interact in other ways with the oxide surfaces
present in the system. Hydrogen peroxide can be catalytically decomposed on most oxide surfaces due to the
strong interaction between the oxide surface and the primary product, the hydroxyl radical. 4 The hydroxyl
radical has also been shown to have a high affinity for oxide surfaces. 5 Hence, the oxide surfaces will
become loaded with radicals displaying somewhat reduced reactivity compared to free radicals in solution.
Of particular interest is the reaction between adsorbed hydroxyl radicals and hydrogen which is a potential
route towards local reducing conditions on the surface exposed to ionizing radiation. Decomposition of
hydrogen peroxide and the reactivity of adsorbed hydroxyl radicals have been studied experimentally and by
using DFT calculations. The DFT calculations were performed using clusters of metal oxides and the
functionals B3LYP, B3LYP-D, B3LYP*, M06, M06-L, PBE0, PBE and PWPW91. 6 In the experiments, the
dynamics of H 2 O 2 consumption in oxide containing systems has been studied along with the formation of
adsorbed hydroxyl radicals. The latter was quantified experimentally by using radical scavengers. The
experimental results on the activation energies and the surface affinity are in qualitative agreement with the
results of the DFT calculations. Recent experimental studies have also shown that H 2 O 2 decomposition on
oxide surfaces can lead to formation of H 2 ; 7 this is however a minor reaction pathway. In addition, autoclave
experiments have shown that H 2 is decomposed to hydrogen atoms on some oxide surfaces.
For some processes it has been shown that aqueous radiation chemistry cannot account for all the chemical
changes observed experimentally. One example of this is radiation induced corrosion of copper under anoxic
conditions where aqueous radiation chemistry only has minor influence on the overall process. 8 In such
systems it is also necessary to account for the radiation energy deposited in the solid phase. This will be
discussed on the basis of recent experimental results and theory.
1. Mats Jonsson, ISRN Materials Science, Volume 2012, Article ID 639520
2. Eriksen, Trygve E.; Shoesmith, David W.; Jonsson, Mats, Journal of Nuclear Materials, 2012, 420, 409-423
3. Åsa Björkbacka, Saman Hosseinpour, Christofer Leygraf and Mats Jonsson, Electrochemical and solid-state letters,
2012, 15, C5-C7.
4. Lousada, Claudio M.; Jonsson, Mats, Journal of Physical Chemistry C, 2010, 114, 11202-11208
5. Lousada, Claudio M.; Johansson, Adam Johannes; Brinck, Tore; Jonsson, Mats, Journal of Physical Chemistry C,
2012, 116, 9533-9543
6. Lousada, Claudio M.; Johansson, Adam Johannes; Brinck, Tore; Jonsson, Mats, Phys. Chem. Chem. Phys., 2013,
15, 5539-5552
7. Lousada, Claudio M.; LaVerne, Jay A.; Jonsson, Mats, Phys. Chem. Chem. Phys. In press, DOI:
10.1039/c3cp51616d
8. Åsa Björkbacka, Saman Hosseinpour, Magnus Johnson, Christofer Leygraf, and Mats Jonsson, Rad. Phys. Chem.
In press, DOI: 10.1016/j.radphyschem.2013.06.033
A4-4
IMPACT OF FE MINERALS ON THE SAFE DISPOSAL OF NUCLEAR WASTE
Carolyn Pearce, Kevin Rosso, Juan Liu, Steven Luksic, Michael Schweiger, John McCloy and Andrew
Felmy
For the safe disposal of radioactive waste in a Geological Disposal Facility, it is necessary to predict
changes in oxidation states of redox active actinide elements and fission products (AFP), such as U and Tc,
which affect their chemical form and attendant risk of mobilization. Over the long time periods needed for
nuclear waste disposal, the chemical environment of the waste package in the subsurface could change in
367

complex ways across the entire continuum from oxidizing to reducing conditions. Both in the repository
near-field and in the far-field subsurface, the oxidation states and chemical form of AFP are closely tied to
changes in other redox-active components. Iron in particular pervades all aspects of the waste package
environment, from the steel in the waste containers, through corrosion products, to the iron minerals present
in the host rock. Variability in redox conditions leads to the expectation that common Fe forms such as the
(oxyhydr)oxides will undergo phase transformations or dissolution. These transformations can impact AFP
through electron exchange, adsorption, and co-precipitation reactions, presenting a challenging system to
understand and quantify. Traditional repository performance assessment models rely upon surface adsorption
or pure phase solubility experiments and do not encompass processes such as incorporation of AFP into Fe
mineral structures and how this impacts AFP residence and reactivity. In this paper, three case studies are
presented to highlight the use of model redox-active materials, as well as field samples, in combination with
high-resolution spectroscopic and microscopic capabilities, to interrogate bulk and interfacial reactivity of U
and Tc with Fe compounds.
The first case study is relevant to near-field environments, involving incorporation of Tc-99 into Febearing
spinel structures to increase its retention during production of nuclear waste glass. Currently, the
melter feed makeup and vitrification process can lead to volatilization of up to 60% of the Tc during glass
melting. Tc has multiple accessible oxidation states depending upon the oxygen partial pressure. Tc(VII) 2 O 7
melts at ~120°C and sublimates above ~300°C in oxygen So to increase the retention in the glass melt, Tc
was converted from KTc(VII)O 4 to Tc(IV)O 2 , which sublimates at ~1000°C. To stabilise the Tc in this
reduced state, it was incorporated into different spinel structures ( Fe 3-x Tc x O 4 , Fe 2-x NiTc x O 4 ) so that it would
be retained in the batch through the glass conversion process and dissolve slowly into the glass at high
temperatures > 950°C.
The second case study, relevant to near- and far-field environments, involves investigating the redox
reactivity of the mixed valence spinel iron oxides, magnetite (Fe 3 O 4 ) and titanomagnetite (Fe 3-x Ti x O 4 ), with
respect to reduction of the highly soluble pertechnetate anion [Tc(VII)O 4 - ] to sparingly soluble
[Tc(IV)O 2·H 2 O]. Sediments with basaltic provenance, such as those at the Hanford nuclear reservation,
Washington, U.S.A, are rich in Fe-bearing minerals of mixed valence. These minerals are redox reactive, and
have the potential to react with important environmental contaminants including Tc and U.
(Titano)magnetite nanoparticles (10 -12 nm) with varying Ti content (0 ≤ x ≤ 0.5) were produced, in aqueous
suspension under ambient conditions, as synthetic analogues having high reactive surface area for batch
studies. Tc(VII) reduction rates by (titano)magnetite nanoparticles were measured and compared with those
for natural, bulk titanomagnetites. The rate controlling factors in the heterogeneous reduction reaction,
including initial Tc(VII) concentration, Ti content, surface oxidation and availability of different Fe(II) pools
in the system were determined. (Titano)magnetite nanoparticles before and after reduction experiments were
characterized using micro X-ray diffraction (μ-XRD), X-ray absorption spectroscopy (XAS) and
transmission electron microscopy (TEM).
The third case study involves using etched silicon microfluidic pore network models (micromodels)
with controlled chemical and redox gradients, mineralogy, and microbiology under continuous flow
conditions to simulate complex far-field microenvironments. Colonization of micromodel pore spaces by an
anaerobic Fe(III)-reducing bacterial species (Geobacter sulfurreducens) and enzymatic reduction of a
bioavailable Fe(III) phase within this environment was demonstrated. X-ray microprobe and XAS were used
to investigate the combined effects of the precipitated Fe(III) phases and the microbial population on
uranium (U(VI)O 2 (NO 3 ) 2 ) biogeochemistry under flow conditions. Precipitated Fe(III) phases within the
micromodel were most effectively reduced in the presence of an electron shuttle (AQDS), and Fe(II) ions
adsorbed onto the precipitated mineral surface without inducing any structural change. In the absence of
Fe(III), U(VI) was microbially reduced to insoluble U(IV),which was precipitated in discrete regions
associated with biomass. In the presence of Fe(III) phases however, both U(IV) and U(VI) could be detected
associated with biomass, suggesting reoxidation of U(IV) by localized Fe(III) phases. These results
demonstrate the importance of the spatial localization of biomass and redox active metals, and illustrate the
key effects of pore-scale processes on contaminant fate and reactive transport.
These examples emphasize the importance of coupled fates of AFP with Fe mineralogy as the
repository environment evolves over time. To develop accurate predictive tools for geologic disposal
scenarios, substantial fundamental research remains to understand mechanistic reaction pathways for AFP in
368

these complex systems. Aspects highlighted here include; knowledge gaps and conceptual advances in
understanding reaction thermodynamic driving forces and reaction products, local chemical environment
effects on the stability of intermediate redox species, structural incorporation of U and Tc into Fe phases and
effects on phase stabilities, and the role of electron transfer in phase transformations.
SESSION 14
C3: DEVELOPMENT AND APPLICATION OF
MODELS
MODELLING OF Tc (IV) INTERACTION WITH DISSOLVED ORGANIC MATTER FROM
BOOM CLAY BY A KINETIC APPROACH
C. Bruggeman, A. Maes, N. Maes, E. Martens, J. Vancluysen, M. Van Gompel (Belgium)
RADIONUCLIDE SOLUBILITY CALCULATIONS IN CRYSTALLINE AND SEDIMENTARY
GROUNDWATERS FROM THE CANADIAN SHIELD
D. García, L. Duro, E. Colàs, V. Montoya (Spain)
DEVELOPING QUALITY ASSURED SORPTION DATABASES FOR PERFORMANCE
ASSESSMENT
E. Klein, S.M.L. Hardie, E.M. Scourse and I.G. McKinley (Switzerland, UK)
DIFFUSION OF CESIUM IN GRIMSEL GRANODIORITE: SIMULATIONS IN TIME DOMAIN
WITH HETEROGENEOUS SORPTION PROPERTIES
M. Voutilainen, P. Sardini , M. Siitari-Kauppi , A. Martin, J. Timonen (Finland, France,
Switzerland)
C3-1
C3-1
C3-2
C3-3
C3-4
MODELLING OF Tc(IV) INTERACTION WITH DISSOLVED ORGANIC MATTER FROM
BOOM CLAY BY A KINETIC APPROACH
C. Bruggeman 1) , A. Maes 2) , N. Maes 1) , E. Martens 1) , J. Vancluysen 2) , M. Van Gompel 1)
1) Belgian Nuclear Research Centre (SCK•CEN), Expert Group Waste & Disposal, Boeretang 200, B-2400
Mol, Belgium
2) Katholieke Universiteit Leuven, Laboratory for Colloid Chemistry, , B-3000 Leuven, Belgium
The redox-sensitive fission product technetium-99 is of great interest in nuclear waste disposal studies
because of its potential to contaminate the geosphere due to its very long half-life and high mobility. In the
case of Boom Clay, which is considered as a potential host rock in Belgium, the stable oxidation state of Tc
is +IV [1,2]. In batch Boom Clay suspensions, the equilibrium Tc(IV) solid-liquid distribution can be
explained by a combination of strong sorption to the solid mineral phases and complexation (or colloidcolloid
interaction) with dissolved organic matter (O.M.) [3]. However, in order to explain long-term
(sequential) migration experiments with Tc kinetic factors needed to be incorporated in order to simulate Tc
percolate concentrations [4].
In this study, we aim to model Tc(IV) geochemical behaviour in both batch sorption and column migration
experiments by a kinetic approach which is based on the Kinetically Controlled Availability Model
(KICAM), considering both fast and slow binding modes of Tc to the dissolved O.M. [5]. The kinetic model
was implemented into PhreeqC v2.18.
Experimental data was obtained initially from a set of batch experiments. Firstly, the long-term (up to 113
days) behaviour of the Tc interaction with dissolved Boom Clay O.M. was studied. It was observed that with
time interaction constants increased from logK = 5.3 to logK = 6.2. This increase corresponded kinetically to
a movement of Tc species from fast to slow binding modes, thereby providing new sorption opportunities for
the remaining Tc.
369

Then, (ir)reversibility effects concerning the interaction of Tc(IV) with dissolved Boom Clay O.M. were
investigated by examining newly installed equilibrium conditions upon interchanging of supernatants of “Tc
spiked” and “not-spiked” Boom Clay suspensions. Mobile dissolved Boom Clay O.M. was seen to be easily
replenished with Tc from “spiked” Boom Clay solid phase. The newly installed equilibrium corresponded to
original solid-liquid distributions indicating that adsorbed Tc(IV) is readily available for dissolved O.M. On
the other hand, Tc associated with dissolved Boom Clay O.M. was relatively unattainable for the solid phase
sinks, and desorption of Tc from this O.M. by competition with the solid Boom Clay phase was kinetically
hindered. From this set of experiments, the different kinetic parameters in the model could be determined.
The obtained kinetic interaction model was then upscaled to a reactive transport model which was used to
simulate both the outflow and the tracer profile in several long-term running percolation experiments (both in
lab and under in situ conditions). In these experiments, Tc was spiked between two Boom Clay cores, which
were percolated with Boom Clay pore water. The outflow of Tc from the 3.5 cm-long Boom Clay cores was
fairly rapid, and could be explained by the slow interactions of Tc(IV) with the mobile dissolved O.M.
fraction. However, only a small fraction of the initial Tc source was percolated. By contrast, most of the Tc
was retained near the source position due to strong sorption of inorganic Tc species and fast dissociation of
organic matter-associated Tc(IV). Due to continuous percolation of this available sorbed Tc(IV), a constantconcentration
outflow was observed in these experiments which could be well explained by the kinetic
model.
Finally, the model's predictive capabilities were tested by performing a blind prediction of a sequential
migration experiment [4]. In this experiment, a second clay core is mounted after the first one and both are
percolated by Boom Clay pore water. The outflow from the first column therefore serves as input source
(constant concentration boundary condition) for the second core.
[1] K. Geraedts, et al., Radiochim. Acta 90, 879 (2002)
[2] A. Maes, et al., Env. Sci. Technol. 38, 2044 (2004)
[3] A. Maes, et al., Env. Sci. Technol. 37, 747 (2003)
[4] N. Maes, et al., Phys. Chem. Earth 36, 1590 (2011)
[5] W. Schuessler, et al., Env. Sci. Technol. 34, 2608 (2000)
C3-2
RADIONUCLIDE SOLUBILITY CALCULATIONS IN CRYSTALLINE AND SEDIMENTARY
GROUNDWATERS FROM THE CANADIAN SHIELD
D. García 1) , L. Duro 1) , E. Colàs 1) 2), 1)
, V. Montoya
1) Amphos 21, Passeig de Garcia i Fària, 49-51, 1-1, 08019 Barcelona (Spain)
2) Karlsruhe Institute of Technology (KIT), Institute for Nuclear Waste Disposal (INE), Hermann-von-
Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)
In a geological repository for spent nuclear fuel, containers are designed to prevent groundwater from getting
in contact with the fuel. In the case of a defective container water may contact the fuel and hence
radionuclides might be mobilized. The assessment of radionuclide solubility limits in the near field of a
repository for used fuel is essential for the safety demonstration of the disposal system. The objective of this
work is then the study and the update of the solubility of 19 radionuclides (Am, As, Bi, C, Cu, Mo, Nb, Np,
Pa, Pb, Pd, Pu, Ra, Se, Sn, Tc, Th, U and Zr) in two different groundwater compositions, considered as
representative groundwaters expected in the geological formation likely to host a future Canadian geological
repository: the saline crystalline groundwater (CR-10) and the highly saline sedimentary groundwater (SR-
270), provided by NWMO (Nuclear Waste Management Organization) [1]. Solubility values for several
radionuclides of interest for Canada were calculated 15-20 years ago based on a range of water chemistries
from Canadian Shield crystalline rock data and using thermodynamic data available at that time [2].
Solubility calculations have been carried out with groundwaters equilibrated with minerals present at
repository depth (CR-10 eq and SR-270 eq) and groundwaters equilibrated with the bentonite buffer and the
carbon steel insert of the copper container (CR-10 NF and SR-270 NF). The main assumptions in these
calculations are: i) the system temperature is fixed at 25ºC, ii) neither sulfate to sulfide reduction nor sulfide
to sulfate oxidation is considered, iii) reduction of carbonate to methane and reduction of nitrate to
370

ammonium or N 2 (g) are not considered , and iv) the basic Ostwald’s principle [3], i.e. the least crystalline
phases are kinetically favored to precipitate, is assumed.
The Yucca Mountain Pitzer (YMP) database [4] has been used in the solubility assessment of Am, C, Cu,
Np, Mo, Pu, Tc, Th, U and Zr. ThermoChimie [5], the thermodynamic database of ANDRA (The French
National Radioactive Waste Management Agency), has been used to compare the results obtained with the
Yucca Mountain Pitzer database. The ThermoChimie database uses SIT (Specific Ion Interaction Theory) for
ionic strength corrections and its most up-to-date version is released as sit.dat with the PhreeqC code [6]. The
SIT database has also been used for the solubility assessment of the elements not included in the Yucca
Mountain Pitzer database (As, Bi, Nb, Pa, Pb, Pd, Ra, Se, Sn and Zr). The solubility assessment done for the
elements contained in both databases led, except in a few cases, to differences in the solubility values below
± 1 logarithmic unit for all elements. Fig.1 shows an example of the work done for Th.
-4
1.0
-4
1.0
-5
Th(OH) 4 (aq)
Th(OH) 2 (CO 3 ) 2
2-
0.9
0.8
-5
Th(OH) 4 (aq)
Th(CO 3 ) 5
6-
0.9
0.8
0.7
0.7
log [Th]T (M)
-6
-7
Th(OH) 3 (CO 3 ) -
0.6
0.5
0.4
Fraction
Th(OH)(CO 3
-9
log [Th]T (M)
-6
-7
Th(OH) 3
+
0.6
0.5
0.4
Fraction
Th(OH) 3
+
-8
Th(OH) 2 (CO 3 )(aq) 0.2
0.1
-9
0.0
-5 -4 -3 -2 -1
log [CO 3 2- ] T (M)
0.3
0.3
-8
0.2
0.1
0.0
-5 -4 -3 -2 -1
log [CO 2- 3 ] T (M)
a) b)
Fig. 1. Thorium solubility (black solid line) and underlying thorium aqueous speciation (dotted lines) as a function of
carbonate concentration in solution at pH = 7.06. Calculations have been made using the a) ThermoChimie SIT
database, and b) Yucca Mountain Pitzer (YMP) database.
Sensitivity analyses, studying the influence of some key parameters (pH, Eh, iron, carbonate, calcium,
chloride, phosphate concentrations, sulfate reduction to sulfide, and carbonate reduction to methane) on the
solubility of radionuclides, and qualitative uncertainty analysis, describing the conceptual uncertainties that
may affect the solubility limits of a radionuclide, are also provided.
[1] L. Duro, V. Montoya, E. Colàs, D. García. NWMO TR-2010-02 (2010)
[2] R.J. Lemire and F. Garisto. Report AECL-10009. (1989)
[3] W. Ostwald. Zeitschrift für Physikalische Chemie, 22, 289–330 (1897)
[4] C. Jove-Colon, T. Wolery, J. Rard, A. Wijesinghe, R. Jareck, K. Helan. Report ANL-EBS-MD-000045 REV 03
(2007)
[5] L. Duro, M. Grivé, E. Giffaut. MRS Proceedings, 1475 (2012)
[6] D. Parkhurts and T. Appelo. USGS Report 99-4259 (1999)
C3-3
DEVELOPING QUALITY ASSURED SORPTION DATABASES FOR PERFORMANCE
ASSESSMENT
E. Klein (1) , T. Beattie (2) , S.M.L. Hardie (1) , E.M. Scourse (2) and I.G. McKinley (1)
(1) MCM Consulting, Baden-Dättwil, Switzerland
(2) MCM Consulting, Bristol, UK
The release and transport of radionuclides in groundwater is usually a focus of the safety cases for geological
disposal concepts and hence the models used to quantify these processes in performance assessments are
particularly important. This is a particular challenge when a volunteering approach is taken to site selection
(e.g. in the UK, Japan or Canada), requiring rapid assessment of the pros and cons of different sites or
371

different formations available at a single site. To assess the current state of the art, the procedures used to
develop the associated sorption databases (SDBs) utilised in recent major safety cases for deep geological
disposal of higher activity wastes at specific sites have been critically reviewed in order to identify lessons
that could be learned. Such reviews focused, in particular, on the transparency / logic of distribution
coefficient (Kd) selection, consistency and associated Quality Assurance (QA).
Some of the key issues identified are associated with the general problem of integration of the huge volumes
of site-specific, multi-disciplinary information that comprise a modern safety case. There are clear signs that
traditional approaches, involving compilation of a hierarchy of documents, begin to show their limitations as
the teams involved in their production become larger, with more specialists and fewer multi-disciplinary
generalists. This often results in difficulties in maintaining both consistency and technical quality control.
For the particular case of quantification of sorption in a fracture-flow system, it is evident that progress over
the last 3 decades has been very limited and attempts to apply a more formal approach to SDB development
have been greatly constrained by the difficulties of sampling and characterising the heterogeneous solute
transport paths present.
For more homogeneous argillaceous rocks in which diffusive solute transport dominates, there have been
significant advances in both understanding the mechanisms of sorption and developing a structured approach
to SDB production. Nevertheless, even here the rigorous application of mechanistic understanding is limited
in terms of the elements to which it can be applied: effectively only for very well studied cations with simple
speciation such as Sr or, with caveats, Cs. Sensibly, there is a close coupling of thermodynamic modelling to
derive solubilities and speciation with the assessment of sorption. Nevertheless, such parameters can be
considered only semi-quantitative indicators for most elements due to limitations of the assumption of
equilibrium applied to many redox systems, the poor quality of required thermodynamic databases
(especially at higher pH) and a general lack of model verification / validation.
A fundamental source of many of the problems identified is the occasional disconnect between the SDB
compilers and the safety assessors, which can lead to inconsistencies in the boundary conditions assumed for
compilation and those for the models in which they are used. Regardless of how the SDB is developed, this
leads also to inconsistent and incomplete quantification of associated model and database uncertainties. Error
propagation to determine the uncertainties on measured sorption data is straightforward, but testing that lab
data are appropriate to the actual geological systems of interest is much trickier. Analogues are often
mentioned in this regard, but their rigorous use has yet to be demonstrated.
The paper will provide a summary of the review, the main lessons learned and discuss how alternative
approaches to safety case documentation (based on knowledge management tools) may help to reduce the
problems associated with production and QA of SDBs. Such approaches can certainly remove problems of
consistency between different documents (as key input is produced once only and hyperlinked to any other
areas where it is needed) and should improve quality and consistency due to the rigorous quality and change
management procedures which can be readily implemented. In addition, by highlighting fundamental
technical problems rather than those caused by the logistics of information management, future research and
development can be better focused and limited resources better utilised.
C3-4
DIFFUSION OF CESIUM IN GRIMSEL GRANODIORITE: SIMULATIONS IN TIME DOMAIN
WITH HETEROGENEOUS SORPTION PROPERTIES
M. Voutilainen (1) , P. Sardini (2) , M. Siitari-Kauppi (1) , A. Martin (3) , J. Timonen (4)
(1) Department of Chemistry, University of Helsinki, P.O. Box 55, 00014 University of Helsinki, Finland
(2)
IC2MP, University of Poitiers, UMR 7285 CNRS, France
(3) Nagra, Hardstrasse 73, 5430 Wettingen, Switzerland
(4) Department of Physics, University of Jyväskylä, P.O. Box 35, 40014 University of Jyväskylä, Finland
In many countries crystalline rock has been chosen to be the host medium for underground repositories of
highly radioactive spent nuclear fuel. For performance assessment and safety analysis of such repositories, it
372

is vital to determine properties linked to migration of radioactive elements in the surrounding geosphere. To
this end, the underground rock laboratory at Grimsel test site in Switzerland provides an opportunity to
develop and test equipment, methodology and models in in-situ conditions which are analogous to those at
repository sites. Some parameters are difficult or impossible to determine in complicated in-situ conditions,
and thus laboratory studies and numerical simulations are often needed when setting the parameters, as well
as when analyzing the results. Some of these features are effects of mineral and structure heterogeneities.
In general, heterogeneity of the rock may have a considerable influence on the overall properties. In addition,
these heterogeneities may also be present in multiple length scales in crystalline rock. In this work we
studied the effect of centimeter scale heterogeneity on diffusion and sorption properties of cesium using time
domain diffusion (TDD) simulations [1]. TDD is a rapid particle-tracking method that gives an opportunity
to simulate diffusion in heterogeneous materials when the local porosities and diffusion coefficients are
known. In our earlier study the TDD method was applied for studying diffusion in heterogeneous porous
rock, and a clear difference was found with the corresponding homogeneous case [2].
In order to determine the heterogeneous structure of a Grimsel granodiorite sample, we applied X-ray
computed microtomography, the 14 C-labeled-polymethylmethacrylate ( 14 C-PMMA) method and mineral
identification [3]. As a result we obtained a 3D mineral map of the sample showing distribution of quartz,
feldspar and biotite (see Fig. 1a). This map can also be considered as a porosity and apparent diffusion
coefficient map when these quantities are known for each mineral. We determined a porosity of 0.67 % for
quartz, 0.67 % for feldspar and 0.88 % for biotite, and used earlier determined distribution coefficients for
cesium: 0.008, 0.047 and 12 m 3 /kg, respectively [4, 5]. The mineral specific apparent diffusion coefficients
were determined from these values, and for the homogeneous reference sample we used the volume
weighted average of these values.
a)
5mm
b)
Figure 1. a) 3D visualization of the Grimsel granodiorite sample (quartz (dark gray), feldspar (gray), biotite
(light gray)) used in sorption simulations. b) Effect of heterogeneous sorption properties on the in-diffusion
of cesium. Implementing the heterogeneous sorption properties (red) to simulations decreased the apparent
diffusion coefficient by 27 % compared to the sample with homogeneous sorption (blue).
Simulations were done as in-diffusion simulations in which particle concentration was kept constant on one
face of the sample, and it was then determined as a function of distance from that face. Determined indiffusion
profiles were fitted by a solution to the corresponding diffusion equation using apparent diffusion
coefficient as the fitting parameter [2]. As a result we obtained apparent diffusion coefficients of 1.5×10 -14
m 2 /s and 1.9×10 -14 m 2 /s for the heterogeneous and homogeneous case, respectively (see Fig. 1b).
Consequently, heterogeneous sorption properties retard diffusive migration by 27 %, and thus make a
relevant contribution to diffusion of cesium in Grimsel granodiorite. Simulation results were compared to
those of in-situ diffusion experiments on cesium [5]. The distribution coefficient determined from in-situ
experiment (8×10 -3 m 3 /kg) seems to be a factor of two lower than the average distribution coefficient from
simulation (15×10 -3 m 3 /kg).
373

[1] Delay, F., Porel, G., and P. Sardini, P. (2002). “Modelling diffusion in a heterogeneous rock matrix with a timedomain
Lagrangian method and an inversion procedure.” C. R. Geosci. 334(13): 967-973
[2] Voutilainen, M., Sardini, P., Siitari-Kauppi, M., Kekäläinen, P., Aho, V., Myllys, M., and Timonen, J. (2013).
“Diffusion of tracer in altered tonalite: experiments and simulations with heterogeneous distribution of porosity.”
Transport Porous Med. 96(2): 319-336
[3] Voutilainen, M., Siitari-Kauppi, M., Sardini, P., Lindberg, A., and Timonen, J. (2012). “Pore-space characterization
of an altered tonalite by X-ray computed microtomography and the 14C-labeled-polymethylmethacrylate method.” J.
Geophys. Res. 117(B1): B01201
[4] Pinnioja, S., Jaakkola, T., and Miettinen, J. K. (1983). ”Comparison of batch and autoradiographic methods in
sorption studies of radionuclides in rock and mineral samples.” MRS Proceedings 26: 1099-1105
[5] Jokelainen, L., Meski, T., Lindberg, A., Soler, J. M., Siitari-Kauppi, M., Martin, A., and Eikenberg, J. (2012). ”The
determination of 134 Cs and 22 Na diffusion profiles in granodiorite using gamma spectroscopy.” J. Radioanal. Nucl.
Chem. 295(3): 2153-2161
SESSION 15
A5: SOLID-LIQUID INTERFACE REACTIONS
THE BIGRAD CONSORTIUM - ALTERATION OF SANDSTONE BY ALKALINE CEMENT
LEACHATES AND ITS EFFECT ON URANIUM SPECIATION
E.B.A. Moyce, K. Morris, T. A. Marshal, S. Shaw (UK)
99-TECHNETIUM BEHAVIOR UPON INTERACTIONS WITH HEMATITE
N.A. Wall, L.C. Gribat (USA)
A5-5
A5-6
DEVELOPMENT AND TESTING OF COMPONENT ADDITIVITY SURFACE
COMPLEXATION MODELS OF NEPTUNIUM AND PLUTONIUM SORPTION TO SOILS
B.A. Powell, S. Herr, N. Conroy, J. Wong, Yu Xie, M. Zavarin, A.B.
Kersting (USA)
SORPTION OF Cs(I) AND Sr(II) ON SMECTITE RICH CLAY: BATCH SORPTION AND
MODELLING
S. Kasar, A.S. Kar, S. Kumar, A.S. Pente, C.P. Kaushik, R.K. Bajpai, B.S. Tomar (India)
A5-7
A5-8
A5-5
THE BIGRAD CONSORTIUM – ALTERATION OF SANDSTONE BY ALKALINE CEMENT
LEACHATES AND ITS EFFECT ON URANIUM SPECIATION.
E.B.A. MOYCE 1 , K. MORRIS 2 , T. A. MARSHALL 2 AND S. SHAW 2
1
Cohen Geochemistry Group, School of Earth and Environment, University of Leeds, Leeds, LS2 9JT
2
Williamson Research Centre, Research Centre for Radwaste and Decommissioning, School of Earth,
Atmospheric and Environmental Sciences, University of Manchester, Manchester, M13 9PL
The interaction of groundwater with cement in a Geological Disposal Facility (GDF) for intermediate level
nuclear waste will result in the formation of a high pH leachate plume upon groundwater resaturation of the
GDF. This plume will react with the surrounding host rock leading to the formation of a Chemically
Disturbed Zone (CDZ) in the area surrounding the GDF. This will cause significant changes in the physical
(e.g. porosity) and chemical (e.g. sorption capacity) properties of the host rock. The situation will be
complicated by the evolution of cement leachate composition over time as different components of the
cement dissolve during the lifetime of the GDF (Atkinson 1985, Berner 1992). Initially a KOH and NaOH
dominated leachate forms with pH ~ 13. At this point the leachate is also in equilibrium with Ca(OH) 2 , but
due to the low solubility of this phase at high pH, the concentration of Ca 2+ in solution will be relatively low.
As these phases are consumed the leachate will become Ca(OH) 2 dominated and buffer to pH ~ 12.5. After
the Ca(OH) 2 supply has been exhausted, Calcium Silicate Hydrate (C-S-H) dissolution will dominate and pH
will buffer at ~ 10.5 until finally, the buffering capacity of the cement will become exhausted. In this study,
model leachates from these different stages are termed Young, Intermediate and Old Cement Leachates
(YCL, ICL and OCL respectively). Previous studies of high pH mineral alteration have found that generally
aluminosilicate minerals undergo dissolution. These dissolved components then react with Ca 2+ /K + in
374

solution to form secondary C/K-(Al)-S-H phases (Gaucher and Blanc, 2006, Savage et al., 1992, Savage
2011). Alteration of the host rock in this way may have an impact (positive or negative) on its ability to
retard radionuclide migration. For example, the high surface area of the secondary phases may enhance
radionuclide sorption and/or radionuclides may become incorporated (i.e. coprecipitation) into the structure
of secondary mineral phases during their formation.
Here, we have investigated the alteration mechanism of sandstone in cement leachates, and the effect of this
process on the uptake and speciation of uranium. Batch experiments were performed to react samples of
disaggregated Hollington Red Sandstone (which is a model rock containing common rock forming minerals)
with cement leachates representing YCL, ICL and OCL both with and without uranium, as U(VI), present.
Experiments have been reacted for 1 year with samples taken at intervals throughout for analysis. The
reacted fluids show that in all three leachate systems without uranium, Si and Al are released and Ca is
removed from solution over the course of the experiment. Characterisation of the solid samples using
electron microscopy indicates dissolution of rock grain surfaces in the YCL and OCL systems, though there
is sparse evidence for formation of any secondary phases. By contrast, in the ICL system by 12 weeks of
reaction, sandstone grains are coated in a secondary C-S-H phase with a sheet-like morphology. In this fluid
system there is also an abundant secondary Ca-Al-O phase (hydrocalumite), present as euhedral, hexagonal
plates, which appears to be overgrown by the same C-S-H phase as that was identified coating the sandstone
grains. Analyses of the fluids taken from parallel leachate experiments in the presence of U(VI) show the
same trends in Si, Al and Ca as the non-uranium experiments and give confidence that these systems are
responding in a similar fashion with time. ICP-MS analysis of the uranium in solution shows that ~ 90% of
the 10 ppm U(VI) spike remains in solution during the YCL experiment and appears unaffected by the
sandstone alteration. In both the ICL and OCL experiments conducted at 1 ppm uranium, uranium is lost
from solution within the first few weeks of reaction. However, EXAFS analyses of the solids from these
experiments show that the uranium is present in different bonding environments indicating that different
mechanisms are controlling the radionuclide speciation in the two systems. This data will be discussed in
terms of mechanisms of alteration, radionuclide behaviour and implications for geological disposal.
ATKINSON, A. 1985. The Time Dependence of pH Within a Repository for Radioactive Waste Disposal. UKAEA,
AERE-R 11777.
BERNER, U. R. 1992. Evolution of Pore Water Chemistry During Degradation of Cement in a Radioactive Waste
Repository Environment. Waste Management, 12, 201-219.
GAUCHER, E. & BLANC, P. 2006. Cement/clay interactions – A review: Experiments, natural analogues, and
modeling. Waste Management, 26, 776-788.
SAVAGE, D., BATEMAN, K., HILL, P., HUGHES, C., MILODOWSKI, A., PEARCE, J., RAE, E. & ROCHELLE, C. 1992. Rate
and mechanism of the reaction of silicates with cement pore fluids. Applied Clay Science, 7, 33-45.
SAVAGE, D. 2011. A review of analogues of alkaline alteration with regard to long-term barrier performance.
Mineralogical Magazine, 75(4), 2401-2418.
A5-6
99 TECHNETIUM BEHAVIOR UPON INTERACTIONS WITH HEMATITE
Nathalie A. Wall and Larissa C. Gribat
Department of Chemistry, Washington State University, Pullman, WA 99164-4630
Technetium-99 ( 99 Tc) (t 1/2 = 2.13∙10 5 y), a notable fission product, represents a significant environmental
contaminant at several nuclear disposal sites (e.g. Hanford, US 1 , Sellafield, UK 2 ). In aqueous solution, the
valence state of Tc can vary from III to VII 3 , but Tc(VII) (as pertechnetate oxyanion, TcO 4 - ) is considered
the most stable under natural oxic conditions and highly mobile in the subsurface environment due to high
solubility in water and essentially nonadsorptive properties towards minerals. Yet, Tc(VII) can be reduced by
biogenic Fe(II) to produce relatively immobile and sparingly soluble Tc(IV) species (e.g. 1) – however,
complexing organic ligands, including naturally ubiquitous humic acids, can dramatically increase Tc(IV)
solubility 4-6 – and the resulting Tc(IV) may be retained by the mineral solids. Fe can influence Tc redox,
even if Tc is embedded inside a matrix; for example Fe present in bentonite reduces Tc to Tc(IV) as it
leaches from spent UO 2 fuel 7 . Conversely, Fe and Fe-bearing minerals can also oxidize part of Tc(IV) to
Tc(VII), while retaining another fraction of Tc(IV).
375

This work presents experimental data for the sorption of Tc(IV) on hematite (Fe 2 O 3 ) and the oxidation of
Tc(IV) to Tc(VII) upon such interaction. Hematite was synthesized and characterized in our laboratory;
Tc(IV) stock solution was prepared according to the literature 8 . Trace concentrations of Tc(IV) or Tc(VII)
were added to batch samples containing controlled size fractions of hematite. Solution pH and ionic strength
were controlled. Tc sorption on hematite was quantified upon equilibration, using ultracentrifugation for
phase separations and liquid scintillation counting for 99 Tc measurements. The concentrations of the IV and
VII Tc oxidation states were quantified using a procedure recently developed by this group, by which
iodonitrotetrazolium chloride isolates Tc(VII) in chloroform, but not Tc(IV) 9 . All experiments were
conducted in an oxygen-free glove box (N 2 99.998%) to maintain Tc in its reduced state.
For experiments initiated with 100% Tc(VII), no Tc sorption was recorded and no Tc(IV) was observed in
the samples. For experiments initiated with 100% Tc(IV) at pH 3.5, 50% of Tc had oxidized to Tc(VII) upon
equilibration with 114 ppm hematite, while only 50% of the original Tc remained as Tc(IV). No Tc(VII) was
found sorbed to hematite. However, not all remaining Tc(IV) was found sorbed to hematite; 65% of the
remaining Tc(IV) was retained by hematite. This work will show the influence of hematite concentration and
solution pH on the distribution of Tc oxidation state and its sorption on the mineral.
[1] J. K. Fredrickson, J. M. Zachara, J. M., Kennedy, D. W., Kukkadapu, R. K., McKinley, J. P., Heald, S. M., Liu, C.
X., and Plymale, A. E. Geochim. Et Cosmochim. Acta 68 3171 (2004)
[2] A. Aarkrog, Topical Studies in Oceanography 50 2597 (2003)
[3] J. A. Rard, J of Nuc and Radiochem. Sc. 6 197 (2005)
[4] M. A. Boggs, W. Dong, B. Gu, N. A. Wall, Radiochim. Acta 98 583 (2010)
[5] M. A. Boggs, T. Minton, W. Dong, S. Lomasney, M. R. Islam, B. Gu, N. A. Wall, Environ. Sci. Technol. 45
2718 (2011)
[6] B. Gu, W. Dong, L. Liang, N. A. Wall, Environ. Sci. Technol. 45 4771 (2011)
[7] H. Rameback, Y. Albinsson, M. Skalberg, U. B. Eklund, L. Kjellberg, L. Werme, L., J. of Nuc. Mat. 277 288 (2000)
[8] N. Hess, O. Qafoku, Y. Xia, A. Felmy, Abstracts of Papers of the American Chemical Society 234 (2007)
[9] M. A. Boggs, L. C. Gribat, C. A. Boele, N. A. Wall, J. of Radioanal. and Nuc. Chem. 293 843 (2012)
A5-7
DEVELOPMENT AND TESTING OF COMPONENT ADDITIVITY SURFACE COMPLEXATION
MODELS FOR NEPTUNIUM AND PLUTONIUM SORPTION TO SOILS
Brian A. Powell 1 , Sarah Herr 1 , Nathan Conroy 1 , Jennifer Wong 1 , Yu Xie 1 , Mavrik Zavarin 2 , Annie B.
Kersting 2
1. Environmental Engineering and Earth Sciences, Clemson University, 342 Computer Court, Anderson,
South Carolina 29625, United States
2. Glenn T. Seaborg Institute, Physical & Life Sciences, Lawrence Livermore National Laboratory, POS Box
808, L-231, Livermore, CA 94550
In this work, we have compiled aqueous complexation constants and surface complexation reactions
describing Np(V), Pu(V), and Pu(IV) aqueous and surface complexation reactions. We used the compiled
database to predict sorption of neptunium and plutonium to soils from the United States Department of
Energy Savannah River Site (SRS) and the Hanford Site. Data from batch sorption experiments describing
Np(V), Pu(V), and Pu(IV) sorption to the pure mineral phases gibbsite, quartz, goethite, kaolinite, and
montmorillonite were used to generate redox coupled surface complexation constants using a two-pKa
double layer model. Many mineral surfaces have been shown to facilitate reduction of Pu(V) to Pu(IV), even
in the absence of an obvious reductant within the mineral. Therefore, oxidation state analysis of Pu was
confirmed using indirect solvent extraction techniques, in situ X-ray absorption near edge spectroscopy
(XANES) and electron microscopy. Data for complexation with inorganic ions was taken from the NEA
thermochemical database 1 . Complexation with Leonardite humic acid (LHA) and Suwannee River fulvic
acid (SRFA) were determined using solvent extraction, ultrafiltration, and solubility studies. Data for the
complexation with citric acid (CA) and desferrioxamine B (DFOB) were taken from available literature 2,3 .
These organic ligands were chosen to represent a wide range of NOM characteristics including molecular
weight, functionality, and hydrophobicity. Batch sorption experiments were used to quantify the influence of
376

natural organic matter on sorption of Pu(IV) to pure mineral surfaces. These studies were supported by
Fourier transform infrared spectroscopic (FTIR) characterization of the mode of interaction between the
ligands and the mineral surfaces.
Batch sorption experiments using soils from DOE sites were conducted using Np(V), Pu(V), and
Pu(IV) as the starting oxidation states. Reduction of Pu(V) to Pu(IV) was found to occur on both SRS and
Hanford soils. Under similar conditions reduction of Np(V) was not observed. Addition of known reductants
such as ferrous iron, ascorbic acid, and dithionite to SRS soil suspensions containing Np(V) did not cause
significant increases in sorption. Therefore, it appears even strong reductants were unable to reduce Np(V)
under oxic conditions in the presence of the SRS soil. Direct (XANES/EXAFS) and indirect (solvent
extraction) oxidation state analyses of Pu in batch sorption experiments with a variety of solid phases present
indicate that the distribution of Pu is dominated by solid phase Pu(IV) and aqueous phase Pu(V). The
thermochemical database described above was used to predict the sorption behavior using a component
additivity approach assuming various fractions of the reactive iron, silica, and alumina sites were active. The
model provided good predictions of the observed sorption data and was also able to capture the reduction of
Pu(V) to Pu(IV) by predicting dominance of Pu(V) in the aqueous phase and Pu(IV) on the solid phase.
Overall sorption to iron oxide surfaces appears to be the dominant sorption reaction despite the soils having
relatively different mineralogy.
1. Guillaumont, R., et al., Update on the Chemical Thermodynamics of Uranium, Neptunium, Plutonium, Americium,
and Technetium, in Chemical Thermodynamics, O.N.E. Agency, Editor. 2003, Elsevier: Amsterdam.
2. Boukhalfa, H., S.D. Reilly, and M.P. Neu, Complexation of Pu(IV) with the natural siderophore desferrioxamine B
and the redox properties of Pu(IV)(siderophore) complexes. Inorganic Chemistry, 2007. 46(3): p. 1018-1026.
3. Boukhalfa, H., et al., EDTA and mixed-ligand complexes of tetravalent and trivalent plutonium. Inorganic
Chemistry, 2004. 43(19): p. 5816-5823.
A5-8
SORPTION OF CS(I) AND SR(II) ON SMECTITE RICH CLAY: BATCH SORPTION AND
MODELLING
S. Kasar 1 , A. S. Kar 1 , S. Kumar 1 , A. S. Pente 2 , C. P. Kaushik 2 , R. K. Bajpai 3 , B. S. Tomar 1
1 Radioanalytical Chemistry Division, 2 Waste Management Division,
3 Technology Development Division
Bhabha Atomic Research Centre, Mumbai-400085, India
Study on the migration of actinides and long lived fission products points to the far-field area in the
environment as an important component in the safety analysis of the deep underground disposal of high level
nuclear waste. Smectite rich natural clay constitutes the engineered barrier in the repository. Interaction of
more mobile fission products with the clay is thus an important aspect of the transport through the clay
matrix. In the present study, sorption of Cs(I) and Sr(II) on smectite rich natural clay was investigated under
varying pH and ionic strength conditions and the data has been modeled to identify the chemical interactions
involved in the sorption process.
Smectite rich natural clay was collected from the western part of India. The physico-chemical
characteristics of the clay are given in table 1. Sorption of Cs(I) and Sr(II) were carried out in batch mode
and percentage retention was determined radiometrically. ~10 -9 M Cs(I) solution in 0.1 N NaCl was placed in
contact with Na-equilibrated clay over pH 3-10 for 48 hrs. After equilibration the suspensions were
centrifuged and 137 Cs activity was assayed using an NaI(Tl) detector. In the Sr(II) experiment ~10 -5 M was
the metal ion concentration used and liquid scintillation counting was used to find the percentage sorption.
Error on the percentage sorption represents the deviations in duplicate measurements.
Modeling of the sorption data was carried out using FITEQL software. Only one ion exchange site
was considered to be involved in the sorption process and competition of Cs + and H + was considered for
surface ion exchange at this site in the modeling process.
377

Figures 1 and 2 show the Cs(I) and Sr(II) sorption data over varying pH and ionic strength. The
significant effect of ionic strength on the sorption percentage is an obvious indicator of the dominance of ion
exchange as the sorption mechanism. This was further corroborated with the lower sorption percentage at
acidic pH values where concentrations of H+ with higher orders of magnitude give strong competition to
Cs+ sorption. Well reproduced modelling results (solid lines in the plots) prove these conclusions. Modelling
results have been given in table 2. Selectivity coefficients are comparable to those in the literature reports [1,
2].
[1] Guo, Z., Xu, J., Shi, K., Tang, Y., Wu, W., Tao, Z., 2009. Eu(III) adsorption/desorption on Na-bentonite:
Experimental and modelling studies. Colloids Surf. A: Physicochem. Eng. Aspects 339, 126-133.
[2] Bradbury, M. H., Baeyens, B., 1999. Modelling the sorption of Zn and Ni on Ca-montmorillonite. Geochem.
Cosmochim. Acta 63, 325-336.
Table 4. Physicochemical characteristics of the natural clay
Characteristics
Value
Surface area (N 2 -BET method) (m 2 g -1 ) 79.6
Pore volume (cc/g) 0.087
Cation exchange capacity (meq/100 g) 76
Total carbon content < 0.05%
Total organic carbon < 0.05%
Composition (by X-ray diffraction), (wt %)
Montmorillonite (~ 90), Kaolinite (~ 10), Silica
(Trace)
Table 2. Modelling results
Species Log K
XH 1.88 ± 0.2
XCs 3.40 ± 0.09
X 2 Sr 4.96 ± 0.04
100
95
0.05 M 0.1 M 0.5 M 1 M
Model Fit
100
90
0.05 M 0.1 M 0.5 M
1 M Model Fit
90
80
85
70
% Sorption
80
75
70
% Sorption
60
50
40
65
30
60
20
55
10
50
2 3 4 5 6 7 8 9
pH
Fig 1. Cs(I) sorption on natural clay
0
2 3 4 5 6 7 8 9 10 11
pH
Fig 2. Sr(II) sorption on natural clay
378

SESSION 16
A2: SOLID SOLUION AND SECONDARY
PHASE FORMATION
URANIUM OXIDATION AND MIGRATION THROUGH CRYSTALLINE ROCK -
TRANSPORT PATHWAYS AND THE ROLE OF SECONDARY PHASES IN RETARDATION
D. Read, S. Black, D. Thornley, A. Milodowski, M. Siitari-Kauppi, W. E. Falck (UK, Finland,
France)
FROM URANOTHORITES TO COFFINITE: A SOLID SOLUTION ROUTE TO THE
THERMODYNAMIC PROPERTIES OF USIO 4
S. Szenknect, D.T. Costin, N. Clavier, A. Mesbah, C. Poinssot, P. Vitorge, N. Dacheux (France)
TETRAVALENT CATION COPRECIPITATION WITH CLAY MINERALS
N. Finck, M. Bouby, K. Dardenne, H. Geckeis (Germany)
A2-1
A2-2
A2-3
INVESTIGATIONS INTO THE FORMATION OF NEPTUNIUM(IV)-SILICA COLLOIDS
R. Husar, S. Weiß, C. Hennig, H. Zänker, G. Bernhard (Germany)
A2-4
LIMITED REACTION OF BENTONITE AND SWELLING CLAYS IN LOW ALKALI CEMENT
LEACHATES: FINAL RESULTS FROM THE CYPRUS NATURAL ANALOGUE PROJECT
(CNAP)
W.R. Alexander, A.E. Milodowski, S.J. Kemp, J.C. Rushton, P. Korkeakoski, S. Norris and P. Sellin
(Switzerland, UK, Finland, Sweden)
A2-1
A2-5
URANIUM OXIDATION AND MIGRATION THROUGH CRYSTALLINE ROCK – TRANSPORT
PATHWAYS AND THE ROLE OF SECONDARY PHASES IN RETARDATION
D. Read (1) , S. Black (2) , D. Thornley (2) , A. Milodowski (3) , M. Siitari-Kauppi (4) and W. E. Falck (5) .
(1) Department of Chemistry, Loughborough University, Loughborough, Leics. LE11 3TU (UK)
(2)
SHES, University of Reading, Whiteknights, Reading, Berks. RG6 6AB (UK)
(3) British Geological Survey, Keyworth, Notts. NG12 5GG (UK)
(4)
Laboratory of Radiochemistry, University of Helsinki, Helsinki (Finland)
(5) University of Versailles, St. Quentin-en-Yvelines (France)
Understanding the mechanisms controlling the migration of uranium in crystalline rocks is a key component
of the safety case for disposal of spent fuel in Sweden and Finland. To further develop our knowledge base, a
novel column experiment was undertaken using a disc of metallic depleted uranium (DU) as a well
characterised source [1] and granodiorite core samples from a former candidate disposal site for spent
uranium fuel at Sievi, Finland, as the host medium. The experiment was conducted over 500 days.
The tests involve placing a pristine, 25 mm diameter DU disc between two 50 x 20 mm drill cores of Sievi
granodiorite in a triaxial cell (Figure 1) and forcing a pre-equilibrated groundwater solution through the
system using over-pressure. Previous studies [2, 3] demonstrate the rapid degradation of the DU discs under
a range of experimental conditions when exposed to excess solution, particularly solutions rich in dissolved
silica. In the experiment described here, rock/solution ratios are much more representative of those likely to
be found in the field.
The drill cores were selected following a series of gas and liquid permeability tests in order to give sufficient,
but not excessive flow on the timescale of the experiment. The drill cores were first vacuum dried and then
saturated by immersing in de-aired stock solution. A ceramic ring was placed just inside the edge of the drill
core, bringing the disc and rock surface flush and preventing marginal flow. A confining sleeve pressure of
35 psi combined with an inlet pressure of 4 psi (27.6 kPa) at the base was applied and samples of the
379

percolating solution were taken at increasing intervals to determine breakthrough. The chemical composition
of the effluent was monitored throughout the experiment as a function of time. On completion, the disc was
examined for corrosion and the cores analysed by X-ray tomography, XRF, SEM/EDAX and digital
autoradiography to determine potential flow paths and the distribution and composition of secondary
minerals formed.
The evolution of total uranium concentration and the 238 U/ 235 U ratio are shown in Figure 2. Evidence of an
increase in aqueous uranium is apparent after just a few hours and it is clear from the isotopic ratio that the
increase is attributable to dissolution of the DU disc. Within one month, the effluent 238 U/ 235 U ratio
corresponds to that of the disc and remains at this value throughout the test. Total uranium levels increase
progressively through a series of stages. The first, lasting approximately four weeks, shows the steepest rise
from a background of less than 10 ppb to ~ 40 ppb. This is followed by a more gradual rise to a peak of 160
ppb after one year. Concentrations then decrease, initially slowly and then more rapidly, falling to only 12
ppb at the point the experiment was stopped.
The pattern observed can be explained in terms of evolving mineralogy bearing in mind that the microcosm
contains a large quantity of uranium in contact with a finite volume of rock. Oxidation of the DU disc has a
marked effect on pH by removing hydroxide from the system, which in turn tends to promote further
dissolution of the disc. The process may be inhibited, however, by the development of a passivating film of
secondary alteration products [3].
The initial and rapid rise in U concentrations corresponds to oxidation and dissolution of the pristine disc
surface. Formation of secondary mineral phases takes longer but as they develop, they increasingly occupy
the surface exposed to percolating water and effluent concentrations rise more slowly. Eventually the entire
surface is covered and aqueous concentrations peak. At this point, one might expect levels to remain constant
as a steady state is established between the solubility of the secondary phase(s) and solution. The fact that
concentrations actually decrease implies an enhanced uranium removal mechanism. The latter could
represent more efficient scavenging by the phases present or reversion of those phases to a more crystalline
form with lower solubility. The final drastic drop in uranium concentration is explained by dilution, as it
corresponds to a marked increase in flow rate.
Initial ‘blind’ modelling efforts using the PHREEQC geochemical code and literature data pointed to several
solid phases that could potentially limit the solubility of uranium. As experimental data became available, the
model was refined, but the initial assumptions were corroborated in principle, pointing to hydrous uranium
silicates as likely solubility controlling phases. A slight supersaturation with respect to iron-oxyhydroxides
gives rise to the possibility that these may also influence uranium concentrations by co-precipitation [4].
It proved difficult to constrain the model and to arrive at unique solutions even for this relatively simple,
controlled experimental system. Several variables, for example the precipitation kinetics of individual solid
phases and the dispersivity of the system, merit further investigation. Comparison of model results with the
measured effluent concentrations clearly demonstrated the limitations of an equilibrium speciation code. The
experiment revealed a number of unexpected features that could not have been predicted in advance.
Consequently, it is very difficult to state a priori whether a given simulation will yield conservative results.
380

Fig 1 Experimental system
180.0
550.0
160.0
500.0
140.0
450.0
400.0
120.0
350.0
238U (ppb)
100.0
80.0
300.0
250.0
238U/235U
60.0
40.0
238U (ppb)
238U/235U
200.0
150.0
100.0
20.0
50.0
0.0
0.0
0 4 8 12 16 20 24 28 32 36 40 44 48 52 56 60 64 68 72
Elapsed Time (weeks)
Fig 2 Evolution of uranium concentration and isotopic ratio
[1] Trueman, E., Black, S. Read, D. (2004). Characterisation of depleted uranium (DU) from an unfired CHARM-3
penetrator. Sci. Tot. Env. 327, 337-340 (2004).
[2] Baumann, N., Arnold, T., Geipel, G., Trueman, E., Black, S. and Read, D. (2006). Detection of U(VI) on the surface
of altered depleted uranium by TRLFS. Sci. Tot. Env. 366, 905-909.
[3] Baumann, N., Arnold, T., Förstendorf, H. and Read, D (2008). Spectroscopic verification of the mineralogy of an
ultra-thin mineral film on depleted DU. Environ. Sci. Technol., 42, 8266-8269.
[4] Falck, W., Read, D., Black, S., Thornley, D. and Siitari-Kauppi, M. (2009). Uranium migration in crystalline rocks.
European Commission, JRC Report EUR 23816 EN.
A2-2
FROM URANOTHORITES TO COFFINITE: A SOLID SOLUTION ROUTE TO THE
THERMODYNAMIC PROPERTIES OF USiO 4
Stéphanie Szenknect 1) , Dan T. Costin 1) , Nicolas Clavier 1) , Adel Mesbah 1) , Christophe Poinssot 2) , Pierre
Vitorge 3) and Nicolas Dacheux 1) .
1)
ICSM, UMR 5257 CEA/CNRS/UM2/ENSCM, Site de Marcoule – Bât. 426, BP 17171, 30207 Bagnolssur-Cèze
cedex, France
2) CEA, Nuclear Energy Division, DRCP/DIR, CEA Marcoule, Bât. 400, BP 17171, 30207 Bagnols-sur-
Cèze cedex, France
3)
CEA, Nuclear Energy Division, DPC/SECR, Site de Saclay, Bât. 391, 91191 Gif-sur-Yvette, France
Coffinite (USiO 4 ) and associated solid solutions are expected to play an important role in the field of
direct storage of spent nuclear fuels in underground repositories since they could control the concentration of
381

actinides in the groundwater. However, the thermodynamic properties associated with coffinite, especially
the solubility constant, remain poorly defined. Very little reliable thermodynamic data related to coffinite
formation or solubility is reported in the literature and none is determined experimentally from solubility
measurements [1]-[4]. Solubility studies require pure single-phase USiO 4 . Most of the natural samples
contain coffinite as very fine grain crystals (≈ 5 µm) [5] and in intimate intergrowths with large amounts of
associated minerals. Moreover, for several decades persistent difficulties have been encountered in the
preparation of pure single-phase synthetic coffinite. An indirect method based on solubility measurements of
Th 1-x U x SiO 4 samples was thus envisaged in this study.
The preparation of synthetic Th 1-x U x SiO 4 uranothorite solid solutions was successfully undertaken under
hydrothermal conditions (T = 523 K) [6], [7]. The formation of a complete solid solution between x = 0
(thorite) and x = 0.8 was evidenced by XRD and EDS analyses. Lattice parameter refinement in the I4 1 /amd
space group confirmed the formation of solid solutions between thorite and coffinite, according to Vegard’s
law. Analysis systematically gave evidence of the formation of pure Th 1-x U x SiO 4 for 0≤ x≤ 0.4. For higher
values of x, a mixed oxide (Th 1-y U y O 2 ) second phase was detected. Phase quantification by the Rietveld
method showed an increase in the amount of the silicate phase with hydrothermal treatment time, indicating
a slow-rate phase transition of the oxide to silicate. The kinetics of this transition were found to slow with the
increase in the uranium bearing.
Experiments on the solubility of intermediate members of the Th 1-x U x SiO 4 solid solution were carried
out to determine the impact of Th-U substitutions on the thermodynamic properties of the solid solution and
then to allow extrapolation to the coffinite end-member. The ion activity products in solutions equilibrated
with Th 1-x U x SiO 4 (0 ≤ x < 0.5) were determined by dissolution experiments conducted in 0.1 mol L -1 HCl
under Ar atmosphere at several temperatures ranging from 298 K to 346 K. For all experiments, dissolution
was congruent and a constant composition of the aqueous solution was reached after 50 to 200 days of
dissolution. The solubility product of thorite was determined (log*K S ,ThSiO 4 = −5.62 ± 0.08) whereas the
solubility product of coffinite was estimated (log*K S ,USiO 4 = −6.1 ± 0.2). The stoichiometric solubility
product of Th 1-x U x SiO 4 reached a maximum value for x = 0.45 ± 0.05. In terms of the variation of the
standard Gibbs free energy of dissolution, solid solutions dissolve more spontaneously than the endmembers.
Variations of standard Gibbs free energy associated with the formation of thorite, coffinite and
intermediate members of the series were then evaluated. Variations of standard Gibbs free energy of
formation were found to increase linearly with the uranium mole fraction. Our data at low temperature
clearly shows that uranothorite solid solutions with x > 0.26, thus coffinite, are less stable than the mixture of
binary oxides, which is consistent with qualitative evidence from petrographic studies of uranium ore
deposits.
[1] Grenthe, I.; Fuger, J.; Konings, R. J. M.; Lemire, R. J.; Muller, A. B.; Nguyen-Trung, C.; Wanner, H., Chemical
Thermodynamics of Uranium. North Holland Elsevier Science Publishers B.V.: Amsterdam, The Netherlands,
1992; Vol. 1, p 715.
[2] Guillaumont, R.; Fanghänel, T.; Fuger, J.; Grenthe, I.; Neck, V.; Palmer, D. A.; Rand, M. H., Update on the
chemical thermodynamics of uranium, Neptunium, Plutonium, Americium and Technetium. North Holland
Elsevier Science Publishers B.V.: Amsterdam, The Netherlands, 2003; Vol. 5, p 919.
[3] Langmuir, D., Geochimica et Cosmochimica Acta 1978, 42, 547-569.
[4] Hemingway, B. S. Thermodynamic properties of selected uranium compounds and aqueous species at 298.15K and
1 bar and at higher temperatures. Preliminary models for the origin of coffinite deposits.; US Geological Survey:
1982; p 89.
[5] Deditius, A. P.; Utsunomiya, S.; Ewing, R. C., Chemical Geology 2008, 251, 33-49.
[6] Costin, D. T.; Mesbah, A.; Clavier, N.; Szenknect, S.; Dacheux, N.; Poinssot, C.; Ravaux, J.; Brau, H. P., Progress
in Nuclear Energy 2012, 57, 155-160.
[7] Costin, D. T.; Mesbah, A.; Clavier, N.; Dacheux, N.; Poinssot, C.; Szenknect, S.; Ravaux, J., Inorganic Chemistry
2011, 50, 11117-11126.
382

A2-3
TETRAVALENT CATION COPRECIPITATION WITH CLAY MINERALS
N. Finck*, M. Bouby, K. Dardenne, H. Geckeis
Institute for Nuclear Waste Disposal (INE), Karlsruhe Institute of Technology (KIT) – Campus North,
Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany)
*nicolas.finck@kit.edu
High level nuclear waste (HLW) is currently being vitrified and the resulting glass will eventually be buried
in permanent geologic repositories. However, over the long time scales needed for the waste to reduce its
radiotoxicity to the level of natural uranium, ground water may move through the barriers and contact the
HLW glass. At the glass / ground water interface, various mechanisms may take place including dissolution
of the waste matrix and subsequent precipitation of secondary phases. Corrosion experiments conducted in
the laboratory showed the formation of various phases including clay minerals [1]. The clay particles formed
can have sizes ranging from the macroscopic scale down to the colloidal scale. For the safety performance of
a HLW repository site, these colloidal particles have a high importance as they may be mobile and carry
radionuclides (RN) over long distances. Depending on the geochemical conditions, they may on the other
hand contribute to RN retention provided that nanoparticulate clay species are immobilized.
In clay-based repositories, reducing conditions are expected to prevail [2] so that the long-lived and
radiotoxic actinides released upon waste matrix corrosion will occur in low oxidation state. Among them,
plutonium may prevail in trivalent and in tetravalent oxidation state. The retention of trivalent actinides (e.g.,
Am, Cm) has been documented in the literature, but the binding of tetravalent f-elements has received far
less attention. This study reports the first results on tetravalent cation retention by coprecipitation with clay
minerals, by using Zr(IV) as proxy for tetravalent actinides.
The magnesian smectite hectorite was synthesized in the presence of Zr(IV) (sample ZrCopHec) following a
multi-step synthesis protocol by using brucite as the precursor phase [3]. Separately, Zr-containing brucite
(sample ZrCopBru) was prepared under identical conditions. XRD data indicated that the samples consist
only of brucite or of clay mineral, and that no other separate phase could be detected. SEM data indicated
that they both consist of small particles with a layered structure.
Zirconium K-edge X-ray absorption spectroscopy (XAS) data were collected to characterize the local
chemical environment. In the XANES region, which is sensitive to the chemical environment and the
coordination geometry, the white line (WL) featured in the samples differ from each other and also differ
from reference spectra of ZrO 2 and the Zr(IV) aqueous ions (Figure 1). This hints to a successive change in
the chemical environment during the hectorite synthesis procedure. The observed broadening and splitting of
the WL can be explained by a decrease in the Zr site symmetry [4]. In the EXAFS region, fits to the data
indicate that Zr is ligated to O atoms at d(Zr-O) = 2.08 Å in ZrCopBru and that the next nearest neighbors
consist of two Mg subshells. This data strongly
suggests that Zr and Mg polyhedra share faces
and
edges. Upon hectorite crystallization, the short
range environment changed. The O shell is
located at d(Zr-O) = 2.08 Å, and the next nearest
neighbors consist of Mg (d(Zr-Mg = 3.04 Å) and
Si
(d(Zr-Si) = 3.22 Å) backscatterers. These bond
lengths match reported values for Mg at the
octahedral site in hectorite [4], meaning that Zr
substitutes for octahedral cations in hectorite.
The size distribution in the hectorite colloidal
fraction was characterized by application of the
asymmetrical flow field-flow fractionation
(AsFlFFF) method and the associated elemental
composition was obtained by ICP-MS. Based on
AsFlFFF data, the mobilized colloidal fraction
Figure 1. XANES data collected for ZrCopHec,
383
ZrCopBru and the reference compounds.

has a size distribution centered at ~60 nm. Furthermore, Zr(IV) is found associated homogeneously to
hectorite. The XAS data collected on the bulk solid and the AsFlFFF data on the colloidal fraction may
indicate that Zr(IV) is structurally incorporated and homogeneously distributed in hectorite from the
macroscopic scale down to the nanoscale. The nanoparticulate fraction thus needs to be considered in
developing a safety case, but the colloidal stability of such a phase as a function of the geochemical
conditions first needs to be investigated.
[1] P. Jollivet, et al., J. Nucl. Mater. 420, 508 (2012).
[2] E. Gaucher, et al., C. R. Geoscience 338 (2006).
[3] N. Finck, et al. Environ. Sci. Technol. 43, 8807 (2009).
[4] P. Li, et al., Phys. Rev. B48, 10063 (1993).
[5] W. Seidl, J. Breu, Z. Kristallogr. 220, 169 (2005).
A2-4
INVESTIGATIONS INTO THE FORMATION OF NEPTUNIUM(IV)-SILICA COLLOIDS
R. Husar, S. Weiß, C. Hennig, H. Zänker, G. Bernhard
Helmholtz-Zentrum Dresden-Rossendorf, Institute of Resource Ecology, P.O. Box 51019
01314 Dresden, Germany
In the near and far field of nuclear waste repositories lower oxidations states of the actinides (An) are
expected to become predominant because of the prevalent reducing conditions. Due to the low solubility at
neutral pH, tetravalent actinides (An(IV)) are assumed to reveal immobile migration behaviour.
Nevertheless, a high environmental mobility has been found, e.g. for Pu(IV) [1, 2], which is obviously
related to the formation of An(IV) eigencolloids or to the sorption of the heavy metal ions onto other
colloids. In the laboratory, An(IV)O 2 ×H 2 O colloids for Th(IV) and Np(IV) have been synthesized [3, 4].
Regarding the erosion processes in repository sites, e.g. on glass molds, surrounding and ubiquitous
occurring matter, the interaction of An(IV) with carbonate and silicate compounds and the potential
formation of colloids has to be considered carefully. In particular, past studies evidenced the formation of
silica-containing U(IV) and Th(IV) colloids [5, 6].
We developed a method to generate aqueous Np(IV) solutions, explicitly excluding the presence of other
oxidation states, for the synthesis of silica-containing colloids. Under anaerobic conditions, Np(IV)
carbonate solutions in the presence and absence of silicate were investigated by UV-vis spectroscopy,
ultrafiltration, LSC and DLS (dynamic light scattering). Figure 1 shows the UV-vis absorption spectra of
solutions containing Np(IV) species determined after 24 h equilibration time. The typical spectrum of
Np(IV) solutions in 1.0 M carbonate is shown as a curve (1) in Figure 1. At lower carbonate concentration
(0.1 M), precipitation of Np(IV) is generally expected and observed [7]. However, when silicate was
admixed to such a solution, Np(IV) is prevented from precipitation and the spectrum obtained is shown in
Fig. 1 (curve 2) which differs from that of dissolved Np(IV) in 1.0 M carbonate. The absorption band at
741 nm is significantly increased and in dependence on the concentration of silica shifted to 747 nm. Thus,
the formation of colloidal Np(IV) silicate is strongly suggested. Furthermore, the colloid-disperse Np(IV)
silicate solution exhibits an increased
(1)
scattering light intensity and diameters of
soluble Np(IV)
0,075 741
(2)
colloidal the particles were determined in the range
Np(IV) silicate
1 kDa ultrafiltrate
(3)
from 1 to 20 nm.
Absorbance
0,050
Ultrafiltration removes these particles and
the respective UV-vis spectrum shows
considerably reduced absorption bands
0,025
(Fig. 1, curve 3). In addition to the
disappearance of the previously observed
(3)
absorption bands around 745 nm, the
0,000
concentration of Np was reduced from
1.0 × 10 −3 to 0.1 × 10 −3 M and 1.8 × 10 −3 to
0.3 × 10 −3 400
M (LSC). All these findings can 450 500 550 600 650 700 750 800 850 900
384
Wavelength [nm]
Fig.1: UV-vis spectra of the soluble Np(IV) carbonate
(1.82 × 10 −3 M Np(IV), 1 M NaHCO 3 ) (1), the colloidal Np(IV)
(1.72 × 10 −3 M Np(IV), 3.2 × 10 −3 M Si, 1 × 10 −1 M NaHCO 3 )
(2), and the ultrafiltrate of colloidal Np(IV) silicate (3).
(1)
(2)

e explained by colloidal behaviour of Np. Moreover, it is ascertained that these colloidal systems are stable
over a period of more than 120 days. In the presence of silicate we observed a stabilized dispersion of
Np(IV) silica colloids. Hence, Np(IV) may become waterborne even if the limit of solubility is exceeded.
The existence of such colloids has never been reported so far.
[1] R. Buddemeier et al., (1988) Appl. Geochem. 3, 535.
[2] S. Utsunomiya et al., (2009) Environmental Science Technology 43 (5), 1293.
[3] M. Altmaier et al., (2004) Radiochim. Acta 92, 537
[4] V. Neck et al., (2001) Radiochim. Acta 89, 439
[5] I. Dreissig et al., (2011) Geochimica et Cosmochimica Acta 75, 352.
[6] C. Hennig et al., (2013) Geochimica et Cosmochimica Acta 103, 197-212.
[7] D. Rai et al., (1999) Radiochim. Acta 84, 159.
A2-5
LIMITED REACTION OF BENTONITE AND SWELLING CLAYS IN LOW ALKALI CEMENT
LEACHATES: FINAL RESULTS FROM THE CYPRUS NATURAL ANALOGUE PROJECT
(CNAP)
W.R. Alexander (1) , A.E. Milodowski (2) , S.J. Kemp (2) , J.C. Rushton (2) , P. Korkeakoski (3) , S. Norris (4) and P.
Sellin (5)
(1) Bedrock Geosciences, Auenstein, Switzerland
(2) British Geological Survey, Keyworth, UK
(3) Posiva, Olkiluoto, Finland
(4) NDA-RWMD, Harwell, UK
(5) SKB, Stockholm, Sweden
Bentonite is an important component in many designs for radioactive waste repositories. The plasticity,
swelling capacity, colloid filtration, low hydraulic conductivity, high retardation of key radionuclides and
stability in relevant geological environments all make bentonite an ideal barrier/buffer material in an
engineered barrier system (EBS). However, bentonite is unstable in higher pH environments and this is a
potential problem for repository designs which mix cement and concrete with bentonite barriers. The alkaline
(initial pH~13) leachates from the cement are expected to degrade the bentonite, potentially resulting in the
loss of some or all of its favourable properties. This has driven recent interest in low alkali cements, because
the pH of the leachate is somewhat lower than standard OPC (Ordinary Portland Cement), lying around pH
10-11. It is hoped that this lower pH will reduce the reactivity of bentonite, so allowing the use of low alkali
cements in close proximity with bentonite.
Assessing the long-term stability of bentonite in contact with such alkaline fluids under conditions
representative of a deep geological repository requires complementary laboratory, modelling and in situ
studies. In particular, to build a robust safety case, it is important to have supporting natural analogue data to
confirm understanding of the likely long-term performance of bentonite.
Natural analogue studies could [1]:
• Provide information on reaction rates, the products of such alteration and their safety-relevance to
the performance of the engineered barrier system
• Allow testing of current models and databases used to assess such reactions
• Provide input to a range of supporting documents for safety cases
There are a number of locations worldwide where an appropriate natural analogue might be found, including
Oman, California, Bosnia, Papua New Guinea and the Philippines [2]. Based on a multi-attribute analysis,
considering factors such as probability of finding suitable locations with alkaline groundwaters and bentonite
together, relevance to European national radioactive waste disposal programmes and logistics, Cyprus was
finally chosen as the focus for the study.
385

Here, the natural bentonites (as analogues of the industrial bentonites in the repository EBS) from the Parsata
area of Cyprus have been examined [3] for evidence of reaction with the underlying, naturally alkaline,
groundwaters (as analogues of low alkali cement leachates) which are produced by reaction with a range of
basic and ultrabasic rocks in the Troodos ophiolite (Figure 1). Two independent lines of evidence will be
presented which suggest that the groundwater system has been running, at least intermittently, for some 10 5
a. There are indications of rock-water reaction at the very base and along some fractures in the bentonite
which are thought to be the product of long-term reaction with the underlying alkali groundwaters [4].
However, any reaction has limited spatial reach with the total volume of bentonite altered (from
montmorillonite to palygorskite) being significantly less than 1% of the total mass present.
These results are supported by recently published data [5] from a medium-term (ca. 15a) laboratory
experiment which indicates that as the cement leachate plume evolves to lower alkalinity, an Mg-rich
palygorskite or sepiolite appears to replace early reaction products such as CSH, ettringite and possibly
apophyllite that originally replaced clay mineral or other phyllosilicates [6]. Overall, the results of the
analyses which will be presented here suggest that there has been very limited alkaline groundwater reaction
with the bentonite over a period of 10 5 a, tending to indicate that any long-term bentonite reaction in low
alkali cement leachates will be minimal. The use of this novel data in support of the safety case of a
repository which combines both bentonite and low alkali cement will be explored in full [cf. 7].
Bentonite
Ophiolite
Surface sediments
D
A
Infiltration / surface (neutral) water flow
B
Zone of active
serpentinisation
C
Hyperalkaline water flow
A – Active serpentisation producing hyperalkaline groundwater and H 2 /CH 4 gas
B – High pH water under bentonite, neutral water above it
C – Potential interaction of high pH waters with base of bentonite (diffusion into it?)
D – Borehole through bentonite and into pillow lavas
E – Dispersed release of high pH waters into deep sediments?
F – Hyperalkaline springs in ophiolite
E
Figure 1: The natural system in Cyprus. Alkaline groundwaters could react with bentonite wherever they
meet [3]
[1] Alexander, W.R., McKinley, I.G. and Kawamura, H. 2013. The process of defining an optimal natural analogue
programme to support national disposal programmes. Proceedings of a workshop on Natural Analogues for Safety
Cases of Repositories in Rock Salt. 4-6 September 2012, Braunschweig, Germany. NEA/OECD, Paris, France (in
press).
[2] Alexander, W.R., Arcilla, C.A. et al. 2008. A new natural analogue study of the interaction of low-alkali cement
leachates and the bentonite buffer. Sci Basis Nucl Waste Manag XXXI, 493-500.
[3] Alexander, W.R., Milodowski, A.E. and Pitty, A.F. (eds) 2011. Cyprus Natural Analogue Project (CNAP) Phase III
Final Report. Posiva Working Report WR 2011-77, Posiva, Eurajoki, Finland.
[4] Alexander, W.R. and Milodowski, A.E. (eds) 2013. Cyprus Natural Analogue Project (CNAP) Phase II Final
Report. Posiva Working Report WR 2013-XX, Posiva, Eurajoki, Finland (in press).
[5] Moyce, E.B.A., Morris, K., Milodowski, A.E., Rochelle, C.A., Brown, A. and Shaw, S. 2013. Long Term Rock
Alteration in the Chemically Disturbed Zone of a Geological Disposal Facility. Min Mag (in press).
[6] Rochelle, C.A., Pearce, J.M., Bateman, K., Coombs, P. and Wetton, P.D. 1997. The evaluation of chemical mass
transfer in the disturbed zone of a deep geological disposal facility for radioactive wastes. X: Interaction between
synthetic cement porefluids and BVG: Observations from experiments of 4, 9 and 15 months duration. British
Geological Survey (BGS), Fluid Processes and Waste Management Group Report, WE/97/16C, 79p. BGS, Keyworth,
UK
[7] Alexander, W.R., Milodowski, A.E. and Norris, S. 2013. The contribution of CNAP to the understanding of
bentonite reaction with alkaline cement leachates. Swiss Journal of Geosciences (in press).
386

THE UNUSUAL CHEMISTRY OF PLUTONIUM
D. L. Clark 1)*
1) Los Alamos National Laboratory, Los Alamos, NM, 87544, USA
Plutonium is one of the most complex elements in the Periodic Table.[1] The pure element exhibits seven
distinct crystal phases, is highly reactive, and is known to form alloys, compounds, or complexes with
virtually every other element. Due to its electropositive nature, elemental plutonium readily oxidizes, and in
aqueous solution plutonium can form positively charged cations in four common oxidation states. The metal
ion in each of these formal oxidation states can form a variety of molecular complexes with man-made and
naturally occurring ligands, each with a characteristic solubility and chemical reactivity.
Because the redox potentials that couple the four common oxidation states in acid solution (III, IV, V, VI)
are all remarkably similar, plutonium cations have a marked tendency to undergo disproportionation and
comproportionation reactions. This redistribution of oxidation states makes plutonium solution chemistry
particularly complex and fascinating, and must be recognized when studying plutonium behavior in solution.
The oxidation state of plutonium affects its chemical behavior. For example, Pu(III) and Pu(IV) are, in
general, relatively insoluble, whereas Pu(V) and Pu(VI) are, in general, more soluble. Each has a different
tendency to undergo hydrolysis and complexation reactions with ligands. This is why knowledge of the
oxidation state under environmental conditions is critically important for the long-term performance of
underground nuclear waste repositories. In oxidation state IV, plutonium strongly hydrolyzes, often to form
small oligomeric clusters or colloids. These intrinsic colloids can attach themselves to natural mineral or
organic colloids that have important consequences for the migration of plutonium in the natural environment.
The unique behavior of plutonium can be traced to its position in the Periodic Table and electronic structure.
The presence of 5f electrons in its valence shell, and low-lying empty 6d orbitals give rise to unusual
opportunities for chemical bonding. In this presentation, I will discuss plutonium chemistry from a basic,
molecular-level perspective, with discussion of the electronic structure, the coordination chemistry and
physicochemical characteristics of common oxidation states, and a survey of how these properties impact the
behavior of plutonium in natural environments including solubility, speciation, and physicochemical
transport.
[1] D. L. Clark, S. S. Hecker, G. D. Jarvinen, M. P. Neu, “Chapter 7, Plutonium” in The Chemistry of the Actinides and
Transactinides, 3 rd Edition, L. R. Morss, N. M. Edelstein, J. Fuger, Eds. 2006, Springer, New York, 813-1264.
SESSION 17
B2: DIFFUSION AND OTHER MIGRATION
PROCESSES
REACTIVE TRANSPORT OF U(VI) THROUGH POROUS MEDIA AMENDED WITH
PHOSPHATE TO INDUCE IN SITU URANIUM IMMOBILIZATION
V.S. Mehta, F. Maillot, Z. Wang, J.G. Catalano, D.E. Giammar (USA)
URANIUM MIGRATION AND RETENTION MECHANISM IN THE PROCESS WASTE OF A
CONVERSION FACILITY
T. Fernandes, L. Duro, Th. Schäfer, P. Masqué, A. Delos, J.S. Flinois, G. Videau (Spain, Germany,
France)
DISSOLVED ORGANIC MATTER AND COLLOID MEDIATED TRANSPORT IN BOOM
CLAY: SIZE EFFECTS
D. Durce, C. Bruggeman, N. Maes (Belgium)
B2-4
B2-5
B2-6
387

B2-4
REACTIVE TRANSPORT OF U(VI) THROUGH POROUS MEDIA AMENDED WITH
PHOSPHATE TO INDUCE IN SITU URANIUM IMMOBILIZATION
Vrajesh S. Mehta 1 , Fabien Maillot 1 , Zheming Wang 2 , Jeffrey G. Catalano 1 and Daniel E. Giammar 1
1 Washington University in St. Louis, MO, USA
2
Pacific Northwest National Laboratory, Richland, USA
Addition of phosphate amendments to U(VI)-contaminated subsurface environments is a promising
approach for in-situ remediation. Batch equilibrium experiments on uranium(VI)-phosphate interactions in
the presence of reactive substrate minerals have demonstrated impacts of phosphate on U(VI) adsorption as
well as the formation of sparingly soluble U(VI) phosphate solids 1-3 . However, the migration of U(VI) in
actual subsurface environments can be strongly influenced by non-equilibrium chemical and physical
processes. U(VI) mobility in subsurface environments can consequently be controlled by advective and
diffusive transport processes as well as by adsorption-desorption and dissolution-precipitation reactions. The
presence of reactive mineral surfaces like those of iron oxides and clays in sediments can potentially limit the
precipitation of U(VI) phosphate solids by adsorbing dissolved U(VI) and phosphate to lower the saturation
ratios of precipitates 4 . On the other hand, these mineral surfaces may facilitate heterogeneous nucleation
through effects on the interfacial energy barrier to nucleation and the formation of ternary U(VI)-phosphate
surface complexes that may serve as precursors to U(VI) phosphate precipitation. Thus, the primary
objective of this study was to understand the impacts of phosphate on the reactive transport of U(VI) through
columns loaded with porous media. Gaining insight into these processes can help identify conditions that
lead to the greatest reductions in U(VI) mobility.
This objective was addressed through a series of column experiments using various porous media.
Laboratory scale glass columns (2.5 cm Diameter x 15 cm Length) were loaded with either uncontaminated
sediments from environmentally relevant field sites in the United States or with goethite-coated sand.
Experiments simulating groundwater flow first introduced 4 µM U(VI) into the columns at a seepage
velocity of 1 m/day. Bromide was added as a conservative tracer for calculating hydrodynamic transport
parameters. Columns were operated for more than 175 pore volumes. For sediments from the Rifle site in
Colorado, U(VI) breakthrough occurred after 60 pore volumes and after the sediments had adsorbed 7.7
µmol U(VI) per kg of sediment. This uptake capacity is within the same order of magnitude reported in other
studies with other sediments 5 . After the sediments had been loaded with U(VI), the columns received U(VI)-
free influents either with 1 mM phosphate added to promote immobilization or without phosphate, and the
release of U(VI) from the columns was monitored by regular monitoring of the column effluents. Adsorption
was significantly faster than desorption. During the desorption phase of experiments, the presence of
phosphate resulted in 2.5 times less U(VI) release than from columns that did not receive phosphate over 82
pore volumes. This indicates that phosphate promoted enhanced uptake of U(VI) by the sediments. The
amount of phosphate adsorbed by sediments was 2960 µmol kg -1 (a factor of 385 higher than that of
uranium). Columns loaded with goethite-coated sand showed much higher affinity for phosphate and
uranium than the field sediments. A stopped flow technique was used to observe the effects of diffusion on
U(VI) adsorption and desorption and to evaluate whether the sediments and groundwater had reached
equilibrium. To identify the dominant U(VI) species in the sediments and their spatial distribution along the
length of the column, sediments were collected in different increments at the end of the experiment and used
for further characterization by electron microscopy, X-ray absorption spectroscopy, and fluorescence
spectroscopy. Finally a 1-dimensional reactive transport model was used to interpret the observations of
U(VI) uptake and release from the sediments. The observations reiterate the importance of studying such a
dynamic system to gain insight on uranium phosphate interactions that can aid in the design of effective insitu
remediation strategies.
1. Cheng, T., et al., Effects of phosphate on uranium(VI) adsorption to goethite-coated sand. Environmental Science
& Technology, 2004. 38(22): p. 6059-6065.
2. Jensen, M.P., et al., Immobilization of actinides in geomedia by phosphate precipitation. Humic and Fulvic Acids:
Isolation, Structure, and Environmental Role, 1996. 651: p. 272-285.
3. Singh, A., K.U. Ulrich, and D.E. Giammar, Impact of phosphate on U(VI) immobilization in the presence of
goethite. Geochimica Et Cosmochimica Acta, 2010. 74(22): p. 6324-6343.
388

4. Fuller, C.C., et al., Mechanisms of uranium interactions with hydroxyapatite: Implications for groundwater
remediation. Environmental Science & Technology, 2002. 36(2): p. 158-165.
5. Qafoku, N.P., et al., Kinetic desorption and sorption of U(VI) during reactive transport in a contaminated Hanford
sediment. Environmental Science & Technology, 2005. 39(9): p. 3157-3165.
B2-5
URANIUM MIGRATION AND RETENTION MECHANISM IN THE PROCESS WASTE OF A
CONVERSION FACILITY
T. FERNANDES 1,2 , L. DURO 1 , TH. SCHÄFER 3 , P. MASQUÉ 2 , A. DELOS 4 , J.S. FLINOIS 5 , G. VIDEAU 5 ,
1 Amphos 21, Barcelona, Spain (*correspondence: teresa.fernandes@amphos21.com) 2 Departament de
Física & Institut de Ciència i Tecnologia Ambientals. Universitat Autònoma de Barcelona, Bellaterra.
Spain 3 Karlsruhe Institute of Technology – Institute for Nuclear Waste Disposal 4 Arcadis, Villeurbanne,
France 5 AREVA, France,
The industrial site of interest is the first step in the treatment of uranium mining concentrate. The process
waste resulting from the conversion of yellowcake into uranium tetrafluoride (UF 4 ) has been managed in
settling ponds. Two historical ponds have been constructed on the mine tailings and waste resulting from the
flotation process of a former sulphur mine. The settling waste, i.e. resulting sludge, contains a variety of
chemicals (nitrates, carbonates, fluorides and sulphates) and radioactive elements.
Source term characterisation suggests U is present in different degrees of crystallinity. The mineral phases of
U have not been identified so far, however Scanning Electron Microscopy (SEM) and Raman Spectroscopy
point to a U-Si phase.
Previous studies of the pond suggest that the mine tailings act as an efficient buffer that minimises migration
to the underlying ground. However, the actual mechanisms by which this control is active are not well
understood.
In order to investigate the mechanism(s) retaining U, migration experiments were carried out. One borehole
core, of total depth 13.2 m, was sampled in the basins and selected core samples of sludge and tailings were
sealed with paraffin wax and transported to the laboratories. Migration experiments of waste and tailings
were carried out on these undisturbed samples to investigate the release of U from the material.
The columns were eluted with artificial rainwater to simulate the release of U under in-situ conditions. In
addition, one column of tailings was eluted with 233 U and the retention of U by the uncontaminated tailings
was investigated. The major chemistry in the eluate was monitored to assist in the interpretation of the
mechanisms behind.
In this contribution, we will present the information obtained on the hydrodynamics of the material and on
the behaviour of U under this unique system. We will discuss how these compare to field conditions and
present an updated conceptual model for the migration of U through the waste and tailings.
B2-6
DISSOLVED ORGANIC MATTER AND COLLOID MEDIATED TRANSPORT IN BOOM CLAY:
SIZE EFFECTS
D. Durce (1) , C. Bruggeman (1) and N. Maes (1)
(1) Belgian Nuclear Research Centre (SCK • CEN), Expert Group Waste & Disposal, Boeretang 200, B-2400
Mol, Belgium
Deep geological disposal of nuclear waste in clay formations is based on low mobility of radionuclides (RN)
resulting from low water permeability and retention phenomena. However, transport of solutes in these
389

formations is also subject to the presence of dissolved organic matter (DOM) which is ubiquitous in natural
environments. For instance, the Boom Clay formation, potential host formation under study in Belgium,
contains relative high amounts of DOM (50-150 mgC/L). Although the influence of DOM on radionuclide
mobility has been proven on many occasions [1], [2] , the involved mechanisms are not yet fully understood.
Indeed, most works were focussed only on the simulation and assessment of the radionuclide behaviour.
However, a better understanding of RN-DOM transport starts with a detailed study of the migration of the
DOM itself.
DOM is a complex assemblage of molecules with variable functionalities and molecular weights.
Consequently, to consider it as a unique entity is likely to lead to only a partial description of its impact.
Previous works, carried out by Put et al. [3], [4] , have already pointed out the existence of a filtration process
for the biggest molecules and have allowed to quantify the average diffusion (D app ) for the mobile fraction.
However, size measurements were missing to identify precisely the cut-off of the clay and to possibly
distinguish several behaviours in the mobile fraction. The objective of the present study is to determine sizedependent
transport behaviours and associated migration parameters for Boom Clay DOM. To this end,
several pools were investigated with different size distributions.
The dissolved organic matter extracted by leaching from dispersed Boom Clay samples displayed a size
distribution ranging from small molecules 100 kDa whereas the DOM sampled
in situ via underground piezometers was dominated by molecules 20 kDa and mobile DOM is shared between a fast fraction (

u.a (280nm)
80
70
60
50
40
30
inlet solution
20days
48days
65days
86days
111days
164 days
20
10
0
1.00E+01 1.00E+02 1.00E+03 1.00E+04 1.00E+05 1.00E+06
-10
M w (Da)
Figure 1. Evolution with time of the size distribution of dissolved organic matter at the outlet of a percolation
experiment performed on an intact Boom Clay core with a solution of piezometer derived DOM in inlet. Size exclusion
chromatography analyses.
[1] Maes, N., Wang, L., Hicks, T., Bennett ,D., Warwick, P., Hall, P., Walker, G., Dierckx, A., (2006). The role of
natural organic matter in the migration behaviour of americium in the Boom Clay-Part 1: migration experiments.
Physics and Chemistry of the Earth, 31, 541-547
[2] Warwick, P., Hall, A., Pashley, V., Bryan, N.D., Griffin, D. ,(2000). Modelling the effect of humic substances on the
transport of europium through porous media: a comparison of equilibrium and equilibrium/kinetic models. J. Contam.
Hydrol. 42, 19-34
[3] Put, M. J., Dierckx, A., Aertsens, M., and De Canniere, P., (1998). Mobility of the dissolved organic matter through
intact boom clay cores. Radiochim Acta 82, 375-378.
[4] Put, M. J., Monsecour, M., and Fonteyne, A., 1992. mobility of the dissolved organic material in the interstitial
boom clay water. Radiochim Acta 58-9, 315-317.
[5] Maes, N., Bruggeman, C., Govaerts, J., Martens, E., Salah, S., and van Gompel, M., 2010. A consistent
phenomenological model for natural organic matter linked migration of Tc(IV), Cm(III), Np(IV), Pu(III,IV) and Pa(V)
in Boom Clay. Physics and Chemistry of the earth, 36, 1590-1599
SESSION 18
ENVIRONMENTAL BEHAVIOUR OF RADIONUCLIDES
AFTER THE FUKUSHIMA ACCIDENT
RADIOCESIUM IN THE ENVIRONMENT IN THE AFTERMATH OF THE FUKUSHIMA
DAIICHI ACCIDENT: STATUS OF RADIOACTIVITY, DECONTAMINATION WORK AND
MIGRATION RESEARCH PROJECT
M. Yui (INVITED) (Japan)
ENVIRONMENTAL BEHAVIOR OF RADIOCESIUM AFTER FUKUSHIMA DAIICHI
NUCLEAR POWER PLANT ACCIDENT
T. Ohnuki, F. Sakamoto, N. Kozai, S. Yamasaki (Japan)
EFFECT OF SOIL PARAMETERS ON SORPTION PROPERTIES OF ACTINIDES AND FISSION
PRODUCTS : DEPTH PROFILE DISTRIBUTION OF FALLOUT RADIONUCLIDES IN SOILS
AFFECTED BY FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT
S. Mishra, A. Sorimachi, M. Hosoda, S. Tokonami, T. Ishikawa, S.K. Sahoo (Japan)
FS-1
FS-2
FS-3
391

FS-1
RADIOCESIUM IN THE ENVIRONMENT IN THE AFTERMATH OF THE FUKUSHIMA DAIICHI
ACCIDENT: STATUS OF RADIOACTIVITY, DECONTAMINATION WORK AND MIGRATION
RESEARCH PROJECT
Mikazu Yui
Japan Atomic Energy Agency (JAEA), Fukushima
The earthquake and subsequent tsunami disasters of March 11, 2011 had serious consequences for the
Fukushima Daiichi nuclear power plant owned by Tokyo Electric Power Company (TEPCO). Substantial
off-site release of volatile radionuclides required many residents in the vicinity of the power plant to be
evacuated. After the accident, monitoring of radioactive contamination was conducted; after the decay of
short lived isotopes of iodine, now the dominant radioisotopes found in the environment are Cs-134 and Cs-
137. Initial decontamination demonstration work was carried out by JAEA and with the knowledge gained
from such work, the Japanese government issued guidelines on how practical decontamination should be
executed. Then, based on the guidelines, decontamination work was initiated by the national government for
areas where the external gamma dose rate was > 20 mSv/y (11 municipalities) and for 104 municipalities
where the dose rate was between 1 and 20 mSv/y.
The most difficult decontamination work we have encountered has been for forested areas, which cover
about 70 % of the Fukushima prefecture. Guidelines focus on the decontamination of forest adjacent to
residential areas, which is both reasonable and practical taking into consideration population dose reduction,
ecosystem conservation and disaster prevention e.g. excessive removal of surface vegetation may lead to
landslips. However, further studies are needed for optimisation of various parts of the decontamination work
programme e.g. decrease of waste production, improvement of waste storage and disposal and the prevention
of recontamination of decontaminated areas.
Even after decontamination work has been completed, JAEA will continue to undertake long-term
investigations into Cs transport and behavior in the environment. The F-trace project will study potential Cs
sources (e.g. from non-decontaminated forest) and any subsequent releases and transport into and through
aquatic systems (rivers, reservoirs, lakes and estuaries). To date, studies have revealed that Cs has rarely
been detected in the aqueous phase as it tends to be strongly sorbed onto surfaces. Mobile Cs is thus
predominantly associated with physical transport e.g. on solid particles such as clay minerals which are well
known to have a high affinity for this element. Based on these continuous long-term investigations, JAEA
will determine if it is possible to develop methodologies to reduce the future dose rates from radiocesium to
residents in affected areas. For example it may be possible to develop countermeasures to reduce Cs
transport by suspended particles in surface waters.
An overview of the accident, the subsequent decontamination activities, the status of Cs distribution in the
environment and ongoing laboratory and field research investigations will be presented.
FS-2
ENVIRONMENTAL BEHAVIOR OF RADIOCESIUM AFTER FUKUSHIMA DAIICHI NUCLEAR
POWER PLANT ACCIDENT
T. Ohnuki 1)* , F. Sakamoto 1) ) , N. Kozai 1) , S. Yamasaki 1)
1) Advanced Science Research Center, Japan Atomic Energy Agency, 2-4,
Muramatsu, Tokai, Ibaraki 319-1195, Japan
Environmental behavior of radioactive Cs deposited by the accident of Fukushima Daiichi Nuclear Power
Plant has been studied by an autoradiography analysis (ARA) combined with a sequential desorption
analysis (SDA). The plant and soil samples were collected in Iitate-village, Fukushima on May, 2011. The
392

ARA conduced to measure spatial distribution of radioactive Cs on/in trees, plants, and surface soil.
Chemical states of radioactive Cs deposited on/in the plants and soil was characterized by the SDA. After the
SDA, the residual soil was treated for size fractionation by sieving. In addition sorption and desorption
experiments of radioactive Cs by different minerals of kaolinite, halloysite, chlorite, montmorillonite,
mordenite, illite, and sericite, and metal oxides of MnO 2 , TiO 2 , Al 2 O 3 , and FeOOH to examine their
contribution to the sorption of radioactive Cs on the soil.
The ARA showed that radioactive Cs was distributed on the branches and leaves of trees that were present
at the accident and that only a small fraction may be transported to new branches and leaves grown after the
accident. Radioactive Cs was present on the grass and rice stubble on the soils, but not in the soils beneath
the grass and rice stubble, indicating that the radioactive Cs deposited on the grass and the rice stubble were
retained on the plants. In addition the fraction of the radioactive Cs penetrated into the soil layer by
weathering was very small for two months after the accident. These results indicate that trees and plants
would be the reservoir of the fallout Cs and function for retardation of the fallout Cs migration with rain
water.
In the SDA, more than 65% of radioactive Cs remained in the residual fraction of the soil samples after
treatment of a 1 mole L -1 NH 4 Cl solution and a 1 mole L -1 CH 3 COOH solution. Approximately 70% of
radioactive Cs in the residual fraction were associated with the size fractions larger than the elutriated one,
even though mica like minerals were contained in the elutriated one. These results strongly suggest that
radioactive Cs was irreversibly associated with soil components other than mica-like minerals in the
contaminated soil.
The adsorption and desorption experiments of Cs on the minerals were carried out at the Cs concentrations
1 x 10 -4 , 1 x 10 -5 and 2 x 10 -9 mole L -1 at pH 5.5. The distribution coefficient (K d ) for the minerals at the Cs
concentration 10 -9 mole L -1 was in the order of mordenite > illite > montmorillonite, sericite, MnO 2 ,
kaolinite, and halloysite > chlorite, TiO 2 , Al 2 O 3 , and FeOOH, differing from the order observed at higher Cs
concentrations. After the SDA by a 0.1 mole L -1 LiCl solution, a 1 mole L -1 KCl solution, and a 1 mole L -1
HCl solution, the residual fraction of Cs was higher at the Cs concentration 10 -9 mole L -1 than at higher
concentrations. Approximately 40%, 40%, 50%, and 25% of the adsorbed Cs were residual in
montmorillonite, mordenite, MnO 2 and kaolinite, respectively after SDA. These results strongly suggest that
(1) radioactive Cs at 10 -9 mole L -1 is more strongly associated with the non-mica minerals than at higher
concentrations of 1 x 10 -4 and 1 x 10 -5 mole L -1 , and (2) the non-mica minerals montmorillonite, mordenite,
kaolinite, and MnO 2 contributed to the irreversible sorption of the radioactive cesium fallout on Fukushima
soil.
FS-3
EFFECT OF SOIL PARAMETERS ON SORPTION PROPERTIES OF ACTINIDES AND FISSION
PRODUCTS : DEPTH PROFILE DISTRIBUTION OF FALLOUT RADIONUCLIDES IN SOILS
AFFECTED BY FUKUSHIMA NUCLEAR POWER PLANT ACCIDENT
S. Mishra 1 , A. Sorimachi 2 , M. Hosoda 3 , S. Tokonami 2 , T. Ishikawa 1 , and S.K.Sahoo 1 *
1 National Institute of Radiological Sciences, 4-9-1 Anagawa, Inage-ku, Chiba 263-8555, Japan
2 Institute of Radiation Emergency Medicine, Hirosaki University, Aomori 036-8564, Japan
3 Graduate School of Health Sciences, Hirosaki University, Aomori 036-8564, Japan
sahoo@nirs.go.jp
Long-lived radionuclides deposited on soil can cause an enhanced radiation exposure even after many years.
For long term predictions of radiation doses, quantitative information on the migration rates of the fallout
radionuclides in the soil is essential. These rates can be obtained by evaluating the observed concentration
depth profiles of a fallout radionuclide in the soil after given time periods following the deposition onto the
soil surface (1). The migration behaviour and depth profile distribution of radionuclides in soils are site
specific and depend on soil characteristics and environmental conditions. Soil acts as a medium for transfer
of radionuclides to biological systems due to various biogeochemical processes. Therefore the knowledge on
the effect of various soil parameters on sorption behaviour and biogeochemical mobilisation of the
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adionuclides is very important for radiation protection and dose assessment. The depth profile distribution
of radionuclides is affected by soil properties like pH, carbonates, organic matter, cation exchange capacity
(CEC), particle size and soil processes such as leaching and adsorption (2).
In the present study the effect of different soil parameters on the sorption property for uranium (U), cesium
(Cs) and strontium (Sr) and depth profile distribution for Cs-134 and Cs-137 have been studied in soil
samples affected by Fukushima Daichi nuclear power plant accident (FDNPP). Soil samples were collected
around the FDNPP accident site and analysed for different isotopes of U, Cs and Sr for finding out the source
of contamination. Chemical characterization with respect to different soil parameters like particle size
distribution, pH, organic content, CEC, CaCO 3 , elemental and oxide composition of soil has been carried out
to understand the geochemical behavior of the radionuclides. Solid-liquid distribution coefficient (K d ), has
been estimated for the three nuclides using laboratory batch method. Stable isotopes are used for Sr and Cs
whereas depleted uranium is used as tracer for uranium sorption experiments. For depth profile distribution,
core soil samples collected up to 20 cm depth at 2 cm and 5 cm intervals. Cs-134 and Cs-137 were measured
in these soil samples using a high purity germanium detector (HPGE) and multi-channel analyser (MCA).
The distribution coefficient values are found to be in the decreasing order U (log K d ≈ 3-4) > Cs (log K d ≈ 2-
3) > Sr (log K d ≈ 1-2). The variation in the distribution coefficient values of these nuclides were explained
based on the soil parameters. At soil pH 5.5, U shows maximum K d , then decreases with increase in pH that
may be due to formation of soluble carbonyl complexes. U-K d shows good correlation with Fe and organic
content . Similarly Cs-K d and Sr-K d show good correlation with CEC and fine particles respectively.
From activity depth profile of Cs-137 and Cs-134 and using convection-dispersion model the migration
parameters were evaluated.
1. K. Bunzl. Vertical random variability of the distribution coefficient in the soil and its effect on the migration of
fallout radionuclides. J. Radioanal. Nucl. Chem., 254 (2002) 15-21.
2. S. Dragović, B. Gajić, R. Dragović, L. Janković-Mandić, L. Slavković-Beškoski, N. Mihailović , M. Momčilović
and M. Ćujić. Edapic factors affecting the vertical distribution of radionuclides in the different soil types of
Belgrade, Serbia. J. Environ. Monit, 14 (2012) 127.
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