The 12th International Conference on Environmental ... - Events

Abstracts Session 51
ings require melting temperature as many as 1350-1400 oC whereas the glass with 50 wt.% waste loading may be produced at 1150-
1250 oC. Wet slurry (~50 and ~70 wt.% water content) in amount of 625.1 kg was processed and glass in amount of 186.6 kg was
produced and poured into 12 canisters for 91 hrs. Average slurry feed rate, glass productivity and specific glass productivity were
6.86 kg/hr, 2.16 kg/hr, and 68.8 kg/(m2 ×hr) respectively.
3) ON FLUIDIZATION OF BOROSILICATE GLASSES IN INTENSE RADIATION FIELDS - 16055
Michael Ojovan, Guenter Möbus, Jim Tsai, Stuart Cook, University of Sheffield (UK);
Guang Yang, University of Chicago (USA)
<strong>The</strong> viscosity is rate-limiting for many processes in glassy materials such as homogenisation and crystallisation. Changes in
the viscous flow behavior in conditions of long-term irradiation are of particular interest for glassy materials used in nuclear installations
as well as for nuclear waste immobilising glasses. We analyse the viscous flow behavior of oxide amorphous materials in
conditions of electron-irradiation using the congruent bond lattice model of oxide materials accounting for the flow-mediating role
of broken bonds termed configurons. An explicit equation of viscosity was obtained which is in agreement with experimental data
for non-irradiated glasses and shows for irradiated glasses, first, a significant decrease of viscosity, and, second, a stepwise reduction
of the activation energy of flow. An equation for glass-transition temperature was derived which shows that irradiated glasses
have lower glass transition temperatures. Intensive electron irradiation of glasses causes their fluidisation due to non-thermal bond
breaking and can occur below the glass transition temperature. Due to surface tension forces fluidisation of glasses at enough high
electron flux densities can result in modification of nano-size volumes and particles such as those experimentally observed under
TEM electron beams.
4) DESIGN INNOVATIONS IN ADVANCED VITRIFICATION SYSTEM AT TARAPUR - 16254
Kalyan Banerjee, Rajendra Gupta, Sudhakaran Nair, Sridutt Misra, Bhabha Atomic Research Centre (India)
Glass has been chosen internationally by most countries as the final waste form for disposal of high level waste (HLW) because
the glass matrix is chemically inert and can accommodate a wide spectrum of radio-nuclides. <strong>The</strong> process, universally referred to,
as vitrification, is very tolerant of expected compositional variations in HLW and has been adopted as a reference process in several
countries over the past fifty years. Liquid high-level wastes are mixed with glassforming chemicals, melted and poured into
robust stainless steel canisters which are then sealed by welding. All these operations are carried out remotely in shielded hot cells
since HLW is highly radioactive. Glass offers sound product characteristics, which are extremely essential for the long-term disposal
in deep geological repositories. <strong>The</strong> last forty years have seen the glass process evolving from small batch operations, to the
presently existing high capacity continuous processes. Single step vitrification process using ceramic melters provides an appreciable
simplification of the whole vitrification process. <strong>The</strong> high energy input into the molten glass pool achieves drying, calcination
and melting in a single piece of process equipment. <strong>The</strong> process is continuous and favours considerably high feed throughputs. With
the Indian atomic energy programme maturing with substantial increase in HLW generation, the ceramic melter technology is
expected to become the preferred process in near future when work begins on the construction of several integrated facilities for
reprocessing and vitrification.
<strong>The</strong> first plant based on the ceramic melter technology named as Advanced Vitrification System (AVS) was designed, installed
and commissioned in the process cells of the Solid Storage Surveillance Facility (SSSF), Tarapur. Since the plant was hot commissioned,
the AVS has successfully produced about 175 m3 of vitrified high level waste. This paper describes the design experience
of AVS and how the lessons learnt during its operation have helped in modifications of future plants.
5) THE EFFECT OF WASTE LOADING ON THE CHARACTERISTICS
OF BOROSILICATE SRS SB4 WASTE GLASSES - 16196
Sergey Stefanovsky, Alexander Kobelev, Vladimir Lebedev, Michael Polkanov,
Dmitriy Suntsov, SIA Radon (Russia); James Marra, SRNL (USA)
<strong>The</strong> glasses containing surrogate of Sludge Batch 4 (SB4) waste with high concentrations of aluminum and ferrous oxides
within the waste loading range of ~50 to 70 wt.% produced in a 236 mm inner diameter cold crucible were examined by X-ray diffraction,
electron microscopy and infra-red spectroscopy. At relatively low waste loadings (50 to 60 w.%) the products remain predominantly
vitreous with minor crystalline spinel structure phase corresponding in chemical composition to trevorite-magnetite
solid solution. At higher waste loadings (>60 wt.%) amount of crystalline phases increased significantly and extra nepheline phase
occurred. <strong>The</strong> products with of 65 to 70 wt.% waste loading are spinel/nepheline glass-ceramics. <strong>The</strong> glass with 60 wt.% waste
loading slowly cooled in accordance with the canister centerline cooling (CCC) conditions contains minor nepheline. Occurrence
of nepheline reduces chemical durability of nuclear waste glasses.
6) DIFFUSION OF HELIUM IN BOROSILICATE GLASSES: A COMPARATIVE APPROACH BETWEEN
RADIOACTIVE AND NON-RADIOACTIVE GLASSES - 16208
Toby Fares, Sylvain Peuget, Atomic Energy Commission (CEAMarcoule) (France);
Jacques Haussy, CEA (France); Xavier Deschanels, ICSM (France)
High-level nuclear wastes consisting mostly of fission products and minor actinides are currently vitrified. A borosilicate glass
has been selected in France as the reference waste conditioning matrix. Because the glass canisters will be placed in a geological
repository, research is carried out to predict the long-term behavior of the glass under disposal conditions. <strong>The</strong> main source of radiation
damage (atomic displacements) inside the glass is caused by the ~-decay mode of the minor actinides (Am and Cm). Each ~disintegration
generates a helium atom that could induce an increase of helium concentration over time in the glass network. Investigating
the diffusion of helium in the glass is therefore important to understand the mechanisms involved in long term helium
behavior. <strong>The</strong>se mechanisms will allow us to predict whether or not helium will be released outside the glass under geological conditions;
if release does not occur, the issue of bubble formation in the glass structure must be evaluated. This paper presents some
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