Euradwaste '08 - EU Bookshop - Europa
Euradwaste '08 - EU Bookshop - Europa Euradwaste '08 - EU Bookshop - Europa
in the European coordinated action project MICADO. This project assesses uncertainties governed by the divergence between the various models and the experimental databases, and uncertainties in predictions that arise by comparing the outcomes of the various models. The waste form is of course only one of the barriers within a disposal concept characterised by the superposition of multiple partly interdependent barriers. Barrier functions concern the establishment of favourable geochemical conditions, limiting water access to the waste, limiting the transfer of contaminants to the geosphere etc. The relative role of the waste form within these barrier systems has been illustrated in the integration research component of NF-PRO [1]. The calculations show that, compared to the absence of any barrier, the presence of a suitable waste form in its container and the low solubility of key radionuclides already reduce hypothetical dose contributions at the near-field/geosphere interface by up to five orders of magnitude. The importance of the waste form depends on the stability of the waste form and on the disposal concept. In a disposal concept in which the transfer of contaminated groundwater to the biosphere takes longer than the time of degradation of the waste form, the waste form stability will have less significance than when nearfield/biosphere transfer rates are fast. On the other hand, if, as it appears today, a waste form is stable for hundreds of thousands or even for more than a million years, then it will provide important safety margins and it sustains the safety case in any disposal concept. 2. Methodology The methodology of the waste form oriented research within NF-PRO consisted in combining bibliographic surveys with modelling and experiments, to obtain missing data in integral near-field environments for various host rock formations, and the evolution with time. Integration with the other research components (processes in the engineered barrier and in the engineering disturbed/damaged zone) was attempted by a performance assessment oriented approach. The programme started from a review of previous projects on glass and spent fuel and considering the material choices, boundary conditions and scenarios fixed by the other research components of NF-PRO. Work on glass was organised in a work package oriented on experiments, and another one on the geochemical modelling of the observed interactions. Work on spent fuel was organised in a work package dealing with the evolution of spent fuel under normal and early failure scenarios, focussing mainly on the instant radionuclide release fraction (IRF), and a work package concerning the behaviour of the fuel matrix under near-field conditions. The methodology of the still ongoing MICADO project consists in bringing together model developers and potential users, experimenters and people with overall system understanding to (1) select models (2) to select and review a common spent fuel/UO2/MOX experimental database, (3) to apply the models with or without re-parameterisation to the repository relevant part of the database, and assess deviations and interdependence of the model parameters, (4) to evaluate all relevant information and approaches for spent fuel performance assessment, including uncertainty propagation to the overall safety analyses, and assess simplification strategies to translate detailed mechanistic models into overall system codes, and (5) to provide an independent regulator point of view on the appreciation of the effects of documented model uncertainties on predictive uncertainties for the repository safety. 3. Results We provide only a summary of the main results. For more details, we refer to reference [2]. 174
3.1 Vitrified waste The simplified reference source term for vitrified waste at the start of NF-PRO was the r0-rr model. In this model, glass dissolution is described to occur in two stages: a first stage of high dissolution rate (r0), which lasts up to the saturation of the metallic overpack products with silica, and a second stage of low residual dissolution rate (rr or rres), which lasts until the glass is completely dissolved. The processes at the basis of this reference source term model were studied in a programme with three components (1) Glass-water interaction, covering the determination of some basic parameters of glass dissolution and the effect of dissolved carbonate on the release of rare earth elements and U. (2) Radionuclide immobilisation in secondary phases. (3) Validation of key mechanisms of glass dissolution in integrated near-field conditions. We present the main results for each of these components. Determination of basic parameters of glass dissolution The dissolution rates r0 and rres are well known for the R7T7 glass (corresponding to the SON68 standard reference glass), but this was not the case for the UK “blended Magnox-UO2” glass, which contains more magnesium and aluminium than the R7T7 glass. These parameters were determined within NF-PRO. The blended Magnox-UO2 glass behaves less favourable than the R7T7 glass: both the forward rate and the residual rate are higher for the blended Magnox-UO2 glass. The reason for the higher residual rate is probably the formation of secondary magnesium phases, triggering the glass dissolution. The lower stability of the blended Magnox-UO2 glass was confirmed by the integrated tests, which are discussed further. The reference source term model can now be applied for the blended magnox-UO2 glass, provided sufficient data are available concerning another important parameter, i.e. the exposed surface area of the glass. Effect of dissolved carbonate on the release of rare earth elements and uranium Long-term dissolution tests with glass GP WAK1 in synthetic Opalinus and Konrad clay pore water solutions showed no clear effect of carbonates on the release of rare earth elements and uranium from the glass in the carbonate concentration range of 73-98 mg/l, but this conclusion cannot be extrapolated as such to higher carbonate concentrations. Crystalline secondary phases were observed, such as powellite, barite, calcite, CaSO4 and clay-like Mg(Ca,Fe) silicates, but there were no distinct uranium phases in the gel. This means that carbonate concentrations do not have much impact on the radionuclide retention in the gel. This should be taken into account if radionuclide retention is considered explicitly in the source term model. In the reference source term model used for NF- PRO, radionuclide retention is not considered explicitly. In this case, retention in the gel contributes to the safety margin. Role of radionuclide immobilisation in secondary phases Under conditions typical for a deep geological nuclear waste repository, secondary alteration phases are formed during dissolution of the waste glass, once their solubility limit has been reached. Radionuclides, which have been released from the waste matrix may co-precipitate with these secondary phases and form thermodynamically stable solid solutions, such as amorphous gel layers and various crystalline phases (see previous paragraph). It was found that trivalent actinides (Am, Pu, Cm) can be structurally incorporated into the host minerals powellite, calcite and clay minerals, forming solid solutions. The quantitative understanding of this solid solution formation has im- 175
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in the European coordinated action project MICADO. This project assesses uncertainties governed<br />
by the divergence between the various models and the experimental databases, and uncertainties in<br />
predictions that arise by comparing the outcomes of the various models.<br />
The waste form is of course only one of the barriers within a disposal concept characterised by the<br />
superposition of multiple partly interdependent barriers. Barrier functions concern the establishment<br />
of favourable geochemical conditions, limiting water access to the waste, limiting the transfer of<br />
contaminants to the geosphere etc. The relative role of the waste form within these barrier systems<br />
has been illustrated in the integration research component of NF-PRO [1]. The calculations show<br />
that, compared to the absence of any barrier, the presence of a suitable waste form in its container<br />
and the low solubility of key radionuclides already reduce hypothetical dose contributions at the<br />
near-field/geosphere interface by up to five orders of magnitude. The importance of the waste form<br />
depends on the stability of the waste form and on the disposal concept. In a disposal concept in<br />
which the transfer of contaminated groundwater to the biosphere takes longer than the time of degradation<br />
of the waste form, the waste form stability will have less significance than when nearfield/biosphere<br />
transfer rates are fast. On the other hand, if, as it appears today, a waste form is stable<br />
for hundreds of thousands or even for more than a million years, then it will provide important<br />
safety margins and it sustains the safety case in any disposal concept.<br />
2. Methodology<br />
The methodology of the waste form oriented research within NF-PRO consisted in combining bibliographic<br />
surveys with modelling and experiments, to obtain missing data in integral near-field environments<br />
for various host rock formations, and the evolution with time. Integration with the other<br />
research components (processes in the engineered barrier and in the engineering disturbed/damaged<br />
zone) was attempted by a performance assessment oriented approach. The programme started from<br />
a review of previous projects on glass and spent fuel and considering the material choices, boundary<br />
conditions and scenarios fixed by the other research components of NF-PRO. Work on glass was<br />
organised in a work package oriented on experiments, and another one on the geochemical modelling<br />
of the observed interactions. Work on spent fuel was organised in a work package dealing with<br />
the evolution of spent fuel under normal and early failure scenarios, focussing mainly on the instant<br />
radionuclide release fraction (IRF), and a work package concerning the behaviour of the fuel matrix<br />
under near-field conditions. The methodology of the still ongoing MICADO project consists in<br />
bringing together model developers and potential users, experimenters and people with overall system<br />
understanding to (1) select models (2) to select and review a common spent fuel/UO2/MOX<br />
experimental database, (3) to apply the models with or without re-parameterisation to the repository<br />
relevant part of the database, and assess deviations and interdependence of the model parameters,<br />
(4) to evaluate all relevant information and approaches for spent fuel performance assessment, including<br />
uncertainty propagation to the overall safety analyses, and assess simplification strategies to<br />
translate detailed mechanistic models into overall system codes, and (5) to provide an independent<br />
regulator point of view on the appreciation of the effects of documented model uncertainties on predictive<br />
uncertainties for the repository safety.<br />
3. Results<br />
We provide only a summary of the main results. For more details, we refer to reference [2].<br />
174