Euradwaste '08 - EU Bookshop - Europa

system robustness, rather than as some sort of alternative behaviour of the system that has a reasonable
expectation of occurring.
For a near field in which cementitious materials are used in significant quantities, the dissolution
rate of HLW glass is poorly known and requires further study in order to assess the radionuclide
retention capacity of HLW glass at high pH.
Regarding the relative significance of the dissolution rate of HLW glass, it has been shown in safety
assessment calculations that for clay host rock disposal systems, in the present state of knowledge,
there would be almost no difference in dose between the cases in which the initial or the residual
rates are used. This is the result of the strong retention of the radionuclides in the host rock, resulting
in substantial decay during transport. Nonetheless, depending on the degree to which an essentially
independent barrier function is considered necessary, there is an argument to be made for
strengthening the scientific basis for and reducing the uncertainty in the low long-term dissolution
rate.
SF
Over the past few years, significant evidence has accumulated that has shown that the potential oxidative
effects of alpha radiolysis may be significantly mitigated by the presence of dissolved hydrogen
and Fe(II). Although a qualitative explanation for these observations can be envisaged, quantification
of the details in terms of the rate constants is clearly difficult. In addition the minimum concentrations
of H2 or Fe(II) needed to mitigate the effects of �-radiolysis are not certain and it is possible
that other dissolved species may also play a role. It is clear that a better understanding of the
radiolytic processes at the SF interface with porewater is needed.
NF-PRO studies have reduced the uncertainties regarding the time evolution of the instantaneous
released fraction (IRF) of the major safety-relevant radionuclides for moderate burnup fuel. There
is now confidence that the different mechanisms previously identified as potentially leading to an
increase with time of the IRF (� self-irradiation enhanced diffusion and helium accumulation) are of
little relevance. The impact of helium accumulation for higher burnup UO2 and MOX fuel may still
warrant further study. Regarding measurements of the IRF, data are still lacking in particular for
high burnup and MOX fuel. More studies should be done, as present models appear to involve
somewhat pessimistic assumptions.
The relative importance of the IRF vs. the matrix dissolution rate has been explored in studies
within and outside of NF-PRO. For IRF values in the range of 5 to 10% (typical values in some assessment
studies), the IRF dominates the dose, with I-129 being by far the most important radionuclide
from the dose perspective. Nonetheless, this is based on matrix dissolution models based on
uranium solubility that estimate that the fuel dissolution process takes millions of years or more.
The assumption of a dissolution control by alpha-radiolysis and thus a dissolution time of ~100,000
years would result in a decrease in the relative contribution of the IRF but also in an increase in
dose. There is thus an important incentive to more clearly understanding the factors that can counter
alpha-radiolytic oxidative dissolution of spent fuel.
The rim region of the fuel pellets contains higher fission products concentrations (due to higher
burnup) and a different microstructure (smaller grains and large pores) compared with the rest of
the pellet. These differences raise doubts regarding the applicability of the UO2 dissolution data to
the rim region. It is necessary to clarify whether rim leaching behaviour is very different from the
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