Leaching experiments with medium and high burn-up spent UOX fuels under anoxic and reducing conditions in highly alkaline media

The disposal of spent nuclear fuel (SNF) in deep geological formations in combination with a resilient
multi-barrier concept is the preferred option for the safe isolation of highly radioactive wastes in Germany,
as well as several other countries (e.g., Belgium, Finland, Sweden, Switzerland). Nevertheless, an intrusion
of ground water, intertwined with the failure of canisters and loss of SNF cladding integrity must be
considered in the long-term evaluation of a deep geological repository. An assessment of SNF performance
in the repository system requires a thorough process understanding of the dissolution rates, the individual
radionuclide source terms as well as the alteration processes of the waste form. The dissolution process of
SNF can be described in two steps: (i) a fast, initial release of radionuclides, which segregated to accessible
structures of the SNF during reactor operation (ii) a slower, long-term release, originating from the
dissolution of the fuel matrix itself.
In the present study, we show results obtained from ongoing leaching experiments with medium
(46.9 GWd/t$_{HM}$) and high burn-up (50.4 GWd/t$_{HM}$) UOX SNF, in highly alkaline, simulated cement water
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solutions under anoxic and reducing conditions induced by H2 overpressure. Both SNF were irradiated in
commercial nuclear power plants in Germany and Switzerland during the 1970s and 1980s. For the volatile
radionuclides, such as the fission gases, 129I and 137Cs, a rapid, initial release is observed, in comparison to
radionuclides assigned to the SNF matrix, e.g., $^{90}$Sr, $^{238}$U or $^{238}$Pu. However, the initially rapid release of
volatile fission products significantly slows down throughout the experiments, unaffected by neither the
pH of the leachant solution nor the presence of reducing H$_2$, albeit a continuous release is observed. In
addition, the data obtained in the current leaching experiments are compared to previous experiments
conducted at KIT-INE with UO$_X$ and (U,Pu)O$_X$ fuels.