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The spent fuel matrix constitutes a dynamic redox system, due to the time-dependent generation of oxidants and reductants at the spent fuel/water interface by alpha, beta and gamma radiolysis. In this context it is critical to understand the main processes and mechanisms that control the radiolytic production of redox components at this interface and their impact on the stability of the UO 2 matrix and in radionuclide release.In order to achieve this, a series of carefully controlled experiments have been carried out in order to determine the time dependence of the radiolytic H 2 , O 2 and H 2 O 2 production and radionuclide release (U, Pu, Np, Tc, Mo, Sr and Cs). These tests have been performed by using PWR Ringhals spent fuel fragments in contact with 10 mM NaHCO 3 solutions in an initially anoxic and closed system.The results of these experiments indicate quite a consistent and reproducible behaviour of the radiolytically generated reductants and oxidants. The results indicate that it is possible to define an equilibrium redox potential for the main redox pairs in the bulk system and that the experimentally determined radionuclide release can be rationalised in terms of the system redox potential.

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The thermodynamic and kinetic dissolution properties of a synthetically obtained soddyite have been determined at different bicarbonate concentrations. This uranium-silicate is expected to be a secondary solid phase of the oxidative alteration pathway of uranium dioxide in waters with low phosphate content and, consequently, it is likely to constitute one of the long-term uranium solubility limiting solid phases.The experimental data obtained at the end of the experiments correspond fairly well to the theoretical model calculated with a log K0S0 of 3.9±0.7.On the other hand, the general trend of the total uranium in solution measured in the experiments as a function of time has been fitted by using a kinetic equation obtained from the principle of detailed balancing of the dissolution reaction. In addition, the EQ3/6 code has also been used to model the uranium concentrations as a function of time. In both modeling exercises comparable results were obtained. The dissolution rate, normalized to the total surface area used in the experiments as measured with the BET method, gave an average value of 6.8 (±4.4) 10−14 mol cm−2 s−1.