###### Abstract  
For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO\({}_{2}\) matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H\({}_{2}\) may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (\(10^{-7}\) year\({}^{-1}\)) is over estimated because it assumes that there is net UO\({}_{2}\) oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel-water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.  
s 2021 11 8566 2021 15 September 2021  
man-made connections to the surface. The repository depth (in Finland, the nominal depth is 430 m below sea level) provides isolation from the surface environment and mitigates the likelihood of inadvertent human intrusion due to common activities, such as underground construction and installation of drinking water wells. The hydrogeochemical conditions at repository depth at the selected disposal sites are typically reducing, supporting the longevity of the disposal canisters.  
However, groundwater will eventually penetrate the disposal canisters and come in contact with the spent fuel. Nuclear fuel used in most commercial nuclear power plants consists of uranium dioxide (UO\({}_{2}\)) enriched with U-235, which is the fissile material. Most radionuclides (e.g., actinides and fission products) produced during the irradiation of nuclear fuel in the reactor are trapped in the UO\({}_{2}\) crystalline matrix. The UO\({}_{2}\) crystalline matrix is a fluorite-type structure with small distortions and a slight excess stoichiometry (UO\({}_{2+\lambda}\)) due to the effect of irradiation and the formation of fission and activation products. Leaching tests of spent nuclear fuel samples (or materials that are designed to simulate spent nuclear fuel) carried out over several decades show that one of the limiting factors in the release of radionuclides into the environment is the intrinsically low dissolution rate of UO\({}_{2}\)[1,2]. Such leaching tests also showed that the release of uranium and other elements seems to be inhibited in the presence of a reducing agent in solution (typically H\({}_{2}\) or metallic iron) conditions. The mechanism for such an inhibiting effect is still being debated.  
Spent nuclear fuel is an intrinsic source of ionizing radiation (alpha, beta, and gamma) and neutrons, and the intensity of the radiation field depends on the irradiation history and the age of the spent fuel. Radiolysis plays an important role in the UO\({}_{2}\) alteration mechanism because the energy deposited within the UO\({}_{2}\) matrix in combination with reactive species produced by radiolysis in the water layer at or near the interface will both lead to the formation of reactive species governing the redox conditions at the UO\({}_{2}\) interface. Understanding oxidation and reduction processes at the spent fuel-water interface will contribute to improving the conceptual model of radionuclide releases and achieve a more realistic assessment of the environmental effects of spent nuclear fuel disposal.