Although spent fuel reprocessing remains the reference scenario in France, direct disposal in deep geological repository is also studied as an option within the framework of the French national plan for the radioactive waste management. Following UO₂ spent fuel that have been intensively studied, the alteration mechanisms of U_1-xPu_xO₂ fuels and especially, Mimas MOX fuels are under investigation to establish long-term evolution models. From the literature, the key mechanism controlling UO₂ dissolution and the associated radionuclides release is an oxidizing dissolution induced by H₂O₂ produced by water radiolysis. Such mechanism can be affected by the MOX microstructure and plutonium content. Moreover, the radionuclides release from MOX spent fuel is affected by the groundwater chemistry. Especially, the presence of cementitious backfill material should create alkaline chemical environment likely to affect the water radiolysis yield and the nature of secondary phases formed at the interface. Furthermore, the use of a non-radioactive surrogate material with comparable properties to MOX fuel has relevant practical advantages. As such, finding a suitable surrogate material allowing multi-parametric studies is a major challenge to improve our knowledge of the MOX fuels alteration in various chemical environments. This work aims to investigate the analogy between U_1-xCe_xO₂ surrogate materials and unirradiated Mimas MOX fuel in the presence of H₂O₂. Once this analogy established, the behavior of Mimas MOX fuel and heterogeneous U_1-xCe_xO₂ surrogate in alkaline solution was compared to evaluate the impact of alpha radiation on the alteration mechanisms.

First, both homogeneous and heterogeneous U_1-xCe_xO₂ dense pellets with x ranging from 0 to 0.25 were prepared through wet and dry chemistry routes, respectively. Surrogate materials were then submitted to dynamic leaching experiments at pH = 7.2 and room temperature. The feeding solution containing 0.20 mmol.L^-1 H₂O₂, was kept under air and renewed every 48 to 72 h to guarantee the H₂O₂ stability during the whole experiment. Normalized alteration rates were determined from uranium concentration measured in the leachates after reaching the steady state. Post-alteration characterizations by Raman spectroscopy, environmental SEM and XPS were achieved. The secondary phase precipitation did not occur at the homogeneous (U,Ce)O₂ materials surface and the dissolution rate was divided by a factor 3 when increasing the Ce molar content from 0.08 to 0.25. However, studtite precipitation was observed all over UO₂ surface, leading to a continuous uranium concentration decreases in the outflow. The same results were obtained with heterogeneous U_0.92Ce_0.08O₂. However, studtite was found to precipitate on UO₂ grains only. This result was consistent with that observed for heterogeneous (U,Pu)O₂ in the same conditions, which confirmed the reliability of cerium as a valuable plutonium analogue.

Then, heterogeneous Mimas MOX fuel and surrogate material with an average composition of U_0.93Pu_0.07O₂ and U_0.92Ce_0.08O₂, respectively were altered over several months under static conditions in alkaline solutions containing 2 mmol.L^-1 of silicates at pH 12, room temperature and under anoxic conditions. The a-activity of the MOX pellet was 1.34 GBq.g_MOX^-1. No matter the presence of alpha-radiation, the uranium concentration in solution reached the same stable value of (5.5 ± 0,6)×10^-6 mmol.L^-1 after 120 days of alteration. Geochemical calculations showed that the uranium concentration measured at equilibrium was compatible with the solubility of the U(VI)-phase, clarkeite (Na(UO₂)O(OH)), or U(IV)-phases coffinite (USiO₄) and UO₂·2H₂O. However, altered surfaces characterizations by SEM and Raman spectroscopy did not reveal the presence of secondary phases. The Raman spectra of the altered MIMAS MOX fuel were characteristic of non-oxidized UO₂ and PuO₂ surfaces. These results rather indicated that oxidative dissolution was inhibited by the presence of silicate ions, as already observed for UO₂ SIMFUEL in the literature.

In safety assessments for disposal of spent nuclear fuel (SNF) in a deep geological repository (DGR), water access, consecutive failure of canisters and loss of integrity of fuel cladding is considered in the long-term. Since most radionuclides produced during irradiation of nuclear fuel in a reactor are trapped within the SNF matrix, it is indispensable to understand the processes leading to the dissolution of the matrix and its dissolution rates to evaluate the performance of SNF in the near-field of such a DGR.

In the present study, irradiated UO₂ and mixed oxide fuel (MOX) specimens were sampled from fuel rod segments with an average burn-up of 50.4 and 38.0 MWd/kg_HM, respectively, and used in the experiments.

Static leaching experiments under anoxic/reducing conditions were performed with cladded pellets and decladded fragments of the sampled SNFs in salt brines, granitic/bentonitic or cementitious groundwaters. The experiments were periodically sampled and solution aliquots and gas phases were analyzed.

As observed in other published studies on SNF dissolution under reducing conditions, a large scatter of the uranium concentration is seen in the initial stages of our experiments. After about one year of leaching, the aqueous concentration of uranium approaches slowly towards the solubility limit of U(IV) independent of the type of studied SNF samples (various irradiated UO₂ and MOX fuels). A similar behaviour is observed for other redox-sensitive actinides. The release rate of uranium decreases significantly in all experiments within first 400 days of leaching. Although the rate is very low, it is > 0 and a continuous release is observed. Comparison of results from SNF experiments under various redox conditions demonstrate that oxidative matrix dissolution is inhibited due to the presence of hydrogen.