In a geological repository for spent nuclear fuel, studtite (UO2)O-2(H2O)(4) may form on the fuel surface when in contact with groundwater under certain conditions. Studtite has a very low solubility and could thereby reduce the reactivity of spent nuclear fuel toward radiolytic oxidants. This would inhibit the dissolution of the fuel matrix and thereby also the spreading of radionuclides. It is therefore important to investigate the stability of studtite under conditions that may influence its stability. In this work we have studied the kinetics of studtite dissolution in aqueous suspensions containing no added HCO3- and with 10 mM HCO3-. The same type of experiment was performed also with solutions containing 0.2 mM H2O2. The solubility of studtite in the suspensions containing no added HCO3- is very low as expected, while the solubility in solutions containing 10 mM HCO3- is significantly higher. This is attributed to the formation of uranyl-carbonate and uranyl-peroxo-carbonate complexes. In 0.2 mM H2O2 and 10 mM HCO3- the observed solubility of U(VI) seems unaffected by the presence of H2O2. Again, this can be rationalized by the formation of uranyl-peroxo-carbonate complexes. It is interesting to note that H2O2 appears to be catalytically decomposed in solutions containing uranyl-carbonate complexes. In addition, gamma-radiation-induced dissolution of studtite in HCO3- deficient solutions and in 10 mM HCO3- was studied. The dissolution rate was found to be extremely high in HCO3- under gamma-irradiation. This is attributed to a combination of radiolytic degradation of H2O2 and the formation of uranyl-peroxo-carbonate complexes keeping the concentration of free H2O2 at a very low level and thereby driving the dissolution process.

Two uranyl peroxides meta-studtite and studtite exist in nature and can form as alteration phases on the surface of spent nuclear fuel upon water intrusion in a geological repository. Meta-studtite and studtite have very low solubility and could therefore reduce the reactivity of spent nuclear fuel toward radiolytic oxidants. This would inhibit the dissolution of the fuel matrix and thereby also the spreading of radionuclides. It is therefore important to investigate the stability of meta-studtite and studtite under conditions that may influence their stability. In the present work, we have studied the dissolution kinetics of meta-studtite in aqueous solution containing 10 mM HCO3-. In addition, the influence of the added H2O2 and the impact of γ-irradiation on the dissolution kinetics of meta-studtite were studied. The results are compared to previously published data for studtite studied under the same conditions. 13C NMR experiments were performed to identify the species present in aqueous solution (i.e., carbonate containing complexes). The speciation studies are compared to calculations based on published equilibrium constants. In addition to the dissolution experiments, experiments focussing on the stability of H2O2 in aqueous solutions containing UO22+ and HCO3- were conducted. The rationale for this is that H2O2 was consumed relatively fast in some of the dissolution experiments.

Hydrogen peroxide is produced upon radiolysis of water and has been shown to be the main oxidant driving oxidative dissolution of UO2-based nuclear fuel under geological repository conditions. While the overall mechanism and speciation are well known for granitic groundwaters, considerably less is known for saline waters of relevance in rock salt or during emergency cooling of reactors using seawater. In this work, the ternary uranyl-peroxo-chioro and uranyl-peroxo-bromo complexes were identified using IR, Raman, and nuclear magnetic resonance (NMR) spectroscopy. Based on Raman spectra, the estimated stability constants for the identified uranyl-peroxo-chloro ((UO2)(O-2)(Cl)(H2O)(2)(-)) and uranyl-peroxo-bromo ((UO2)(O-2)(Br)(H2O)(2)(-)) complexes are 0.17 and 0.04, respectively, at ionic strength approximate to 45 mol/ L. It was found that the uranyl-peroxo-chloro complex is more stable than the uranyi-peroxo- bromo complex, which transforms into studtite at high uranyl and H2O2 concentrations. Studtitc is also found to be dissolved at a high ionic strength, implying that this may not be a stable solid phase under very saline conditions. The uranyl-peroxo-bromo complex was shown to facilitate H2O2 decomposition via a mechanism involving reactive intermediates.

Studtite and meta-studtite are the only two uranyl peroxides found in nature. Sparsely soluble studtite has been found in natural uranium deposits, on the surface of spent nuclear fuel in contact with water and on core material from major nuclear accidents such as Chernobyl. The formation of studtite on the surface of nuclear fuel can have an impact on the release of radionuclides to the biosphere. In this work, we have experimentally studied the formation of studtite as function of HCO₃- concentration and pH. The results show that studtite can form at pH = 10 in solutions without added HCO₃-. At pH <= 7, the precipitate was found to be mainly studtite, while at 8 = pH = 9.8, a mixture of studtite and meta-schoepite was found. Studtite formation from UO22+ and H₂O₂ was observed at [HCO3-] <= 2 mM and studtite was only found to dissolve at [HCO₃-] > 2 mM.

Understanding the possible change in UO2 surface reactivity after exposure to oxidants is of key importance when assessing the impact of spent nuclear fuel dissolution on the safety of a repository for spent nuclear fuel. In this work, we have experimentally studied the change in UO2 reactivity after consecutive exposures to O2 or γ-radiation in aqueous solutions containing 10 mM HCO3-. The experiments show that the reactivity of UO2 toward O2 decreases significantly with time in a single exposure. In consecutive exposures, the reactivity also decreases from exposure to exposure. In γ-radiation exposures, the system reaches a steady state and the rate of uranium dissolution becomes governed by the radiolytic production of oxidants. Changes in surface reactivity can therefore not be observed in the irradiated system. The potential surface modification responsible for the change in UO2 reactivity was studied by XPS and UPS after consecutive exposures to either O2, H2O2, or γ-radiation in 10 mM HCO3- solution. The results show that the surfaces were significantly oxidized to a stoichiometric ratio of O/U of UO2.3 under all the three exposure conditions. XPS results also show that the surfaces were dominated by U(V) with no observed U(VI). The experiments also show that U(V) is slowly removed from the surface when exposed to anoxic aqueous solutions containing 10 mM HCO3-. The UPS results show that the outer ultrathin layer of the surfaces most probably contains a significant amount of U(VI). U(VI) may form upon exposure to air during the rinsing process with water prior to XPS and UPS measurements.

H2O2 is one of the oxidants responsible for driving the process of radiation-induced dissolution of spent nuclear fuel in geological repositories for spent nuclear fuel. As the groundwater composition will vary depending on geographical location as well as on the age of the repository (in relation to glacial cycles, etc.), it is important to elucidate the impact of different groundwater constituents. While several studies have addressed the impact of HCO3− and halide ions on the radiation chemistry of water in general and radiation-induced oxidative dissolution of spent nuclear fuel in particular, very few studies have addressed the impact of halide ions on the mechanism of the reaction between H2O2 and UO2. In this work, the impact of Cl−, Br− and ClO4− on the mechanism and kinetics of H2O2-induced oxidative dissolution of UO2-powder in aqueous suspensions with and without added HCO3− has been studied experimentally. The experiments reveal both ionic strength effects and specific ion effects on the kinetics of the reactions involved. These are discussed in connection to the results.

The interfacial radiation chemistry of UO2 is of key importance in the development of models to predict the corrosion rate of spent nuclear fuel in contact with groundwater. Here, the oxidative dissolution of UO2 induced by radiolytically produced H2O2 is of particular importance. The difficulty of fitting experimental data to simple first-order kinetics suggests that additional factors need to be considered when describing the surface reaction between H2O2 and UO2. It has been known for some time that UO22+ forms stable uranyl peroxo-carbonato complexes in water containing H2O2 and HCO3-/CO32-, yet this concept has largely been overlooked in studies where the oxidative dissolution of UO2 is considered. In this work, we show that uranyl peroxocarbonato complexes display little to no reactivity toward the solid UO2 surface in 10 mM bicarbonate solution (pH 8-10). The rate of peroxide consumption and UO22+ dissolution will thus depend on the UO22+ concentration and becomes limited by the free H2O2 fraction. The rate of peroxide consumption and the subsequent UO22+ dissolution can be accurately predicted based on the firstorder kinetics with respect to free H2O2, taking the initial H2O2 surface coverage into account.