Cesium sorption on studtite (UO2O2·4H2O)

Americium incorporation into studtite: a theoretical and experimental study

Studtite, [UO₂(η²-O₂)(H₂O)₂]·2H₂O, and metastudtite, [UO₂(η²-O₂)(H₂O)₂], are important phase alterations of UO₂ in a spent nuclear fuel repository and have previously been shown to react with Np(v). In this work we extend the study to Am(v) on a tracer scale and show spectroscopic evidence that the Am is incorporated into the structure of studtite as Am(iii). A computational study on the possible mechanisms for the incorporation of Np and Am shows that protonation of the -yl oxygen is the favoured route and the calculated incorporation energies are large and positive. The results suggest that Am is less favoured compared to Np but energetically more favoured to incorporate both actinide ions into metastudtite rather than studtite. Finally, we have shown that once incorporated, Am readily leaches into water but spectroscopic measurements suggest subtle changes in the structure of studtite.

Dissolution of studtite [UO2(O2)(H2O)4] in various geochemical conditions

This study determined the dissolution rate of studtite, (UO₂)O₂(H₂O)₄, which can be formed by reaction between H₂O₂ and UO₂²⁺ that leaks from spent nuclear fuel (SNF) in deep geological repositories. The batch dissolution experiments were conducted using synthesized studtite under different solution conditions with varying pHs and concentrations of HCO₃^- and [H₂O₂] in synthetic groundwater. The experimental results suggested that carbonate ligand and H₂O₂ in groundwater accelerated the dissolution of studtite and uranium (U) release. Above 10^-5 M of H₂O₂ initial concentration, the released uranium concentration in solution decreased, possibly as a result of reprecipitation of studtite due to reaction between uranium and H₂O₂. The results will be useful to assess the comprehensive transport of uranium from both nuclear waste and SNF stored in deep geological repositories.

Retention of cesium and strontium by uranophane, Ca(UO2)2(SiO3OH)2·5H2O

This work determines the capacity of uranophane, one of the long-term uranyl secondary solid phases formed on the spent nuclear fuel (SNF), to retain radionuclides (cesium and strontium) released during the dissolution of the SNF. Sorption was fast in both cases, and uranophane had a high sorption capacity for both radionuclides (maximum sorption capacities of 1.53·10^-5 mol m^-2 for cesium and 3.45·10^-3 mol m^-2 for strontium). The high sorption capacity of uranophane highlights the importance of the formation of uranyl silicates as secondary phases during the SNF dissolution, especially in retaining the release of radionuclides not retarded by other mechanisms such as precipitation.

Radionuclide behaviour in the near-field of a geological repository for spent nuclear fuel

Even though chemical processes related to the corrosion of spent nuclear fuel in a deep geological repository are of complex nature, knowledge on underlying mechanisms has very much improved over the last years. As a major result of numerous studies it turns out that alteration of irradiated fuel is significantly inhibited under the strongly reducing conditions induced by container corrosion and consecutive H2 production. In contrast to earlier results, radiolysis driven fuel corrosion and oxidative dissolution appears to be less relevant for most repository concepts. The protective hydrogen effect on corrosion of irradiated fuel has been evidenced in many experiments. Still, open questions remain related to the exact mechanism and the impact of potentially interfering naturally occurring groundwater trace components. Container corrosion products are known to offer considerable reactive surface area in addition to engineered buffer and backfill material. In combination, waste form, container corrosion products and backfill material represent strong barriers for radionuclide retention and retardation and thus attenuate radionuclide release from the repository near-field.