Corrosion Mechanisms of Spent Fuel Under Oxidizing Conditions

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.

Effects of ionizing radiation and temperature on uranyl silicates: soddyite (UO2)2(SiO4)(H2O)2 and Uranophane Ca(UO2)2(SiO3OH)2·5H2O

The stability of soddyite under electron irradiation has been studied over the temperature range of 25-300 °C. At room temperature, soddyite undergoes a crystalline-to-amorphous transformation (amorphization) at a total dose of 6.38 × 10(8) Gy. The electron beam irradiation results suggest that the soddyite structure is susceptible to radiation-induced nanocrystallization of UO(2). The temperature dependence of amorphization dose increases linearly up to 300 °C. A thermogravimetric and calorimetric analysis (TGA-DSC) combined with X-ray diffraction (XRD) indicates that soddyite retains its water groups up to 400 °C, followed by the collapse of the structure. Based on thermal analysis of uranophane, the removal of some water groups at relatively low temperatures provokes the collapse of the uranophane structure. This structural change appears to be the reason for the increase of amorphization dose at 140 °C. According to the results obtained, radiation field of a nuclear waste repository, rather than temperature effects, may cause changes in the crystallinity of soddyite and affect its stability during long-term storage.

Formation of studtite during the oxidative dissolution of UO2 by hydrogen peroxide: a SFM study

Understanding the formation of alteration phases on the surface of spent nuclear fuel, such as those observed during leaching experiments, is necessary in order to predict the concentration of radionuclides in the near-field of a final repository. Hydrogen peroxide has been identified as one of the oxidants formed by the radiolysis of water in the presence of spent nuclear fuel; especially due to alpha activity. The presence of this species in solution can contribute to the formation of uranium peroxide secondary phases. In this work, we have studied the oxidative dissolution of synthetic UO2 disks in hydrogen peroxide solutions of two different concentrations (5 x 10(-4) and 5 x 10(-6) mol dm(-3)), both at pH 5.8 +/- 0.1. The solid surface evolution of the disks has been followed by means of ex-situ scanning force microscope (SFM) measurements, and uranium concentration in solution has been determined by inductively coupled plasma mass spectrometry. During the first stage of the experiment, SFM images indicate that only UO2 dissolution is occurring. After 142 h, a secondary phase is observed on the surface of the solid at 5 x 10(-4) mol dm(-3) hydrogen peroxide concentration. This secondary phase has been identified by X-ray diffraction as studtite (UO4 x 4H2O). From the analysis of SFM topographic profiles at different elapsed times, a precipitation rate for the studtite has been estimated to be in the range of (8-32) x 10(-10) mol m(-2) s(-1).

The geochemical behaviour of Tc, Np and Pu in spent nuclear fuel in an oxidizing environment

Spent fuel from commercial nuclear reactors consists mainly of uranium oxide. However, the changes that occur during reactor operations have a profound effect on chemical and physical properties of this material. Heat build-up in the fuel pellet during reactor operations can cause redistribution of fission products. The fission products may aggregate in one of three types of precipitates; gaseous, metallic, or oxide, depending on the burn-up and in-core treatment. Radiation damage and variations in fission and neutron capture yields across the fuel pellets lead to Pu enrichment and increased porosity with increasing burn-up. A more porous surface may make the fuel more susceptible to oxidative dissolution. As the level of actinides and fission products increases, the fuel may become more resistant to oxidation. These changes may limit the usefulness of natural uraninite (UO2) analogues for predicting the geological behaviour of spent fuel disposed in a high-level waste (HLW) repository. In this Chapter, an overview of spent fuel microstructure, radiolytic effects, and alteration processes is presented. Evidence for Np incorporation into U6+ phases, the nature of Pu surface precipitates on spent fuel, and evidence for the preferential removal of 4d-metals from ε-particles in corroded spent fuel is discussed. Understanding the potential mechanisms of radionuclide attenuation through sorption and/or incorporation requires techniques with both high spatial resolution and excellent elemental sensitivity.