Medium-Temperature Glass-Composite Phosphate Materials for the Immobilization of Chloride Radioactive Waste

Among the many radiochemical problems, the search for new materials and technologies for the immobilization of radioactive waste remains relevant, and the range continues to change and expand. The possibility of immobilizing the spent chloride electrolyte after the pyrochemical processing of the mixed uranium-plutonium spent nuclear fuel of the new fast reactor BREST-OD-300 on lead coolant into glass-composite phosphate materials synthesized at temperatures of 650–750 °C was studied. The structure of the obtained samples was studied using XRD and SEM/EDS methods. It has been shown that the spent electrolyte simulator components create stable mixed pyrophosphate phases in the glass composite structure. The materials were found to have high hydrolytic stability. This indicates the promise of using phosphate glass composites as materials for the reliable immobilization of the spent electrolyte.