Immobilization of spent electrolyte–radioactive waste (RW) generated during the pyrochemical processing of mixed nitride uranium–plutonium spent nuclear fuel is an acute task for further development of the closed nuclear fuel cycle with fast neutron reactors. The electrolyte is a mixture of chloride salts that cannot be immobilized directly in conventional cement or glass matrix. In this work, a low-temperature magnesium potassium phosphate (MPP) matrix and two types of high-temperature matrices (sodium aluminoironphosphate (NAFP) glass and ceramics based on bentonite clay) were synthesized. Two systems (Li_0.4K_0.28La_0.08Cs_0.016Sr_0.016Ba_0.016Cl and Li_0.56K_0.40Cs_0.02Sr_0.02Cl) were used as spent electrolyte imitators. The phase composition and structure of obtained materials were studied by XRD and SEM-EDS methods. The differential leaching rate of Cs from MPP compound and ceramic based on bentonite clay was about 10⁻⁵ g/(cm²·day), and the rate of Na from NAFP glass was about 10⁻⁶ g/(cm²·day). The rate of ²³⁹Pu from MPP compound (leaching at 25 °C) and NAFP glass (leaching at 90 °C) was about 10⁻⁶ and 10⁻⁷ g/(cm²·day), respectively. All the synthesized materials demonstrated high hydrolytic, mechanical compression strength (40–50 MPa) even after thermal (up to 450 °C) and irradiation (up to 10⁹ Gy) tests. The characteristics of the studied matrices correspond to the current requirements to immobilized high-level RW, that allow us to suggest these materials for industrial processing of the spent electrolyte.

Possibility of using a low-temperature magnesium-potassium phosphate matrix to solve the problem of immobilizing the radioactive wastes containing radioactive carbon (14C) in the form of calcium...

Pyrochemistry is a promising technology that can provide benefits for the safe reprocessing of relatively fresh spent nuclear fuel with a short storage time (3–5 years). The radioactive waste emanating from this process is an electrolyte (LiCl–KCl) mixture with fission products included. Such wastes are rarely immobilized through common matrices such as cement and glass. In this study, samples of ceramic materials, based on natural bentonite clay, were studied as matrices for radioactive waste in the form of LiCl–KCl eutectic. The phase composition of the samples, and their mechanical, hydrolytic, and radiation resistance were characterized. The possibility of using bentonite clay as a material for immobilizing high-level waste arising from pyrochemical processing of spent nuclear fuel is further discussed in this paper.