Perspective compounds for immobilization of spent electrolyte from pyrochemical processing of spent nuclear fuel

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 (Li0.4K0.28La0.08Cs0.016Sr0.016Ba0.016Cl and Li0.56K0.40Cs0.02Sr0.02Cl) 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.