Separating nuclides from spent nuclear fuel is crucial to reduce the final disposal area. The use of molten salt offers a potential method for nuclide separation without requiring electricity, similar to the oxide reduction process in pyroprocessing. In this study, a molten salt leaching technique was evaluated for its ability to separate nuclides from simulated oxide fuel in MgCl2 molten salts at 800°C. The simulated oxide fuel contained 2wt% Sr, 3wt% Ba, 2wt% Ce, 3wt% Nd, 3wt% Zr, 2wt% Mo, and 89wt% U. The separation of Sr from the simulated oxide fuel was achieved by loading it into a porous alumina basket and immersing it in the molten salt. The concentration of Sr in the salt was measured using ICP analysis after sampling the salt outside the basket with a dip-stick technique. The separated nuclides were analyzed with ICP-OES up to a duration of 156 hours. The results indicate that Ba and Sr can be successfully separated from the simulated fuel in MgCl2, while Ce, Nd, and U were not effectively separated.

• Sang-Kwon Lee(Korea Atomic Energy Research Institute (KAERI)) Corresponding author
• Dokyu Kang(Korea Atomic Energy Research Institute (KAERI))
• Ju Ho Lee(Korea Atomic Energy Research Institute (KAERI))
• Eun-Young Choi(Korea Atomic Energy Research Institute (KAERI))
• Chang Hwa Lee(Korea Atomic Energy Research Institute (KAERI))