Abstract: The molten salt reactor, utilizing molten fluoride salt as fuel solvent and coolant, is one of the generation-Ⅳ nuclear reactors for advanced nuclear energy. The generated fluoride nuclear waste during the operation of molten salt reactor and reprocessing of spent fuel should be immobilized in a stable matrix before disposal. At present, vitrification is still the only technology in the world that could industrially immobilize highlevel waste. However, such fluoride waste cannot be directly vitrified into borosilicate glass since it contains a large amount of F, which could lead to oversaturation in the glass. In order to provide an effective solution for the safe treatment of molten fluoride salt waste, H₂C₂O₄ was used as a defluorination agent to defluorinate the simulated salt waste mainly consisting of alkali fluorides. The mixed sample of H₂C₂O₄ and simulated fluoride salt waste was characterized with TG-DSC-MS and XRD to analyze the endothermic and exothermic behaviors, released gases, and phase transitions at elevated temperatures. Afterward, the effects of thermal treatment temperature and the molar ratio of H₂C₂O₄ to F on the fluorine removal efficiency were investigated, and the optimal process parameters of defluorination were finally determined. The results show that H₂C₂O₄ could react with alkali fluoride salt to release HF gas besides being decomposed into H₂O, CO and CO₂ at temperatures between 100 ℃ and 300 ℃, while the alkali fluorides turned to alkali oxalates, which could decompose to alkali carbonates at temperatures of 500 ℃. The fluorine removal efficiency would reach up to 93% when the molar ratio of H₂C₂O₄ to F was 2 and the thermal treatment temperature was 300 ℃. The defluorinated waste thus obtained was then immobilized in a borosilicate glass waste form at about 1 200 °C with a waste loading of 25%, and the normalized elemental (B, Li, Na, K, Cs, Sr, and Ce) releases of the waste glass conducted with 7-day product consistency test were lower than 2.0 g/m2, showcasing acceptable durability for nuclear waste glass. The above results indicate that the proposed approach which defluorination with H₂C₂O₄ in the first step at temperatures of below 300 °C and vitrifying the remaining waste into a borosilicate glass in the second step, would provide a practical way to safely treat the molten fluoride salt nuclear waste.