Yuanzineng kexue jishu (Mar 2023)
Investigation of Defluorination of Molten Fluoride Salt Nuclear Waste with Oxalic Acid
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 highlevel 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.