Dechlorination and vitrification of electrochemical processing salt waste

Abstract

Electrochemical reprocessing of spent nuclear fuel in molten chloride salt is a promising strategy for the advanced nuclear fuel cycle. However, the generated chloride nuclear waste cannot be vitrified into borosilicate glass due to the low solubility of Cl in the glass. To dechlorinate salt waste before immobilizing it in the borosilicate glass, H2C2O4 could be used at a temperature below 300 °C. The chlorine removal efficiency could reach up to 99% if the dechlorination parameters, such as thermal treatment temperature and the molar ratio of H2C2O4 to Cl, were optimized. The Cl was removed as HCl during dechlorination, while the remaining cations which formed oxalates were decomposed into carbonates at a temperature above 500 °C. The residual carbonates were finally vitrified into a borosilicate waste glass form, and the waste loading could reach up to 35 wt%, at which the PCT-7 normalized elemental (Li, Na, K, Cs, and B) releases were lower than 1.5 g/m2. This work can be treated as a practical approach to treat chloride nuclear waste.