Dechlorination Of Molten Salt Nuclear Waste And Vitrification Of The Remaining Waste

Electrochemical reprocessing of spent nuclear fuel in molten salt has become one of the most promising technologies for advanced fuel cycle,owing to its simple process and less produced radioactive wastes.During this process,chloride molten salt is usually used as electrolyte,which results in a large amount of molten chloride salt nuclear waste after electrolysis.Such waste is classified as high-level waste and 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,the low solubility of Cl in glass limits the application of vitrification technology for the treatment of molten chloride salt nuclear waste.Thus,the treatment of molten chloride salt nuclear waste is still one of the difficulties in the dry reprocessing of spent fuel.In order to solve the above problem,this work proposes a two-step treatment process,firstly,dechlorinating the molten salt nuclear waste with oxalic acid(H2C2O4)and then immobilizing the remaining waste in a waste glass form.The dechlorination process and the mechanism of Cl removal of simulated molten chloride salt nuclear waste using H2C2O4 were mainly investigated.Vitrification experiments using borosilicate glass additives were carried out based on the chemical composition of the dechlorinated waste,and a series of glass formulations were obtained according to the chemical durability of waste glasses.Finally,the simulated molten fluoride salt nuclear waste was taken to verify the feasibility of defluorination with H2C2O4,and the defluorinated waste was immobilized in a borosilicate glass.The specific research results are as follows:(1)H2C2O4 could effectively remove Cl from the simulated molten chloride salt nuclear waste.When the molar ratio of H2C2O4 to Cl is 2,amd the thermal treatment temperature is 300 °C,the chlorine removal efficiency could reach up to 99%;During the dechlorination,the Cl was removed from the waste in the form of HCl gas,while metal cations were left in the remaining waste and form metal oxalates,which were converted to metal carbonates at temperatures above 500 °C.(2)Based on the composition of dechlorinated waste,vitrification experiments using Si O2,B2O3,Ca O,and Al2O3 as glass additives were carried out with waste loadings ranging from 25 wt% to 45 wt%.The 7-day product consistency test shows that the normalized elemental releases of the waste glass with a waste loading of less than 35 wt% were lower than 2.0 g/m2.The glass formula with a fixed waste loading of 30 wt% was optimized by adjusting the composition of glass additives.The 28-day static leaching test suggested that the mass loss per unit surface area of the optimized waste glasses was reduced from ~30 g/m2 to lower than 15 g/m2,and the waste glasses meeting the durability requirements of the domestic nuclear industry could be obtained.(3)The validation experiment of defluorination shows that H2C2O4 could also remove F from simulated molten fluoride salt nuclear waste.When the molar ratio of H2C2O4 to F was 2 and the thermal treatment temperature was 300 °C,the F removal efficiency could reach up to 93%.And the metal cations in the waste formed oxalates that were completely converted to metal carbonates above 500 °C.The defluorinated waste thus obtained was then immobilized in a borosilicate glass waste form,and the waste loading could reach more than 25 wt%.