Removal of Cs+ from chloride-containing spent salt by combined precipitation and ion exchange and its glass solidification

In order to reduce the volume of nuclear waste, the spent salt after electrolysis refining of spent fuel needs to be purified and the removed radioactive nuclide need to be vitrified. In this work, the radioactive nuclides Cs⁺ in spent salt was removed by a two-step method, and the waste containing Cs after removal was solidified. The results showed that the total removal rate of Cs⁺ in spent salt could reach 94 %, and the maximum inclusion ratio of borosilicate glass to precipitated cesium phosphate (Cs₃PO₄) and zeolite-4A containing cesium could reach 30 wt% and 35 wt%, respectively. The glass samples with the embedding ratio less than or equal to the maximum were found to be all amorphous, uniform and dense. The normalized leaching rate (NR_i) of Cs⁺ in the two types of glass samples were 3.245 × 10⁻⁴ g m⁻² d⁻¹ and 1.145 × 10⁻⁵ g m⁻² d⁻¹ on the 28th day, respectively, showing good chemical durability, which met the geological disposal standards.