Abstract
Purification of U by fused salt electrolysis is one method of pyrometallurgically reprocessing irradiated nuclear fuel. In the present study, dummy spent fuel constituted of U alloys containing Zr and Mo to simulate fission products were used for the anode. Molten zinc was chosen for cathode, for its adaptability to continuous operation. The electrolytic refining was carried out in an NaCl-KCl-UF4 fused salt system at 750°C, and a KCl-LiCl-UCl3 system at 500°600°C.
In the case of NaCl-KCl-UF4 system, at 750°C and with cathodic current densities of 190400mA/cm2, the cathodic current efficiencies obtained were 2060% with a bath voltage in the range 13V. Continuous operation however was disturbed by dendritic deposits on the cathode container. In the case of KCl-LiCl-UCl3 system, the data were UCl3 concentration 10_??_ at 500°C, cathodic current density 100250mA/cm2, bath voltage 1.22.5V, and cathodic current efficiencies obtained 4090%. The electrolysis in this case could be maintained during an appreciable period.
The U was recovered by vacuum distillation of the U-Zn alloy obtained on the cathode. Calculating the results of chemical analysis of the Zr and Mo contents before and after electrolysis, the decontamination factors were determined to be 1040 for Zr, and 30 for Mo.
