Investigation of the Distillation Characteristics of LiCl-Li 2 O Molten Electrolyte

The metal product of the electrolytic reduction of oxide spent nuclear fuel in molten LiCl-Li2O mixtures retains about 8-30 wt% of the residual LiCl-Li2O salt. The salt occluded in the uranium metal should be removed by vacuum distillation. The purpose of this work is to study the distillation of the LiCl-Li2O mixture (without uranium). We have tested different distillation regimes. The distillation of LiCl-Li2O melts (3 wt.%) was carried out in quartz tubes at 750-800°C and the pressure of (1-3)*10-2 mm Hg for 30-70 min. Herewith, the initial mixture was kept in the MgO crucible. The crucible was weighed before and after the experiment. The salt remaining in the crucible and the sublimates were analyzed for Li2O content. The amount of Li2O was determined by titration of the aqueous solutions of the salts with a pH meter, and also by ICP-AES. The results of our studies do not confirm the conclusions of ref. [1] on the significant co-evaporation of hardly volatile lithium oxide together with more volatile LiCl during distillation. In our experiments, the fraction of evaporated lithium oxide remained insignificant (less than 1-3%), regardless of the fraction of evaporated LiCl (25-95%). The loss of large amounts of Li2O from the initial mixture of LiCl-Li2O is possible either with spattering of the melt in the case of rapid evaporation of LiCl, or as a result of a side reaction of the uranium metal oxidation with Li2O in residual salts: U + 2Li2O * UO2 + 4Li. This reaction becomes probable at temperatures above 850-900°C due to the evaporation of metallic lithium at such temperatures.