Influence of plutonium content in dissolver solutions derived from irradiated fast reactor fuels on plutonium stripping in multistage countercurrent liquid–liquid extraction with acid split flowsheet

In the acid split flowsheet, the Pu content in the dissolver solution of nuclear fuel has a significant impact on the configuration of flowsheet and the Pu content in the products. The influence of Pu content in dissolver solutions of nuclear fuel on the Pu stripping has been studied. The decontamination performance of Cs as major fission products was assessed, and Cs was decontaminated well in the acid split flowsheet. High HNO3 concentration in the feed solution and at the co-decontamination section was adjusted for the co-recovery of Np with U and Pu, contributing to Np oxidation and the increase in the distribution ratio of Np. The experimental results demonstrate that the Pu content in the U/Pu and U products increased with an increase in the Pu content in the feed solution. The calculated results indicate that the Pu leakage into the U product is suppressed with the Pu stripping solution only at low temperature under our experimental conditions, and Pu is recovered into the U/Pu product efficiently. This study proves that the Pu leakage into the U product is controlled and the acid split flowsheet is flexible with change in the Pu content in the feed solution.