Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant

A solvent extraction flowsheet for Pu partitioning, based on the acid split method without reductant, originally proposed by the Oak Ridge National Laboratory (ORNL), was tested for sodium-cooled fast reactor fuel reprocessing. To enhance resistance to nuclear proliferation, a flowsheet for co-processing was developed that controls Pu content in the products while avoiding Pu polymerization and formation of a third phase during extraction. In this method, Pu is partitioned using the difference in distribution coefficients of U and Pu. It is effective for selective Pu stripping from U at low temperatures and HNO₃ concentrations. The flowsheet with a supply of 0.15 mol/dm³ HNO₃ solution at 21°C for Pu partitioning was tested experimentally using miniature centrifugal contactors and a highly radioactive solution. Neither a Pu(IV) polymer nor a third phase was observed during the experiment. The Pu content in the U/Pu product increased to 2.28 times that in the feed solution. The leakage ratio of Pu to the U product was slightly less in the U stripping section. Some fission products (FPs) were effectively decontaminated; e.g., decontamination factors (DFs) of Cs in U/Pu and U products were 4.51×10⁵ and 2.42×10⁵, respectively.