Title: Co-processing of uranium and plutonium for sodium-cooled fast reactor fuel reprocessing by acid split method for plutonium partitioning without reductant
Abstract: 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 ...
Publication Year: 2013
Publication Date: 2013-11-01
Language: en
Type: article
Indexed In: ['crossref']
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Cited By Count: 4
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