Plutonium recovery from spent reactor fuel by uranium displacement

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Description

This report discusses a process for separating uranium values and transuranic values from fission products containing rare earth values when the values which are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the … continued below

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18 p.

Creation Information

Ackerman, J. P. December 31, 1991.

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Description

This report discusses a process for separating uranium values and transuranic values from fission products containing rare earth values when the values which are contained together in a molten chloride salt electrolyte. A molten chloride salt electrolyte with a first ratio of plutonium chloride to uranium chloride is contacted with both a solid cathode and an anode having values of uranium and fission products including plutonium. A voltage is applied across the anode and cathode electrolytically to transfer uranium and plutonium from the anode to the electrolyte while uranium values in the electrolyte electrolytically deposit as uranium metal on the solid cathode in an amount equal to the uranium and plutonium transferred from the anode causing the electrolyte to have a second ratio of plutonium chloride to uranium chloride. Then the solid cathode with the uranium metal deposited thereon is removed and molten cadmium having uranium dissolved therein is brought into contact with the electrolyte resulting in chemical transfer of plutonium values from the electrolyte to the molten cadmium and transfer of uranium values from the molten cadmium to the electrolyte until the first ratio of plutonium chloride to uranium chloride is re-established.

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18 p.

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OSTI; NTIS; GPO Dep.

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  • Other Information: PBD: 1991

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  • Other: DE93011998
  • Report No.: PATENTS-US--A7703641
  • Grant Number: W-31109-ENG-38
  • Office of Scientific & Technical Information Report Number: 10147537
  • Archival Resource Key: ark:/67531/metadc1314596

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

Office of Scientific and Technical Information (OSTI) is the Department of Energy (DOE) office that collects, preserves, and disseminates DOE-sponsored research and development (R&D) results that are the outcomes of R&D projects or other funded activities at DOE labs and facilities nationwide and grantees at universities and other institutions.

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  • December 31, 1991

Added to The UNT Digital Library

  • Nov. 3, 2018, 11:47 a.m.

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  • March 11, 2019, 1:26 p.m.

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Ackerman, J. P. Plutonium recovery from spent reactor fuel by uranium displacement, patent, December 31, 1991; Illinois. (https://digital.library.unt.edu/ark:/67531/metadc1314596/: accessed July 16, 2024), University of North Texas Libraries, UNT Digital Library, https://digital.library.unt.edu; crediting UNT Libraries Government Documents Department.

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