Preconceptual design for separation of plutonium and gallium by ion exchange

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The disposition of plutonium from decommissioned nuclear weapons, by incorporation into commercial UO{sub 2}-based nuclear reactor fuel, is a viable means to reduce the potential for theft of excess plutonium. This fuel, which would be a combination of plutonium oxide and uranium oxide, is referred to as a mixed oxide (MOX). Following power generation in commercial reactors with this fuel, the remaining plutonium would become mixed with highly radioactive fission products in a spent fuel assembly. The radioactivity, complex chemical composition, and large size of this spent fuel assembly, would make theft difficult with elaborate chemical processing required for plutonium ... continued below

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33 p.

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DeMuth, S. F. September 30, 1997.

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Description

The disposition of plutonium from decommissioned nuclear weapons, by incorporation into commercial UO{sub 2}-based nuclear reactor fuel, is a viable means to reduce the potential for theft of excess plutonium. This fuel, which would be a combination of plutonium oxide and uranium oxide, is referred to as a mixed oxide (MOX). Following power generation in commercial reactors with this fuel, the remaining plutonium would become mixed with highly radioactive fission products in a spent fuel assembly. The radioactivity, complex chemical composition, and large size of this spent fuel assembly, would make theft difficult with elaborate chemical processing required for plutonium recovery. In fabricating the MOX fuel, it is important to maintain current commercial fuel purity specifications. While impurities from the weapons plutonium may or may not have a detrimental affect on the fuel fabrication or fuel/cladding performance, certifying the effect as insignificant could be more costly than purification. Two primary concerns have been raised with regard to the gallium impurity: (1) gallium vaporization during fuel sintering may adversely affect the MOX fuel fabrication process, and (2) gallium vaporization during reactor operation may adversely affect the fuel cladding performance. Consequently, processes for the separation of plutonium from gallium are currently being developed and/or designed. In particular, two separation processes are being considered: (1) a developmental, potentially lower cost and lower waste, thermal vaporization process following PuO{sub 2} powder preparation, and (2) an off-the-shelf, potentially higher cost and higher waste, aqueous-based ion exchange (IX) process. While it is planned to use the thermal vaporization process should its development prove successful, IX has been recommended as a backup process. This report presents a preconceptual design with material balances for separation of plutonium from gallium by IX.

Physical Description

33 p.

Notes

INIS; OSTI as DE98001666

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  • Other Information: PBD: 30 Sep 1997

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  • Other: DE98001666
  • Report No.: LA-UR--97-3769-Rev.1
  • Grant Number: W-7405-ENG-36
  • DOI: 10.2172/555275 | External Link
  • Office of Scientific & Technical Information Report Number: 555275
  • Archival Resource Key: ark:/67531/metadc699183

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Office of Scientific & Technical Information Technical Reports

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  • September 30, 1997

Added to The UNT Digital Library

  • Aug. 14, 2015, 8:43 a.m.

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  • Feb. 26, 2016, 3:24 p.m.

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DeMuth, S. F. Preconceptual design for separation of plutonium and gallium by ion exchange, report, September 30, 1997; New Mexico. (digital.library.unt.edu/ark:/67531/metadc699183/: accessed November 20, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.