SRS vitrification studies in support of the U.S. program for disposition of excess plutonium

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Many thousands of nuclear weapons are being retired in the U.S. and Russian as a result of nuclear disarmament activities. These efforts are expected to produce a surplus of about 50 MT of weapons grade plutonium (Pu) in each country. In addition to this inventory, the U.S. Department of Energy (DOE) has more than 20 MT of Pu scrap, residue, etc., and Russian is also believed to have at least as much of this type of material. The entire surplus Pu inventories in the U.S. and Russian present a clear and immediate danger to national and international security. It is ... continued below

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15 p.

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Wicks, G.G.; McKibben, J.M.; Plodinec, M.J. & Ramsey, W.G. September 1, 1995.

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Description

Many thousands of nuclear weapons are being retired in the U.S. and Russian as a result of nuclear disarmament activities. These efforts are expected to produce a surplus of about 50 MT of weapons grade plutonium (Pu) in each country. In addition to this inventory, the U.S. Department of Energy (DOE) has more than 20 MT of Pu scrap, residue, etc., and Russian is also believed to have at least as much of this type of material. The entire surplus Pu inventories in the U.S. and Russian present a clear and immediate danger to national and international security. It is important that a solution be found to secure and manage this material effectively and that such an effort be implemented as quickly as possible. One option under consideration is vitrification of Pu into a safe, durable, accountable and proliferation-resistant form. As a result of decades to experience within the DOE community involving vitrification of a variety of hazardous and radioactive wastes, this existing technology can now be expanded to include mobilization of large amounts of Pu. This technology can then be implemented rapidly using the many existing resources currently available. An overall strategy to vitrify many different types of Pu will be already developed throughout the waste management community can be used in a staged Pu vitrification effort. This approach uses the flexible vitrification technology already available and can even be made portable so that it may be brought to the source and ultimately, used to produce a consistent and common borosilicate glass composition for the vitrified Pu. The final composition of this product can be made similar to nationally and internationally accepted HLW glasses.

Physical Description

15 p.

Notes

INIS; OSTI as DE95017489

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  • North Atlantic Treaty Organization (NATO) international scientific exchange program advanced research workshop, St. Petersburg (Russian Federation), 14-17 May 1995

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  • Other: DE95017489
  • Report No.: WSRC-MS--95-0151
  • Report No.: CONF-9505238--3
  • Grant Number: AC09-89SR18035
  • Office of Scientific & Technical Information Report Number: 102406
  • Archival Resource Key: ark:/67531/metadc619394

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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.

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Creation Date

  • September 1, 1995

Added to The UNT Digital Library

  • June 16, 2015, 7:43 a.m.

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  • Feb. 9, 2016, 7:05 p.m.

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Wicks, G.G.; McKibben, J.M.; Plodinec, M.J. & Ramsey, W.G. SRS vitrification studies in support of the U.S. program for disposition of excess plutonium, article, September 1, 1995; Aiken, South Carolina. (digital.library.unt.edu/ark:/67531/metadc619394/: accessed May 24, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.