PLUTONIUM-238 RECOVERY FROM IRRADIATED NEPTUNIUM TARGETS USING SOLVENT EXTRACTION

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The United States Department of Energy proposes to re-establish a domestic capability for producing plutonium-238 (238Pu) to fuel radioisotope power systems primarily in support of future space missions. A conceptual design report is currently being prepared for a new 238Pu, and neptunium-237 (237Np) target fabrication and processing facility tentatively to be built at the Idaho National Laboratory (INL) in the USA. The facility would be capable of producing at least 5 kg of 238Pu-oxide powder per year. Production of 238Pu requires fabrication of 237Np targets with subsequent irradiation in the existing Advanced Test Reactor (ATR) located at the INL. The ... continued below

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Herbst, Scott; Todd, Terry; Law, Jack; Mincher, Bruce; Frank, Steve & Swanson, John October 1, 2006.

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The United States Department of Energy proposes to re-establish a domestic capability for producing plutonium-238 (238Pu) to fuel radioisotope power systems primarily in support of future space missions. A conceptual design report is currently being prepared for a new 238Pu, and neptunium-237 (237Np) target fabrication and processing facility tentatively to be built at the Idaho National Laboratory (INL) in the USA. The facility would be capable of producing at least 5 kg of 238Pu-oxide powder per year. Production of 238Pu requires fabrication of 237Np targets with subsequent irradiation in the existing Advanced Test Reactor (ATR) located at the INL. The targets are 237Np oxide dispersed in a compact of powdered aluminum and clad with aluminum metal. The 238Pu product is separated and purified from the residual 237Np, aluminum matrix, and fission products. The unconverted 237Np is also a valuable starting material and is separated, purified and recycled to the target fabrication process. The proposed baseline method for separating and purifying 238Pu and unconverted 237Np post irradiation is by anion exchange (IX). Separation of Pu from Np by IX was chosen as the baseline method because of the method’s proven ability to produce a quality Pu product and because it is amenable to the relatively small scale, batch type production methods used (small batches of ~200g 238Pu are processed at a time). Multiple IX cycles are required involving substantial volumes of nitric acid and other process solutions which must be cleaned and recycled or disposed of as waste. Acid recycle requires rather large evaporator systems, including one contained in a hot cell for remote operation. Finally, the organic based anion exchange resins are rapidly degraded due to the high a-dose and associated heat production from 238Pu decay, and must be regularly replaced (and disposed of as waste). In summary, IX is time consuming, cumbersome, and requires substantial tankage to accommodate the process. The primary purpose of the preliminary study discussed here is to develop an alternative process flowsheet using well-known solvent extraction (SX) techniques based on decades of experience with PUREX processing of nuclear materials. Ultimately, this initial study will be used to determine if an SX approach would offer any significant processing advantages relative to the currently proposed anion exchange process.

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  • 15th Pacific Basin Nuclear Conference,Sydney, Australia,10/15/2006,10/20/2006

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  • Report No.: INL/CON-06-11850
  • Grant Number: DE-AC07-99ID-13727
  • Office of Scientific & Technical Information Report Number: 911818
  • Archival Resource Key: ark:/67531/metadc889460

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  • October 1, 2006

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  • Sept. 22, 2016, 2:13 a.m.

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  • Nov. 7, 2016, 6:56 p.m.

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Herbst, Scott; Todd, Terry; Law, Jack; Mincher, Bruce; Frank, Steve & Swanson, John. PLUTONIUM-238 RECOVERY FROM IRRADIATED NEPTUNIUM TARGETS USING SOLVENT EXTRACTION, article, October 1, 2006; [Idaho Falls, Idaho]. (digital.library.unt.edu/ark:/67531/metadc889460/: accessed August 24, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.