Preliminary evaluation of solvent-extraction and/or ion-exchange process for meeting AAA program multi-tier systems recovery and purification goals.

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Description

Several potential processes are described and evaluated for their suitability in a multitier aqueous-based approach to processing dissolved spent nuclear fuel under the Advanced Accelerator Applications (AAA) program. The evaluation is focused on solvent extraction and ion exchange technologies that have been demonstrated to varying degrees. The goals of the program are to separate uranium (U), technetium (Tc), and the transuranic (TRU) elements from the fission products that are to be vitrified for disposal as high-level waste (HLW). Uranium will be disposed as low-level waste (LLW); Tc and TRU will be transmuted in an accelerator. A number of processes have ... continued below

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27 pages

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Pereira, C.; Vandegrift, G. F. & Swanson, J. L. October 17, 2002.

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Description

Several potential processes are described and evaluated for their suitability in a multitier aqueous-based approach to processing dissolved spent nuclear fuel under the Advanced Accelerator Applications (AAA) program. The evaluation is focused on solvent extraction and ion exchange technologies that have been demonstrated to varying degrees. The goals of the program are to separate uranium (U), technetium (Tc), and the transuranic (TRU) elements from the fission products that are to be vitrified for disposal as high-level waste (HLW). Uranium will be disposed as low-level waste (LLW); Tc and TRU will be transmuted in an accelerator. A number of processes have been examined. The focus was on liquid-liquid solvent extraction processes because of their relatively high state of development and their suitability for high-throughput-rate processing. Ion exchange processes were also examined. PUREX and UREX were evaluated as options for recovery of uranium; UREX is also an option for Tc recovery. Solvent extraction options examined for TRU recovery included TRUEX, DIAMEX, and TRPO, as well as some based on TBP extraction. Processes for trivalent actinide separation from lanthanides were also examined. The PUREX processes have been developed over many years, and have been refined to a significant degree. In the first cycle, U and plutonium (Pu) are co-extracted from dissolved spent fuel solutions by TBP in a diluent, selectively stripped, and purified in additional extraction/strip cycles. Neptunium (Np) can be extracted or driven into the raffinate by adjusting the oxidation state. Trivalent actinides and fission products remain predominantly in the raffinate, although a significant Tc fraction will co-extract with the U and Pu.

Physical Description

27 pages

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  • Other Information: PBD: 17 Oct 2002

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  • Report No.: ANL-02/29
  • Grant Number: W-31-109-ENG-38
  • DOI: 10.2172/805261 | External Link
  • Office of Scientific & Technical Information Report Number: 805261
  • Archival Resource Key: ark:/67531/metadc734485

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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

  • October 17, 2002

Added to The UNT Digital Library

  • Oct. 18, 2015, 6:40 p.m.

Description Last Updated

  • March 25, 2016, 2:20 p.m.

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Pereira, C.; Vandegrift, G. F. & Swanson, J. L. Preliminary evaluation of solvent-extraction and/or ion-exchange process for meeting AAA program multi-tier systems recovery and purification goals., report, October 17, 2002; Illinois. (digital.library.unt.edu/ark:/67531/metadc734485/: accessed October 20, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.