TRUEX partitioning from radioactive ICPP sodium bearing waste Metadata

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  • Main Title TRUEX partitioning from radioactive ICPP sodium bearing waste


  • Author: Herbst, R. S.
    Creator Type: Personal
  • Author: Brewer, K. N.
    Creator Type: Personal
  • Author: Tranter, T. J.
    Creator Type: Personal
  • Author: Todd, T. A.
    Creator Type: Personal


  • Sponsor: United States. Department of Energy.
    Contributor Type: Organization
    Contributor Info: USDOE, Washington, DC (United States)


  • Name: Fermi National Accelerator Laboratory
    Place of Publication: Batavia, Illinois


  • Creation: 1995-03


  • English


  • Content Description: The Idaho Chemical Processing Plant (ICPP) located at the Idaho National Engineering Laboratory in Southeast Idaho is currently evaluating several treatment technologies applicable to waste streams generated over several decades of-nuclear fuel reprocessing. Liquid sodium bearing waste (SBW), generated primarily during decontamination activities, is one of the waste streams of interest. The TRansUranic EXtraction (TRUEX) process developed at Argonne National Laboratory is currently being evaluated to separate the actinides from SBW. On a mass basis, the amount of the radioactive species in SBW are low relative to inert matrix components. Thus, the advantage of separations is a dramatic decrease in resulting volumes of high activity waste (HAW) which must be dispositioned. Numerous studies conducted at the ICPP indicate the applicability of the TRUEX process has been demonstrated; however, these studies relied on a simulated SBW surrogate for the real waste. Consequently, a series of batch contacts were performed on samples of radioactive ICPP SBW taken from tank WM-185 to verify that actual waste would behave similarly to the simulated waste. The test results with SBW from tank WM-185 indicate the TRUEX solvent effectively extracts the actinides from the samples of actual waste. Gross alpha radioactivity, attributed predominantly to Pu and Am, was reduced from 3.14E+04 dps/mL to 1.46 dps/mL in three successive batch contacts with fresh TRUEX solvent. This reduction corresponds to a decontamination factor of DF = 20,000 or 99.995% removal of the gross a activity in the feed. The TRUEX solvent also extracted the matrix components Zr, Fe, and Hg to an appreciable extent (D{sub Zr} > 10, D{sub Fe} {approx} 2, D{sub Hg} {approx}6). Iron co-extracted with the actinides can be successfully scrubbed from the organic with 0.2 M HNO{sub 3}. Mercury can be selectively partitioned from the actinides with either sodium carbonate or nitric acid ({ge} 5 M HNO{sub 3}) solutions.
  • Physical Description: 23 p.


  • Keyword: Alpha-Bearing Wastes
  • Keyword: Reprocessing
  • Keyword: Idaho Chemical Processing Plant
  • Keyword: Separation Processes
  • STI Subject Categories: 05 Nuclear Fuels
  • Keyword: Performance Testing
  • Keyword: Truex Process
  • Keyword: Spent Fuels
  • Keyword: Radioactive Waste Management
  • Keyword: Evaluation


  • Other Information: PBD: Mar 1995


  • Name: Office of Scientific & Technical Information Technical Reports
    Code: OSTI


  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Report


  • Text


  • Other: DE96001188
  • Report No.: INEL--95/0224
  • Grant Number: AC07-94ID13223
  • DOI: 10.2172/114583
  • Office of Scientific & Technical Information Report Number: 114583
  • Archival Resource Key: ark:/67531/metadc626091


  • Display Note: INIS; OSTI as DE96001188