Evaluation of Co-precipitation Processes for the Synthesis of Mixed-Oxide Fuel Feedstock Materials Metadata

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Title

  • Main Title Evaluation of Co-precipitation Processes for the Synthesis of Mixed-Oxide Fuel Feedstock Materials

Creator

  • Author: Collins, Emory D
    Creator Type: Personal
    Creator Info: ORNL
  • Author: Voit, Stewart L
    Creator Type: Personal
    Creator Info: ORNL
  • Author: Vedder, Raymond James
    Creator Type: Personal
    Creator Info: ORNL

Contributor

  • Sponsor: United States. Office of the Assistant Secretary for Nuclear Energy.
    Contributor Type: Organization
    Contributor Info: NE USDOE - Office of Nuclear Energy

Publisher

  • Name: Oak Ridge National Laboratory
    Place of Publication: United States

Date

  • Creation: 2011-06-01

Language

  • English

Description

  • Content Description: The focus of this report is the evaluation of various co-precipitation processes for use in the synthesis of mixed oxide feedstock powders for the Ceramic Fuels Technology Area within the Fuels Cycle R&D (FCR&D) Program's Advanced Fuels Campaign. The evaluation will include a comparison with standard mechanical mixing of dry powders and as well as other co-conversion methods. The end result will be the down selection of a preferred sequence of co-precipitation process for the preparation of nuclear fuel feedstock materials to be used for comparison with other feedstock preparation methods. A review of the literature was done to identify potential nitrate-to-oxide co-conversion processes which have been applied to mixtures of uranium and plutonium to achieve recycle fuel homogeneity. Recent studies have begun to study the options for co-converting all of the plutonium and neptunium recovered from used nuclear fuels, together with appropriate portions of recovered uranium to produce the desired mixed oxide recycle fuel. The addition of recycled uranium will help reduce the safeguard attractiveness level and improve proliferation resistance of the recycled fuel. The inclusion of neptunium is primarily driven by its chemical similarity to plutonium, thus enabling a simple quick path to recycle. For recycle fuel to thermal-spectrum light water reactors (LWRs), the uranium concentration can be {approx}90% (wt.), and for fast spectrum reactors, the uranium concentration can typically exceed 70% (wt.). However, some of the co-conversion/recycle fuel fabrication processes being developed utilize a two-step process to reach the desired uranium concentration. In these processes, a 50-50 'master-mix' MOX powder is produced by the co-conversion process, and the uranium concentration is adjusted to the desired level for MOX fuel recycle by powder blending (milling) the 'master-mix' with depleted uranium oxide. In general, parameters that must be controlled for co-precipitation processes include (1) feed solution concentration adjustment, (2) precipitant concentration and addition methods, (3) pH, temperature, mixing method and time, (4) valence adjustment, (5) solid precipitate separation from the filtrate 'mother liquor,' generally by means of centrifugation or filtration, and (6) temperatures and times for drying, calcination, and reduction of the MOX product powder. Also a recovery step is necessary because of low, but finite solubility of the U/TRU metals in the mother liquor. The recovery step usually involves destruction of the residual precipitant and disposal of by-product wastes. Direct denitrations of U/TRU require fewer steps, but must utilize various methods to enable production of MOX with product characteristics that are acceptable for recycle fuel fabrication. The three co-precipitation processes considered for evaluation are (1) the ammonia co-precipitation process being developed in Russia, (2) the oxalate co-precipitation process, being developed in France, and (3) the ammonium-uranyl-plutonyl-carbonate (AUPuC) process being developed in Germany. Two direct denitration processes are presented for comparison: (1) the 'Microwave Heating (MH)' automated multi-batch process developed in Japan and (2) the 'Modified Direct Denitration (MDD)' continuous process being developed in the USA. Brief comparative descriptions of the U/TRU co-conversion processes are described. More complete details are provided in the references.

Subject

  • Keyword: By-Products
  • Keyword: Calcination
  • Keyword: Centrifugation
  • Keyword: Filtration
  • Keyword: Neptunium
  • Keyword: Oxides
  • Keyword: Synthesis
  • Keyword: Milling
  • Keyword: Depleted Uranium
  • Keyword: Oxalates
  • Keyword: Wastes
  • Keyword: Uranium
  • Keyword: Mixed Oxide Fuels
  • STI Subject Categories: 11 Nuclear Fuel Cycle And Fuel Materials
  • Keyword: Nuclear Fuels
  • Keyword: Denitration
  • Keyword: Mixtures
  • Keyword: Fabrication
  • Keyword: Proliferation
  • Keyword: Plutonium
  • Keyword: Ceramics

Collection

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

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Report

Format

  • Text

Identifier

  • Report No.: ORNL/TM-2011/164
  • Grant Number: DE-AC05-00OR22725
  • DOI: 10.2172/1024695
  • Office of Scientific & Technical Information Report Number: 1024695
  • Archival Resource Key: ark:/67531/metadc829973