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Title: Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

Here, an investigation of molybdate melts containing sodium molybdate (Na 2MoO 4) and molybdenum trioxide (MoO 3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate the feasibility of UO 2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D > 10). A melt consisting of 20 wt% MoO 3-50 wt% Na 2MoO 4-30 wt% UO 2 heated to 1313 K and cooled to 1123 K for the physical separation ofmore » the UO 2 product from the melt, and washed once with Na 2MoO 4 displays optimum conditions for separation of the UO 2 from the fission products.« less
Authors:
 [1] ; ORCiD logo [2] ;  [3] ;  [2] ;  [2]
  1. Oregon State Univ., Corvallis, OR (United States); Argonne National Lab. (ANL), Argonne, IL (United States)
  2. Argonne National Lab. (ANL), Argonne, IL (United States)
  3. Oregon State Univ., Corvallis, OR (United States)
Publication Date:
Grant/Contract Number:
AC02-06CH11357
Type:
Accepted Manuscript
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 486; Journal Issue: C; Journal ID: ISSN 0022-3115
Publisher:
Elsevier
Research Org:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE), Nuclear Reactor Technologies (NE-7)
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS; modified open fuel cycle; recrystallization; uranium behavior in molybdate melts; uranium separation; uysed nuclear fuel recycle
OSTI Identifier:
1363800
Alternate Identifier(s):
OSTI ID: 1419390