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

Journal Article · Tue Jan 17 00:00:00 EST 2017 · Journal of Nuclear Materials

DOI:https://doi.org/10.1016/j.jnucmat.2017.01.027· OSTI ID:1363800

Hames, Amber L. ^[1]; Tkac, Peter  ^[2]; Paulenova, Alena ^[3]; Willit, James L. ^[2]; Williamson, Mark A. ^[2]

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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)

Here, an investigation of molybdate melts containing sodium molybdate (Na₂MoO₄) and molybdenum trioxide (MoO₃) 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 UO₂ 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% MoO₃-50 wt% Na₂MoO₄-30 wt% UO₂ heated to 1313 K and cooled to 1123 K for the physical separation of the UO₂ product from the melt, and washed once with Na₂MoO₄ displays optimum conditions for separation of the UO₂ from the fission products.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE Office of Nuclear Energy (NE); USDOE Office of Nuclear Energy (NE), Reactor Fleet and Advanced Reactor Development. Office of Nuclear Energy Technologies
• Grant/Contract Number: AC02-06CH11357
• OSTI ID: 1363800
• Alternate ID(s): OSTI ID: 1419390
• Journal Information: Journal of Nuclear Materials, Vol. 486, Issue C; ISSN 0022-3115
• Publisher: Elsevier
• Country of Publication: United States
• Language: English

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
modified open fuel cycle
recrystallization
uranium behavior in molybdate melts
uranium separation
uysed nuclear fuel recycle