Dissolution of Zirconium-Bonded, Monolithic, Uranium-Molybdenum Fuel for Uranium Recovery

Technical Report · Sun May 31 04:00:00 EDT 2009

DOI:https://doi.org/10.2172/1132246· OSTI ID:1132246

Jerden, James ^[1]; Fortner, Jeffrey ^[1]; Stepinski, Dominique ^[1]

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1. Argonne National Laboratory (ANL), Argonne, IL (United States)

This report presents results from a process development study of dissolution scenarios that would facilitate the recovery of uranium from both irradiated (spent) and unirradiated (scrap) U-10Mo fuel using a tributyl-phosphate-based solvent extraction process. A specific focus of this study is to assess how metals added as diffusion/bonding barriers will influence the chemistry of the fuel dissolution process. Of particular interest are fuels containing zirconium or niobium as barrier materials, because these fuels may contain an intermetallic uranium-zirconium or uranium-niobium phase (formed during fabrication and enhanced during irradiation) that could react with explosive violence during oxidative dissolution. The first part of this report will evaluate the explosive potential of this intermetallic phase and discuss means to inhibit explosive reactions in the dissolver when it is present (e.g., by the addition of fluoride to the dissolver solution). The second part of the report will discuss dissolution scenarios for both spent and scrap U-10Mo fuel and will discuss the possibility of dissolver corrosion in the fluoride bearing dissolver solutions.

• Research Organization: Argonne National Laboratory (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Office of Defense Nuclear Nonproliferation
• DOE Contract Number: AC02-06CH11357
• OSTI ID: 1132246
• Report Number(s): ANL/CSE--13/30
• Country of Publication: United States
• Language: ENGLISH