Actinide oxide dissolution in tributyl phosphate

Conference · Wed Feb 05 23:00:00 EST 2025

DOI:https://doi.org/10.1051/epjconf/202531701003· OSTI ID:2511252

Gogolski, Jarrod ^[1]; Goetzman, Chelsea ^[1]; Mills, Matthew ^[1]; Nguyen, Vinh ^[1]; Rudisill, Tracy ^[1]; Lascola, Jr., Robert J. ^[1]

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1. Savannah River National Laboratory (SRNL), Aiken, SC (United States)

An alternative to dissolving used nuclear fuel (UNF) in an acidic solution during reprocessing is direct dissolution in an organic solution, which would eliminate an aqueous dissolution step, decrease the amount of nitrate needed, and reduce the facility size. The flowsheet for this potentially less expensive alternative is first to voloxidize the UNF to remove fission product gases and form an oxide. After voloxidation, the UNF is then dissolved in an organic solution containing an extractant mixed with an aliphatic diluent and pre-equilibrated with nitric acid. The organic solution then goes through a solvent extraction process to recover the uranium and/or other desired radionuclides. This work qualitatively studied the dissolution of actinide oxides (UO2, NpO2, and PuO2) in tributyl phosphate using UV-Vis-NIR absorbance spectroscopy to ascertain dissolution behavior. Initial studies included material that is otherwise difficult to dissolve in only nitric acid, specifically CeO2, that is sometimes used as a dissolution surrogate for PuO2. This work confirmed that CeO2, NpO2, and PuO2 are difficult to dissolve in 30 vol% TBP-dodecane pre-equilibrated with 10 M HNO3 and will readily dissolve when co-precipitated with U (i.e., the mixed oxides U-Ce, U-Np, and U-Pu), surrogates for voloxidized nuclear fuel.

• Research Organization: Savannah River National Laboratory (SRNL), Aiken, SC (United States)
• Sponsoring Organization: USDOE Office of Environmental Management (EM)
• DOE Contract Number: 89303321CEM000080
• OSTI ID: 2511252
• Report Number(s): SRNL-STI-2024-00496-2
• Country of Publication: United States
• Language: English