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A Conceptual Process Study for Recovery of Uranium Alone from Spent Nuclear Fuel by Using High-Alkaline Carbonate Media

Kwang-Wook Kim, Dong-Yong Chung, Han-Bum Yang, Jea-Kwan Lim, Eil-Hee Lee, Kee-Chan Song, Kyuseok Song

Nuclear Technology

Volume 166 / Number 2 / May 2009 / Pages 170-179


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This work studied a conceptual process to recover uranium alone from spent nuclear fuel using high-alkaline carbonate media with hydrogen peroxide for the purposes of reducing the volume of high-level active waste and recycling of uranium from the spent fuel with greatly enhanced proliferation resistance, environmental friendliness, and operational safety. The transuranium (TRU) elements were evaluated to be undissolved and precipitated together with other fission products during the oxidative leaching of uranium from the spent fuel. The leaching ratio of uranium dioxide to TRU dioxide from spent fuel in the carbonate solution with H2O2 was estimated to be more than about 108. Only Cs, Tc, Mo, and Te among the major fission products in the spent fuel were dissolved together in the carbonate solution. In the carbonate solution with H2O2, UO2 was dissolved in the form of uranyl peroxo-carbonato complex ions, which could be recovered in the form of uranium peroxide precipitate with a very low solubility by acidification of the solution in a succeeding step. All the inorganic salts of Na2CO3, NaOH, and HNO3 used in the process suggested could be almost completely recovered and recycled into the process again without any generation of secondary wastes.

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