Patent ID: 7749469

Claim:
A process for recovering uranium (U) alone from a spent nuclear fuel using a highly alkaline carbonate solution, the process comprising: (a) placing a spent nuclear fuel comprising U and at least one Transuranium (TRU) nuclide in a solution comprising carbonate and H 2 O 2 at a pH of about 11 to about 13 to provide a U-containing carbonate solution, wherein the solution dissolves and leaches the U without dissolving or leaching a TRU nuclide present in the spent nuclear fuel; (b) precipitating at least one of cesium (Cs), technetium (Tc), or a combination thereof from the U-containing carbonate solution by adding an organic precipitant to the U-containing carbonate solution; (c) adjusting a pH of the U-containing carbonate solution to about 2 to about 4, wherein the adjusting precipitates and separates U from the U-containing carbonate solution in the form of UO 4 , and a carbonate salt is recovered during the adjusting by flowing carbon dioxide gas generated during the adjusting into a gas absorption column in which an alkaline solution is circulated; and (d) separating an impurity nuclide partially co-dissolved with U in the U-containing carbonate solution by electrodialysis using a cation exchange membrane and an anion exchange membrane to recover the carbonate solution, wherein the impurity nuclide is selected from: molybdenum (Mo), tellurium (Te), and combinations thereof.