2006 AIChE Annual Meeting

(526b) Separation of Uranium from Fission Products in Spent Nuclear Fuel Using Aqueous Hydrogen Peroxide-Carbonate Solutions

Authors

George S. Goff - Presenter, Los Alamos National Laboratory
Felicia L. Taw, Los Alamos National Laboratory
Shane M. Peper, Los Alamos National Laboratory
Lia F. Brodnax, Los Alamos National Laboratory
Stephanie E. Field, Los Alamos National Laboratory
Chris Wakefield, Los Alamos National Laboratory
Wolfgang H. Runde, Los Alamos National Laboratory
As an alternative to the traditional PUREX process for recycling spent nuclear fuel (SNF), a new process is being developed that utilizes hydrogen peroxide to dissolve SNF in aqueous carbonate solutions. Some of the primary benefits of the new peroxide promoted carbonate process include a smaller footprint, fewer separations steps, elimination of hazardous organic waste, and enhanced proliferation resistance by elimination of the pure plutonium stream produced in the PUREX process.

The rate of dissolution of several uranium oxides, fired at various temperatures, was studied over a range of peroxide and carbonate concentrations. Generally, dissolution rates decreased with increasing carbonate concentration and increased with increasing peroxide concentrations. Stable mixed peroxo-carbonate complexes have been identified in solution using various methods such as UV-VIS, Raman, and 13C NMR spectroscopy. U(VI) forms complexes of the form UO2(O2)x(CO3)y2-2x-2y (x/y = 1/2, and 2/1), and the monoperoxo-biscarbonato complex (UO2(O2)(CO3)24-) has been crystallized and characterized by single crystal x-ray diffraction studies. Solubilities and equilibrium constants for the various uranium species are being determined in order to accurately predict solution speciation under industrial process conditions. Studies are also underway to determine the behavior of neptunium and plutonium in aqueous peroxide-carbonate solutions.