In the fuel fabrication process for light water nuclear power reactors, enriched uranium hexafluoride (UF6) is converted to UO2 as illustrated in Fig. 14. 7. Disposable solid waste generated by the process is reduced by a factor of 1/25 through a carefully controlled incineration process. The incin­erator ash contains approximately 10% by weight of enriched uranium (about 3.5% 235U). Some gadolinium (Gd) which is added to the fuel as a neutron absorber is also present in the waste ash. In 2003, AREVA NP started looking into the possibility of using sc-CO2 as a medium for recover­ing enriched uranium from the incinerator ash waste generated by the fuel fabrication process in collaboration with the University of Idaho. With the aid of the TBP-HNO3 complex described in the previous section, enriched uranium and gadolinium in the incinerator ash can be effectively extracted

Incoming UF6 cylinders

14.7 Illustration of a typical light water reactor fuel fabrication facility (Source: US Nuclear Regulatory Commission. "Light Water Reactor Low-Enriched Uranium Fuel," Fuel Fabrication. ONLINE. 2009. Nuclear Regulatory Commission. Available: http://www. nrc. gov/materials/fuel- cycle-fac/fuel-fab. html [5 Aug. 2009].

able commodity (enriched uranium) from a material that previously was considered waste.

Another industrial demonstration of the sc-CO2 extraction technology was conducted in Japan by Mitsubishi Heavy Industries, Japan Nuclear Cycle Corp. and Nagoya University testing the feasibility of reprocessing spent nuclear fuel using this green solvent (Shimada et al. 2002). The process is called “Super-DIREX” process which stands for supercritical fluid for direction extraction. The main purpose of this project is to evaluate the fea­sibility of direct dissolution of uranium and plutonium in irradiated uranium oxides and in spent fuel by sc-CO2 containing the TBP(HNO3)18(TBP)06 complex. After dissolution, uranium and plutonium are stripped from the sc-CO2 phase with water. Recovery of UO2 from the strip solution follows the conventional PUREX process. The details of the Super-DIREX project are not fully known. The nitric acid concentration in the acid droplets sur­rounded by TBP (like reverse micelles) is estimated to be greater than 12 mol/L (Sawada et al. 2005). A recent report by Sawada et al. (2009) described the distribution coefficients of U(VI) and some simulated fission products measured between aqueous phase of high nitrate concentrations (4-15 mol/L) and dodecane with the purpose of understanding the extrac­tion behavior of uranium and fission products in the Super-DIREX process. Reprocessing spent nuclear fuel in sc-CO2 is obviously a complicated process and requires many efforts in chemical and engineering studies to evaluate its feasibility. Nevertheless, the Super-DIREX project should provide some valuable information regarding the chemistry of uranium, plutonium and fission products in the sc-CO2 dissolution process and safe handling of highly radioactive materials in a pressurized system.