Nuclear fuel recycling methods

Uranium oxide [1344-57-6] from mills is converted into uranium hexafluoride [7783-81-5] FJF, for use in gaseous diffusion isotope separation plants (see Diffusion separation methods). The wastes from these operations are only slightly radioactive. Both uranium-235 and uranium-238 have long half-Hves, 7.08 x 10 and 4.46 x 10 yr, respectively. Uranium enriched to around 3 wt % is shipped to a reactor fuel fabrication plant (see Nuclear REACTORS, NUCLEAR FUEL reserves). There conversion to uranium dioxide is foUowed by peUet formation, sintering, and placement in tubes to form fuel rods. The rods are put in bundles to form fuel assembHes. Despite active recycling (qv), some low activity wastes are produced. 

It is apparent from the foregoing discussion that both ILs and supercritical carbon dioxide do indeed offer promise as alternative solvents in the reprocessing of spent nuclear fuel and the treatment of nuclear wastes. It is equally apparent, however, that considerable additional work lies ahead before this promise can be fully realized. Of particular importance in this context is the need for an improved understanding of the fundamental aspects of metal ion transfer into ILs, for a thorough evaluation of the desirability of extractant functionalization of ILs, and for the development of new methods for both the recovery of extracted ions (e.g., uranium) and the recycling of extractants in supercritical C02-based systems. Only after such issues have been addressed might these unique solvents reasonably be expected to provide the basis of improved approaches to An or FP separations. 

Many hydrometallurgical processes or process steps are used to upgrade concentrates, process recycled scrap metal, or purify aqueous process steams. Examples are (I) the leaching of molybdenite concentrate to remove Knpurities ,ft (2) leaching of tungsten carbide and molybdenum scrap-, (3) removal of copper impurities in nickel anolyte by cementation on metallic ruckel and (4) various methods for treating nuclear fuel elements. 

The simplified PUREX method, in which purification processes for Pu and U are eliminated, could be considered. Centralized reprocessing is assumed. The spent nuclear fuel (SNF) will be reprocessed and only HLW will be discarded minor actinides (MAs) could be recycled, if MA recovery and MA fuel fabrication processes are established. 

Accurate prediction of criticality is essential not only for the quality design of plutonium recycle cores tmt also for the safety of handling fdutonium assemblies. To assess the ffllL capability in this regard, multigroup cross sections based on the latest nuclear data file, ENDF/B-IV, were used in conjunction with the lattice analysis code, HAMMER, and the two-dimensional (2-D) mnltigroup discrete ordinate code, TW0TRAN, to analyze a series of mixed-oxide-fneled critical experiments. These critical experiments which covered a wide range of water-to-fiiel ratios and Pu compositions represent a set of excellent benchmarks for validation of both the analytical methods and nuclear data for criticality prediction of mixed-oxide fuel assemblies typical of pin tonium recycle cores in LWRs. 

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