Abstract
Using the fresh FBR mixed (U,Pu) dioxide, monocarbide and mononitride fuel pieces (pellets), the dissolution (HNO3), solution digestion (storage) and extractive plutonium recovery (TBP) operations were studied in high radiation conditions.lt was established, that well known nitric acid procedure of mixed U-Pu dioxide dissolution is applicable to advanced breeder fuels.The extractive procedure of the “classic” PUREX-technology (out of Pu(VI) build-up, extraction kinetics and equilibrium view-points) can be used for quantitative actinoids recovery from above mentioned fuel solutions.The possible problems as regards spent advanced fuel reprocessing are discussed as well.
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