~(232)U CONTENT DETERMINATION IN SPENT FA FROM FAST REACTOR WITH A URANIUM LOAD

摘要

Experiments performed at the Mayak Production Assiciation to determine the U-232 content in spent BN-600 fuel are analyzed. Calculations show that to within the limits of error the concentration largely agrees with the measured value, which makes it possible to use the calculations for a preliminary assessment of the U-232 content in batches of uranium regenerated from reprocessed spent BN-600 FA.