Burnup studies of spent fuels of varying types and enrichment

This paper describes the results of fuel burnup measurements, made over a period of several years on discharged fuel from nuclear power plant and research reactor. The measured and calculated burnup of different spent fuel types, viz.: Natural uranium CANDU fuel bundles; 10.5% enriched booster rods; 20% enriched MTR fuel elements have been presented. High-resolution gamma spectrometry, using 137Cs and 134Cs burnup monitors was employed in different reactors to estimate the amount of 235U depletion in the respective fuel. The experimental data was compared with those of calculations to optimize fuel-scheduling programme. The burnup data is useful for assessment of fuel performance in the core and resolving design issues related to long-term storage facilities. It has been observed that the gamma spectrometry is very effective in identifying exact position of individual booster bundles inside the discharged booster assemblies, which is useful in safeguard applications. It is concluded that the distribution of measured isotopic activity ratios of 134Cs/137Cs along the height of the spent fuel gives accurate estimate of the axial neutron flux profiles in the core. The activity ratio technique therefore provides a useful method to determine flux peaking factors at the particular locations of the fuel assemblies in the reactor.