Burnup Measurement of Spent Fuel Assembly by CZT-based Gamma-Ray Spectroscopy for Input Nuclear Material Accountability of Pyroprocessing

Input nuclear material accountability is crucial for a pyroprocessing facility safeguards. Until a direct Pu measurement technique is established, an indirect method based on code calculations with burnup measurement and neutron counting for ²⁴⁴Cm could be a practical option. Burnup can be determined by destructive analysis (DA) for final dispositive accuracy or by nondestructive assay (NDA) for near-real time accountancy. In the present study, an underwater burnup measurement system based on gamma-ray spectroscopy with the CZT detector was developed and tested on a spent fuel assembly. Burnup was determined according to the ¹³⁴Cs/¹³⁷Cs activity ratio with efficiency correction by Geant4 Monte Carlo simulations. The activity ratio as a function of burnup was obtained by ORIGEN calculations. The measured burnup error was 8.6%, which was within the measurement uncertainty. It is expected that the underwater burnup measurement system could fulfill an important role as a means of near-real time accountancy at a future pyroprocessing facility.