A Monte Carlo Simulation Study for Designing Collimators for a CZT-based Spent Nuclear Fuel Characterization System

As an attempt to develop a spent nuclear fuel characterization system that can acquire both high-resolution gamma-ray energy spectrum and neutron flux simultaneously at multiple positions around the spent nuclear fuel assembly and provide more information compared to conventional equipment, simulation study was conducted to design a collimator around a gamma-ray detector to obtain high-resolution gamma-ray spectra in a high-flux gamma-ray environment. The ORIGEN depletion code was used to evaluate the source terms of spent nuclear fuel, and MCNP 6.2, the Monte Carlo simulation code, was used to simulate the detector response. Considering the realistic operating conditions of Korea's pressurized light water reactor, the source term for the ACE7 17x17 nuclear fuel assembly was derived under the conditions of initial enrichment ranging from 3 to 5%, burnup ranging from 30 to 60 GWd/MTU, and maximum cooling time of 30 years. MCNP 6.2, the Monte Carlo simulation code, was used to design a spent nuclear fuel characterization system using a CZT detector as a gamma-ray detector to obtain a high-resolution gamma-ray energy spectrum. In particular, the research focused on designing collimators to be used in high-flux radiation environments, such as spent nuclear fuel inspection, with CZT detectors. A simulation model of the CZT detector to be installed internally was made and the validity of the detector simulation model was confirmed by comparing simulation results with the measurement experiment results. Finally, the response of the CZT detector was calculated by changing various setting parameters of the collimator. By comparing the calculated detector response with the design standard count rate, the design of a concentrator that can utilize the CZT detector in a spent nuclear fuel inspection environment was derived.