Journal of Nuclear Fuel Cycle and Waste Technology

To non-destructively determine the burnup of a spent nuclear fuel assembly, it is essential to analyze the nuclear isotopes present in the assembly and detect the neutrons and gamma rays emitted from these isotopes. Specifically, gamma-ray measurement methods can utilize a single radiation measurement value of ¹³⁷Cs or measure based on the energy peak ratio of Cs isotopes such as ¹³⁴Cs/¹³⁷Cs and ¹⁵⁴Eu/¹³⁷Cs. In this study, we validated the extent to which the results of gamma-ray measurements using cadmium zinc telluride (CZT) sensors based on ¹³⁷Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we propose equations that represent radiation behavior that reaches the detector by assuming “Direct hit” and “Penetration with attenuation” situations. The results obtained from MCNP confirmed an increase in measurement efficiency by 0.47 times when using the CZT detector, demonstrating the efficacy of the measurement system.

Keywords

Spent nuclear fuel, Burnup, MCNP, CZT, Reaction rate