A Comparative Study on Gamma-ray Measurement and MCNP Simulation for Precise Measurement of Spent Nuclear Fuel Burnup

Sohee Cha, Kwangheon Park^*

Kyung Hee University, 1732, Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 17104, Republic of Korea

Received 2024-01-12 ; Revised 2024-01-25 ; Approved 2024-04-18

Abstract

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 results from gamma-ray measurement experiments utilizing CZT sensors based on 137Cs could be accurately simulated by implementing identical conditions on MCNP. To simulate measurement scenarios using a lead collimator, we proposed equations representing the radiation behavior reaching the detector by assuming ‘Direct hit’ and ‘Penetration with attenuation situations. The results obtained from MCNP confirmed a measurement efficiency of 0.47 times when using the CZT detector, showcasing the efficacy of the measurement system.

Key Words : Spent Nuclear fuel, Burnup, MCNP, CZT, Reaction rate