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

• 발행기관: 한국방사성폐기물학회
• 저자명: Sohee Cha Kwangheon Park
• 간행물 정보: 『Journal of Nuclear Fuel Cycle and Waste Technology (JNFCWT)』Vol.22 No.2, 129~137쪽, 전체 9쪽
• 주제분류: 공학 > 공학일반
• 파일형태: PDF
• 발행일자: 2024.06.30

국문 초록

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.

키워드

Spent nuclear fuel Burnup MCNP CZT Reaction rate