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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2021.7
- 수록면
- 2,104 - 2,125 (22page)
- DOI
- https://doi.org/10.1016/j.net.2021.01.009
이용수
초록· 키워드
오류제보하기
This paper presents the verification and validation (V&V) of a calculation module for isotope inventoryprediction to control the back-end cycle of spent nuclear fuel (SNF). The calculation method presentedherein was implemented in a two-step code system of a lattice code STREAM and a nodal diffusion codeRAST-K. STREAM generates a cross section and provides the number density information using branch/history depletion branch calculations, whereas RAST-K supplies the power history and three historyindices (boron concentration, moderator temperature, and fuel temperature). As its primary feature, thismethod can directly consider three-dimensional core simulation conditions using history indices of theoperating conditions. Therefore, this method reduces the computation time by avoiding a recalculation ofthe fuel depletion. The module for isotope inventory calculates the number densities using the Lagrangeinterpolation method and power history correction factors, which are applied to correct the effects of thedecay and fission products generated at different power levels. To assess the reliability of the developedcode system for back-end cycle analysis, validation study was performed with 58 measured samples ofpressurized water reactor (PWR) SNF, and code-to-code comparison was conducted with STREAM-SNF,HELIOS-1.6 and SCALE 5.1. The V&V results presented that the developed code system can providereasonable results with comparable confidence intervals. As a result, this paper successfully demonstratesthat the isotope inventory prediction code system can be used for spent nuclear fuel analysis
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참고문헌 (27)
참고문헌 신청
B. Ebiwonjumi / 2018 / Validation of lattice physics code STREAM for predicting pressurized water reactor spent nuclear fuel isotopic inventory / Ann. Nucl. Energy 120:431~449
A. Mohammadi / 2016 / Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors / Appl. Radiat. Isot 108:129~132
N. Mai / 2020 / Extension of Monte Carlo code MCS to spent fuel cask shielding analysis / Int. J. Energy Res 44:8089~8102
J. Jang / 2018 / Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask / Ann. Nucl. Energy 114:495~509
2001 / Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit
A. Mohammadi / 2016 / Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors / Appl. Radiat. Isot 108:129~132
N. Mai / 2020 / Extension of Monte Carlo code MCS to spent fuel cask shielding analysis / Int. J. Energy Res 44:8089~8102
J. Jang / 2018 / Validation of UNIST Monte Carlo code MCS for criticality safety analysis of PWR spent fuel pool and storage cask / Ann. Nucl. Energy 114:495~509
2001 / Computational Benchmark for Estimation of Reactivity Margin from Fission Products and Minor Actinides in PWR Burnup Credit
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