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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2020.9
- 수록면
- 1,907 - 1,916 (10page)
- DOI
- 10.1016/j.net.2020.02.017
이용수
초록· 키워드
오류제보하기
In this work, the depletion capability implemented in Monte Carlo code MCS is investigated to predict the isotopic compositions of spent nuclear fuel (SNF). By comparison of MCS calculation results to post irradiation examination (PIE) data obtained from one pressurized water reactor (PWR), the validation of this capability is conducted. The depletion analysis is performed with the ENDF/B-VII.1 library and a fuel assembly model. The transmutation equation is solved by the Chebyshev Rational Approximation Method (CRAM) with a depletion chain of 3820 isotopes. 18 actinides and 19 ?ssion products are analyzed in 14 SNF samples. The effect of statistical uncertainties on the calculated number densities is discussed. On average, most of the actinides and ?ssion products analyzed are predicted within ±6% of the experiment. MCS depletion results are also compared to other depletion codes based on publicly reported information in literature. The code-to-code analysis shows comparable accuracy. Overall, it is demonstrated that the depletion capability in MCS can be reliably applied in the prediction of SNF isotopic inventory
목차
등록된 정보가 없습니다.
참고문헌 (37)
참고문헌 신청
L. M. Fensin / 2010 / The enhancements and testing for the MCNPX 2. 6. 0 depletion capability / Nucl. Technol 170:68
G. G. Davidson / 2018 / Nuclide depletion capabilities in the SHIFT Monte Carlo code / Ann. Nucl. Energy 114:259
심형진 / 2012 / MCCARD: MONTE CARLO CODE FOR ADVANCED REACTOR DESIGN AND ANALYSIS / Nuclear Engineering and Technology 44(2):161~176
J. Leppanen / 2015 / development and applications in 2013 / Ann. Nucl. Energy 82:142
K. Okumura / 2000 / Validation of a continuousenergy Monte Carlo code MVP-BURN and its application to analysis of post irradiation experiment / J. Nucl. Sci. Technol 37(2):128
G. G. Davidson / 2018 / Nuclide depletion capabilities in the SHIFT Monte Carlo code / Ann. Nucl. Energy 114:259
심형진 / 2012 / MCCARD: MONTE CARLO CODE FOR ADVANCED REACTOR DESIGN AND ANALYSIS / Nuclear Engineering and Technology 44(2):161~176
J. Leppanen / 2015 / development and applications in 2013 / Ann. Nucl. Energy 82:142
K. Okumura / 2000 / Validation of a continuousenergy Monte Carlo code MVP-BURN and its application to analysis of post irradiation experiment / J. Nucl. Sci. Technol 37(2):128
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