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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2020.12
- 수록면
- 2,725 - 2,732 (8page)
- DOI
- https://doi.org/10.1016/j.net.2020.05.027
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초록· 키워드
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This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurposeresearch reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlocode and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as acoolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems arederived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS.
The calculated results, such as the neutron effective multiplication factor (k) value and the control rodworth are compared with the experimental data. Moreover, additional calculated results, including theneutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, andkinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to beslightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction crosssections of 16O, 9Be, 235U, 238U, and S(a,b) of 1H in H2O from ENDF/B-VIII.0 have substantial updates;hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes inRSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, andfission rate distribution are similar and consistent among the nuclear data libraries
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참고문헌 (20)
참고문헌 신청
D. A. Brown / 2018 / ENDF/B-VIII.0: the 8th major release of the nuclear reaction data library with CIELO-project cross sections, new standards and thermal scattering data / Nucl. Data Sheets 148:1~42
M. B. Chadwick / 2011 / ENDF/B-VII.1 nuclear data for science and technology:cross sections, covariances, fission product yields and decay data / Nucl. Data Sheets 112(12):2887~2996
Kabach Ouadie / 2017 / Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016 / Nuclear Engineering and Technology 49(8):1610~1616
O. Kabach / 2019 / Processing of JEFF-3. 3 and ENDF/B-VIII. 0 and testing with critical benchmark experiments and TRIGA mark II research reactor using MCNPX / Appl. Radiat. Isot 150:146~156
H. J. Park / 2019 / Comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 in criticality, depletion benchmark, and uncertainty analyses by McCARD / Ann. Nucl. Energy 131:443~459
M. B. Chadwick / 2011 / ENDF/B-VII.1 nuclear data for science and technology:cross sections, covariances, fission product yields and decay data / Nucl. Data Sheets 112(12):2887~2996
Kabach Ouadie / 2017 / Processing and benchmarking of evaluated nuclear data file/b-viii.0β4 cross-section library by analysis of a series of critical experimental benchmark using the monte carlo code MCNP(X) and NJOY2016 / Nuclear Engineering and Technology 49(8):1610~1616
O. Kabach / 2019 / Processing of JEFF-3. 3 and ENDF/B-VIII. 0 and testing with critical benchmark experiments and TRIGA mark II research reactor using MCNPX / Appl. Radiat. Isot 150:146~156
H. J. Park / 2019 / Comparison of ENDF/B-VIII.0 and ENDF/B-VII.1 in criticality, depletion benchmark, and uncertainty analyses by McCARD / Ann. Nucl. Energy 131:443~459
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