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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2020.12
- 수록면
- 2,733 - 2,742 (10page)
- DOI
- https://doi.org/10.1016/j.net.2020.05.029
이용수
초록· 키워드
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This paper presents an assessment of applicability of the multigroup cross sections generated withMonte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section setswere generated for simple one- (1-D) and two-dimensional (2-D) sodium-cooled fast reactor problemsusing the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative tothe reference continuous-energy SERPENT results, with the transport corrected P0 scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, sinceanisotropic scattering is important in fast reactors. When the scattering order was increased to P5, the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertaintyanalysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scatteringmatrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC2-3 code and merged with the SERPENT cross sections. The mixed crosssection set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example,in the 2-D calculation with P3 scattering order, the k-eff error was reduced from 634 pcm to 223 pcm.
The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reducedfrom 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion thatoccurred in 237Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reducedconsistently, for example, from 1.1% to 0.1% for 235U, from 2.2% to 0.7% for 238Pu, and from 1.6% to 0.2% for 241Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SERPENT may not be adequate for application to fast reactors where anisotropic scattering is important
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참고문헌 (21)
참고문헌 신청
E. Fridman / 2011 / On the use of the serpent Monte Carlo code for fewgroup cross section generation / Ann. Nucl. Energy 38:1399~1405
H. J. Park / 2012 / Generation of few-group diffusion constants by Monte Carlo code McCARD / Nucl. Sci. Eng 172:66~77
J. Navarro / 2012 / MCNP and SCALE 6. 1 cross section evaluation for ATRC verification and validation applications / Trans. Am. Nucl. Soc 107:1002~1006
A. Scopatz / 2012 / Interpolations of nuclide-specific scattering kernels generated with SERPENT / Proc. of PHYSOR (2012). ANS International Topical Meeting on Advances in Reactor Physics
A. G. Nelson / 2015 / Improved Monte Carlo tallying of multi-group scattering moments using the NDPP code / Trans. Am. Nucl. Soc 113:645~648
H. J. Park / 2012 / Generation of few-group diffusion constants by Monte Carlo code McCARD / Nucl. Sci. Eng 172:66~77
J. Navarro / 2012 / MCNP and SCALE 6. 1 cross section evaluation for ATRC verification and validation applications / Trans. Am. Nucl. Soc 107:1002~1006
A. Scopatz / 2012 / Interpolations of nuclide-specific scattering kernels generated with SERPENT / Proc. of PHYSOR (2012). ANS International Topical Meeting on Advances in Reactor Physics
A. G. Nelson / 2015 / Improved Monte Carlo tallying of multi-group scattering moments using the NDPP code / Trans. Am. Nucl. Soc 113:645~648
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