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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2018.1
- 수록면
- 340 - 355 (16page)
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The methods and performance of a fast reactor multigroup cross section (XS) generation code EXUS-F aredescribed that is capable of directly processing Evaluated Nuclear Data File format nuclear data files. RECONR of NJOY is used to generate pointwise XS data, and Doppler broadening is incorporated by theGausseHermite quadrature method. The self-shielding effect is incorporated in the ultrafine group XSs inthe resolved and unresolved resonance ranges. Functions to generate scattering transfer matrices andfission spectrum matrices are realized. The extended transport approximation is used in zerodimensionalcalculations, whereas the collision probability method and the method of characteristicsare used for one-dimensional cylindrical geometry and two-dimensional hexagonal geometry problems,respectively. Verification calculations are performed first for various homogeneous mixtures and cylindricalproblems. It is confirmed that the spectrum calculations and the corresponding multigroup XSgenerations are performed adequately in that the reactivity errors are less than 50 pcmwith the McCARDMonte Carlo solutions. The nTRACER core calculations are performed with the EXUS-Fegenerated 47group XSs for the two-dimensional Advanced Burner Reactor 1000 benchmark problem. The reactivityerror of 160 pcm and the root mean square error of the pin powers of 0.7% indicate that EXUF-F generatesproperly the broad-group XSs.
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참고문헌 (23)
참고문헌 신청
유재운 / 2016 / Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea / Nuclear Engineering and Technology 48(5):1059~1070
Yeon Sang Jung / 2013 / Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers / Ann. Nuclear Energy 62:357~374
W. S. YANG / 2012 / FAST REACTOR PHYSICS AND COMPUTATIONAL METHODS / Nuclear Engineering and Technology 44(2):177~198
C.S. Gil / 2009 / ZZ KAFAX-E70, 150 and 12 groups cross section library in MATXS format based on ENDF/B-VII.0 for fast reactors, NEA-1815, OECD/NEA Data Bank, Issy-les-Moulineaux
B.J. Toppel / 1978 / $ETOE-2/MC^{2}-2/SDX$ multi-group cross-section processing / Conf-780334-5, Proc. of RSIC Seminarworkshop on Multi-group Cross Sections
Yeon Sang Jung / 2013 / Practical numerical reactor employing direct whole core neutron transport and subchannel thermal/hydraulic solvers / Ann. Nuclear Energy 62:357~374
W. S. YANG / 2012 / FAST REACTOR PHYSICS AND COMPUTATIONAL METHODS / Nuclear Engineering and Technology 44(2):177~198
C.S. Gil / 2009 / ZZ KAFAX-E70, 150 and 12 groups cross section library in MATXS format based on ENDF/B-VII.0 for fast reactors, NEA-1815, OECD/NEA Data Bank, Issy-les-Moulineaux
B.J. Toppel / 1978 /
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