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RISS 인기검색어
- 검색키워드
- 주제어 : Monte Carlo burnup analysis (검색결과 3 건)
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Monte Carlo burnup and its uncertainty propagation analyses for VERA depletion benchmarks by McCARD
박호진,이동혁,전병규,심형진 한국원자력학회 2018 Nuclear Engineering and Technology Vol.50 No.7
For an efficient Monte Carlo (MC) burnup analysis, an accurate high-order depletion scheme to considerthe nonlinear flux variation in a coarse burnup-step interval is crucial accompanied with an accuratedepletion equation solver. In a Seoul National University MC code, McCARD, the high-order depletionschemes of the quadratic depletion method (QDM) and the linear extrapolation/quadratic interpolation(LEQI) method and a depletion equation solver by the Chebyshev rational approximation method (CRAM)have been newly implemented in addition to the existing constant extrapolation/backward extrapolation(CEBE) method using the matrix exponential method (MEM) solver with substeps. In this paper, thequadratic extrapolation/quadratic interpolation (QEQI) method is proposed as a new high-order depletionscheme. In order to examine the effectiveness of the newly-implemented depletion modules inMcCARD, four problems in the VERA depletion benchmarks are solved by CEBE/MEM, CEBE/CRAM, LEQI/MEM, QEQI/MEM, and QDM for gadolinium isotopes. From the comparisons, it is shown that the QEQI/MEM predicts kinf's most accurately among the test cases. In addition, statistical uncertainty propagationanalyses for a VERA pin cell problem are conducted by the sensitivity and uncertainty and the stochasticsampling methods.
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윤형주,Do-Yeon Kim,박광헌,홍서기 한국원자력학회 2016 Nuclear Engineering and Technology Vol.48 No.3
Nuclear criticality safety analyses (NCSAs) considering burnup credit were performed forthe GBC-32 cask. The used nuclear fuel assemblies (UNFAs) discharged from Hanbit NuclearPower Plant Unit 3 Cycle 6 were loaded into the cask. Their axial burnup distributionsand average discharge burnups were evaluated using the DeCART and Multi-purposeAnalyzer for Static and Transient Effects of Reactors (MASTER) codes, and NCSAs wereperformed using SCALE 6.1/STandardized Analysis of Reactivity for Burnup Credit usingSCALE (STARBUCS) and Monte Carlo N-Particle transport code, version 6 (MCNP 6). Theaxial burnup distributions were determined for 20 UNFAs with various initial enrichmentsand burnups, which were applied to the criticality analysis for the cask system. The UNFAsfor 20- and 30-year cooling times were assumed to be stored in the cask. The criticalityanalyses indicated that keff values for UNFAs with nonuniform axial burnup distributionswere larger than those with a uniform distribution, that is, the end effects were positive butmuch smaller than those with the reference distribution. The axial burnup distributions for20 UNFAs had shapes that were more symmetrical with a less steep gradient in the upperregion than the reference ones of the United States Department of Energy. These differencesin the axial burnup distributions resulted in a significant reduction in end effectscompared with the reference.
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Garcia, Manuel,Vocka, Radim,Tuominen, Riku,Gommlich, Andre,Leppanen, Jaakko,Valtavirta, Ville,Imke, Uwe,Ferraro, Diego,Uffelen, Paul Van,Milisdorfer, Lukas,Sanchez-Espinoza, Victor Korean Nuclear Society 2021 Nuclear Engineering and Technology Vol.53 No.10
This work deals with the validation of a high-fidelity multiphysics system coupling the Serpent 2 Monte Carlo neutron transport code with SUBCHANFLOW, a subchannel thermalhydraulics code, and TRANSURANUS, a fuel-performance analysis code. The results for a full-core pin-by-pin burnup calculation for the ninth operating cycle of the Temelín II VVER-1000 plant, which starts from a fresh core, are presented and assessed using experimental data. A good agreement is found comparing the critical boron concentration and a set of pin-level neutron flux profiles against measurements. In addition, the calculated axial and radial power distributions match closely the values reported by the core monitoring system. To demonstrate the modeling capabilities of the three-code coupling, pin-level neutronic, thermalhydraulic and thermomechanic results are shown as well. These studies are encompassed in the final phase of the EU Horizon 2020 McSAFE project, during which the Serpent-SUBCHANFLOW-TRANSURANUS system was developed.
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