Parallelization and application of SACOS for whole core thermal-hydraulic analysis

Minyang Gui; Wenxi Tian; Di Wu; Ronghua Chen; Mingjun Wang; G.H. Su

doi:https://doi.org/10.1016/j.net.2021.06.029

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자료유형: 학술저널

저자정보: Minyang Gui (Xi'an Jiaotong University) Wenxi Tian (Xi'an Jiaotong University) Di Wu (Xi'an Jiaotong University) Ronghua Chen (Xi'an Jiaotong University) Mingjun Wang (Xi'an Jiaotong University) G.H. Su (Xi'an Jiaotong University)

저널정보: 한국원자력학회 Nuclear Engineering and Technology Nuclear Engineering and Technology 제53권 제12호

발행연도: 2021.12

수록면: 3,902 - 3,909 (8page)

DOI: https://doi.org/10.1016/j.net.2021.06.029

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SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and arebeing used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole corepin-level analysis, the input preprocessing and parallel capabilities of the code have been developed inthis study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less errorprone input; parallelization is established based on the domain decomposition method with the hybridof MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which candetermine the appropriate task division of the core domain according to the number of processors of theserver. By performing the calculation time evaluation for the several PWR assembly problems, the codeparallelization has been successfully verified with different number of processors. Subsequent analysisresults for rectangular- and hexagonal-assembly core imply that the code can be used to model andperform pin-level core safety analysis with acceptable computational efficiency

#SACOS #Pin-level thermal-hydraulic analysis #Input preprocessing #Parallelization whole Core

참고문헌 (22)

참고문헌 신청

C.W. Stewart / 1989 / VIPRE-01: a Thermal-Hydraulic Code for Reactor Cores

D. Basile / 1999 / COBRA-EN : an upgraded version of the COBRA-3C/MIT code for thermal hydraulic transient analysis of light water reactor fuel assemblies and cores

Y. J. Yoo / 1999 / Development of a subchannel analysis code MATRA applicable to PWRs and ALWRs / J. Korean Nucl. Soc 31(3):314~327

Y. F. Rao / 2014 / ASSERT-PV 3. 2 : advanced subchannel thermalhydraulics code for CANDU fuel bundles / Nucl. Eng. Des 275:69~79

C. Chauliac / 2011 / NURESIMeA European simulation platform for nuclear reactor safety: multi-scale and multi-physics calculations, sensitivity and uncertainty analysis / Nucl. Eng. Des 241(9):3416~3426

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