A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN_K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN_K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation.
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2014 22nd International Conference on Nuclear Engineering
July 7–11, 2014
Prague, Czech Republic
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4593-6
PROCEEDINGS PAPER
Coupled Three Dimensional Neutronics/Thermal-Hydraulics Code STTA for SCWR Core Transient Analysis
Lianjie Wang,
Lianjie Wang
Nuclear Power Institute of China, Chengdu, China
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Wenbo Zhao,
Wenbo Zhao
Nuclear Power Institute of China, Chengdu, China
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Ping Yang,
Ping Yang
Nuclear Power Institute of China, Chengdu, China
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Bingde Chen,
Bingde Chen
Nuclear Power Institute of China, Chengdu, China
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Dong Yao
Dong Yao
Nuclear Power Institute of China, Chengdu, China
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Lianjie Wang
Nuclear Power Institute of China, Chengdu, China
Wenbo Zhao
Nuclear Power Institute of China, Chengdu, China
Ping Yang
Nuclear Power Institute of China, Chengdu, China
Bingde Chen
Nuclear Power Institute of China, Chengdu, China
Dong Yao
Nuclear Power Institute of China, Chengdu, China
Paper No:
ICONE22-30078, V003T05A001; 6 pages
Published Online:
November 17, 2014
Citation
Wang, L, Zhao, W, Yang, P, Chen, B, & Yao, D. "Coupled Three Dimensional Neutronics/Thermal-Hydraulics Code STTA for SCWR Core Transient Analysis." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security. Prague, Czech Republic. July 7–11, 2014. V003T05A001. ASME. https://doi.org/10.1115/ICONE22-30078
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