Gas cooled fast reactors are one of the Generation 4 nuclear power plants with hard neutron spectrum and high conversion ratio. In the study a long life Supercritical CO2 (S-CO2) cooled fast reactor core design with 300 MWth is presented. Physical calculation was carried out based on Dragon and CITATION, and thermal hydraulic analysis was performed based on the single channel code. The MOX fuel was utilized in the core design, and the tube-in-duct (TID) assembly was chosen for its excellent characteristics. According to the physical and thermal hydraulic coupling calculation, the reactor in the study can be operated with 300MWth for 20Ys without shuffling or refueling. Through the core life power peaking was kept relatively low, and the fuel temperature was kept below the 1800 degree centigrade.
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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
Preliminary Core Design of the Long Life S-CO2 Cooled Fast Reactor With 300WMth
Baolin Liu,
Baolin Liu
Xi’an Jiaotong University, Xi’an, China
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Hongchun Wu,
Hongchun Wu
Xi’an Jiaotong University, Xi’an, China
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Youqi Zheng,
Youqi Zheng
Xi’an Jiaotong University, Xi’an, China
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Liangzhi Cao,
Liangzhi Cao
Xi’an Jiaotong University, Xi’an, China
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Xianbao Yuan
Xianbao Yuan
Xi’an Jiaotong University, Xi’an, China
China Three Gorges University, Yichang, China
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Baolin Liu
Xi’an Jiaotong University, Xi’an, China
Hongchun Wu
Xi’an Jiaotong University, Xi’an, China
Youqi Zheng
Xi’an Jiaotong University, Xi’an, China
Liangzhi Cao
Xi’an Jiaotong University, Xi’an, China
Xianbao Yuan
Xi’an Jiaotong University, Xi’an, China
China Three Gorges University, Yichang, China
Paper No:
ICONE22-30390, V003T05A014; 6 pages
Published Online:
November 17, 2014
Citation
Liu, B, Wu, H, Zheng, Y, Cao, L, & Yuan, X. "Preliminary Core Design of the Long Life S-CO2 Cooled Fast Reactor With 300WMth." 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. V003T05A014. ASME. https://doi.org/10.1115/ICONE22-30390
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