The China EU cooperation project SCWR-FQT small experimental reactor is selected as the research object. The new SCWR-FQT calculation code is made. Thermal performance of four channels are analyzed. The results show that: During normal condition, the coolant temperature increases gradually and pressure reduces gradually. The highest temperature of the coolant does not exceed 450°C working limit, meets the design safety requirement. When transient of entrance temperature occurs, lateral two channels are affected significantly. The inside channel has maximum temperature stability because of the minimum affection. The increasing of the entrance temperature will shorten the time of the relative inner two channels to reach stable. When transient of entrance flow occurs, the inside channel has the maximum affection. The decreasing of the entrance flow will increase the time of the inside channel to reach stable. The entrance flow should not lower than 49.2% of the normal operating mode.
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2014 22nd International Conference on Nuclear Engineering
July 7–11, 2014
Prague, Czech Republic
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4595-0
PROCEEDINGS PAPER
Thermal-Hydraulic Analysis for Experimental Device of Supercritical Water Reactor-Fuel Qualification Test
Liang Liu,
Liang Liu
North China Electric Power University, Beijing, China
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Tao Zhou,
Tao Zhou
North China Electric Power University, Beijing, China
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Yu Li,
Yu Li
North China Electric Power University, Beijing, China
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Wanxu Cheng,
Wanxu Cheng
North China Electric Power University, Beijing, China
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Juan Chen,
Juan Chen
North China Electric Power University, Beijing, China
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Shanfang Huang,
Shanfang Huang
Tsinghua University, Beijing, China
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Yanping Huang
Yanping Huang
National Key Laboratory of Bubble Physics and Natural Circulation, Chengdu, Sichuan, China
Search for other works by this author on:
Liang Liu
North China Electric Power University, Beijing, China
Tao Zhou
North China Electric Power University, Beijing, China
Yu Li
North China Electric Power University, Beijing, China
Wanxu Cheng
North China Electric Power University, Beijing, China
Juan Chen
North China Electric Power University, Beijing, China
Shanfang Huang
Tsinghua University, Beijing, China
Yanping Huang
National Key Laboratory of Bubble Physics and Natural Circulation, Chengdu, Sichuan, China
Paper No:
ICONE22-30176, V005T17A025; 11 pages
Published Online:
November 17, 2014
Citation
Liu, L, Zhou, T, Li, Y, Cheng, W, Chen, J, Huang, S, & Huang, Y. "Thermal-Hydraulic Analysis for Experimental Device of Supercritical Water Reactor-Fuel Qualification Test." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 5: Innovative Nuclear Power Plant Design and New Technology Application; Student Paper Competition. Prague, Czech Republic. July 7–11, 2014. V005T17A025. ASME. https://doi.org/10.1115/ICONE22-30176
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