Software CFX is used to build a typical natural circulation loop to study flow and heat transfer characteristics of water vapor, the vapor-liquid two-phase and supercritical water under natural circulation. During the process of natural circulation, the variation of parameters, heat transfer coefficient and mass flow is compared. It is found that when formed a natural circulation, the steam has a lower mass flow and heat transfer coefficient, while the two parameters of two-phase and supercritical water are higher. Indicates that the heat transfer capability of steam is weak, the steam cannot transfer heat out opportunely when serious accidents take place. The two-phase water is of high heat transfer coefficient. Supercritical water is of strong exchange capacity, supercritical water under natural circulation is a promising flow pattern.
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2014 22nd International Conference on Nuclear Engineering
July 7–11, 2014
Prague, Czech Republic
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4591-2
PROCEEDINGS PAPER
Study on Flow and Heat Transfer Characteristics of Different States of Water Under Natural Circulation
Zhongyun Ju,
Zhongyun Ju
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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Jingjing Li,
Jingjing Li
North China Electric Power University, Beijing, China
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Zejun Xiao
Zejun Xiao
Nuclear Reactor Thermal-Hydraulic Technology Laboratory, Chengdu, Sichuan, China
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Zhongyun Ju
North China Electric Power University, Beijing, China
Tao Zhou
North China Electric Power University, Beijing, China
Jingjing Li
North China Electric Power University, Beijing, China
Zejun Xiao
Nuclear Reactor Thermal-Hydraulic Technology Laboratory, Chengdu, Sichuan, China
Paper No:
ICONE22-31227, V02BT09A060; 6 pages
Published Online:
November 17, 2014
Citation
Ju, Z, Zhou, T, Li, J, & Xiao, Z. "Study on Flow and Heat Transfer Characteristics of Different States of Water Under Natural Circulation." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 2B: Thermal Hydraulics. Prague, Czech Republic. July 7–11, 2014. V02BT09A060. ASME. https://doi.org/10.1115/ICONE22-31227
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