This paper addresses the numerical simulation of two-phase flow heat transfer among the tube bundles with tube support plate (TSP) of an integral type pressurized water reactor steam generator using RPI wall boiling model. The subcooled nucleate boiling phenomenon and the coupled heat transfer between the SG primary side and secondary side were obtained. Also, the effects of tube support plate (TSP) and the different inlet subcooling on the thermal-hydraulic characteristics of SG were studied. From the results of the present numerical simulation, it reasonably revealed the subcooled flow boiling occurred in the SG secondary side and the distributions of key parameters around TSP, elucidating that this model can provide useful information to the design of the steam generator.
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2014 22nd International Conference on Nuclear Engineering
July 7–11, 2014
Prague, Czech Republic
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4594-3
PROCEEDINGS PAPER
CFD Analysis of Subcooled Wall Boiling at Shell Side of Steam Generator With TSP
Chenglong Wang,
Chenglong Wang
Xi’an Jiaotong University, Xi’an, China
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Dalin Zhang,
Dalin Zhang
Xi’an Jiaotong University, Xi’an, China
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Suizheng Qiu,
Suizheng Qiu
Xi’an Jiaotong University, Xi’an, China
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Wenxi Tian,
Wenxi Tian
Xi’an Jiaotong University, Xi’an, China
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Guanghui Su
Guanghui Su
Xi’an Jiaotong University, Xi’an, China
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Chenglong Wang
Xi’an Jiaotong University, Xi’an, China
Dalin Zhang
Xi’an Jiaotong University, Xi’an, China
Suizheng Qiu
Xi’an Jiaotong University, Xi’an, China
Wenxi Tian
Xi’an Jiaotong University, Xi’an, China
Guanghui Su
Xi’an Jiaotong University, Xi’an, China
Paper No:
ICONE22-30334, V004T10A011; 9 pages
Published Online:
November 17, 2014
Citation
Wang, C, Zhang, D, Qiu, S, Tian, W, & Su, G. "CFD Analysis of Subcooled Wall Boiling at Shell Side of Steam Generator With TSP." Proceedings of the 2014 22nd International Conference on Nuclear Engineering. Volume 4: Radiation Protection and Nuclear Technology Applications; Fuel Cycle, Radioactive Waste Management and Decommissioning; Computational Fluid Dynamics (CFD) and Coupled Codes; Reactor Physics and Transport Theory. Prague, Czech Republic. July 7–11, 2014. V004T10A011. ASME. https://doi.org/10.1115/ICONE22-30334
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