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Thermal‐hydraulic analysis of wire‐wrapped rod bundle in lead‐based fast reactor with non‐uniform heat flux

2022; Wiley; Volume: 46; Issue: 12 Linguagem: Inglês

10.1002/er.8316

1099-114X

Kejian Dong, Shakeel Ahmad, Shahid Ali Khan, Peng Ding, Wenhuai Li, Jiyun Zhao,

Nuclear Engineering Thermal-Hydraulics

International Journal of Energy ResearchVolume 46, Issue 12 p. 16538-16549 RESEARCH ARTICLE Thermal-hydraulic analysis of wire-wrapped rod bundle in lead-based fast reactor with non-uniform heat flux Kejian Dong, Kejian Dong Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong KongSearch for more papers by this authorShakeel Ahmad, Shakeel Ahmad Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong KongSearch for more papers by this authorShahid Ali Khan, Shahid Ali Khan Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong KongSearch for more papers by this authorPeng Ding, Peng Ding Lead-Bismuth Fast Reactor Project Department, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, ChinaSearch for more papers by this authorWenhuai Li, Wenhuai Li Lead-Bismuth Fast Reactor Project Department, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, ChinaSearch for more papers by this authorJiyun Zhao, Corresponding Author Jiyun Zhao [email protected] orcid.org/0000-0002-9425-6036 Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong Kong Correspondence Jiyun Zhao, Department of Mechanical Engineering, City University of Hong Kong, Tat Chee Avenue, Kowloon Tong, Hong Kong. Email: [email protected]Search for more papers by this author Kejian Dong, Kejian Dong Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong KongSearch for more papers by this authorShakeel Ahmad, Shakeel Ahmad Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong KongSearch for more papers by this authorShahid Ali Khan, Shahid Ali Khan Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong KongSearch for more papers by this authorPeng Ding, Peng Ding Lead-Bismuth Fast Reactor Project Department, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, ChinaSearch for more papers by this authorWenhuai Li, Wenhuai Li Lead-Bismuth Fast Reactor Project Department, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, ChinaSearch for more papers by this authorJiyun Zhao, Corresponding Author Jiyun Zhao [email protected] orcid.org/0000-0002-9425-6036 Department of Mechanical Engineering, City University of Hong Kong, Kowloon Tong, Hong Kong Correspondence Jiyun Zhao, Department of Mechanical Engineering, City University of Hong Kong, Tat Chee Avenue, Kowloon Tong, Hong Kong. Email: [email protected]Search for more papers by this author First published: 28 June 2022 https://doi.org/10.1002/er.8316 Funding information: Guangdong Provincial Key R&D Program, Grant/Award Number: 2021B0101250002; Natural Science Foundation of Guangdong Province, China, Grant/Award Number: 2020a15110753 Read the full textAboutPDF ToolsRequest permissionExport citationAdd to favoritesTrack citation ShareShare Give accessShare full text accessShare full-text accessPlease review our Terms and Conditions of Use and check box below to share full-text version of article.I have read and accept the Wiley Online Library Terms and Conditions of UseShareable LinkUse the link below to share a full-text version of this article with your friends and colleagues. Learn more.Copy URL Share a linkShare onEmailFacebookTwitterLinkedInRedditWechat Summary As one of the most promising advanced reactors, the lead-based fast reactor has drawn great attention due to economic and safety advantages. Investigating thermal hydraulics is essential for the design of a lead-based reactor. In this paper, a CFD simulation of a 19-wire-wrapped-rod bundle with lead-bismuth eutectic (LBE) as coolant is carried out using the Reynolds-averaged Navier-Stokes equation method, and four different types of axial non-uniform heat flux are applied. Excellent validation of results for thermal and hydraulic aspects is obtained first by comparing simulation results using turbulent model k-ω SST with experimental data and high-fidelity large eddy simulation data. The mechanism of transverse flow variation in subchannels and faces resulting from changing locations of wires is studied. The strong transverse flow at the edge and corner subchannels lead to a more distinct oscillation in peripheral cladding temperature under non-uniform heat flux. The hot spot issue for blockage conditions is studied, and it is found that the temperature increment at blockage is linear to the local heat flux. When the blockages locate at the peak normalized heat flux of 1.56, there is no evident difference in the maximum temperature when changing the heat flux pattern, with an averaged Tmax of 732 K. Even though the normalized heat flux is high at 1.89, the flatter and lower temperature distribution would not lead to a remarkable high hot spot temperature compared with that of 1.56. The distribution of hot spots, as well as the inhomogeneity of heat transfer in different heat flux patterns, provide a reference for lead-based reactor design. REFERENCES 1Kelly JE. Generation IV international forum: a decade of progress through international cooperation. Prog Nucl Energy. 2014; 77: 240-246. 2Glazov A, Leonov V, Orlov V, et al. Brest reactor and plant-site nuclear fuel cycle. Atomic Energy. 2007; 103(1): 501-508. 3Maes D. Mechanical design of the small-scale experimental ADS: MYRRHA. Energy Convers Manag. 2006; 47(17): 2710-2723. 4Grasso G, Petrovich C, Mattioli D, et al. 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