Numerical investigation of the critical heat flux in a 5 × 5 rod bundle with multi-grid

Wei Liu; Zemin Shang; Shihao Yang; Lixin Yang; Zihao Tian; Yu Liu; Xi Chen; Qian Peng

Nuclear Engineering and Technology (May 2022)

Numerical investigation of the critical heat flux in a 5 × 5 rod bundle with multi-grid

Wei Liu,
Zemin Shang,
Shihao Yang,
Lixin Yang,
Zihao Tian,
Yu Liu,
Xi Chen,
Qian Peng

Affiliations

Wei Liu: State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, 610213, China
Zemin Shang: School of Mechanical, Electronic and Control Engineering, Beijing Jiaotong University, Beijing, 100044, China
Shihao Yang: School of Mechanical, Electronic and Control Engineering, Beijing Jiaotong University, Beijing, 100044, China
Lixin Yang: School of Mechanical, Electronic and Control Engineering, Beijing Jiaotong University, Beijing, 100044, China; Corresponding author.
Zihao Tian: School of Mechanical, Electronic and Control Engineering, Beijing Jiaotong University, Beijing, 100044, China
Yu Liu: State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, 610213, China; Corresponding author.
Xi Chen: State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, 610213, China
Qian Peng: State Key Laboratory of Reactor System Design Technology, Nuclear Power Institute of China, Chengdu, 610213, China

Journal volume & issue: Vol. 54, no. 5
pp. 1914 – 1928

Abstract

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To improve the heat transfer efficiency of the reactor fuel assembly, it is necessary to accurately calculate the two-phase flow boiling characteristics and the critical heat flux (CHF) in the fuel assembly. In this paper, a Eulerian two-fluid model combined with the extended wall boiling model was used to numerically simulate the 5 × 5 fuel rod bundle with spacer grids (four sets of mixing vane grids and four sets of simple support grids without mixing vanes). We calculated and analyzed 11 experimental conditions under different pressure, inlet temperature, and mass flux. After comparing the CHF and the location of departure from the nucleate boiling obtained by the numerical simulation with the experimental results, we confirmed the reliability of computational fluid dynamic analysis for the prediction of the CHF of the rod bundle and the boiling characteristics of the two-phase flow. Subsequently, we analyzed the influence of the spacer grid and mixing vanes on the void fraction, liquid temperature, and secondary flow distribution. The research in this article provides theoretical support for the design of fuel assemblies.

Published in Nuclear Engineering and Technology

ISSN: 1738-5733 (Print)
Publisher: Elsevier
Country of publisher: Korea, Republic of
LCC subjects: Technology: Electrical engineering. Electronics. Nuclear engineering: Nuclear engineering. Atomic power
Website: http://www.journals.elsevier.com/nuclear-engineering-and-technology

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