Kejia Li
Xiong Zheng
MENG Shuqi
desheng Jin
*The final, formatted version of the article will be published soon.
ORIGINAL RESEARCH article
Front. Energy Res.
Sec. Nuclear Energy
Volume 12 - 2024 |
doi: 10.3389/fenrg.2024.1366902
This article is part of the Research Topic Experimental and Simulation Research on Nuclear Reactor Thermal-Hydraulics, Volume II View all 9 articles
Prediction of Critical Heat Flux in a Rod Bundle Channel with Spacer Grids Based on the Eulerian Two-Fluid Model
Provisionally accepted- China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, China
The critical heat flux (CHF) is a vital parameter influencing the safety and efficiency of reactor cores. In this study, the Eulerian two-fluid model coupled with the extended wall boiling model in STAR-CCM+ was employed to simulate the departure from nucleate boiling (DNB) phenomenon in a 5×5 pressurized water reactor (PWR) fuel rod bundle channel with spacer grids under non-uniform heating conditions. The transition in boiling curves was used as the criterion of DNB occurrence, while the temperature distribution of rod surfaces was utilized for CHF location predictions. The predicted CHF value and CHF location exhibited good agreement with the experimental data. The deviation between calculated and experimental CHF values was within 15% and the deviation between predicted and experimental CHF locations was within one grid-to-grid span length. The results of this study suggested good prospects for the application of two-phase CFD model in predicting CHF in fuel assemblies with spacer grids.
Keywords: Critical heat flux, CFD, Rod bundle, two phase flow, spacer grids
Received: 07 Jan 2024; Accepted: 13 Dec 2024.
Copyright: © 2024 Li, Zheng, Shuqi, Jin, Mao, Hu, Zhou and Chen. This is an open-access article distributed under the terms of the Creative Commons Attribution License (CC BY). The use, distribution or reproduction in other forums is permitted, provided the original author(s) or licensor are credited and that the original publication in this journal is cited, in accordance with accepted academic practice. No use, distribution or reproduction is permitted which does not comply with these terms.
* Correspondence:
desheng Jin, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, China
Yulong Mao, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, China
Yisong Hu, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, China
Youxin Zhou, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, China
jun Chen, China Nuclear Power Technology Research Institute Co., Ltd, Shenzhen, China
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