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ORIGINAL RESEARCH article

Front. Energy Res.
Sec. Nuclear Energy
Volume 12 - 2024 | doi: 10.3389/fenrg.2024.1432836
This article is part of the Research Topic Numerical and Experimental Studies on Small/Micro Nuclear Reactors, Volume II View all 3 articles

Application of Flow and Heat Transfer Network Analysis Method to the Core of Prismatic Gas-Cooled Micro Reactor

Provisionally accepted
Shuoting ZHANG Shuoting ZHANG *Jianhua Dong Jianhua Dong Zhe Zhou Zhe Zhou Zheng Huang Zheng Huang Guoming Liu Guoming Liu Qiaoyan Chen Qiaoyan Chen
  • China Nuclear Power Engineering Co Ltd, Beijing, China

The final, formatted version of the article will be published soon.

    The prismatic gas-cooled micro reactor core is composed of graphite prismatic blocks with gaps between them. The coolant passes through the gap to form flows that affect the flow and temperature distribution of the core. Conventionally, computational fluid dynamics methods have been used for prismatic gas-cooled reactor core analysis. However, they require considerable computational time and cost. For rapid and accurate calculation, in this study, a flow and heat transfer network analysis method is developed for evaluation of the core flow and temperature distribution. Finally, the calculated results of the flow heat transfer network analysis method are compared with the calculated results of computational fluid dynamics. The results show that the results of the flow heat transfer network analysis method are in good agreement with the results of computational fluid dynamics.

    Keywords: flow network, Bypass flow, Maximum fuel temperature, Prismatic gas-cooled reactor, code V&V

    Received: 14 May 2024; Accepted: 31 Jul 2024.

    Copyright: © 2024 ZHANG, Dong, Zhou, Huang, Liu 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: Shuoting ZHANG, China Nuclear Power Engineering Co Ltd, Beijing, China

    Disclaimer: All claims expressed in this article are solely those of the authors and do not necessarily represent those of their affiliated organizations, or those of the publisher, the editors and the reviewers. Any product that may be evaluated in this article or claim that may be made by its manufacturer is not guaranteed or endorsed by the publisher.