Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Beryllium oxide utilized in nuclear reactors: Part I: Application history, thermal properties, mechanical properties, corrosion behavior and fabrication methods

  • Ming-dong Hou (Institute of Nuclear and New Energy Technology, Tsinghua University) ;
  • Xiang-wen Zhou (Institute of Nuclear and New Energy Technology, Tsinghua University) ;
  • Bing Liu (Institute of Nuclear and New Energy Technology, Tsinghua University)
  • Received : 2022.03.04
  • Accepted : 2022.07.21
  • Published : 2022.12.25
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Abstract

In recent years, beryllium oxide has been widely utilized in multiple compact nuclear reactors as the neutron moderator, the neutron reflector or the matrix material with dispersed nuclear fuels due to its prominent properties. In the past 70 years, beryllium oxide has been studied extensively, but rarely been systematically organized. This article provides a systematic review of the application history, thermal properties, mechanical properties, corrosion behavior and fabrication methods of beryllium oxide. Data from previous literature are extracted and sorted out, and all of these original data are attached as the supplementary material, so that subsequent researchers can utilize this paper as a database for beryllium oxide research in reactor design or simulation analysis, etc. In addition, this review article also attempts to point out the insufficiency of research on beryllium oxide, and the possible key research areas about beryllium oxide in the future.

Keywords

Acknowledgement

This work was funded by the National S&T Major Project (Grant No. ZX06901) and Key R&D Plan of Shandong Province (major scientific and technological innovation project, 2020CXGC010306).

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