Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Conceptual design of a copper-bonded steam generator for SFR and the development of its thermal-hydraulic analyzing code

  • Im, Sunghyuk (Center for Thermometry and Fluid Flow Metrology, Korea Research Institute of Standards and Science) ;
  • Jung, Yohan (Versatile Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Hong, Jonggan (Versatile Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Choi, Sun Rock (Versatile Reactor Technology Development, Korea Atomic Energy Research Institute)
  • Received : 2021.06.22
  • Accepted : 2021.12.04
  • Published : 2022.06.25
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Abstract

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

Keywords

Acknowledgement

This work was supported by the National Research Foundation of Korea, Republic of Korea (NRF) grant and National Research Council of Science & Technology (NST) grant funded by the Korean government (MSIT) [grant numbers 2021M2E2A2081061, CAP-20-03-KAERI].

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