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RESEARCH ACTIVITIES ON A SUPERCRITICAL PRESSURE WATER REACTOR IN KOREA
Bae, Yoon-Yeong,Jang, Jin-Sung,Kim, Hwan-Yeol,Yoon, Han-Young,Kang, Han-Ok,Bae, Kang-Mok Korean Nuclear Society 2007 Nuclear Engineering and Technology Vol.39 No.4
This paper presents the research activities performed to date for the development of a supercritical pressure water-cooled reactor (SCWR) in Korea. The research areas include a conceptual design of an SCWR with an internal flow recirculation, a reactor core conceptual design, a heat transfer test with supercritical $CO_2$, an adaptation of an existing safety analysis code to the supercritical pressure condition, and an evaluation of candidate materials through a corrosion study. Methods to reduce the cladding temperature are introduced from two different perspectives, namely, thermal-hydraulics and core neutronics. Briefly described are the results of an experiment on the heat transfer at a supercritical pressure, an experiment that is essential for the analysis of the subchannels of fuel assemblies and the analysis of a system safety. An existing system code has been adapted to SCWR conditions, and the process of a first-hand validation is presented. Finally, the corrosion test results of the candidate materials for an SCWR are introduced.
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