Current research activities on thermal hydraulics in Japan Sodium-cooled Fast Reactor (JSFR)

Kamide, Hideki ; Hayafune, Hiroki

Design study and related researches are carried out for Japan Sodium-cooled Fast Reactor (JSFR). Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are undergoing. Here progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed.

Language	:	English
Journal	:	Proceedings of 7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7) (CD-ROM)
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Pages	:	7 Pages
Publication Year/Month	:	2010/11
Meeting title	:	7th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-7)
Held date	:	2010/11
Location (city)	:	Chuncheon
Location (country)	:	Korea
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Registration No. : BB20102518
JAEA Abstracts No. : 39000098
Paper Submission No. : 8339

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