Thermal-hydraulics testing needs for sodium-cooled fast reactors in Japan

Yamano, Hidemasa

The contents are Demonstration Sodium-cooled Fast Reactor (SFR), Analysis code V&V and testing for design and safety analysis toward licensing, Facilities for SFR Development, Facilities for SFR Development.

Language	:	English
Journal	:
Volume	:
Number	:
Pages	:
Publication Year/Month	:
Meeting title	:	21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21)
Held date	:	2025/08
Location (city)	:	Busan
Location (country)	:	Korea
Patent information	:
PDF	:
Paper URL	:
DOI for research data	:	Link to research data related to this paper/report.
Keywords	:	ナトリウム冷却高速炉; 熱流動
Research Facility	:
Press Release	:
JAEA R&D Navigator	:
Cooperating Institute	:	経済産業省

Accesses	:	- Accesses
Web of Science® Times Cited Count	:	Times Cited Count： If you would like to get the latest times cited, please access the Web of Science®. http://www.webofknowledge.com/wos
InCites™	:
Altmetrics	:

Registration No. : BB20250955
JAEA Abstracts No. :
Paper Submission No. :

[CLARIVATE ANALYTICS], [WEB OF SCIENCE], [HIGHLY CITED PAPER & CUP LOGO] and [HOT PAPER & FIRE LOGO] are trademarks of Clarivate Analytics, and/or its affiliated company or companies, and used herein by permission and/or license.