文献
J-GLOBAL ID:202102249743870854
整理番号:21A0157067
ナトリウム冷却高速炉のシステム安全性解析のための過渡熱流力コードの研究開発【JST・京大機械翻訳】
Research and development of a transient thermal-hydraulic code for system safety analysis of sodium cooled fast reactor
著者 (7件):
Yu Yang
(Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610213, China)
,
Liu Dong
(Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610213, China)
,
Song Xiaoming
(Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610213, China)
,
Li Zhongchun
(Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610213, China)
,
Guo Fengchen
(Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610213, China)
,
Wang Yafeng
(Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610213, China)
,
Pang Bo
(Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China, Chengdu, Sichuan 610213, China)
資料名:
Annals of Nuclear Energy
(Annals of Nuclear Energy)
巻:
152
ページ:
Null
発行年:
2021年
JST資料番号:
C0325D
ISSN:
0306-4549
資料種別:
逐次刊行物 (A)
記事区分:
原著論文
発行国:
イギリス (GBR)
言語:
英語 (EN)