Thermal-hydraulic analysis of lead-bismuth cooled reactors core based on single channel model

被引：0
|
作者：
Mei, Huaping [1 ]
Liu, Shuyong [1 ,2 ]
Chen, Chao [1 ,2 ]
Zhang, Jiansong [1 ,2 ]
Li, Taosheng [1 ,2 ]
机构：
[1] Institute of Nuclear Energy Safety Technology, Hefei Institutes of Physical Science, Chinese Academy of Sciences, Anhui Province, Hefei, China
[2] University of Science and Technology of China, Anhui, Hefei, China
来源：
International Journal of Nuclear Energy Science and Technology | 2024年 / 17卷 / 2-3期
关键词：
Compendex;
D O I：
10.1504/IJNEST.2024.142757
中图分类号：
学科分类号：
摘要：
Uranium dioxide
引用
收藏
页码：180 / 196
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