POWER-GENERATING UNIT WITH DUAL-LOOP SODIUM-COOLED FAST REACTOR AND DUAL-WALL STEAM GENERATOR

被引：3
|
作者：
Bagdasarov, Yu. E. [1 ]
Kamaev, A. A. [1 ]
机构：
[1] State Sci Ctr Russian Federat, Leipunskii Inst Phys & Power Engn, GNTs RF FEI, Obninsk, Russia
来源：
ATOMIC ENERGY | 2009年 / 107卷 / 03期
关键词：
Steam; Steam Generator; Water Tube; Sodium Bismuth; Main Circulation Pump;
D O I：
10.1007/s10512-010-9210-6
中图分类号：
TL [原子能技术]; O571 [原子核物理学];
学科分类号：
0827 ; 082701 ;
摘要：
Due to advances in the technology of liquid metals, the idea, advanced in the 1950s, of replacing the intermediate sodium loop in a fast reactor with sodium coolant in the first loop by an intermediate stationary heat-transferring medium between the sodium flows in the first loop and water-steam is being revived. The mass/size characteristics and safety of such steam generators are evaluated from the standpoint of water-steam getting into the first loop of the reactor. Intermediate media that could be used are analyzed. An example of the implementation of dual-loop schemes for sodium-cooled fast reactors is proposed for a 900 MW(e) power-generating unit.
引用
收藏
页码：155 / 161
页数：7
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