Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

被引:1
作者
Oya, Yasuhisa [1 ]
Oda, Takuji [2 ]
Tanaka, Satoru [3 ]
Okuno, Kenji [1 ]
机构
[1] Shizuoka Univ, Fac Sci, Radiochem Res Lab, Suruga Ku, Shizuoka 4228529, Japan
[2] Univ Tokyo, Sch Engn, Dept Nucl Engn & Management, Bunkyo Ku, Tokyo 1138656, Japan
[3] Univ Tokyo, Sch Engn, Dept Quantum Engn & Syst Sci, Bunkyo Ku, Tokyo 1138656, Japan
关键词
D O I
10.13182/FST08-A1826
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
The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D-2 gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing. Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 11200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment.
引用
收藏
页码:337 / 340
页数:4
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