Water corrosion resistance of ODS ferritic-martensitic steel tubes

被引：5
|
作者：
Narita, Takeshi [1 ]
Ukai, Shigeharu [1 ]
Kaito, Takeji [1 ]
Ohtsuka, Satoshi [1 ]
Matsuda, Yasuji [2 ]
机构：
[1] Japan Atom Energy Agcy, Oarai, Ibaraki 3111393, Japan
[2] Sumitomo Met Technol Inc, Amagasaki, Hyogo 6600856, Japan
来源：
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY | 2008年 / 45卷 / 02期
关键词：
F0800 (fusion reactor materials); C0500 (cladding materials); F0200 (fast reactor materials); S1000; (steels; ferritic/martensitic; low activation) spent fuel pool cooling system; water corrosion; pH; alkaline corrosion resistance; alloy elements;
D O I：
10.3327/jnst.45.99
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：
A water corrosion test was conducted on oxide dispersion strengthened (ODS) steels in consideration of the spent fuel-pool cooling condition. The two types of ODS steel tubes tested were 9Cr-ODS and 12Cr-ODS. A test temperature of 333K was selected for the test by considering the environment of spent fuel-pool cooling system. The water corrosion test was performed for 1000h under a degassed and Cl-controlled water environment at pH values of 8.4, 10 and 12. The result demonstrated that at a pH value of 12, the weight losses of 9Cr- and 12Cr-ODS are larger than those of PNC-FMS and PNC316. The test specimens lose their metallic gloss and a slightly blackish surface is observed. It is considered that 9Cr-ODS and 12Cr-ODS claddings keep their strength and are comparable to that of conventional austenitic stainless steel and ferritic-martensitic stainless steel claddings in consideration of the spent-fuel-pool cooling condition.
引用
收藏
页码：99 / 102
页数：4
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