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Title
Japanese: 
English:Corrosion Experiments of the Candidate Materials for Liquid Lithium Lead Blanket of Fusion Reactor 
Author
Japanese: Qunying HUANG, Sheng GAO, ZhiqiangZHU, Zhihui GUO, Xinzhen LING, ZilinYAN, 近藤 正聡, Valentyn TSISAR, 室賀 健夫, Yican WU.  
English: Qunying HUANG, Sheng GAO, ZhiqiangZHU, Zhihui GUO, Xinzhen LING, ZilinYAN, Masatoshi KONDO, Valentyn TSISAR, Takeo MUROGA, Yican WU.  
Language English 
Journal/Book name
Japanese: 
English:Advances in Science and Technology 
Volume, Number, Page Vol. 73        pp. 1841-1846
Published date Oct. 2010 
Publisher
Japanese: 
English:Trans Tech Publications, Switzerland 
Conference name
Japanese: 
English: 
Conference site
Japanese: 
English: 
DOI https://doi.org/10.4028/www.scientific.net/AST.73.41
Abstract Liquid lithium lead (LiPb) eutectic is considered as one of the promising candidates of tritium breeder materials for fusion reactors. Series experiments on compatibility of LiPb with candidate structural materials such as CLAM steel and SiC f /SiC composites have been done in DRAGON serious experimental devices in FDS team such as DRAGON-RTand stirred pot device in NIFS at 500 o C and 600 o C, respectively. The weight loss of CLAM specimens exposed in flowing LiPb with the velocity of 0.17m/s increased with temperature, and the morphology and composition of the corroded surfaces were done by SEM observation and EDX analysis. The coating specimens including Al 2O3 and FeAl/Al 2O3 coatings prepared on the CLAM specimens were also exposed in the DRAGON-RT device, the results revealed that there was no obvious thinning observed on the outer surface of the protective coating. Preliminary analysis of SiC f /SiC composites specimens indicated that the mullite coating with plasma spray method on the SiC f /SiC composites specimen corroded in the high temperature LiPb, but no obvious corrosionattack was observed on the specimen surface, while the matrix and fiber of reaction-sintered composites showed slightly corrosion attack after exposure in static LiPb at 800°C for 200 hrs. Further corrosion experiment will be carried out in the near future.

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