Zobrazeno 1 - 10
of 21
pro vyhledávání: '"Kenichiro Satoh"'
Publikováno v:
The Proceedings of the Materials and Mechanics Conference. 2016:OS09-19
Publikováno v:
The Proceedings of the Materials and Mechanics Conference. 2015:GS0101-10
Publikováno v:
Nuclear Science and Engineering. 178:76-85
The primary hot-leg piping system of the advanced sodium-cooled fast reactor under conceptual study in Japan (named Japan sodium-cooled fast reactor: JSFR) utilizes large-diameter and thin-walled p...
Publikováno v:
Journal of Pressure Vessel Technology. 138
Creep strength enhanced ferritic (CSEF) steels including ASME Gr.91 are widely used in fossil power plants. In the advanced loop-type sodium-cooled fast reactor (SFR), modified 9Cr–1Mo steel (ASME Gr.91) is going to be adopted as a structural mater
Publikováno v:
Journal of Pressure Vessel Technology. 138
Type 316 stainless steel with low-carbon and medium-nitrogen contents called 316FR stainless steel is a candidate structural material for reactor vessels and internals of future-generation fast breeder reactors (FBRs). The reactor vessel cannot be ma
Publikováno v:
Volume 3: Design and Analysis.
This paper proposes provisional welded joint strength reduction factors (WJSRF) of modified 9Cr-1Mo steel (ASME Gr.91) applicable to the structural designing of “Japan sodium cooled fast reactor (JSFR)”. In the welded joints of creep strength enh
Autor:
Kenichiro Satoh, Masanori Ando, Tomomi Otani, Kazuyuki Tsukimori, Koichi Kikuchi, Tai Asayama, Sota Watanabe
Publikováno v:
Volume 1B: Codes and Standards.
New 2012 edition of JSME code for design and construction of fast reactors (FRs code) was published by Japan society of mechanical engineers (JSME). Main topic of the current JSME FRs code 2012 edition is registration of the two new materials, 316FR
Autor:
Hideki Takasho, Shingo Date, Masanori Ando, Koichi Kikuchi, Kazuyuki Tsukimori, Kenichiro Satoh, Hiroshi Kanasaki
Publikováno v:
Volume 6: Materials and Fabrication, Parts A and B.
In a component design at elevated temperature, fatigue and creep-fatigue is one of the most important failure modes, and fatigue and creep-fatigue life assessment in structural discontinuities is important issue to evaluate structural integrity of th
Publikováno v:
ASME 2010 Pressure Vessels and Piping Conference: Volume 1.
Main loadings of reactor vessels in fast reactor plants are thermal stresses induced by fluid temperature change at transient operation. Structures respond to them with elastic plastic creep deformation under high temperature conditions. It can induc
Publikováno v:
The Proceedings of Mechanical Engineering Congress, Japan. 2015:S0820104