Zobrazeno 1 - 10
of 53
pro vyhledávání: '"Koji DOZAKI"'
Publikováno v:
Nihon Kikai Gakkai ronbunshu, Vol 81, Iss 827, Pp 15-00063-15-00063 (2015)
To assess the integrity of piping under the detonation pressure of accumulated hydrogen-oxygen, the fracture strain of the pipe material must be identified. In carbon steel pipe specimens that had been experimentally ruptured by hydrogen-oxygen deton
Externí odkaz:
https://doaj.org/article/b9c8506c5084432c8c795cdf17c1e917
Publikováno v:
Nihon Kikai Gakkai ronbunshu, Vol 81, Iss 827, Pp 14-00665-14-00665 (2015)
Carbon steel pipe specimens, which had been experimentally ruptured or deformed by hydrogen-oxygen detonation, were analysed by dynamic analysis using a finite element method (FEM). Pressure load histories induced by detonation were estimated by comp
Externí odkaz:
https://doaj.org/article/d234d96fdd7b4b599fbc5d245538aa94
Publikováno v:
Transactions of the Atomic Energy Society of Japan. 15:21-31
Publikováno v:
Volume 1B: Codes and Standards.
Maintenance of nuclear power plant facilities involves activities comprising a large system composed of both plant hardware and human subsystems to assure safe and reliable operation. Maintenance activities are composed of inspection, evaluation and
Publikováno v:
The Proceedings of the National Symposium on Power and Energy Systems. :157-160
Autor:
Seiichi Koshizuka, Masanori Naitoh, Shunsuke Uchida, Hisashi Ninokata, Koji Dozaki, Akihiko Minato, Minoru Akiyama, Naoki Anahara, Hiroaki Saitoh, Koji Hotta
Publikováno v:
Heat Transfer Engineering. 29:712-720
The troubles of major components and structural materials in nuclear power plants have often been caused by flow-induced vibration, corrosion, and their overlapping effects. In order to establish safe and reliable plant operation, it is required to f
Autor:
Yoshinori Matsui, Yoshiyuki Kaji, Junichi Nakano, Akira Shibata, Takashi Tsukada, Koji Dozaki, Masao Ohmi, Kazuo Kawamata, Hirokazu Ugachi, Nobuaki Nagata, Hideki Takiguchi
Publikováno v:
Journal of Nuclear Science and Technology. 45:725-734
Irradiation-assisted stress corrosion cracking (IASCC) is one of the critical concerns when stainless steel components have been in service in light water reactors for a long period. In-core IASCC growth tests have been carried out using the compact
Publikováno v:
Volume 1B: Codes and Standards.
Chapter of Repair, Replacement Activities (RRA) in Fitness-For-Service (FFS) Code of the Japanese Society of Mechanical Engineers (JSME) includes rules of article RB-3000 for temporary repair techniques in the use of covering leakage during operation
Publikováno v:
Nuclear Science and Engineering. 149:312-324
Hydrogen injection has been applied as a preventive measure against the stress corrosion cracking (SCC) phenomenon in many boiling water reactors. However, it can be applied only during normal plant operation since hydrogen is usually injected into t
Autor:
Koji Dozaki
Publikováno v:
The Proceedings of Mechanical Engineering Congress, Japan. 2017:F031004