Zobrazeno 1 - 10
of 18
pro vyhledávání: '"Ill Seok Jeong"'
Publikováno v:
Nuclear Engineering and Technology. 43:83-88
To define the effect of strain rate variation from 0.04% to 0.004%/s on environmental fatigue of CF8M cast stainless steel, which is used as a primary piping material in nuclear power plants, low-cycle fatigue tests were conducted at operating pressu
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::87721d5cb94b4ca8235a960fa991d306
https://doi.org/10.5516/net.2011.43.1.083
https://doi.org/10.5516/net.2011.43.1.083
Publikováno v:
Journal of Loss Prevention in the Process Industries. 22:879-883
To develop a fatigue design curve of cast stainless steel CF8M that has been used in piping material of nuclear power plants, a low cycle fatigue tests were conducted in the environment of a pressurized water reactor at 15 MPa and 315 °C. Cylindrica
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::af46f7136be7b55e50930bad9223af65
https://doi.org/10.1016/j.jlp.2009.07.011
https://doi.org/10.1016/j.jlp.2009.07.011
Publikováno v:
Key Engineering Materials. :1011-1014
For developing fatigue design curve of cast stainless steel that is used in piping material of nuclear power plants, a low-cycle fatigue test rig was built. It is capable of performing tests in pressurized high temperature water environment of PWR. C
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::9ca5ab507093e1f2213f2777fa8ec4a7
https://doi.org/10.4028/www.scientific.net/kem.326-328.1011
https://doi.org/10.4028/www.scientific.net/kem.326-328.1011
Publikováno v:
Key Engineering Materials. :968-973
Environmental fatigue crack propagation of CF8M and CF8A steels used in the domestic nuclear power plants (NPPs) were investigated on the simulated pressurized water reactor (PWR) condition (temperature: 316°C, pressure: 15MPa). The test equipment f
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::1e60931b0c3722c924885f2de6e14c7c
https://doi.org/10.4028/www.scientific.net/kem.297-300.968
https://doi.org/10.4028/www.scientific.net/kem.297-300.968
Publikováno v:
Key Engineering Materials. :102-107
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::199209dc6f8cef088092dd9e4b7c8068
https://doi.org/10.4028/www.scientific.net/kem.270-273.102
https://doi.org/10.4028/www.scientific.net/kem.270-273.102
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers A. 28:460-466
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::5d75df74fc10617e730272e025d87027
https://doi.org/10.3795/ksme-a.2004.28.4.460
https://doi.org/10.3795/ksme-a.2004.28.4.460
Publikováno v:
Nuclear Engineering and Design. 226:127-140
Fracture mechanics analysis is the key element of the integrity evaluation of the nuclear reactor pressure vessel (RPV), such as the pressurized thermal shock (PTS) analysis and P–T limit curve construction. However, the existence of stainless stee
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::4323db5597c7a16523434fcfa94656e5
https://doi.org/10.1016/s0029-5493(03)00190-0
https://doi.org/10.1016/s0029-5493(03)00190-0
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers A. 27:94-101
The pressure tubes made of cold-worked Zr-2.5Nb alloy are subjected to creep deformation during service period resulting in changes to their geometry such as longitudinal elongation, diameter increase and sagging. To evaluate integrity of them, infor
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::d15326af1f1b3bf33eea2d204b3755e8
https://doi.org/10.3795/ksme-a.2003.27.1.094
https://doi.org/10.3795/ksme-a.2003.27.1.094
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers A. 26:2390-2398
For integrity analysis of nuclear reactor pressure vessel, including the Pressurized thermal shock analysis, the fast and accurate calculation of the stress intensity factor at the crack tip is needed. For this, a simple approximation scheme is devel
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::ff5fdd9b332b455546f1b5c65459cb93
https://doi.org/10.3795/ksme-a.2002.26.11.2390
https://doi.org/10.3795/ksme-a.2002.26.11.2390
Publikováno v:
Nuclear Engineering and Design. 214:163-172
The reactor pressure vessel (RPV) is the key component of pressurized water reactor. It has to comply with various rules and regulatory guides to ensure sufficient safety and operating margins during the plant lifetime. Thus, it is crucial to assure
Externí odkaz:
https://explore.openaire.eu/search/publication?articleId=doi_________::163458dd79ba7ed4b4067565af4d1db2
https://doi.org/10.1016/s0029-5493(02)00025-0
https://doi.org/10.1016/s0029-5493(02)00025-0