Zobrazeno 1 - 10
of 119
pro vyhledávání: '"Hyeong-Yeon Lee"'
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 9, Pp 3764-3774 (2024)
In accordance with Regulatory Guide 1.207, Rev.1, fatigue assessments must be conducted considering the influence of primary coolant environment in nuclear reactors. Environmental fatigue, resulting from corrosion in the primary coolant, is evaluated
Externí odkaz:
https://doaj.org/article/a2e569e69a15451685114998fbb8335f
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 1, Pp 1-8 (2024)
In this study, the influence of thermal aging on structural integrity is investigated for Gr. 91 steel. A commercial grade Gr. 91 steel is used for the virgin material, and service-exposed Gr. 91 steel is sampled from a steam pipe of a super critical
Externí odkaz:
https://doaj.org/article/b35e920a39514aa6823892f612f95043
Creep and creep crack growth behaviors for base, weld, and heat affected zone in a grade 91 weldment
Publikováno v:
Nuclear Engineering and Technology, Vol 53, Iss 2, Pp 572-582 (2021)
This study investigated the creep and creep crack growth (CCG) behavior of the base metal (BM), weld metal (WM), and heat affected zone (HAZ) in a Gr. 91 weldment, which was made by a shield metal arc weld process. A series of tensile, creep, and CCG
Externí odkaz:
https://doaj.org/article/5a90f03457f7453d9c65fe9ceb05b7f4
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers - A. 46:1033-1039
Publikováno v:
Journal of the Korea Academia-Industrial cooperation Society. 23:657-667
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers - A. 46:655-662
Publikováno v:
Crystal Growth & Design.
Publikováno v:
Transactions of the Korean Society of Mechanical Engineers - A. 46:545-551
Autor:
Hyeong-Yeon Lee, Jewhan Lee
Publikováno v:
Materials at High Temperatures. 39:436-445
Publikováno v:
Nuclear Engineering and Technology, Vol 48, Iss 2, Pp 376-385 (2016)
A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic
Externí odkaz:
https://doaj.org/article/72473978cf6a41639c161ba1734caeb7