Zobrazeno 1 - 10
of 16
pro vyhledávání: '"Sangyeob Lim"'
Publikováno v:
Materials, Vol 16, Iss 24, p 7557 (2023)
This study investigates the tensile behaviors of additively manufactured (AM) 17-4PH stainless steels heat-treated within various temperature ranges from 400 °C to 700 °C in order to identify the effective aging temperature. Despite an aging treatm
Externí odkaz:
https://doaj.org/article/9bc3ebb6ed294c24a5d5ed8d87943906
Publikováno v:
Nuclear Engineering and Technology, Vol 49, Iss 4, Pp 760-768 (2017)
The transition temperature shift (TTS) of the reactor pressure vessel materials is an important factor that determines the lifetime of a nuclear power plant. The prediction of the TTS at the end of a plant’s lifespan is calculated based on the equa
Externí odkaz:
https://doaj.org/article/fb222ad778e1415e8383cd318c94ffd9
Publikováno v:
Journal of Nuclear Materials. 523:458-471
In order to study the mechanical properties and the anisotropy of a Zr-2.5 wt%Nb pressure tube material, tensile tests along the principal tube directions, which are axial, transverse and radial, were performed at room temperature using a pressure tu
Publikováno v:
Metals and Materials International. 25:838-845
We analyzed the microstructural characteristics such as number density and length and width of hydrides in Zr–2.5 wt%Nb pressure tube. The hydrogen was charged cathodically and the hydride-contained sample was evaluated using the advanced analysis
Publikováno v:
Corrosion Science. 145:109-118
The purpose of this work is to provide insights into the lead-accelerated stress corrosion cracking mechanisms of Alloy 600 from the perspective of the electrochemical role of lead. Lead-species were electrochemically reduced at the oxide/matrix inte
Autor:
Dongchan Jang, Hyung-Ha Jin, Chansun Shin, Junhyun Kwon, Sangyeob Lim, Hwanuk Guim, Sangeun Kim, Jungwoo Heo
Publikováno v:
Journal of Nuclear Materials. 503:263-270
HT9, a ferritic/martensitic steel, is a candidate structural material for next-generation advanced reactors. Its microstructure is a typical tempered martensite showing a hierarchical lath-block-and-packet structure. We investigate the specimen size
Publikováno v:
Journal of Nuclear Materials. 552:152978
Canada Deuterium Uranium (CANDU) reactors use Zr-2.5Nb alloys as the pressure tube material. Ensuring optimal performance of Zr-2.5Nb pressure tubes is vital to the safety of CANDU-type heavy water reactors. Long-term use of the pressure tube induces
Autor:
Hariharan Krishnaswamy, Jayant Jain, Hobyung Chae, Hyongjoon Lee, Soo Yeol Lee, Kali Prasad, Min Ji Song, Ke An, Sangyeob Lim, You Sub Kim
Publikováno v:
Mechanics of Materials. 158:103899
We examined the repetitive stress relaxation behavior of extruded AA7075 subjected to different heat-treatments (solid-solutionized (SS), under-aged (UA), peak-aged (PA)). Compared to SS and UA, the PA showed the highest stress drop (Δσs), plastic
Autor:
Weon-Ju Kim, Daejong Kim, Sung-Woo Kim, Sangyeob Lim, Arang Do, Heon Jin Choi, Hyeon-Geun Lee
Publikováno v:
Journal of Nuclear Materials. 549:152903
In order to improve the hydrothermal corrosion resistance of nuclear fuel cladding, TiN and TiCrN were deposited on the surface of a Zr-based alloy tube as an environmental barrier coating using arc ion plating and DC sputtering under an Ar+N2 atmosp
Publikováno v:
Nuclear Instruments and Methods in Physics Research Section B: Beam Interactions with Materials and Atoms. 409:318-322
Microstructural changes in austenitic stainless steel caused by hydrogen ion irradiation were investigated using transmission electron microscopy (TEM). It has been confirmed that the irradiation induced the formation of martensite along the grain bo