Zobrazeno 1 - 10
of 186
pro vyhledávání: '"Changheui Jang"'
Publikováno v:
Journal of Materials Research and Technology, Vol 32, Iss , Pp 49-58 (2024)
This study aimed at investigation of thermal aging effects on evolution of microstructure and mechanical properties of ER316L austenitic stainless-steel weld (ASSW). The ASSW was subjected to thermal aging at 400 °C for up to 30,000 h. Initially, th
Externí odkaz:
https://doaj.org/article/2e99aeb4f8954119b3d82cc3637419d9
Autor:
Muthu Shanmugam Mannan, Changheui Jang
Publikováno v:
Journal of Materials Research and Technology, Vol 32, Iss , Pp 4278-4292 (2024)
In this decade, the working temperature of the power plants significantly increased to above 700 °C to enhance efficiency. The corrosive species deposits on the hot section components were prone to corrosion damage at elevated temperatures. This stu
Externí odkaz:
https://doaj.org/article/0b2938b868e744f596c705b6925b181e
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
Autor:
Chaewon Jeong, Ji Ho Shin, Byeong Seo Kong, Junjie Chen, Qian Xiao, Changheui Jang, Yun-Jae Kim
Publikováno v:
Nuclear Engineering and Technology, Vol 56, Iss 6, Pp 2131-2140 (2024)
The chloride-induced stress corrosion cracking (CISCC) is one of the major integrity concerns in dry storage canisters made of austenitic stainless steels (ASSs). In this study, an advanced duplex stainless steel (DSS) with a composition of Fe–19Cr
Externí odkaz:
https://doaj.org/article/4cc17309f06847ddb9a5592d065f9718
Publikováno v:
Nuclear Engineering and Technology, Vol 55, Iss 8, Pp 2823-2834 (2023)
Austenitic stainless steel welds (ASSWs) of nuclear components undergo aging-related degradations caused by high temperature and neutron radiation. Since irradiation leads to the change of material characteristics, relevant quantification is importan
Externí odkaz:
https://doaj.org/article/a42c1206ab904cd9926071b259fafcdf
Publikováno v:
Heliyon, Vol 9, Iss 12, Pp e22538- (2023)
In this study, we developed and optimized a trivalent chromium coating electrodeposited on 304L stainless steel (SS) from a Cr-trivalent bath. The results reveal that the Cr coatings at all bath temperatures except for 80 °C showed clusters of polyh
Externí odkaz:
https://doaj.org/article/e91697b64edc424aa66bbaf9d3d6c7bf
Autor:
Hongsheng Chen, Jiandong Luo, Tengfei Zhang, Changheui Jang, Rui Tang, Baojun Dong, Jin Li, Xuesong Leng
Publikováno v:
Journal of Materials Research and Technology, Vol 23, Iss , Pp 5246-5259 (2023)
The microstructures, tensile properties and corrosion behaviors of the Fe20Cr25NiNb stainless steels with different Al contents were investigated in the supercritical carbon dioxide. Results show that the Fe20Cr25NiNb stainless steels transform from
Externí odkaz:
https://doaj.org/article/5d678357eda74c1c91e45dcfdd95cd10
Publikováno v:
Journal of Materials Research and Technology, Vol 23, Iss , Pp 4990-5003 (2023)
The deformation behavior of δ-ferrite in an aged austenitic stainless-steel weld was investigated to obtain insights on the contribution of aging-induced nanofeatures, such as spinodal decomposition and G-phase precipitation, to aging embrittlement.
Externí odkaz:
https://doaj.org/article/7c9321870bb449bc9d1dc231599e6ad3
Publikováno v:
Nuclear Engineering and Technology, Vol 55, Iss 2, Pp 555-565 (2023)
Despite many advantages as structural materials, austenitic stainless steels (SSs) have been avoided in many next generation nuclear systems due to poor void swelling resistance. In this paper, we report the results of heavy ion irradiation to the re
Externí odkaz:
https://doaj.org/article/dd6762da0bfb4c7caa1c82c13efd07ee
Autor:
Shanmugam Mannan Muthu, Hyeon-Bae Lee, Bright O. Okonkwo, Dong Wang, Changheui Jang, Taehyung Na
Publikováno v:
Materials, Vol 16, Iss 24, p 7589 (2023)
For the dry storage of Canada Deuterium Uranium (CANDU) spent nuclear fuels, the integrity of Zircaloy-4 fuel cladding has to be verified. However, the formation of ~10 µm-thick oxide layers in typical CANDU reactor operating conditions takes severa
Externí odkaz:
https://doaj.org/article/62dd74de8e67482f9bb0283f15c265ed