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원전 이종금속 용접부의 장기 열적 시효에 따른 미세조직 및 기계적 특성변화에 관한 고찰
최경준,유승창,김지현,Choi, Kyoung Joon,Yoo, Seung Chang,Kim, Ji Hyun 한국압력기기공학회 2014 한국압력기기공학회 논문집 Vol.10 No.1
In this study, the metallurgical analysis and mechanical property measurement have been performed to investigate the effect of long-term thermal aging on the microstructural evolution in the fusion boundary region between weld metal and low alloy steel in dissimilar metal welds. A representative dissimilar weld mock-up made of Alloy 690-Alloy 152-A533 Gr. B was fabricated and aged at $450^{\circ}C$ for 2,750 hours. The microstructural characterization was conducted mainly near in a weld root region by using optical microscopy, scanning electron microscopy, transmission electron microscopy. And the mechanical properties were measured with Vickers microhardness test and nanoindentation method. A steep gradient was shown in the chemical composition profile across the interface between A533 Gr. B and Alloy 152. Type-II boundaries were found in weld side of DMW and the hardness was the highest at the narrow zone between Type-II boundary and fusion boundary.
삼축 응력이 인가된 Alloy 600의 일차수 응력부식균열 거동에 대한 연구
유승창(Seung Chang Yoo),최경준(Kyoung Joon Choi),김지수(Ji-Soo Kim),최병호(Byoung Ho Choi),김윤재(Yun-jae Kim),김종성(Jong Sung Kim),김지현(Ji Hyun Kim) 대한기계학회 2016 대한기계학회 춘추학술대회 Vol.2016 No.12
In this study, the effects of triaxial stress was investigated in terms of primary water stress corrosion susceptibility. Thermally aged Alloy 600 specimens were prepared via accelerated heat treatment at 400 °C which is one of the highest temperature which will not cause the formation of excessive carbides or precipitates which will not be formed at nuclear power plant primary circuit environment. Triaxial stress was then applied by making round notch at the surface of round tensile specimen. The crack initiation time ere measured in-situ by direct current potential drop method during slow strain rate test at primary water condition. At the notched specimen, cracks were mainly observed at the region where shear stress is focused. Also the stair shape of fracture surface was observed, which might cause by enhanced localized plasticity and multiple number of activated slip system under triaxial stress state.
원전 일차 수화학 환경에 노출된 지르코늄 핵연료 피복관의 기계적 물성 변화
김태호(Taeho Kim),최경준(Kyoung Joon Choi),유승창(Seung Chang Yoo),김지현(Ji Hyun Kim) 대한기계학회 2015 대한기계학회 춘추학술대회 Vol.2015 No.11
The effect of oxidation time on mechanical properties of zirconium fuel cladding is investigated in this study. The zirconium fuel cladding tubes were oxidized in pressurized light water reactor primary water condition with 360 ℃ and 200 bar. The dissolved hydrogen concentration was maintained 30 ㏄/㎏ and the other water chemistry was simulated the operating nuclear power plant condition. The tensile and burst pressure test were conducted with two different oxidation conditions, 50 days and 100 days from start-up. The ultimate tensile strength and yield strength were obtained from tensile test, and the rupture pressure was obtained from the burst pressure test using high pressure water injection system. The mechanical properties of two different specimens were considered and the oxide structure of zirconium fuel cladding tube specimens were also investigated using transmission electron microscopy.