지원사업
학술연구/단체지원/교육 등 연구자 활동을 지속하도록 DBpia가 지원하고 있어요.
커뮤니티
연구자들이 자신의 연구와 전문성을 널리 알리고, 새로운 협력의 기회를 만들 수 있는 네트워킹 공간이에요.
논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2021.8
- 수록면
- 189 - 195 (7page)
- DOI
- 10.14773/cst.2021.20.4.189
이용수
초록· 키워드
오류제보하기
A technology for designing and licensing a dedicated radiation shielding facility needs to be developed for safe and efficient operation an R&D center. Technology development is important for smooth operation of such facilities. Causes of damage to internal structures (such as baffle former bolt (BFB) of pressurized water reactor) of a nuclear power reactor should be analyzed along with prevention and countermeasures for similar cases of other plants. It is important to develop technologies that can comprehensively analyze various characteristics of internal structures of long term operated reactors. In high-temperature, high-pressure operating environment of nuclear power plants, cases of BFB cracks caused by irradiated assisted stress corrosion cracks (IASCC) have been reported overseas. The integrity of a reactor’s internal structure has emerged as an important issue. Identifying the cause of the defect is requested by the Korean regulatory agency. It is also important to secure a foundation for testing technology to demonstrate the operating environment for medium-level irradiated testing materials. The demonstration testing facility can be used for research on material utilization of the plant, which might have highest fluence on the internal structure of a reactor globally.
목차
등록된 정보가 없습니다.
참고문헌 (3)
참고문헌 신청
Z. Jiao / 2007 / Deformation microstructure of proton-irradiated stainless steels / Journal of Nuclear Materials 361(2-3):218~227
O.K. Chopra / 2011 / A review of irradiation effects on LWR core internal materials – IASCC susceptibility and crack growth rates of austenitic stainless steels / Journal of Nuclear Materials 409(3):235~256
임연수 / 2021 / 양성자 조사가 316 스테인리스강의 미세조직과 표면산화 특성에 미치는 영향 / CORROSION SCIENCE AND TECHNOLOGY 20(3):158~168
O.K. Chopra / 2011 / A review of irradiation effects on LWR core internal materials – IASCC susceptibility and crack growth rates of austenitic stainless steels / Journal of Nuclear Materials 409(3):235~256
임연수 / 2021 / 양성자 조사가 316 스테인리스강의 미세조직과 표면산화 특성에 미치는 영향 / CORROSION SCIENCE AND TECHNOLOGY 20(3):158~168
함께 읽어보면 좋을 논문
논문 유사도에 따라 DBpia 가 추천하는 논문입니다. 함께 보면 좋을 연관 논문을 확인해보세요!
-
원자로 내부구조물 재료열화이력 및 관리방안
한국압력기기공학회 논문집
2016 .06
-
원자로 내부구조물 균열개시 민감도에 미치는 영향인자 고찰
Corrosion Science and Technology
2021 .08
-
AM600: A New Look at the Nuclear Steam Cycle
Nuclear Engineering and Technology
2017 .01
-
TOP-MOUNTED IN-CORE INSTRUMENTATION : CURRENT STATUS AND TECHNICAL ISSUES
에너지공학
2015 .06
-
Technology Selection for Offshore Underwater Small Modular Reactors
Nuclear Engineering and Technology
2016 .01
-
원자로 내부구조물의 개선된 동특성 규명 방법
한국소음진동공학회 학술대회논문집
2015 .10
-
Inverse method to obtain reactivity in nuclear reactors with P1 point reactor kinetics model using matrix formulation
Nuclear Engineering and Technology
2021 .02
-
Enhancing Utilization and Ensuring Security: Insights to Compromise Contradicting Conditions in New Research Reactors
Nuclear Engineering and Technology
2021 .05
-
Robust feedback-linearization control for axial power distribution in pressurized water reactors during load-following operation
Nuclear Engineering and Technology
2018 .01
-
Assessment of N-16 activity concentration in Bangladesh Atomic Energy Commission TRIGA Research Reactor
Nuclear Engineering and Technology
2018 .01
-
IASCC growth analysis of reactor internal structure considering welding residual stress and neutron irradiation effect
대한기계학회 춘추학술대회
2021 .11
-
Measurement of vibration and stress for APR-1400 reactor internals .
Nuclear Engineering and Technology
2018 .01
-
Thermal-hydraulic analysis of a new conceptual heat pipe cooled small nuclear reactor system
Nuclear Engineering and Technology
2020 .01
-
신규 개발 저출력 연구용 원자로 구조물 집합체의 동특성 해석
대한기계학회 춘추학술대회
2018 .12
-
Risk-informed approach to the safety improvement of the reactor protection system of the AGN-201K research reactor
Nuclear Engineering and Technology
2020 .01
-
Development of classification criteria for non-reactor nuclear facilities in Korea
Nuclear Engineering and Technology
2023 .02
-
Conceptual design for a 5 kWe space nuclear reactor power system
Nuclear Engineering and Technology
2024 .09
-
Development of a new CVAP structural analysis methodology of APR1400 reactor internals using scaled model tests
Nuclear Engineering and Technology
2024 .01
-
INSTRUMENTATION AND CONTROL STRATEGIES FOR AN INTEGRAL PRESSURIZED WATER REACTOR
Nuclear Engineering and Technology
2015 .01
-
Vessel failure sensitivities of an advanced reactor for SBLOCA
Nuclear Engineering and Technology
2020 .01
이 논문의 저자 정보
최근 본 자료
댓글(0)
0