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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 저널정보
- 대한기계학회 Journal of Mechanical Science and Technology Journal of Mechanical Science and Technology Vol.19 No.2
- 발행연도
- 2005.2
- 수록면
- 634 - 648 (15page)
이용수
초록· 키워드
오류제보하기
Performed here o a comparative assessment study for the probabilistic fracture mechanics approach of the pressurized thermal shock of the reactor pressure vessel A round consisting of one prerequisite deterministic study and five cases for probabilistic approaches io proposed, and all organizations interested are invited The problems are solved by the participants and their results are compared to issue some recommendation of best practices and to assure an understanding of the key parameters in this type of approach, like transient description and frequency, material properties, defect type and distribution, fracture mechanics methodology etc, which will be useful in the justification through a probabilistic approach for the case of a plant over-passing the screening criteria Six participants from 3 organizations responded to the problem and their results are compiled and analyzed in this study.
목차
Abstract
1. Introduction
2. Problem Definition
3. Analysis
4. Results and Discussion
5. Conclusions
References
참고문헌 (12)
참고문헌 신청
ASME, 1998, ASME Boiler and Pressure Vessel Code, Section XI Rules for Inspection of Nuclear Power Plant Components, Appendix A Analysis of Flaws, The American Society of Mechanical Engineers
Bass, B R, Pugh, C E, Keeney, J, Schulz, H and Sievers, J, 1996, "CSNI Project for Fracture Analysis of Large-Scale International Reference Experiments (FALSIRE II)," NEA/CSNI/R(96) 1, OECD/NEA, November
Dickson, T L and Williams, P T, 2003, Fracture Analysis of Vessels-Oak Ridge FAVOR, v02 4, Computer Code Theory and Implementation of Algorithms, Methods and Correlations, draft NUREG, US Nuc1ear Regulatory Commission
Faidy, C, 2003, PROSIR-Probabilistic Structural Integrity of a PWR Reactor Pressure Vessel, Proposed to OECD/NEA PWG3-Metal Group, EDF-SEPTEN, July
Jang, C, Kang, S.C, Moonn, H R, Jeong, I S and Kim, T R, 2003, "The Effects of the Stainless Steel Cladding in Pressurized Thermal Shock Evaluation," Nuclear Engineering and Design, Vol 226, pp 127~140
Bass, B R, Pugh, C E, Keeney, J, Schulz, H and Sievers, J, 1996, "CSNI Project for Fracture Analysis of Large-Scale International Reference Experiments (FALSIRE II)," NEA/CSNI/R(96) 1, OECD/NEA, November
Dickson, T L and Williams, P T, 2003, Fracture Analysis of Vessels-Oak Ridge FAVOR, v02 4, Computer Code Theory and Implementation of Algorithms, Methods and Correlations, draft NUREG, US Nuc1ear Regulatory Commission
Faidy, C, 2003, PROSIR-Probabilistic Structural Integrity of a PWR Reactor Pressure Vessel, Proposed to OECD/NEA PWG3-Metal Group, EDF-SEPTEN, July
Jang, C, Kang, S.C, Moonn, H R, Jeong, I S and Kim, T R, 2003, "The Effects of the Stainless Steel Cladding in Pressurized Thermal Shock Evaluation," Nuclear Engineering and Design, Vol 226, pp 127~140
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UCI(KEPA) : I410-ECN-0101-2009-550-014473069