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논문 기본 정보
- 자료유형
- 학술저널
- 저자정보
- 발행연도
- 2016.1
- 수록면
- 1,036 - 1,046 (11page)
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초록· 키워드
오류제보하기
We propose a primary water stress corrosion cracking (PWSCC) initiation model of Alloy600 that considers the stress triaxiality factor to apply to finite element analysis. Weinvestigated the correlation between stress triaxiality effects and PWSCC growth behaviorin cold-worked Alloy 600 stream generator tubes, and identified an additional stresstriaxiality factor that can be added to Garud's PWSCC initiation model. By applying theproposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growthsimulations based on the macroscopic phenomenological damage mechanics approachwere carried out on the PWSCC growth tests of various cold-worked Alloy 600 steamgenerator tubes and compact tension specimens. As a result, PWSCC growth behaviorresults from the finite element prediction are in good agreement with the experimentalresults.
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참고문헌 (17)
참고문헌 신청
USNRC / 2010 / Final ReportdGeneric Aging Lessons Learned (GALL)Report
W.C. Moshier / 2000 / Effect of cold work and processing orientation on stress corrosion cracking behavior of Alloy 600 / Corrosion 56:307~320
R. Bandy / 1984 / Stress corrosion cracking of Inconel Alloy 600 in high temperature water-an update / Corrosion 40:425~430
Y. Yamamoto / 2005 / Evaluation of crack growth rate for alloy 600TT SG tubing in primary and faulted secondary water environments / Proc. of the 12th Int. Conf. on Environmental Degradation of Materials in Nuclear Power SystemdWater ReactorsdTMS :1243~1252
ASME / 2007 / ASME Boiler & Pressure Vessel Code, Section XI. Rules for Inservice Inspection of Nuclear Power Plant Components
W.C. Moshier / 2000 / Effect of cold work and processing orientation on stress corrosion cracking behavior of Alloy 600 / Corrosion 56:307~320
R. Bandy / 1984 / Stress corrosion cracking of Inconel Alloy 600 in high temperature water-an update / Corrosion 40:425~430
Y. Yamamoto / 2005 / Evaluation of crack growth rate for alloy 600TT SG tubing in primary and faulted secondary water environments / Proc. of the 12th Int. Conf. on Environmental Degradation of Materials in Nuclear Power SystemdWater ReactorsdTMS :1243~1252
ASME / 2007 / ASME Boiler & Pressure Vessel Code, Section XI. Rules for Inservice Inspection of Nuclear Power Plant Components
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