ENVIRONMENTAL FATIGUE OF METALLIC MATERIALS IN NUCLEAR POWER PLANTS - A REVIEW OF KOREAN TEST PROGRAMS

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dc.contributor.authorJang, Changheuiko
dc.contributor.authorJang, Hunko
dc.contributor.authorHong, Jong-Daeko
dc.contributor.authorCho, Hyunchulko
dc.contributor.authorKim, Tae Soonko
dc.contributor.authorLee, Jae-Gonko
dc.date.accessioned2014-08-26T08:22:00Z-
dc.date.available2014-08-26T08:22:00Z-
dc.date.created2014-01-13-
dc.date.created2014-01-13-
dc.date.issued2013-12-
dc.identifier.citationNUCLEAR ENGINEERING AND TECHNOLOGY, v.45, no.7, pp.929 - 940-
dc.identifier.issn1738-5733-
dc.identifier.urihttp://hdl.handle.net/10203/187124-
dc.description.abstractEnvironmental fatigue of the metallic components in light water reactors has been the subject of extensive research and regulatory interest in Korea and abroad. Especially, it was one of the key domestic issues for the license renewal of operating reactors and licensing of advanced reactors during the early 2000s. To deal with the environmental fatigue issue domestically, a systematic test program has been initiated and is still underway. The materials tested were SA508 Gr.1a low alloy steels, 316LN stainless steels, cast stainless steels, and an Alloy 690 and 52M weld. Through tests and subsequent analysis, the mechanisms of reduced low cycle fatigue life have been investigated for those alloys. In addition, the effects of temperature, dissolved oxygen level, and dissolved hydrogen level on low cycle fatigue behaviors have been investigated. In this paper, the test results and key analysis results are briefly summarized. Finally, an on-going test program for hot-bending of 347 stainless steel is introduced.-
dc.languageEnglish-
dc.publisherKOREAN NUCLEAR SOC-
dc.subjectLOW-CYCLE FATIGUE-
dc.subjectAUSTENITIC STAINLESS-STEEL-
dc.subjectLOW-ALLOY STEEL-
dc.subjectDEOXYGENATED WATER-
dc.subjectCRACK-GROWTH-
dc.subjectDEGREES-C-
dc.subject310-DEGREES-C-
dc.subjectDEFORMATION-
dc.subjectBEHAVIORS-
dc.subjectHYDROGEN-
dc.titleENVIRONMENTAL FATIGUE OF METALLIC MATERIALS IN NUCLEAR POWER PLANTS - A REVIEW OF KOREAN TEST PROGRAMS-
dc.typeArticle-
dc.identifier.wosid000329681900009-
dc.identifier.scopusid2-s2.0-84892729264-
dc.type.rimsART-
dc.citation.volume45-
dc.citation.issue7-
dc.citation.beginningpage929-
dc.citation.endingpage940-
dc.citation.publicationnameNUCLEAR ENGINEERING AND TECHNOLOGY-
dc.identifier.doi10.5516/NET.07.2013.040-
dc.embargo.liftdate9999-12-31-
dc.embargo.terms9999-12-31-
dc.contributor.localauthorJang, Changheui-
dc.contributor.nonIdAuthorJang, Hun-
dc.contributor.nonIdAuthorHong, Jong-Dae-
dc.contributor.nonIdAuthorCho, Hyunchul-
dc.contributor.nonIdAuthorKim, Tae Soon-
dc.contributor.nonIdAuthorLee, Jae-Gon-
dc.type.journalArticleArticle-
dc.subject.keywordAuthorEnvironmental Fatigue-
dc.subject.keywordAuthorLow Cycle Fatigue-
dc.subject.keywordAuthorMicrostructure Effect-
dc.subject.keywordAuthorDynamic Strain Ageing-
dc.subject.keywordAuthorHydrogen Induced Cracking-
dc.subject.keywordPlusLOW-CYCLE FATIGUE-
dc.subject.keywordPlusAUSTENITIC STAINLESS-STEEL-
dc.subject.keywordPlusLOW-ALLOY STEEL-
dc.subject.keywordPlusDEOXYGENATED WATER-
dc.subject.keywordPlusCRACK-GROWTH-
dc.subject.keywordPlusDEGREES-C-
dc.subject.keywordPlus310-DEGREES-C-
dc.subject.keywordPlusDEFORMATION-
dc.subject.keywordPlusBEHAVIORS-
dc.subject.keywordPlusHYDROGEN-
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