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Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants
Tan, Jibo1; Zhang, Ziyu1; Zheng, Hui2; Wang, Xiang1; Gao, Jun1; Wu, Xinqiang1; Han, En-Hou1; Yang, Shuangliang2; Huang, Pengtao2
Corresponding AuthorWu, Xinqiang(xqwu@imr.ac.cn)
2020-12-01
Source PublicationJOURNAL OF NUCLEAR MATERIALS
ISSN0022-3115
Volume541Pages:10
AbstractLow cycle fatigue behavior of 316LN stainless steel (SS), 304 SS and 308L weld metal was investigated in air and high-temperature pressurized water environment. The effects of mechanical factors and environmental factors on fatigue lives were considered. The fatigue mean curve and design curve for nucleargrade austenitic SSs in air are obtained by fitting fatigue data. The corrosion fatigue model for nucleargrade austenitic SSs in high-temperature pressurized water environment is proposed based on environmental fatigue correction factor method, which mainly considers the effects of strain rate, temperature and dissolved oxygen concentration. (C) 2020 Elsevier B.V. All rights reserved.
KeywordCorrosion fatigue Austenitic stainless steel High temperature corrosion Model
Funding OrganizationNational Key R&D Program of China ; National Science and Technology Major Project ; National Natural Science Foundation of China ; Chinese Academy of Sciences ; Institute of Metal Research, Chinese Academy of Sciences
DOI10.1016/j.jnucmat.2020.152407
Indexed BySCI
Language英语
Funding ProjectNational Key R&D Program of China[2017YFB0702103] ; National Science and Technology Major Project[2015ZX06002005] ; National Science and Technology Major Project[2017ZX06002003-004-002] ; National Natural Science Foundation of China[51671201] ; Chinese Academy of Sciences[ZDRW-CN-2017-1] ; Institute of Metal Research, Chinese Academy of Sciences[SCJJ-2013-ZD-02]
WOS Research AreaMaterials Science ; Nuclear Science & Technology
WOS SubjectMaterials Science, Multidisciplinary ; Nuclear Science & Technology
WOS IDWOS:000575165800008
PublisherELSEVIER
Citation statistics
Cited Times:2[WOS]   [WOS Record]     [Related Records in WOS]
Document Type期刊论文
Identifierhttp://ir.imr.ac.cn/handle/321006/140827
Collection中国科学院金属研究所
Corresponding AuthorWu, Xinqiang
Affiliation1.Chinese Acad Sci, Inst Met Res, CAS Key Lab Nucl Mat & Safety Assessment, Liaoning Key Lab Safety & Assessment Tech Nucl Ma, Shenyang 110016, Peoples R China
2.State Nucl Power Plant Serv Co, Shanghai 200233, Peoples R China
Recommended Citation
GB/T 7714
Tan, Jibo,Zhang, Ziyu,Zheng, Hui,et al. Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants[J]. JOURNAL OF NUCLEAR MATERIALS,2020,541:10.
APA Tan, Jibo.,Zhang, Ziyu.,Zheng, Hui.,Wang, Xiang.,Gao, Jun.,...&Huang, Pengtao.(2020).Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants.JOURNAL OF NUCLEAR MATERIALS,541,10.
MLA Tan, Jibo,et al."Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants".JOURNAL OF NUCLEAR MATERIALS 541(2020):10.
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