Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants

doi:10.1016/j.jnucmat.2020.152407

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	Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants
	Tan, Jibo 1; Zhang, Ziyu 1; Zheng, Hui 2; Wang, Xiang 1; Gao, Jun 1; Wu, Xinqiang 1; Han, En-Hou 1; Yang, Shuangliang 2; Huang, Pengtao 2
刊名	JOURNAL OF NUCLEAR MATERIALS
	2020-12-01
卷号	541 页码:10
关键词	Corrosion fatigue Austenitic stainless steel High temperature corrosion Model
ISSN号	0022-3115
DOI	10.1016/j.jnucmat.2020.152407
通讯作者	Wu, Xinqiang(xqwu@imr.ac.cn)
英文摘要	Low 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.
资助项目	National 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研究方向	Materials Science ; Nuclear Science & Technology
语种	英语
出版者	ELSEVIER
WOS记录号	WOS:000575165800008
资助机构	National 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
内容类型	期刊论文
源URL	[http://ir.imr.ac.cn/handle/321006/140826]
专题	金属研究所_中国科学院金属研究所
通讯作者	Wu, Xinqiang
作者单位	1.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
推荐引用方式 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.