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浏览/检索结果: 共19条，第1-10条 帮助 已选(0)清除 条数/页：5101520253035404550556065707580859095100   排序方式：请选择作者升序作者降序题名升序题名降序发表日期升序发表日期降序提交时间升序提交时间降序 Grain boundary engineering for improving stress corrosion cracking of 304 stainless steel 期刊论文 MATERIALS SCIENCE AND TECHNOLOGY, 2019, 卷号: 35, 页码: 477-487 作者:  Liu, Tingguang[1];  Bai, Qin[2];  Ru, Xiangkun[3];  Xia, Shuang[4];  Zhong, Xiangyu[5] 收藏  |  浏览/下载：6/0  |  提交时间：2019/04/22 Stainless steel  grain boundary engineering  slow strain rate testing  twin boundary  stress corrosion cracking   Twins and twin-related domains in a grain boundary-engineered 304 stainless steel 期刊论文 MATERIALS SCIENCE AND TECHNOLOGY, 2018, 卷号: 34, 页码: 561-571 作者:  Liu, Tingguang[1];  Xia, Shuang[2];  Ru, Xiangkun[3];  Bai, Qin[4];  Zhou, Bangxin[5] 收藏  |  浏览/下载：15/0  |  提交时间：2019/04/24 Stainless steel  grain boundary engineering  EBSD  twins  twin-related domain  three-dimensional  grain boundary network  twin-chain   Characterization of microstructure of A508III/309L/308L weld and oxide films formed in deaerated high-temperature water 期刊论文 JOURNAL OF NUCLEAR MATERIALS, 2018, 卷号: 498, 页码: 227-240 作者:  Xiong, Qi[1];  Li, Hongjuan[2];  Lu, Zhanpeng[3];  Chen, Junjie[4];  Xiao, Qian[5] 收藏  |  浏览/下载：8/0  |  提交时间：2019/04/24 Microstructure  Ferrite  Low alloy steel  Stainless steel  Oxide film  High temperature water   Distribution and Characteristics of Oxide Films Formed on Stainless Steel Cladding on Low Alloy Steel in Simulated PWR Primary Water Environments 会议论文 PROCEEDINGS OF THE 18TH INTERNATIONAL CONFERENCE ON ENVIRONMENTAL DEGRADATION OF MATERIALS IN NUCLEAR POWER SYSTEMS - WATER REACTORS, VOL 2, 2018-01-01 作者:  Xiong, Qi[1];  Li, Hongjuan[2];  Lu, Zhanpeng[3];  Chen, Junjie[4];  Xiao, Qian[5] 收藏  |  浏览/下载：11/0  |  提交时间：2019/04/22 Stainless steel cladding  Pressurized water reactor  Low alloy steel  Oxide film  High temperature water   Properties of Oxide Films on Ni-Cr-xFe Alloys in a Simulated PWR Water Environment 会议论文 PROCEEDINGS OF THE 18TH INTERNATIONAL CONFERENCE ON ENVIRONMENTAL DEGRADATION OF MATERIALS IN NUCLEAR POWER SYSTEMS - WATER REACTORS, VOL 2, 2018-01-01 作者:  Ru, Xiangkun[1];  Lu, Zhanpeng[2];  Chen, Junjie[3];  Han, Guangdong[4];  Zhang, Jinlong[5] 收藏  |  浏览/下载：17/0  |  提交时间：2019/04/22 Nickel-based alloys  Simulated PWR water  TEM  Iron content  Oxide film  Local oxidation penetration   Effects of iron content in Ni-Cr-Fe alloys on the oxide films formed in an oxygenated simulated PWR water environment 期刊论文 JOURNAL OF NUCLEAR MATERIALS, 2018, 卷号: 509, 页码: 29-42 作者:  Ru, Xiangkun[1];  Ma, Jiarong[2];  Lu, Zhanpeng[3];  Chen, Junjie[4];  Han, Guangdong[5] 收藏  |  浏览/下载：19/0  |  提交时间：2019/04/22 Nickel-based alloys  Simulated PWR water  TEM  Iron content  Oxide film   MICROSTRUCTURE ANALYSIS OF STAINLESS STEEL CLADDING ON LOW ALLOY STEEL WELD IN A PWR SAFE END NOZZLE AND CORROSION IN HIGH TEMPERATURE WATER 会议论文 PROCEEDINGS OF THE 25TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2017, VOL 2, 2017-01-01 作者:  Xiong, Qi[1];  Lu, Zhanpeng[2];  Chen, Junjie[3];  Ru, Xiangkun[4];  Li, Hongjuan[5] 收藏  |  浏览/下载：7/0  |  提交时间：2019/04/24 DATA COMPILING AND ANALYSIS OF OXIDE FILM GROWTH ON NICKEL-BASE ALLOYS IN HIGH TEMPERTURE WATER 会议论文 PROCEEDINGS OF THE 25TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2017, VOL 2, 2017-01-01 作者:  Ru, Xiangkun[1];  Lu, Zhanpeng[2];  Chen, Junjie[3];  Peng, Hao[4];  Xiong, Qi[5] 收藏  |  浏览/下载：20/0  |  提交时间：2019/04/24 EFFECTS OF CHARGED HYDROGEN ON OXIDATION BEHAVIOR OF ALLOY 690 IN HIGH TEMPERATURE WATER ENVIRONMENTS 会议论文 PROCEEDINGS OF THE 25TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2017, VOL 2, 2017-01-01 作者:  Peng, Hao[1];  Lu, Zhanpeng[2];  Ru, Xiangkun[3];  Chen, Junjie[4];  Ma, Jiarong[5] 收藏  |  浏览/下载：7/0  |  提交时间：2019/04/24 Retarding effect of prior-overloading on stress corrosion cracking of cold rolled 316L SS in simulated PWR water environment 期刊论文 JOURNAL OF NUCLEAR MATERIALS, 2017, 卷号: 496, 页码: 313-324 作者:  Chen, Junjie[1];  Lu, Zhanpeng[2];  Xiao, Qian[3];  Ru, Xiangkun[4];  Ma, Jiarong[5] 收藏  |  浏览/下载：8/0  |  提交时间：2019/04/24 Stress corrosion cracking  316L stainless steel  Pressurized water reactor  Single tensile overloading  Cold rolling