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Establishment of fretting maps of Zr alloy cladding tube mated with Zr alloy dimple in simulated primary water of pressurized water reactor
期刊论文
TRIBOLOGY INTERNATIONAL, 2023, 卷号: 178, 页码: 13
Authors:
Zhang, Yusheng
;
Lai, Jiang
;
Wang, Jiazhen
;
Gao, Lixia
;
Ming, Hongliang
;
Wang, Jianqiu
;
Han, En-Hou
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View/Download:6/0
  |  
Submit date:2023/05/09
Fretting wear
Running condition fretting map
Material response fretting map
Zr alloy cladding tube
High temperature pressurised water
Effect of Normal Force on Fretting Wear Behavior of Zirconium Alloy Tube in Simulated Primary Water of PWR
期刊论文
ACTA METALLURGICA SINICA-ENGLISH LETTERS, 2022, 页码: 16
Authors:
Zhang, Yusheng
;
Lai, Jiang
;
Ming, Hongliang
;
Gao, Lixia
;
Wang, Jianqiu
;
Han, En-Hou
Favorite
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View/Download:10/0
  |  
Submit date:2023/05/09
Zirconium alloy tube
Fretting wear
Grid-to-rod
High-temperature pressurized water
Normal force
Effect of the frequency on fretting corrosion behavior between Alloy 690TT tube and 405 stainless steel plate in high temperature pressurized water
期刊论文
TRIBOLOGY INTERNATIONAL, 2021, 卷号: 164, 页码: 16
Authors:
Zhang, Yusheng
;
Ming, Hongliang
;
Tang, Lichen
;
Wang, Jianqiu
;
Qian, Hao
;
Han, En-Hou
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View/Download:78/0
  |  
Submit date:2021/11/22
Fretting corrosion
Fretting frequency
Alloy 690TT
High temperature pressurized water
Environmentally assisted cracking in the fusion boundary region of a SA508-Alloy 52M dissimilar weld joint in simulated primary pressurized water reactor environments
期刊论文
CORROSION SCIENCE, 2021, 卷号: 190, 页码: 13
Authors:
Dong, Lijin
;
Zhang, Yan
;
Han, Yaolei
;
Peng, Qunjia
;
Han, En-Hou
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View/Download:66/0
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Submit date:2021/10/15
A
Nickel
A
Low alloy steel
B
TEM
C
Stress corrosion
C
Corrosion fatigue
C
Welding
Grain boundary oxidation of proton-irradiated nuclear grade stainless steel in simulated primary water of pressurized water reactor (vol 11, 1371, 2021)
期刊论文
SCIENTIFIC REPORTS, 2021, 卷号: 11, 期号: 1, 页码: 1
Authors:
Deng, Ping
;
Peng, Qunjia
;
Han, En-Hou
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Submit date:2021/10/15
Grain boundary oxidation of proton-irradiated nuclear grade stainless steel in simulated primary water of pressurized water reactor
期刊论文
SCIENTIFIC REPORTS, 2021, 卷号: 11, 期号: 1, 页码: 9
Authors:
Deng, Ping
;
Peng, Qunjia
;
Han, En-Hou
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Submit date:2021/10/15
Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants
期刊论文
JOURNAL OF NUCLEAR MATERIALS, 2020, 卷号: 541, 页码: 10
Authors:
Tan, Jibo
;
Zhang, Ziyu
;
Zheng, Hui
;
Wang, Xiang
;
Gao, Jun
;
Wu, Xinqiang
;
Han, En-Hou
;
Yang, Shuangliang
;
Huang, Pengtao
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View/Download:102/0
  |  
Submit date:2021/02/02
Corrosion fatigue
Austenitic stainless steel
High temperature corrosion
Model
Corrosion fatigue model of austenitic stainless steels used in pressurized water reactor nuclear power plants
期刊论文
JOURNAL OF NUCLEAR MATERIALS, 2020, 卷号: 541, 页码: 10
Authors:
Tan, Jibo
;
Zhang, Ziyu
;
Zheng, Hui
;
Wang, Xiang
;
Gao, Jun
;
Wu, Xinqiang
;
Han, En-Hou
;
Yang, Shuangliang
;
Huang, Pengtao
Favorite
  |  
View/Download:119/0
  |  
Submit date:2021/02/02
Corrosion fatigue
Austenitic stainless steel
High temperature corrosion
Model
Electrochemistry and In Situ Scratch Behavior of 690 Alloy in Simulated Nuclear Power High Temperature High Pressure Water
期刊论文
ACTA METALLURGICA SINICA, 2020, 卷号: 56, 期号: 11, 页码: 1474-1484
Authors:
Li Xiaohui
;
Wang Jianqiu
;
Han En-Hou
;
Guo Yanjun
;
Zheng Hui
;
Yang Shuangliang
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  |  
View/Download:110/0
  |  
Submit date:2021/02/02
alloy 690
high temperature high pressure water
corrosion
in situ scratch
Electrochemistry and In Situ Scratch Behavior of 690 Alloy in Simulated Nuclear Power High Temperature High Pressure Water
期刊论文
ACTA METALLURGICA SINICA, 2020, 卷号: 56, 期号: 11, 页码: 1474-1484
Authors:
Li Xiaohui
;
Wang Jianqiu
;
Han En-Hou
;
Guo Yanjun
;
Zheng Hui
;
Yang Shuangliang
Favorite
  |  
View/Download:106/0
  |  
Submit date:2021/02/02
alloy 690
high temperature high pressure water
corrosion
in situ scratch