Corrosion fatigue crack growth behavior of 316NG heat affected zone in simulated pressurized water reactor environment

Xiao Jun; Xiao Ting; Zheng Yuqi; Wang Hao; Chen Yong; Qiu Shaoyu; Gong Xing

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Corrosion fatigue crack growth behavior of 316NG heat affected zone in simulated pressurized water reactor environment

机译：模拟加压水反应器环境中316NG热影响区的腐蚀疲劳裂纹生长行为

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来源
《Journal of Nuclear Materials: Materials Aspects of Fission and Fusion》 |2020年第1期|共11页
作者
Xiao Jun; Xiao Ting; Zheng Yuqi; Wang Hao; Chen Yong; Qiu Shaoyu; Gong Xing;
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作者单位

Nucl Power Inst China Sci &

Technol Reactor Fuel &

Mat Lab Chengdu 610213 Peoples R China;

Nucl Power Inst China Sci &

Technol Reactor Fuel &

Mat Lab Chengdu 610213 Peoples R China;

Nucl Power Inst China Sci &

Technol Reactor Fuel &

Mat Lab Chengdu 610213 Peoples R China;

Nucl Power Inst China Sci &

Technol Reactor Fuel &

Mat Lab Chengdu 610213 Peoples R China;

Nucl Power Inst China Sci &

Technol Reactor Fuel &

Mat Lab Chengdu 610213 Peoples R China;

Nucl Power Inst China Sci &

Technol Reactor Fuel &

Mat Lab Chengdu 610213 Peoples R China;

Shenzhen Univ Coll Phys &

Optoelect Engn Adv Nucl Energy Res Team Shenzhen 518060 Peoples R China;

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正文语种 eng
中图分类原子能技术;
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1. Corrosion fatigue crack growth behavior of 316NG heat affected zone in simulated pressurized water reactor environment [J] . Xiao Jun, Xiao Ting, Zheng Yuqi, Journal of Nuclear Materials: Materials Aspects of Fission and Fusion . 2020,第1期

机译：模拟加压水反应器环境中316NG热影响区的腐蚀疲劳裂纹生长行为
2. Synergistic effects of local strain-hardening and dissolved oxygen on stress corrosion cracking of 316NG weld heat-affected zones in simulated BWR environments [J] . Lu Z., Shoji T., Xue H., Journal of Nuclear Materials: Materials Aspects of Fission and Fusion . 2012,第1a3期

机译：模拟BWR环境中局部应变硬化和溶解氧对316NG焊接热影响区应力腐蚀开裂的协同效应
3. Effects of crack tip plastic zone on corrosion fatigue cracking of alloy 690(TT) in pressurized water reactor environments [J] . Xiao J., Qiu S. Y., Chen Y., Journal of Nuclear Materials: Materials Aspects of Fission and Fusion . 2015,第Null期

机译：压水堆环境中裂纹尖端塑性区对690（TT）合金腐蚀疲劳开裂的影响
4. FATIGUE CRACK GROWTH BEHAVIOR OF SENSITIZED TYPE 304 STAINLESS STEEL IN SIMULATED BOILING WATER REACTOR ENVIRONMENTS [C] . 14th Asian-Pacific Corrosion Control Conference论文集 . 2006

机译：模拟沸水反应器环境中敏化304不锈钢的疲劳裂纹扩展行为
5. The Role of Deformation and Microchemistry in the Corrosion Processes of Type 304 Stainless Steel in Simulated Pressurized Water Reactor Environments [D] . Fisher, Kevin B. 2017

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6. Stress-corrosion crack growth of Si-Na-K-Mg-Ca-P-O bioactive glasses in simulated human physiological environment [O] . D. R. Bloyer, J. M. McNaney, R. M. Cannon, -1

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7. Corrosion fatigue studies in a nuclear pressure vessel steel in simulated pressurized water reactor environments [O] . Hicks Peter David 2015

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8. The Stress Corrosion Crack Growth Rate of Alloy 600 Heat Affected Zones Exposed to High Purity Water [R] . 2003

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Corrosion fatigue crack growth behavior of 316NG heat affected zone in simulated pressurized water reactor environment

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