Stress corrosion cracking behavior of cold-deformed 316 stainless steel in high temperature water

被引:0
作者
Du, Dong-Hai [1 ]
Lu, Hui [1 ]
Chen, Kai [1 ]
Zhang, Le-Fu [1 ]
Shi, Xiu-Qiang [2 ]
Xu, Xue-Lian [2 ]
机构
[1] Corrosion Laboratory for Nuclear Power Materials, Shanghai Jiao Tong University, Shanghai
[2] Shanghai Key Laboratory for Nuclear Power Engineering, Shanghai
来源
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | 2015年 / 49卷 / 11期
关键词
Cold deformation; Dissolved oxygen; Stress corrosion cracking;
D O I
10.7538/yzk.2015.49.11.1977
中图分类号
学科分类号
摘要
The stress corrosion cracking (SCC) behaviors of different deformations of cold-deformed nuclear grade 316SS and 316L SS in high temperature water were studied. The effects of dissolved oxygen, Cl- and temperature on the crack growth rate were analyzed detailedly. Test results show that dissolved oxygen and Cl- can significantly increase the crack growth rate of the material. Moreover, the crack growth rates are faster at 325℃ than that at 288℃ when water chemical conditions are the same. © 2015, Editorial Board of Atomic Energy Science and Technology. All right reserved.
引用
收藏
页码:1977 / 1983
页数:6
相关论文
共 16 条
[1]  
Ford F.P., Andresen P.L., Development and use of a predictive model of crack propagation in 304/316L, A533B/A508 and Inconel 600/182 alloys in 288℃ water, Proceedings of the Third International Symposium on Environmental Degradation of Materials in Nuclear Power Systems, (1988)
[2]  
Andresen P.L., Ford F.P., Life prediction by mechanistic modelling and system monitoring of environmental cracking of Fe and Ni alloys in aqueous systems, Materials Science and Engineering, 103, 1, pp. 167-184, (1988)
[3]  
Ford F.P., Taylor D.F., Andresen P.L., Et al., Corrosion-assisted cracking of stainless and low-alloy steels in LWR environments, (1987)
[4]  
Andresen P.L., Perspective and direction of stress corrosion cracking in hot water, Proceedings of Tenth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, (2001)
[5]  
Andresen P.L., Modeling of water and material chemistry effects on crack tip chemistry and resulting crack growth kinetics, Proceedings of the Third International Symposium on Environmental Degradation of Materials in Nuclear Power System, (1988)
[6]  
Andresen P.L., Young L.M., Characterization of the roles of electrochemistry, convection and crack chemistry in stress corrosion cracking, Proceedings of the 7th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, pp. 579-596, (1995)
[7]  
Andresen P.L., Conceptual similarities and common predictive approaches for SCC in high temperature water systems, (1996)
[8]  
Andresen P.L., Emigh P.W., Morra M.M., Et al., Effects of PWR primary water chemistry and deaerated water on SCC, Proceedings of the 12th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, (2005)
[9]  
Shoji T., Progress in mechanistic understanding of BWR SCC and its implication to prediction of SCC growth behavior in plant, Proceeding of 11th International Conference on Environmental Degradation of Materials in Nuclear Systems Held at Stevenson, (2003)
[10]  
Lu Z., Shoji T., Meng F., Et al., Characterization of microstructure and local deformation in 316NG weld heat-affected zone and stress corrosion cracking in high temperature water, Corrosion Science, 53, 5, pp. 1916-1932, (2011)