316L奥氏体不锈钢在高温水中的应力腐蚀
Stress Corrosion of Austenitic Stainless Steel 316L in High Temperature Water
摘 要
为了得到裂纹尖端应力强度因子（K）、塑性变形量及水化学对316L奥氏体不锈钢在核电厂高温高压水中应力腐蚀开裂（SCC）的影响规律，采用dK/dt和dK/da两种不同的降K方式，在含氧与含氢高温水中，研究了冷变形316L不锈钢应力腐蚀裂纹扩展速率（CGR）与K值的关系。结果表明：在高变形量和苛刻的腐蚀环境条件下，316L不锈钢的应力腐蚀CGR对K值的依赖性降低；同时，降K加载过程得到的CGR比恒K加载时的低。
应力强度因子
应力腐蚀
变形量
裂纹扩展速率
316L stainless steel
stress intensity factor
stress corrosion cracking
deformation level
crack growth rate (CGR)
Abstract
In order to get the effects of stress intensity factor at crack tip (K), plastic deformation and hydrochemistry on the stress corrosion cracking (SCC) of austenitic stainless steel 316L in high temperature and pressure water of nuclear power plant, the relationship between stress corrosion crack growth rate (CGR) and K value of cold deformed 316L stainless steel in high temperature water containing oxygen and hydrogen were investigated using two K decreasing methods of dK/dt and dK/da. The results show that in high deformation and harsh corrosive environment, the dependence of stress corrosion CGR on the K value of 316L stainless steel decreased. At the same time, the CGR tested in the descending K loading process was lower than that tested in the constant K loading process.
中图分类号 TL341 DOI 10.11973/fsyfh-201804006
所属栏目 试验研究
基金项目
收稿日期 2016/7/28
修改稿日期
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引用该论文: DU Donghai,LAI Ping,ZHANG Hao,CHEN Kai,WANG Jiamei,GUO Xianglong,ZHANG Lefu. Stress Corrosion of Austenitic Stainless Steel 316L in High Temperature Water[J]. Corrosion & Protection, 2018, 39(4): 274
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参考文献
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【2】ZINKLE S J,WAS G S. Materials challenges in nuclear energy[J]. Acta Materialia,2013,61(3):735-758.
【3】SCOTT P M. An overview of materials degradation by stress corrosion in PWRs[M]//Corrosion Issues in Light Water Reactors. Cambridge,England:Woodhead Publishing,2007:3-24.
【4】杜东海,沈朝,陈凯,等. 冷变形316L不锈钢在高温高压水中的应力腐蚀[J]. 工程科学学报,2015(2):196-203.
【5】杜东海,余论,陈凯,等. 直流电压降法应力腐蚀裂纹扩展速率在线测定试验系统[J]. 理化检验(物理分册),2014(1):25-30.
【6】杜东海,陆辉,陈凯,等. 溶解氧对高温水中冷变形316L应力腐蚀开裂的影响规律[J]. 上海交通大学学报,2014(11):1644-1649.
【7】ANDRESEN P L. Emerging issues and fundamental processes in environmental cracking in hot water[J]. Corrosion,2008,64(5):439-464.
【8】ANDRESEN P L. K/Size effects on SCC in irradiated,cold worked and unirradiated stainless steel[C]//Proceedings of 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Chicago,Illinois,USA:ANS,2003:132-138.
【9】ANDRESEN P L. Unusual cold work and strain rate effects on SCC[C]//Proceedings of the 14th Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Chicago,Illinois,USA:ANS,2009:79-86.
【10】LU Z,SHOJI T,TAKEDA Y,et al. Effects of loading mode and water chemistry on stress corrosion crack growth behavior of 316L HAZ and weld metal materials in high temperature pure water[J]. Corrosion Science,2008,50(3):625-638.
【11】LU Z,XUE H,SHOJI T. Crack branching and its effect on environmentally assisted cracking in high temperature water environments[C]//Proceedings of ASME 2010 Pressure Vessels and Piping Division/K-PVP Conference.[S.l.]:American Society of Mechanical Engineers,2010:917-927.
【12】MAGDOWSKI R M,UGGOWITZER P J,SPEIDEL M O. The effect of crack branching on the residual lifetime of machine components containing stress corrosion cracks[J]. Corrosion Science,1985,25(8/9):745-756.
【13】WILSON W K,CHEREPKO J. Analysis of cracks with multiple branches[J]. International Journal of Fracture,1983,22(4):303-315.
【14】CODRINGTON J,KOTOUSOV A. Crack growth retardation following the application of an overload cycle using a strip-yield model[J]. Engineering Fracture Mechanics,2009,76(11):1667-1682.
【15】ANDRESEN P L,MORRA M M. Effects of positive and negative dK/da on SCC growth rates[C]//Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power System-Water Reactors. Chicago,Illinois,USA:ANS,2005:167-183.
【16】TERACHI T,YAMADA T,MIYAMOTO T,et al. SCC growth behaviors of austenitic stainless steels in simulated PWR primary water[J]. Journal of Nuclear Materials,2012,426(1/3):59-70.
【17】ANDRESEN P L,YOUNG L M. Characterization of the roles of electrochemistry,convection and crack chemistry in stress corrosion cracking[C]//7th International Symposium on Environment Degradation of Materials in Nuclear Power System:Water Reactors. Chicago, Illinios, USA:ANS,1995:664-678.
【18】刘伟,李志斌,王翔,等. 应变速率对奥氏体不锈钢应变诱发α'-马氏体转变和力学行为的影响[J]. 金属学报,2009,45(3):285-291.
【19】陈海云,盛水平. S31608奥氏体不锈钢应变强化诱发马氏体相变实验[J]. 压力容器,2013(8):7-14.
【20】SÁEZ-MADERUELO A,CASTRO L,DE DIEGO G. Plastic strain characterization in austenitic stainless steels and nickel alloys by electron backscatter diffraction[J]. Journal of Nuclear Materials,2011,416(1/2):75-79.
【21】WAS G S,FARKAS D,ROBERTSON I M. Micromechanics of dislocation channeling in intergranular stress corrosion crack nucleation[J]. Current Opinion in Solid State and Materials Science,2012,16(3):134-142.
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