晶界碳化物和冷变形对600合金应力腐蚀开裂的影响规律
Effect of Grain Boundary Carbides and Cold Work on Stress Corrosion Cracking of Alloy 600
摘 要
采用直流电位降(DCPD)在线监测方法研究了晶界碳化物和冷变形对600合金在高温水环境中应力腐蚀开裂(SCC)裂纹扩展速率的影响规律，并结合高分辨微观表征技术阐明了作用机理。结果表明:无冷变形时，晶界碳化物能通过阻挡裂纹向基体内部扩展和抑制晶界氧化而有效降低600合金的SCC敏感性；冷变形后，晶界残余应变和晶界碳化物周围产生的局部应变集中，促进了氧元素扩散、加速晶界氧化，进而加速裂纹扩展。
碳化物
冷变形
应力腐蚀开裂
裂纹扩展速率
alloy 600
carbide
cold work
stress corrosion cracking
crack growth rate
Abstract
The stress corrosion cracking (SCC) behavior of alloy 600 in simulated pressurized water reactor (PWR) high-temperature water environment was investigated by direct current potential drop (DCPD) on-line monitoring. Analytical electron microscopy was utilized to characterize the cracking process to better understand the cold work and grain boundary (GB) carbides effect. Synergistic effect of GB carbides and cold work on SCC behavior was identified. A beneficial effect of carbides was identified in non-cold worked alloy 600, which was attributed to the physical impediment and reduced GB internal oxidation. In cold worked alloy 600, GB carbides had a detrimental effect due to the combination of the enhanced local strain at GBs and also enhanced GB internal oxidation.
中图分类号 TL341 DOI 10.11973/fsyfh-202204001
所属栏目 试验研究
基金项目 国家重点研发项目（2017YFB0702203；YS2018YFE010246）；自然科学基金（51871153）
收稿日期 2020/6/12
修改稿日期
网络出版日期
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引用该论文: WANG Jiamei,SU Haozhan,CHEN Kai,GUO Xianglong,ZHANG Lefu,WANG Yuanhua,MA Wujiang. Effect of Grain Boundary Carbides and Cold Work on Stress Corrosion Cracking of Alloy 600[J]. Corrosion & Protection, 2022, 43(4): 1
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参考文献
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【2】STAEHLE R W,GORMAN J A. Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors:Part 1[J]. Corrosion,2003,59(11):931-994.
【3】ZHAO L,CHU F,LUO K,et al. Failure analysis of branch connection on the reactor primary pipeline,Eng. Fail. Anal.,2018(85):137-148.
【4】FÉRON D,GUERRE C,HERMS E,et al. Chapter 9-Stress corrosion cracking of Alloy 600:overviews and experimental techniques,stress corrosion cracking of nickel based alloys in water-cooled nuclear reactors[M].[s.n.],Woodhead Publishing,2016:325-353.
【5】ARIOKA K. 2014 WR Whitney award lecture:change in bonding strength at grain boundaries before long-term SCC initiation[J]. Corrosion,2014,71:403-419.
【6】ECONOMY G,JACKO R J,PEMENT F W. IGSCC behavior of alloy 600 steam generator tubing in water or steam tests above 360℃[J]. Corrosion,1987,43:727-734.
【7】BULISCHECK T S,ROOYEN D V. Stress corrosion cracking of alloy 600 using the constant strain rate test[J]. Corrosion,1981,37:597-607.
【8】ARIOKA K,YAMADA T,MIYAMOTO T,et al. SCC initiation of CW alloy 690TT and alloy 600 in PWR water[C]//Proc. 17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors Toronto,Canada:CNS,2015.
【9】ARIOKA K. Change in bonding strength at grain boundaries before long-term SCC initiation[J]. Corrosion,2014,71(4):403-419.
【10】SHEN Z,KARAMCHED P,ARIOKA K,et al. Observation and quantification of the diffusion-induced grain boundary migration ahead of SCC crack tips[J]. Corrosion Science,2019,147:163-168.
【11】PERSAUD S Y,KORINEK A,HUANG J,et al. Internal oxidation of Alloy 600 exposed to hydrogenated steam and the beneficial effects of thermal treatment[J]. Corrosion Science,2014,86:108-122.
【12】杜东海,陆辉,陈凯,等. 冷变形316不锈钢在高温水中的应力腐蚀开裂行为[J]. 原子能科学技术,2015,49(11):1977-1983.
【13】CHEN K,WANG J,DU D,et al. dK/da effects on the SCC growth rates of nickel base alloys in high-temperature water[J]. Journal of Nuclear Materials,2018,503:13-21.
【14】KIM Y J,ANDRESEN P L. Data quality,issues,and guidelines for electrochemical corrosion potential measurement in high-temperature water[J]. Corrosion,2003,59(7):584-596.
【15】ANDRESEN P L,HICKLING J,AHLUWALIA A,et al. Effects of hydrogen on stress corrosion crack growth rate of nickel alloys in high-temperature water[J]. Corrosion,2008,64(9):707-720.
【16】ANDRESEN P L,HICKLING J,AHLUWALIA A. Mitigation of PWSCC in nickel-base alloys by optimizing H2[J]. EPRI Final Report,2007,1015288.
【17】ANDRESEN P,CHOU P. Crack initiation of alloy 600 in PWR water[C]//Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Springer:Cham,2019:121-135.
【18】ZHAI Z,TOLACZKO M,BRUEMMER S. Microstructural effects on SCC initiation in PWR primary water for cold-worked alloy 600[C]//Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Springer:Cham,2019:217-229.
【19】HOU J,WANG J Q,KE W,et al. Effect of mono-tension and biaxial tension on microstructure and stress corrosion cracking of alloy 690TT[J]. Materials Science and Engineering:A,2009,518(1/2):19-26.
【20】张子龙,夏爽,曹伟,等. 晶界特征对316不锈钢沿晶应力腐蚀开裂裂纹萌生的影响[J]. 金属学报,2016,52(3):313-319.
【21】CASSAGNE T,GELPI A. Crack growth rate measurements on alloy 600 steam generator tubing in primary and hydrogenated AVT water[C]//Proceedings of the Sixth Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors.[S.l.]:[s.n.],1993:679-685.
【22】REBAK R B,SZKLARSKA-SMIALOWSKA Z MCLLREE A R. Effects of pH and stress intensity on crack growth rate in Alloy 600 in lithiated + borated water at high temperatures[C]//Proceedings of the 5th Intl. Symp. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. La Grange Park:American Nuclear Society,1991:511-517.
【23】KIM H P,CHOI M J,KIM S W,et al. Effects of grain boundary morphologies on stress corrosion cracking of alloy 600[J]. Archives of Metallurgy and Materials,2017,62(2):1415-1419.
【24】SHEN Z,ARIOKA K,LOZANO-PEREZ S. A mechanistic study of SCC in Alloy 600 through high-resolution characterization[J]. Corrosion Science,2018,132:244-259.
【25】SHEN Z,MEISNAR M,ARIOKA K,et al. Mechanistic understanding of the temperature dependence of crack growth rate in alloy 600 and 316 stainless steel through high-resolution characterization[J]. Acta Materialia,2019,165:73-86.
【26】DONG L,PENG Q,HAN E H,et al. Stress corrosion cracking in the heat affected zone of a stainless steel 308L-316L weld joint in primary water[J]. Corrosion science,2016,107:172-181.
【27】BRUEMMER S M,THOMAS L E. Comparison of IGSCC crack-tip characteristics produced in BWR oxidizing water and BWR hydrogen water chemistry conditions[C]//Proceedings of 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors,[S.l.]:[s.n.],2007.
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