反应堆压力容器用508Ⅲ低合金钢在高温高压硼锂水环境中的应力腐蚀开裂行为
Stress Corrosion Cracking Behavior of 508Ⅲ LAS for RPV in High Temperature and High Pressure B/Li Water Environment
摘 要
通过直流电压降法在线测量了反应堆压力容器用508Ⅲ低合金钢在模拟压水堆高温高压硼锂水环境中应力腐蚀开裂的裂纹扩展曲线，评估了其在含氧含氯离子环境中的应力腐蚀开裂敏感性。结果表明：在较低的恒载荷条件下，508Ⅲ低合金钢在含氧或除氧的300 ℃硼锂水环境中的裂纹扩展速率均低于BWRVIP-60 Line 1预测曲线，且对氧化性介质和低含量的杂质离子具有较高的应力腐蚀开裂容忍度，但腐蚀性介质或载荷的突变会造成裂纹扩展速率激增的现象；在含氧及除氧水环境中该合金的应力腐蚀开裂均由穿晶开裂主导，且裂纹尖端存在剧烈的金属溶解和氧化，大量的氧化物形成是造成裂纹尖端变“钝”，裂纹扩展速率降低的主要原因。
压水堆
应力腐蚀开裂
裂纹扩展速率
腐蚀电位
508Ⅲ low alloy steel (LAS)
pressurized water reactor (PWR) water
stress corrosion cracking
crack growth rate
corrosion potential
Abstract
The stress corrosion crack growth curves of 508Ⅲ low alloy steel (LAS) for reactor pressure vessel (RPV) in a high temperature and high pressure boron-lithium water environment simulating pressurized water reactor (PWR) were measured online by direct current potential drop method. The stress corrosion cracking susceptibility of the LAS in an oxygen-containing and chloride-containing environment was evaluated. The results show that under relatively low constant load conditions, the crack growth rate of the 508Ⅲ LAS was lower than the predicted curve of BWRVIP-60 Line 1 in a 300 ℃ boron-lithium water environment whether it contained oxygen or not. And the 508Ⅲ LAS had a high stress corrosion cracking tolerance to oxidizing media and low concentration impurity ions, but the sudden change of corrosive medium or load resulted in a rapid increase in crack growth rate. The stress corrosion cracking of the LAS in oxygen-containing or de-oxygenated water environment was dominated by transcrystalline cracking, and there was violent metal dissolution and oxidation at the crack tip. The formation of a large amount of oxide was the main reason for being “blunt” of crack tip and the decrease of crack growth rate.
中图分类号 TL341 DOI 10.11973/fsyfh-202108005
所属栏目 试验研究
基金项目 国家重点研发项目（2017YFB0702203；YS2018YFE010246）；国家自然科学基金（51871153）；浙江省科技计划项目（2018C37035）
收稿日期 2019/9/17
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引用该论文: WANG Jiamei,SU Haozhan,ZHANG Lefu,CHEN Kai,DU Donghai,XIONG Tuan,WANG Hui,SHEN Kanghua,WU Lihua. Stress Corrosion Cracking Behavior of 508Ⅲ LAS for RPV in High Temperature and High Pressure B/Li Water Environment[J]. Corrosion & Protection, 2021, 42(8): 27
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参考文献
【1】SEIFERT H P,RITTER S. Stress corrosion cracking of low-alloy reactor pressure vessel steels under boiling water reactor conditions[J]. Journal of Nuclear Materials,2008,372(1):114-131.
【2】SEIFERT H P,RITTER S,SHOJI T,et al. Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment[J]. Journal of Nuclear Materials,2008,378(2):197-210.
【3】SEIFERT H P,RITTER S. Corrosion fatigue crack growth behaviour of low-alloy reactor pressure vessel steels under boiling water reactor conditions[J]. Corrosion Science,2008,50(7):1884-1899.
【4】SEIFERT H P,HICKLING J,LISTER D. Corrosion and environmentally-assisted cracking of carbon and low-alloy steels[M]//Comprehensive Nuclear Materials. Amsterdam:Elsevier,2012:105-142.
【5】MATSUNAGA T,MATSUNAGA K. Stress corrosion cracking of CRD stub tube joint and repair at Hamaoka unit 1[C/CD]//The Proceedings of the 11th International Conference on Nuclear Engineering. Tokyo:The Japan Society of Mechanical Engineers,2003:ICONE 11-36056.
【6】YAMASHITA A. Stress corrosion cracking at shroud support Tsuruga unit 1[C/CD]//Proceedings of the 9th International Conference on Nuclear Engineering. Tokyo:The Japan Society of Mechanical Engineers,2001:ICONE 9-66.
【7】HORN R M,ANDRESEN P L,HICKLING J. BWR alloy 182 stress corrosion cracking experience[C/CD]//the Procedings of the 5th International Symposium on Contribution of Materials Investigation to the Resolution of Problems Encountered in PWR plants. [S.l.]:[s.n.],2002.
【8】SEIFERT H P,RITTER S,HICKLING J. Research and service experience with environmentally assisted cracking of low-alloy pressure-boundary components under LWR conditions[J]. Power Plant Chemistry,2004,6(2):111-123.
【9】SCOTT P M,TICE D R. Stress corrosion in low alloy steels[J]. Nuclear Engineering and Design,1990,119(2/3):399-413.
【10】HICKLING J,BLIND D. Strain-induced corrosion cracking of low-alloy steels in LWR systems—case histories and identification of conditions leading to susceptibility[J]. Nuclear Engineering and Design,1986,91(3):305-330.
【11】BRUEMMER S,FORD P,WAS G. Ninth international symposium on environmental degradation of materials in nuclear power systems—water reactors[M]. Hoboken,NJ,USA:John Wiley & Sons,Inc.,1999.
【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 M,DU D H,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】ZHANG L F,CHEN K,WANG J M,et al. Effects of zinc injection on stress corrosion cracking of cold worked austenitic stainless steel in high-temperature water environments[J]. Scripta Materialia,2017,140:50-54.
【15】CHEN K,WANG J M,DU D H,et al. Stress corrosion crack growth behavior of type 310S stainless steel in supercritical water[J]. Corrosion,2018,74(7):776-787.
【16】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.
【17】LIN C C,SMITH F R,ICHIKAWA N,et al. Electrochemical potential measurements under simulated BWR water chemistry conditions[J]. Corrosion,1992,48(1):16-28.
【18】DUAN Z G,ARJMAND F,ZHANG L F,et al. Investigation of the corrosion behavior of 304L and 316L stainless steels at high-temperature borated and lithiated water[J]. Journal of Nuclear Science and Technology,2016,53(9):1435-1446.
【19】WANG J M,SU H Z,CHEN K,et al. Effect of δ-ferrite on the stress corrosion cracking behavior of 321 stainless steel[J]. Corrosion Science,2019,158:108079.
【2】SEIFERT H P,RITTER S,SHOJI T,et al. Environmentally-assisted cracking behaviour in the transition region of an Alloy182/SA 508 Cl.2 dissimilar metal weld joint in simulated boiling water reactor normal water chemistry environment[J]. Journal of Nuclear Materials,2008,378(2):197-210.
【3】SEIFERT H P,RITTER S. Corrosion fatigue crack growth behaviour of low-alloy reactor pressure vessel steels under boiling water reactor conditions[J]. Corrosion Science,2008,50(7):1884-1899.
【4】SEIFERT H P,HICKLING J,LISTER D. Corrosion and environmentally-assisted cracking of carbon and low-alloy steels[M]//Comprehensive Nuclear Materials. Amsterdam:Elsevier,2012:105-142.
【5】MATSUNAGA T,MATSUNAGA K. Stress corrosion cracking of CRD stub tube joint and repair at Hamaoka unit 1[C/CD]//The Proceedings of the 11th International Conference on Nuclear Engineering. Tokyo:The Japan Society of Mechanical Engineers,2003:ICONE 11-36056.
【6】YAMASHITA A. Stress corrosion cracking at shroud support Tsuruga unit 1[C/CD]//Proceedings of the 9th International Conference on Nuclear Engineering. Tokyo:The Japan Society of Mechanical Engineers,2001:ICONE 9-66.
【7】HORN R M,ANDRESEN P L,HICKLING J. BWR alloy 182 stress corrosion cracking experience[C/CD]//the Procedings of the 5th International Symposium on Contribution of Materials Investigation to the Resolution of Problems Encountered in PWR plants. [S.l.]:[s.n.],2002.
【8】SEIFERT H P,RITTER S,HICKLING J. Research and service experience with environmentally assisted cracking of low-alloy pressure-boundary components under LWR conditions[J]. Power Plant Chemistry,2004,6(2):111-123.
【9】SCOTT P M,TICE D R. Stress corrosion in low alloy steels[J]. Nuclear Engineering and Design,1990,119(2/3):399-413.
【10】HICKLING J,BLIND D. Strain-induced corrosion cracking of low-alloy steels in LWR systems—case histories and identification of conditions leading to susceptibility[J]. Nuclear Engineering and Design,1986,91(3):305-330.
【11】BRUEMMER S,FORD P,WAS G. Ninth international symposium on environmental degradation of materials in nuclear power systems—water reactors[M]. Hoboken,NJ,USA:John Wiley & Sons,Inc.,1999.
【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 M,DU D H,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】ZHANG L F,CHEN K,WANG J M,et al. Effects of zinc injection on stress corrosion cracking of cold worked austenitic stainless steel in high-temperature water environments[J]. Scripta Materialia,2017,140:50-54.
【15】CHEN K,WANG J M,DU D H,et al. Stress corrosion crack growth behavior of type 310S stainless steel in supercritical water[J]. Corrosion,2018,74(7):776-787.
【16】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.
【17】LIN C C,SMITH F R,ICHIKAWA N,et al. Electrochemical potential measurements under simulated BWR water chemistry conditions[J]. Corrosion,1992,48(1):16-28.
【18】DUAN Z G,ARJMAND F,ZHANG L F,et al. Investigation of the corrosion behavior of 304L and 316L stainless steels at high-temperature borated and lithiated water[J]. Journal of Nuclear Science and Technology,2016,53(9):1435-1446.
【19】WANG J M,SU H Z,CHEN K,et al. Effect of δ-ferrite on the stress corrosion cracking behavior of 321 stainless steel[J]. Corrosion Science,2019,158:108079.
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