核用304不锈钢辐照促进应力腐蚀开裂研究
|
摘要
采用2 MeV质子束在360℃对国产核用304不锈钢试样进行了辐照实验,利用高温高压水环境的慢应变速率拉伸实验(SSRT)和SEM、EBSD、TEM等研究了核用304不锈钢辐照促进应力腐蚀开裂(IASCC)机理。结果表明,慢应变速率拉伸过程中辐照促进材料晶界和表面滑移台阶处形成应变集中,且其程度随辐照剂量增加而增加。滑移台阶穿过或终止于晶界,终止于晶界的台阶造成晶界处产生不连续滑移,易将位错传输到晶界,在晶界区域形成位错塞积和残余应变集中。而台阶不连续滑移的形成则受毗邻晶粒的Schmidt因子对的类型影响。另一方面,辐照促进晶界发生贫Cr富Ni元素偏析,其偏析程度随辐照剂量增加而增加。SSRT实验后辐照试样表面发生明显的沿晶应力腐蚀开裂,且裂纹数量随辐照剂量和外加应变增加而增加。同时,裂纹尖端区域发生明显晶界腐蚀,且氧化物宽度和长度随辐照剂量增加而增加。分析认为,辐照致晶界应变集中和元素偏析的协同作用造成材料变形行为和晶界腐蚀行为变化是IASCC发生的关键因素。
Irradiation assisted stress corrosion cracking(IASCC) of austenitic stainless steel core components is one major concern for maintenance of nuclear power plants. Previous studies on the IASCC had mainly focused on the effect of irradiation on changes in deformation modes and interaction of dislocation channels with grain boundary. The role of corrosion in IASCC, however, has not received sufficient attentions. In the process of stress corrosion cracking(SCC), corrosion occurs simultaneously with localized deformation in the vicinity of the crack tip. This indicates that corrosion is one of the potential contributors to IASCC. In this work, IASCC of proton-irradiated nuclear grade 304 stainless steel(304 SS)was investigated. The IASCC tests were conducted by interrupted slow strain rate tensile(SSRT) tests at320 ℃ in simulated primary water of pressurized water reactor containing 1200 mg/L B as H_3BO_3 and2.3 mg/L Li as LiOH · H_2O, with a dissolved hydrogen concentration of 2.6 mg/L. Following the SSRT tests, the localized deformation, corrosion and IASCC of the specimens were characterized. The results revealed that increasing the irradiation dose promoted residual strain accumulation at slip steps and grain boundaries of nuclear grade 304 SS. Since the slip step usually transmitted or terminated at the grain boundary, it eventually promoted localized deformation at the grain boundary. Specially, the slip step transmitted at grain boundary led to slip continuity at the grain boundary. In contrast, a slip discontinuity was observed at the grain boundary where the slip step terminated, which caused a much higher strain accumulation by feeding dislocations to the grain boundary region. Further, formation of the slip discontinuity was related to the Schmidt factor pair type of the adjacent grains. The irradiation resulted in a depletion of Cr and an enrichment of Ni at grain boundary, while the magnitude of Cr depletion and Ni enrichment increased with increasing the irradiation dose. Following the SSRT tests, intergranular cracking was observed on surfaces of the irradiated specimens, while the number of the cracks was increased by a higher irradiation dose and applied strain. This suggested a higher IASCC susceptibility of nuclear grade304 SS in the primary water. Meanwhile, significant intergranular oxidation ahead of the crack tip was observed, while both the width and length of the oxide were larger at a higher irradiation dose. The synergic effect of irradiation-promoted deformation and intergranular corrosion was the primary cause for the IASCC of the irradiated steel.
Irradiation assisted stress corrosion cracking(IASCC) of austenitic stainless steel core components is one major concern for maintenance of nuclear power plants. Previous studies on the IASCC had mainly focused on the effect of irradiation on changes in deformation modes and interaction of dislocation channels with grain boundary. The role of corrosion in IASCC, however, has not received sufficient attentions. In the process of stress corrosion cracking(SCC), corrosion occurs simultaneously with localized deformation in the vicinity of the crack tip. This indicates that corrosion is one of the potential contributors to IASCC. In this work, IASCC of proton-irradiated nuclear grade 304 stainless steel(304 SS)was investigated. The IASCC tests were conducted by interrupted slow strain rate tensile(SSRT) tests at320 ℃ in simulated primary water of pressurized water reactor containing 1200 mg/L B as H_3BO_3 and2.3 mg/L Li as LiOH · H_2O, with a dissolved hydrogen concentration of 2.6 mg/L. Following the SSRT tests, the localized deformation, corrosion and IASCC of the specimens were characterized. The results revealed that increasing the irradiation dose promoted residual strain accumulation at slip steps and grain boundaries of nuclear grade 304 SS. Since the slip step usually transmitted or terminated at the grain boundary, it eventually promoted localized deformation at the grain boundary. Specially, the slip step transmitted at grain boundary led to slip continuity at the grain boundary. In contrast, a slip discontinuity was observed at the grain boundary where the slip step terminated, which caused a much higher strain accumulation by feeding dislocations to the grain boundary region. Further, formation of the slip discontinuity was related to the Schmidt factor pair type of the adjacent grains. The irradiation resulted in a depletion of Cr and an enrichment of Ni at grain boundary, while the magnitude of Cr depletion and Ni enrichment increased with increasing the irradiation dose. Following the SSRT tests, intergranular cracking was observed on surfaces of the irradiated specimens, while the number of the cracks was increased by a higher irradiation dose and applied strain. This suggested a higher IASCC susceptibility of nuclear grade304 SS in the primary water. Meanwhile, significant intergranular oxidation ahead of the crack tip was observed, while both the width and length of the oxide were larger at a higher irradiation dose. The synergic effect of irradiation-promoted deformation and intergranular corrosion was the primary cause for the IASCC of the irradiated steel.
引文
[1]Ge?rard R,Somville F.Situation of the baffle-former bolts in Bel‐gian units[A].Proceeding of the 17th International Conference on Nuclear Engineering[C].Brussels,Belgium:American Society of Mechanical Engineers,2009:521
[2]Andresen P L,Ford F P,Murphy S M,et al.State of knowledge of radiation effects on environmental cracking in light water reactor core materials[A].Proceeding of the 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Sys‐tems-Water Reactors[C].Houston:National Association of Cor‐rosion Engineers,1989:83
[3]Nishioka H,Fukuya K,Fujii K,et al.IASCC initiation in highly ir‐radiated stainless steels under uniaxial constant load conditions[J].J.Nucl.Sci.Technol.,2008,45:1072
[4]Zhou R S,West E A,Jiao Z J,et al.Irradiation-assisted stress cor‐rosion cracking of austenitic alloys in supercritical water[J].J.Nu‐cl.Mater.,2009,395:11
[5]Was G S,Bruemmer S M.Effects of irradiation on intergranular stress corrosion cracking[J].J.Nucl.Mater.,1994,216:326
[6]Stephenson K J,Was G S.Comparison of the microstructure,de‐formation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons[J].J.Nucl.Mater.,2015,456:85
[7]Jiao Z,Was G S.Impact of localized deformation on IASCC in austenitic stainless steels[J].J.Nucl.Mater.,2011,408:246
[8]Jiao Z,Was G S.Localized deformation and IASCC initiation in austenitic stainless steels[J].J.Nucl.Mater.,2008,382:203
[9]Jiao Z,Was G S,Busby J T.The role of localized deformation in IASCC of proton-irradiated austenitic stainless steels[A].Proceed‐ing of the 13th International Conference on Environmental Degra‐dation of Materials in Nuclear Power Systems-Water Reactors[C].Whister,British Columia,2007(CD-ROM)
[10]Jiao Z,Was G S.The role of irradiated microstructure in the local‐ized deformation of austenitic stainless steels[J].J.Nucl.Mater.,2010,407:34
[11]Was G S,Busby J T.Role of irradiated microstructure and micro‐chemistry in irradiation-assisted stress corrosion cracking[J].Phi‐los.Mag.,2005,85:443
[12]West E A,Was G S.Strain incompatibilities and their role in inter‐granular cracking of irradiated 316L stainless steel[J].J.Nucl.Ma‐ter.,2013,441:623
[13]McMurtrey M D,Was G S,Cui B,et al.Strain localization at dislo‐cation channel-grain boundary intersections in irradiated stainless steel[J].Int.J.Plast.,2014,56:219
[14]Was G S,Farkas D,Robertson I M.Micromechanics of dislocation channeling in intergranular stress corrosion crack nucleation[J].Curr.Opin.Solid State Mater.Sci.,2012,16:134
[15]Fukuya K,Nishioka H,Fujii K,et al.Charateriation of IASCCcrack tip in highly irradiated stainless steels[A].Proceeding of the14th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C].Virgin‐ia Beach,Virginia,2009:1248(CD-ROM)
[16]Thomas L E,Bruemmer S M.Microstructural and microchemical characterization of intergranular stress corrosion cracks in irradiat‐ed type 304SS removed from a BWR top guide[A].Proceeding of the 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C].Ste‐venson,WA,2003:1049(CD-ROM)
[17]Edwards D J,Thomas L E,Asano K,et al.Microstructure,micro‐chemistry and stress corrosion crack characteristics in a BWR316L SS core shroud weld[A].Proceeding of the 13th Internation‐al Conference on Environmental Degradation of Materials in Nu‐clear Power Systems-Water Reactors[C].Whister,British Colu‐mia,2007(CD-ROM)
[18]Thomas L E,Edwards D J,Asano K,et al.Crack-tip characteris‐tics in BWR service components[A].Proceeding of the 13th Inter‐national Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C].Whister,British Columia,2007(CD-ROM)
[19]Deng P,Peng Q J,Han E H,et al.Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary wa‐ter[J].Corros.Sci.,2017,127:91
[20]Kamaya M.Characterization of microstructural damage due to low-cycle fatigue by EBSD observation[J].Mater.Charact.,2009,60:1454
[21]Deng P,Peng Q J,Han E H,et al.Study of irradiation damage in domestically fabricated nuclear grade stainless steel[J].Acta Metall.Sin.,2017,53:1588(邓平,彭群家,韩恩厚等.国产核用不锈钢辐照损伤研究[J].金属学报,2017,53:1588)
[22]Shen Z,Arioka K,Lozano-Perez S.A mechanistic study of SCC in Alloy 600 through high-resolution characterization[J].Corros.Sci.,2018,132:244
[23]Lu Y H,Peng Q J,Sato T,et al.An ATEM study of oxidation be‐havior of SCC crack tips in 304L stainless steel in high tempera‐ture oxygenated water[J].J.Nucl.Mater.,2005,3477:52
[24]Lozano-Perez S,Kruska K,Iyengar I,et al.The role of cold work and applied stress on surface oxidation of 304 stainless steel[J].Corros.Sci.,2012,56:78
[25]Gertsman V Y,Bruemmer S M.Study of grain boundary character along intergranular stress corrosion crack paths in austenitic alloys[J].Acta Mater.,2001,49:1589
[26]Hu C L,Xia S,Li H,et al.Effect of grain boundary network on the intergranular stress corrosion cracking of 304 stainless steel[J].Acta Metall.Sin.,2011,47:939(胡长亮,夏爽,李慧等.晶界网络特征对304不锈钢晶间应力腐蚀开裂的影响[J].金属学报,2011,47:939)
[27]Ming H L,Zhang Z M,Wang J Q,et al.Microstructure and local properties of a domestic safe-end dissimilar metal weld joint by us‐ing hot-wire GTAW[J].Acta Metall.Sin.,2017,53:57(明洪亮,张志明,王俭秋等.国产核电安全端异种金属焊接件的微观结构及局部性能研究[J].金属学报,2017,53:57)
[28]Hou J,Peng Q J,Shoji T,et al.Effects of cold working path on strain concentration,grain boundary microstructure and stress cor‐rosion cracking in Alloy 600[J].Corros.Sci.,2011,53:2956
[29]Hou J,Peng Q J,Lu Z P,et al.Effects of cold working degrees on grain boundary characters and strain concentration at grain bound‐aries in Alloy 600[J].Corros.Sci.,2011,53:1137
[30]Alexandreanu B,Was G S.Grain boundary deformation-induced intergranular stress corrosion cracking of Ni-16Cr-9Fe in 360℃Water[J].Corrosion,2003,59:705
[2]Andresen P L,Ford F P,Murphy S M,et al.State of knowledge of radiation effects on environmental cracking in light water reactor core materials[A].Proceeding of the 4th International Symposium on Environmental Degradation of Materials in Nuclear Power Sys‐tems-Water Reactors[C].Houston:National Association of Cor‐rosion Engineers,1989:83
[3]Nishioka H,Fukuya K,Fujii K,et al.IASCC initiation in highly ir‐radiated stainless steels under uniaxial constant load conditions[J].J.Nucl.Sci.Technol.,2008,45:1072
[4]Zhou R S,West E A,Jiao Z J,et al.Irradiation-assisted stress cor‐rosion cracking of austenitic alloys in supercritical water[J].J.Nu‐cl.Mater.,2009,395:11
[5]Was G S,Bruemmer S M.Effects of irradiation on intergranular stress corrosion cracking[J].J.Nucl.Mater.,1994,216:326
[6]Stephenson K J,Was G S.Comparison of the microstructure,de‐formation and crack initiation behavior of austenitic stainless steel irradiated in-reactor or with protons[J].J.Nucl.Mater.,2015,456:85
[7]Jiao Z,Was G S.Impact of localized deformation on IASCC in austenitic stainless steels[J].J.Nucl.Mater.,2011,408:246
[8]Jiao Z,Was G S.Localized deformation and IASCC initiation in austenitic stainless steels[J].J.Nucl.Mater.,2008,382:203
[9]Jiao Z,Was G S,Busby J T.The role of localized deformation in IASCC of proton-irradiated austenitic stainless steels[A].Proceed‐ing of the 13th International Conference on Environmental Degra‐dation of Materials in Nuclear Power Systems-Water Reactors[C].Whister,British Columia,2007(CD-ROM)
[10]Jiao Z,Was G S.The role of irradiated microstructure in the local‐ized deformation of austenitic stainless steels[J].J.Nucl.Mater.,2010,407:34
[11]Was G S,Busby J T.Role of irradiated microstructure and micro‐chemistry in irradiation-assisted stress corrosion cracking[J].Phi‐los.Mag.,2005,85:443
[12]West E A,Was G S.Strain incompatibilities and their role in inter‐granular cracking of irradiated 316L stainless steel[J].J.Nucl.Ma‐ter.,2013,441:623
[13]McMurtrey M D,Was G S,Cui B,et al.Strain localization at dislo‐cation channel-grain boundary intersections in irradiated stainless steel[J].Int.J.Plast.,2014,56:219
[14]Was G S,Farkas D,Robertson I M.Micromechanics of dislocation channeling in intergranular stress corrosion crack nucleation[J].Curr.Opin.Solid State Mater.Sci.,2012,16:134
[15]Fukuya K,Nishioka H,Fujii K,et al.Charateriation of IASCCcrack tip in highly irradiated stainless steels[A].Proceeding of the14th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C].Virgin‐ia Beach,Virginia,2009:1248(CD-ROM)
[16]Thomas L E,Bruemmer S M.Microstructural and microchemical characterization of intergranular stress corrosion cracks in irradiat‐ed type 304SS removed from a BWR top guide[A].Proceeding of the 11th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C].Ste‐venson,WA,2003:1049(CD-ROM)
[17]Edwards D J,Thomas L E,Asano K,et al.Microstructure,micro‐chemistry and stress corrosion crack characteristics in a BWR316L SS core shroud weld[A].Proceeding of the 13th Internation‐al Conference on Environmental Degradation of Materials in Nu‐clear Power Systems-Water Reactors[C].Whister,British Colu‐mia,2007(CD-ROM)
[18]Thomas L E,Edwards D J,Asano K,et al.Crack-tip characteris‐tics in BWR service components[A].Proceeding of the 13th Inter‐national Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors[C].Whister,British Columia,2007(CD-ROM)
[19]Deng P,Peng Q J,Han E H,et al.Effect of irradiation on corrosion of 304 nuclear grade stainless steel in simulated PWR primary wa‐ter[J].Corros.Sci.,2017,127:91
[20]Kamaya M.Characterization of microstructural damage due to low-cycle fatigue by EBSD observation[J].Mater.Charact.,2009,60:1454
[21]Deng P,Peng Q J,Han E H,et al.Study of irradiation damage in domestically fabricated nuclear grade stainless steel[J].Acta Metall.Sin.,2017,53:1588(邓平,彭群家,韩恩厚等.国产核用不锈钢辐照损伤研究[J].金属学报,2017,53:1588)
[22]Shen Z,Arioka K,Lozano-Perez S.A mechanistic study of SCC in Alloy 600 through high-resolution characterization[J].Corros.Sci.,2018,132:244
[23]Lu Y H,Peng Q J,Sato T,et al.An ATEM study of oxidation be‐havior of SCC crack tips in 304L stainless steel in high tempera‐ture oxygenated water[J].J.Nucl.Mater.,2005,3477:52
[24]Lozano-Perez S,Kruska K,Iyengar I,et al.The role of cold work and applied stress on surface oxidation of 304 stainless steel[J].Corros.Sci.,2012,56:78
[25]Gertsman V Y,Bruemmer S M.Study of grain boundary character along intergranular stress corrosion crack paths in austenitic alloys[J].Acta Mater.,2001,49:1589
[26]Hu C L,Xia S,Li H,et al.Effect of grain boundary network on the intergranular stress corrosion cracking of 304 stainless steel[J].Acta Metall.Sin.,2011,47:939(胡长亮,夏爽,李慧等.晶界网络特征对304不锈钢晶间应力腐蚀开裂的影响[J].金属学报,2011,47:939)
[27]Ming H L,Zhang Z M,Wang J Q,et al.Microstructure and local properties of a domestic safe-end dissimilar metal weld joint by us‐ing hot-wire GTAW[J].Acta Metall.Sin.,2017,53:57(明洪亮,张志明,王俭秋等.国产核电安全端异种金属焊接件的微观结构及局部性能研究[J].金属学报,2017,53:57)
[28]Hou J,Peng Q J,Shoji T,et al.Effects of cold working path on strain concentration,grain boundary microstructure and stress cor‐rosion cracking in Alloy 600[J].Corros.Sci.,2011,53:2956
[29]Hou J,Peng Q J,Lu Z P,et al.Effects of cold working degrees on grain boundary characters and strain concentration at grain bound‐aries in Alloy 600[J].Corros.Sci.,2011,53:1137
[30]Alexandreanu B,Was G S.Grain boundary deformation-induced intergranular stress corrosion cracking of Ni-16Cr-9Fe in 360℃Water[J].Corrosion,2003,59:705