Please wait a minute...
金属学报  2019, Vol. 55 Issue (3): 349-361    DOI: 10.11900/0412.1961.2018.00359
  本期目录 | 过刊浏览 |
核用304不锈钢辐照促进应力腐蚀开裂研究
邓平1,2,孙晨3,彭群家1,4(),韩恩厚1,柯伟1,焦治杰5
1. 中国科学院金属研究所中国科学院核用材料与安全性评价重点实验室 沈阳 110016
2. 中国科学技术大学材料科学与工程学院 沈阳 110016
3. 国家电投集团科学技术研究院有限公司 北京 102209
4. 苏州热工研究院有限公司 苏州 215004
5. Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109, USA
Study on Irradiation Assisted Stress Corrosion Cracking of Nuclear Grade 304 Stainless Steel
Ping DENG1,2,Chen SUN3,Qunjia PENG1,4(),En-Hou HAN1,Wei KE1,Zhijie JIAO5
1. CAS Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, China
2. School of Materials Science and Engineering, University of Science and Technology of China, Shenyang 110016, China
3. State Power Investment Corporation Research Institute, Beijing 102209, China
4. Suzhou Nuclear Power Research Institute, Suzhou 215004, China
5. Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109, USA
全文: PDF(24156 KB)   HTML
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。

关键词 核用不锈钢质子辐照局部变形腐蚀辐照促进应力腐蚀开裂    
Abstract

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 (304SS) was investigated. The IASCC tests were conducted by interrupted slow strain rate tensile (SSRT) tests at 320 ℃ in simulated primary water of pressurized water reactor containing 1200 mg/L B as H3BO3 and 2.3 mg/L Li as LiOH·H2O, 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 304SS. 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 grade 304SS 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.

Key wordsnuclear grade stainless steel    proton irradiation    localized deformation    corrosion    irradiation assisted stress corrosion cracking
收稿日期: 2018-07-31      出版日期: 2018-11-14
ZTFLH:  TG139.4  
基金资助:科技部国际合作专项项目(2014DFA50800);国家自然科学基金项目(51571204);国家核电技术公司基础研究项目(2015SN010-007)
通讯作者: 彭群家     E-mail: pengqunjia@cgnpc.com.cn
Corresponding author: Qunjia PENG     E-mail: pengqunjia@cgnpc.com.cn
作者简介: 邓 平,男,1989年生,博士生

引用本文:

邓平,孙晨,彭群家,韩恩厚,柯伟,焦治杰. 核用304不锈钢辐照促进应力腐蚀开裂研究[J]. 金属学报, 2019, 55(3): 349-361.
Ping DENG,Chen SUN,Qunjia PENG,En-Hou HAN,Wei KE,Zhijie JIAO. Study on Irradiation Assisted Stress Corrosion Cracking of Nuclear Grade 304 Stainless Steel. Acta Metall Sin, 2019, 55(3): 349-361.

链接本文:

http://www.ams.org.cn/CN/10.11900/0412.1961.2018.00359      或      http://www.ams.org.cn/CN/Y2019/V55/I3/349

图1  慢应变速率拉伸(SSRT)试样形状和尺寸示意图
图2  0.5 dpa辐照试样经3%应变SSRT实验后应力腐蚀开裂(SCC)区域的EBSD分析
图3  5.0 dpa辐照试样经3%应变SSRT实验后SCC区域的EBSD分析
图4  5.0 dpa辐照试样经3%应变SSRT实验后晶界处滑移台阶的EBSD分析
图5  核用304不锈钢中晶粒的Schmidt因子分布和Schmidt因子对类型与晶界开裂的关系
图6  0.5 dpa辐照试样经3%应变SSRT实验后变形结构的TEM像及分析
图7  核用304不锈钢3%应变SSRT实验后SCC行为的表面SEM像及滑移台阶与晶界作用示意图
图8  核用304不锈钢间断SSRT实验后表面SCC裂纹统计结果
图9  核用304不锈钢SSRT实验至断裂失效后表面二次裂纹的SEM像
图10  0.5 dpa辐照试样经3%应变SSRT实验后裂纹尖端TEM像与分析
图11  5.0 dpa辐照试样经3%应变SSRT实验后裂纹尖端TEM像与分析
图12  0.5和5.0 dpa辐照试样在一回路水中SSRT实验后断口的SEM像
图13  Schmidt因子和台阶滑移促进晶界应变集中示意图
图14  晶界局部变形和局部腐蚀促进IASCC机制示意图
[1] Gérard R, Somville F. Situation of the baffle-former bolts in Belgian 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 Systems—Water Reactors [C]. Houston: National Association of Corrosion Engineers, 1989: 83
[3] Nishioka H, Fukuya K, Fujii K, et al. IASCC initiation in highly irradiated 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 corrosion cracking of austenitic alloys in supercritical water [J]. J. Nucl. 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, deformation 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]. Proceeding of the 13th International Conference on Environmental Degradation 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 localized deformation of austenitic stainless steels [J]. J. Nucl. Mater., 2010, 407: 34
[11] Was G S, Busby J T. Role of irradiated microstructure and microchemistry in irradiation-assisted stress corrosion cracking [J]. Philos. Mag., 2005, 85: 443
[12] West E A, Was G S. Strain incompatibilities and their role in intergranular cracking of irradiated 316L stainless steel [J]. J. Nucl. Mater., 2013, 441: 623
[13] McMurtrey M D, Was G S, Cui B, et al. Strain localization at dislocation 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 IASCC crack tip in highly irradiated stainless steels [A]. Proceeding of the 14th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Virginia Beach, Virginia, 2009: 1248 (CD-ROM)
[16] Thomas L E, Bruemmer S M. Microstructural and microchemical characterization of intergranular stress corrosion cracks in irradiated 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]. Stevenson, WA, 2003: 1049 (CD-ROM)
[17] Edwards D J, Thomas L E, Asano K, et al. Microstructure, microchemistry and stress corrosion crack characteristics in a BWR 316L SS core shroud weld [A]. Proceeding of the 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [C]. Whister, British Columia, 2007 (CD-ROM)
[18] Thomas L E, Edwards D J, Asano K, et al. Crack-tip characteristics in BWR service components [A]. Proceeding of the 13th International 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 water [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
[21] 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 behavior of SCC crack tips in 304L stainless steel in high temperature 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
doi: 10.3724/SP.J.1037.2011.00184
[26] 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
doi: 10.3724/SP.J.1037.2011.00184
[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 using hot-wire GTAW [J]. Acta Metall. Sin., 2017, 53: 57
[27] 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 corrosion 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 boundaries 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
[1] 李鑫,董月成,淡振华,常辉,方志刚,郭艳华. 等通道角挤压制备超细晶纯Ti的腐蚀性能研究[J]. 金属学报, 2019, 55(8): 967-975.
[2] 蔡超,李煬,李劲风,张昭,张鉴清. 2A97 Al-Li合金薄板时效析出与电位及晶间腐蚀的相关性研究[J]. 金属学报, 2019, 55(8): 958-966.
[3] 刘灿帅,田朝晖,张志明,王俭秋,韩恩厚. 地质处置低氧过渡期X65低碳钢腐蚀行为研究[J]. 金属学报, 2019, 55(7): 849-858.
[4] 李亚东,李强,唐晓,李焰. X80管线钢焊接接头的模拟重构及电偶腐蚀行为表征[J]. 金属学报, 2019, 55(6): 801-810.
[5] 陈星晨, 王杰, 陈德任, 钟舜聪, 王向峰. Na对于Al早期大气腐蚀的影响[J]. 金属学报, 2019, 55(4): 529-536.
[6] 马宏驰, 杜翠薇, 刘智勇, 李永, 李晓刚. E690高强低合金钢焊接热影响区典型组织在含SO2海洋环境中的应力腐蚀行为对比研究[J]. 金属学报, 2019, 55(4): 469-479.
[7] 廖依敏, 丰敏, 陈明辉, 耿哲, 刘阳, 王福会, 朱圣龙. TiAl合金表面搪瓷基复合涂层与多弧离子镀NiCrAlY涂层的抗热腐蚀行为对比研究[J]. 金属学报, 2019, 55(2): 229-237.
[8] 陈闽东, 张帆, 刘智勇, 杨朝晖, 丁国清, 李晓刚. 金属材料在三亚海水中的腐蚀电位序及合金成分对耐蚀性的影响[J]. 金属学报, 2018, 54(9): 1311-1321.
[9] 曾荣昌, 崔蓝月, 柯伟. 医用镁合金:成分、组织及腐蚀[J]. 金属学报, 2018, 54(9): 1215-1235.
[10] 梁秀兵, 范建文, 张志彬, 陈永雄. 铝基非晶纳米晶复合涂层显微组织与腐蚀性能研究[J]. 金属学报, 2018, 54(8): 1193-1203.
[11] 刘金辉, 宋影伟, 单大勇, 韩恩厚. 铸态和锻造态Mg-5Y-7Gd-1Nd-0.5Zr合金腐蚀行为对比研究[J]. 金属学报, 2018, 54(8): 1141-1149.
[12] 屈少鹏, 程柏璋, 董丽华, 尹衍升, 杨丽景. 2205钢在模拟深海热液区中的腐蚀行为[J]. 金属学报, 2018, 54(8): 1094-1104.
[13] 吴俊升, 张博威, 李晓刚, 黄一中. 纳米金属腐蚀[J]. 金属学报, 2018, 54(8): 1087-1093.
[14] 范丽, 陈海龑, 董耀华, 李雪莹, 董丽华, 尹衍升. 激光熔覆铁基合金涂层在HCl溶液中的腐蚀行为[J]. 金属学报, 2018, 54(7): 1019-1030.
[15] 马政, 陆喜, 高明, 谭丽丽, 杨柯. 缝隙腐蚀对Fe-30Mn-1C合金降解速率的影响[J]. 金属学报, 2018, 54(7): 1010-1018.