Review of Stress Corrosion Crack Initiation of Nuclear Structural Materials in High Temperature and High Pressure Water

doi:10.11902/1005.4537.2021.130

Journal of Chinese Society for Corrosion and protection

2022, Vol. 42

Issue (4): 513-522 DOI: 10.11902/1005.4537.2021.130

Current Issue | Archive | Adv Search

Review of Stress Corrosion Crack Initiation of Nuclear Structural Materials in High Temperature and High Pressure Water

LIU Baoping¹^,², ZHANG Zhiming¹^,³(

), WANG Jianqiu¹, HAN En-Hou¹, KE Wei¹

^1.Key Laboratory of Nuclear Materials and Safety Assessment, Liaoning Key Laboratory for Safety and Assessment Technique of Nuclear Materials, 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.Institute of Corrosion Science and Technology, Guangzhou 510700, China

Download:

HTML

PDF(2715KB)
Export: BibTeX | EndNote (RIS)

Abstract

In this paper, the test methods, evaluation indexes, influencing factors and initiation mechanism related with the stress corrosion cracking in high temperature and high pressure water for structural materials such as stainless steels and nickel-based alloys commonly used in nuclear power plants are reviewed, and the shortcomings of current research and the future research trends are also pointed out.

Key words: nuclear power plant structural material high temperature and high pressure water stress corrosion cracking initiation mechanism

Received: 10 June 2021

ZTFLH:

TG174

Fund: National Key Research and Development Program of China(2017YFB0702100)

Corresponding Authors: ZHANG Zhiming E-mail: zmzhang@imr.ac.cn

About author: ZHANG Zhiming, E-mail: zmzhang@imr.ac.cn

	Service

	E-mail this article
	Add to citation manager
	E-mail Alert
	RSS
	（）
	Articles by authors
	Baoping LIU
	Zhiming ZHANG
	Jianqiu WANG
	En-Hou HAN
	Wei KE

Cite this article:

LIU Baoping, ZHANG Zhiming, WANG Jianqiu, HAN En-Hou, KE Wei. Review of Stress Corrosion Crack Initiation of Nuclear Structural Materials in High Temperature and High Pressure Water. Journal of Chinese Society for Corrosion and protection, 2022, 42(4): 513-522.

URL:

https://www.jcscp.org/EN/10.11902/1005.4537.2021.130 OR https://www.jcscp.org/EN/Y2022/V42/I4/513

Fig.1 Detection of crack initiation in 600MA alloy by DCPD: (a) relation diagram of referenced strain and time, (b) crack on sample surface, (c) crack on sample section^[26]

Fig.2 Effects of DH on crack growth rate of 600 alloy and 82/182 alloy at 325 ℃ (a) and the crack initiation time of 600 alloy at 360 ℃ (b)^[6,25]

Fig.3 Schematic diagram of internal oxidation mechan-ism: carbide formed at grain boundary on the left and not formed on the right^[76]

Fig.4 Schematic diagram of formation of cavities (a) and cavity morphologies of 690TT alloy (21%-31% cold-rolled, BNCT specimen, 360 ℃, 26 cc/kg H₂) in precursor crack at 5352 h (b), the shallow IG crack at 10049 h (c), the mature SCC crack at 11709 h (d)^[53,54]

Fig.5 Schematics of the SCC initiation stages of alloy 690 in simulated PWR primary environment under dynamic straining: (a) formation of compact Cr₂O₃ above the migrated grain boundary, (b) intergranular oxidation as the surface Cr₂O₃ is breached by straining, (c) crack nucleation within the intergranular oxide as the oxidation proceeds along the migration zone^[81]

1	Berg H P. Corrosion mechanisms and their consequences for nuclear power plants with light water reactors [J]. Reliab. Risk Anal.: Theory Appl., 2009, 2: 57
2	Liu X, Zhao J C, Wang G G, et al. Failure analysis of pipelines and welding joints in nuclear power plant [J]. Failure Anal. Prev., 2013, 8: 300
	刘肖, 赵建仓, 王淦刚等. 核电厂管道及焊接接头失效案例综述 [J]. 失效分析与预防, 2013, 8: 300
3	Was G S, Ashida Y, Andresen P L. Irradiation-assisted stress corrosion cracking [J]. Corros. Rev., 2011, 29: 7
4	Sun H T, Ling L G, Lv Y H, et al. Stress corrosion problems and safety management of equipment and materials in domestic pressurized water reactor nuclear power plants [J]. Corros. Sci. Prot. Technol., 2016, 28: 283
	孙海涛, 凌礼恭, 吕云鹤等. 国内压水堆核电站设备材料应力腐蚀问题及安全管理 [J]. 腐蚀科学与防护技术, 2016, 28: 283
5	Hojná A. Environmentally assisted cracking initiation in high-tem-perature water [J]. Metals, 2021, 11: 199 doi: 10.3390/met11020199
6	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: 707 doi: 10.5006/1.3278508
7	Andresen P L. Stress corrosion cracking of current structural materials in commercial nuclear power plants [J]. Corrosion, 2013, 69: 1024 doi: 10.5006/0801
8	Amzallag C, Vaillant F. Stress corrosion crack propagation rates in reactor vessel head penetrations in alloy 600 [A]. BruemmerS, FordP, WasG. Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors [M]. The Minerals, Metals and Materials Society, Newport Beach, 1999: 235
9	Zhu R L, Wang J Q, Zhang L T, et al. Stress corrosion cracking of 316L HAZ for 316L stainless steel/Inconel 52M dissimilar metal weld joint in simulated primary water [J]. Corros. Sci., 2016, 112: 373 doi: 10.1016/j.corsci.2016.07.031
10	Zhang L T, Wang J Q. Effect of dissolved oxygen content on stress corrosion cracking of a cold worked 316L stainless steel in simulated pressurized water reactor primary water environment [J]. J. Nucl. Mater., 2014, 446: 15 doi: 10.1016/j.jnucmat.2013.11.027
11	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]. J. Nucl. Mater., 2018, 503: 13 doi: 10.1016/j.jnucmat.2018.02.032
12	Du D H, Wang J M, Chen K, et al. Environmentally assisted cracking of forged 316LN stainless steel and its weld in high temperature water [J]. Corros. Sci., 2019, 147: 69 doi: 10.1016/j.corsci.2018.10.032
13	Chen K, Wang J M, Shen Z, et al. Effect of intergranular carbides on the cracking behavior of cold worked alloy 690 in subcritical and supercritical water [J]. Corros. Sci., 2020, 164: 108313 doi: 10.1016/j.corsci.2019.108313
14	Zhang L T, Wang J Q. Stress corrosion crack propagation behavior of domestic forged nuclear grade 316L stainless steel in high temperature and high pressure water [J]. Acta Metall. Sin., 2013, 49: 911 doi: 10.3724/SP.J.1037.2013.00171
	张利涛, 王俭秋. 国产锻造态核级管材316L不锈钢在高温高压水中的应力腐蚀裂纹扩展行为 [J]. 金属学报, 2013, 49: 911 doi: 10.3724/SP.J.1037.2013.00171
15	Zhu R L, Wang J Q, Zhang Z M, et al. Stress corrosion cracking of fusion boundary for 316L/52M dissimilar metal weld joints in borated and lithiated high temperature water [J]. Corros. Sci., 2017, 120: 219 doi: 10.1016/j.corsci.2017.01.024
16	Guo S, Han E-H, Wang H T, et al. Life prediction for stress corrosion behavior of 316L stainless steel elbow of nuclear power plant [J]. Acta Metall. Sin., 2017, 53: 455
	郭舒, 韩恩厚, 王海涛等. 核电站316L不锈钢弯头应力腐蚀行为的寿命预测 [J]. 金属学报, 2017, 53: 455
17	Zhu R L, Zhang Z M, Wang J Q, et al. Review on SCC crack growth behavior of dissimilar metal welds for nuclear power reactors [J]. J. Chin. Soc. Corros. Prot., 2015, 35: 189
	朱若林, 张志明, 王俭秋等. 核电异种金属焊接接头的应力腐蚀裂纹扩展行为研究进展 [J]. 中国腐蚀与防护学报, 2015, 35: 189
18	Zhang K Q, Hu S L, Tang Z M, et al. Review on stress corrosion crack propagation behavior of cold worked nuclear structural materials in high temperature and high pressure water [J]. J. Chin. Soc. Corros. Prot., 2018, 38: 517
	张克乾, 胡石林, 唐占梅等. 冷加工核电结构材料在高温高压水中应力腐蚀裂纹扩展行为的研究进展 [J]. 中国腐蚀与防护学报, 2018, 38: 517
19	Zhang K Q, Tang Z M, Hu S L, et al. The research status of SCC crack propagation in structural materials of nuclear reactors [J]. Corros. Prot., 2019, 40: 157
	张克乾, 唐占梅, 胡石林等. 核电用结构材料SCC裂纹扩展的研究现状 [J]. 腐蚀与防护, 2019, 40: 157
20	Ma C, Peng Q J, Han E-H, et al. Review of stress corrosion cracking of structural materials in nuclear power plants [J]. J. Chin. Soc. Corros. Prot., 2014, 34: 37
	马成, 彭群家, 韩恩厚等. 核电结构材料应力腐蚀开裂的研究现状与进展 [J]. 中国腐蚀与防护学报, 2014, 34: 37
21	Jiao Y, Zhang S H, Tan Y. Research progress on stress corrosion cracking of stainless steel for nuclear power plant in high-temperature and high-pressure water [J]. J. Chin. Soc. Corros. Prot., 2021, 41: 417
	焦洋, 张胜寒, 檀玉. 核电站用不锈钢在高温高压水中应力腐蚀开裂行为的研究进展 [J]. 中国腐蚀与防护学报, 2021, 41: 417
22	Dozaki K, Akutagawa D, Nagata N, et al. Effects of dissolved hydrogen content in PWR primary water on PWSCC initiation property [J]. E-J. Adv. Maint., 2010, 2: 65
23	Moss T, Kuang W J, Was G S. Stress corrosion crack initiation in Alloy 690 in high temperature water [J]. Curr. Opin. Solid State Mater. Sci., 2018, 22: 16 doi: 10.1016/j.cossms.2018.02.001
24	Boursier J M, Desjardins D, Vaillant F. The influence of the strain-rate on the stress corrosion cracking of alloy 600 in high temperature primary water [J]. Corros. Sci., 1995, 37: 493 doi: 10.1016/0010-938X(94)00158-3
25	Richey E, Morton D, Schurman M. SCC initiation testing of nickel-based alloys using in-situ monitored uniaxial tensile specimens [A]. AllenT R, KingP J, NelsonL. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. The Minerals, Metals & Materials Society, 2005: 947
26	Zhai Z Q, Toloczko M B, Olszta M J, et al. Stress corrosion crack initiation of alloy 600 in PWR primary water [J]. Corros. Sci., 2017, 123: 76 doi: 10.1016/j.corsci.2017.04.013
27	Toloczko M, Zhai Z Q, Bruemmer S. SCC initiation behavior of alloy 182 in PWR primary water [A]. JacksonJH, ParaventiD, WrightM. Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. Cham: Springer, 2019: 137
28	Pemberton S R, Chatterton M A, Griffiths A S, et al. The effect of surface condition on primary water stress corrosion cracking initiation of alloy 600 [A]. JacksonJ H, ParaventiD, WrightM. Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. Cham, 2019: 203
29	Wu W B. Study of stress corrosion cracking initiation and propagation behavior of two typical nuclear key metal materials [D]. Hefei: University of Science and Technology of China, 2019
	吴文博. 两种典型核电关键金属材料应力腐蚀裂纹萌生与扩展行为研究 [D]. 合肥: 中国科学技术大学, 2019
30	Arioka K, Staehle R W, Yamada T, et al. Degradation of alloy 690 after relatively short times [J]. Corrosion, 2016, 72: 1252 doi: 10.5006/2107
31	Kuang W J, Was G S. The effects of grain boundary carbide density and strain rate on the stress corrosion cracking behavior of cold rolled Alloy 690 [J]. Corros. Sci., 2015, 97: 107 doi: 10.1016/j.corsci.2015.04.020
32	Zhong X Y, Bali S C, Shoji T. Accelerated test for evaluation of intergranular stress corrosion cracking initiation characteristics of non-sensitized 316 austenitic stainless steel in simulated pressure water reactor environment [J]. Corros. Sci., 2017, 115: 106 doi: 10.1016/j.corsci.2016.11.019
33	Hong S L. Influence of surface condition on primary water stress corrosion cracking initiation of alloy 600 [J]. Corrosion, 2001, 57: 323 doi: 10.5006/1.3290356
34	Chang L T, Volpe L, Wang Y L, et al. Effect of machining on stress corrosion crack initiation in warm-forged type 304L stainless steel in high temperature water [J]. Acta Mater., 2019, 165: 203 doi: 10.1016/j.actamat.2018.11.046
35	Chang L T, Burke M G, Scenini F. Stress corrosion crack initiation in machined type 316L austenitic stainless steel in simulated pressurized water reactor primary water [J]. Corros. Sci., 2018, 138: 54 doi: 10.1016/j.corsci.2018.04.003
36	Wang S, Shoji T, Kawaguchi N. Initiation of environmentally assisted cracking in high-temperature water [J]. Corrosion, 2005, 61: 137 doi: 10.5006/1.3278168
37	Hong S L, Amzallag C, Gelpi A. Modelling of stress corrosion crack initiation on alloy 600 in primary water of PWRs [A]. BruemmerS, FordP, WasG. Proceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. The Minerals, Metals and Materials Society, Newport Beach, 1999: 115
38	Rebak R B, Smialowska Z S. Influence of stress intensity and loading mode on intergranular stress corrosion cracking of alloy 600 in primary waters of pressurized water reactors [J]. Corrosion, 1994, 50: 378 doi: 10.5006/1.3294347
39	Kuang W J, Was G S, Miller C, et al. The effect of cold rolling on grain boundary structure and stress corrosion cracking susceptibility of twins in alloy 690 in simulated PWR primary water environment [J]. Corros. Sci., 2018, 130: 126 doi: 10.1016/j.corsci.2017.11.002
40	Moss T, Was G S. Accelerated stress corrosion crack initiation of alloys 600 and 690 in hydrogenated supercritical water [J]. Metall. Mater. Trans., 2017, 48A: 1613
41	Etien R A, Richey E, Morton D S, et al. SCC initiation testing of alloy 600 in high temperature water [A]. BusbyJ T, IlevbareG, AndresenP L. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. Cham: Springer, 2011: 2407
42	Maeng W Y, Choi M S, Kim U C. Effect of dissolved oxygen on PWSCC susceptibility of Alloy 600 in high temperature water [J]. J. Mater. Sci., 2004, 39: 655 doi: 10.1023/B:JMSC.0000011524.41986.4c
43	Tsutsumi K, Couvant T. Evaluation of the susceptibility to SCC initiation of alloy 690 in simulated PWR primary water [A]. BusbyJ T, IlevbareG, AndresenP L. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. Cham: Springer, 2011: 41
44	Zhang W Q, Fang K W, Hu Y J, et al. Effect of machining-induced surface residual stress on initiation of stress corrosion cracking in 316 austenitic stainless steel [J]. Corros. Sci., 2016, 108: 173 doi: 10.1016/j.corsci.2016.03.008
45	Santarini G. Comprehensive interpretation of CERTs: a method for the characterization and the prediction of IGSCC [J]. Corrosion, 1989, 45: 369 doi: 10.5006/1.3582031
46	Isselin J, Kai A, Sakaguchi K, et al. Assessment of the effects of cold work on crack initiation in a light water environment using the small-punch test [J]. Metall. Mater. Trans., 2008, 39A: 1099
47	Zhang K Q, Tang Z M, Hu S L, et al. Effect of cold work and slow strain rate on 321SS stress corrosion cracking in abnormal conditions of simulated PWR primary environment [J]. Nucl. Mater. Energy, 2019, 20: 100697
48	Kuniya J, Masaoka I, Sasaki R. Effect of cold work on the stress corrosion cracking of nonsensitized AISI 304 stainless steel in high-temperature oxygenated water [J]. Corrosion, 1988, 44: 21 doi: 10.5006/1.3582020
49	Arioka K, Miyamoto T, Yamada T, et al. Role of cavity formation in crack initiation of cold-worked carbon steel in high-temperature water [J]. Corrosion, 2013, 69: 487 doi: 10.5006/0821
50	Zhang H B, Li S J, Hu Y H, et al. Research status of Inconel 690 alloy in steam generator heat transfer tubes abroad [J]. Spec. Steel Technol., 2003, 8(4): 2
	张红斌, 李守军, 胡尧和等. 国外关于蒸汽发生器传热管用Inconel 690合金研究现状 [J]. 特钢技术, 2003, 8(4): 2
51	Yang Y Z, Cai Z G, Wang Y D, et al. Effects of thermal treatment on microstructure and properties of 690 alloy heat transfer tubes [J]. Hot Work. Technol., 2017, 46(22): 209
	杨义忠, 蔡志刚, 王永东等. TT处理对690合金传热管显微组织和性能的影响 [J]. 热加工工艺, 2017, 46(22): 209
52	Zhang Y Y. Effects of deformation and hot treatment on microstructure and mechanical properties of 690 alloy [D]. Dalian: Dalian University of Technology, 2017
	张雨樾. 变形和热处理对690合金微观组织与力学性能的影响 [D]. 大连: 大连理工大学, 2017
53	Arioka K. 2014 W. R. Whitney Award Lecture: change in bonding strength at grain boundaries before long-term SCC initiation [J]. Corrosion, 2015, 71: 403 doi: 10.5006/1573
54	Zhai Z Q, Toloczko M, Kruska K, et al. Grain boundary damage evolution and SCC initiation of cold-worked alloy 690 in simulated PWR primary water [A]. JacksonJH, ParaventiD, WrightM. Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. Cham: Springer, 2019: 457
55	Yu W W, Meng X M, Jiang J W, et al. Investigation on sensitive to thermal aging for key materials used in primary circuit of nuclear power plants [J]. Nucl. Power Eng. Technol., 2014, 27(3): 22
	余伟炜, 蒙新明, 姜家旺等. 核电站一回路关键设备材料热老化敏感性分析 [J]. 核电工程与技术, 2014, 27(3): 22
56	Lin X D, Peng Q J, Han E-H, et al. Review of thermal aging of nuclear grade stainless steels [J]. J. Chin. Soc. Corros. Prot., 2017, 37: 81
	林晓冬, 彭群家, 韩恩厚等. 核级不锈钢的热老化研究进展 [J]. 中国腐蚀与防护学报, 2017, 37: 81
57	Yoo S C, Choi K J, Kim T, et al. Microstructural evolution and stress-corrosion-cracking behavior of thermally aged Ni-Cr-Fe alloy [J]. Corros. Sci., 2016, 111: 39 doi: 10.1016/j.corsci.2016.04.051
58	Yoo S C, Choi K J, Kim T, et al. Effects of thermal aging and stress triaxiality on PWSCC initiation susceptibility of nickel-based Alloy 600 [J]. J. Mater. Sci. Technol., 2016, 30: 4403
59	Li S L, Wang Y L, Wang H, et al. Effects of long-term thermal aging on the stress corrosion cracking behavior of cast austenitic stainless steels in simulated PWR primary water [J]. J. Nucl. Mater., 2016, 469: 262 doi: 10.1016/j.jnucmat.2015.11.043
60	Lai C L, Lu W F, Huang J Y. Effect of δ-ferrite content on the stress corrosion cracking behavior of cast austenitic stainless steel in high-temperature water environment [J]. Corrosion, 2014, 70: 591 doi: 10.5006/1155
61	Kim H S, Hong J D, Lee J, et al. Effects of hydrogen on the PWSCC initiation behaviours of alloy 182 weld in PWR environments [J]. Corros. Sci. Technol., 2015, 14: 113 doi: 10.14773/cst.2015.14.3.113
62	Li Y C, Zhu Z P, Yang D W, et al. Hydrochemical Control Conditions in Nuclear Power Plants [M]. Beijing: Chemical Industry Press, 2008
	李宇春, 朱志平, 杨道武等. 核电站水化学控制工况 [M]. 北京: 化学工业出版社, 2008
63	Kim Y J, Andresen P L, Moran E, et al. Modification of surface property for controlling the type 304 stainless steel electrochemical corrosion potential in 288℃ water [J]. Corrosion, 2005, 61: 648 doi: 10.5006/1.3278200
64	Soustelle C, Foucault M, Framatome P C, et al. PWSCC of alloy 600: a parametric study of surface film effects [A]. BruemmerS, FordP, WasG. Proceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. The Minerals, Metals and Materials Society, Newport Beach, 1999: 105
65	Zhong X Y, Bali S C, Shoji T. Effects of dissolved hydrogen and surface condition on the intergranular stress corrosion cracking initiation and short crack growth behavior of non-sensitized 316 stainless steel in simulated PWR primary water [J]. Corros. Sci., 2017, 118: 143 doi: 10.1016/j.corsci.2017.02.003
66	Nakagawa K, Nono M, Kimura A. Effect of dissolved hydrogen on the SCC susceptibility of SUS316L stainless steel [J]. Mater. Sci. Forum, 2010, 654-656: 2887 doi: 10.4028/www.scientific.net/MSF.654-656.2887
67	Choi K J, Yoo S C, Kim T, et al. Effects of dissolved hydrogen on crack growth rate of warm-rolled 316L austenitic stainless steel in primary water condition [A]. Proceedings of the ASME 2015 Pressure Vessels and Piping Conference [C]. Boston, 2015: 1
68	Matocha K, Wozniak J. Stress corrosion cracking initiation in austenitic stainless steel in high temperature water [A]. BruemmerS, FordP, WasG. Proceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. The Minerals, Metals and Materials Society, Newport Beach, 1999: 383
69	Kawamura H, Hirano H, Shirai S, et al. Inhibitory effect of zinc addition to high-temperature hydrogenated water on mill-annealed and prefilmed alloy 600 [J]. Corrosion, 2000, 56: 623 doi: 10.5006/1.3280565
70	Febrianto, Sriyono, Widodo S, et al. The effect of zinc injection on the increasing of Inconel 600 TT corrosion resistances [J]. J. Phys.: Conf. Ser., 2018, 962: 012049
71	Ford F P. Quantitative prediction of environmentally assisted cracking [J]. Corrosion, 1996, 52: 375 doi: 10.5006/1.3292125
72	Andresen P L. Emerging issues and fundamental processes in environmental cracking in hot water [J]. Corrosion, 2008, 64: 439 doi: 10.5006/1.3278483
73	Macdonald D D, Lu P C, Urquidi-Macdonald M, et al. Theoretical estimation of crack growth rates in type 304 stainless steel in boiling-water reactor coolant environments [J]. Corrosion, 1996, 52: 768 doi: 10.5006/1.3292070
74	MacDonald D D, Urquidi-MacDonald M. A coupled environment model for stress corrosion cracking in sensitized type 304 stainless steel in LWR environments [J]. Corros. Sci., 1991, 32: 51 doi: 10.1016/0010-938X(91)90063-U
75	Peng Q J, Hou J, Takeda Y, et al. Effect of chemical composition on grain boundary microchemistry and stress corrosion cracking in Alloy 182 [J]. Corros. Sci., 2013, 67: 91 doi: 10.1016/j.corsci.2012.10.012
76	Panter J, Viguier B, Cloué J M, et al. Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600 [J]. J. Nucl. Mater., 2006, 348: 213 doi: 10.1016/j.jnucmat.2005.10.002
77	Scott P M. An overview of internal oxidation as a possible explanation of intergranular stress corrosion cracking of alloy 600 in PWRS [A]. BruemmerS, FordP, WasG. Proceedings of the Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. The Minerals, Metals and Materials Society, Newport Beach, 1999: 3
78	Arioka K, Miyamoto T, Yamad T, et al. Role of cavity formation on crack growth of cold-worked carbon steel, TT 690 and MA 600 in high temperature water [A]. BusbyJ T, IlevbareG, AndresenP L. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [M]. Cham: Springer, 2011: 55
79	Arioka K, Miyamoto T, Yamada T, et al. Formation of cavities prior to crack initiation and growth on cold-worked carbon steel in high-temperature water [J]. Corrosion, 2010, 66: 015008
80	Van Bueren H G. Theory of the formation of lattice defects during plastic strain [J]. Acta Metall., 1955, 3: 519 doi: 10.1016/0001-6160(55)90109-7
81	Kuang W J, Was G S. A high-resolution characterization of the initiation of stress corrosion crack in Alloy 690 in simulated pressurized water reactor primary water [J]. Corros. Sci., 2020, 163: 108243 doi: 10.1016/j.corsci.2019.108243
82	Volpe L, Burke M G, Scenini F. Correlation between grain boundary migration and stress corrosion cracking of alloy 600 in hydrogenated steam [J]. Acta Metall., 2020, 186: 454
83	Terachi T, Totsuka N, Yamada T, et al. Influence of dissolved hydrogen on structure of oxide film on alloy 600 formed in primary water of pressurized water reactors [J]. J. Nucl. Sci. Technol., 2003, 40: 509 doi: 10.1080/18811248.2003.9715385
84	Ferguson J B, Lopez H F. Oxidation products of INCONEL alloys 600 and 690 in pressurized water reactor environments and their role in intergranular stress corrosion cracking [J]. Metall. Mater. Trans., 2006, 37A: 2471
85	Han E-H. Research trends on micro and nano-scale materials degradation in nuclear power plant [J]. Acta Metall. Sin., 2011, 47: 769
	韩恩厚. 核电站关键材料在微纳米尺度上的环境损伤行为研究—进展与趋势 [J]. 金属学报, 2011, 47: 769

[1]	LIU Yutong, CHEN Zhenyu, ZHU Zhongliang, FENG Rui, BAO Hansheng, ZHANG Naiqiang. SCC Susceptibility of 2.25Cr1Mo Steel and Its Weld Joints in High Temperature Steam[J]. 中国腐蚀与防护学报, 2022, 42(4): 647-654.
[2]	LIU Haochen, FAN Lin, ZHANG Haibing, WANG Yingying, TANG Junlei, BAI Xuehan, SUN Mingxian. Research Progress of Stress Corrosion Cracking of Ti-alloy in Deep Sea Environments[J]. 中国腐蚀与防护学报, 2022, 42(2): 175-185.
[3]	ZHANG Ziyu, WU Xinqiang, HAN En-Hou, KE Wei. A Review on Corrosion Fatigue Crack Growth Behavior of Structural Materials in Nuclear Power Plants[J]. 中国腐蚀与防护学报, 2022, 42(1): 9-15.
[4]	YU Deyuan, LIU Zhiyong, DU Cuiwei, HUANG Hui, LIN Nan. Research Progress and Prospect of Stress Corrosion Cracking of Pipeline Steel in Soil Environments[J]. 中国腐蚀与防护学报, 2021, 41(6): 737-747.
[5]	SUN Baozhuang, ZHOU Xiaocheng, LI Xiaorong, SUN Weilu, LIU Zirui, WANG Yuhua, HU Yang, LIU Zhiyong. Stress Corrosion Cracking Behavior of 316L Stainless Steel with Varying Microstructure in Ammonium Chloride Environment[J]. 中国腐蚀与防护学报, 2021, 41(6): 811-818.
[6]	JIAO Yang, ZHANG Shenghan, TAN Yu. Research Progress on Stress Corrosion Cracking of Stainless Steel for Nuclear Power Plant in High-temperature and High-pressure Water[J]. 中国腐蚀与防护学报, 2021, 41(4): 417-428.
[7]	LIN Zhaohui, MING Nanxi, HE Chuan, ZHENG Ping, CHEN Xu. Effect of Hydrostatic Pressure on Corrosion Behavior of X70 Steel in Simulated Sea Water[J]. 中国腐蚀与防护学报, 2021, 41(3): 307-317.
[8]	WANG Xintong, CHEN Xu, HAN Zhenze, LI Chengyuan, WANG Qishan. Stress Corrosion Cracking Behavior of 2205 Duplex Stainless Steel in 3.5%NaCl Solution with Sulfate Reducing Bacteria[J]. 中国腐蚀与防护学报, 2021, 41(1): 43-50.
[9]	MA Mingwei, ZHAO Zhihao, JING Siwen, YU Wenfeng, GU Yien, WANG Xu, WU Ming. Corrosion Behavior of 17-4 PH Stainless Steel in Simulated Seawater Containing SRB[J]. 中国腐蚀与防护学报, 2020, 40(6): 523-528.
[10]	ZHU Lixia, JIA Haidong, LUO Jinheng, LI Lifeng, JIN Jian, WU Gang, XU Congmin. Effect of Applied Potential on Stress Corrosion Behavior of X80 Pipeline Steel and Its Weld Joint in a Simulated Liquor of Soil at Lunnan Area of Xinjiang[J]. 中国腐蚀与防护学报, 2020, 40(4): 325-331.
[11]	ZHANG Zhen, WU Xinqiang, TAN Jibo. *Review of Electrochemical Noise Technique for in situ* Monitoring of Stress Corrosion Cracking**[J]. 中国腐蚀与防护学报, 2020, 40(3): 223-229.
[12]	CHEN Xu,MA Jiong,LI Xin,WU Ming,SONG Bo. Synergistic Effect of SRB and Temperature on Stress Corrosion Cracking of X70 Steel in an ArtificialSea Mud Solution[J]. 中国腐蚀与防护学报, 2019, 39(6): 477-483.
[13]	Baojie WANG,Jiyu LUAN,Shidong WANG,Daokui XU. Research Progress on Stress Corrosion Cracking Behavior of Magnesium Alloys[J]. 中国腐蚀与防护学报, 2019, 39(2): 89-95.
[14]	Keqian ZHANG,Shilin HU,Zhanmei TANG,Pingzhu ZHANG. Review on Stress Corrosion Crack Propagation Behavior of Cold Worked Nuclear Structural Materials in High Temperature and High Pressure Water[J]. 中国腐蚀与防护学报, 2018, 38(6): 517-522.
[15]	Ruolin ZHU, Litao ZHANG, Jianqiu WANG, Zhiming ZHANG, En-Hou HAN. Stress Corrosion Crack Propagation Behavior of Elbow Pipe of Nuclear Grade 316LN Stainless Steel in High Temperature High Pressure Water[J]. 中国腐蚀与防护学报, 2018, 38(1): 54-61.

No Suggested Reading articles found!

Viewed

Full text

Abstract

Cited

Shared

Discussed