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中国腐蚀与防护学报  2014, Vol. 34 Issue (1): 37-45    DOI: 10.11902/1005.4537.2013.175
  本期目录 | 过刊浏览 |
核电结构材料应力腐蚀开裂的研究现状与进展
马成, 彭群家(), 韩恩厚, 柯伟
中国科学院金属研究所 金属腐蚀与防护国家重点实验室 沈阳 110016
Review of Stress Corrosion Cracking of Structural Materials in Nuclear Power Plants
MA Cheng, PENG Qunjia(), HAN En-Hou, KE Wei
State Key Laboratory of Corrosion and Protection, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, China
全文: PDF(1404 KB)   HTML
摘要: 

围绕应力腐蚀行为的实验研究方法、影响因素以及应力腐蚀机制的理论分析等几个方面综述了核电结构材料应力腐蚀研究的现状,讨论了研究中亟待解决的问题,指出了研究的发展方向与趋势。

关键词 核电结构材料高温高压水应力腐蚀冷加工焊接件    
Abstract

The structural materials used in light water reactors (LWR) such as nickel based alloys and stainless steels have been found to be susceptible to stress corrosion cracking (SCC) in high temperature water. In this review, we summarized the research progress and current status on SCC of the structural materials used in LWR in terms of experimental methods, factors influencing SCC and the mechanisms of SCC. The research hotspots like the influence of cold-working and the SCC of weld joints were discussed. Some of the challenges and perspectives for the research of SCC in the future were also briefly addressed.

Key wordsstructural materials of light water reactor    high temperature water    stress corrosion cracking    cold working    weld joint
收稿日期: 2013-08-19     
ZTFLH:  TG174  
基金资助:中国科学院“百人计划”项目资助
作者简介: null

马成,男,1988年生,硕士生,研究方向为核电结构材料的应力腐蚀

引用本文:

马成, 彭群家, 韩恩厚, 柯伟. 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(1): 37-45.
Cheng MA, Qunjia PENG, En-Hou HAN, Wei KE. Review of Stress Corrosion Cracking of Structural Materials in Nuclear Power Plants. Journal of Chinese Society for Corrosion and protection, 2014, 34(1): 37-45.

链接本文:

https://www.jcscp.org/CN/10.11902/1005.4537.2013.175      或      https://www.jcscp.org/CN/Y2014/V34/I1/37

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[1] Staehle R,Gorman J. 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
[2] Horn R M, Gordon G M, Ford F P, et al. Experience and assessment of stress corrosion cracking in L-grade stainless steel BWR internals[J]. Nucl. Eng. Des., 1997, 174(3): 313-325
[3] Andresen P L, Ford F P, Solomon H D, et al. Monitoring and modeling stress-corrosion and corrosion fatigue damage in nuclear-reactors[J]. JOM-J. Min. Met. Mat. Soc., 1990, 42(12): 7-11
[4] Bamford W, Hall J. Cracking of alloy 600 nozzles and welds in PWRs: review of cracking events and repair service experience [A]. Proceedings of 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Warrendale, PA: TMS, 2005: 959-966
[5] Zinkle S J, Was G S. Materials challenges in nuclear energy[J]. Acta Mater., 2013, 61(3): 735-758
[6] EPRI. Steam Generator Progress Report [M]. Palo Alto, CA: Electric Power Research Institute, 2011
[7] Crum J R, Nagashima T. Review of Alloy 690 Steam Generator Studies [M]. LaGrange Park: American Nuclear Society, 1997
[8] Hwang S S, Kim H P, Lee D H, et al. The mode of stress corrosion cracking in Ni-base alloys in high temperature water containing lead[J]. J. Nucl. Mater., 1999, 275(1): 28-36
[9] Peng Q J, Teysseyre S, Andresen P L, et al. Stress corrosion crack growth in type 316 stainless steel in supercritical water[J]. Corrosion, 2007, 63(11): 1033-1041
[10] Andresen P L, Emigh P W, Morra M M, et al. Effects of PWR primary water chemistry and deaerated water on SCC [A]. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Houston, TX: NACE, 2005: 989-1008
[11] Andersen P L, Morra M M, Hickling J, et al. PWSCC of alloys 690, 52 and 152 [A]. Proceedings of the 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Toronto: CNS, 2007
[12] Ritter S, Seifert H P. Effect of corrosion potential on the corrosion fatigue crack growth behaviour of low-alloy steels in high-temperature water[J]. J. Nucl. Mater., 2008, 375(1): 72-79
[13] 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]. J. Nucl. Mater., 2008, 378(2): 197-210
[14] Was G S, Sung J K, Angeliu T M. Effects of grain-boundary chemistry on the intergranular cracking behavior of Ni-16Cr-9Fe in high-temperature water[J]. Metall. Mater. Trans., 1992, 23(12)A: 3343-3359
[15] Was G S, Rajan V B. The mechanism of intergranular cracking of Ni-Cr-Fe alloys in sodium tetrathionate[J]. Metall. Mater. Trans., 1987, 18(7)A: 1313-1323
[16] Bruemmer S M. Linking grain boundary structure and composition to intergranular stress corrosion cracking of austenitic stainless steels[A]. Materials Research Society Symposium Proceedings [C]. Warrendale, PA: Materials Research Society; 2004: 101-110
[17] Bruemmer S M, Was G S. Microstructural and microchemical mechanisms controlling intergranular stress-corrosion cracking in light-water-reactor systems[J]. J. Nucl. Mater., 1994, 216: 348-363
[18] Peng Q J, Yamauchi H, Shoji T. Investigation of dendrite-boundary microchemistry in alloy 182 using auger electron spectroscopy analysis[J]. Metall. Mater. Trans., 2003, 34(9)A: 1891-1899
[19] Was G S, Lian K. Role of carbides in stress corrosion cracking resistance of alloy 600 and controlled-purity Ni-16%Cr-9%Fe in primary water at 360 ℃[J]. Corrosion, 1998, 54: 675-688
[20] Leonard F,Cottis R A,Vaillant F,et al. Mechanistic studies of stress corrosion cracking of nickel-based alloys in high temperature high pressure PWR environment [A]. Proceedings of the 14th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. LaGrange Park, IL: American Nuclear Society, 2009: 45-54
[21] 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: 91-99
[22] Bruemmer S M, Charlot L A, Henager C H. Microstructure and microdeformation effects on IGSCC of alloy-600 steam-generator tubing[J]. Corrosion, 1988, 44(11): 782-788
[23] Kozaczek K J, Sinharoy A, Ruud C O, et al. Micromechanical modelling of microstress fields around carbide precipitates in alloy 600[J]. Model Simul. Mater. Sci., 1995, 3(6): 829-843
[24] Randle V. The coincidence site lattice and the 'sigma enigma'[J]. Mater. Charact., 2001, 47(5): 411-416
[25] Lin P, Palumbo G, Erb U, et al. Influence of grain-boundary-character-distribution on sensitization and intergranular corrosion of alloy-600[J]. Scr. Mater., 1995, 33(9): 1387-1392
[26] Randle V. Electron backscatter diffraction: strategies for reliable data acquisition and processing[J]. Mater. Charact., 2009, 60(9): 913-922
[27] 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(9): 1589-1598
[28] Lehockey E M, Brennenstuhl A M, Thompson I. On the relationship between grain boundary connectivity, coincident site lattice boundaries and intergranular stress corrosion cracking[J]. Corros. Sci., 2004, 46: 2383-2404
[29] Kumar M, King W E, Schwartz A J. Modifications to the microstructural topology in F.C.C. materials through thermomechanical processing[J]. Acta Mater., 2000, 48(9): 2081-2091
[30] Terachi T, Yamada T, Miyamoto T, et al. SCC growth behaviors of austenitic stainless steels in simulated PWR primary water[J]. J. Nucl. Mater., 2012, 426(1-3): 59-70
[31] Jiao Z, Was G S. Impact of localized deformation on IASCC in austenitic stainless steels[J]. J. Nucl. Mater., 2011, 408(3): 246-256
[32] Couvant T, Moulart P, Legras L, et al. Effect of strain-hardening on stress corrosion cracking of AISI 304Lstainless steel in PWR environment at 360 degree [A]. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Warrendale, PA: TMS, 2005: 1069-1079
[33] Kamaya M. Influence of bulk damage on crack initiation in low-cycle fatigue of 316 stainless steel[J]. Fatigue Fract. Eng. Mater. Struc., 2010, 33(2): 94-104
[34] Couvant T, Legras L, Pokor C, et al. Investigations on the mechanisms of PWSCC of strain hardened austenitic stainless steels [A]. Proceedings of the 13th International Conference on Environmental Degradation of Materials in Nuclear Power Systems [C]. Toronto, CN: CNS, 2007: 1-16
[35] Hou J, Shoji T, Lu Z P, et al. Residual strain measurement and grain boundary characterization in the heat-affected zone of a weld joint between alloy 690TT and alloy 52 [J]. J. Nucl. Mater., 2010, 397(1-3): 109-115
[36] Couvant T, Vaillant F. Initiation of PWSCC of weld alloy 182 [A]. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Warrendale, PA: TMS, 2011: 1141-1151
[37] Andresen P L, Ford F P. Fundamental modeling of environmental cracking for improved design and lifetime evaluation in BWRs[J]. Int. J. Pres. Ves. Pip., 1994, 59(1-3): 61-70
[38] Andresen P L, Ford F P. Life prediction by mechanistic modeling and system monitoring of environmental cracking of iron and nickel-alloys in aqueous systems[J]. Mater. Sci. Eng., 1988, A103(1): 167-184
[39] Peng Q J, Kwon J, Shoji T. Development of a fundamental crack tip strain rate equation and its application to quantitative prediction of stress corrosion cracking of stainless steels in high temperature oxygenated water[J]. J. Nucl. Mater., 2004, 324(1): 52-61
[40] Shoji T, Lu Z, Murakami H. Formulating stress corrosion cracking growth rates by combination of crack tip mechanics and crack tip oxidation kinetics[J]. Corros. Sci., 2010, 52(3): 769-779
[41] Andresen P L, Reid R, Wilson J. SCC mitigration of Ni alloys and weld metals by op-timizing dissolved hydrogen [A]. Proceedings of the 14th Internati-onal Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Warrendale, PA: TMS, 2009: 345-372
[42] Peng Q, Hou J, Sakaguchi K, et al. Effect of dissolved hydrogen on corrosion of inconel alloy 600 in high temperature hydrogenated water[J]. Electrochim. Acta, 2011, 56(24): 8375-8386
[43] Combrade P, Scott P, Foucault M, et al. Oxidation of Ni base alloys in PWR water oxide layers and associated damage to the base metal [A]. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power System Water Reactors [C]. Warrendale, PA: TMS, 2005: 883-890
[44] Meng F, Lu Z, Shoji T, et al. Stress corrosion cracking of uni-directionally cold worked 316NG stainless steel in simulated PWR primary water with various dissolved hydrogen concentrations[J]. Corros. Sci., 2011, 53(8): 2558-2565
[45] Rocher A, Cassagne T, Durbec V, et al. The influence of chemical factors on the initiation of primary side IG-SCC in alloy 600 steam generator tubing [A]. Colloque International [C]. Fontevraud, 1994: 337-346
[46] Norring K. Influence of LiOH and H2 on Primary side IGSCC of Alloy 600 Steam Generator Tubes [M]. Studsvik AB: Studsvik Energy, 1990
[47] Jacko R, Economy G, Pement F. The influence of dissolved hydrogen on primary water stress corrosion cracking of alloy 600 at PWR steam generator operating temperatures [A]. Proceedings of the 5th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [C]. LaGrange Park, IL: American Nuclear Society, 1992: 613-620
[48] Cassange T, Fleury S, Vaillant F, et al. An update on the influence of hydrogen on the PWSCC of nickel base alloys in high temperature water [A]. Proceedings of the 9th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. LaGrange Park, IL: American Nuclear Society, 1997: 307-315
[49] Rebak R B,Szklarskasmialowska Z. Influence of stress intensity and loading mode on IASCC of alloy 600 in primary water of pressurized water reactors [J]. Corrosion, 1994, 50(5): 378-393
[50] 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
[51] Hwang S S, Kim H P, Lim Y S, et al. Transgranular SCC mechanism of thermally treated alloy 600 in alkaline water containing lead[J]. Corros. Sci., 2007, 49(10): 3797-3811
[52] Agrawal A K, Paine J P N. Lead cracking of alloy 600-a review [A]. Proceedings of the 4th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Houston, TX: NACE, 1989: 7-1-7-17
[53] Yang I J. Effect of sulphate and chloride ions on the crevice chemistry and stress corrosion cracking of alloy 600 in high temperature aqueous solutions[J]. Corros. Sci., 1992, 33(1): 25-37
[54] 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, 347(1/2): 52-68
[55] Stellwag B. The mechanism of oxide film formation on austenitic stainless steels in high temperature water[J]. Corros. Sci., 1998, 40(2/3): 337-370
[56] Wang S C, Takeda Y, Shoji T, et al. Observation of the oxide film formed in high temperature water by applying electroless Ni-P coating[J]. J. Nucl. Sci. Technol., 2004, 41(7): 777-779
[57] Terachi T, Fujii K, Arioka K. Microstructural characterization of SCC crack tip and oxide film for SUS 316 stainless steel in simulated PWR primary water at 320 ℃[J]. J. Nucl. Sci. Technol., 2005, 42(2): 225-232
[58] Soulas R, Cheynet M, Rauch E, et al. TEM investigations of the oxide layers formed on a 316L alloy in simulated PWR environment[J]. J. Mater. Sci., 2013, 48(7): 2861-2871
[59] Kuang W J, Wu X Q, Han E-H. Influence of dissolved oxygen concentration on the oxide film formed on 304 stainless steel in high temperature water[J]. Corros. Sci., 2012, 63: 259-266
[60] Kuang W J, Han E-H, Wu X Q, et al. Microstructural characteristics of the oxide scale formed on 304 stainless steel in oxygenated high temperature water[J]. Corros. Sci., 2010, 52(11): 3654-3660
[61] Kuang W J, Wu X Q, Han E-H. The oxidation behaviour of 304 stainless steel in oxygenated high temperature water[J]. Corros. Sci., 2010, 52(12): 4081-4087
[62] Li X H, Wang J Q, Han E-H, et al. Corrosion behaviour for alloy 690 and alloy 800 tubes in simulated primary water[J]. Corros. Sci., 2013, 67: 169-178
[63] Liu X H, Wu X Q, Han E-H. Influence of Zn injection on characteristics of oxide film on 304 stainless steel in borated and lithiated high temperature water[J]. Corros. Sci., 2011, 53(10): 3337-3345
[64] Ziemniak S E, Hanson M. Corrosion behavior of 304 stainless steel in high temperature, hydrogenated water[J]. Corros. Sci., 2002, 44(10): 2209-2230
[65] Neves C F C, Alvial G M, Schvartzman M M A, et al. Characterisation of oxide films formed on alloy 600 in simulated PWR primary water[J]. Energ. Mat., 2008, 3(2): 126-131
[66] Liu J H, Mendonca R, Bosch R W, et al. Characterization of oxide films formed on alloy 182 in simulated PWR primary water[J]. J. Nucl. Mater., 2009, 393(2): 242-248
[67] Machet A, Galtayries A, Marcus P, et al. XPS study of oxides formed on nickel-base alloys in high-temperature and high-pressure water[J]. Surf. Interface Anal., 2002, 34(1): 197-200
[68] Machet A, Galtayries A, Zanna S, et al. XPS and STM study of the growth and structure of passive films in high temperature water on a nickel-base alloy[J]. Electrochim. Acta, 2004, 49(22/23): 3957-3964
[69] Panter J, Viguier B, Cloue J M, et al. Influence of oxide films on primary water stress corrosion cracking initiation of alloy 600[J]. J. Nucl. Mater., 2006, 348(1/2): 213-221
[70] Zhang Z M, Wang J Q, Han E-H, et al. Influence of dissolved oxygen on oxide films of alloy 690TT with different surface status in simulated primary water[J]. Corros. Sci., 2011, 53(11): 3623-3635
[71] Huang F, Wang J, Han E-H, et al. Microstructural characteristics of the oxide films formed on alloy 690 in pure and primary water at 325 ℃[J]. Corros. Sci., 2013, 76: 52-59
[72] Vermilye D A. A theory for propagation of stress-corrosion cracks in metals[J]. J. Electrochem. Soc., 1972, 119(4): 405-407
[73] Turnbull A. Modeling of environment assisted cracking[J]. Corros. Sci., 1993, 34(6): 921-960
[74] Ford F P. Quantitative prediction of environmentally assisted cracking[J]. Corrosion, 1996, 52(5): 375-395
[75] Macdonald D D, Urquid-Macdonald M. A coupled environment model for stress-corrosion cracking in sensitized type-304 stainless-steel in LWR environments[J]. Corros. Sci., 1991, 32(1): 51-81
[76] Rebak R B, Szklarskasmialowska Z. The mechanism of stress corrosion cracking of alloy 600 in high temperature water[J]. Corros. Sci., 1996, 38(6): 971-988
[77] Scenini F, Newman R C, Cottis R A, et al. Alloy oxidation studies related to PWSCC [A]. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power System Water Reactors [C]. Warrendale, PA: TMS, 2005: 891-902
[78] Bruemmer S M, Thomas L E. Insights into Environmental Degradation Mechanisms from High-Resolution Characterization of Crack Tips [M]. Warrendale: Minerals, Metals & Materials Society, 2001
[79] Bruemmer S M,Thomas L. Insights into stress corrosion cracking mechanisms from high-resolution measurements of crack-tip structures and compositions [A]. MRS Proceedings [C]. Cambridge University Press, 2010: 1264-BB01-09
[80] Kamaya M, Haruna T. Influence of local stress on initiation behavior of stress corrosion cracking for sensitized 304 stainless steel[J]. Corros. Sci., 2007, 49(8): 3303-3324
[81] 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(3): 1137-1142
[82] Lu B T, Chen Z K, Luo J L, et al. Pitting and stress corrosion cracking behavior in welded austenitic stainless steel[J]. Electrochim. Acta, 2005, 50(6): 1391-1403
[83] Peng Q J, Shoji T, Ritter S, et al. SCC behaviour in the transition region of an alloy 182-SA 508 Cl.2 dissimilar weld joint under simulated BWR-NWC conditions [A]. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors [C]. Warrendale, PA: TMS, 2005: 589-599
[84] Kim J W, Lee K, Kim J S, et al. Local mechanical properties of alloy 82/182 dissimilar weld joint between SA508 Gr.1a and F316 SS at RT and 320 ℃[J]. J. Nucl. Mater., 2009, 384(3): 212-221
[85] Lee H T, Wu J L. Correlation between corrosion resistance properties and thermal cycles experienced by gas tungsten arc welding and laser beam welding alloy 690 butt weldments[J]. Corros. Sci., 2009, 51(4): 733-743
[86] Peng Q J, Xue H, Hou J, et al. Role of water chemistry and microstructure in stress corrosion cracking in the fusion boundary region of an alloy 182-A533Blow alloy steel dissimilar weld joint in high temperature water[J]. Corros. Sci., 2011, 53(12): 4309-4317
[87] Han E-H. Research trends on micro and nano-scale materials degradation in nuclear power plant[J]. Acta Metall. Sin., 2011, 47(7): 769-776
[87] (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(7): 769-776)
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