Please wait a minute...
Just Accepted | Current Issue | Archive | Featured Articles | Special Issue | Most Read | Most Downloaded | Most Cited
Journal of Chinese Society for Corrosion and protection  2015, Vol. 35 Issue (3): 189-198    DOI: 10.11902/1005.4537.2014.101
Current Issue | Archive | Adv Search |
Review on SCC Crack Growth Behavior of Dissimilar Metal Welds for Nuclear Power Reactors
Ruolin ZHU,Zhiming ZHANG,Jianqiu WANG(),En-Hou HAN
Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, China
Download:  HTML  PDF(3326KB) 
Export:  BibTeX | EndNote (RIS)      
Abstract  

Recent progress on stress corrosion cracking (SCC) crack growth behavior of dissimilar metal welds between reactor pressure vessel and reactor coolant piping is reviewed in this paper. The worldwide studies concerning the effect of materials, stress and environmental conditions on SCC crack growth behavior are described. Meanwhile, the direction of further research is also forecasted.
Key words:  light water reactor      dissimilar metal weld      stress corrosion crack      crack growth      high temperature and high pressure water     
Service
E-mail this article
Add to citation manager
E-mail Alert
RSS
（）
Articles by authors
Ruolin ZHU
Zhiming ZHANG
Jianqiu WANG
En-Hou HAN

Cite this article: 

Ruolin ZHU,Zhiming ZHANG,Jianqiu WANG,En-Hou HAN. Review on SCC Crack Growth Behavior of Dissimilar Metal Welds for Nuclear Power Reactors. Journal of Chinese Society for Corrosion and protection, 2015, 35(3): 189-198.

URL: 

https://www.jcscp.org/EN/10.11902/1005.4537.2014.101     OR     https://www.jcscp.org/EN/Y2015/V35/I3/189

Fig.1  Representative nozzle to piping weld indicating the location of the dissimilar mental weld[2]
Fig.2  J-groove weld on a reactor-vessel head[3]
Fig.3  Generic location of Alloy 182/82 welds[3]
Fig.4  Intergranular SCC fracture morphologies of annealed (or as-deposited): (a) Type 304L stainless steel+20% cold work at -55 ℃, (b) annealed Alloy 600+20% cold work, (c) Alloy 182 weld metal [8]
Plant Inspection date Nozzle Type of indication Indication depth a / cm OD indication length l / cm a: thickness l: circumference
Calvert cliffs 2 2005 CL drain Cric 0.142 1.595 10% 10%
Calvert cliffs 2 2005 HL drain Axial 0.996 0.000 70% 0%
DC cook 2005 Safety Axial 3.129 0.000 88% 0%
Calvert cliffs 1 2006 HL drain Cric 0.254 1.143 19% 5%
Calvert cliffs 1 2006 Relief Axial 0.254 0.000 8% 0%
Calvert cliffs 1 2006 Surge Circ 1.016 6.096 25% 6%
Davis besse 2006 CL drain Axial 0.142 0.000 7% 0%
San onofre 2 2006 Safety Axial 1.067 0.000 30% 0%
San onofre 2 2006 Safety Axial 1.067 0.000 30% 0%
Wolf creek 2006 Relief Cric 0.864 29.210 25.8% 46%
Wolf creek 2006 Safety Cric 0.754 6.350 22.5% 10%
Wolf creek 2006 Surge Cric 1.181 22.225 32.1% 19%
Farley 2 2007 Surge Circ 1.270 7.620 33% 6%
Davis besse 2008 --- Axial --- --- --- ---
Crystal river 3 2008 --- Cric --- --- --- ---
Table 1  Cracking indications detected in reactor coolant loop alloy 182/82 butt welds, 2005 through mid-2008[16]
Fig.5  SEM images showing propagation of stress corrosion crack along the type-II and type-I boundaries before (a) and after (b) the crack reached the FB[25]
Fig.6  Ni-H2O Pourbaix diagram at 300 ℃[53]
Fig.7  Effect of H2 fugacity on the crack growth rate of Ni alloys in high-temperature water[53]
[1] Rebak R B. 2012 Research topical symposium proceedings "Corrosion degradation of materials in nuclear power reactors-lessons lear-ned future challenges" introduction[J]. Corrosion, 2013, 69(10):951
[2] Kerr M, Hill M R, Olson M D. Study of residual stresses in compact tension specimens fabricated from weld metal[J]. Corrosion, 2013, 69(10): 975
[3] Celin R,Tehovnik F.Degradation of a Ni-Cr-Fe alloy in a pressurised-water nuclear power plant [J]. Mater. Technol, 2011, 45(2): 151
[4] Tsuruta T,Sato K,Asada S,et al. PWSCC of nickel base alloys in vapor phase environment of pressurizer [A]. Icone16: Proceeding of the 16th International Conference on Nuclear Engineering [C]. Orlando: ASME, 2008: 571
[5] Chung W C, Huang J Y, Tsay L W, et al. Microstructure and stress corrosion cracking behavior of the weld metal in alloy 52-A508 dissimilar welds[J]. Mater. Trans. JIM, 2011, 52(1): 12
[6] 0 Li G F, Congleton J. Stress corrosion cracking of a low alloy steel to stainless steel transition weld in PWR primary waters at 292 ℃[J]. Corros. Sci., 2000, 42(6): 1005
[7] Muransky O, Smith M C, Bendeich P J, et al. Validated numerical analysis of residual stresses in Safety Relief Valve (SRV) nozzle mock-ups[J]. Comput. Mater. Sci., 2011, 50(7): 2203
[8] Andresen P L.Emerging issues and fundamental processes in environmental cracking in hot water (Reprinted from proceedings of the CORROSION/2007 research topical symposium"Advances in environmentally assisted cracking", 2007)[J]. Corrosion, 2008, 64(5): 439
[9] Cattant F, Crusset D, Feron D. Corrosion issues in nuclear industry today[J]. Mater. Today, 2008, 11(10): 32
[10] Andresen P L, Morra M M. Stress corrosion cracking of stainless steels and nickel alloys in high-temperature water[J]. Corrosion, 2008, 64(1): 15
[11] Paraventi D J, Moshier W C.The effect of cold work and dissolved hydrogen in the stress corrosion cracking of Alloy 82 and Alloy 182 weld metal [A]. Proc. 12th Int. Symp. Environmental Degradation of Materials in Nuclear Power System [C]. Warrendale: TMS, 2005: 543
[12] Alexandreanu B,Chen Y,Natesan K,et al. SCC behavior of Alloy 690 HAZ in a PWR environment [A]. Proceedings of the Asme Pressure Vessels and Piping Conference [C]. Baltimore: ASME, 2012: 385
[13] Amzallag C,Boursier J,Pages C,et al. Stress corrosion life experience of 182 and 82 welds in French PWRs [A]. 5th Fontevraud Conf. Contribution of Material Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors [C]. Fontevraud, 2002: 22
[14] Scott P M.An overview of materials degradation by stress corrosion in PWRs [A]. European Corrosion Conference: Long Term Prediction and Modelling of Corrosion, EUROCORR 2004 [C]. Nice: Cefracor, 2004: 3
[15] Bamford W, Hall J.A review of alloy 600 cracking in operating nuclear plants including alloy 82 and 182 weld behavior [A]. 12th International Conference on Nuclear Engineering (ICONE12) [C]. Arlington: ASME, 2004: 131
[16] Gorman J, Hunt S, Riccardella P. PWR Reactor Vessel Alloy 600 Issues [M]. New York: ASME, 2009: 63
[17] Yvon P, Carre F.Structural materials challenges for advanced reactor systems[J]. J. Nucl. Mater., 2009, 385(2): 217
[18] Young G A, Etien R A, Hackett M J, et al. Physical metallurgy, weldability, and in-service performance of nickel-chromium filler metals used in nuclear power systems [A]. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors [C]. Colorado Sprin-gs: Wiley Online Library, 2009: 2431
[19] Szklarskasmialowska S, Cragnolino G. Stress-corrosion cracking of sensitized type-304 stainless-steel in oxygenated pure water at elevated-temperatures (review)[J]. Corrosion, 1980, 36(12): 653
[20] Huang J Y, Liu R F, Chiang M F, et al. Corrosion fatigue behavior of dissimilar metal weldments under nominal constant Delta K loading mode in a simulated BWR coolant environment[J]. Corros. Sci., 2011, 53(6): 2289
[21] Huang J Y, Chiang M F, Jeng S L, et al. Environmentally assisted cracking behavior of dissimilar metal weldments in simulated BWR coolant environments[J]. J. Nucl. Mater., 2013, 432(1-3): 189
[22] Huang J Y, Yung T Y, Huang J S, et al. Effects of heat treatment and chromium content on the environmentally assisted cracking behavior of the dissimilar metal welds in simulated BWR coolant environments[J]. Corros. Sci., 2013, 75: 386
[23] Ozawa M, Yamamoto Y, Nakata K, et al. Evaluation of SCC crack growth rate in alloy 600 and its weld metals in simulated BWR environments [A]. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors [C]. Salt Lake City: TMS, 2006: 651
[24] Lee H T, Wu J L. Intergranular corrosion resistance of nickel-based alloy 690 weldments[J]. Corros. Sci., 2010, 52(5): 1545
[25] Hou J, Peng Q, Takeda Y, et al. Microstructure and stress corrosion cracking of the fusion boundary region in an alloy 182-A533B low alloy steel dissimilar weld joint[J]. Corros. Sci., 2010, 52(12): 3949
[26] Peng Q, 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-A533B low alloy steel dissimilar weld joint in high temperature water[J]. Corros. Sci., 2011, 53(12): 4309
[27] White G A, Nordmann N S, Hickling J, et al. Development of crack growth rate disposition curves for primary water stress corrosion cracking (PWSCC) of alloy 82, 182, and 132 weldments [A]. Proceedings of the 12th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors [C]. Salt Lake City: TMS, 2006: 511
[28] 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
[29] Guerre C, Chaumun E, Crepin J, et al. Stress corrosion cracking of nickel base alloys in PWR primary water [A]. 1st International Workshop on Materials Innovation for Nuclear Optimized Systems[C].CEA Saclay: EPJ Web of Conferences, 2013
null
[30] Yeh T K, huang G R, Wang M Y, et al. Stress corrosion cracking in dissimilar metal welds with 304L stainless steel and Alloy 82 in high temperature water[J]. Prog. Nucl. Energy, 2013, 63: 7
[31] Huang J Y, Chiang M F, Kuo R C, et al. Stress corrosion cracking behavior of dissimilar metal weldments in high temperature water environments [A]. Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [C]. Colorado Springs: TMS, 2011: 1105
[32] Saito N, Tanaka S, Sakamoto H. Effect of corrosion potential and microstructure on the stress corrosion cracking susceptibility of nickel-base alloys in high-temperature water[J]. Corrosion, 2003, 59(12): 1064
[33] Alexandreanu B,Chopra O K,Shack W J. The stress corrosion cracking behavior of alloys 690 and 152 weld in a PWR environment [A]. Pressure Vessel and Piping Division of the American Society of Mechanical Engineers [C]. Chicago: ASME, 2009: 153
[34] Alexandreanu B,Chopra O K,Shack W J. Crack growth rates of nickel alloy welds in a PWR environment [A]. ASME 2006 Pressure Vessels and Piping/ICPVT-11 Conference [C]. Vancouver: ASME, 2006: 153
[35] 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
[36] Kim S W, Kim H P, Jeong J U, et al. Effect of residual stress of dissimilar metal welding on stress corrosion cracking of bottom-mounted instrumentation penetration mock-up[J]. Corrosion, 2010 66(10): 106001
[37] Zhang T,Brust F W,Wilkowski G,et al. Welding residual stress in a large diameter nuclear reactor pressure vessel nozzle [J]. J. Press Vess-T ASME, 2013, 135(2): 021208
[38] Liu R F, Huang C C. Welding residual stress analysis for weld overlay on a BWR feedwater nozzle[J]. Nucl. Eng. Des., 2013, 256: 291
[39] Lu Z, Shoji T, Takeda Y, et al. The dependency of the crack growth rate on the loading pattern and temperature in stress corrosion cracking of strain-hardened 316L stainless steels in a simulated BWR environment[J]. Corros. Sci., 2008, 50(3): 698
[40] Andresen P L, Young L M, Emigh P W, et al. Stress corrosion crack growth rate behavior of ni alloys 182 and 600 in high temperature water [A]. Corrosion/2002 [C]. Denver: NACE, 2002: 1
[41] 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(1-3): 15
[42] Toloczko M B, Olszta M J, Bruemmer S M. Stress corrosion crack growth of alloy 52m in simulated PWR primary water [A]. 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-water Reactors [C]. Colorado Sprin-gs: TMS, 2011: 225
[43] Lu Z, Shoji T, Xue H, et al. Synergistic effects of local strain-hardening and dissolved oxygen on stress corrosion cracking of 316NG weld heat-affected zones in simulated BWR environments[J]. J. Nucl. Mater., 2012, 423(1-3): 28
[44] Lu Z, Shoji T, Meng F, et al. Characterization of microstructure and local deformation in 316NG weld heat-affected zone and stress corrosion cracking in high temperature water[J]. Corros. Sci., 2011, 53(5): 1916
[45] Alexandreanu B,Chen Y,Natesan K,et al. Cyclic and SCC behavior of alloy 152 weld in a PWR environment [A]. Proceedings of the ASME Pressure Vessels and Piping Conference [C]. Baltimore: AMSE, 2012: 639
[46] 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
[47] Zhanpeng L, Shoji T, Takeda Y, et al. Transient and steady state crack growth kinetics for stress corrosion cracking of a cold worked 316L stainless steel in oxygenated pure water at different temperatures[J]. Corros. Sci., 2008, 50(2): 561
[48] Andresen P L. Stress corrosion cracking of current structural materials in commercial nuclear power plants[J]. Corrosion, 2013, 69(10): 1024
[49] Stjarnsater J, Jenssen A, Jansson C, et al. The effect of temperature on the crack growth rate of stainless steel and ni-alloys in simulated BWR environment [A]. 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [C]. Colorado Springs: TMS, 2011: 827
[50] Hoang P H, Gangadharan A, Ramalingam S C. Primary water stress corrosion cracking inspection ranking scheme for alloy 600 components[J]. Nucl. Eng. Des., 1998, 181(1-3): 209
[51] 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(7): 648
[52] Lima L I L,Schvartzman M M A M,Figueiredo C A,et al. Stress corrosion cracking behavior of alloy 182 weld in pressurized water reactor primary water environment at 325 ℃ [J]. Corrosion, 2011, 67(8): 085004
[53] 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
[54] Andresen P L, Young L M. Crack-tip microsampling and growth-rate measurements in low-alloy steel in high-temperature water[J]. Corrosion, 1995, 51(3): 223
[55] 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]. Salt Lake City : TMS, 2006: 989
[56] Korb J, Stellwag B. Thermodynamics of zinc chemistry in PWRs: effects and alternatives to zinc[J]. Nucl. Energ-J. Br. Nucl., 1997, 36(5): 377
[57] 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
[58] Liu X H, Han E H, Wu X Q.Effect of Zn injection on established surface oxide films on 316 L stainless steel in borated and lithiated high temperature water[J]. Corros. Sci., 2012, 65: 136
[1] 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.
[2] 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.
[3] 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.
[4] 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.
[5] 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.
[6] 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.
[7] Jiapeng LIAO,Xinqiang WU. Review of Notch Effect on Fatigue Behavior of Materials for LWR Plants in High Temperature High Pressure Water[J]. 中国腐蚀与防护学报, 2018, 38(6): 511-516.
[8] 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.
[9] 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.
[10] Xiaocheng ZHOU, Qiaoqi CUI, Jinghuan JIA, Zhiyong LIU, Cuiwei DU. Influence of Cl- Concentration on Stress Corrosion Cracking Behavior of 316L Stainless Steel in Alkaline NaCl/Na2S Solution[J]. 中国腐蚀与防护学报, 2017, 37(6): 526-532.
[11] Naiqiang ZHANG,Guoqiang YUE,Fabin LV,Qi CAO,Mengyuan LI,Hong XU. Crack Growth Rate of Stress Corrosion Cracking of Inconel 625 in High Temperature Steam[J]. 中国腐蚀与防护学报, 2017, 37(1): 9-15.
[12] Jinheng LUO,Congmin XU,Dongping YANG. Stress Corrosion Cracking of X100 Pipeline Steel in Acid Soil Medium with SRB[J]. 中国腐蚀与防护学报, 2016, 36(4): 321-327.
[13] Ping DENG,Chen SUN,Qunjia PENG,En-Hou HAN,Wei KE. Review of Irradiation Assisted Stress Corrosion Cracking of Core Structural Materials[J]. 中国腐蚀与防护学报, 2015, 35(6): 479-487.
[14] Yueling GUO,En-Hou HAN,Jianqiu WANG. Effect of Post-forging Heat Treatment on Stress Corrosion Cracking of Nuclear Grade 316LN Stainless Steel in Boiling MgCl2 Solution[J]. 中国腐蚀与防护学报, 2015, 35(6): 488-495.
[15] Zhiming ZHANG,Qingjiao PENG,Jianqiu WANG,En-Hou HAN,Wei KE. Stress Corrosion Cracking Behavior of Forged 316L Stainless Steel Used for Nuclear Power Plants in Alkaline Solution at 330 ℃[J]. 中国腐蚀与防护学报, 2015, 35(3): 205-212.
No Suggested Reading articles found!

Copyright © 2017 Journal of Chinese Society for Corrosion and protection, All Rights Reserved.
Editorial Office: Journal of Chinese Society for Corrosion and Protection, 72 Wenhua Rd., Shenyang 110016, China
Tel: +86- 024-23971318, Fax: +86-024-23971819 　E-mail: jcscp@imr.ac.cn
Powered by Beijing Magtech