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| CORROSION BEHAVIOR OF ALLOY 690TT IN HIGH TEMPERATURE LEAD-CONTAINING CAUSTIC SOLUTION |
| HU Yisong, WANG Jianqiu, KE Wei, HAN En-hou |
| State Key Laboratory for Corrosion and Protection, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 |
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Abstract A small amount of PbO could induce the passive film of alloy 690TT unstable in the caustic solutions at room temperature. The immersion tests were carried out in high temperature lead-containing caustic solutions in a static autoclave. The results showed that alloy 690TT immersed in 10% NaOH+10 g/L PbO solution suffered from intergranular attack (IGA) at 330℃, and the specimens lost weight. The coarser the original surface of the samples, the more serious the samples suffered from corrosion. IGA was observed and developed as intergranular stress corrosion (IGSCC) with respect of residual stress. While no IGA was found for the samples immersed in 10% NaOH solution at 330℃.
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Received: 29 September 2009
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Corresponding Authors:
WANG Jianqiu
E-mail: jiqwang@imr.ac.cn
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| [1] Was G S. Grain-boundary chemistry and intergranular fracture in austenitic nickel-base alloys-a review [J]. Corrosion, 1990, 46(4):319-330[2] Ding X S. Corrosion and protection of nuclear power plants steam generator tubes [J].Corros. Prot., 2000, 21(1): 15-18 (丁训慎. 核电站蒸汽发生器传热管的腐蚀与防护 [J]. 腐蚀与防护, 2000, 21(1): 15-18)[3] Ding X S. Cleaning technique for steam generators in nuclear power station [J]. Cleaning World, 2004, 20(4): 32-36 (丁训慎. 核电站蒸汽发生器传热管的清洗技术 [J]. 清洗世界, 2004,20(4): 32-36)[4] Ding X S. The operation safety of steam generators in NPP [J]. Nucl. Safety, 2004, (4): 29-34(丁训慎. 核电厂蒸汽发生器运行中的安全问题, 核安全, 2004, (4): 29-34)[5] Staehle R W, Gorman J A. 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[6] Staehle R W, Gorman J A. Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors: Part 2 [J].Corrosion, 2004, 60(1): 5-63[7] Staehle R W, Gorman J A.Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors: Part3 [J]. Corrosion, 2004, 60(2): 115-180[8] Miglin B P, Sarver J M. Preliminary studies of lead stress corrosion cracking of alloy 690 [A]. Proceeding of the Fourth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors [C]. Jekyll Island, GA: 1989: 7-18[9] Gomez B D, Castno M L, Garcia M S. Stress corrosion cracking susceptibility of steam generator tube materials inAVT(all volatile treatment) chemistry contaminated with lead [J]. Nucl. Eng. Design,1996, 165(1): 161-169[10] Kim U C, Kim K M, Lee E H. Effects of chemical compounds on the stress corrosion cracking of steam generator tubing materials in a caustic solution [J]. J. Nucl. Mater., 2005, 341: 169-174[11] Cao C N. Corrosion Electrochemistry [M]. Beijing: Chemical Industry Press, 2004(曹楚南. 腐蚀电化学原理 [M]. 北京：化学工业出版社, 2004)[12] Ahn S J, Rao V S, Kwon H S, et al. Effects of PbO on the repassivation kinetics of alloy 690 [J]. Corros. Sci.,2006, 48(5): 1137-1153[13] 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[14] Hwang S S, Kim J S. Electrochemical interaction of lead with Alloy 600 and Alloy 690 in high-temperature water [J]. Corrosion, 2002, 58(5): 392-398[15] Montemor M F, Ferreira M G S, Walls M, et al. Influnence of pH on properties of oxide films formed on type 316L stainless steel, alloy 600, and alloy 690 in high-temperature aqueous environments [J]. Corrosion,2003, 59(1): 11-21[16] Peng B, Lu B T, Lu Y C, et al. Investigation of passive films on nickel alloy 690 in lead-containing environments [J]. J. Nucl. Mater., 2008, 378(3): 333-340[17] Ding X S. IGA/IGSCC and protection on secondary side of steam generator tubes in nuclear power plants [J]. Corros. Prot., 2002,23(10): 441-444 (丁训慎. 核电站蒸汽发生器传热管二次侧晶间腐蚀和晶间应力腐蚀及防护 [J]. 腐蚀与防护, 2002, 23(10): 441-444)[18] Thomas L E, Bruemmer S M. High-resolution characterization of intergranular attack and stress corrosion cracking of alloy 600 in high-temperature primary water [J]. Corrosion, 2000,56(6): 572-587[19] Miglin B P, Sarver J M, Psaila-Dombrowksi M J, et al. Lead assisted stress corrosion cracking of nuclear steam generator tubing materials [A]. Proceeding of Improving the Understanding and Control of Corrosion on the Secondary Side of Steam Generator [C]. Airlie, NACE: 1995: 305-320 |
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