The microstructure and stress corrosion cracking (SCC) behavior of scratched zone on alloy 690TT were studied by using microhardness, TEM, EBSD-OIM and immersion experiment in caustic solution. It was found that a deformed hardening layer with a dimension range of 100 $\mu$m was produced near the scratch. TEM and EBSD-OIM observations showed that the grains at shallow surface of scratch groove were refined to nano-size. SCC tests for scratched alloy 690TT were performed in caustic solution at high temperature with or without addition of lead oxides. The results showed that SCC cracks initiated and propagated at scratch banks and scratch grooves. Grain boundaries, twin boundaries deformed and microcracks produced during scratching process are preferential sites for SCC. The oxide films formed on scratch groove were loosed by lead. The SCC crack length increased with increase of lead content. Scratched alloy 690TT is susceptible to SCC.
YU Jian-Qiu
,
MENG Fan-Jiang
. MICROSTRUCTURE NEAR SCRATCH ON ALLOY 690TT AND STRESS CORROSION INDUCED BY SCRATCHING[J]. Acta Metall Sin, 2011
, 47(7)
: 839
-846
.
DOI: 10.3724/SP.J.1037.2011.00213
[1] Ding X S. Nucl Power Plants, 2003; 4: 11
(丁训慎. 核电站, 2003; 4: 11)
[2] Blanchet J, Coriou H, Grall L, Mahieu C, Otter C, Turluer G. In: Staehle R W, Hochmann J, McCright R D, Slater J E, eds., Stress Corrosion Cracking and Hydrogen Embrittlement of Iron Base Alloys, No.5, Houston: NACE, 1977: 1149
[3] Staehle R W, Gorman J A. Corrosion, 2003; 59: 931
[4] Laboratory Evaluation of Steam Generator Tube Sections from the McGuire Nuclear Station. Babcock & Wilcox, Report RDD:91:5087–01:01, October, 1990: 1
[5] Examination of Steam Generator Tube Sections from the McGuire Unit 2 Nuclear Station. Babcock & Wilcox–NESI, Report LTC:93:0021–01:01, January, 1993: 1
[6] Analysis of Steam Generator Tubing from Oconee Unit 1 Nuclear Station. EPRI report TR–106484
[7] Oconee 2 Steam Generator Tube Examination. EPRI report TR–106863
[8] MacDonald P E, Shah V N,Ward LW, Ellison P G. Steam Generator Tube Failures. NUREG/CR–6365
[9] Rochester D P. Water Chemistry of Nuclear Reactor Systems. 1st Ed., London: Thomas Telford Publishing, 2001: 1
[10] Staehle R W, Gorman J A. Corrosion, 2004; 60: 115
[11] Staehle R W, Gorman J A. Corrosion, 2004; 60: 5
[12] Meng F J, Wang J Q, Han E H, Ke W. Corros Sci, 2009; 51: 2761
[13] Tkaya M B, Zidi M, Mezlini S, Zahouani H, Kapsa P. Mater Des, 2008; 29: 98
[14] Tao N R, Wang Z B, Tong W P, Sui M L, Lu J, Lu K. Acta Mater, 2002; 50: 4603
[15] Symons D. Metall Mater Trans, 1997; 28A: 655
[16] Tao N R, Zhang H W, Lu J, Lu K. Metall Trans, 2003; 44: 1919
[17] Hwang S, Kim H, Lim Y, Kim Y, Thomas L. Corros Sci, 2007; 49: 3797
[18] Hwang S, Kim H, Lee D, Kim U, Kim J. J Nucl Mater, 1999; 275: 28
[19] Lumsden J, McIlree A, Eaker R, Thompson R, Slosnerick S. Proc 12th Int Conf Environmental Degradation of Materials in Nuclear Power System–Water Reactors, Salt Lake City, Utah, USA, 2005: 1
[20] Meng F J, Han E H, Wang J Q, Zhang Z M, Ke W. Electrochim Acta, 2011; 56: 1781
