Browse > Article

Stress Corrosion Cracking of Alloy 600 and Alloy 690 in Caustic Solution  

Kim, Hong Pyo (Korea Atomic Energy Research Institute)
Lim, Yun Soo (Korea Atomic Energy Research Institute)
Kim, Joung Soo (Korea Atomic Energy Research Institute)
Publication Information
Corrosion Science and Technology / v.2, no.2, 2003 , pp. 82-87 More about this Journal
Abstract
Stress corrosion cracking of Alloy 600 and Alloy 690 has been studied with a C-ring specimen in 1%, 10% and 40% NaOH at $315^{\circ}C$. SCC test was performed at 200 mV above corrosion potential. Initial stress on the apex of C-ring specimen was varied from 300 MPa to 565 MPa. Materials were heat treated at various temperatures. SCC resistance of Ni-$_\chi$Cr-10Fe alloy increased as the Cr content of the alloy increased if the density of an intergranular carbide were comparable. SCC resistance of Alloy 600 increased in caustic solution as the product of coverage of an intergranular carbide in grain boundary, intergranular carbide thickness and Cr concentration at grain boundary increased. Low temperature mill annealed Alloy 600 with small grain size and without intergranular carbide was most susceptible to SCC. TT Alloy 690 was most resistant to SCC due to the high value of the product of coverage of an intergranular carbide in grain boundary, intergranular carbide thickness and Cr concentration at grain boundary. Dependency of SCC rate on stress and NaOH concentration was obtained.
Keywords
stress corrosion cracking; Alloy 600; Alloy 690; NaOH;
Citations & Related Records
연도 인용수 순위
  • Reference
1 W. Staehle, Basis for Predicting the Earliest Penetrations due to SCC for Alloy 600 on the Secondary Side of PWR Steam Generators, Presented at Seminar at KAERI, Korea, 2001
2 R. J. Jacko, in Corrosion Evaluation of Thermally Treated Alloy 600 Tubing in Primary and Faulted Secondary Water Environments, EPRI NP-6721-ISD, 1990
3 K. Yamanaka and J. Murayama, in Proc. of the 4th Intern. Symp. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Georgia, USA, 1989, p.5
4 F. Vaillant, E-M Pavageau, M, Bouchcourt, J-M Boursier, and P. Lemaire, in Proceedings of 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Amelia Island, FL, 1997, p.679
5 PWR Secondary Water Chemistry Guidelines Revision Committee, PWR Secondary Water Chemistry Guidelines - Revision 3, EPRI TR-102134, Revision 3, 1993
6 R. Bandy and D. van Rooyen, Nuclear Engineering and Design, 86, 49 (1985)
7 T. Yonezawa, N. Sagaguri, and K, Inimura, in Proceedings of JAIF International Conference in Water Chemistry in Nuclear Power Plants, Tokyo, Japan, 1988, p.490
8 D. C. Crawford and G. S. Was, Met. Trans. A, 23A(4), 1195 (1992)
9 L. E. Murr, Interfacial Phenomena in a Metals and Alloys, Addison Wesley Publishing Company 1975, p.145
10 P. Lin, G. Palumbo, U. Erb, and K. T. Aust, Scripta Metallurgical Materialia, 33(9), 1387 (1995)
11 C. Goffin, J. Jadot, and L. Duvivier, in Proceedings of International Symposium Fontevraud IV, Fontevraud France, 1998, p.441
12 J. K. Sung, J. Koch, T. Angeliu, and G. S. Was, Metall. Trans.A, 23A, 2887 (1992)
13 J. P. N. Paine, S. A. Hobart, and S. G. Sawochka, in Proceedings of the 5th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, Monterey, CA, 1991, p.739
14 Korea Instutute of Nuclear Safety, Actions for Steam generator Tube IGSCC in YoungKwang Unit 4, 2001
15 P. E. Doherty, D. M. Doyle, j. M. Sarver, and B. P. Miglin, in Proceedings of a Conf. on Control of Corrosion on the Secondary Side of Steam Generators, Airlie, Virginia, USA, 1995 p.401
16 P.M. Scott and P.Combrade, in Proceedings of the 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-WaterReactors, Amelia Island, FL,1997, p.65