Correlation of laboratory and field stress corrosion results in the power generation industry

NACE - International Corrosion Conference Series

Volumn 1999-April, Issue , 1999, Pages

  Speidel, Markus O ^a   Magdowski, Ruth ^a  

^a ETH ZURICH   (Switzerland)

Author keywords

Alloy 600; Austenitic steel; Crack growth rate; Service experience; Stress corrosion cracking

Indexed keywords

AUSTENITE; AUSTENITIC STAINLESS STEEL; AUSTENITIC STEEL; BOILING WATER REACTORS; CRACK PROPAGATION; GROWTH RATE; HIGH STRENGTH ALLOYS; PIPELINE CORROSION; RESIDUAL STRESSES; STEEL CORROSION; STRESS CORROSION CRACKING;

ALLOY 600; BOILING WATER REACTOR (BWR); FIELD EXPERIENCE; GRAPHITE-MODERATED; NUCLEAR REACTOR COOLANT; POWER GENERATION INDUSTRIES; POWER-GENERATING INDUSTRY; SERVICE EXPERIENCE;

PRESSURIZED WATER REACTORS;

EID: 0013006179     PISSN: 03614409     EISSN: None     Source Type: Book Series    
DOI: None     Document Type: Conference Paper

References (13)

1. Ruth Magdowski and Markus O. Speidel: "Clean steam turbine rotor steels - their stress corrosion resistance", Metall. Trans, vol. 19A, 1988, pp. 1583-1596.
2. M. O. Speidel and R. Magdowski, "Stress Corrosion Cracking of TT- and Nb-stabilized Austenitic Stainless Steels in 288°C Water", CORROSION '94, Paper 132, NACE, 1994.
3. Ruth Magdowski and Markus O. Speidel, "Effect of cold work on the growth rates of stress corrosion cracks in structural materials of nuclear systems", CORROSION '96, Paper 112, NACE, 1996.
4. Ruth Magdowski, François Vaillant, Claude Amzallag, and Markus O. Speidel, "Stress corrosion crack growth rates of alloy 600 in simulated PWR coolant", Proceedings of the Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Aug. 10-14, 1997, Amelia Island, Florida, USA, pp. 333-339.
5. Proceedings of International Symposium Fontevraud III, Contribution of Materials Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors, SFEN, Fontevraud, France, September 1994.
6. Proceedings of the Fourth EPRI Workshop on PWSCC of alloy 600 in PWR's, Feb. 25-27, 1997, Daytona Beach, FL, USA.
7. C. Amzallag, F. Vaillant, A. Teissier," Derivation of stress corrosion crack growth laws for alloy 600 PWR components", Proceedings of the Fourth EPRI Workshop on PWSCC of alloy 600 in PWR's, Feb. 25-27, 1997, Daytona Beach, FL, USA.
8. P. M. Scott, "An analysis of primary water stress corrosion cracking in PWR steam generator, NEA/CSNI UNIPEDE specialist meeting on operating experience with steam generators, Bruxelles, Belgium, 1991.
9. F. Vaillant, C. Amzallag, J. Champredonde, "Crack growth rate measurements of alloy 600 vessel head penetrations", Proceedings of the Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, Aug. 10-14, 1997, Amelia Island, Florida, USA, pp. 357-365.
10. P. M. Scott et al, "Stress corrosion crack growth of nickel alloys in PWR primary coolant", AEA Technology report 1205, December 1996.
11. W. H. Bamford, J.-P. Foster, and R. S. Pathania, "Characterization of alloy 600 static load crack growth in PWR environments", Proceedings of International Symposium Fontevraud IV, Contribution of Materials Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors, SFEN, Fontevraud, France, September 1998, pp. 939-957.
12. Markus O. Speidel, "Intergranular and transgranular stress corrosion cracking of mestallic structural materials in nuclear power plants", Proceedings of VGB-Conference Corrosion and corrosion protection in power generation technology, VGB-TB 513, VGB Kraftwerkstechnik GmbH, Essen, 1996, pp. 1-12.
13. Ruth Magdowski and Markus O. Speidel, "Generator rotor retaining rings: Stress corrosion problems and solurtions", Proceedings of International Symposium Fontevraud IV, Contribution of Materials Investigation to the Resolution of Problems Encountered in Pressurized Water Reactors, SFEN, Fontevraud, France, September 1998, pp. 755-763.