Stress corrosion cracking assessment

The technique used for determining SCC effects on component life is in-service inspection by volumetric means, such as ultrasonic imaging to detect and size flaws, and subsequent fracture mechanics evaluation to predict life remaining after initiation of the detected flaw.

Analytical evaluation is a useful tool for dispositioning detected and sized flaws, this method most often used for component life prediction is crack growth assessment, using laboratory crack growth data and analytical fracture mechanics stress intensity calculations.

SCC crack growth rates under various environmental conditions have been generated for the austenitic steels used for internals. The BWRVIP programme has also provided recommended IGSCC crack growth rates for use in dispositioning detected and sized flaws [6.25 – 6.28].

These data, in conjunction with geometry specific stress intensity solutions, can determine subsequent crack growth, assuming that a crack has been detected and has reached a measurable initial crack size. The method does not consider the time to crack initiation; it relies on inspection to detect cracks or on historical data to predict the time to initiate cracking.

For component life to be determined, a limit on the amount of crack growth (allowable crack size) for the particular component must be established. Crack growth calculations can then be performed to verify that the predicted crack size, for some period of continued

operation, is less than the allowable crack size. Component life can then be estimated by calculating the time at which the allowable crack size is reached. Repair or replacement options can be implemented at the predicted end of allowable component life.

REFERENCES IN SECTION 6

[6.1] Code of Federal Regulations, Title 10 — Energy Part 50, Domestic Licensing of (Nuclear Power) Production and Utilization Facilities.” Published by the Office of Federal Register, National Archives and Records Administration, Washington, DC.

Same as Reference [3.3].

[6.2] AMERICAN SOCIETY OF MECHANICAL ENGINEERS, ASME Boiler and Pressure Vessel Code, Section III, “Nuclear Power Plant Components”, Appendix G,

“Protection Against Non-ductile Failure”, ASME, New York (1998).

Same as reference [3.2].

[6.3] Nuclear Regulatory Guides, RG 1.99 (Revision 1), “Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials”, April 1987.

Same as reference [3.8] and [4.2]

[6.4] Nuclear Regulatory Guides, RG 1.99 (Revision 2), “Radiation Embrittlement Reactor Vessel Materials”, May 1988.

Same as reference [3.9] and [4.3]

[6.5] ASTM E 185-82, Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, “Annual Book of ASTM Standards, American Society for Testing and Materials, 1916 Race Street, Philadelphia PA 14103, July 1982.

Same as reference [3.7]

[6.6] WRC Bulletin 175, PVRC Recommendations on Toughness Requirements for Ferritic Materials, “Welding Research Council (WRC), 345 East 47^th Street, New York NY 10017, August 1972.

[6.7] NRC BTP MTEB5-2Rev.1, “Fracture Toughness Requirements”, US NRC Branch Technical Position, July 1981.

[6.8] EPRI NP-4797, “Nuclear Plant Irradiated Steel Handbook”, Electrical Power Research Institute, 3412 Hillview Avenue, P.O.Box 10412, Palo Alto CA 943303, September 1986.

[6.9] GE NEDC-31 1 40, ‘BWR Owners’ Group Evaluation of Regulatory Guide 1.99, Revision 2, Impact on BWRs, “GE Nuclear Energy, 175 Curtner Avenue, San Jose CA 95125, January 1986.

[6.10] BWRVIP-86-A: BWR Vessel and Internals Project, “Updated BWR Integrated Surveillance Program (ISP) Implementation Plan”, EPRI, Palo Alto, CA: 2002.

1003346.

[6.11] USNRC, Generic Letter 92-01 and RPV integrity Workshop Handouts, K. Wichman, M. Mitchell, and A. Hiser, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.

[6.12] Mehta, H.S. et. al., “10CFR50 Appendix G Equivalent Margin Analysis for Low Upper Shelf Energy in BWR/2 Through BWR/6 Vessels", GE Report NEDO-32205-A, Revision 1, February 1994.

[6.13] BWRVIP-102: “BWR Vessel and Internals Project, BWR Integrated Surveillance Program Implementation Guidelines," EPRI Technical Report 1003551, June 2002.

[6.14] ASTM Standard E 1921-02, “Test Method for the Determination of Reference Temperature, To, for Ferritic Steels in the Transition Range,” 2003 Annual Book of ASTM Standards, Vol. 03.01, American Society for Testing and Materials, West Conshohocken, PA.

[6.15] AMERICAN SOCIETY OF MECHANICAL ENGINEERS, ASME Code Case N-629, “Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials”, ASME, New York (2002).

[6.16] AMERICAN SOCIETY OF MECHANICAL ENGINEERS, ASME Code Case N-588, “Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels”, ASME, New York (2001).

[6.17] AMERICAN SOCIETY OF MECHANICAL ENGINEERS, ASME Code Case N-640, “Alternative Reference Fracture Toughness for Development of P-T Limit Curves”, ASME, New York (2002).

[6.18] Safety Standards of the Nuclear Safety Standards Commission (KTA); KTA 3201.2 Components of the Reactor Coolant Pressure Boundary of Light Water Reactors; Part 2: Design and Analysis, Edition 06/1996.

[6.19] Safety Standards of the Nuclear Safety Standards Commission (KTA); KTA 3203 Surveillance of the Irradiation Behaviour of Reactor Pressure Vessel Materials of LWR Facilities, Edition 06/2001.

[6.20] Japanese Industrial Technical Standards: The Reactor Vessel Material Surveillance Test Methods, JEAC 4201-2000, Japan Electric Association, 2000.

Same as Reference [5.38]

[6.21] ASTM SIP 513-Extensive Study of Low Fatigue Crack Growth Rates in A533 and A508 Steels, “P.C. Paris, et al., American Society of Testing and Materials, 1916 Race Street, Philadelphia PA 19103, 1972.

[6.22] Notification of Establishing Technical Standards on Structures etc., of Nuclear Power Generation Facilities, Ministry of Economy and Industry Notification No. 501, Japan, October 1980.

Same as reference [3.13]

[6.23] Japanese Society of Mechanical Engineers, JSME Codes for Nuclear Power Generation Facilities, JSME S NC1-2001, Rules on Design and Construction for Nuclear Power Plant, JSME Tokyo, August 2001.

Same as reference [3.14]

[6.24] Thermal and Nuclear Power Engineering Society, Guidelines on Environmental Fatigue Evaluation for LWR Component, June 2002.

[6.25] BWRVIP-14: “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals,” EPRI Report TR-105873, March 1996.

[6.26] BWRVIP-59: “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Nickel Base Austenitic Alloys in RPV Internals,” EPRI Report TR-108710, December 1998.

[6.27] BWRVIP-60-A: “BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment,” EPRI Technical Report 1008871, June 2003.

[6.28] BWRVIP-99: BWR Vessel and Internals Project, Crack Growth Rates in Irradiated Stainless Steels in BWR Internal Components,” EPRI Technical Report 1003018, December 2001.