XI.M19 (NUREG-1801 R0)
XI.M19 STEAM GENERATOR TUBE INTEGRITY
Program Description
Steam generator (SG) tubes have experienced tube degradation related to corrosion phenomena, such as primary water stress corrosion cracking (PWSCC), outside diameter stress corrosion cracking (ODSCC), intergranular attack (IGA), pitting, and wastage, along with other mechanically induced phenomena, such as denting, wear, impingement damage, and fatigue. Nondestructive examination (NDE) techniques are used to identify tubes that are defective and need to be removed from service or repaired in accordance with the guidelines of the plant technical specifications. In addition, operational leakage limits are included to ensure that, should substantial tube leakage develop, prompt action is taken to avoid rupture of the leaking tubes. These limits are included in plant technical specifications, such as standard technical specifications of NUREG-1430, Rev. 1, for Babcock & Wilcox pressurized water reactors (PWRs); NUREG-1431, Rev. 1, for Westinghouse PWRs; and NUREG-1432, Rev. 1, for Combustion Engineering PWRs.
The technical specifications specify SG inspection scope and frequency, and acceptance criteria for the plugging and repair of flawed tubes. The U.S. Nuclear Regulatory Commission (US NRC) Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded Steam Generator Tubes," provides guidelines for determining the tube repair criteria and operational leakage limits. Acceptance criteria for the plugging and repair of flawed tubes are incorporated in the plant technical specifications.
However, plants may apply for changes in their technical specifications to provide an alternate regulatory basis for SG degradation management. The US NRC has approved changes in the technical specification tube repair criteria at certain plants. Examples include the alternate voltage-based repair criteria of US NRC Generic Letter (GL) 95-05 and certain sleeving process. In addition, all PWR licensees have committed voluntarily to a SG degradation management program described in the Nuclear Energy Institute (NEI 97-06, "Steam Generator Program Guidelines." This program references a number of industry guidelines and incorporates a balance of prevention, inspection, evaluation, repair, and leakage monitoring measures. These guidelines are currently under NRC review. The NEI 97-06 document (a) includes performance criteria that are intended to provide assurance that tube integrity is being maintained consistent with the plant's licensing basis, and (b) provides guidance for monitoring and maintaining the tubes to provide assurance that the performance criteria are met at all times between scheduled inspections of the tubes. The NEI 97-06 program includes an assessment of degradation mechanisms that considers operating experience from similar SGs to identify degradation mechanisms and, for each mechanism, defines the inspection techniques, measurement uncertainty, as well as the sampling strategy. The industry guidelines provide criteria for the qualification of personnel, specific techniques, and the associated acquisition and analysis of data, including procedures, probe selection, analysis protocols, and reporting criteria. The performance criteria pertain to structural integrity, accident-induced leakage, and operational leakage. The SG monitoring program includes guidance on assessment of degradation mechanisms, inspection, tube integrity assessment, maintenance, plugging, repair, and leakage monitoring, as well as procedures for monitoring and controlling secondary-side and primary-side water chemistry. The water chemistry program for PWRs relies on monitoring and control of reactor water chemistry and secondary water chemistry.
As evaluated below, the plant technical specifications, incorporating NEI 97-06 as approved by the staff and any other alternate regulatory bases for SG degradation management that have been previously approved by the staff for that plant, are adequate to manage the effects of aging on the SG tubes. However, because NEI 97-06 is still under staff review, until the staff has approved NEI 97-06, the applicant's program should be reviewed on a plant-specific basis.
Evaluation and Technical Basis
- 1. Scope of Program: The scope of the program is specific to SG tubes. The program includes preventive measures to mitigate degradation related to corrosion phenomena; assessment of degradation mechanisms; inservice inspection (ISI) of steam generator tubes to detect degradation; evaluation and plugging or repair, as needed; and leakage monitoring to maintain the structural and leakage integrity of the pressure boundary. Tube inspection scope and frequency, plugging or repair, and leakage monitoring are in accordance with the plant technical specifications.
- 2. Preventive Actions: The program includes preventive measures to mitigate degradation related to corrosion phenomena. The guidelines in NEI 97-06 include foreign material exclusion as a means to inhibit fretting and wear degradation. The water chemistry program for PWRs relies on monitoring and control of reactor water chemistry based on the EPRI guidelines in TR-105714(Archived) for primary water chemistry and TR-102134(Archived) for secondary water chemistry. The program description and the evaluation and technical basis of monitoring and maintaining reactor water chemistry are presented in Chapter XI.M2, 'Water Chemistry," of this report.
- 3. Parameters Monitored/Inspected: The inspection activities in the program detect flaws in tubing or degradation of secondary side internals needed to maintain tubing integrity. Flaws are removed based on technical specification repair criteria. Degradation of steam generator internals is evaluated for corrective actions.
- 4. Detection of Aging Effects: The inspection requirements in the technical specifications are intended to detect tube degradation (i.e., aging effects), if it should occur. The NEI 97-06 document, which is currently under US NRC staff review, provides additional guidance on inspection programs to detect degradation. The intent of the inspection and repair criteria is to provide assurance of continued tube integrity between inspections.
- 5. Monitoring and Trending: Condition monitoring assessments are performed to determine whether structural and accident leakage criteria have been satisfied. Operational assessments are performed after inspections to verify that structural and leakage integrity are maintained during the operating interval until the next required inspection, which is selected in accordance with the technical specifications and staff approved NEI 97-06 guidelines. Comparison of the results of the condition monitoring assessment with the predictions of the previous operational assessment provides feedback for evaluation of the adequacy of the operational assessment and additional insights that can be incorporated into the next operational assessment.
- 6. Acceptance Criteria: Assessment of tube integrity and plugging or repair criteria of flawed tubes is in accordance with the plant technical specifications. The criteria for plugging or repairing SG tubes are based on US NRC RG 1.121 or other criteria previously reviewed and approved by the staff and incorporated into the plant technical specifications. Some examples that are applicable under certain circumstances include P*, F*, L*, or US NRC GL 95-05.
- For general and pitting corrosion, the acceptance criteria are in accordance with staff approved NEI 97-06 guidelines. Also, loose parts or foreign objects that are found are removed from the SGs unless it can be shown by evaluation that these objects do not cause unacceptable tube damage. The evaluation is to define an acceptable operating interval.
- For Westinghouse steam generator tube plugs, limits for the life of the plug and correlations for estimating their life are contained in WCAP-1 2244 and WCAP-12245.
- 7. Corrective Actions: Tubes containing flaws that do not meet the acceptance criteria are plugged or repaired. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable in addressing corrective actions.
- 8. Confirmation Process: Site quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B. As discussed in the appendix to this report, the staff finds the requirements of 10 CFR Part 50, Appendix B, acceptable in addressing the confirmation process and administrative controls.
- 9. Administrative Controls: See Item 8, above.
- 10. Operating Experience: Failures to detect some flaws, uncertainties in flaw sizing, inaccuracies in flaw locations, and the inability to detect some cracks at locations with dents have been reviewed in US NRC Information Notice (IN) 97-88. Recent experience indicates the importance of performing a complete inspection by using appropriate techniques and equipment for the reliable detection of tube degradation and to provide assurance that new forms of degradation are detected. Implementation of the program provides reasonable assurance that SG tube integrity is maintained consistent with the plant's licensing basis for the period of extended operation. Experience with the condition and operational assessments required for plants that have implemented the alternate repair criteria in US NRC GL 95-05 has shown that the predictions of the operational assessments have generally been consistent with the results of the subsequent condition monitoring assessments. In cases where discrepancies have been noted, adjustments have been made in the operational assessment models to improve agreement in subsequent assessments. In addition, NEI has prepared NEI 97-06 to incorporate lessons learned from plant operation experience and SG inspections and is under staff review.
References
EPRI TR-102134(Archived), PWR Secondary Water Chemistry Guidelines: Revision 3, Electric Power Research Institute, Palo Alto, CA, May 1993.
EPRI TR-105714(Archived), PWR Primary Water Chemistry Guidelines: Revision 3, Electric Power Research Institute, Palo Alto, CA, November 1995.
EPRI TR-107569(Archived), PWR Steam Generator Examination Guidelines: Revision 5, Electric Power Research Institute, Palo Alto, CA, September 1997.
NEI 97-06, Rev. 1, Steam Generator Program Guidelines, Nuclear Energy Institute, January 2000.
US NRC Generic Letter 95-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside-Diameter Stress-Corrosion Cracking, U.S. Nuclear Regulatory Commission, August 3, 1995.
US NRC Information Notice, 97-88, Experiences during Recent Steam Generator Inspections, U.S. Nuclear Regulatory Commission, December 12, 1997.
US NRC Regulatory Guide, 1.83, Rev. 1, Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes, U.S. Nuclear Regulatory Commission, July 1975.
US NRC Regulatory Guide, 1.121, Bases for Plugging Degraded PWR Steam Generator Tubes, U.S. Nuclear Regulatory Commission, August 1976.
NUREG-1430, Rev. 1, Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors, U.S. Nuclear Regulatory Commission, April 1995.
NUREG-1431, Rev. 1, Standard Technical Specifications for Westinghouse Pressurized Water Reactors, U.S. Nuclear Regulatory Commission, April 1995.
NUREG-1432, Rev. 1, Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors, U.S. Nuclear Regulatory Commission, April 1995.
WCAP-1 2244 and WCAP-1 2245, Steam Generator Tube Plug Integrity Summary Report, Addendum 2 to Revision 3, Westinghouse Electric Corporation, PA, May 1991.