XI.M19 (NUREG 2191, R0)

XI.M19 STEAM GENERATORS

Program Description

The Steam Generator program is applicable to managing the aging of steam generator tubes, plugs, sleeves, divider plate assemblies, tube-to-tubesheet welds, heads (interior surfaces of channel or lower/upper heads), tubesheet(s) (primary side), and secondary side components that are contained within the steam generator (i.e., secondary side internals). The aging of steam generator pressure vessel welds is managed by other programs such as Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report aging management program (AMP) XI.M1, “ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD,” and AMP XI.M2, “Water Chemistry.”

The establishment of a steam generator program for ensuring steam generator tube integrity is required by plant technical specifications (TS). The steam generator tube integrity portion of the TS at each pressurized water reactor (PWR) contains the same fundamental requirements as outlined in the standard TS of NUREG–1430, Volume 1, Revision 4, for Babcock & Wilcox (B&W) PWRs; NUREG–1431, Volume 1, Revision 4, for Westinghouse PWRs; and NUREG–1432, Volume 1, Revision 4, for Combustion Engineering PWRs. The requirements pertaining to steam generators in these three versions of the standard TS are essentially identical. The TS require tube integrity to be maintained and specify performance criteria, condition monitoring requirements, inspection scope and frequency, acceptance criteria for the plugging or repair of flawed tubes, acceptable tube repair methods, and leakage monitoring requirements.

The nondestructive examination techniques used to inspect steam generator components covered by this program are intended to identify components (e.g., tubes, plugs) with degradation that may need to be removed from service (e.g., tubes), repaired, or replaced, as appropriate.

The Steam Generator program at PWRs is modeled after Nuclear Energy Institute (NEI) 97-06, Revision 3, “Steam Generator Program Guidelines.” This program references a number of industry guidelines (e.g., the Electric Power Research Institute (EPRI) PWR Steam Generator Examination Guidelines, PWR Primary-to-Secondary Leak Guidelines(revision referenced is archived), PWR Primary Water Chemistry Guidelines(revision referenced is archived), PWR Secondary Water Chemistry Guidelines(revision referenced is archived), Steam Generator Integrity Assessment Guidelines(revision referenced is archived), Steam Generator In Situ Pressure Test Guidelines(revision referenced is archived)) and incorporates a balance of prevention, mitigation, inspection, evaluation, repair, and leakage monitoring measures. 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. Steam generator tube integrity can be affected by degradation of steam generator plugs, sleeves, and secondary side components. The NEI 97-06 program has been effective in managing the aging effects associated with steam generator tubes, plugs, sleeves, and secondary side components.

Degradation of divider plate assemblies, tube-to-tubesheet welds, heads (internal surfaces), or tubesheets (primary side) may have safety implications. Therefore, all of these components and the steam generator tubes, plugs, sleeves and secondary side components are addressed by this AMP.

Evaluation and Technical Basis

1. Scope of Program: This program addresses degradation associated with steam generator tubes, plugs, sleeves, divider plate assemblies, tube-to-tubesheet welds, heads (interior surfaces of channel or lower/upper heads), tubesheet(s) (primary side), and secondary side components that are contained within the steam generator (i.e., secondary side internals). The program does not cover the steam generator secondary side shell, any nozzles attached to the secondary side shell or steam generator head, or the welds associated with these components. In addition, the program does not cover steam generator head welds (other than general corrosion of these welds caused as a result of degradation (defects/flaws) in the primary side cladding).

2. Preventive Actions: This program includes preventive and mitigative actions for addressing degradation. Preventive and mitigative measures that are part of the Steam Generator program include foreign material exclusion programs, and other primary and secondary side maintenance activities. The program includes foreign material exclusion as a means to inhibit wear degradation and secondary side maintenance activities, such as sludge lancing, for removing deposits that may contribute to degradation. Guidance on foreign material exclusion is provided in NEI 97-06. Guidance on maintenance of secondary side integrity is provided in the EPRI Steam Generator Integrity Assessment Guidelines(revision referenced is archived). Primary side preventive maintenance activities include replacing plugs made with corrosion susceptible materials with more corrosion resistant materials and preventively plugging tubes susceptible to degradation.

Extensive deposit buildup in the steam generators could affect tube integrity. The EPRI Steam Generator Integrity Assessment Guidelines(revision referenced is archived), which are referenced in NEI 97-06, provide guidance on maintaining the secondary side of the steam generator, including secondary side cleaning. Secondary side water chemistry plays an important role in controlling the introduction of impurities into the steam generator and potentially limiting their deposition on the tubes. Maintaining high water purity reduces susceptibility to stress corrosion cracking (SCC) or intergranular stress corrosion cracking (IGSCC). Water chemistry is monitored and maintained in accordance with the Water Chemistry program. The program description and evaluation and technical basis of monitoring and maintaining water chemistry are addressed in the GALL-SLR Report AMP XI.M2, “Water Chemistry.”

3. Parameters Monitored or Inspected: There are currently three types of steam generator tubing used in the United States: (i) mill annealed Alloy 600, (ii) thermally treated Alloy 600, and (iii) thermally treated Alloy 690. Mill annealed Alloy 600 steam generator tubes have experienced degradation due to corrosion (e.g., primary water SCC, outside diameter SCC, intergranular attack, pitting, and wastage) and mechanically induced phenomena (e.g., denting, wear, impingement damage, and fatigue). Thermally treated Alloy 600 steam generator tubes have experienced degradation due to corrosion (primarily cracking) and mechanically induced phenomena (primarily wear). Thermally treated Alloy 690 tubes have only experienced tube degradation due to mechanically induced phenomena (primarily wear).

Degradation of tube plugs, sleeves, heads, tubesheet(s), and secondary side internals has also been observed, depending, in part, on the material of construction of the specific component. The potential for degradation exists for divider plate assemblies and tube-to-tubesheet welds, depending, in part, on the materials of construction for the components.

The program includes an assessment of the forms of degradation to which a component is susceptible and implementation of inspection techniques capable of detecting those forms of degradation. The parameter monitored is specific to the component and the acceptance criteria for the inspection. For example, the severity of tube degradation may be evaluated in terms of the depth of degradation or measured voltage, dependent on whether a depth-based or voltage-based tube repair criteria (acceptance criteria) is being implemented for that specific degradation mechanism. Other parameters monitored include signals of excessive deposit buildup (e.g., steam generator water level oscillations), which may result in fatigue failure of tubes or corrosion of the tubes; water chemistry parameters, which may indicate unacceptable levels of impurities; primary-to-secondary leakage, which may indicate excessive tube, plug, or sleeve degradation; and the presence of loose parts or foreign objects on the primary and secondary side of the steam generator, which may result in tube damage.

Water chemistry parameters are also monitored and controlled, as discussed in GALL-SLR Report AMP XI.M2. The EPRI PWR Steam Generator Primary-to-Secondary Leakage Guidelines (EPRI 1022832(Archived)) provides guidance on monitoring primary-to-secondary leakage. The EPRI Steam Generator Integrity Assessment Guidelines (EPRI 3002007571(Archived)) provide guidance on secondary side activities.

In summary, the NEI 97-06 program provides guidance on parameters to be monitored or inspected except for steam generator divider plate assemblies, tube-to-tubesheet welds, heads (channel or lower/upper heads), and tubesheets. For these latter components, visual inspections are performed at least every 72 effective full power months or every third refueling outage, whichever results in more frequent inspections. These inspections of the steam generator head interior surfaces including the divider plate are intended to identify signs that cracking or loss of material may be occurring (e.g., through identification of rust stains).

4. Detection of Aging Effects: The TS require that a Steam Generator program be established and implemented to maintain the integrity of the steam generator tubes. In accordance with this requirement, components that could compromise tube integrity are properly evaluated or monitored (e.g., degradation of a secondary side component that could result in a loss of tube integrity is managed by this program). The inspection requirements in the TS are intended to detect degradation (i.e., aging effects), if they should occur.

The TS are performance-based, and the actual scope of the inspection and the expansion of sample inspections are justified based on the results of the inspections. The goal is to perform inspections at a frequency sufficient to provide reasonable assurance of steam generator tube integrity for the period of time between inspections.

The general condition of some components (e.g., plugs, secondary side components, divider plates, and primary side cladding of channel heads and tubesheets) is monitored. It may be monitored visually, and, subsequently, more detailed inspections may be performed if degradation is detected.

NEI 97-06 provides additional guidance on inspection programs to detect degradation of tubes, sleeves, plugs, and secondary side internals. The frequencies of the inspections are based on technical assessments. Guidance on performing these technical assessments is contained in NEI 97-06 and the associated industry guidelines.

The inspections and monitoring are performed by qualified personnel using qualified techniques in accordance with approved licensee procedures. The EPRI PWR Steam Generator Examination Guidelines (EPRI 3002007572) contains guidance on the qualification of steam generator tube inspection techniques.

The primary-to-secondary leakage monitoring program provides a potential indicator of a loss of steam generator tube integrity. NEI 97-06 and the associated EPRI guidelines provide information pertaining to an effective leakage monitoring program.

5. Monitoring and Trending: Condition monitoring assessments are performed to determine whether the structural- and accident-induced leakage performance criteria were satisfied during the prior operating interval. Operational assessments are performed to verify that structural and leakage integrity will be maintained for the planned operating interval before the next inspection. If tube integrity cannot be maintained for the planned operating interval before the next inspection, corrective actions are taken in accordance with the plant’s corrective action program. Comparisons of the results of the condition monitoring assessment to the predictions of the previous operational assessment are performed to evaluate the adequacy of the previous operational assessment methodology. If the operational assessment was not conservative in terms of the number and/or severity of the condition, corrective actions are taken in accordance with the plant’s corrective action program.

The TS require condition monitoring and operational assessments to be performed (although the TS do not explicitly require operational assessments, these assessments are necessary to ensure that the tube integrity will be maintained until the next inspection). Condition monitoring and operational assessments are done in accordance with the TS requirements and guidance in NEI 97-06 and the EPRI Steam Generator Integrity Assessment Guidelines(revision referenced is archived).

The goal of the inspection program for all components covered by this AMP is to ensure that the components continue to function consistent with the design and licensing basis of the facility (including regulatory safety margins).

Assessments of the degradation that may occur in the components covered by this AMP, except for steam generator divider plate assemblies, tube-to-tubesheet welds, heads, and tubesheets as noted above, are performed in accordance with the guidance in the EPRI Steam Generator Integrity Assessment Guidelines(revision referenced is archived). All assessments of component degradation are performed to confirm that the components continue to function consistent with the design and licensing basis and to confirm that TS requirements are satisfied.

6. Acceptance Criteria: Assessment of tube and sleeve integrity and plugging or repair criteria of flawed and sleeved tubes is in accordance with plant TS. The criteria for plugging or repairing steam generator tubes and sleeves are based on the U.S. Nuclear Regulatory Commission (US NRC) [Regulatory Guide (RG) 1.121 and are incorporated into plant TS. Guidance on assessing the acceptability of flaws is also provided in NEI 97-06 and the associated EPRI guidelines, including the EPRI PWR Steam Generator Examination Guidelines (EPRI 3002007572), EPRI Steam Generator In-Situ Pressure Test Guidelines (EPRI 1025132(Archived)) and EPRI Steam Generator Integrity Assessment Guidelines (EPRI 3002007571(Archived)).

Degraded plugs, divider plate assemblies, tube-to-tubesheet welds, heads (interior surfaces), tubesheets (primary side), and secondary side internals are evaluated for continued acceptability on a case-by-case basis, as is done for leaving a loose part or a foreign object in a steam generator. NEI 97-06 and the associated EPRI guidelines provide guidance on the performance of some of these evaluations. The intent of all evaluations is to ensure that the components will continue to perform their functions consistent with the design and licensing basis of the facility, and will not affect the integrity of other components (e.g., by generating loose parts).

Guidance on the acceptability of primary-to-secondary leakage and water chemistry parameters also are discussed in NEI 97-06 and the associated EPRI guidelines.

7. Corrective Actions: Results that do not meet the acceptance criteria are addressed in the applicant’s corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, “Corrective Action,” of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the corrective actions element of this AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

For degradation of steam generator tubes and sleeves (if applicable), the TS provide requirements on the actions to be taken when the acceptance criteria are not met. For degradation of other components, the appropriate corrective action is evaluated per NEI 97-06 and the associated EPRI guidelines, the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section XI, 10 CFR 50.65, and 10 CFR Part 50, Appendix B, as appropriate.

8. Confirmation Process: The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

The adequacy of the preventive measures in the Steam Generator program is confirmed through periodic inspections.

9. Administrative Controls: Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR Part 50, Appendix B, associated with managing the effects of aging. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

10. Operating Experience: Several generic communications have been issued by the US NRC related to the steam generator programs implemented at plants. The reference section lists many of these generic communications. In addition, NEI 97-06 provides guidance to the industry for routinely sharing pertinent steam generator operating experience (OE) and for incorporating lessons learned from plant operation into guidelines referenced in NEI 97-06. The latter includes providing interim guidance to the industry, when needed.

The NEI 97-06 program has been effective at managing the aging effects associated with steam generator tubes, plugs, sleeves, and secondary side components that are contained within the steam generator (i.e., secondary side internals), such that the steam generators can perform their intended safety function.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry OE including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

References

10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.55a, “Codes and Standards.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.65, “Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” New York, New York: The American Society of Mechanical Engineers. 2008.

EPRI. EPRI 1014986(Archived), “PWR Primary Water Chemistry Guidelines.” Revision 7, Volumes 1 and 2. Palo Alto, California: Electric Power Research Institute. April 2014.

_____. EPRI 1016555(Archived), “PWR Secondary Water Chemistry Guidelines.” Revision 7. Palo Alto, California: Electric Power Research Institute. February 2009.

_____. EPRI 1022832(Archived), “PWR Primary-to-Secondary Leak Guidelines.” Revision 4. Palo Alto, California: Electric Power Research Institute. November 2011.

_____. EPRI 1025132(Archived), “Steam Generator In-Situ Pressure Test Guidelines.” Revision 4. Palo Alto, California: Electric Power Research Institute. October 2012.

_____. EPRI 3002007571(Archived), “Steam Generator Integrity Assessment Guidelines.” Revision 4. Palo Alto, California: Electric Power Research Institute. June 2016.

_____. EPRI 3002007572, “PWR Steam Generator Examination Guidelines.” Revision 8. Palo Alto, California: Electric Power Research Institute. June 2016.

NEI. NEI 97-06, “Steam Generator Program Guidelines.” Revision 3. Washington, DC: Nuclear Energy Institute. January 2011.

US NRC. Bulletin 88-02, “Rapidly Propagating Cracks in Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1988.

_____. Bulletin 89-01, “Failure of Westinghouse Steam Generator Tube Mechanical Plugs.” Washington, DC: U.S. Nuclear Regulatory Commission. May 1989.

_____. Bulletin 89-01, “Failure of Westinghouse Steam Generator Tube Mechanical Plugs.” Supplement 1. Washington, DC: U.S. Nuclear Regulatory Commission. November 1990.

_____. Bulletin 89-01, “Failure of Westinghouse Steam Generator Tube Mechanical Plugs.” Supplement 2. Washington, DC: U.S. Nuclear Regulatory Commission. June 1991.

_____. Draft Regulatory Guide DG-1074, “Steam Generator Tube Integrity.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1998.

_____. Generic Letter 95-03, “Circumferential Cracking of Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. April 1995.

_____. Generic Letter 95-05, “Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking.” Washington, DC: U.S. Nuclear Regulatory Commission. August 1995.

_____. Generic Letter 97-06, “Degradation of Steam Generator Internals.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1997.

_____. Generic Letter 2004-01, “Requirements for Steam Generator Tube Inspections.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2004.

_____. Generic Letter 2006-01, “Steam Generator Tube Integrity and Associated Technical Specifications.” Washington, DC: U.S. Nuclear Regulatory Commission. January 2006.

_____. Information Notice 85-37, “Chemical Cleaning of Steam Generators at Millstone 2.” Washington, DC: U.S. Nuclear Regulatory Commission. May 1985.

_____. Information Notice 88-06, “Foreign Objects in Steam Generators.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1988.

_____. Information Notice 88-99, “Detection and Monitoring of Sudden and/or Rapidly Increasing Primary-to-Secondary Leakage.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1988.

_____. Information Notice 89-65, “Potential for Stress Corrosion Cracking in Steam Generator Tube Plugs Supplied by Babcock and Wilcox.” Washington, DC: U.S. Nuclear Regulatory Commission. September 1989.

_____. Information Notice 90-49, “Stress Corrosion Cracking in PWR Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. August 1990.

_____. Information Notice 91-19, “Steam Generator Feedwater Distribution Piping Damage.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1991.

_____. Information Notice 91-43, “Recent Incidents Involving Rapid Increases in Primary-to-Secondary Leak Rate.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1991.

_____. Information Notice 91-67, “Problems with the Reliable Detection of Intergranular Attack (IGA) of Steam Generator Tubing.” Washington, DC: U.S. Nuclear Regulatory Commission. October 1991.

_____. Information Notice 92-80, “Operation with Steam Generator Tubes Seriously Degraded.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1992.

_____. Information Notice 93-52, Draft NUREG–1477, “Voltage-Based Interim Plugging Criteria for Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1993.

_____. Information Notice 93-56, “Weaknesses in Emergency Operating Procedures Found as a Result of Steam Generator Tube Rupture.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1993.

_____. Information Notice 94-05, “Potential Failure of Steam Generator Tubes Sleeved With Kinetically Welded Sleeves.” Washington, DC: U.S. Nuclear Regulatory Commission. January 1994.

_____. Information Notice 94-43, “Determination of Primary-to-Secondary Steam Generator Leak Rate.” Washington, DC: U.S. Nuclear Regulatory Commission. June 1994.

_____. Information Notice 94-62, “Operational Experience on Steam Generator Tube Leaks and Tube Ruptures.” Washington, DC: U.S. Nuclear Regulatory Commission. August 1994.

_____. Information Notice 94-87, “Unanticipated Crack in a Particular Heat of Alloy 600 Used for Westinghouse Mechanical Plugs for Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1994.

_____. Information Notice 94-88, “Inservice Inspection Deficiencies Result in Severely Degraded Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1994.

_____. Information Notice 95-40, “Supplemental Information to Generic Letter 95-03, Circumferential Cracking of Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. September 1995.

_____. Information Notice 96-09, “Damage in Foreign Steam Generator Internals.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1996.

_____. Information Notice 96-09, “Damage in Foreign Steam Generator Internals.” Supplement 1. Washington, DC: U.S. Nuclear Regulatory Commission. July 1996.

_____. Information Notice 96-38, “Results of Steam Generator Tube Examinations.” Washington, DC: U.S. Nuclear Regulatory Commission. June 1996.

_____. Information Notice 97-26, “Degradation in Small-Radius U-Bend Regions of Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. May 1997.

_____. Information Notice 97-49, “B&W Once-Through Steam Generator Tube Inspection Findings.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1997.

_____. Information Notice 97-79, “Potential Inconsistency in the Assessment of the Radiological Consequences of a Main Steam Line Break Associated with the Implementation of Steam Generator Tube Voltage-Based Repair Criteria.” Washington, DC: U.S. Nuclear Regulatory Commission. November 1997.

_____. Information Notice 97-88, “Experiences During Recent Steam Generator Inspections.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1997.

_____. Information Notice 98-27, “Steam Generator Tube End Cracking.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1998.

_____. Information Notice 2000-09, “Steam Generator Tube Failure at Indian Point Unit 2.” Washington, DC: U.S. Nuclear Regulatory Commission. June 2000.

_____. Information Notice 2001-16, “Recent Foreign and Domestic Experience with Degradation of Steam Generator Tubes and Internals.” Washington, DC: U.S. Nuclear Regulatory Commission. October 2001.

_____. Information Notice 2002-02, “Recent Experience with Plugged Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. January 2002.

_____. Information Notice 2002-02, “Recent Experience with Plugged Steam Generator Tubes.” Supplement 1. Washington, DC: U.S. Nuclear Regulatory Commission. July 2002.

_____. Information Notice 2002-21, “Axial Outside- Diameter Cracking Affecting Thermally Treated Alloy 600 Steam Generator Tubing.” Washington, DC: U.S. Nuclear Regulatory Commission. June 2002.

_____. Information Notice 2002-21, “Axial Outside-Diameter Cracking Affecting Thermally Treated Alloy 600 Steam Generator Tubing.” Supplement 1. Washington, DC: U.S. Nuclear Regulatory Commission. April 2003.

_____. Information Notice 2003-05, “Failure to Detect Freespan Cracks in PWR Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. June 2003.

_____. Information Notice 2003-13, “Steam Generator Tube Degradation at Diablo Canyon.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2003.

_____. Information Notice 2004-10, “Loose Parts in Steam Generators.” Washington, DC: U.S. Nuclear Regulatory Commission. May 2004.

_____. Information Notice 2004-16, “Tube Leakage Due to a Fabrication Flaw in a Replacement Steam Generator.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2004.

_____. Information Notice 2004-17, “Loose Part Detection and Computerized Eddy Current Data Analysis in Steam Generators.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2004.

_____. Information Notice 2005-09, “Indications in Thermally Treated Alloy 600 Steam Generator Tubes and Tube-to-Tubesheet Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. April 2005.

_____. Information Notice 2005-29, “Steam Generator Tube and Support Configuration.” Washington, DC: U.S. Nuclear Regulatory Commission. October 2005.

_____. Information Notice 2007-37, “Buildup of Deposits in Steam Generators.” Washington, DC: U.S. Nuclear Regulatory Commission. November 2007.

_____. Information Notice 2008-07, “Cracking Indications in Thermally Treated Alloy 600 Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. April 2008.

_____. Information Notice 2010-05, “Management of Steam Generator Loose Parts and Automated Eddy Current Data Analysis.” Washington, DC: U.S. Nuclear Regulatory Commission. February 2010.

_____. Information Notice 2010-21, “Crack-Like Indication in the U-Bend Region of a Thermally Treated Alloy 600 Steam Generator Tube.” Washington, DC: U.S. Nuclear Regulatory Commission. October 2010.

_____. Information Notice 2012-07, “Tube-To-Tube Contact Resulting in Wear in Once-Through Steam Generators.” Washington, DC: U.S. Nuclear Regulatory Commission. July 2012.

_____. Information Notice 2013-11, “Crack-Like Indications at Dents/Dings and in the Freespan Region of Thermally Treated Alloy 600 Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. July 2013.

_____. Information Notice 2013-20, “Steam Generator Channel Head and Tubesheet Degradation.” Washington, DC: U.S. Nuclear Regulatory Commission. July 2013.

_____. NUREG–1430, “Standard Technical Specifications for Babcock and Wilcox Pressurized Water Reactors.” Volume 1, Revision 4. Washington DC: U.S. Nuclear Regulatory Commission. April 2012.

_____. NUREG–1431, “Standard Technical Specifications for Westinghouse Pressurized Water Reactors.” Volume 1, Revision 4. Washington DC: U.S. Nuclear Regulatory Commission. April 2012.

_____. NUREG–1432, “Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors.” Volume 1, Revision 4. Washington DC: U.S. Nuclear Regulatory Commission. April 2012.

_____. Regulatory Guide 1.121, “Bases for Plugging Degraded PWR Steam Generator Tubes.” Washington, DC: U.S. Nuclear Regulatory Commission. August 1976.

_____. Regulatory Issue Summary 2000-22, “Issues Stemming from NRC Staff Review of Recent Difficulties Experienced in Maintaining Steam Generator Tube Integrity.” Washington DC: U.S. Nuclear Regulatory Commission. November 2000.

_____. Regulatory Issue Summary 2007-20, “Implementation of Primary-to-Secondary Leakage Performance Criteria.” Washington DC: U.S. Nuclear Regulatory Commission. August 2007.

_____. Regulatory Issue Summary 2009-04, “Steam Generator Tube Inspection Requirements.” Washington DC: U.S. Nuclear Regulatory Commission. April 2009.

XI.M19 (NUREG 2191, R0)

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