2191 R0 XI.M11B

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XI.M11B CRACKING OF NICKEL-ALLOY COMPONENTS AND LOSS OF MATERIAL DUE TO BORIC ACID-INDUCED CORROSION IN REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS (PWRs ONLY)

Program Description

This program addresses operating experience (OE) of degradation due to primary water stress corrosion cracking (PWSCC) of components or welds constructed from certain nickel alloys (e.g., Alloy 600/82/182) and exposed to pressurized water reactor (PWR) primary coolant at elevated temperature. The initiation and growth of PWSCC cracks have been shown to be a function of several variables, including but not limited to: (i) temperature, (ii) stress, (iii) microstructure, (iv) time, and (v) water chemistry. As a result, this program is informed by Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report aging management program (AMP) XI.M2, “Water Chemistry.”

In addition to inspections designed to identify cracking of nickel alloy components, this program also contains inspections designed to potentially identify the presence of boric acid residues, which has been demonstrated by OE to lead to loss of material in susceptible carbon and low alloy steel components. Thus, this program is used in conjunction with GALL-SLR Report AMP XI.M10, “Boric Acid Corrosion.” Except as required in Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, it is not the general intent of this program to manage the aging of components and welds constructed from PWSCC-resistant nickel alloys (e.g., Alloy 690/52/152).

Plants have implemented and maintained existing programs to manage cracking due to PWSCC for nickel alloy components and welds, consistent with Electric Power Research Institute (EPRI) Materials Reliability Program (MRP)-126. The scope of subsequent license renewal may identify additional nickel alloy components or welds to be included in the applicant’s aging management program.

Evaluation and Technical Basis

1. Scope of Program: The scope of this program includes three basic groups of components and materials: (i) all nickel alloy components and welds which are identified at the plant in accordance with the guidelines of EPRI MRP-126; (ii) nickel alloy components and welds identified in American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Cases N-770, N-729 and N-722, as incorporated by reference in 10 CFR 50.55a; and (iii) components that are susceptible to corrosion by boric acid and may be impacted by leakage of boric acid from nearby or adjacent nickel alloy components previously described. This program manages cracking due to PWSCC and loss of material due to boric acid corrosion.

2. Preventive Actions: This program is primarily a condition monitoring program. Since the cracking of nickel alloys is affected by water quality this program is used in conjunction with GALL-SLR Report AMP XI.M2, “Water Chemistry.” Additionally, in accordance with 10 CFR 50.55a, an applicant may choose to mitigate component degradation in lieu of performing required inspections.

3. Parameters Monitored or Inspected: Components and welds within the scope of this program are inspected for evidence of PWSCC by volumetric, surface, or visual testing. In the event boric acid residues or corrosion products are discovered during these inspections, the potential for, or extent of, loss of material is evaluated by visual and quantitative methods.

4. Detection of Aging Effects: For nickel alloy components and welds addressed by regulatory requirements contained in 10 CFR 50.55a, inspections are conducted in accordance with 10 CFR 50.55a. Other nickel alloy components and welds within the scope of this program are inspected in accordance with the guidance in the EPRI MRP-126 report.

The program also performs a baseline volumetric or inner-diameter surface inspection of all susceptible nickel alloy branch line connections and associated welds as identified in Table 4-1 of EPRI MRP-126 if such components or welds are of a sufficient size to create a loss of coolant accident through a complete failure (guillotine break) or ejection of the component and the normal operating temperature of the components is 274 °C (Celsius) (525 °F (Fahrenheit)) or greater. The baseline inspection is performed prior to the subsequent period of extended operation using a qualified method in accordance with Appendix IV or VIII of ASME Code Section XI as incorporated by reference in 10 CFR 50.55a, or equivalent. Existing periodic inspections using volumetric or surface examination methods may be credited for the baseline inspection. If the baseline inspection indicates the occurrence of PWSCC, periodic volumetric or inner-diameter surface inspections are performed with adequate periodicity.

5. Monitoring and Trending: Reactor coolant leakage is calculated and trended on a routine basis in accordance with technical specifications to detect changes in the leakage rates (Regulatory Guide (RG) 1.45). Flaw evaluation through 10 CFR 50.55a is a means to monitor cracking. Detected flaws are monitored and trended by performing periodic and successive inspections in accordance with ASME Code Cases N-770, N-729 and N-722, as incorporated by reference in 10 CFR 50.55a, and the guidelines in EPRI MRP-126.

6. Acceptance Criteria: Acceptance criteria are in accordance with applicable sections of Section XI of the ASME Code, as incorporated by reference in 10 CFR 50.55a. If any boric acid residue or corrosion product is detected, additional actions are performed to determine the source of leakage and the impact of boric acid corrosion on adjacent components.

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 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.

Components with relevant unacceptable flaw indications are corrected for further services through an implementation of appropriate repair or replacement as dictated by 10 CFR 50.55a and industry guidelines (e.g., EPRI MRP-126). In addition, detection of leakage or evidence of cracking in susceptible components within the scope of this program require a scope expansion of current inspection and increased inspection frequencies for some components, as required by 10 CFR 50.55a and industry guidelines (e.g., EPRI MRP-126).

Repair and replacement procedures and activities must either comply with ASME Code Section XI, as incorporated in 10 CFR 50.55a or conform to applicable ASME Code Cases that have been endorsed in 10 CFR 50.55a by referencing the latest version of RG 1.147.

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.

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: This program addresses review of related OE, including plant-specific information, generic industry findings, and international data. Within the current regulatory requirements, as necessary, the applicant maintains a record of OE through the required update of the facility’s inservice inspection program in accordance with 10 CFR 50.55a. Additionally, the applicant follows mandated industry guidelines developed to address OE in accordance with Nuclear Energy Institute (NEI)-03-08, “Guideline for the Management of Materials Issues.”

PWSCC of Alloy 600 components has been observed in domestic and foreign PWRs (US NRC Information Notice (IN) 90-10). The ingress of demineralizer resins also has occurred in operating plants (US NRC IN 96-11). The Water Chemistry program, GALL-SLR Report AMP XI.M2, manages the effects of such excursions through monitoring and control of primary water chemistry. US NRC Generic Letter 97-01 is effective in managing the effect of PWSCC. PWSCC has occurred in the vessel head penetration nozzles of U.S. PWRs as described in US NRC Bulletins 2001-01, 2002-01 and 2002-02. In addition, PWSCC was observed in reactor vessel bottom mounted instrument nozzles (US NRC IN 2003-11, Supplement 1, and Licensee Event Report 530/2013-001-00).

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.

ASME. ASME Code Case N-722-1, “Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials. New York, New York: The American Society of Mechanical Engineers. January 2009.

_____. ASME Code Case N-729-1, “Alternative Examination Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds.” New York, New York: The American Society of Mechanical Engineers. March 2006.

_____. ASME Code Case N-770, “Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities.” New York, New York: The American Society of Mechanical Engineers. January 2009.

EPRI. MRP-126, “Generic Guidance for Alloy 600 Management.” Palo Alto, California: Electric Power Research Institute. November 2004.

Licensee Event Report 530/2013-001-00, “Leakage on Reactor Vessel Bottom-Mounted Instrumentation Nozzle 3.” https://lersearch.inl.gov/LERSearchCriteria.aspx. December 2013.

NEI. NEI 03-08, “Guideline for the Management of Materials Issues.” Revision 2. Washington, DC: Nuclear Energy Institute. January 2010.

US NRC. Bulletin 2001-01, “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2001.

_____. Bulletin 2002-01, “Reactor Pressure Vessel Head Degradation and Reactor Coolant Pressure Boundary Integrity.” Agencywide Documents Access and Management System (ADAMS) Accession No. ML020770497. Washington, DC: U.S. Nuclear Regulatory Commission. March 2002.

_____. Bulletin 2002-02, “Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection Programs.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2002.

_____. Generic Letter 97-01, “Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations.” Washington, DC: U.S. Nuclear Regulatory Commission. April 1997.

_____. Information Notice 90-10, “Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1990.

_____. Information Notice 96-11, “Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1996.

_____. Information Notice 2003-11, “Leakage Found on Bottom-Mounted Instrumentation Nozzles.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2003.

_____. Information Notice 2003-11, “Leakage Found on Bottom-Mounted Instrumentation Nozzles.” Supplement 1. Washington, DC: U.S. Nuclear Regulatory Commission. January 2004.

_____. Inspection Manual, Inspection Procedure 71111.08, “Inservice Inspection Activities.” Washington, DC: U.S. Nuclear Regulatory Commission. January 2015.

_____. NUREG–1823, “U.S. Plant Experience with Alloy 600 Cracking and Boric Acid Corrosion of Light-Water Reactor Pressure Vessel Materials.” Washington, DC: U.S. Nuclear Regulatory Commission. April 2005.

_____. Regulatory Guide 1.45, Revision 1, “Guidance on Monitoring and Responding to Reactor Coolant System Leakage.” Washington, DC: U.S. Nuclear Regulatory Commission. May 2008.

_____. Regulatory Guide 1.147, Revision 17, “Inservice Inspection Code Case Acceptability.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2014.

_____. Regulatory Information Summary 2008-25, “Regulatory Approach for Primary Water Stress Corrosion Cracking of Dissimilar Metal Butt Welds in Pressurized Water Reactor Primary Coolant System Piping.” Washington, DC: U.S. Nuclear Regulatory Commission. October 2008.

Operating Experience

Primary Water Stress Corrosion Cracking

Pressurized Water Reactor

Electric Power Research Institute

American Society of Mechanical Engineers

Quality Assurance

Aging Management Program

Nuclear Energy Institute

United States of America

Nuclear Regulatory Commission - The US federal government agency that regulates nuclear power plants

Agency wide Documents Access and Management System - The official recordkeeping system, through which the U.S. Nuclear Regulatory Commission (US NRC) provides access to the "libraries" or collections of publicly available documents