2191 R0 XI.S4

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XI.S4 10 CFR PART 50, APPENDIX J

Program Description

A typical primary reactor containment system consists of a containment structure (containment), and a number of electrical, mechanical, equipment hatch, and personnel air lock penetrations. As described in Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” (Appendix J) periodic containment leak rate tests are required to assure that (a) leakage through these containments or systems and components penetrating these containments does not exceed allowable leakage rates specified in the Technical Specification (TS) and (b) integrity of the containment structure is maintained during its service life.

This aging management program (AMP) credits the existing program required by 10 CFR Part 50 Appendix J, and augments it to ensure that all containment pressure-retaining components are managed for age-related degradation.

Appendix J provides two options, Option A and Option B, to meet the requirements of a containment leak rate test (LRT) program. Option A is prescriptive with all testing performed on specified periodic intervals. Option B is a performance-based approach. The U.S. Nuclear Regulatory Commission (US NRC) Regulatory Guide 1.163, “Performance-Based Containment Leak-Test Program” and Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option for 10 CFR Part 50, Appendix J, as approved by the US NRC final safety evaluation for NEI 94-01, Revision 2-A and Revision 3-A, provide additional information regarding Option B. Three types of tests are performed under either Option A or Option B, or a mix as adopted by licensees on a voluntary basis.

Type A integrated leak rate tests determine the overall containment integrated leakage rate, at the calculated peak containment internal pressure related to the design basis loss of coolant accident. Type B (containment penetration leak rate) tests detect local leaks and measure leakage across each pressure-containing or leakage-limiting boundary of containment penetrations. Type C (containment isolation valve leak rate) tests detect local leaks and measure leakage across containment isolation valves installed in containment penetrations or lines penetrating the containment.

Appendix J requires a general visual inspection of the accessible interior and exterior surfaces of the containment structures and components (SCs) to be performed prior to any Type A test and at periodic intervals between tests based on the performance of the containment system. The visual inspections required by American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Section XI, Subsections IWE and IWL are acceptable substitutes for the general visual inspection. The purpose of the Appendix J general visual inspection is to uncover any evidence of structural deterioration that may affect the containment structure leakage integrity or the performance of the Type A test.

Evaluation and Technical Basis

1. Scope of Program: The scope of the containment LRT program includes the containment system and related systems and components penetrating the containment pressure-retaining or leakage-limiting boundary. The aging effects associated with containment pressure-retaining boundary components within the scope of subsequent license renewal and excluded from Type B or C Appendix J testing must still be managed.

Other programs may be credited for managing the aging effects associated with these components; however, the component and the proposed AMP should be clearly identified.

2. Preventive Action: The containment LRT program is a performance monitoring program with no specific preventive actions.

3. Parameters Monitored or Inspected: The monitored parameters are leakage rates through the containment shell, containment liner, penetrations, associated welds, access openings, and associated pressure boundary components.

4. Detection of Aging Effects: A containment LRT program is effective in detecting leakage rates of the containment pressure boundary components, including seals and gaskets, and in identifying and correcting sources of leakage. While the calculation of leakage rates and satisfactory performance of containment leak rate testing demonstrates the leakage integrity of the containment, it does not by itself provide information that would indicate that age-related degradation has initiated or that the capacity of the containment may have been reduced for other types of loading conditions. This would be achieved with the implementation of acceptable containment inservice inspection (ISI) programs such as ASME Code Section XI, Subsection IWE ( Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report AMP XI.S1), and ASME Code Section XI, Subsection IWL ( GALL-SLR Report AMP XI.S2).

5. Monitoring and Trending: Because the containment LRT program is repeated periodically throughout the operating license period, the entire containment pressure boundary is monitored over time. The frequency of these tests depends on which option (A or B) is selected. With Option A, testing is performed on a regular fixed time interval as defined in Appendix J. In the case of Option B, acceptable performance in prior tests meeting leakage rate limits serves as a basis to adjust the testing interval. For valves and penetrations administrative leakage rate limits may be set lower than the regulatory acceptance criteria for early detection of age-related degradation.

6. Acceptance Criteria: Plant TS define the regulatory acceptance criteria for leakage rate limits. The regulatory acceptance criteria meet the requirements as set forth in Appendix J, and are part of each plant’s licensing basis.

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 SCs within the scope of this program.

Corrective actions are taken in accordance with 10 CFR Part 50, Appendix J and NEI 94-01. When leakage rates do not meet the acceptance criteria, an evaluation is performed to identify the cause of the unacceptable performance and appropriate corrective actions are taken.

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: To date, Appendix J, containment LRT program, in conjunction with the containment ISI program, have been effective in preventing unacceptable leakage through the containment pressure boundary. Implementation of Option B for testing frequency must be consistent with plant-specific operating experience (OE).

US NRC Information Notice 92-20, “Inadequate Local Leak Rate Testing,” describes OE of inadequate local leak rate testing of two-ply steel expansion bellows that were used on some piping penetrations.

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 Part 50, Appendix J, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.” 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 50.72, “Immediate Notification Requirements for Operating Nuclear Power Reactors.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.73, “Licensee Event Report System.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWE, Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants.” New York, New York: The American Society of Mechanical Engineers. 2008.

_____. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components, Subsection IWL, Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants.” New York, New York: The American Society of Mechanical Engineers. 2008.

NEI. NEI 94-01, “Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J.” Revision 2-A. Washington, DC: Nuclear Energy Institute. October 2008.

_____. NEI 94-01, “Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J.” Revision 3-A. Agencywide Documents Access and Management System (ADAMS) Accession No. ML12221A202. Washington, DC: Nuclear Energy Institute. July 2012.

US NRC. “Final Safety Evaluation for Electric Power Research Institute (EPRI) Report No. 1009325(Archived), Revision 2, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals.” ADAMS Accession ML072970208. Washington, DC: U.S. Nuclear Regulatory Commission. August 2007.

_____. “Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR, Part 50, Appendix J.” ADAMS Accession No. ML081140105. Washington, DC: U.S. Nuclear Regulatory Commission. June 2008.

_____. Information Notice 92-20, “Inadequate Local Leak Rate Testing.” ADAMS Accession No. ML031200473. Washington, DC: U.S. Nuclear Regulatory Commission. March 1992.

_____. Regulatory Guide 1.163, “Performance-Based Containment Leak-Test Program.” Revision 0. ADAMS Accession No. ML003740058. Washington, DC: U.S. Nuclear Regulatory Commission. September 1995.

Aging Management Program

United States of America

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

Nuclear Energy Institute

American Society of Mechanical Engineers

Quality Assurance

Operating Experience

Code of Federal Regulations - The US federal government regulations. 10 CFR 54 is the regulation for renewal of nuclear power plant operating licenses.

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

Electric Power Research Institute