ML19325D824

R. E. Ginna Nuclear Power Plant Issuance of Amendment No. 136 to Revise Technical Specification 5.5.15, Containment Leakage Rate Testing Program, to Extend Containment Integrated Leak Rate Test Frequency

ML19325D824
Person / Time
Site:	Ginna
Issue date:	12/23/2019
From:	V Sreenivas Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co
Sreenivas V, NRR/DORL/LPL1, 415-2596
References
EPID L-2019-LLA-0027
Download: ML19325D824 (35)

Text

December 23, 2019 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

R. E. GINNA NUCLEAR POWER PLANT - ISSUANCE OF AMENDMENT NO. 136 TO REVISE TECHNICAL SPECIFICATION 5.5.15, CONTAINMENT LEAKAGE RATE TESTING PROGRAM, TO EXTEND CONTAINMENT INTEGRATED LEAK RATE TEST FREQUENCY (EPID L-2019-LLA-0027)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 136 to Renewed Facility Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant in response to your application dated February 15, 2019.

The amendment revises Technical Specification 5.5.15, Containment Leakage Rate Testing Program, to adopt Nuclear Energy Institute (NEI) 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J. Specifically, the amendment allows the maximum interval for the Integrated Leakage Rate Test (ILRT), also known as Type A test, to be extended permanently from once in 10 years to once in 15 years, and an administrative change to remove the exception under Technical Specification 5.5.15 for the one-time 15-year Type A test interval being performed prior to May 31, 2011.

A copy of the related safety evaluation is also enclosed. Notice of issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

V. Sreenivas, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1. Amendment No. 136 to DPR-18

2. Safety Evaluation cc: Listserv

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-244 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 136 Renewed License No. DPR-18

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated February 15, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 136, are hereby incorporated in the renewed license. Exelon Generation shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA - J. Kim for/

James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 23, 2019

ATTACHMENT TO LICENSE AMENDMENT NO. 136 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following pages of Renewed Facility Operating License No. DPR-18 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License Remove Insert 3 3 Technical Specifications Remove Insert 5.5-11 5.5-11

(b) Exelon Generation pursuant to the Act and 10 CFR Part 70, to possess and use four (4) mixed oxide fuel assemblies in accordance with the RG&E's application dated December 14, 1797 (transmitted by letter dated December 20, 1979), as supplemented February 20, 1980, and March 5, 1980; (3) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70 to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required.

(4) Exelon Generation pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Exelon Generation pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at steady-state power levels up to a maximum of 1775 megawatts (thermal}.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 136, are hereby incorporated in the renewed license.

Exelon Generation shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection Exelon Generation shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee's amendment request dated March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015.

Except where NRC approval for changes or deviations is required R. E. Ginna Nuclear Power Plant Amendment No. 136

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 136 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 EXELON GENERATION COMPANY, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244

1.0 INTRODUCTION

By application dated February 15, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19045A282), Exelon Generation Company, LLC (Exelon, the licensee) submitted a license amendment request (LAR) for R. E. Ginna Nuclear Power Plant (Ginna, GNPP). The LAR would revise Technical Specification (TS) 5.5.15, Containment Leakage Rate Testing Program, to adopt Nuclear Energy Institute (NEI) 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (ADAMS Accession No. ML100620847), as the guidance document for the implementation of the performance-based Option B of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J.

The amendment would allow the maximum interval for the integrated leakage rate test (ILRT),

also known as Type A test, to be extended permanently from once in 10 years to once in 15 years, provided acceptable performance history and other requirements as stated in NEI 94-01, Revision 2-A, are maintained, and an administrative change to remove the exception under TS 5.5.15, paragraph a, for the one-time 15-year Type A test interval being performed after May 31, 1996, and prior to May 31, 2011, since it is no longer applicable.

2.0 REGULATORY EVALUATION

2.1 Regulatory Requirements and Guidance The U.S. Nuclear Regulatory Commissions (NRC, the Commission) regulatory requirements related to the content of the TSs are set forth in 10 CFR 50.36, Technical specifications. This regulation requires that the TSs include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in plant TSs.

Enclosure 2

The regulations in 10 CFR 50.54(o) require that the primary reactor containments for water-cooled power reactors shall be subject to the requirements set forth in 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. This appendix includes two options: Option A Prescriptive Requirements and Option B Performance-Based Requirements, either of which can be chosen for meeting the requirements of Appendix J.

The testing requirements in 10 CFR Part 50, Appendix J, ensure that: (a) leakage through the containments or systems and components penetrating the containments does not exceed allowable leakage rates specified in the TSs, and (b) the integrity of the containment structure is maintained during its service life.

Option B of 10 CFR Part 50, Appendix J, specifies performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by performing a Type A test to measure the containment system overall integrated leakage rate of the primary containments; a Type B test consisting of a pneumatic test to detect and measure local leakage rates across pressure-retaining leakage-limiting boundaries; and a Type C test consisting of a pneumatic test to measure containment isolation valve leakage rates. After the preoperational tests, these tests are required to be conducted at periodic intervals based on the historical performance of the overall containment system (for Type A tests) and based on the safety significance and historical performance of each penetration boundary and isolation valve (for Type B and Type C tests) to ensure the integrity of the overall containment system as a barrier to fission product release.

The overall integrity (structural and leaktight integrity) of the primary containment is verified by a Type A ILRT, and the integrity of the penetrations and isolation valves is verified by Type B and Type C local leak rate tests (LLRT), as required by 10 CFR Part 50, Appendix J. These tests are performed to verify the essential leaktight characteristics of the containment structure at the design-basis accident pressure. The Type A test also provides a verification of structural integrity.

The leakage rate test results must not exceed the maximum allowable leakage rate (La) with margin, as specified in the TSs. Option B also requires that a general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration that may affect the containment leaktight integrity must be conducted prior to each Type A test, and at a periodic interval between tests.

Section V.B.3 of 10 CFR Part 50, Appendix J, Option B, requires that the regulatory guide (RG) or other implementation document used by a licensee to develop a performance-based leakage-testing program must be included, by general reference, in the plant TSs. Further, the submittal for TS revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the NRC and endorsed in an RG.

The NRC staffs final safety evaluation (SE) for NEI 94-01, Revision 2, and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, dated August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated June 25, 2008 (ADAMS Accession No. ML081140105), was incorporated into NEI 94-01, Revision 2-A, dated November 19, 2008. NEI 94-01, Revision 2-A, describes an NRC-approved approach for implementing the optional performance-based requirements of Option B described in 10 CFR Part 50, Appendix J, which includes provisions for extending Type A ILRT intervals to up to 15 years, and incorporates the regulatory positions stated in RG 1.163,

Performance-Based Containment Leak-Test Program, dated September 1995 (ADAMS Accession No. ML003740058). NEI 94-01, Revision 2-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. This method uses industry performance, plant-specific data, and risk insights in determining the appropriate testing frequency, and also discusses the performance factors that licensees must consider in determining test intervals.

The NRC staffs final SE dated June 8, 2012 (ADAMS Accession No. ML121030286), of NEI 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, was incorporated into NEI 94-01, Revision 3-A, dated July 2012 (ADAMS Accession No. ML12221A202). NEI 94-01, Revision 3-A, documents the NRC staffs evaluation and acceptance of NEI 94-01, Revision 3.

The regulations in 10 CFR 50.55a contain the containment inservice inspection program requirements that, in conjunction with the requirements of 10 CFR Part 50, Appendix J, ensure the continued leaktight and structural integrity of the containment during its service life.

The regulations in 10 CFR 50.65(a) state, in part, that the licensee:

shall monitor the performance or condition of structures, systems, or components, against licensee-established goals, in a manner sufficient to provide reasonable assurance that these structures, systems, and components, as defined in paragraph (b) of this section, are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience.

EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, provides a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance (ADAMS Accession No. ML100620847). NEI 94-01, Revision 3-A, Section 9.2.3.1, states that Type A ILRT intervals of up to 15 years are allowed by this guideline. The risk impact assessment of extended integrated leak rate testing intervals, EPRI Report No. 1018243, 1 Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (i.e., formerly Report No. 1009325, Revision 2-A), indicates that, in general, the risk impact associated with ILRT interval extensions for intervals up to 15 years is small. However, plant-specific confirmatory analyses are required.

The NRC staff reviewed NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2. For NEI 94-01, Revision 2, the NRC staff determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals up to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff found that the Type A testing methodology as described in ANSI/ANS-56.8-2002, and the modified testing frequencies recommended by NEI 94-01, Revision 2, serve to ensure the continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual penetrations are essentially leaktight. In addition, aggregate Type B and Type C leakage rates support the leakage tightness of the primary containment by minimizing potential leakage paths.

1 EPRI Report No. 1018243, is also identified as EPRI Report No. 1009325, Revision 2-A. This report is publicly available and can be found at www.epri.com by typing 1018243 in the search field box.

For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant-specific risk insights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRC staff found that the proposed methodology satisfies the key principles of risk-informed decisionmaking applied to changes to TSs as delineated in RG 1.177, Revision 1, An Approach to Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications (ADAMS Accession No. ML100910008), and RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML17317A256). The NRC staff found that this guidance was acceptable for referencing by licensees proposing to amend their TSs regarding containment leakage rate testing subject to the limitations and conditions noted in Section 4.2 of the safety evaluation report (SER) for EPRI Report No. 1009325, Revision 2.

In 2012, NEI 94-01, Revision 3, was issued. The NRC staff reviewed NEI 94-01, Revision 3, and determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J, as modified by the conditions and limitations summarized in Section 4.0 of the associated SE. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual containment isolation valves are essentially leaktight. In addition, aggregate Type C leakage rates support the leakage tightness of the primary containment by minimizing potential leakage paths. The NRC staff, therefore, found that this guidance, as modified to include two limitations and conditions, was acceptable for referencing by licensees proposing to amend their TSs regarding containment leakage rate testing. Any applicant may reference NEI 94-01, Revision 3-A, as modified by the associated SER and approved by the NRC, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, in a licensing action to satisfy the requirements of Option B to 10 CFR Part 50, Appendix J. Exelon evaluated the additional extension of Type C intervals afforded by NEI 94-01, Revision 3-A, and chose not to adopt NEI 94-01, Revision 3-A, for Ginna at this time.

However, the risk assessment performed to permanently extend the currently allowed containment Type A ILRT to 15 years used the methodology currently endorsed by NEI 94-01, Revision 3-A, for the required confirmatory risk impact assessment, as this is the most up-to-date guidance available.

3.0 TECHNICAL EVALUATION

3.1 Background By Amendment No. 61 dated February 13, 1996 (ADAMS Accession No. ML010640012), the NRC approved a change to Ginna TS 5.5.15 to permanently increase the period of the primary containment Type A test interval to 10 years. However, before the next Type A test became due, the licensee applied for a one-time extension of the ILRT to 15 years, which was approved by Amendment No. 93, dated December 8, 2005 (ADAMS Accession No. ML053130384).

Based on this approval, the last ILRT at Ginna was performed in 2011. Therefore, the exception to NEI 94-01-1995, Section 9.2.3, denoted by TS 5.5.15, paragraph a, stating that the next Type A test performance for Ginna shall be by May 31, 2011, is no longer necessary as that date has passed with the test having been completed in 2011.

After the completion of the ILRT in 2011, the test frequency reverted to a 10-year interval, and the licensee stated that the next Type A test is required by the spring 2020 outage. The proposed change will allow Ginna to perform the next Type A test approximately 5 years later than the spring 2020 outage, and on a 15-year interval from then onward.

With the proposed change, Ginna will implement NEI 94-01, Revision 2-A, and the limitations and conditions contained within Section 4.1 of the SE for NEI 94-01, Revision 2. NEI 94-01, Revision 2-A, provides that extension of the Type A test interval to 15 years shall be based on two consecutive successful Type A tests (performance history) and other requirements stated in Section 9.2.3 of NEI 94-01. The basis for acceptability of extending the Type A test interval also includes implementation of robust Type B and Type C testing of the penetration barriers where most containment leakage has historically been shown to occur and is expected to continue to be the pathway for a majority of potential primary containment leakage, and of a robust containment visual inspection program where deterioration of the primary containment boundary away from penetrations can be detected and remediated before any actual significant leakage potential can occur.

The proposed change would replace the TS reference to RG 1.163 with a reference to NEI 94-01, Revision 2-A. This is allowed by the provision in 10 CFR Part 50, Appendix J, Option B,Section V.B.3, regarding the TSs referencing the NRC staff-approved guidance document for program implementation.

Ginna TS 5.5.15 contains the following definitions for Pa and La:

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 60 psig [pounds per square inch gauge].

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.2% of containment air weight per day.

Containment leakage rate acceptance criterion is 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests.

3.2 System Description The following description of the containment is a condensed version of the information provided in the LAR.

Ginna is a Westinghouse two-loop pressurized water reactor type design. The reactor containment structure is a reinforced concrete, vertical right cylinder with a flat base and a hemispherical dome. It ensures that leakage of radioactive materials to the environment is minimized, even if gross failure of the reactor coolant system were to occur. A welded steel liner is attached to the inside face of the concrete shell to ensure a high degree of leaktightness.

The thickness of the liner in the cylinder and dome is 3/8 inch (in.) and in the base it is 1/4 in.

The cylindrical reinforced concrete walls are 3 feet (ft.), 6 in. thick, and the concrete hemispherical dome is 2 ft., 6 in. thick. The concrete base slab is 2 ft. thick with an additional 2-ft.-thick concrete fill over the bottom liner plate. The containment structure is 99 ft. high to the spring line of the dome and has an inside diameter of 105 ft. The containment vessel provides a minimum free volume of approximately 972,000 cubic feet (ft3). The reactor vessel is in the center of the containment structure below ground level.

The containment leakage pressure boundary is provided by the single steel liner in the containment vessel. Each system for which its piping penetrates this boundary is designed to maintain isolation of the containment from the outside environment. The containment structure

and all penetrations are designed to withstand, within design limits, the combined loadings of the design-basis accident and design seismic conditions. All piping systems that penetrate the containment are anchored in the penetration sleeve or the structural concrete of the containment structure. The penetrations for the main steam, feedwater, blowdown, and sample lines are designed so that the penetration is stronger than the piping system and the containment will not be breached due to a postulated pipe rupture. The liner thickness near typical penetrations is increased to a minimum of 3/4 in. All lines connected to the primary coolant system that penetrate the containment are also anchored in the secondary shield walls (i.e., walls surrounding the steam generators and reactor coolant pumps) and are each provided with at least one valve between the anchor and the reactor coolant system. For mechanical penetrations that interface with hot fluid systems, a containment penetration cooling system is used to prevent the bulk concrete temperature surrounding the penetrations from exceeding 150 degrees Fahrenheit (°F). Containment electrical penetrations are designed so the containment structure can, without exceeding the design leakage rate, accommodate the postulated environment resulting from a loss-of-coolant accident. The electrical penetrations have been shown to maintain structural integrity when subjected to mechanical stresses caused by large magnitude fault currents.

3.2.1 Fuel Transfer Penetration A fuel transfer penetration is provided for fuel movement between the refueling transfer canal in the reactor containment and the spent fuel pool. The penetration consists of a stainless-steel pipe installed inside a larger pipe. The inner pipe acts as the transfer tube and connects the refueling canal with the spent fuel pool. The tube is fitted with a standard stainless-steel flange in the refueling canal and a stainless-steel sluice gate valve in the spent fuel pool. The outer pipe is welded to the containment liner. The fuel transfer penetration, like all other penetrations, is anchored in the containment shell. Because this anchor point moves when the containment vessel is subjected to load, expansion joints are provided where the penetration is connected to structures inside and outside of the containment vessel. Since the penetration is located on a skewed angle not normal to the containment shell, the expansion joints are subjected to both radial and tangential (lateral) motions. The expansion bellows inside the containment vessel provide a water seal for the refueling canal and accommodate thermal growth of the penetration from the anchor, as well as the pressure and earthquake produced motion of the anchor (the containment shell). The expansion joint accommodates motion of the sleeve within the containment shell relative to the portion of the sleeve anchored in the wall of the refueling canal in the auxiliary building.

3.2.2 Equipment Hatch and Personnel Hatch An equipment hatch constructed of welded steel and having a double-gasketed flange and bolted dished door is located near grade. The equipment hatch has a diameter of 14 ft. and is used for transportation of equipment through the containment wall.

Two personnel accesses are provided. One personnel hatch penetrates the dished door of the equipment hatch. The other is located diametrically opposite the equipment hatch. Each personnel hatch is a hydraulically-latched double door, welded steel assembly. An equalizing valve connects each personnel hatch with the interior of the containment vessel for equalizing pressure in the personnel hatch with that in the containment. Hatch closures are the double-tongue, single gasket type. The access locks are properly interlocked to ensure door closure at all times.

3.3 Testing Requirements Implementation Section 50.54(o) of 10 CFR requires that the primary reactor containments for water-cooled power reactors shall be subject to the requirements set forth in Appendix J to 10 CFR Part 50.

Appendix J to 10 CFR Part 50 includes two options: Option A - Prescriptive Requirements and Option B - Performance-Based Requirements, either of which can be chosen for meeting the requirements of the appendix. The testing requirements in 10 CFR Part 50, Appendix J, ensure that (a) leakage through the containments or systems and components penetrating the containments does not exceed allowable leakage rates specified in the TSs, and (b) the integrity of the containment structure is maintained during its service life.

Option B of 10 CFR Part 50, Appendix J, specifies performance-based requirements and criteria for preoperational and subsequent leakage-rate testing. These requirements are met by performance of Type A tests to measure the containment system overall integrated leakage rate; Type B pneumatic tests to detect and measure local leakage rates across pressure-retaining leakage-limiting boundaries such as penetrations; and Type C pneumatic tests to measure containment isolation valve (CIV) leakage rates. After the preoperational tests, these tests are required to be conducted at periodic intervals based on the historical performance of the overall containment system (for Type A tests) and based on the safety significance and historical performance of each boundary and isolation valve (for Type B and Type C tests) to ensure the integrity of the overall containment system as a barrier to fission product release.

In 1995, 10 CFR Part 50, Appendix J, was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach.

Performance-based test intervals are based on consideration of the operating history of the component and the resulting risk from its failure. The use of the term performance-based in 10 CFR Part 50, Appendix J, refers to both the performance history necessary to extend test intervals, as well as to the criteria necessary to meet the requirements of Option B.

Also, RG 1.163 endorsed NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 21, 1995 (ADAMS Accession No. ML11327A025), with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on the NRC risk assessment contained in NUREG-1493, Performance Based Containment Leak-Test Program, dated January 1995, and EPRI Report No. 104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, dated August 1994, both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months were considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, with indication that this extension of interval should be used only in cases where refueling schedules have been changed to accommodate other factors.

Exelon has voluntarily adopted and has been implementing Option B for meeting the testing requirements of Appendix J to 10 CFR Part 50 for Ginna. The NRC staff issued the associated amendment to TS 5.5.15 by Amendment No. 61, dated February 13, 1996. In addition, the licensee also applied for a one-time extension to the interval for completing the next Type A test

pursuant to 10 CFR Part 50, Appendix J, from 10 years to 15 years, which was approved by Amendment No. 93 on December 8, 2005.

Based on the issuance of Amendment Nos. 61 and 93, current Ginna TS 5.5.15 states, in part:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J:

a. Section 9.2.3: The first Type A test performed after the May 31, 1996 Type A test shall be performed by May 31, 2011.

NEI 94-01, Revision 2, has been reviewed by the NRC and approved for use. The final SE for Revision 2, issued by letter dated June 25, 2008, states that NEI 94-01, Revision 2, describes an acceptable approach for implementing the optional performance-based requirements of 10 CFR Part 50, Appendix J, Option B. NEI 94-01, Revision 2, incorporates the regulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals up to 15 years. The final approved version, NEI 94-01, Revision 2-A, was published on November 19, 2008, and includes the NRC SE. The proposed change for Ginna reflects an increase to the existing Type A ILRT program interval from 10 to 15 years in accordance with NEI 94-01, Revision 2-A. In addition, the LAR proposes an administrative change to remove the exception in the current TSs for a one-time extension of the ILRT that was previously implemented by Amendment No. 93.

NEI 94-01, Revision 3-A, also provides that a maximum interval of 120 months could be allowed for Type B tests and a maximum interval of 75 months could be allowed for Type C tests with a limited provision for extension or grace period up to 9 months allowed for these LLRTs. The licensee has chosen not to adopt NEI 94-01, Revision 3-A, at this time. However, the licensee stated that the risk assessment performed in support of permanently extending the containment Type A test used the methodology currently endorsed by NEI 94-01, Revision 3-A, for the required confirmatory risk impact assessment, as it is the most up-to-date guidance available.

3.4 Licensees Proposed Changes The current TS 5.5.15 states, in part:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J:

a. Section 9.2.3: The first Type A test performed after the May 31, 1996 Type A test shall be performed by May 31, 2011.

The proposed changes to TS 5.5.15 would revise the above statements as follows:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 2-A, dated October 2008.

The remaining text of TS 5.5.15 would be unaffected by the proposed changes.

3.3 Historical Leakage Rate Test Results The licensee provided historical results of ILRT Type A tests and Type B and Type C LLRTs.

3.3.1 Historical Type A Test (ILRT) Results In LAR Table 3.3.4-1, GNPP Type A Testing History, the licensee presented the historical results of the Type A ILRT tests starting from the pre-operational test in November 1969. The results are summarized in Table 3.2-1 below (the first two tests in November 1969 and October 1972 were omitted, as they do not contain any information pertinent to this SE):

Table 3.2-1, GNPP Type A ILRT History Test Date 95% Upper As-Found Acceptance As-Left (AL) Acceptance Confidence (AF) Criteria Leakage Criteria, 0.75La Limit Leakage (weight (weight (weight %/day)

(weight (weight  %/day)  %/day)

%/day)  %/day) 2/1976 (1) N/A (2) 0.1146 0.0440 0.1146 3/1978 (1) 0.04900 0.05092 0.1146 0.0490 0.1146 3/1982 (1) 0.01970 (3) 0.1146 0.0197 0.1146 3/1986 (1) 0.06407 0.06741 0.1146 0.06407 0.1146 5/1989 (1) 0.04631 0.04632 0.1146 0.04631 0.1146 4/15/1993 (1) 0.05383 0.05554 0.1146 0.05550 0/.1146 6/10/1996 0.11967 0.11969 0.2000 (4) & 0.11967 0.1500 (4) & (5)

(5) 6/2/2011 0.10260 0.1313 0.2000 0.1078 0.1500 Notes:

(1) Reduced pressure tests performed at 35 psig. The only full pressure (60 psig) and reduced pressure test completed was in 11/1969, which was a pre-operational test. Subsequent tests were conducted at reduced pressure (35 psig), until 1994, when ANSI/ANS 56.8 required ILRT test pressure to be greater than (0.96) times (peak post-accident pressure (Pa)).

(2) The ILRT that began on February 7, 1976, was aborted after approximately seven hours into the equalization period due to leakage that exceeded the

test acceptance criteria of 0.75La (0.1146 wt.%/day). The failure of this Type A test was the result of leakage through the purge exhaust dampers and check valve 1713. Abnormally high leakage rates through the purge dampers have historically occurred only when the plant is in the cold shutdown condition. The pathway from containment through check valve 1713 leads into a normally closed nitrogen system outside containment, which is pressurized to 90 psig. Venting of this system both inside and outside containment was performed for the Type A test.

(3) The containment purge system at penetration 204 (purge air supply) and penetration 300 (purge air exhaust) were modified in September 1987, to provide greater assurance of containment integrity. The 48 in. inboard CIV at each of these two penetrations was removed and replaced with a blind flange. The blind flanges on penetrations 204 and 300 are removed and reassembled during plant outages, when required. These blind flanges remain in place when the reactor is in a mode where containment integrity is required.

(4) The Ginna As-Found Leakage rate exceeded its limit of 0.75La (0.1146 wt.%/day). After plant shutdown for the 1982 refueling outage (RFO), AF Types B and C tests were performed. During this testing, the leakage rates for the Purge Supply (5870) and Exhaust (5878), as well as check valve 1599, were found to be excessive. The measured leak rates on the Purge Supply (5870) and Purge Exhaust (5878) were 15,593 cubic centimeters per minute (cc/min) and 244,142 cc/min, respectively. The leak rate on check valve 1599 was not quantified. Because of this, the Leakage Savings for this ILRT could not be determined and the AF leak rate could not be calculated. This ILRT was performed in March 1982. It should be noted that the Type C total leak rate summation at the end of 1981 was 1,016 cc/min. The measured leak rates for the Purge Supply (5870) and Exhaust (5878) were found to be 174 cc/min and 362 cc/min, respectively, on November 18, 1981.

(5) Check valve 1599 was subsequently replaced in November 1982 with an Air-Operated Valve (AOV) to provide greater assurance of proper valve operability and leak tight closure.

As stated above in Note 3, the containment purge system at penetration 204 (purge air supply) and penetration 300 (purge air exhaust) were modified in 1987 to provide greater assurance of containment integrity. The blind flanges will replace the valves when the reactor is in a mode where containment integrity is required (6) NEI 94-01, Revision 0, issued on July 26, 1995, specifies the acceptance criteria of 1.0 La for the AF ILRT leakage. Prior to this, the acceptance criterion for the AF ILRT leakage was 0.75 La. Therefore, for ILRT tests performed after July 26, 1995, at Ginna, the acceptance criteria specified as wt.%/day was increased from 0.15 wt.%/day to 0.2 wt.%/day. This is assuming a test pressure of Pa. See Note 1 regarding testing at reduced pressures where test pressure is less than Pa.

(7) ANSI/ANS 56.8 (1994) was approved on August 4, 1994. In Section 3.2.11, it states The Type A test pressure shall not be less than 0.96Pa. This eliminated the option to perform Type A testing at a reduced pressure. ILRT tests performed subsequent to ANSI/ANS 56.8 (1994) were tested at Pa (60 psig). Therefore, tests performed after 1994 did not require a partial pressure calculation to adjust the wt%/day for the reduced test pressure. See Note 1 above.

The LAR provided the historical results of the two most recent ILRTs in 1996 and 2011. All the results show margins to La. Option B does not allow reduced pressure Type A tests.

Section 9.2.3, Extended Test Intervals, of NEI 94-01, Revision 2-A, states that [i]n the event where previous Type A tests were performed at reduced pressure (as described in 10 CFR [Part] 50, Appendix J, Option A), at least one of the two consecutive periodic Type A tests shall be performed at peak accident pressure (Pa). The NEI 94-01, Revision 2-A, requirement for allowing the extended test interval is that the past two tests meet the performance criterion, demonstrating a leakage of La or less. The 1996 and 2011 ILRT results indicate leakage less than the allowable performance criterion of 0.2 weight % per day, with as-found leakages of 0.11969 and 0.1313 weight % per day, equivalent to approximately 60 percent and 66 percent of the allowable leakage for the 1996 and 2011 ILRT tests, respectively. Similarly, the as-left leakages of 0.11967 and 0.1078 are equivalent to approximately 80 percent and 72 percent of the allowable leakage of 0.15 weight % per day for the 1996 and 2011 ILRT tests. Since both tests were satisfactory, the Ginna ILRT remains on an extended frequency. The current ILRT frequency for Ginna is 10 years.

The results provided in Section 3.3.4 of the LAR and the Ginna Type A ILRT testing history demonstrate a history of adequately managing the leakage rates, thereby continuing to provide a high degree of assurance that containment leaktight integrity is being satisfactorily maintained in accordance with the requirements of TS 5.5.15, and support the licensees proposed permanent extension request of the current Type A ILRT frequency from 10 years to 15 years.

3.3.2 Historical Type B and Type C Test (LLRTs) Results Ginna TS 5.5.15 states, in part:

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is 1.0 La. During the first plant startup following testing in accordance with this program, the leakage rate acceptance criteria are 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests.

The containment performance is demonstrated by the as-found minimum pathway summations, whereas the as-left maximum pathway summations signify the acceptance criteria for restart.

The licensee submitted the local leak rate summaries in Table 3.5.5-1, GNPP Type B and C LLRT Combined As-Found/As-Left Trend Summary, of Attachment 1 to the LAR. The summary was provided for eight refueling outages, from year 2008 to year 2018. As noted earlier, La in TS 5.5.15 is 0.2 percent of primary containment air weight per day. The acceptance criteria in TS 5.5.15 for combined Type B and Type C test total is 0.6 La. In Table 3.5.5-1 of Attachment 1 to the LAR, the licensee provided the as-found minimum pathway and as-left maximum pathway leakage rates for the eight refueling outages as a percent fraction of 0.6La. The results show that:

• The as-found minimum pathway leakage rates for the last eight refueling outages since 2008 vary from a minimum of 2.84 percent of 0.6La to a maximum of 13.49 percent of 0.6La.

• The as-left maximum pathway leakage rates for the last eight refueling outages since 2008 vary from a minimum of 4.40 percent of 0.6La to a maximum of 7.73 percent of 0.6La.

Section 3.5.6 of Attachment 1 to the LAR provided the following additional details regarding Type B and Type C component performance.

Type B Components The percentage of the total number of Ginna Type B tested components that are on a 60-month extended performance-based test interval is 64.0 percent.

GNPP did not adopt the maximum test frequency of a 120-month interval allowed under NEI 94-01, Revision 0. The Type B components not on a 60-month test frequency are either used during RFOs and, therefore, must be as-left tested each RFO subsequent to use, or are limited to a test frequency of 30 months per TS 5.5.15.

Type C Components The percentage of the total number of Ginna Type B tested components that are on a 60-month extended performance-based test interval is 76.04 percent.

The Type C penetrations not on a 60-month test frequency are either on a 30-month frequency following valve replacement or major maintenance to reestablish their performance history of two satisfactory sequential as-found tests, or are used or removed during RFOs to support flex or outage requirements, or tested on an RFO frequency to satisfy inservice testing 24-month test frequency requirements.

The licensee stated that no Type B or Type C components on an extended test frequency have exceeded their administrative limits over the last five refueling outages (RFOs). The results suggest that performance criteria are unlikely to be exceeded by the use of extended LLRT intervals.

Based on the review of the data contained in Table 3.5.5-1 of Attachment 1 to the LAR, the NRC staff concludes that the aggregate results of the as-found/as-left trend summary for the Type B and Type C tests from 2008 to 2018 demonstrate a history of successful tests.

Furthermore, the table shows that there were no as-found failures that resulted in exceeding the TS 5.5.15, paragraph a, acceptance limit of 0.6La. The as-found minimum pathway summations represent an acceptable quality of maintenance of Type B and Type C components. Since the Type B and Type C leakages constitute a significant portion of the Type A leakages, the substantial margin apparent in the results of the Type A and Type B tests suggests that performance criteria are unlikely to be exceeded as a result of an extended interval for the Type A ILRT tests to 15 years.

3.3.3 Containment Inservice Inspection Program The Ginna containment is a steel-lined, reinforced concrete, vertical right cylinder with a flat base and a hemispherical dome that employs the use of a post-tensioned pre-stressing tendons. The containment is discussed in Section 3.8.1.1 of the Ginna Updated Final Safety Analysis Report (UFSAR).

In Section 3.5.3 of the LAR, the licensee stated that it is implementing its containment inservice inspection (CISI) program in accordance with the applicable edition/addenda of Subsections IWE/IWL of Section XI, Division 1 of the American Society of Mechanical Engineers (ASME) Code, subject to the applicable regulatory conditions as required by 10 CFR 50.55a(g)(4)(iv). The CISI plan for ASME Class metal containment (MC) and concrete containment (CC) components for the second 10-year CISI interval has been developed using the ASME Code,Section XI, 2004 Edition, except where specific written alternatives from Code requirements has been requested by Exelon and granted by the NRC, or as amended by the NRC in 10 CFR 50.55a. As of February 15, 2019 (the submittal date of the LAR), the licensee stated in Section 3.5.3.1 that the third 10-year CISI plan, which begins in January 2020, is currently under development and that the code of record will be the 2013 Edition of ASME Section XI. The present second 10-year CISI interval concludes on December 31, 2019, several months before the planned performance of the extended ILRT scheduled for the spring 2020 RFO. Subsection IWL provides the rules and requirements for inservice inspection of Class CC components. Subsection IWE provides the rules and requirements for inservice inspection of Class MC pressure-retaining components and requires general visual examination of 100 percent of accessible metallic surfaces of the containment pressure boundary three times over a 10-year inspection interval, pursuant to 10 CFR 50.55a(b)(2)(ix)(E). Pursuant to IWL-2410(a) and (c), general visual examinations of accessible surfaces of containment concrete and post-tensioning system components of the containment are conducted every 5 years, which would be two examinations over a 10-year interval.

In LAR Section 3.6.6, the licensee provided a discussion of the results of recent containment inspections performed of Service Level I primary containment coated surfaces by Underwater Engineering Services, Inc. during RFO G1R39 (fall 2015) and RFO G1R40 (spring 2017).

Interior surfaces of the primary containment, components, and equipment were inspected and assessed. Areas included the containment liner wall, basement level, intermediate level, operation levels, and containment dome. Although deficiencies were noted during both assessments, the licensee concluded that the overall coating system inside the primary containment building is in fair to good condition with no current coating conditions observed that could impact structural integrity, plant operations, or safe shutdown.

In 2014 and 2017, the licensee also performed Appendix J containment visual structural inspections of the accessible interior and exterior surfaces of the containment structure with existing liner plate insulation panels in place. The inspection is designed to uncover any evidence of structural deterioration affecting the structural integrity of the containment. The inspections were found to be satisfactory with no structural issues identified.

Several containment structure post-tensioning system tendon surveillances were performed in 2010, 2011, and 2016 by Precision Surveillance Corporation to assess the quality and structural performance of the post-tensioning system and consisted of periodic inspections of a selected group of tendons. The examination was performed in accordance with the requirements of the ASME Code,Section XI, 2004 Edition, and the applicable amendments specified in 10 CFR 50.55a. The surveillance was conducted in accordance with IWL-3221, Un-bonded

Post-Tensioning Systems. The licensee stated that all the tendon liftoffs and tendon wire samples were found to be within acceptable levels required by the Ginna Tendon Surveillance Program, STP-O-27.2.

Recent IWE examinations performed to ensure that the structural integrity of the ASME Class MC containment liner was maintained were performed during RFO G1R39 (fall 2015) and RFO G1R40 (spring 2017). During RFO G1R39, the moisture barrier was 100 percent examined. A VT-3 inspection of the moisture barrier identified unacceptable heavy rusting in eight spots.

Following VT-1 and ultrasonic examination, the area was recoated, and a subsequent examination was performed with acceptable results. A portion of the containment liner plate was found to have a lack of complete coating with no structural concerns. Following nondestructive examination, the affected area was determined to be acceptable. A VT-3 examination was performed after coating repairs were made to the liner and found to be acceptable with no reportable indications. Issues were also identified with degraded caulking and signs of discoloration on the concrete floor of the containment building basement and rust observed on the carbon steel containment liner plate. The licensee stated that the most likely cause for the observed degradation of the liner was the exposure to borated water leakage from the reactor cavity during refueling activities. Following ultrasonic examinations of the containment liner and recoating and re-caulking of the moisture barrier, the licensee stated that the identified degradation of the liner has stopped.

During RFO G1R40, the containment vessel dome liner and several mechanical penetrations were examined. Indications of staining (discoloration, rust, and oxidation) were identified on the containment dome surface and compared to indications recorded during the 2015 RFO examinations. Additional areas of rust and staining identified during the 2018 RFO examinations were compared against the 2015 results and were found to be acceptable. As a result of VT-3 examinations performed for mechanical penetrations 321A (steam generator blowdown) and mechanical penetration 412 (main steam from B steam generator), light rust on welds and minor corrosion on the penetration to the liner weld were identified, with no material loss observed.

IWL examinations were performed in 2010 and 2015 to ensure that the structural integrity of the ASME Class CC reinforced concrete was maintained. Condition assessment of the concrete was achieved by the performance of visual examinations of the accessible surfaces. For example, during the assessment, grease was noted at the base and top of tendon can No. 130, along with insufficiently engaged nuts for tendon can Nos.111, 117, and 118. Other observations included minor intermittent concrete cracking, efflorescence, loss of material from base of tendon can Nos. 38 through 47, grouted areas between tendon cans cracking and breaking away, and degraded gaskets on several tendon cans. GNPP Engineering concluded, with few exceptions, that the indications noted were acceptable.

In Section 3.8 of the LAR, the licensee also identified several license renewal aging management programs for the Ginna containment. As part of the license renewal effort, the licensee demonstrated that commitments related to the aging effects applicable for the systems, structures, and components within the scope of license renewal would be adequately managed during the period of extended operation. The renewed operating license for Ginna was issued on May 19, 2004 (ADAMS Accession No. ML041330109), extending the original licensed operating term by 20 years. The following aging management programs, consistent with the corresponding programs described in NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report (ADAMS Accession No. ML103490041), and related activities are discussed in the UFSAR and credited with the aging management of the Primary Containment:

10 CFR 50 Appendix J Program, which monitors leakage rates through the containment pressure boundary, including penetrations and access openings; UFSAR Section 18.2.1.3, IS-IWE Program, which manages aging effects for the containment liners and its integral attachments, including connecting penetrations and parts forming the leaktight boundary; ISI-IWL Program, which manages the reinforced concrete and unbonded post-tensioning systems of the containment structures; and Protective Coating Monitoring and Maintenance Program, which is essential to ensure operability of post-accident safety systems that rely on water recirculated through the containment emergency sump B.

Based on the licensees evaluation, the inspection findings were deemed acceptable by the licensee, thereby not impacting the structural integrity of the containment. Therefore, based on the above, the NRC staff finds that the licensee has an adequate CISI program in place to periodically examine, monitor, and manage structural deterioration and aging degradation of the Ginna containment pressure boundary and can perform its intended function as a leaktight barrier, consistent with the guidance contained in RG 1.163.

The NRC staff finds that the licensee is satisfactorily monitoring and managing the Ginna containment and performing supplemental inspections to periodically examine and monitor aging degradation, thereby providing reasonable assurance that the containment structural and leaktight integrity will continue to be maintained. The licensee justified the proposed change to extend the performance-based Type A ILRT test interval by demonstrating adequate performance of the containment based on plant-specific Type A ILRT test program results.

Therefore, based on its review, the staff finds the requested permanent extension for the Type A ILRT leakage rate test frequency from 10 years to 15 years acceptable.

3.4 NRC Conditions in NEI 94-01, Revision 2-A In the NRC SER dated June 25, 2008, the NRC staff concluded that the guidance in NEI 94-01, Revision 2, is acceptable for reference by licensees proposing to amend their TSs regarding containment leakage rate testing, subject to six conditions. The provisions of NEI 94-01 stayed essentially the same from the original version through Revision 2, except that the regulatory positions of RG 1.163 were incorporated and the maximum ILRT interval extended to 15 years.

Industry review and familiarization with these changes were extensive during the NEI 94-01 revision process. The accepted version of NEI 94-01, Revision 2, was subsequently issued as Revision 2-A. To ensure that the licensee acknowledged and satisfied the limitations and conditions in the NEI 94-01, Revision 2, SER, the NRC staff evaluated LAR Section 3.9, NRC-SER Limitations and Conditions, Table 3.9.1.1, NEI 94-01, Revision 2-A, Limitations and Conditions.

In Table 3.9.1-1, the licensee provided the following responses to the limitations and conditions.

a. NRC Condition 1 For calculating the Type A leakage rate, the licensee should use the definition in

[NEI 94-01, Revision 2], in lieu of that in ANSI/ANS-56.8-2002. (Refer to SE Section 3.1.1.1).

Ginna Response to Limitation/Condition 1 GNPP will utilize the definition in NEI 94-01, Revision 2-A, Section 5.0.

Staff Assessment Section 3.2.9, Type A test performance criterion, of ANSI/ANS-56.8-2002 defines the performance leakage rate and states, in part:

The performance criterion for a Type A test is met if the performance leakage rate is less than La. The performance leakage rate is equal to the sum of the measured Type A test UCL and the total as-left MNPLR of all Type B or Type C pathways isolated during performance of the Type A test.

Section 3.1.1.1 of the NRC staff SE of NEI 94-01, Revision 2, states, in part:

Section 5.0 of NEI TR 94-01, Revision 2, uses a definition of performance leakage rate for Type A tests that is different from that of ANSI/ANS-56.8-2002. The definition contained in NEI TR 94-01, Revision 2, is more inclusive because it considers excessive leakage in the performance determination. In defining the minimum pathway leakage rate, NEI TR 94-01, Revision 2, includes the leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position prior to the performance of the Type A test. Additionally, the NEI TR 94-01, Revision 2, definition of performance leakage rate requires consideration of the leakage pathways that were isolated during performance of the test because of excessive leakage in the performance determination. The NRC staff finds this modification of the definition of performance leakage rate used for Type A tests to be acceptable.

Section 5.0, Definitions, of NEI 94-01, Revision 2-A, states, in part:

The performance leakage rate is calculated as the sum of the Type A upper confidence limit (UCL) and as-left minimum pathway leakage rate (MNPLR) leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position (i.e., drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0La.

The NRC staff reviewed the definitions of performance leakage rate contained in NEI 94-01, Revisions 2 and 2-A. The staff determined that the definitions contained in the two revisions are identical. Therefore, the staff concludes that Ginna will use the definition found in Section 5.0 of NEI 94-01, Revision 2, for calculating the Type A leakage rate in the Containment Leakage Rate Testing Program.

Based on the above review, the NRC staff finds that the licensee has adequately addressed NRC Condition 1.

b. NRC Condition 2 The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. (Refer to SE Section 3.1.1.3).

Ginna Response to Limitation/Condition 2 Reference Sections 3.5.3 and 3.5.4 of this submittal.

Staff Assessment Section 9.2.3.2, Supplemental Inspection Requirements, of NEI 94-01, Revision 2-A, states:

To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

As stated in the LAR, Section 3.5.3, Containment Inservice Inspection Plan, the GNPP CISI plan includes ASME Code Section XI, Subsections IWE and IWL, for CISI Class MC components (metallic containments and metallic shell and penetration liners of Class CC containments) and CISI Class CC components (concrete containments) of light-water cooled nuclear power plants. This containment ISI plan also includes information related to augmented examination areas, component accessibility, and examination review.

Programmatically, each 10-year CISI interval is divided into three successive inspection periods as determined by calendar year of plant service within the inspection interval. Based on the information in the LAR, Section 3.5.3, the Ginna first CISI interval was effective from September 9, 2001, through December 31, 2009, and the Ginna second CISI interval spans from January 1, 2010, through December 31, 2019. The LAR also provided a detailed second CISI interval plan for Ginna in Table 3.5.3.1-1, GNPP Fifth Containment MC/CC Inspection Interval. The LAR further stated that the Ginna third 10-year plan is currently under development and begins in January 2020. The code of record for the next 10-year CISI interval will be the 2013 Edition of ASME Section XI. The NRC staff finds this acceptable, since a third interval CISI program plan is required to be submitted to the NRC prior to its implementation.

Section 3.5.4, Supplemental Inspection Requirements, of the LAR acknowledged that the proposed TS changes require a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leaktight integrity. In addition, the LAR also acknowledged that this inspection must be conducted prior to each Type A test and during at least three outages before the next Type A test if the interval for the Type A test is extended to 15 years in accordance with the NEI 94-01, Revision 2-A, guidance in Sections 9.2.1, Pretest Inspection and Test Methodology, and 9.2.3.2, Supplemental Inspection Requirements. The LAR also stated that inspections are required at Ginna to ensure compliance with the visual inspection requirements in the current TS SR 3.6.1.1. In summary, the licensee expects to comply with NEI 94-01, Revision 2-A, by a combination of IWE and IWL examinations scheduled in accordance with CISI, and the Appendix J to 10 CFR Part 50 primary containment inspection in accordance with the current TS SR 3.6.1.1. Therefore, the NRC staff finds that the licensee addressed and satisfied NRC Condition 2.

c. NRC Condition 3 The licensee addresses the areas of the containment structure potentially subjected to degradation (Refer to SE Section 3.1.3).

Ginna Response to Limitation/Condition 3 Reference Section 3.5.3 of this submittal.

Staff Assessment Section 3.1.3 of the NRC staff SE of NEI 94-01, Revision 2, states, in part:

In approving for Type A tests the one-time extension from 10 years to 15 years, the NRC staff has identified areas that need to be specifically addressed during the IWE and IWL inspections including a number of containment pressure-retaining boundary components (e.g., seals and gaskets of mechanical and electrical penetrations, bolting, penetration bellows) and a number of the accessible and inaccessible areas of the containment structures (e.g., moisture barriers, steel shells, and liners backed by concrete, inaccessible areas of ice-condenser containments that are potentially subject to corrosion).

In Section 3.5.3 of the LAR, the licensee stated that concrete examinations were conducted as described in IWL-2410(a) and (c) and detailed the examinations during the 2010 and 2015 IWL examinations in Table 3.6.6-1, 2010 and 2015 Concrete Indications, in Attachment 1 to the LAR.

Based on the information provided by the licensee, the NRC staff finds that the licensee has adequately addressed NRC Condition 3.

d. NRC Condition 4 The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4).

Ginna Response to Limitation/Condition 4 There are no major modifications planned. No containment or containment isolation system modifications were required at GNPP to comply with the NRC orders for FLEX. Reference Section 3.7 of this submittal.

Staff Assessment Section 3.1.4 of the NRC staff SE of NEI 94-01, Revision 2, states, in part:

Section 9.2.4 of NEI TR 94-01, Revision 2, states that: Repairs and modifications that affect the containment leakage integrity require LLRT or short duration structural tests as appropriate to provide assurance of containment integrity following the modification or repair. This testing shall be performed prior to returning the containment to operation. Article IWE-5000 of the ASME Code,

Section XI, Subsection IWE (up to the 2001 Edition and the 2003 Addenda),

would require a Type A test after major repair or modifications to the containment.

In general, the NRC staff considers the cutting of a large hole in the containment for replacement of steam generators or reactor vessel heads, replacement of large penetrations, as major repair or modifications to the containment structure.

This condition is intended to verify that any major modification or maintenance repair of the primary containment since the last ILRT has been appropriately accompanied by either a structural integrity test or an ILRT and that any plans for such major modification also include appropriate structural and leak testing. Section 3.7, Containment Modifications, of the LAR states that no containment modifications have been performed since the last ILRT in 2011.

Based on the information provided by the licensee, the NRC staff finds that the licensee has adequately addressed NRC Condition 4.

e. NRC Condition 5 The normal Type A test interval should be less than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE Section 3.1.1.2).

Ginna Response to Limitation/Condition 5 GNPP will follow the requirements of NEI 94-01 Revision 2-A, Section 9.1. In accordance with the requirements of NEI 94-01, Revision 2-A, SER Section 3.1.1.2, GNPP will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

Staff Assessment Section 3.1.1.2 of the NRC staff SE of NEI 94-01, Revision 2, states:

As noted above, Section 9.2.3, NEI TR 94-01, Revision 2, states, Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history. However, Section 9.1 states that the required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes. The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

The licensees statement indicates acknowledgement and acceptance of the NRC staff position that extensions of the Type A test interval beyond the upper-bound

performance-based limit of 15 years should be infrequent and that any request for an extension should demonstrate to the NRC staff that an unforeseen emergent condition exists.

Based on the information provided by the licensee, the NRC staff finds that the licensee has addressed and satisfied NRC Condition 5.

f. NRC Condition 6 For plants licensed under 10 CFR Part 52, applications requesting a permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past ILRT data.

Ginna Response to Limitation/Condition 6 Not applicable. GNPP was not licensed under 10 CFR 52.

Based on the information provided by the licensee, the NRC staff finds that the licensee has addressed and satisfied NRC Condition 6.

Based on the NRC staffs review of the LAR and its above technical evaluations, the staff finds that the licensee has adequately addressed the NRC conditions to demonstrate the acceptability of adopting NEI 94-01, Revision 2-A, as the 10 CFR Part 50, Appendix J, Option B, implementation documents. Therefore, the proposed change to TS 5.5.15 to replace RG 1.163 with references to NEI 94-01, Revision 2-A, is acceptable.

The exception in TS 5.5.15.a is no longer necessary, as that date has passed with the test having been completed. Therefore, the proposed deletion of the exception from the TSs is acceptable.

Therefore, the proposed changes to Ginna TS 5.5.15 are acceptable.

3.5 Risk Evaluation 3.5.1 Background Section 9.2.3.1, General Requirements for ILRT Interval Extensions beyond Ten Years, of NEI 94-01, Revision 3-A, states that plant-specific confirmatory analyses are required when extending the Type A ILRT interval beyond 10 years. Section 9.2.3.4, Plant-Specific Confirmatory Analyses, of NEI 94-01, Revision 3-A, states that the assessment should be performed using the approach and methodology described in EPRI TR-10182431, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals. The analysis is to be performed by the licensee and retained in the plant documentation and records as part of the basis for extending the ILRT interval.

In the SER dated June 25, 2008, the NRC staff found the methodology in NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, to be acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT interval to 15 years, provided that certain

conditions are satisfied. These conditions, set forth in Section 4.2 of the SER for EPRI TR-1009325, Revision 2, provide that:

1. The licensee submit documentation indicating that the technical adequacy of its Probabilistic Risk Assessment (PRA) is consistent with the requirements of Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, relevant to the ILRT extension application. Additional application-specific guidance on the technical adequacy of a PRA used to extend ILRT intervals is provided in the SER for EPRI TR-1009325, Revision 2.

2. The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small and consistent with the clarification provided in Section 3.2.4.62 of the SER for EPRI TR-1009325, Revision 2.

3. The methodology in EPRI TR-1009325, Revision 2, is acceptable provided the average leak rate for the pre-existing containment large leak accident case (i.e.,

accident case 3b) used by licensees is assigned a value of 100 times the maximum allowable leakage rate (La) instead of 35 La.

4. An LAR is required in instances where containment over-pressure is relied upon for emergency core cooling system (ECCS) performance.

3.5.2 Plant-Specific Risk Evaluation The licensee performed a risk impact assessment for extending the Type A containment ILRT interval to once in 15 years in Attachment 4 to the LAR.

In Section 3.3.1 of Attachment 1 to the LAR, the licensee stated that the plant-specific risk assessment follows the guidance in NEI 94-01, Revision 3-A; the methodology described in EPRI TR-1018243 (also identified as EPRI TR-1009325, Revision 2-A); and the NRC regulatory guidance outlined in RG 1.174.

The licensee addressed each of the four conditions for the use of EPRI TR-1009325, Revision 2, listed in Section 4.2 of the NRC SER. A summary of how each condition is met is provided in Sections 3.5.2.1 through 3.5.2.4 below.

3.5.2.1 Technical Adequacy of the PRA The first condition in Section 4.2 of the SER for EPRI TR-1009325, Revision 2, provides that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with RG 1.200, Revision 2, relevant to the ILRT extension application.

Consistent with the information provided in Regulatory Issue Summary 2007-06, Regulatory Guide 1.200 Implementation (ADAMS Accession No. ML070650428), the NRC staff uses Revision 2 of RG 1.200 (ADAMS Accession No. ML090410014) to assess technical adequacy of the PRA used to support risk-informed applications received after March 2010. In 2

Section 4.2 of the SER for EPRI TR-1009325, Revision 2, indicates that the clarification regarding small increases in risk is provided in Section 3.2.4.5; however, the clarification is actually provided in Section 3.2.4.6.

Section 3.2.4.1 of the SER for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, the NRC staff states that Capability Category (CC) I of the ASME PRA standard shall be applied as the standard for assessing PRA quality for ILRT extension applications, since approximate values of core damage frequency (CDF) and large early release frequency (LERF) and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies.

As discussed in Section 5.1.2 of Attachment 1 to the LAR, the Ginna risk assessment performed to support the ILRT application uses the current Ginna Level 1 and Level 2 internal events PRA model of record, which the licensee completed in March 2016. The internal events PRA also includes internal flooding hazards. In Appendix A1 of Attachment 1 to the LAR, the licensee describes the process used for controlling the model and for ensuring that the model reflects the as-built and as-operated plant configuration.

The licensee has a process for continued PRA maintenance and updates, including procedures for regularly scheduled and interim PRA model updates and for tracking issues identified as potentially affecting the PRA model. The licensee performed a review of the plant modifications and changes and concluded that the aggregate of open updating requirements evaluation (UREs) leads to the current PRA model being conservative. Exelon maintains a URE database to track all enhancements, corrections, and unincorporated plant changes. During the normal screening conducted as part of the plant change process, if a potential model update is identified, a new URE database item is created. Depending on the potential impact of the identified change, the requirements for incorporation will vary.

As part of the LAR, the licensee reviewed the open items in the URE database for Ginna, and an assessment of the impact on the results of the LAR was made. Exelon identified a few UREs that might affect the LERF containment modeling. Open URE 834 pertains to finding and observation (F&O) LEC10-01, which states that credit for scrubbing was not taken. Scrubbing may be applicable to three containment bypass conditions: (1) a steam generator tube rupture event with feedwater available, (2) internal flood scenarios with an interfacing system loss-of-coolant accident (LOCA) and the affected auxiliary building room flooded, or (3) sequences where the interfacing system LOCA (ISLOCA) break is in the residual heat removal pits. Since these are all Class 8 (steam generator tube rupture (SGTR) or ISLOCA) sequences, the licensee indicated, and the NRC staff agrees, that there would be no effect from the change in the ILRT program impacting the computed change in LERF (LERF), whereas a change in Class 3b sequences would impact LERF.

Open URE 837 pertains to LERF quantification, which is used in this LAR, and tracks F&O LE-C9a. The NRC staff notes that LERF at Ginna has significant margin to the RG 1.174 acceptance guidelines, so any model update from this URE is judged to be sufficiently small so as to not affect this ILRT extension analysis.

The licensee also identified that additional open UREs may also affect overall CDF and LERF quantification results, which are used to calculate change in risk metrics for the ILRT extension evaluation. After evaluating all open UREs for their effect on CDF and LERF, Exelon concluded that the aggregate of open UREs leads to the current PRA model being conservative. A conservative CDF alone would lead to conservative calculations for the ILRT extension. Since LERF is subtracted from CDF to calculate the risk increase due to the ILRT extension, if the LERF conservatism is greater than the CDF conservatism, the ILRT extension calculations would not be conservative. As the magnitude of CDF and LERF conservatisms are unknown, a sensitivity study was performed by the licensee where LERF is not subtracted from CDF when

calculating the change in risk for the ILRT extension. (This is described in Section 5.2.1 in Appendix B of Attachment 2 to the LAR). Sensitivity results are shown in Appendix B of to the LAR. The increase in the overall probability of LERF is less than 10-7.

Therefore, the LERF is considered very small.

The peer review process demonstrates the technical acceptability of the Ginna PRA models.

The purpose of the industry PRA peer review process is to provide a method for establishing the technical capability and adequacy of a PRA relative to expectations of knowledgeable practitioners using a set of guidance that establishes a set of minimum requirements. PRA peer reviews continue to be performed as PRAs are updated (and upgraded) to ensure the ability to support risk-informed applications. There have been three relevant peer reviews conducted on the current Ginna PRA model:

• The 2009 peer review for the PRA ASME model update identified 309 supporting requirements applicable to the Ginna PRA. Of these, 29 were not met, 2 met CC 1, 13 met CC 1/2, 31 met CC 2, 22 met CC 2/3, 14 met CC 3, and 198 fully met all capability requirements. There were 24 F&Os issued to address the identified gaps to compliance with the PRA standard. Subsequent to the peer review, 13 of the findings have been addressed and 11 are still open, pending the next model update. The open F&Os are listed in Table A-5 of Attachment 2 to the LAR, and the results show an insignificant impact to the CDF, LERF, and this ILRT extension.

The 2012 fire PRA peer review for the PRA ASME model update identified 183 supporting requirements to be reviewed for the Ginna PRA. All the findings that impact the fire PRA were closed prior to the initial National Fire Protection Association (NFPA) 805 submittal. NRC staff questions relating to the fire PRA were resolved in the NFPA-805 LAR and request for additional information responses.

• A peer review was conducted to assess actions taken to address existing finding-level F&Os. The June 2017 full power internal event review performed an independent assessment of finding-level F&Os from previous peer reviews. Finding-level F&Os that were reviewed and were determined to have been adequately addressed through this technical review are considered closed. These closed F&Os are no longer relevant to the current PRA model. The technical review team determined that 17 of the 23 finding-level F&Os were resolved. Four of the finding-level F&Os remain open. The remaining two finding-level F&Os were partially resolved but require further documentation (i.e., all technical aspects were resolved). The remaining gaps are documented in the URE database, and the results show an insignificant impact to the CDF, LERF, and this ILRT extension.

Based on the peer reviews, independent assessment of F&O resolutions, and the focused scope peer reviews, the licensee determined that the current Ginna internal events and fire PRA models mostly conform to CC 2 of ASME RASb-2009, ASME/ANS Standard for Probabilistic Risk Assessment of Nuclear Power Plant Applications, as endorsed by RG 1.200, Revision 2 (with the remaining few items conforming to CC I of ASME RA-Sb-2009).

As provided in the regulatory analysis in Section 2.0 above, CC I of the ASME PRA standard shall be applied as the standard for assessing PRA quality for ILRT extension applications, since approximate values of CDF and LERF and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies. Since the licensee demonstrated that the technical adequacy of its PRA is consistent with the requirements of

RG 1.200 relevant to the ILRT extension application, the NRC staff finds that using these models for this ILRT LAR meets the technical adequacy requirements.

In Section 3.2.4.2 of the SER for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, the NRC staff states, in part, that:

Although the emphasis of the quantitative evaluation is on the risk impact from internal events, the guidance in EPRI Report No. 1009325, Revision 2, Section 4.2.7, External Events, states that: Where possible, the analysis should include a quantitative assessment of the contribution of external events (e.g., fire and seismic) in the risk impact assessment for extended ILRT intervals. This section also states that: If the external event analysis is not of sufficient quality or detail to directly apply the methodology provided in this document [(i.e., EPRI Report No. 1009325, Revision 2)], the quality or detail will be increased or a suitable estimate of the risk impact from the external events should be performed.

This assessment can be taken from existing previously submitted and approved analyses or other alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval.

The Ginna individual plant examination of external events seismic risk analysis did not quantify a CDF impact. Thus, a new seismic core damage frequency (SCDF) calculation is summarized in Section 5.2.7 and detailed in Appendix B of Attachment 2 to the LAR. Ginna also submitted a seismic hazard screening report (ADAMS Accession No. ML14099A196) to the NRC in accordance with the requirements of Near-Term Task Force (NTTF) Recommendation 2.1:

Seismic (ADAMS Accession No. ML12053A340). The seismic hazard screening report confirmed the current Ginna seismic hazard (i.e., ground motion response spectrum) is bounded by the Ginna seismic capability (i.e., safe shutdown earthquake in the frequency range of 1 to 10 hertz (Hz)), except for a narrow band exceedance between 9 and 10 Hz. As a result, the NRC staff issued a staff assessment report (ADAMS Accession No. ML15153A026) and concluded a seismic risk evaluation (i.e., seismic PRA or seismic margins assessment) is not required for Ginna.

Furthermore, the NRC staff also concluded that for NTTF Recommendation 2.1: Seismic, a spent fuel pool assessment was not required, but a high frequency confirmation was required.

The licensee submitted a high frequency confirmation (ADAMS Accession No. ML15338A003) that was accepted by the NRC in a separate staff assessment (ADAMS Accession No. ML15364A544). Since these updated evaluations do not include a quantitative evaluation of seismic risk (and were not required to), another analysis is used to estimate CDF. To estimate SCDF, the plant level high-confidence-of-low-probability-of-failure seismic capacity is used to convolve the corresponding failure probabilities as a function of seismic hazard level with the seismic hazard curve. Further details of the SCDF calculation, including the seismic hazard input (shown in Table B-1) and seismic hazard intervals are used in the analysis with their representative peak ground acceleration and occurrence frequency (shown in Table B-2). The SCDF is estimated to be 3.86E-06 per year. Applying the internal event LERF/CDF ratio to the SCDF yields an estimated seismic LERF of 1.56E-7.

Based on its review of the above information, the NRC staff finds the licensees analysis of the impact of external events acceptable for the ILRT extension application since it complies with Section 3.2.4.2 of the SER for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2.

Furthermore, the licensee has evaluated its internal events PRA against the current PRA standard and Revision 2 of RG 1.200. The NRC staff finds that the licensee has addressed the relevant findings and gaps from the peer reviews and that they have no impact on the results of this application. Therefore, the NRC staff concludes that the internal events PRA model used by the licensee is of sufficient quality to support the evaluation of changes to ILRT frequencies.

Accordingly, the first condition is met.

3.5.2.2 Estimated Risk Increase The second condition provides that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small and consistent with the guidance in RG 1.174 and the clarification provided in Section 3.2.4.5 of the NRC SER for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive.

In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage points. Additionally, for plants that rely on containment overpressure for net positive suction head for ECCS injection, both CDF and LERF will be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.174. As discussed in Section 3.5.2.4 of this SE, Ginna does not rely on containment over-pressure for ECCS performance. Thus, the associated risk metrics include LERF, population dose, and CCFP.

The licensee-reported results from this ILRT extension risk assessment for Ginna are summarized in Table 6-1 of Attachment 4 to the LAR. Details of the risk assessment are provided in the Ginna PRA application notebook provided in Attachment 4 to the LAR. The reported risk impacts are based on a change in the Type A containment ILRT frequency from three tests in 10 years (the test frequency under 10 CFR Part 50, Appendix J, Option A) to one test in 15 years and also account for the risk from undetected containment leaks due to steel liner corrosion. The following conclusions can be drawn from the licensees analysis associated with extending the Type A ILRT frequency:

1. RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-06/year and increases in LERF less than 1.0E-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF.

The licensee estimated the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years to be 9.52E-8/year using the EPRI guidance. This value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Therefore, the estimated change in LERF is determined to be very small using the acceptance guidelines of RG 1.174.

The risk change resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change.

Considering the increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 3.96E-8/year, the risk increase is very small using the acceptance guidelines of RG 1.174. When external

event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated by Exelon as 3.91E-7/year using the EPRI guidance, and total LERF is 1.61E-6/year. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174. As discussed in Sections 5.1.3 and 5.2.7 of Attachment 2 to the LAR, the EPRI methodology used to estimate the increase in LERF and the models used to estimate total LERF are conservative. Therefore, the conservative methodology adds margin.

2. The effect resulting from changing the Type A test frequency to 1 in 15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.29 person-rem/year. NEI 94-01, Revision 3-A, states that a small population dose is defined as an increase of 1.0 person-rem per year, or 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. Thus, the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

3. The licensee estimated that the increase in CCFP due to the change in test frequency from 3 in 10 years to 1 in 15 years is 0.881 percent. This value is below 1.5 percent, which is the acceptance guidelines in Section 3.2.4.6 of the NRC SER for NEI 94-01, Revision 2. Thus, the increase in CCFP due to the change in test frequency is well within the guidelines and supportive of the proposed change.

Based on the risk assessment results, the NRC staff concludes that the increase in LERF is small and consistent with the acceptance guidelines of RG 1.174, and that the increase in the total population dose and the magnitude of the change in the CCFP for the proposed change are small. The defense-in-depth philosophy is maintained, as the independence of barriers will not be degraded as a result of the requested change, and the use of the three quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. Accordingly, the second condition is met.

3.5.2.3 Leak Rate for the Large Preexisting Containment Leak Rate Case The third condition provides that in order to make the methodology in EPRI TR-1009325, Revision 2, acceptable, the average leak rate for the preexisting containment large leak rate accident case (i.e., Accident Case 3b) used by licensees shall be 100 La instead of 35 La. As noted by the licensee in Section 4.0 of Attachment 4 to the LAR, the methodology in EPRI TR-1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the preexisting containment large leak rate accident case (Accident Case 3b), and this value has been used in the Ginna plant-specific risk assessment. Accordingly, the third condition is met.

3.5.2.4 Applicability if Containment Overpressure is Credited for ECCS Performance The fourth condition provides that in instances where containment overpressure is relied upon for ECCS performance, an LAR is required to be submitted. In Table 3.4.1-1 of Attachment 1 to

the LAR, the licensee stated that containment overpressure is not relied upon for ECCS performance at Ginna. Accordingly, the fourth condition is not applicable.

3.6 Technical Evaluation Summary The NRC staff finds that the licensee is satisfactorily monitoring and managing the Ginna containment and performing inspections to periodically examine and monitor aging degradation, thereby providing reasonable assurance that the containment structural and leaktight integrity will continue to be maintained. The licensee justified the proposed change to extend the performance-based Type A ILRT test interval by demonstrating adequate performance of the containment based on plant-specific Type A ILRT test program results, consistent with the guidance in NEI 94-01.

As noted, recent IWE examinations performed to ensure that the structural integrity of the ASME Class MC containment liner was maintained were performed during RFO G1R39 (fall 2015) and RFO G1R40 (spring 2017). IWL examinations were performed in 2010 and 2015 to ensure that the structural integrity of the ASME Class CC reinforced concrete was maintained. The licensee also demonstrated satisfactory containment inspection results consistent with the inservice inspection program requirements of ASME Section XI, Subsections IWE and IWL. In summary, the NRC staff finds that the licensee has an adequate CISI program in place to periodically examine, monitor, and manage structural deterioration and aging degradation of the Ginna containment pressure boundary and that it can perform its intended function as a leaktight barrier, consistent with the guidance contained in RG 1.163.

Therefore, the NRC staff concludes that the proposed change to Ginna TS 5.5.15 regarding the primary containment leakage rate testing program is acceptable.

In the SER dated June 25, 2008, the NRC staff found that the methodology in NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, is acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT interval to 15 years, provided certain conditions are satisfied. These conditions are set forth in Section 4.2 of the SER for EPRI TR-1009325, Revision 2. Based on the NRC staffs review of the LAR and the technical evaluations, the staff finds that the licensee has adequately addressed the applicable conditions to satisfy the key principles of risk-informed decisionmaking.

In summary, the NRC staff finds that the licensee has adequately implemented its primary containment leakage rate testing program for the integrated leak rate tests and the local leak rate tests. The results of the recent ILRTs and LLRTs demonstrate acceptable performance and that the structural and leaktight integrity of the containment structure is adequately managed and will continue to be periodically monitored and managed by the ILRTs and LLRTs.

The staff finds that the licensee has addressed the NRC conditions to demonstrate the acceptability of adopting NEI 94-01, Revision 2-A, and the limitations and conditions identified in the staffs SE incorporated in NEI 94-01, Revision 2-A, without undue risk to the public health and safety. Therefore, the staff concludes that the proposed change to implement the requested extension of the ILRT interval for Ginna TS 5.5.15 is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on November 25, 2019. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on April 9, 2019 (84 FR 14148). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: R. Pettis J. Dozier N. Karipineni J. Bettle Date: December 23, 2019

ML19325D824 *by memorandum **by e-mail OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DSS/SCPB/BC*

NAME VSreenivas LRonewicz BWittick DATE 11/21/2019 11/22/2019 11/08/2019 OFFICE NRR/DEX/ESEB/BC* NRR/DRA/APLB/BC* NRR/DSS/STSB/BC**

NAME BWittick KHsueh VCusumano DATE 07/11/2019 09/27/2019 12/03/2019 OFFICE OGC - NLO** NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME GWachutka JDanna (JKim for) VSreenivas DATE 12/17/2019 12/23/2019 12/23/2019