ML20213C726

Issuance of Amendment No. 300 to Revise the Containment Leakage Rate Testing Program

ML20213C726
Person / Time
Site:	Davis Besse
Issue date:	08/24/2020
From:	Blake Purnell Plant Licensing Branch III
To:	Tony Brown Energy Harbor Nuclear Corp
Purnell B
References
EPID L-2019-LLA-0186
Download: ML20213C726 (29)

Text

August 24, 2020 Mr. Terry J. Brown Site Vice President Energy Harbor Nuclear Corp.

Mail Stop P-DB-3080 5501 North State Route 2 Oak Harbor, OH 43449-9760

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 - ISSUANCE OF AMENDMENT NO. 300 TO REVISE THE CONTAINMENT LEAKAGE RATE TESTING PROGRAM (EPID L-2019-LLA-0186)

Dear Mr. Brown:

The U.S. Nuclear Regulatory Commission (NRC or Commission) has issued the enclosed Amendment No. 300 to Renewed Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse). The amendment is in response to the FirstEnergy Nuclear Operating Company application dated August 26, 2019 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML19241A267), as supplemented by letter dated February 3, 2020 (ADAMS Accession No. ML20034D804). Effective February 27, 2020, the facility operating license for Davis-Besse was transferred from FirstEnergy Nuclear Generation, LLC (owner) and FirstEnergy Nuclear Operating Company (operator) to Energy Harbor Nuclear Generation LLC (owner) and Energy Harbor Nuclear Corp. (operator) (ADAMS Accession No. ML20030A440). Upon completion of this license transfer, Energy Harbor Nuclear Corp. assumed the responsibility for all licensing actions under NRC review at the time of the transfer and requested that the NRC continue its review of these actions (ADAMS Accession No. ML20054B733).

The amendment revises the Davis-Besse Technical Specification 5.5.15, Containment Leakage Rate Testing Program, to allow the Type A test interval to be extended to 15 years based on acceptable performance history.

A copy of the NRC staffs Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Blake Purnell, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-346

Enclosures:

1. Amendment No. 300 to NPF-3

2. Safety Evaluation cc: Listserv

ENERGY HARBOR NUCLEAR CORP.

AND ENERGY HARBOR NUCLEAR GENERATION LLC DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE DOCKET NO. 50-346 Amendment No. 300 Renewed License No. NPF-3

1.

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

A.

The application for amendment filed by FirstEnergy Nuclear Operating Company dated August 26, 2019, as supplemented by letter dated February 3, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-3 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300, are hereby incorporated in the renewed license. Energy Harbor Nuclear Corp. shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications and Renewed Facility Operating License Date of Issuance: August 24, 2020 Nancy L.

Salgado Digitally signed by Nancy L. Salgado Date: 2020.08.24 11:00:57 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 300 RENEWED FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 Replace the following pages of the Renewed Facility Operating License and Appendix A, Technical Specifications (TSs), with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert License NPF-3 License NPF-3 L-5 L-5 TSs TSs 5.5-12 5.5-12

L-5 Renewed License No. NPF-3 Amendment No. 300 2.C.

This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: 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 to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level Energy Harbor Nuclear Corp. is authorized to operate the facility at steady state reactor core power levels not in excess of 2817 megawatts (thermal). Prior to attaining the power level, Toledo Edison Company shall comply with the conditions identified in Paragraph (3) (o) below and complete the preoperational tests, startup tests and other items identified in Attachment 2 to this license in the sequence specified. Attachment 2 is an integral part of this renewed license.

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 300, are hereby incorporated in the renewed license.

Energy Harbor Nuclear Corp. shall operate the facility in accordance with the Technical Specifications.

(3)

Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission:

(a)

Energy Harbor Nuclear Corp. shall not operate the reactor in operational Modes 1 and 2 with less than three reactor coolant pumps in operation.

(b)

Deleted per Amendment 6 (c)

Deleted per Amendment 5

Programs and Manuals 5.5 Davis-Besse 5.5-12 Amendment 300 5.5 Programs and Manuals 5.5.14 Safety Function Determination Program (continued)

3.

Provisions to ensure that an inoperable supported systems Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4.

Other appropriate limitations and remedial or compensatory actions.

b.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable; and

1.

A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

2.

A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

3.

A required system redundant to the support system(s) for the supported systems described in Specifications 5.5.14.b.1 and 5.5.14.b.2 above is also inoperable.

c.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.15 Containment Leakage Rate Testing Program

a.

A program shall establish 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 Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1.

A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 300 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-3 ENERGY HARBOR NUCLEAR CORP.

ENERGY HARBOR NUCLEAR GENERATION LLC DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346

1.0 INTRODUCTION

By application dated August 26, 2019 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML19241A267), as supplemented by letter dated February 3, 2020 (ADAMS Accession No. ML20034D804), FirstEnergy Nuclear Operating Company (FENOC) submitted a license amendment request (LAR) for Davis-Besse Nuclear Power Station, Unit No. 1 (Davis-Besse). Effective February 27, 2020, the facility operating license for Davis-Besse was transferred from FirstEnergy Nuclear Generation, LLC (owner) and FENOC (operator) to Energy Harbor Nuclear Generation LLC (owner) and Energy Harbor Nuclear Corp. (operator) (ADAMS Accession No. ML20030A440). Upon completion of this license transfer, Energy Harbor Nuclear Corp. assumed the responsibility for all licensing actions under U.S. Nuclear Regulatory Commission (NRC or Commission) review at the time of the transfer and requested that the NRC continue its review of these actions (ADAMS Accession No. ML20054B733).

The proposed amendment would revise the Davis-Besse Technical Specification (TS) 5.5.15, Containment Leakage Rate Testing Program, by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program (ADAMS Accession No. ML003740058), with a reference to the Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR [Title 10 of the Code of Federal Regulations] Part 50, Appendix J (ADAMS Accession No. ML12221A202), and the limitations and conditions specified in NEI 94-01, Revision 2-A (ADAMS Accession No. ML100620847). The proposed amendment would allow the containment Type A test interval to be extended to 15 years based on acceptable performance history, as defined in NEI 94-01, Revision 3-A.

The February 3, 2020, supplement was in response to an NRC staff request for additional information issued on January 21, 2020 (ADAMS Accession No. ML20021A316). The supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on October 22, 2019 (84 FR 56481).

2.0 REGULATORY EVALUATION

2.1 Background

The Davis-Besse containment consists of three basic structures: a steel containment vessel, a reinforced concrete shield building, and the internal structures. The containment vessel is a cylindrical steel pressure vessel with hemispherical dome and ellipsoidal bottom. It is completely enclosed by a reinforced concrete shield building having a cylindrical shape with a shallow dome roof. An annular space is provided between the wall of the containment vessel and the shield building, and clearance is also provided between the containment vessel and the dome of the shield building. The containment vessel and shield building are supported on a concrete foundation founded on a firm rock structure. Except for the concrete under the containment vessel, there are no structural ties between the containment vessel and the shield building above the foundation slab. Above this, there is unlimited freedom of differential movement between the containment vessel and the shield building. The containment internal structures are constructed of reinforced concrete and structural steel. These structures are isolated from the containment vessel by steel grating panels with sliding supports which allows free differential movement between the internal structures and the vessel. The internal structures are supported by the massive concrete fill within the containment vessel bottom head.

The containment vessel inside diameter is 130 feet and the net free volume is approximately 2,834,000 cubic feet. The cylindrical shell and bottom head thickness, exclusive of reinforced areas, is 1-1/2 inches with a dome thickness of 13/16 inches. Access to the containment is provided by an equipment hatch, a personnel air lock, and an emergency air lock. Electrical and mechanical penetrations are provided for services to the containment.

The shield building is a reinforced concrete structure of right cylinder configuration with a shallow dome roof. An annular space is provided between the steel containment vessel and the interior face of the concrete shield building of approximately 4.5 feet to permit construction operations and periodic visual inspection of the steel containment vessel. The volume contained within this annulus is approximately 678,700 cubic feet. The shield building has a height of 279.5 feet measured from the top of the foundation ring to the top of the dome. The thicknesses of the wall and the dome are approximately 2.5 feet and 2 feet, respectively.

The shield building completely encloses the containment vessel, the personnel access openings, the equipment hatch, and that portion of penetrations that are associated with primary containment. The design of the shield building provides for biological shielding, controlled release of the annulus atmosphere under accident condition, and environmental protection of the containment vessel.

2.2 Regulatory Requirements Section (o) of 10 CFR 50.54, Conditions of licenses, requires that primary reactor containments for water-cooled power reactors be subject to the requirements in Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, to 10 CFR Part 50. Appendix J contains two options, Option A - Prescriptive Requirements and Option B - Performance-Based Requirements, either of which can be used to meet the Appendix J requirements. The testing requirements in Appendix J ensure that: (1) leakage through containment or systems and components penetrating containment does not exceed allowable leakage rates specified in the TSs and (2) integrity of the containment structure is maintained during the service life of the containment.

Option B specifies performance-based requirements and criteria for preoperational and subsequent leakage rate testing. These requirements are met by:

1. Type A tests to measure the containment system overall integrated leakage rate,

2. Type B pneumatic tests to detect and measure local leakage rates across pressure retaining leakage-limiting boundaries such as penetrations, and

3. Type C pneumatic tests to measure containment isolation valve leakage rates.

The Type A test is also referred to as an integrated leak rate test (ILRT), and the Type B and C tests are referred to as local leak rate tests (LLRTs). Davis-Besse implemented Option B of 10 CFR Part 50, Appendix J, for the ILRT and LLRTs through Amendment No. 205 (ADAMS Accession No. ML021210135) and Amendment No. 240 (ADAMS Accession No. ML003698061), respectively, to its license.

Section V.B.3 of 10 CFR Part 50, Appendix J, Option B, requires the TSs to include, by general reference, the RG or other implementation document used by the licensee to develop a performance-based leakage-testing program. The submittal for TS revisions must also contain justification, including supporting analyses, if the licensee deviates from methods approved by the NRC and endorsed in RG 1.163.

After the containment system has been completed and is ready for operation, Type A tests are conducted at periodic intervals based on the performance history of the overall containment system to measure the overall integrated leakage rate. The leakage rate test results must not exceed the maximum allowable leakage rate (La) at the calculated peak containment internal pressure for the design-basis loss-of-coolant accident (Pa) with margin, as specified in the TSs.

Option B also requires that a general visual inspection for structural deterioration of the accessible interior and exterior surfaces of the containment system, which may affect the containment leaktight integrity, be conducted prior to each Type A test and at a periodic interval between tests based on the performance of the containment system.

Type B and Type C tests are performed based on the safety significance and performance history of each boundary and isolation valve to ensure integrity of the overall containment system as a barrier to fission product release.

Section 50.55a, Codes and standards, of 10 CFR contains the containment inservice inspection (ISI) requirements, which, in conjunction with the requirements of 10 CFR Part 50, Appendix J, ensure the continued leaktightness and structural integrity of the containment during its service life. Paragraph 50.55a(g)(4)(ii) of 10 CFR requires, in part, that inservice examination of components and system pressure tests conducted during successive 10-year ISI intervals (i.e., after the initial 10-year interval) must comply with the latest edition and addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) (or the optional ASME Code Cases) incorporated by reference in 10 CFR 50.55a(a) 12 months before the start of the 10-year interval, subject to the conditions listed in 10 CFR 50.55a(b). The 2007 Edition with the 2008 Addenda of Section XI to the ASME Code is applicable to the current 10-year ISI interval at Davis-Besse, which ends on September 20, 2022.

Paragraph (a)(1) of 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, states, 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, are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience.

The regulations in 10 CFR 50.36, Technical specifications, establish the regulatory requirements related to the content of TSs. In accordance with 10 CFR 50.36(c)(5), the TSs must include administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

2.3 Regulatory Guidance NEI 94-01, Revision 0 (ADAMS Accession No. ML11327A025), provides methods for complying with Option B of 10 CFR Part 50, Appendix J, and allows for the extension of the performance-based Type A test interval up to 10 years, based upon two consecutive successful tests. NEI 94-01, Revision 0, was endorsed by the NRC in RG 1.163 with some conditions.

NEI 94-01, Revision 2 (ADAMS Accession No. ML072970206), incorporated the NRC conditions in RG 1.163 and added provisions for extending Type A test intervals up to 15 years.

This revision of NEI 94-01 was supported by Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated August 2007 (ADAMS Accession No. ML072970208). The EPRI report provides a generic assessment of the risks associated with permanently extending the ILRT interval to 15 years, and it provides a risk-informed methodology to be used to confirm the risk impact of the ILRT extension on a plant-specific basis. Probabilistic risk assessment (PRA) methods are used in combination with ILRT performance data and other considerations to justify the extension of the ILRT interval.

The NRC staffs review of both NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, is described in an NRC safety evaluation (SE) dated June 25, 2008 (ADAMS Accession No. ML081140105). The SE 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. In addition, the SE states that the methodology described in EPRI Report No. 1009325, Revision 2, satisfies the key principles of risk-informed decision-making described in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ADAMS Accession No. ML17317A256), and RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications (ADAMS Accession No. ML100910008). The NRC staff concluded that NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, are acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to the conditions listed in Section 4.0 of the SE. The SE was incorporated into NEI 94-01, Revision 2, and subsequently issued as NEI 94-01, Revision 2-A, on November 19, 2008. In October 2008, EPRI Report No. 1009325, Revision 2-A, was published, which incorporated the applicable changes identified in the SE.1 NEI 94-01, Revision 3 (ADAMS Accession No. ML112920567), added guidance for extending Type C LLRT intervals beyond 60 months. In an SE dated June 8, 2012 (ADAMS Accession No. ML121030286), the NRC staff concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of Appendix J, and is acceptable for reference by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions related to Type C testing. The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012. Davis-Besse was approved to use NEI 94-01, Revision 3-A, for Type C testing by Amendment No. 288 issued on October 9, 2015 (ADAMS Accession No. ML15239B293).

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities, (ADAMS Accession No. ML090410014),

provides guidance for determining the technical adequacy of the base PRA used in a risk-informed regulatory activity, and endorses standards and industry peer-review guidance.

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

2.4 Licensees Proposed Changes The requirements for the Davis-Besse containment leakage rate testing program are specified in TS 5.5.15. Currently, Davis-Besse TS 5.5.15.a states:

A program shall establish 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. For Type C tests, this program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 2012.

For Type A and Type B tests, 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 exceptions:

1. A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

2. The fuel transfer tube blind flanges (containment penetrations 23 and 24) will not be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However, as-found testing will not be required.

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

The February 3, 2020, letter from FENOC states that the proposed change will revise TS 5.5.15.a to state:

A program shall establish 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 Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1. A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

2. The fuel transfer tube blind flanges (containment penetrations 23 and 24) will not be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However, as-found testing will not be required.

With the proposed change, Davis-Besse can extend the current Type A test interval from 10 years to 15 years in accordance with the guidelines in NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A. For Type A and Type B tests, the proposed change would also result in a more conservative test interval extension of 9 months for emergent conditions. The LAR did not request any changes related to Type C tests or the exceptions listed in TS 5.5.15.a.

3.0 TECHNICAL EVALUATION

3.1 Type A Integrated Leak Rate Test Section 9.2.3, Extended Test Intervals, of NEI 94-01, Revision 3-A, provides criteria for extending the ILRT interval. Section 9.2.3 states, in part:

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. Acceptable performance history is defined as successful completion of two consecutive periodic Type A tests where the calculated performance leakage rate was less than 1.0 La. Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

The performance leakage rate is calculated as the sum of the Type A upper confidence limit and as-left minimum pathway 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 performing the Type A test. Section 9.2.3 of NEI 94-01, Revision 3-A, provides additional details regarding the methodology for calculating the performance leakage rate. However, Section 9.2.3 also allows the performance history to be established using previously acceptable Type A test results that used a different methodology.

The Davis-Besse TS 5.5.15 states that the peak calculated containment internal pressure for the design-basis loss of coolant accident (Pa) is 38 pounds per square inch, gauge (psig).

TS 5.5.15 also states that the maximum allowable containment leakage rate (i.e., La) at Pa is 0.5 percent of the containment air weight per day. TS 5.5.15 requires the as-left leakage rate to be less than 0.375 percent of containment air weight per day (0.75 La) during the first startup following completion of Type A testing.

LAR Section 3.3.5 provided the results from the last two Type A tests at Davis-Besse, which are summarized in Table 1 below. The results show substantial margin has been maintained relative to the TS 5.5.15 acceptance criterion for the last two Type A tests. These results demonstrate acceptable performance history as defined in Section 9.2.3 of NEI 94-01, Revision 3-A. Therefore, the Type A test interval at Davis-Besse could be extended to 15 years under the NEI 94-01, Revision 3-A, guidance. In addition, no adverse trend is apparent that would suggest the performance criterion might be exceeded if the Type A test interval is extended to 15 years.

Table 1: Summary of Recent Type A Test Results at Davis-Besse (All values are expressed in percent of the containment air weight per day)

Test Date April 2003 November 2011 As-Found Leakage 0.16714 0.0722 Upper 95% Confidence Level 0.1627 0.0675 Level Corrections 0.0 0.0 As-Left Minimum Pathway Penalty for Isolation Pathways (Leakage Improvements) 0.00444 0.0031 (0.0016)

TS 5.5.15 Acceptance Criterion (0.75La) 0.375 0.375 Adjusted As-Left Leak Rate 0.16714 0.0706 3.2 Type B and Type C Leak Rate Testing Program Industry experience has shown that most ILRT failures result from leakage that is detectable by Type B and C testing. Section 10.2 of NEI 94-01, Revision 3-A, states that the acceptance criteria for the combined leakage rate for all penetrations subject to Type B or Type C testing shall be defined in accordance with American National Standards Institute (ANSI) and American Nuclear Society (ANS) joint standard ANSI/ANS 56.8-2002, Containment System Leakage Testing Requirements. Specifically, the combined as-found minimum pathway leakage rate of all Type B and C tests shall be less than 0.60 La when containment operability is required. The combined as-found leakage criterion ensures that penetration leakage is kept to acceptable levels throughout each operating cycle. In addition, if Type B or C testing occurred during an outage, the combined as-left maximum pathway leakage rate for all penetrations subject to Type B or C tests shall be less than 0.60 La before entering a mode where containment operability is required. The criterion for the as-found and as-left leakage rates ensures that margin is available to accommodate potential increases in leakage between outages where leakage testing is performed.

In accordance with Davis-Besse TS 5.5.15, the combined as-left maximum pathway leakage rate for all Type B and C tests must be less than 0.60 La during the first unit startup following testing in accordance with the containment leakage rate testing program. LAR Section 3.5.5 provided a summary of the combined Type B and C test results for both the as-found and as-left tests performed during the seven refueling outages between 2008 and 2018. The as-found minimum pathway leakage rates were on average 5.23 percent of 0.60 La with a high of 10.01 percent of 0.60 La. The as-left maximum pathway leakage rates were on average 10.66 percent of 0.60 La with a high of 22.69 percent of 0.60 La. The NRC staff determined that the results for the as-found leakage rates demonstrate a history of adequate maintenance since the results are significantly less than the acceptance criterion of 0.6 La in NEI 94-01, Revision 3-A. In addition, the results for the as-left leakage rates are significantly less than the acceptance criterion of 0.6 La in TS 5.5.15 and in NEI 94-01, Revision 3-A.

LAR Section 3.5.6 states that there are 33 components at Davis-Besse subject to Type B testing. Of the 33 components, 18 mechanical penetration components are eligible for an extended test frequency and are tested on an extended frequency. Electrical penetrations 101 and 102 are tested on a refueling outage frequency and are not on extended test intervals due to their inability to establish good performance. Additionally, Type B testing is performed on penetrations 101 and 102 to support the nitrogen check valve reverse-flow testing that is required every outage. The remaining 13 mechanical penetration components are not eligible for testing on an extended test frequency.

LAR Section 3.5.6 states that, except for three valves, the components at Davis-Besse subject to Type C testing that are eligible for extended intervals, following implementation of the 75-month Type C test intervals in 2015, are on extended intervals. The three valves not currently on extended intervals failed the as-found LLRTs during the 2016 or 2018 refueling outages. The three valves were replaced during the 2018 refueling outage and returned to acceptable leakage. Based on the information provided in LAR Section 3.5.6, the NRC staff finds that the corrective actions taken regarding the three valves are consistent with Section 10.2.3 of NEI 94-01, Revision 3-A.

Based on the review of LAR Sections 3.5.5 and 3.5.6, the NRC staff concludes that Davis-Besse is effectively implementing its performance-based containment leakage rate testing program for Type B and C testing. The combined Type B and C test results for both the as-found and as-left tests performed during the seven refueling outages between 2008 and 2018 are substantially below the acceptance criteria. In addition, appropriate corrective actions have been taken to address poor performing valves and penetrations. This conclusion supports extending the Type A test interval to 15 years because the LLRT program at Davis-Besse will provide continuing assurance that the most likely sources of leakage will be identified and repaired.

3.3 Containment Inspection and Testing Programs LAR Section 3.5 provides evaluations of other non-risk considerations related to the proposed amendment. This includes the inspection and testing programs that ensure the containment structure remains capable of meeting the design functions and identification of degraded conditions which may affect the containment capability.

3.3.1 Containment Inservice Inspection Program The containment ISI program at Davis-Besse monitors, in part, the integrity of the containment vessel. As required by 10 CFR 50.55a, this program must comply with Subsection IWE, Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants, of the ASME Code,Section XI (2007 Edition with the 2008 Addenda), for the current 10-year ISI interval at Davis-Besse. Subsection IWL, Requirements for Class CC Concrete Components of Light-Water Cooled Plants, of the ASME Code,Section XI, is not applicable to Davis-Besse. In addition, for the current ISI interval, Davis-Besse must comply with the conditions on metal containment examinations in paragraphs (A)(2), (B), and (J) of 10 CFR 50.55a(b)(2)(ix). Paragraph (A)(2) specifies conditions on the evaluation of inaccessible areas of the containment structure. Paragraph (B) specifies conditions on the performance of remote visual examinations of the containment structure. Paragraph (J) specifies requirements for containment leakage rate testing following major containment modification, repair, or replacement activities.

LAR Section 3.5.2 provides a summary of the containment examinations required by Subsection IWE of the ASME Code,Section XI, for the current ISI interval at Davis-Besse. As required by paragraph IWE-2411 of the ASME Code,Section XI, the 10-year ISI interval is divided into three inspection periods. Subsection IWE specifies three examination categories:

Category E-A (containment surfaces), Category E-C (containment surfaces requiring augmented examination), and Category E-G (pressure retaining bolting). For each category, the LAR describes the parts examined, examination method, extent of examination, examination percentage, total population of parts to be examined, total required examinations, and the examinations for each inspection period. Based on the information provided in LAR Section 3.5.2, the NRC staff finds there is reasonable assurance that the Davis-Besse containment ISI program complies with Subsection IWE of the ASME Code,Section XI, as required by 10 CFR 50.55a.

3.3.2 Nuclear Safety-Related Protective Coatings Program LAR Section 3.5.1 states, in part, that:

Protective coating material has been applied to carbon steel and concrete surfaces within the containment. The function of the protective coating material is to resist exposure to conditions including ionizing radiation, high temperature, and impingement from sprays, that result from normal operating and design basis accident conditions.

LAR Section 3.5.1 states that Davis-Besse is committed to RG 1.54 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants, dated June 1973 (Accession No. ML003740187), with clarifications.2 RG 1.54 endorses the ANSI Standard N101.4-1972, Quality Assurance for Protective Coatings Applied to Nuclear Facilities, subject to specific conditions. LAR Section 3.5.1 states that, in lieu of the full requirements of RG 1.54 and ANSI N101.4, Davis-Besse shall impose the following requirements:

a. The quality assurance requirements of Section 3 of ANSI N101.4 applicable to the coating manufacturer shall be imposed on the coating manufacturer through the procurement process.

b. Coating application procedures shall be developed based on the manufacturer's recommendations for application of the selected coating systems.

2 The commitment to RG 1.54, with clarifications, is described in Section 6.1.1, Protective Coating Systems (Paints) - Organic Materials (ADAMS Accession No. ML18283A911), of the Davis-Besse Updated Final Safety Analysis Report.

c. Coating applicators shall be qualified to demonstrate their ability to satisfactorily apply the coatings in accordance with the manufacturers recommendations.

d. Quality control personnel shall perform inspections to verify conformance of the coating application procedure. Section 6 of ANSI N101.4 shall be used as guidelines in the establishment of the inspection program.

e. Quality control personnel shall be qualified to the requirement of Regulatory Guide 1.58 (Revision 1) [Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel (ADAMS Accession No. ML12216A006)].

f.

Documentation demonstrating conformance to the above requirements shall be maintained.

LAR Section 3.5.1 states that non-design-basis-accident qualified coating materials have been applied to structures and components within the containment. These coating materials have been quantified and are tracked, and the documented quantity of these materials must remain below the limit of coating material debris identified by the emergency core cooling system (ECCS) emergency sump debris analysis. Coating condition assessment inspections are performed each refueling outage to identify and correct degraded coating materials. The LAR also states that: The Nuclear Safety-Related Protective Coatings Program requires that an inventory of unqualified coatings within containment be maintained, and that degraded coating conditions be documented using the corrective action process. LAR Table 3.5.1-1 lists the current amount of degraded and unqualified coatings at Davis-Besse and their associated limits.

Based on the information provided in LAR Section 3.5.1, the NRC staff has reasonable assurance that the Davis-Besse nuclear safety-related protective coating program is being properly implemented.

3.3.3 Conclusion Regarding Containment Inspection and Testing Programs Based on its review of the information provided in LAR Sections 3.5.1 and 3.5.2, the NRC staff finds that the containment ISI program and the nuclear safety-related protective coating program at Davis-Besse are being properly implemented. Therefore, the NRC staff has reasonable assurance that the structural integrity of the primary containment at Davis-Besse will be maintained if the ILRT interval is extended to 15 years.

3.4 Conditions on NEI 94-01, Revision 2 The Davis-Besse containment is subject to the requirements in 10 CFR Part 50, Appendix J, Option B, which allows the Type A, B, and C test intervals to be determined using a performance-based approach. Currently, the Type A and B testing at Davis-Besse is implemented in accordance with RG 1.163 and the Type C testing is implemented in accordance with NEI 94-01, Revision 3-A. The proposed change would revise TS 5.5.12 to require Type A and B testing to be implemented in accordance with NEI 94-01, Revision 3-A, along with the conditions and limitations of NEI 94-01, Revision 2-A, to govern the test frequencies and the grace periods for the containment leakage rate tests.

RG 1.163 endorses NEI 94-01, Revision 0, with some conditions. NEI 94-01, Revision 2, incorporated the NRC conditions in RG 1.63 and added provisions for extending Type A test intervals up to 15 years. In an SE 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 revise their containment leakage rate testing program, subject to the six conditions listed in Section 4.1 of the SE. The SE with these conditions is incorporated into NEI 94-01, Revision 2-A. The NRC staff evaluated the information in the LAR, as supplemented, to determine whether the NRC conditions on the use of NEI 94-01, Revision 2, have been adequately addressed.

3.4.1 Condition 1 on NEI 94-01, Revision 2 Condition 1 on NEI 94-01, Revision 2, states that the definition of the performance leakage rate in NEI 94-01, Revision 2, should be used when calculating the Type A leakage rate in lieu of the definition in ANSI/ANS 56.8-2002.

LAR Section 3.8.1 states that Davis-Besse will use the definition of the performance leakage rate in Section 5.0 of NEI 94-01, Revision 3-A. The definition of the performance leakage rate in Revision 2, Revision 2-A, and Revision 3-A of NEI 94-01 are identical. Therefore, the NRC staff concludes that Condition 1 on NEI 94-01, Revision 2, has been adequately addressed.

3.4.2 Condition 2 on NEI 94-01, Revision 2 Condition 2 stipulates that the licensee submit a schedule of containment inspections to be performed prior to and between Type A tests.

Section 9.2.1, Pretest Inspection and Test Methodology, of NEI 94-01, Revision 2, 2-A, and 3-A, states, in part:

Prior to initiating a Type A test, a visual examination shall be conducted of accessible interior and exterior surfaces of the containment system for structural problems that may affect either the containment structure leakage integrity or the performance of the Type A test. This inspection should be a general visual inspection of accessible interior and exterior surfaces of the primary containment and components. 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.

Section 9.2.3.2, Supplemental Inspection Requirements, of NEI 94-01, Revision 2, 2-A, and 3-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 discussed in Section 3.3.1 of this SE, LAR Section 3.5.2 provides a summary of the containment examinations required by Subsection IWE of the ASME Code,Section XI, for the current ISI interval at Davis-Besse. Subsection IWL of the ASME Code,Section XI, is not applicable to Davis-Besse. LAR Section 3.5.3 states, in part:

Inspections of the interior and exterior steel containment vessel surfaces are performed. The visual examination of the steel containment vessel surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified in Table 3.5.2-3, E-A, Containment Surfaces. This will require the performance of a minimum of four (4) visual inspections in accordance with ASME Code,Section XI, Subsection IWE surpassing the requirements of NEI 94-01, Revision 3-A Section 9.2.3.2, Supplemental Inspection Requirements, of three (3) inspections over a 15-year interval.

An additional visual inspection of the steel containment vessel is conducted prior to the conduct of each Type A test.

The NRC staffs review of LAR Sections 3.5.2 and 3.5.3 confirmed there is reasonable assurance that the pretest and supplemental inspection requirements for Type A tests in Sections 9.2.1 and 9.2.3.2 of NEI 94-01, Revision 3-A, will be satisfied for Davis-Besse.

Therefore, the NRC staff concludes that Condition 2 on NEI 94-01, Revision 2, has been adequately addressed.

3.4.3 Condition 3 on NEI 94-01, Revision 2 Condition 3 stipulates that the licensee address the areas of the containment structure potentially subjected to degradation.

As discussed in Section 3.3.1 of this SE, LAR Section 3.5.2 provides a summary of the containment examinations required for the current ISI interval at Davis-Besse. Paragraph (A)(2) of 10 CFR 50.55a(b)(2)(ix) specifies requirements for the evaluation of inaccessible areas of containment, which includes evaluation of degradation. The LAR indicates that all accessible surface areas and moisture barriers for the containment are examined each inspection period for the current interval.

LAR Section 3.5.2 provided the following information relative to containment surfaces requiring augmented examinations:

A condition report identified that a gap had formed at two areas between the containment vessel and the concrete ledge on the inside of containment at the 565-foot elevation. It was determined that these areas should be considered surface areas requiring augmented examination as required by [paragraph]

IWE-1240 [of the ASME Code, Section XI].

Another condition report documented the occurrence of general rust and corrosion noted on the sleeves of the service water and component cooling water piping penetrations into containment. From 2002 through 2014, these components were examined under an owner elected category. These penetrations have been placed in the E-C examination category as they meet the criteria noted by [subparagraph] IWE-1241(a) [of the ASME Code, Section XI].

These locations shall have a VT-1 visual examination performed once per Period.

LAR Section 3.6.3 discusses corrosion of the Davis-Besse containment vessel that was identified in 2002 during the cycle 13 refueling outage. During the outage, the corrosion was investigated using ISI methods to identify material loss due to corrosion. The LAR states that ultrasonic measurements verified that the minimum recorded vessel wall thickness was greater than the minimum required wall thickness. The containment vessel area behind the interior concrete structure was designated as an area susceptible to corrosion and the augmented examination requirements of Section IWE of the ASME Code,Section XI, were imposed. The augmented inspection areas received ultrasonic thickness examinations every 3-1/3 years until three successive examinations showed no evidence of ongoing corrosion. Ultrasonic test readings taken in 2008 showed that the containment vessel wall thicknesses in these areas remained essentially unchanged since 2002. The augmented inspections were performed in 2008, 2012, and 2014 and the results were found acceptable. The LAR states that no further corrective actions are required.

LAR Section 3.5.4 provides the results of containment visual examination of surfaces performed during the 2018 refueling outage. In several areas of the containment vessel interior, the examinations identified:

paint scrapes with no corrosion noted; areas of paint flaking or peeling with no degradation of the liner surface detected; or areas with nicks, chips, and abrasions on painted liner surfaces but with no significant corrosion detected.

Based on the review of LAR Sections 3.5.2, 3.6.3, and 3.5.4, the NRC staff finds that there is no evidence, to date, of unacceptable degradation of the Davis-Besse containment vessel.

Therefore, the NRC staff finds that Condition 3 on NEI 94-01, Revision 2, has been adequately addressed.

3.4.4 Condition 4 on NEI 94-01, Revision 2 Condition 4 stipulates that the licensee address any tests and inspections performed following major modifications to the containment structure, as applicable.

LAR Section 3.2 describes the containment tests and inspections performed following three major modificationsreplacement of the reactor vessel head in 2003 and 2011 and replacement of the steam generators in 2014that involved cutting a large hole in the Davis-Besse containment. The LAR states that a Type A test was performed in April 2003 and November 2011 after the containment vessel was restored following the reactor vessel head replacements. The results of these tests (see Table 1 of this SE) demonstrate that the containment leakage rate limits in Davis-Besse TS 5.5.15 were met.

LAR Section 3.2 states that following replacement of the steam generators in 2014 the containment vessel was restored to its original design requirements. Subsequently, a post-repair test of containment structural and leaktight integrity was performed by an alternative containment leakage test method approved by the NRC by letter dated May 8, 2013 (ADAMS Accession No. ML13121A404). The LAR states that the result of this alternative test met the acceptance criterion of zero detectable leakage.

Based on the review of LAR Section 3.2, the NRC staff finds that FENOC performed the required containment tests and achieved acceptable results following major modifications of the Davis-Besse containment vessel. Therefore, the NRC staff finds that Condition 4 on NEI 94-01, Revision 2, has been adequately addressed.

3.4.5 Condition 5 on NEI 94-01, Revision 2 Condition 5 on NEI 94-01, Revision 2, stipulates that the normal Type A test interval should be less than 15 years. If a licensee wants to use the provision of Section 9.1 of NEI 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.

Section 9.1 of NEI 94-01, Revisions 2, 2-A, and 3-A, states, in part, that: 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. Section 3.1.1.2 of the NRC staffs SE dated June 25, 2008, states, in part, that: 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 [topical report]

NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists. The NRC Regulatory Issue Summary 2008-27, Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50 (ADAMS Accession No. ML080020394), provides guidance regarding justification for such an extension request.

The February 3, 2020, letter from FENOC revised LAR Section 2.4 to state, in part, that the Type A test interval extensions of up to 9 months are permissible to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes.

LAR Table 3.8.1-1 states that Davis-Besse will follow the requirements in Section 9.1 of NEI 94-01, Revision 3-A, which are the same as NEI 94-01, Revision 2, and will demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required. Therefore, the NRC staff determined that the licensee understands that any extension of the Type A test interval beyond the performance-based limit of 15 years should be infrequent and that a request for such an extension will demonstrate to the NRC staff that an unforeseen emergent condition exists.

Based on its review of LAR Table 3.8.1-1 and the February 3, 2020, letter, the NRC staff determined that Davis-Besse will implement Condition 5 on NEI 94-01, Revision 2, consistent with the NRC staffs position described in Section 3.1.1.2 of the NRC staffs SE dated June 25, 2008. In addition, the proposed TS 5.5.15 will require Davis-Besse to comply with the conditions and limitations specified in NEI 94-01, Revision 2-A, which includes Condition 5 on NEI 94-01, Revision 2. Therefore, the NRC staff finds that Condition 5 on NEI 94-01, Revision 2, has been adequately addressed.

3.4.6 Condition 6 on NEI 94-01, Revision 2 Condition 6 applies only to plants licensed under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. Condition 6 is not applicable to Davis-Besse because it was licensed under 10 CFR Part 50.

3.4.7 Conclusion Regarding the Conditions on NEI 94-01, Revision 2 The NRC staff evaluated each of the conditions on NEI 94-01, Revision 2, listed in Section 4.1 of the NRC staffs SE dated June 25, 2008, and determined that each condition was adequately addressed in the LAR, as supplemented. The SE with these conditions is incorporated into NEI 94-01, Revision 2-A. Therefore, the NRC staff finds it acceptable for Davis-Besse to adopt the conditions and limitations specified in NEI 94-01, Revision 2-A, as part of the implementation documents listed in TS 5.5.15.

3.5 NEI 94-01, Revision 3-A As discussed in Section 2.2 of this SE, NEI 94-01, Revision 3-A, added guidance for extending Type C LLRT intervals beyond 60 months. In the NRC staffs SE for NEI 94-01, Revision 3, the staff concluded that NEI 94-01, Revision 3, describes an acceptable approach for implementing the optional performance-based requirements of Appendix J, and is acceptable for reference by licensees proposing to amend their containment leakage rate testing TSs, subject to two conditions related to Type C testing.

By letter December 19, 2014 (ADAMS Accession No. ML14353A349), as supplemented by letter dated June 26, 2015 (ADAMS Accession No. ML15180A040), FENOC submitted a LAR for Davis-Besse to adopt NEI 94-01, Revision 3-A, for Type C testing. On October 9, 2015, the NRC staff issued Amendment No. 288 to the Davis-Besse license, which approved the LAR to revise the Type C testing. In the SE for Amendment No. 288, the NRC staff determined that FENOC had adequately addressed the two NRC conditions for the use of NEI 94-01, Revision 3-A. The August 26, 2019, LAR does not propose any changes to the Type C testing at Davis-Besse. Therefore, the NRC staff determined that the two NRC conditions on the use of NEI 94-01, Revision 3, are not applicable to this LAR and the proposed changes do not impact the staffs previous findings regarding Amendment No. 288. Thus, the NRC staff finds it acceptable for Davis-Besse to adopt NEI 94-01, Revision 3-A, for Type A and B tests.

Section 8.0 of NEI 94-01, Revision 3-A, states that Type A, B, and C tests should be performed using the technical methods and techniques specified in ANSI/ANS-56.8-2002, or other alternative testing methods that have been approved by the NRC. Additionally, Section 2.0 of NEI 94-01, Revision 3-A, states that where differences exist between NEI 94-01, Revision 3-A, and ANSI/ANS-56.8-2002, the guidance in NEI 94-01, Revision 3-A, takes precedence.

Section 2.5 of the August 26, 2019, LAR, states that ANSI/ANS-56.8-2002 will be used for the performance of Type A and Type B testing at Davis-Besse. However, with the proposed change, Davis-Besse would be required to follow the stipulation that the guidance in NEI 94-01, Revision 3-A, takes precedence over ANSI/ANS-56.8-2002. The use of ANSI/ANS-56.8-2002 for the performance of Type A and Type B testing at Davis-Besse is acceptable to the NRC staff because the proposed change is consistent with Sections 2.0 and 8.0 of NEI 94-01, Revision 3-A.

3.6 Probabilistic Risk Assessment of the Proposed Extension of the ILRT Test Intervals Section 9.2.3.1 of NEI 94-01, Revision 3-A, states, in part, that the risk impact associated with extending the ILRT interval up to 15 years is small, but plant-specific confirmatory analyses are required when extending the ILRT interval beyond 10 years. Section 9.2.3.4 of NEI 94-01, Revision 3-A, states that a risk assessment should be performed using the approach and methodology described in EPRI Report No. 1009325, Revision 2-A.3 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 SE dated June 25, 2008, the NRC staff found the methodology described in EPRI Report No. 1009325, Revision 2, satisfies the key principles of risk-informed decision-making described in RG 1.174 and RG 1.77. In addition, the NRC staff found the methodology acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT interval to 15 years, provided the following conditions from Section 4.2 of the SE are satisfied:

1. The licensee submits documentation indicating that the technical adequacy of their PRA is consistent with the guidance in RG 1.200 relevant to the ILRT extension application.

2. The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small and consistent with the clarification provided in Section 3.2.4.6 of the NRC staffs SE dated June 25, 2008.4

3. The methodology in EPRI Report No. 1009325, Revision 2, is acceptable except for the calculation of the increase in expected population dose (per year of reactor operation).

In order to make the methodology acceptable, the average leak rate for the pre-existing containment large leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

4. A LAR is required in instances where containment overpressure is relied upon for ECCS performance.

Attachment B to the LAR provided a plant-specific risk assessment for permanently extending the ILRT interval at Davis-Besse from 10 years to 15 years. LAR Section 3.4.1 states, in part, that the plant-specific risk assessment follows the guidance in NEI 94-01, Revision 3-A; the methodology in EPRI Report No. 1009325, Revision 2-A; and the risk insights outlined in RG 1.174. A summary of how Davis-Besse met the four conditions for the use of EPRI Report No. 1009325, Revision 2, is provided in Sections 3.6.1 through 3.6.4 of this SE.

3.6.1 Technical Adequacy of PRA NRC Condition 1 on EPRI Report No. 1009325, Revision 2, stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the guidance in RG 1.200 relevant to the ILRT extension application. RG 1.200 describes one acceptable approach for determining whether the technical adequacy of the PRA (in total or the parts that are used to support an application) is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-water reactors.

The NRC staff used RG 1.200, Revision 2, to assess technical adequacy of the PRA used to support the LAR. Section 3.2.4.1 of the NRC staffs SE dated June 25, 2008, states that capability category I of the ASME PRA standard shall be applied as the standard for assessing 3 EPRI Report No. 1009325, Revision 2-A, incorporated the applicable changes to EPRI Report No. 1009325, Revision 2, identified in the NRC staffs SE dated June 25, 2008.

4 Condition 2 in Section 4.2 of the SE incorrectly referenced Section 3.2.4.5 instead of Section 3.2.4.6.

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 for use in the methodology described in EPRI Report No. 1009325, Revision 2.

The technical adequacy of the Davis-Besse PRA is addressed in Section 3.4.2 and Attachment B of the LAR. The risk assessment performed to support the LAR used the 2019 Davis-Besse internal events PRA model of record. The 2019 versions of the Davis-Besse PRA models are the most recent risk profile evaluations at Davis-Besse for internal events. The LAR states, in part, that: The Davis-Besse PRA model of record and supporting documentation have been maintained as a living program, with updates directed every other refueling cycle (approximately every four years) to reflect the as-built, as-operated plant. Additionally, interim updates to the PRA model may be prepared and issued as needed. The LAR also explains the approach used to establish and maintain the technical adequacy and plant fidelity of the Davis-Besse PRA models. This approach includes a PRA maintenance and update process, the use of self-assessments, and independent peer reviews.

Section 3.4.2.4 and Attachment B, Section A.2.1, of the LAR describe the peer reviews of the Davis-Besse PRA model for internal events. The PRA model received a formal industry peer review in 2008, and all the facts and observations (F&Os) identified during this review have been addressed. The internal events PRA model had been updated several times since 2008, and the F&Os from the focused-scope peer reviews associated with these updates have been addressed. A focused-scope peer review addressing the LERF was performed in 2011 using guidance provided in RG 1.200, Revision 2, and ASME/ANS RA-Sa-2009, Standard for Level l/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications. In 2012, a focused-scope PRA peer review was performed addressing internal flooding. Lastly, in October 2017, an independent assessment of F&Os and closeout review was performed. One F&O from this assessment remains open because one supporting requirement does not meet capability category II. The NRC staff determined that the one open F&O does not impact this LAR because the associated supporting requirement meets capability category I, which is acceptable in accordance with Section 3.2.4.1 of the NRC staffs SE dated June 25, 2008.

With respect to external events, RG 1.174 stipulates that established acceptance guidelines are intended for comparison with a full-scope assessment of the change in the applicable risk metrics. The guidance recognizes that many PRAs are not full scope and PRA information of less than full scope may be acceptable. The methodology described in EPRI Report No. 1009325, Revision 2-A, indicates that if the external event analysis is not of sufficient quality or detail to allow direct application of the methodology, 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 another alternate method of assessing an order-of-magnitude estimate for contribution of the external event to the impact of the changed interval. Section 5.2.7 of LAR Attachment B provides a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension. This analysis references the currently available information for external events models and information to develop an external events multiplier to be applied to the internal events results.

Section 3.4.2.5 and Attachment B, Section A.2.2, of the LAR describe the peer review of the Davis-Besse fire PRA model. The fire PRA model was peer reviewed in 2013, and the associated F&Os were addressed using FENOCs PRA program to disposition each individual F&O to ensure the model satisfied the PRA standard requirements. The October 2017 independent assessment and closeout review addressed FENOC's disposition to the F&Os from the fire PRA peer review. There were five finding-level F&Os that remained open after the independent assessment which are described in Attachment B, Table A-1, of the LAR. The independent assessment also identified a suggestion-level F&O to improve the PRA documentation. The LAR indicated that four of these F&Os were resolved by making appropriate changes to the PRA models. The remaining F&O was identified by the independent assessment team as a documentation issue. The independent assessment for fire PRA determined that each associated supporting requirement met at least capability category II.

Section 3.4.2.6 and Attachment B, Section A.2., of the LAR describe the peer review of the Davis-Besse seismic PRA model. The seismic PRA Model was peer reviewed in July 2014, and the associated F&Os were addressed using FENOC's PRA program to disposition each individual F&O to ensure the model satisfies the PRA standard requirements. The October 2017 independent assessment and closeout review addressed FENOC's disposition of the F&Os from the seismic PRA peer review. The independent assessment for seismic PRA closed the F&Os and determined that each associated supporting requirement meets at least capability category II.

Based on the review of Section 3.4.2 and Attachment B of the LAR, as discussed above, the NRC staff finds the licensee has addressed the relevant findings and gaps from the peer reviews and determined that they have no impact on this LAR. Therefore, the NRC staff concludes that the Davis-Besse PRA models are of sufficient quality to support the evaluation of changes to ILRT intervals, and Condition 1 on EPRI Report No. 1009325, Revision 2, has been adequately addressed.

3.6.2 Estimated Risk Increase NRC Condition 2 on EPRI Report No. 1009325, Revision 2, stipulates that the licensee demonstrate that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small and consistent with the clarification provided in Section 3.2.4.6 of the NRC staffs SE dated June 25, 2008. 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.

LAR Table 3.4.1-1 provides information in response to NRC Condition 2 on EPRI Report No. 1009325, Revision 2. The results of the Davis-Besse risk assessment and sensitivity calculations for the extension of the ILRT interval to 15 years are provided in Sections 5.2 and 5.3, respectively, of LAR Attachment B. LAR Section 3.4.3 provides a summary of these results and FENOCs conclusions based on this assessment. The assessment included an evaluation of risk impacts for a change in the Type A containment test frequency from three tests in 10 years (the test frequency specified in 10 CFR Part 50, Appendix J, Option A) to one test in 15 years. The risk impacts from this evaluation bounds the risk impacts for the proposed change in the Type A test frequency from one test in 10 years to one test in 15 years.

The NRC staff determined that the most relevant risk criterion for the proposed change is LERF because a change in the ILRT frequency does not impact CDF and Davis-Besse does not rely upon containment overpressure for ECCS performance (see Section 3.6.3 of this SE). The LAR states that the increase in LERF resulting from a change in the Type A test frequency from three tests in 10 years to one test in 15 years is estimated to be 4.83x10-8 per year. The LAR states that 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. The risk increase is less than 10-7 which is considered very small under the acceptance guidelines in RG 1.174. When external events are included in this evaluation, the increase in LERF is estimated to be 6.06x10-7 per year and total LERF is 5.50x10-6 per year. The risk increase is between 10-7 and 10-6 per year with a total LERF less than 10-5 per year, which is considered small under the acceptance guidelines in RG 1.174.

LAR Section 3.4.3 states that the calculated change in population dose resulting from decreasing the ILRT test frequency to once per 15 years is 0.016 person-rem per year. NRC Condition 2 on EPRI Report No. 1009325, Revision 2, states that 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.

Therefore, the NRC staff determined that the calculated change in population dose resulting from the proposed change is small per the definition in Condition 2.

Section 5.2.5 of LAR Appendix B evaluates the impact on CCFP from a change in the ILRT frequency from three tests in 10 years to one test in 15 years. Section 5.2.5 indicates that the CCFP would increase by 0.870 percentage points as a result of the evaluated change.

Condition 2 on EPRI Report No. 1009325, Revision 2, states the increase in CCFP should be less than or equal to 1.5 percentage points. Therefore, the NRC staff determined that the increase in CCFP for Davis-Besse is acceptable per Condition 2.

Based on the risk assessment results provided in the LAR, the NRC staff concludes that the increase in LERF for the proposed change is small and consistent with the acceptance guidelines of RG 1.174. The increase in the total population dose and CCFP for the proposed change are also small. The defense-in-depth philosophy is maintained as the independence of barriers will not be degraded because of the requested change and the use of the quantitative risk metrics collectively ensures that the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. Therefore, the NRC staff finds that Condition 2 on EPRI Report No. 1009325, Revision 2, has been adequately addressed.

3.6.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case NRC Condition 3 on EPRI Report No. 1009325, Revision 2, stipulates that to make the methodology in EPRI Report No. 1009325, Revision 2 acceptable, the average leak rate for the pre-existing containment large leak rate accident case (accident case 3b) used by the licensee shall be 100 La instead of 35 La. LAR Table 3.4.1-1 states that 100 La is used as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b).

Section 4.0 of LAR Attachment B also states this was an assumption used in the plant-specific risk assessment for Davis-Besse. Therefore, the NRC staff finds that Condition 3 on EPRI Report No. 1009325, Revision 2, has been adequately addressed.

3.6.4 Applicability if Containment Overpressure is Credited for ECCS Performance NRC Condition 4 on EPRI Report No. 1009325, Revision 2, stipulates that a LAR is required in instances where containment overpressure is relied upon for ECCS performance.

LAR Table 3.4.1-1 states that Davis-Besse does not rely upon containment overpressure for ECCS performance. LAR Section 3.1 states that the ECCS and containment heat removal system pumps at Davis-Besse are the high-pressure core injection pumps, low-pressure core injection pumps, and containment spray pumps. The calculation for the net positive suction head for the low-pressure core injection pumps and the containment spray pumps following a loss-of-coolant accident assumes that the vapor pressure is equal to the containment pressure; thus, only a static head (difference in elevations between the pump centerline and emergency sump) was available. In addition, for added conservatism, only a minimum water level inside containment was used. Thus, no credit was taken for the containment pressure. LAR Section 3.1 further states: If required in the recirculation mode, the high-pressure core injection pump takes suction from the discharge of the low-pressure core injection pump. Therefore, no credit has been taken for the containment pressure in determining the [net positive suction head] for the high-pressure core injection pumps.

The NRC staff determined that a separate LAR is not needed to satisfy Condition 4 on EPRI Report No. 1009325, Revision 2, because Davis-Besse does not rely upon containment overpressure for ECCS performance.

3.6.5 Conclusion Regarding Conditions on EPRI Report No. 1009325, Revision 2 The NRC staff evaluated each of the conditions on EPRI Report No. 1009325, Revision 2, listed in Section 4.2 of the NRC staffs SE dated June 25, 2008, and determined that each condition was adequately addressed in the LAR. These conditions are incorporated into NEI 94-01, Revision 2-A. Therefore, the NRC staff finds it acceptable for Davis-Besse to adopt the conditions and limitations specified in NEI 94-01, Revision 2-A, as part of the implementation documents listed in TS 5.5.15.

3.7 Technical Conclusion Based on the preceding regulatory and technical evaluations, the NRC staff finds that the licensee has adequately implemented its existing primary containment leakage rate testing program consisting of ILRTs and LLRTs. The results of the recent ILRTs and the combined totals for LLRTs demonstrate acceptable performance and support a conclusion that the structural integrity and leaktightness of the primary containment is adequately managed and will continue to be periodically monitored and managed effectively with the proposed changes. The NRC staff finds that FENOC has adequately addressed the NRC conditions for adopting NEI 94-01, Revision 2-A. The staff previously found that FENOC had adequately addressed the NRC conditions for adopting NEI 94-01, Revision 3-A, for Type C testing when it issued Amendment No. 288 to the Davis-Besse license. Lastly, the NRC staff finds that the Davis-Besse PRA is of sufficient technical adequacy to support the evaluation of changes to ILRT frequency. Therefore, the NRC staff concludes that the proposed changes to the primary containment leakage rate testing program at Davis-Besse are acceptable.

The NRC staff determined that the revised TS 5.5.15 will include the implementation documents used by Davis-Besse to develop the performance-based leakage-testing program, as required by Section V.B.3 of 10 CFR Part 50, Appendix J, Option B. In addition, the NRC staff determined that the Davis-Besse TSs will continue to meet 10 CFR 50.36(c)(5) by specifying provisions necessary to assure operation of the facility in a safe manner. Therefore, the NRC staff concludes that the proposed changes to TS 5.5.15 are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment on July 1, 2020. 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 areas as defined in 10 CFR Part 20 and changes surveillance requirements. 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, which was published in the Federal Register on October 22, 2019 (84 FR 56481), that the amendment involves no significant hazards consideration, and there has been no public comment on such finding. 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 Commission's 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: Brian E. Lee, NRR Jerry Dozier, NRR Zach Coffman, NRR Samual Cudrado de Jesus, NRR Date of Issuance: August 24, 2020

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