ML20217L185

Issuance of Amendment Nos. 232 and 220 Increase the Integrated Leak Rate Test Program Type a and Type C Test Frequency

ML20217L185
Person / Time
Site:	Prairie Island
Issue date:	10/02/2020
From:	Robert Kuntz Plant Licensing Branch III
To:	Sharp S Northern States Power Co
Kuntz R
References
EPID L-2019-LLA-0226
Download: ML20217L185 (37)

Text

October 2, 2020 Mr. Scott Sharp Site Vice President Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota 1717 Wakonade Drive East Welch, MN 55089

SUBJECT:

PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 -

ISSUANCE OF AMENDMENT NOS. 232 AND 220 RE: INCREASE THE INTEGRATED LEAK RATE TEST PROGRAM TYPE A AND TYPE C TEST FREQUENCY (EPID L-2019-LLA-0226)

Dear Mr. Sharp:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 232 to Renewed Facility Operating License No. DPR-42 and Amendment No. 220 to Renewed Facility Operating License No. DPR-60 for the Prairie Island Nuclear Generating Plant, Units 1 and 2, respectively. The amendments consist of changes to the technical specifications (TSs) in response to your application dated October 7, 2019.

The amendments revise TS 5.5.14, Containment Leakage Rate Testing Program, to increase the containment integrated leakage rate test program Type A test interval from 10 to 15 years and extend the containment isolation valve Type C leakage rate test frequency from 60 to up to 75 months.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA/

Robert F. Kuntz, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-282 and 50-306

Enclosures:

1. Amendment No. 232 to DPR-42

2. Amendment No. 220 to DPR-60

3. Safety Evaluation cc: Listserv

NORTHERN STATES POWER COMPANY - MINNESOTA DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 232 Renewed License No. DPR-42

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

A. The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated October 7, 2019, 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.

Enclosure 1

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. DPR-42 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 232, are hereby incorporated in the renewed operating license. The licensee 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 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Scott P. Scott P. Wall Date: 2020.10.02 Wall 13:30:33 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 2, 2020

NORTHERN STATES POWER COMPANY - MINNESOTA DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 220 Renewed License No. DPR-60

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

A. The application for amendment by Northern States Power Company, a Minnesota Corporation (NSPM, the licensee), dated October 7, 2019, 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.

Enclosure 2

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. DPR-60 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 220, are hereby incorporated in the renewed operating license. The licensee 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 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Scott P. Scott P. Wall Date: 2020.10.02 Wall 13:30:57 -04'00' Nancy L. Salgado, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: October 2, 2020

ATTACHMENT TO LICENSE AMENDMENT NOS. 232 AND 220 RENEWED FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Renewed Facility Operating License Nos. DPR-42 and DPR-60 with the attached revised pages. The changed areas are identified by a marginal line.

REMOVE INSERT Page 3 Page 3 Page 3 Page 3 Technical Specifications Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contain marginal lines indicating the areas of change.

REMOVE INSERT 5.0-28 5.0-28

(3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM 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 and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purpose of volume reduction and decontamination.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:

Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.232, are hereby incorporated in the renewed operating license. NSPM shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-42 Amendment No. 232

(3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NSPM 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) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, NSPM 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 and equipment calibration or associated with radioactive apparatus or components; (5) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (6) Pursuant to the Act and 10 CFR Parts 30 and 70, NSPM to transfer byproduct materials from other job sites owned by NSPM for the purposes of volume reduction and decontamination.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter l:

Part 20, Section 30.34 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; 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 NSPM is authorized to operate the facility at steady state reactor core power levels not in excess of 1677 megawatts thermal.

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 220, are hereby incorporated in the renewed operating license.

NSPM shall operate the facility in accordance with the Technical Specifications.

(3) Physical Protection NSPM shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Renewed Operating License No. DPR-60 Amendment No. 220

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.14 Containment Leakage Rate Testing Program

a. 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 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exception:

1. Unit 1 and Unit 2 (steam generator (SG) replacement commencing Fall 2013) are excepted from post-modification integrated leakage rate testing requirements associated with SG replacement.

b. The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal design pressure, Pa, of 46 psig.

c. The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.15% of primary containment air weight per day. For pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.06% of primary containment air weight per day at pressure Pa. For pipes connected to systems that are exterior to both the shield building and the auxiliary building special ventilation zone, the total leakage past isolation valves shall be less than 0.006% of primary containment air weight per day at pressure Pa.

Prairie Island Unit 1 - Amendment No. 232 Units 1 and 2 5.0-28 Unit 2 - Amendment No. 220

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 232 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. 220 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-60 NORTHERN STATES POWER COMPANY - MINNESOTA PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By application dated October 7, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19280B335), Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (NSPM or licensee), requested changes to the technical specifications (TSs) for the Prairie Island Nuclear Generating Plant (PINGP),

Units 1 and 2. The license amendment request (LAR) proposes changes to TS 5.5.14, Containment Leakage Rate Testing Program, by replacing the TS 5.5.14 reference to Regulatory Guide (RG) 1.163, Performance Based Containment Leak-Test Program (ADAMS Accession No. ML003740058), with a reference to Nuclear Energy Institute (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 conditions and limitations specified in NEI 94-01, Revision 2-A, of the same name, dated October 2008 (ADAMS Accession No. ML100620847). The proposed amendment would allow extension of the Type A test interval from 10 to 15 years and extension of the Type C test interval from 60 to up to 75 months, based on acceptable performance history as defined in NEI 94-01, Revision 3-A. The LAR also proposes to revise TS 5.5.14 by deleting one of the two listed exceptions to program guidelines because it is no longer necessary.

2.0 REGULATORY EVALUATION

2.1 System Description PINGP Units 1 and 2 consist of two pressurized water reactors, located in Welch, MN, approximately 28 miles southeast of Minneapolis, MN. The total containment for each PINGP unit consists of two systems, the Primary Containment and the Secondary Containment.

Enclosure 3

The Primary Containment system consists of a steel structure, also referred to as the Reactor Containment Vessel, and its associated engineered safety feature systems. The Reactor Containment Vessel is a low leakage steel shell. The system, including penetrations, is designed to confine radioactive material that could be released by accidental loss of integrity.

The Secondary Containment system for each PINPG unit consists of a Shield Building with its associated engineered safety features systems and the Category I Ventilation Zone with its associated engineered safety features systems. The Shield Building is a concrete structure surrounding the Reactor Containment Vessel.

2.2 Licensees Proposed Changes The LAR proposes changes to TS 5.5.14, Containment Leakage Rate Testing Program, by replacing the TS 5.5.14 reference to RG 1.163 with a reference to NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A. This proposed change would allow a Type A test interval of up to 15 years and a Type C interval of up to 75 months.

The LAR also proposes to delete an exception listed in TS 5.5.14 which required that the PINGP, Unit 1 Type A test performed after December 1, 1997 be performed by December 1, 2012 and the PINGP, Unit 2 Type A test performed after March 7, 1997 be performed by March 7, 2012. The LAR proposes removal of this exception as an administrative change because the required tests have been completed.

2.3 Regulatory Requirements The LAR requested a change to the Renewed Facility Operating Licenses for PINGP, Units 1 and 2, in accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit.

The regulations in 10 CFR 50.36(c)(5), Administrative controls, require, in part, the inclusion of administrative controls in TSs that are necessary to ensure operation of the facility in a safe manner. The LAR requested a change to Administrative Controls section of the PINGP, Units 1 and 2, TSs.

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 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. Appendix J contains two options: Option A - Prescriptive Requirements and Option B - Performance-Based Requirements, either of which can be used to meet Appendix J requirements.

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

Section V.B.3 of 10 CFR Part 50, Appendix J, Option B, requires the licensee to develop a performance-based leakage-testing program using the RG or other implementation document and referencing it in the plant TS. The submittal for TS revisions must also contain justification, including supporting analyses, if the licensee deviates from methods approved by the U.S. Nuclear Regulatory Commission (NRC or Commission) and endorsed in RG 1.163, Performance-Based Containment Leak-Test Program.

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 (CIV) leakage rates.

After the containment system has been completed and is ready for operation, Type A tests are conducted at periodic intervals based on the historical performance of the overall containment system to measure the overall integrated leakage rate. The leakage rate test results must not exceed the maximum allowable leakage (La) at design-basis loss-of-coolant accident (DBLOCA) pressure (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 leak-tight 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 historical performance 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 leak-tight and structural integrity of the containment during its service life.

Section 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, paragraph (a)(1), 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.

2.4 Regulatory Guidance 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), provides an acceptable approach for developing risk-informed applications for licensing basis changes that considers engineering issues and applies risk insights.

RG 1.174 provides general guidance concerning analysis of the risk associated with proposed changes in plant design and operation.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment [PRA] Results for Risk-Informed Activities (ADAMS Accession

No. ML090410014), provides guidance for determining the technical adequacy and quality of the PRA.

Regulatory Issue Summary (RIS) 2007-06, Regulatory Guide 1.200 Implementation (ADAMS Accession No. ML070650428), provides information regarding how the NRC will implement its technical adequacy review of plant-specific PRAs in support of RG 1.200.

NEI 94-01, Revision 0 (ADAMS Accession No. ML11327A025), provides methods for complying with the provisions of 10 CFR Part 50, Appendix J, Option B, and includes provisions that address the extension of the performance-based Type A test interval for up to 10 years, based upon two consecutive successful tests.

The final safety evaluation (SE) for NEI Topical Report (TR) 94-01, Revision 2, Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated June 25, 2008 (ADAMS Accession No. ML081140105), 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. The NRC staff concluded that NEI 94-01, Revision 2, is acceptable for referencing by licensees proposing to amend their containment leakage rate testing TSs, subject to the specific limitations and conditions listed in Section 4.1 of the SE.

NEI 94-01, Revision 2-A, incorporates the regulatory positions stated in RG 1.163, and includes provisions for extending Type A test intervals up to 15 years.

EPRI TR-1009325, Revision 2-A, a risk-informed methodology using plant-specific risk insights and industry ILRT performance data to extend ILRT surveillance frequencies to 15 years. The NRC staff found the proposed methodology satisfies the key principles of risk-informed decision-making applied to changes to the TS as delineated in RG 1.177, An Approach to Plant-Specific, Risk-Informed Decision Making: Technical Specifications, and RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-specific Changes to the Licensing Basis. The NRC staff found that this guidance was acceptable for referencing by licensees proposing to amend their TS regarding containment leakage rate testing subject to the limitations and conditions noted in Section 4.2 of the June 25, 2008 safety evaluation report (SER) for EPRI Report No. 1009325, Revision 2. Revision 2-A to NEI technical report 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, was revised to incorporate the June 25, 2008 NRC staffs Final Safety Evaluation Report (SER) and has been reissued with an "-A" (designating accepted) following the report number. The June 25, 2008 SER is referred to as the SER for EPRI TR-1009325, Revision 2 in the Technical Evaluation.

NEI 94-01, Revision 3-A, July 2012, provides guidance for extending Type C local leak rate test (LLRT) intervals beyond 60 months. The NRC staff published an SE with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012 (ADAMS Accession No. ML121030286). In the SE, 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. The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012.

TS 5.5.14.c requires that Types A, B, and C, test results must not exceed the maximum allowable primary containment leakage rate, La, with margin. Option B of Appendix J to 10 CFR

Part 50 requires that a general visual inspection of the accessible interior and exterior surfaces of the containment system for structural deterioration, which may affect the containment leak-tight integrity, be conducted prior to each Type A test and at a periodic interval between tests, based on the performance of the containment system. A Type A ILRT is currently required to be performed once every 10 years. The LAR asks to extend the interval to 15 years.

3.0 TECHNICAL EVALUATION

3.1 Deletion of Exception in TS 5.5.14 The LAR proposed to delete the exception to perform post-modification leakage rate testing by a certain date which was associated with steam generator replacements at PINGP, Units 1 and 2. This provision was related to leakage testing completed for each unit in 2012. Because the requirement is related to testing which has been performed, the requirement is no longer relevant and, therefore, the NRC staff finds the deletion acceptable.

3.2 Integrated Leak Rate Testing History (Type A Testing)

On January 22, 2013, the NRC approved Amendments 206 and 193, for PINGP, Units 1 and 2, respectively, which modified the PINGP licensing basis and TSs to reflect adoption of the Alternate Source Term (AST) methodology. With AST implementation, the allowable TS La was reduced going forward.

Per TS 5.5.14, PINGP specified La to be 0.15 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pa. Prior to January 22, 2013, La was 0.25 percent of containment air weight per day at the calculated peak pressure. TS 5.5.14 indicates that the peak calculated containment internal pressure for a DBLOCA, Pa, is 46 pounds per square inch (psig).

There have been four ILRTs performed on the PINGP, Unit 1, containment since June 1991 and a total of three ILRTs performed on the PINGP, Unit 2, containment since January 1993. The LAR provided the ILRT results which show substantial margin has been maintained relative to the performance criterion for the most recent Type A tests for both units, so the extended interval would be allowed by program guidance for PINGP. In addition, no adverse trend is apparent that would suggest the performance criterion might be exceeded with the requested interval extension to 15 years. The test results of these ILRTs were documented in LAR Section 3.2.1. The four test results for PINGP, Unit 1, and three test results for PINGP, Unit 2, are summarized in Table 3.1.1 and Table 3.1.2 below.

TABLE 3.1.1 PINGP, Unit 1, Type A ILRT History Test Test Pressure Design As-Found ILRT As-Left Date (psig) Pressure Leakage Acceptance Leakage (3) (psig) (wt%/day) Criteria(1) (wt%/day)

(2) (wt%/day) (2)

June 1991 23 46 0.0667 0.11569 (0.75 0.0597 L T)

June 1994 23 46 0.0998 0.11569 (0.75 0.0772 L T)

Dec. 1997 46 46 0.0413 0.25 (1.0 La) 0.0445 Oct. 2012 46 46 0.0116 0.25 (1.0 La) 0.0173 Table 3.1.1 Notes:

(1)

LT = Maximum Allowable Test Leakage Rate at reduced test pressure (23 psig). This allowance was removed when 10 CFR Part 50, Appendix J, Option B was implemented.

(2)

Data source LAR Section 3.3.5.

(3)

The original 10 CFR, Appendix J regulation allowed reduced pressure testing (half the design pressure).

TABLE 3.1.2 PINGP, Unit 2 ,Type A ILRT History Test Test Pressure Design As-Found ILRT As-Left Date (psig) Pressure Leakage Acceptance Leakage (3) (psig) (wt%/day) Criteria(1) (wt%/day)

(2) (wt%/day) (2)

Jan. 1993 23 46 0.0307 0.13258 (0.75 0.0158 L T)

Mar. 46 46 0.0435 0.25 (1.0 La) 0.0418 1997 Mar. 46 46 0.0284 0.25 (1.0 La) 0.0288 2012 Table 3.1.2 Notes:,

(1)

LT = Maximum Allowable Test Leakage Rate at reduced test pressure (23 psig). This allowance was removed when 10 CFR Part 50, Appendix J, Option B was implemented.

(2)

Data source LAR Section 3.3.5.

(3)

The original 10 CFR, Appendix J regulation allowed reduced pressure testing (half the design pressure).

The NRC staff notes that Section 9.1.2 of NEI 94-01, Revision 3-A, reads in part that [t]he elapsed time between the first and the last tests in a series of consecutive passing tests used to determine performance shall be at least 24 months. As can be seen in both Table 3.1.1 and Table 3.1.2, the recommendation of NEI 94-01, Section 9.1.2, has been satisfied.

The Appendix J, Option B, in PINGP TS 5.5.14 references document RG 1.163. Regulatory Position C of RG 1.1.63 in turn states that NEI 94-01, Revision 0, provides methods acceptable to the NRC staff for complying with the provisions of Option B in Appendix J to 10 CFR Part 50, . The third paragraph of Section 9.2.3 Extended Test Intervals of NEI 94-01, Revision 0, reads in part:

In reviewing past performance history, Type A test results may have been calculated and reported using computational techniques other than the Mass Point method from ANSI/ANS-[American National Standards Institute/American Nuclear Society] 56.8-1994 (e.g., Total Time or Point-to-Point). Reported test

results from these previously acceptable Type A tests can be used to establish the performance history. Additionally, a licensee may recalculate past Type A Upper Confidence Limit (UCL) (using the same test intervals as reported) in accordance with ANSI/ANS-56.8-1994 Mass Point methodology and its adjoining Termination criteria in order to determine acceptable performance history.

NEI 94-01, Revision 3-A reads nearly identically except the test standard invoked is ANSI/ANS-56.8-2002.

The NRC staff notes that Section 9.2.3 does not mandate that a licensee recalculate past Type A test results to demonstrate conformance with the definition of performance leakage rate contained in NEI 94-01, Revision 3-A. The NRC staff also notes that the ILRT results since December 1997 (PINGP, Unit 1) and March 1997 (PINGP, Unit 2) demonstrated ample margin (i.e., > 82 percent) between each As-found leakage value and La. Accordingly, Type A Test results from earlier than the ILRTs of December 1997 (PINGP, Unit 1) and March 1997 (PINGP, Unit 2) are not necessary.

PINGP, Units 1 and 2, TS 5.5.14 establishes the maximum limit for the as-left leakage rate for startup following completion of Type A testing at 0.75 La, which currently equals 0.1875 percent of containment air weight per day.

Prior to January 22, 2013, the PINGP, Units 1 and 2, TS specified a leakage rate La not to exceed 0.25% by weight of containment air per day at the calculated peak pressure, Pa. As displayed in both Table 3.1.1 and Table 3.1.2, there has been adequate margin to the performance limit as described in TS 5.5.14 of La for the historical ILRTs spanning a period of at least nineteen years.

3.2.1 Integrated Leak Rate Testing History (Type A Testing) Conclusion The past two ILRT results for PINGP, Units 1 and 2, dating back to 1997 have confirmed that the primary containment leakage rates are acceptable with respect to the design criterion leakage of containment air weight (La) per day. Since the last two Type A tests for PINGP, Units 1 and 2, had as found test results well within the current maximum allowable containment leakage rate specified in TS 5.5.14 of 0.15 weight-percent per day, a test frequency of 15 years in accordance with NEI 94-01, Revision 3-A, and the conditions and limitations of NEI 94-01, Revision 2-A, would be acceptable. Based on the last two PINGP, Units 1 and 2, ILRT test results, the NRC staff concludes that the requirements of Section 9.1.2 of NEI 94-01, Revision 3-A, have been satisfied.

3.3 Type B and C Testing PINGP, Units 1 and Unit 2, Type B and C leakage is comprised of three different zones: the Auxiliary Building Special Ventilation Zone (ABSVZ); the Exterior Zone of containment; and the containment Annulus. The ABSVZ and the Exterior Zone have leakage rate limits specified within the TS. The Annulus does not have a separate TS specified leakage limit, so Annulus leakage is included in the leakage summation together with that from the ABSVZ and Exterior Zone of containment and is compared against 0.6 La.

As noted in SE Section 3.1 above, on January 22, 2013, the NRC approved Amendments 206 and 193, for PINGP Units 1 and 2, respectively, which modified the PINGP, Units 1 and 2,

licensing basis and TS to reflect adoption of the AST methodology. With AST implementation, the allowable TS La was reduced going forward.

The NRC staff reviewed the local leak rate summaries contained in LAR Section 3.2.2 Type B and Type C Testing. For PINGP, Units 1 and 2, respectively, the combined Type B and Type C leakage acceptance criterion is 0.60 La or 154,800 standard cubic centimeters per minute (sccm) prior to 2013 and 92,880 sccm thereafter. NEI 94-01, Revision 3-A, Section 10.2, indicates that this criterion is to be evaluated for the combined Type B and C as-found minimum pathway test total and as a restart permissive criterion for the combined Type B and C as-left maximum pathway test total. The NRC staff notes that the Type B and Type C test results show a large amount of margin between the actual as-found and the as-left outage summations and the respective TS leakage rate acceptance criteria.

With the use of these La values and the data contained in LAR Section 3.2.2, the NRC staff confirmed the accuracy of the Fraction of 0.6 La values contained in the LAR and concluded that:

The PINGP, Unit 1, as-found minimum pathway leakage rates for the last 10 refueling outages since 2002 have an average of 8.8 percent of 0.6 La with a high of 13.5 percent 0.6 La.

The PINGP, Unit 1, as-left maximum pathway leakage rates for the last 10 refueling outages since 2002 have an average of 12.8 percent of 0.6 La with a high of 18.0 percent 0.6 La.

The PINPG, Unit 2, as-found minimum pathway leakage rates for the last nine refueling outages since 2003 have an average of 4.2 percent of 0.6 La with a high of 9.1 percent 0.6 La.

The PINGP, Unit 2, as-left maximum pathway leakage rates for the last nine refueling outages since 2003 have an average of 7.3 percent of 0.6 La with a high of 15.4 percent 0.6 La.

As conveyed in LAR Section 3.2.2, for PINGP, Unit 1, there have been no LLRT failures in the past 36 months (two outages). However, for PINGP Unit 2, there were two penetrations that exceeded the administrative limit and one penetration that had a test failure in the past 48 months (two outages). These Type B or Type C penetration test failures during the two most recent refueling outages for PINGP, Unit 2, are described below:

During Prairie Island, Unit 2 Refueling Outage 29 (2R29) (Fall 2015), CV-31736 (CIV) exceeded the administrative limit due to the leakage on the downstream side of CV-31735. Both CV-31736 and CV-31735 had a valve overhaul and actuator replaced. The as-left leakage was 12 sccm (CV-31736) and 3 sccm (CV-31735).

During 2R30 (Fall 2017), Valve MV-32210 failed due to disc degradation. The valve was disassembled; the disc was repaired and retested with an as-left leakage of 2 sccm.

During 2R30, the as-found leakage for Penetration 34 was 3,347 sccm, which exceeded the administrative limit due to a loose union connection on one electrical penetration assembly. The connection was tightened and the as-left leakage test for the electrical penetration was completed satisfactory (1,967 sccm).

The LAR indicated that following satisfactory performance of two consecutive as-found LLRTs, these penetrations will return to an extended testing frequency. The NRC staff notes that this course of action is consistent with the guidance of NEI 94-01, Revision 0, Section 10.2.3 Type C Test Interval.

PINGP has a total of 28 Type B tested penetrations on each unit. Of the 28 penetrations, 19 penetrations for PINGP, Unit 1, and 18 penetrations for PINGP, Unit 2, are currently on extended test frequency. For both PINGP, Units 1 and 2, as-left testing is performed when the penetrations are opened. If these penetrations are not opened for multiple outages, the penetrations are eligible for extended frequency testing. Measured leakage for these penetrations has not changed significantly over 120 months.

PINGP has a total of 28 Type C tested penetrations on each unit. Of the 28 penetrations, 16 penetrations for PINGP, Unit 1, and 14 penetrations for PINGP, Unit 2, are currently on extended test frequency. The vent and purge valves have been modified to include blind flanges inside containment. The number of penetrations on extended frequency is adjusted periodically based on valve performance and other plant testing requirements.

Based on the NRC staffs review of the historical information provided in LAR Section 3.2.2 Type B and Type C Testing, the NRC staff observed that there was no indication of the licensees failure to adequately implement the requirements of its Appendix J, Option B, performance-based testing program.

In summary, the licensee provided an adequate explanation of the cause of failure for the LLRT Type C penetration experienced during refueling outage 2R29. Furthermore, based on the review of LAR Section 3.2.2, the NRC staff concluded that the aggregate leakage rate results of the As Found Minimum Pathway for all PINGP, Units 1 and 2, Type B and C tests from the last two outages respectively, demonstrates a history of adequate maintenance since the aggregate test results at the end of each operating cycle were all well below (i.e., > 95 percent margin) the Type B and Type C test TS Leakage Rate acceptance criteria of < 0.60 La contained in TS 5.5.14.

3.3.1 Type B and Type C Test Program Conclusion In summary, the NRC staff concludes that:

PINGP, Units 1 and 2, LLRT have been completed consistent with the guidance of RG 1.163 and NEI 94-01, Revision 0, the recent historical combined total Type B and C test results are substantially below the acceptance limit of TS 5.5.14, and the licensees corrective action program has appropriately addressed poor performing valves and penetrations.

Therefore, the NRC staff finds that the licensee is effectively implementing the PINGP, Units 1 and 2, Type B and Type C leakage rate test program, as required by Option B of 10 CFR Part 50, Appendix J.

3.4 Containment Inspection 3.4.1 Containment ISI Program Section 3.3.1 of the LAR lists the applicable editions and addenda of the American Society of Mechanical Engineers (ASME) Code,Section XI, Subsection IWE for the containment ISIs, as required by 10 CFR 50.55a (g)(4)(ii), and states that NSPM has established a containment ISI program for PINGP, Units 1 and 2, in accordance with 10 CFR 50.55a for Class MC components.

The NRC staff reviewed the information provided in the Section 3.3.1 of the LAR and noted that each 10-year inspection interval consists of three examination periods, and inspection periods are divided into three refueling outages. The NRC staff confirmed that containment ISI intervals for the PINGP, Units 1 and 2, are in accordance with the requirements of the ASME code,Section XI, Subsection IWE, examinations. The NRC staff also reviewed the results of recent containment IWE inspections described in the Section 3.3.1 of the LAR and noted that no indications of significant degradations have been identified in the past ASME Section XI, Subsection IWE inspections, and observed degradations were appropriately evaluated.

3.4.2 Containment Visual Inspection Program Section 3.3.2 of the LAR describes the general visual examinations performed for the accessible interior and exterior surfaces of the containment vessel in accordance with SRs and TS 5.5.14, and describes the proposed change to TS 5.5.14 by replacing the reference to RG 1.163 with a reference to NEI 94-01, Revision 3-A. A general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity is required by NEI 94-01, Revision 3-A, 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.

The NRC staff reviewed information provided in the Section 3.3.2 of the LAR, and noted that the PINGP, Units 1 and 2, containment leak rate testing program credits these visual examinations which were performed in accordance with the ASME Section XI, Subsection IWE, containment inspection programs.

3.4.3 Inaccessible Areas Section 3.3.3 of the LAR states that the PINGP, Units 1 and 2, ISI Summary Report, as required by 10 CFR 50.55a(b)(2)(ix)(A), provides:

• a description of the type and estimated extent of degradation, and the conditions that led to the degradation,

• an evaluation of each area, and the result of the evaluation, and

• a description of necessary corrective actions The regulation in 10 CFR 50.55a(b)(2)(ix)(A) requires that for Class MC applications, the licensee must evaluate the acceptability of inaccessible areas when conditions exist in

accessible areas that could indicate the presence of or could result in degradation to such inaccessible areas, and that the specified information be provided for each inaccessible area identified for evaluation. The NRC staff reviewed the ASME code,Section XI, Subsection IWE 2013, edition and finds that inaccessible surface areas identified in the IWE component database are exempted from examination because they have met the requirements of the original construction code. Therefore, the licensee is not required to perform inspections of inaccessible surface areas identified in the IWE component database.

3.4.4 Containment Coatings Inspection Program Section 3.3.4 of the LAR describes the PINGP containment coatings program and its requirements for the purpose of assessing the condition of the protective coatings on structures and equipment in the reactor containment building.

The NRC staff reviewed the results of recent coatings inspections provided in Section 3.3.4 of the LAR, and noted that the condition of the containment coatings was acceptable, that no immediate corrective actions were required to meet design and licensing basis requirements, and that the total quantity of degraded qualified and unqualified coatings remains within the bounds required by the design basis, with margins remaining below administrative limits.

3.4.5 Maintenance Rule Section 3.3.5 of the LAR states that the maintenance rule is implemented in accordance with the PINGP, Units 1 and 2, maintenance rule program. The proposed change to TS 5.5.14 changes the IRLT and LLRT test intervals and does not impact the PINGP, Units 1 and 2, maintenance rule program. Therefore, the PINGP, Units 1 and 2, maintenance rule program is not affected by the LAR.

3.4.6 Containment Inspection Conclusion The NRC staff reviewed the containment inspection information provided in the LAR. The NRC staff concludes that containment inspections have been completed consistent with the requirements for PINGP, Units 1 and 2.

3.5 NEI 94-01, Revision 2-A, Limitations and Conditions In the SE issued by the NRC staff dated June 25, 2008, the staff concluded that the methodology in NEI 94-01, Revision 2, is acceptable for referencing by licensees proposing to amend their TS to permanently extend the Type A surveillance test interval to 15 years, subject to the limitations and conditions noted within the SE. Table 20 of the LAR provides a response to each of these limitations and conditions.

Limitation and Condition 1 Limitation and Condition 1 specifies that 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.

Licensees Response to Limitation and Condition 1 The response to this limitation and condition provided in the LAR stated that NSPM will use the definition in Section 5.0 of NEI 94-01, Revision 3-A, and notes that the definition is unchanged from Revision 2 to Revision 3 of the NEI guidance.

NRC Staff Assessment of Licensees Response to Limitation and Condition 1 Section 3.2.9 Type A test performance criterion of ANSI/ANS-56.8-2002 defines the performance leakage rate and reads 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 [upper confidence limit] and the total as-left MNPLR

[minimum pathway leakage rate] of all Type B or Type C pathways isolated during performance of the Type A test.

NRC staff SE Section 3.1.1.1 Enclosure Page 6, for NEI 94-01 Revision 2, reads 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 3-A reads 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 inservice, 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, Revision 2 and Revision 3-A. The NRC staff concluded that the definitions contained in both documents are identical.

Therefore, the NRC staff concludes that PINGP, Units 1 and 2, will use the definition found in Section 5.0 of NEI 94-01, Revision 2, for calculating the Type A leakage rate in the PINGP, Units 1 and 2, Containment Leakage Rate Testing Program.

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

Limitation and Condition 2 Limitation and Condition 2 stipulates that the licensee submit a schedule of containment inspections to be performed prior to and between Type A tests.

Licensees Response to Limitation and Condition 2 The requested schedule was provided in Subsection 3.3.1 of Enclosure 1 to the LAR.

NRC Staff Assessment of Licensees Response to Limitation and Condition 2 The NRC staffs SE Section 3.1.1.3, Enclosure Page 7, for NEI 94-01 Revision 2, reads in part:

NEI TR 94-01, Revision 2, Section 9.2.3.2, states that: 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. NEI TR 94-01, Revision 2, recommends that these inspections be performed in conjunction or coordinated with the examinations required by ASME Code,Section XI, Subsections IWE and IWL. The NRC staff finds that these visual examination provisions, which are consistent with the provisions of regulatory position C.3 of RG 1.163, are acceptable considering the longer 15-year interval. Regulatory Position C.3 of RG 1.163 recommends that such examination be performed at least two more times in the period of 10 years.

The NRC staff agrees that as the Type A test interval is changed to 15 years, the schedule of visual inspections should also be revised. Section 9.2.3.2 in NEI TR 94-01, Revision 2, addresses the supplemental inspection requirements that are acceptable to the NRC staff.

Page 10 of NEI 94-01, Revision 3-A, Section 9.2.1, Pretest Inspection and Test Methodology reads, 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.

Page 12 of NEI 94-01, Revision 3-A, Section 9.2.3.2, Supplemental Inspection Requirements reads:

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.

The NRC staff reviewed LAR Section 3.3, Containment Inspection, and the table contained in LAR Section 3.3.1 Containment Inservice Inspection Program. Table 18 of LAR Section 3.3.1 provides the scheduled dates for 100 percent completion of each required IWE inspection and each pre-ILRT inspection. ASME Code Section XI, Subsection IWL, does not apply to PINGP, Units 1 and 2. Based on this review, the NRC staff has confirmed that the IWE inspection requirements and the pre-ILRT primary containment inspection requirement of NRC staff SE Section 3.1.1.3 for NEI 94-01 Revision 2-A, is satisfied for PINGP, Units 1 and 2.

Summary Based on the above discussion, the NRC staff concludes that the schedule of inspections planned for PINGP, Units 1 and 2, complies with the guidance contained in NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2 and satisfies the provisions contained in the NRC staff SE Section 3.1.1.3. Accordingly, the NRC staff finds that the licensee has adequately addressed Limitation and Condition 2.

Limitation and Condition 3 Limitation and Condition 3 stipulates that the licensee address the areas of the containment structure potentially subjected to degradation.

Licensees Response to Limitation and Condition 3 The LAR stated that general visual observations of the accessible interior and external surfaces of the containment structure will continue to be performed in accordance with ASME Code Section XI, Subsection IWE, and NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2.

NRC Staff Assessment of Licensees Response to Limitation and Condition 3 The NRC staff reviewed the information contained in LAR Section 3.1 Containment Inspection.

The NRC staff SE, Section 3.1.3, Enclosure Page 9, for NEI 94-01, Revision 2 reads, 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).

General visual examinations of the accessible surfaces of containment are performed to assess the general condition of the containment surfaces. In conformance with 10 CFR 50.55a, the current applicable code edition and addenda for the PINGP, Units 1 and 2, third 10-year ISI interval is the 2013 Edition, Subsections IWE. This plan applies to the containment vessel. In particular: IWE deals with Class MC pressure retaining components and their integral attachment; and Class CC metallic shell and penetration liners.

INACCESSIBLE AREAS/AUGMENTED EXAMINATIONS The programmatic requirements for Class MC inaccessible areas of components and structures, as specified in 10 CFR 50.55a(b)(2)(ix)(A) are:

(1) the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.

(2) For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report as required by IWA-6000:

i. A description of the type and estimated extent of degradation, and the conditions that led to the degradation; ii. An evaluation of each area, and the result of the evaluation; and iii. A description of necessary corrective actions.

LAR Section 3.3 indicated that the PINGP, Units 1 and 2, primary containment examinations are performed in accordance with the IWE program and satisfy the general visual examination requirements specified in 10 CFR Part 50, Appendix J, Option B.

LAR Section 3.3.1 Containment Inservice Inspection Program states, in part:

The scope of the program includes the accessible pressure retaining containment surface areas including: Containment vessel surfaces and integral attachments, surfaces requiring augmented examination, mechanical/ electrical penetrations, moisture barriers, pressure retaining bolting and Appendix J tested IWE components.

IWE examination Category E-C includes IWE component areas that were selected for augmented examination because of known existing degraded conditions. Surface areas likely to experience accelerated degradation and aging require augmented examination. In addition, interior containment surfaces that are subject to excessive wear causing a loss of protective coatings, deformation or material loss are also examined. Examination methods are detailed visual examinations (VT-1) and ultrasonic testing (UT). The PINGP Units 1 and 2 do not currently have any areas requiring augmented examination identified.

BELLOWS The PINGP, Units 1 and 2, Type B and Type C testing program consists of LLRT of penetrations with expansion bellows that serve as a barrier to the release of the post-accident containment atmosphere. The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life.

ELECTRICAL PENETRATIONS The Type B and Type C testing program requires testing of electrical penetrations in accordance with 10 CFR Part 50, Appendix J, Option B, and RG 1.163. The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life. Refer to SE Section 3.2 Types B and C Leak Rate Test History for additional details about the monitoring of these electrical penetrations.

BOLTING The licensee performs bolting examinations in accordance with the IWE program. This program satisfies the general visual examination requirements specified in 10 CFR Part 50, Appendix J, Option B. Examination of pressure-retaining bolted connections and evaluation of containment bolting flaws or degradation are performed in accordance with the requirements of 10 CFR 50.55a(b)(ix)(G) and 10 CFR 50.55a(b)(ix)(H).

MOISTURE BARRIERS LAR Section 3.3.1 Containment Inservice Inspection Program states, in part:

Unit 1 - The containment moisture barriers were examined during the 2016 fall outage. There were three indications of lack of adhesion and one indication of tearing which were repaired and accepted by preservice examination.

Unit 2 - The containment moisture barriers were examined during the 2017 fall outage. There were no indications.

CONTAINMENT VESSEL LAR Section 3.3.1 Containment Inservice Inspection Program states, in part:

Unit 1 - Visual Examination of the Containment Vessel was performed during the 2016 Unit 1 fall outage. The containment vessel showed indications of gouges in three locations. The indications were determined accepted by evaluation since they did not reduce the nominal wall thickness of 1.5 inch by more than 10 percent. The indications are thought to be from original construction as the paint is intact with no evidence of repainting and there is no feasible damage mechanism.

Unit 2 - Visual Examination of the Containment Vessel was performed during the 2017 fall outage. The containment vessel showed no indications.

Summary In summary, the NRC staff concludes that based on the information contained in LAR Section 3.3.1, a containment ISI program has been established that satisfies the issues presented SE Section 3.1.3.

Accordingly, the NRC staff finds that the licensee has adequately addressed Limitation and Condition 3.

Limitation and Condition 4 Limitation and Condition 4 specifies that the licensee address any tests and inspections performed following major modifications to the containment structure, as applicable.

Licensees Response to Limitation and Condition 4 The LAR states that no major repairs or modifications have been performed to the PINGP, Units 1 and 2, containment structures. The LAR further states that no major modifications are planned that would affect the PINGP, Units 1 and 2, containment structures.

NRC Staff Assessment of Licensees Response to Limitation and Condition 4 The NRC staff SE Section 3.1.4, Enclosure Page 9, for NEI 94-01 Revision 2, reads, 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 any major modification or maintenance repair of the containment since the last ILRT has been appropriately accompanied by either a structural integrity test (SIT) or ILRT and that any plans for such major modification also includes appropriate pressure testing.

Summary As stated in the licensees response to Limitation and Condition 4 in the LAR, no major repairs or modifications have been performed or are planned to the PINGP, Units 1 and 2, primary containment. The NRC staff notes that by adopting the limitations and conditions specified in NEI 94-01, Revision 2-A as a basis for its 10 CFR Part 50, Appendix J, Option B, program, the licensee is also bound by the incorporated related guidance of SE Section 3.1.4.

Therefore, the NRC staff concludes that the licensee has adequately addressed the issues of SE, Section 3.1.4, and Limitation and Condition 4.

Limitation and Condition 5 Limitation and Condition 5 specifies that 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 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.

Licensees Response to Limitation and Condition 5 The LAR states that NSPM will follow the guidance in Section 9.1 of NEI 94-01, Revision 3-A. In accordance with Section 3.1.1.2 of the NRC staffs SE dated June 25, 2008, NSPM 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, justification for such an extension request will be in accordance with the staff position in RIS 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).

NRC Staff Assessment of Licensees Response to Limitation and Condition 5 The NRC staffs SE Section 3.1.1.2, Enclosure Page 6, for NEI 94-01 Revision 2, reads:

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 LAR stated that NSPM acknowledges and accepts the NRC staff position, as communicated to the nuclear industry in RIS 2008-27. The above passage from SE, Section 3.1.1.2, accurately reflects the RIS NRC staff position.

Summary The LAR indicated that any amendment requesting an extension of the Type A test interval beyond the upper-bound performance-based limit of 15 years should be infrequent and that any requested permission (i.e., for such an extension) will demonstrate to the NRC staff that an unforeseen emergent condition exists.

Based on the above review, the NRC staff finds that the licensee has adequately addressed Limitation and Condition 5.

Limitation and Condition 6 Limitation and Condition 6 specifies that 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 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.

Licensees Response to Limitation and Condition 6 The LAR stated this item is not applicable for PINGP, Units 1 and 2, because the units were not licensed under 10 CFR Part 52.

NRC Staff Assessment of Licensees Response to Limitation and Condition 6 This limitation and condition is applicable for plants licensed under 10 CFR Part 52. The PINGP, Units 1 and 2, licenses were issued under Part 50 and, therefore, this item is not applicable.

Conclusion Related to the Limitations and Conditions Listed in NEI 94-01, Revision 2-A, Section 4.1, of the NRC staffs SE The NRC staff evaluated each of the six limitations and conditions listed above and determined that the LAR adequately addressed all of the limitations and conditions identified in NEI 94-01, Revision 2-A, Section 4.1, of the NRC staffs SE. Therefore, the NRC staff finds it acceptable for PINGP, Units 1 and 2, to adopt the conditions and limitations of NEI 94-01, Revision 2-A, SE, Section 4.1, as part of the implementation documents listed in TS 5.5.14.

3.6 NEI 94-01, Revision 3-A, Conditions The NRC staff published an SE with limitations and conditions for NEI 94-01, Revision 3, by letter dated June 8, 2012. In the SE, 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. The SE was incorporated into Revision 3 and subsequently issued as NEI 94-01, Revision 3-A, on July 31, 2012.

The LAR proposes to use NEI 94-01, Revision 3-A, as the implementation document for the leak rate testing program. Accordingly, PINGP, Units 1 and 2, will be adopting, in part, the testing criteria in ANSI/ANS 56.8-2002 as part of its licensing basis. As stated in NEI 94-01, Revision 3-A, Section 2.0, Purpose and Scope, where technical guidance overlaps between NEI 94-01, Revision 3-A, and ANSI/ANS 56.8-2002, the guidance in NEI 9401, Revision 3-A, takes precedence.

Topical Report Condition 1 The June 8, 2012, NEI 94-01, Revision 3, SE, Section 4.0, Condition 1, stipulates that:

NEI TR 94-01, Revision 3-A, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is

allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g. BWR MSIVs),

and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

Condition 1 identifies three issues that are required to be addressed:

(1) The allowance of an extended interval for Type C LLRTs of 75 months requires that a licensees post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit; (2) A corrective action plan is to be developed to restore the margin to an acceptable level; and (3) Use of the allowed 9-month extension for eligible Type C valves is only allowed for non-routine emergent conditions, but not for valves categorically restricted and other excepted valves.

Licensees Response to Condition 1 The LAR states for item (1) above that the post-outage report shall include the margin between the Type B and Type C minimum pathway leak rate summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.

The LAR states for item (2) above that when the potential leakage understatement adjusted, Type B and Type C minimum pathway leak rate total is greater than the PINGP, Units 1 and 2, administrative leakage summation limit of 0.50 La, but less than the TS limit of 0.60 La, then an analysis and a corrective action plan will be prepared to restore the leakage summation margin to less than the PINGP, Units 1 and 2, administrative leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and the manner of timely corrective action (as deemed appropriate) that best focuses on the prevention of future component leakage performance issues.

The LAR states for item (3) above that the nine-month grace period will only apply to eligible Type C components and only for non-routine emergent conditions for the PINGP, Units 1 and 2.

The LAR further states that such occurrences would be documented in the record of the tests.

NRC Staff Assessment of Licensees Response to Condition 1 The NRC staff has reviewed the requirements of NEI TR 94-01, Revision 3-A against the NSPM response for Issues (1), (2), and (3) of TR Condition 1. The LAR indicated that following approval of the subject amendment, NSPMs actions will be consistent with the guidance of NEI TR 94-01, Revision 3-A. The NRC staff notes that revised guidance contained in Revision 3-A, Section 10.1 Introduction; Section 10.2.3 Corrective Actions; Section 11.3.2 Programmatic

Controls, and Section 12.1 Report Requirements reflects the NRC staffs SE input pertaining to Issues (1), (2), and (3). The NRC staff concludes that for Condition 1, PINGP, Units 1 and 2, will comply with these requirements.

Topical Report Condition 2 The NRC staffs SE dated June 8, 2012, Section 4.0, Condition 2, stipulates that:

The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves which, in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for. Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1.

When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B & C total and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Condition 2 identifies two issues that are required to be addressed:

(1) Extending the Type C LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative, provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1; and

(2) When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the Primary Containment Leakage Rate Testing Program trending or monitoring must include an estimate of the amount of understatement in the Type B and Type C total, and must be included in a licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Licensees Response to Condition 2 The LAR stated for item (1) above that a potential leakage understatement adjustment factor of 1.25 will be applied to the as-left leakage total for each Type C component currently on the greater than 60-month extended test interval. The LAR further stated that when the potential leakage understatement adjusted leak rate total for those Type C components being tested on a greater than 60-month extended interval is summed with the non-adjusted total of those Type C components being tested at less than the 60-month interval and the total of the Type B tested components, if the minimum pathway leak rate is greater than the PINGP, Units 1 and 2, administrative leakage summation limit of 0.50 La, but less than the regulatory limit of 0.60 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the administrative leakage limit. The corrective action plan shall focus on those components that have contributed the most to the increase in the leakage summation value and the manner of timely corrective action (as deemed appropriate) that best focuses on the prevention of future component leakage performance issues.

The LAR stated for item (2) above that if the potential leakage understatement adjusted minimum pathway leak rate is less than the administrative leakage summation limit of 0.50 La, then the acceptability of the 75-month local leak rate test extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

NRC Staff Assessment of Licensees Response to Condition 2 The NRC staff has reviewed the requirements of NEI TR 94-01, Revision 3-A, against the NSPM response to issues (1) and (2) of TR Condition 2. The LAR indicated that following approval of the subject amendment, NSPMs actions will be consistent with the guidance of NEI TR 94-01, Revision 3-A. The NRC staff concludes that for Condition 2, PINGP, Units 1 and 2, will comply with these requirements.

NEI 94-01, Revision 3-A, SE Section 4.0, Conclusion Based on the above evaluation of each condition, the NRC staff determined that the licensee has adequately addressed both conditions in Section 4.0 of the NRC staffs SE for TR NEI 94-01, Revision 3. Therefore, the NRC staff finds it acceptable for NSPM to adopt TR NEI 94-01, Revision 3-A, as the implementation document in TS 5.5.14 Containment Leakage Rate Testing Program for PINGP.

3.7 Probabilistic Risk Assessment of the Proposed Extension of the ILRT Test Intervals The LAR provided a plant specific risk assessment for permanently extending the currently allowed containment Type A ILRT interval from 10 years to 15 years in Enclosure 2.

The LAR 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 LAR addressed each of the four conditions for the use of EPRI TR-1009325, Revision 2, which are listed in Section 4.2 of the associated NRC staff safety evaluation report (SER). The four conditions that are being evaluated are 1) PRA Quality (referred to as PRA Adequacy),

2) Estimated Risk Increases, 3) Leak Rate for the Large Pre-Existing Containment Leak Rate Case, and 4) Determining if Containment Overpressure is Relied Upon for ECCS Performance.

A summary of how each condition is met is provided in Sections 3.7.1 through 3.7.4 below.

3.7.1 PRA Quality - Condition 1 The first condition in Section 4.2 of the SER for EPRI TR 1009325, Revision 2, stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the RG 1.200 relevant to the ILRT extension application. This regulatory guide 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.

Consistent with the information provided in RIS 2007-06, Regulatory Guide 1.200 Implementation, the NRC staff will use Revision 2 of RG 1.200 to assess technical adequacy of the PRA used to support risk informed applications received after March 2010. In 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.

The PINGP, Units 1 and 2, PRA technical adequacy is addressed in Section 3.6 of the LAR and , Prairie Island Nuclear Generating Plant: Evaluation of Risk Significance of Permanent ILRT Extension. As discussed in Enclosure 1, Section 3.6 of the LAR, the PINGP, Units 1 and 2, risk assessment performed to support the ILRT application used Revision 5.3 (the most recent evaluation of internal event risk) of the PINGP, Units 1 and 2, PRA model. The LAR described the PINGP, Units 1 and 2, PRA model being assessed against RG 1.200 and describes a November 2010 internal PRA model that was subjected to a peer review in accordance with RG 1.200. The LAR stated that a September 2012 focused peer review was conducted on internal flooding events for the internal events PRA model. The LAR also discussed that additional peer review of the internal events PRA model was conducted during May 2014. This peer review evaluated the changes to their PRA model to address incorporation of the FLOWserve N9000 Reactor Coolant Pump seals against the guidance in RG 1.200.

The LAR described that the PINGP fire PRA model was subjected to a full scope peer review in May of 2012. The LAR also described two focused scope peer reviews conducted in November 2013 and December 2017.

The LAR described the results of the October 2017 finding closure reviews for internal events, including internal flooding and fire PRA models. One of the outstanding findings was resolved during a focused fire PRA review. An additional closure review was conducted on April 31 to May 1, 2019, in which five additional findings were determined to be closed. The closure review did not address the finding SY-A17-01, which remains open for the licensee. The licensee provides the details of SY-A17-01 in section A.2 of enclosure 2. To address this finding, the LAR described the small increase in risk when assuming actions to trip reactor coolant pumps was required but does not occur in Section 5.3.3 of enclosure 2 of the LAR. The LAR also stated that the small risk increase resulted in a negligible source of uncertainty for the PRA model, and does not significantly affect the plant CDF. The NRC staff reviewed the licensees justification and agrees with the licensees assessment that this small increase in risk has negligible impact on the PINGP, Units 1 and 2, CDF and is not risk significant.

Based on its review of the above information, the NRC staff finds 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 PRA models used by the licensee are of sufficient quality to support the evaluation of changes to ILRT frequencies.

Accordingly, the first condition is met.

3.7.2 Estimated Risk Increase - Condition 2 The second condition in Section 4.2 of the SER for EPRI TR 1009325, Revision 2, stipulates that the licensee submit documentation indicating that the estimated risk increase associated with permanently extending the ILRT interval to 15 years is small, consistent with the guidance in RG 1.174 and the clarification provided in the NRC staffs 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.

Lastly, for plants that rely on containment over-pressure for net positive suction head (NPSH) 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. This last point is not applicable given that PINGP, Units 1 and 2, does not rely on containment overpressure for ECCS performance. RG 1.174 defines very small changes in risk as resulting in increases of CDF and LERF of less than 1.0E-6/year and 1.0E-07/year respectively. Thus, the associated risk metrics include LERF, population dose and CCFP.

The LAR reported the results of the plant-specific risk assessment in Section 3.6.3 of and provided the details of the calculations in LAR Enclosure 2, Section 5.2. The LAR stated that 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 to be 1.17E-7/year for PINGP, Unit 1, and 1.13E-7/year for PINGP, Unit 2, using the EPRI guidance. Containment overpressure is not required in support of ECCS performance to mitigate design basis accidents and there is no equipment in the shield building being credited in the CDF model at PINGP, Units 1 and 2. As a result, the LAR stated that the ILRT change does not impact CDF, so the relevant criterion is

LERF. The total internal events LERF is 3.32E-7 and 2.99E-7 for PINGP, Units 1 and 2, respectively. Given RG 1.174s criteria for a small change to LERF being greater than 1.0E-7/year but less than 1.0E-6/year, the LAR concluded that the increase is small. Also noted in the LAR is the LERF change for a Type A ILRT interval from 1 in 10 years to 1 in 15 years. The resulting LERF values of 4.86E-8/year for PINGP, Unit 1 and 4.72E-8/year for PINGP, Unit 2 are provided and noted to be bounded by the previous interval change.

Including external event risk, the increase in LERF for a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated to be 7.38E-7/year and 7.33E-7/year for PINGP Units 1 and 2, respectively. Given that the change in LERF is greater than 1.0E-7/year and less than 1.0E-6/year, the LAR categorized the change in risk as small. Also noted in the LAR is the LERF change for a Type A ILRT interval from 1 in 10 years to 1 in 15 years. The resulting LERF values of 3.08E-7/year for PINGP, Unit 1 and 3.05E-7/year for PINGP, Unit 2, are provided and noted to be bounded by the previous interval change. Given the licensees analysis showing an increase of greater than 1.0E-7/year and less than 1.0E-6/year for PINGP, Units 1 and 2, the increase to LERF is small.

The LAR provided the resulting population doses for the Type A test frequency change to 1 per 15 years to be 0.028 person-rem/year for PINGP, Unit 1, and 0.027 person-rem/year for PINGP, Unit 2. The guidance contained within NEI 94-01 states that a small total population dose is defined as an increase of 1.0 person-rem/year, or 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals.

The reported increase in total population dose is below the acceptance criteria provided in EPRI TR-1009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC staffs SER for NEI 94-01, Revision 2. Given the analysis showing an increase of less than one person-rem/year for PINGP, Units 1 and 2, the increase to population dose is small.

The CCFP is discussed in LAR Section 5.2.5 of enclosure 2. The increase in the CCFP due to the change in test frequency from three in 10 years to once in 15 years is 0.908 percent for PINGP, Unit 1 and 0.910 percent for PINGP, Unit 2. NEI 94-01 states that increases in CCFP of 1.5 percent is considered small. The values provided in the LAR are below the acceptance guidelines in Section 3.2.4.6 of the NRC staffs SER for NEI 94-01.

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 the increase in the total population dose and the magnitude of the change in the 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. Accordingly, the second condition is met.

3.7.3 Leak Rate for the Large Pre-Existing Containment Leak Rate Case - Condition 3 The third condition in Section 4.2 of the SER for EPRI TR 1009325, Revision 2, stipulates that to make the methodology acceptable, the average leak rate for the pre-existing containment large leak rate accident case (i.e., accident case 3b) used by the licensees shall be 100 La instead of 35 La. As noted in LAR Section 3.5.1 of Enclosure 1, the methodology in EPRI TR-1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has

been used in the PINGP, Units 1 and 2, risk assessments. Accordingly, the third condition is met.

3.7.4 Containment Overpressure is Relied Upon for ECCS Performance - Condition 4 The fourth condition in Section 4.2 of the SER for EPRI TR 1009325, Revision 2, stipulates that in instances where containment over-pressure is relied upon for ECCS performance, a LAR is required to be submitted. Section 3.5.1 of Enclosure 1 of the LAR stated that PINGP, Units 1 and 2, do not rely on containment overpressure for ECCS performance. Thus, no reliance is placed on pressure and/or temperature transients to ensure adequate net-positive suction head.

Accordingly, the fourth condition is not applicable.

3.7.5 Seismic PRA The LAR stated that a seismic PRA model is not maintained for PINGP, Units 1 and 2. NRC Generic Letter 199 (GI-199), Appendix D: Seismic Core-Damage Frequencies, provides the seismic core- damage frequency estimates developed in the Safety/Risk Assessment. The LAR stated that the 2014 Seismic Reevaluations for operating reactor sites found in, The Nuclear Energy Institute - Seismic Risk Evaluations for Plants in the Central and Eastern United States (ADAMS Accession No. ML14083A596), states that the conclusions made by GI-199 remain valid for estimating Seismic CDF at plants. Table D-1 of GI-199 provides seismic CDFs using 2008 USGS (United States Geological Survey) Seismic Hazard Curves. The LAR claimed to use the average SCDF (seismic core damage frequency} of 1.84E-6 from Table D-1.

Since no Seismic LERF value is calculated, the licensee assumes the LERF/CDF ratio will be similar for the seismic risk as used for internal event risk and calculates seismic LERF values of 3.07E-8 for PINGP, Unit 1, and 2.74E-8 for PINPG, Unit 2. RG 1.174 defines very small changes in risk as resulting in increases of CDF and LERF of less than 1.0E-6/year and 1.0E-07/year respectively. The chosen seismic CDF is judged to be sufficient to support PINGP, Units 1 and 2, ILRT external events risk impact assessment because the licensee demonstrated that the chosen seismic LERF contribution to the total external event results in a very small impact as described by the criteria in RG 1.174.

3.8 Technical Conclusion Based on the preceding regulatory and technical evaluations, the NRC staff finds that the exception in TS 5.5.14 related to previously completed testing is no longer relevant and therefore its deletion is acceptable. Further, the NRC staff finds that the licensee has adequately implemented its existing primary containment leakage rate testing program consisting of ILRT and LLRT. The results of the recent ILRTs and of the LLRTs combined totals demonstrate acceptable performance and support a conclusion that the structural and leak-tight integrity 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 the licensee has addressed the NRC conditions to demonstrate acceptability of adopting NEI 94-01, Revision 3-A, and the limitations and conditions identified in the NRC staffs SE incorporated in NEI 94-01, Revision 2-A. Therefore, the NRC staff finds that the proposed changes to PINGP TS 5.5.14 regarding the primary containment leakage rate testing program are acceptable.

Lastly, the NRC staff finds that the licensee has adequately addressed the four conditions for the use of EPRI TR-1009325, Revision 2, by addressing PRA quality, demonstrated through calculation the small increases in LERF, used the 100 La instead of 35 La for average leak rates, and addressed containment overpressure by not relying on overpressure for ECCS

performance. Accordingly, the NRC staff finds the PRA assessment for the extension of the ILRT intervals acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations on, the Minnesota State official was notified of the proposed issuance of the amendments on August 3, 2020. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change the requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration and there has been no public comment on such finding (84 FR 66234). Accordingly, the amendments meet 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 amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: Zachary Gran Brian Lee George Wang Date of Issuance: October 2, 2020

ML20217L185 *via e-mail OFFICE NRR/DORL/LPL3/PM* NRR/DORL/LPL3/LA* NRR/DSS/SCPB/BC* NRR/DSS/STSB/BC*

NAME RKuntz SRohrer BWittick VCusumano DATE 08/04/2020 08/05/2020 07/21/2020 08/05/2020 OFFICE NRR/DEX/ESEB/BC* NRR/DEX/EMIB/BC* NRR/DRA/ARCB/BC* OGC NLO*

NAME SKrepel ABuford KHsueh STurk DATE 04/01/2020 08/04/2020 08/25/2020 09/09/2020 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM*

NAME NSalgado (SWall for) RKuntz DATE 10/02/2020 10/02/2020