ML20161A000

Issuance of Amendment No. 276 Regarding Revision to the Integrated Leak Rate Type a and Type C Test Intervals

ML20161A000
Person / Time
Site:	Millstone
Issue date:	07/15/2020
From:	Richard Guzman NRC/NRR/DORL/LPL1
To:	Stoddard D Dominion Energy Nuclear Connecticut
Guzman R
References
EPID L-2019-LLA-0165
Download: ML20161A000 (35)

Text

July 15, 2020 Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Dominion Nuclear Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

MILLSTONE POWER STATION, UNIT NO. 3 ISSUANCE OF AMENDMENT NO. 276 REGARDING REVISION TO THE INTEGRATED LEAK RATE TYPE A AND TYPE C TEST INTERVALS (EPID L-2019-LLA-0165)

Dear Mr. Stoddard:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 276 to Renewed Facility Operating License No. NPF-49 for the Millstone Power Station, Unit No. 3 (Millstone 3), in response to your application dated July 30, 2019.

The amendment revises Millstone 3 Technical Specification (TS) 6.8.4.f, Containment Leakage Rate Testing Program, by replacing the reference to Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, with a reference to Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and the conditions and limitations specified in NEI 94-01, Revision 2-A. The amendment allows Dominion Energy Nuclear Connecticut, Inc. to extend the Type A primary containment integrated leak rate test interval for Millstone 3 from 10 years to 15 years and the Type C local leak rate test interval from 60 months to 75 months, and incorporates the regulatory positions stated in Regulatory Guide 1.163.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.

Sincerely,

/RA/

Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-423

Enclosures:

1. Amendment No. 276 to NFP-49

2. Safety Evaluation cc: Listserv

DOMINION ENERGY NUCLEAR CONNECTICUT, INC., ET AL DOCKET NO. 50-423 MILLSTONE POWER STATION, UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 276 Renewed License No. NPF-49

1.

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

A.

The application for amendment by Dominion Energy Nuclear Connecticut, Inc.

(the licensee) dated July 30, 3019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations, and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-49 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 276 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: July 15, 2020 James G.

Danna Digitally signed by James G. Danna Date: 2020.07.15 16:12:53 -04'00'

ATTACHMENT TO LICENSE AMENDMENT NO. 276 MILLSTONE POWER STATION, UNIT NO. 3 RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOCKET NO. 50-423 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 4

4 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Insert 6-17 6-17 (2)

Technical Specifications The Technical Specifications contained in Appendix A, revised through Amendment No. 276 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated into the license. DENC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3)

DENC shall not take any action that would cause Dominion Energy, Inc.

or its parent companies to void, cancel, or diminish DENCs Commitment to have sufficient funds available to fund an extended plant shutdown as represented in the application for approval of the transfer of the licenses for MPS Unit No. 3.

(4)

Immediately after the transfer of interests in MPS Unit No. 3 to DNC*, the amount in the decommissioning trust fund for MPS Unit No. 3 must, with respect to the interest in MPS Unit No. 3, that DNC* would then hold, be at a level no less than the formula amount under 10 CFR 50.75.

(5)

The decommissioning trust agreement for MPS Unit No. 3 at the time the transfer of the unit to DNC* is effected and thereafter is subject to the following:

(a)

The decommissioning trust agreement must be in a form acceptable to the NRC.

(b)

With respect to the decommissioning trust fund, investments in the securities or other obligations of Dominion Energy, Inc. or its affiliates or subsidiaries, successors, or assigns are prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are prohibited.

(c)

The decommissioning trust agreement for MPS Unit No. 3 must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee until the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The decommissioning trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(d)

The decommissioning trust agreement must provide that the agreement cannot be amended in any material respect without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

• On May 12, 2017, the name Dominion Nuclear Connecticut, Inc. changed to Dominion Energy Nuclear Connecticut, Inc.

Renewed License No. NPF-49 Amendment No. 270, 271, 272, 273, 275, 276

MILLSTONE - UNIT 3 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 2)

Pre-planned operating procedures and backup instrumentation to be used if one or more monitoring instruments become inoperable, and 3)

Administrative procedures for returning inoperable instruments to OPERABLE status as soon as practicable.

f.

Containment Leakage Rate Testing Program 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, Revision 3-A Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 2012 and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 41.9 psig.

The maximum allowable containment leakage rate La, at Pa, shall be 0.3 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 />.

Leakage rate acceptance criteria are:

1)

Containment overall leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and 0.06 La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests; 2)

Air lock testing acceptance criteria are:

a.

Overall air lock leakage rate is 0.05 La when tested at Pa.

b.

For each door, seal leakage rate is < 0.01 La when pressurized to Pa.

The provisions of Specification 4.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of Specification 4.0.3 are applicable to the Containment Leakage Rate Testing Program.

An exemption to Appendix J, Option A, paragraph III.D.2(b)(ii), of 10 CFR Part 50, as approved by the NRC on December 6, 1985.

Amendment No. 69, 186, 232, 239, 242, 259, 6-17 276

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 276 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION, UNIT NO. 3 DOCKET NO. 50-423

1.0 INTRODUCTION

By letter dated July 30, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19217A208), Dominion Energy Nuclear Connecticut, Inc. (DENC, the licensee), submitted a license amendment request (LAR) to revise the Technical Specifications (TSs) for Millstone Power Station, Unit No. 3 (MPS3).

The amendment would revise MPS3 TS 6.8.4.f, Containment Leakage Rate Testing Program, by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995 (ADAMS Accession No. ML003740058), with a reference to Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (ADAMS Accession No. ML12221A202), and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated November 19, 2008 (ADAMS Accession No. ML100620847).

The revision would allow DENC to extend the Type A primary containment integrated leak rate test interval (ILRT) for MPS3 from 10 years to 15 years and the Type C local leak rate test (LLRT) interval from 60 months to 75 months and incorporate the regulatory positions stated in RG 1.163.

2.0 REGULATORY EVALUATION

2.1 Description of Containment In Section 4.1, Description of Containment, of Attachment 1 to the LAR dated July 30, 2019, the licensee provided the following as a description of the MPS3 containment:

The MPS3 containment structure is a steel-lined, conventionally reinforced concrete structure designed to operate under sub-atmospheric conditions. The structure has a vertical cylindrical wall and hemispherical dome supported on a flat base mat which is founded on bedrock.

The base mat or foundation slab is 10 [feet] ft. thick with a diameter of 158 ft.

The floor liner plate is 0.25 [inch] in. thick. A reinforced concrete slab approximately 2 ft. thick was placed over and anchored through the mat liner to stiffen it against negative pressures and to protect it from heat associated with a design basis accident. This slab also serves as anchorage and support for equipment located on the lowest level of the containment.

The cylindrical wall is 4 ft. 6 in. thick with an inside diameter of 140 ft. and a height from mat to spring line of 131 ft. 3 in. The inside radius of the 2 ft. 6 in.

thick dome is 70 ft. The internal height from base mat to the center of the dome is 201 ft. 3 in. The dome liner plate is 0.5 in. thick.

The liner plate is a continuously welded steel membrane supported by and anchored to the inside of the containment at sufficiently close intervals with anchor studs and deformed bars so that the overall deformation of the liner under the parameters derived from the design basis accident (DBA) and normal operation is essentially the same as that of the concrete containment structure.

The function of the liner is to act as a gas-tight membrane under conditions that can be encountered throughout the operating life of the plant. The liner is designed to resist all direct loads and accommodate deformation of the concrete containment structure without jeopardizing leak-tight integrity.

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

The testing requirements in 10 CFR Part 50, 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. MPS3 has voluntarily adopted and has implemented Option B for meeting the requirements of 10 CFR Part 50, Appendix J.

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

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

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

The NRC staffs final safety evaluation (SE) for NEI 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and Electric Power Research Institute (EPRI) TR-1009325, Revision 2, dated August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated June 25, 2008 (ADAMS Accession No. ML081140105) was incorporated into NEI 94-01, Revision 2-A, dated November 19, 2008. NEI 94-01, Revision 2-A, describes an NRC-approved approach for implementing the optional performance-based requirements described in 10 CFR Part 50, Appendix J, Option B, which includes provisions for extending Type A ILRT intervals for up to 15 years, and incorporates the regulatory positions stated in RG 1.163. NEI 94-01, Revision 2-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. This method uses industry performance and plant-specific data and risk insights in determining the appropriate testing frequency, and also discusses the performance factors that licensees must consider in determining test intervals. NEI 94-01, Revision 2-A, includes six specific limitations and conditions listed in Section 4.1 of the SE.

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

NEI 94-01, Revision 3-A, documents the NRCs evaluation and acceptance of NEI 94-01, Revision 3, and includes two specific limitations and conditions listed in Section 4.0 of the SE.

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

The regulations in 10 CFR 50.65(a), Requirements for monitoring the effectiveness of maintenance at nuclear power plants, require that the licensee shall monitor the performance or condition of structures, systems, or components against licensee-established goals in a manner sufficient to provide reasonable assurance that these structures, systems, and components, as defined in paragraph (b) of this section, are capable of fulfilling their intended functions. These goals shall be established commensurate with safety and, where practical, take into account industrywide operating experience.

The regulations in 10 CFR 50.36, Technical specifications, state that the TSs include items in five specific categories. These categories include: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting condition for operations; (3) surveillance requirements; (4) design features; and (5) administrative controls. NUREG-1431, Standard Technical Specifications - Westinghouse Plants, Revision 4.0, Volume 1, Specifications (ADAMS Accession No. ML12100A222), incorporates Standard Technical Specifications Task Force (TSTF) Traveler TSTF-52, Revision 3, Implement 10 CFR 50, Appendix J, Option B (ADAMS Accession No. ML040400371), which includes guidance for specific changes to TSs for implementation of 10 CFR Part 50, Appendix J, Option B.

A Type A test is an overall ILRT of the containment structure. NEI 94-01, Revision 0 (ADAMS Accession No. ML11327A025), specifies an initial test interval of 48 months, but allows an extended interval of 10 years based upon two consecutive, successful tests. There is also a provision for extending the test interval an additional 15 months, but this should be used only in cases where refueling schedules have been changed to accommodate other factors.

Amendment No. 239 to Facility Operating License No. NPF-49 for MPS3, dated June 29, 2007 (ADAMS Accession No. ML071690523), allowed a one-time extension of the ILRT interval to 15 years. However, subsequent to this one-time extension, the long-term ILRT test interval requirement in MPS3 TS 6.8.4.f remained at 10 years.

Guidance for extending Type A ILRT surveillance intervals beyond 10 years is provided in Section 9.2.3, Extended Test Intervals, of NEI 94-01, Revision 3-A.

Guidance for extending Type C test LLRT surveillance intervals beyond 60 months is provided in Section 10.2.3, Type C Test Interval, of NEI 94-01, Revision 3-A.

The Type A and the combined Type B and Type C test results must not exceed the La with margin, as specified in MPS3 TS 6.8.4.f(1). Option B also 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 leaktight integrity, be conducted prior to each Type A test and at a periodic interval between tests based on the performance of the containment system.

DENC proposes to extend the MPS3 interval for the containment ILRT to 15 years from the last ILRT. The last MPS3 ILRT was performed on November 8, 2011; therefore, the next ILRT for MPS3 is due no later than November 8, 2021. Using the proposed interval of 15 years, the next ILRT for MPS3 would need to be completed by November 8, 2026.

3.0 TECHNICAL EVALUATION

3.1 Licensees Proposed Changes The first paragraph of MPS3 TS 6.8.4.f, Containment Leakage Rate Testing Program, currently states:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, Industry Performance-Based Option of 10 CFR Part 50, Appendix J: The first Type A test performed after the January 6, 1998 Type A test shall be performed no later than January 6, 2013.

The proposed change would revise TS 6.8.4.f by deleting the exception to NEI 94-01, Revision 0, and replacing the reference to RG 1.163 with a reference to NEI 94-01, Revision 3-A, and the limitations and conditions specified in NEI 94-01, Revision 2-A.

Specifically, the first paragraph of TS 6.8.4.f would be changed to read as follows:

A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B as modified by approved exemptions*. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 2012 and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008.

3.2 Deterministic Considerations: Structural and Leaktight Integrity of the Containment The licensee proposed to revise TS 6.8.4.f by replacing the 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. Versions of NEI 94-01 would provide for:

Adopting the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, Containment System Leakage Testing Requirements, and Adopting a more conservative grace interval of 9 months, for Type A, Type B, and Type C leakage tests in accordance with NEI 94-01, Revision 3-A.

With respect to the deletion of the exception currently contained in MPS3 TS 6.8.4.f, the NRC staff concurs with the licensee that the deletion of the one-time exception to NEI 94-01, Revision 0, Section 9.2.3, would be appropriate with the staffs approval of the subject LAR dated July 30, 2019. In particular, this exception states, in part, The first Type A test performed after the January 6, 1998 Type A test shall be performed no later than January 6, 2013. As indicated in LAR Section 4.2, the first MPS3 Type A (ILRT) test after the 1998 ILRT was completed on November 7, 2011. Accordingly, the NRC staff approves the deletion of this exception associated with MPS3 TS 6.8.4.f.

Consistent with the guidance contained in both NEI 94-01, Revision 2-A, and NEI 94-01, Revision 3-A, the licensee justified the proposed changes by demonstrating adequate performance of the MPS3 containment based on (a) the historical plant-specific containment leakage testing program results, (b) the CISI program results, and (c) an MPS3 plant-specific risk assessment.

The NRC staff reviewed the MPS3 LAR from the perspective of deterministic considerations regarding containment leaktight integrity. The NRC staffs review and analysis of these changes is conveyed in Sections 3.2.1 and 3.2.2 below.

3.2.1 Type A Integrated Leak Rate Test History Per TS 6.8.4.f, the MPS3 containment was designed for a maximum allowable containment leakage rate La of 0.3 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 the calculated peak pressure, Pa. TS 6.8.4.f indicates that the peak calculated containment internal pressure for the design-basis loss-of-coolant accident, Pa, is 41.9 pounds per square inch gauge (psig).

Since October 1993, a total of three ILRTs have been performed on the MPS3 containment.

These three Unit 3 ILRTs all had satisfactory leakage rate results. However, the ILRT in 1998 was performed following an extended unit shutdown, and the as-found Type A test was considered a failure due to degraded valve (Type C) performance of containment purge supply (penetration 86) identified in LER 96-012-00, B15752, dated June 14, 1996 (ADAMS Accession No. ML20117M216). The test results of these three ILRTs are documented in LAR Section 4.2.

The three test results are summarized in Table 3.1 below:

Table 1 MPS3 Type A ILRT History1 Test Date As-Found Leakage (wt%/day)

Upper 95%

Confidence Level (wt%/day)

Line Up Penalties (wt%/day)

Leakage Savings (wt%/day)

As-Found Corrected Results (wt%/day)

ILRT Acceptance Criteria(2), La (wt%/day)

As-Left Corrected Results (wt%/day)

October 1993 0.0942 0.1311 0.0002 0.0014 0.1339 0.3 0.1333 October 1998 0.0601

>0.3 0.0061

>0.3

>0.3 0.3 0.1158 November 2011 0.0435 0.0462 0.0044 0.0025 0.0531 0.3 0.0506 (1) Data source: LAR Section 4.2 (2) Per TS 6.8.4.f Section 9.1.2 of NEI 94-01, Revision 3-A, states, in part, The 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. The NRC staff confirmed that the requirement of NEI 94-01, Section 9.1.2, has been satisfied based on the test dates shown in Table 1.

TS 6.8.4.f references RG 1.163. Regulatory Position C of RG 1.163 states that NEI 94-01, Revision 0, provides methods acceptable to the NRC staff for complying with the provisions of 10 CFR Part 50, Appendix J, Option B. The third paragraph of Section 9.2.3, Extended Test Intervals, of NEI 94-01, Revision 0, states, 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-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 UCL [Upper Confidence Limit] (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, is nearly identical except the test standard invoked is ANSI/ANS-56.8-2002. The NRC staff notes that NEI 94-01, Revision 3-A, 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 staff also notes that the MPS3 ILRT results since October 1993 demonstrated ample margin (i.e., greater than 68 percent) between each as-found ILRT value and La. Therefore, the staff did not request that the licensee reconstitute the MPS3 Type A test results from before the ILRT of October 1998.

The requirement of TS 6.8.4.f.1 (i.e., leakage rate acceptance criteria) establishes the maximum limit for the MPS3 as-left leakage rate for unit startup following completion of Type A testing at less than or equal to () 0.75 La, which equals 0.225 percent of containment air weight per day.

The MPS3 containment was designed for a leakage rate La not to exceed 0.3 percent by weight of containment air per day at the calculated peak pressure, Pa. As shown in Table 1 above, there has been adequate margin to the performance limit as described in TS 6.8.4.f for La for the historical ILRTs spanning a period of 18 years.

The past three MPS3 ILRT results dating back to 1993 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 MPS3 had as found test results of less than 1.0La, 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, are acceptable for MPS3. The NRC staff finds that the last two MPS3 ILRT test results satisfy the requirements of Sections 9.1.2 of NEI 94-01, Revision 3-A.

3.2.2 Type B and Type C Leak Rate Test History MPS3 TS 6.8.4.f, Containment Leakage Rate Testing Program, states, in part:

Leakage rate acceptance criteria are:

1)

Containment overall leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the combined Type B and Type C tests, and 0.06La for all penetrations that are Secondary Containment bypass leakage paths, and < 0.75 La for Type A tests; The NRC staff reviewed the local leak rate summaries contained in LAR Section 4.3, Type B and Type C Testing. The licensee indicated in the LAR that For MPS3, the combined Type B and Type C leakage rate acceptance criterion is 0.60 La or 307,520 standard cubic centimeters per minute (sccm).

NEI 94-01, Revision 3-A, Section 10.2, indicates that this criterion is to be evaluated to the combined Type B and Type C as-found minimum pathway test total and as a restart permissive criterion to the combined Type B and Type C as-left maximum pathway test total.

With the use of these La values and the data contained in LAR Section 4.3, the NRC staff confirmed the accuracy of the fraction of 0.6La values contained in the LAR and concluded that (1) the MPS3 as-found minimum pathway leakage rates for the last three refueling outages since 2016 have an average of 6.622 percent of 0.6 La with a high of 8.076 percent 0.6 La, and (2) the MPS3 as-left maximum pathway leakage rates for the last three refueling outages since 2016 have an average of 16.068 percent of 0.6 La with a high of 21.470 percent 0.6 La.

As indicated in LAR Section 4.3, the Type B or Type C penetration test failures during the three most recent refueling outages 3R17, 3R18, or 3R19 are described below:

During 3R19, the Chilled Water Return (penetration 72) inside containment isolation valve, 3CDS*CTV91B, failed its LLRT due to an extruded T-ring seat seal. The valve was disassembled and repaired and the subsequent leakage rate was satisfactory. The redundant outside containment isolation valve, 3CDS*CTV38B, was tested and found leak tight, maintaining minimum pathway leakage criteria. Leakage rate testing is performed each refueling outage until compliance with leakage limits is restored.

During 3R18, the inside containment isolation check valve 3CHS*V58 failed its LLRT, requiring disassembly and inspection. Leakage rate testing was performed until compliance with leakage limits was restored.

During 3R17, containment penetration 38, Chilled Water Return, exhibited both inside valve (3CDS*CTV91A) and outside valve (3CDS*CTV38A) leakage rates in excess of 0.60La for the penetration. As a result, repairs were made and leakage rate testing was performed until compliance with leakage limits was restored.

The NRC staff notes that this course of action is consistent with the guidance in NEI 94-01, Revision 0, Section 10.2.3, Type C Test Interval.

MPS3 has a total of 83 Type B tested components, including 80 electrical penetrations. All electrical penetrations are on the extended test interval and are tested within the 120-month performance-based test interval. The air lock and manway seals, fuel transfer tube, and containment equipment hatch and manway are tested every refueling outage. Air lock door gaskets are tested following containment entries during power operation.

The Type C tested components at MPS3 consist of 67 penetrations and 172 components. Five valves are currently on a refueling frequency for testing, whereas 97 percent of all Type C tested components are on the extended interval. The containment purge and exhaust penetrations are not on the performance-based extended interval and are, therefore, tested on a refueling frequency. Approximately 50 percent of the Type C penetrations are tested each refueling outage due to scheduled maintenance and train-specific alignments.

Based on the NRC staffs review of the historical information provided in LAR Section 4.3, Type B and Type C Testing, the staff observed that there was no indication of the licensees failure to adequately implement the requirements of its 10 CFR Part 50, Appendix J, Option B performance-based testing program.

In summary, the licensee provided an adequate explanation about the cause of failure for the LLRT Type C penetration failures experienced since refueling outage 3R17 (i.e., April 2016).

Furthermore, based on the review of LAR Section 4.3, the NRC staff concludes that the aggregate leakage rate results of the as-found minimum pathway for all MPS3 Type B and Type C tests from 2016 through 2019 demonstrate 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 6.8.4.f.1. Therefore, the NRC staff has reasonable assurance that the licensee is compliant with the guidance of both Section 10.2.1, Type B Test Intervals, and Section 10.2.3, Type C Test Interval, of NEI 94-01, Revision 0.

Based on the above, the NRC staff concludes that (1) the licensee has been compliant with the guidance of RG 1.163 and NEI 94-01, Revision 0; (2) the recent historical cumulative Type B and Type C test results are substantially below the acceptance limit of TS 6.8.4.f.1; and (3) the licensee has a corrective action program that appropriately addresses poor performing valves and penetrations. Therefore, the NRC staff finds that the licensee is effectively implementing the MPS3 Type B and Type C leakage rate test program as required by to 10 CFR Part 50, Appendix J, Option B.

3.2.3 Containment Inservice Inspection Program Previous Type A ILRT Results In its LAR, the licensee states that MPS3 TS 6.8.4.f currently requires Type A, Type B, and Type C testing in accordance with RG 1.163, which endorses the methodology for complying with 10 CFR Part 50, Appendix J, Option B. Since the adoption of Option B, the performance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1, for Type A testing.

The following table was provided in the LAR, which lists the past periodic Type A ILRT results for MPS3.

Table 2 Type A - Integrated Leakage Rate Testing History Test Date As Found (wt.%/day)

As Left (wt.%/day)

Maximum Allowable (TS limit)

(wt.%/day)

October 10, 1993 0.1339 0.1333 0.1339 < 0.3 or 45%

October 28, 1998

> 0.3 0.1158

> 0.3 November 7, 2011 0.0531 0.0506 0.0531 < 0.3 or 18%

Containment Inservice Inspection Program In its LAR, the licensee states that prior to initiating a Type A test, a general visual examination 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 is performed. This inspection is typically conducted in accordance with the Millstone CISI plan, which implements the requirements of American Society of Mechanical Engineers (ASME) Code,Section XI, Subsection IWE/IWL. The 2001 Edition with the 2003 Addenda is the current applicable code edition and addenda for the second 10-year IWE/IWL program interval.

IWE Examinations In its LAR, the licensee states that MPS3 has completed the examination requirements of Interval 2, Period 3 of the containment IWE inservice inspection program. Examinations were performed to the requirements of the 2001 Edition through 2003 Addenda of ASME Section XI, as modified by the 10 CFR 50.55a(b) limitations. At this time, no augmented Category E-C examinations are planned for MPS3.

The licensee also stated that a complete inspection of all accessible containment liner surface areas is performed each period. In addition, each refueling outage, a trained coating inspector specifically examines the containment liner coating, independent of the IWE examination.

There is no active degradation mechanism present. Coating degradation of the liner has been primarily the result of mechanical damage incurred during outages. To prevent further mechanical damage, bubble wrap is placed on the outer annulus walls near staged equipment and high traffic areas to prevent interaction with the liner. There are no primary containment surface areas that require augmented examination in accordance with ASME Section XI, IWE-1240. However, each period, 100 percent of the accessible primary containment surface area is inspected. Any significant changes or potential concerns receive a detailed inspection at that time.

IWL Examinations The licensee states that the second 10-year interval of concrete containment examinations (IWL) has been performed for MPS3. In accordance with Category L-A of the 2001 Edition with 2003 Addenda of ASME Section XI, general and detailed visual examinations were completed by the required due date (March 28, 2016). In accordance with Category L-A of the ASME Code, the licensee states it would perform the similar examination (100 percent of accessible areas) for the third 10-year interval by March 28, 2020, and 2025 (plus or minus 1 year).

The licensee also stated that the 2016 examination indications noted minor spalls, efflorescence, pop-outs, cracks, stains, nails, or metal trapped within the concrete and abandoned anchors/anchor holes. Due to the controlled environment within the enclosure building, there have been no changes in the indications. All indications and conditions identified were minor in nature and did not require excavation or repair. Based on these inspections, the licensee concludes that the MPS3 containment structure is in good material condition. The containment structure continues to retain its ability to perform, as designed.

Table 3 below provides an approximate schedule for MPS3 containment surface examinations, assuming the Type A test frequency is extended to 15 years.

Table 3 MPS3 Containment (CTMT) Surface Examination Schedule Calendar Year Type A Test (ILRT)

General Visual Examination of Accessible Exterior Surfaces (IWL)

CTMT Basemat Settlement Inspection General Visual Examination of Accessible Interior Liner Surfaces (IWE)

MR Structural Inspection CTMT&

Enclosure Building CTMT Enclosure Roof Inspection 1998 1/17 10/3 (Pre-ILRT) 1/13 10/3 (Pre-ILRT) 12/18 1999 5/27 5/31 2000 4/7 9/7 2001 First required IWL Exam (8/21) 2/21 First required IWE Exam (2/1) 2002 11/7 9/25 6/18 2003 4/17 7/15 2004 7/6 4/28 6/15 2005 10/6 10/6 6/15 2006 3/27 5/1 6/14 2007 6/19 5/31 6/12 2008 11/3 7/11 2009 4/23 6/10 2010 5/8 6/8 2011 11/6 1/31 IWL 11/4 (Pre-ILRT) 6/30 11/4 (Pre-ILRT) 11/3 6/8 2012 6/11 2013 5/13 9/1 5/6 2014 10/14 2015 2016 1/18 4/27 6/14 5/16 4/20 2017 10/31 10/17 2018 2019 (4/13)

(6/1)

(4/11) 2020 (2/18) (+/- 1 year)

(10/4) 2021 (5/12)

Note 1 2022 2023 Spring (6/1)

(5/16) 2024 2025 (2/18) (+/- 1 year)

Spring (Fall) 2026 (15 Year)

(Pre-ILRT)

(Pre-ILRT)

Note 1: The inspection frequency is each Refueling Outage. Inspections are rescheduled upon completion of the current inspection.

Type B and Type C LLRT Program In its LAR, the licensee states that the MPS3 Appendix J Type B and Type C testing program requires testing of electrical penetrations, airlocks, hatches, flanges, and valves within the scope of the program as required by 10 CFR Part 50, Appendix J, Option B, and TS 6.8.4.f.

A review of the most recent Type B and Type C test results and their comparison with the allowable leakage rate was performed. For MPS3, the combined Type B and Type C leakage acceptance criterion is 0.60 La or 307,520 sccm. The maximum and minimum pathway leak rate summary totals for the last three refueling outages are shown in Table 4 below.

Table 4 Type B and Type C LLRT Combined As-Found/As-Left Trend Summary Refueling Outage (Year) 3R19 May 2019 3R18 April 2017 3R17 April 2016 As-Found Minimum Pathway (sccm) 15,896 (5.169% of 0.6La) 24,838 (8.076% of 0.6La)

Undetermined (>0.6La)

As-Left Minimum Pathway (sccm) 12,299 (3.999% of 0.6La) 24,131 (7.847% of 0.6La) 18,240 (5.931% of 0.6La)

As-Left Maximum Pathway (sccm) 25,780 (8.383% of 0.6La) 66,025 (21.470% of 0.6La) 66,025 (18.352% of 0.6La) 3.2.4 Deterministic Review Summary Based on the preceding regulatory and technical evaluations, the NRC staff finds that the licensee has submitted its CISI schedule and has adequately implemented its primary containment leakage rate testing program consisting of ILRT and LLRT. The results of the recent ILRT and LLRT combined totals demonstrate acceptable performance and support a conclusion that the structural and leaktight integrity of the primary containment vessel is adequately managed and will continue to be periodically monitored and managed by the ILRTs and LLRTs.

The NRC staff finds that there is reasonable assurance that the structural and leaktight integrity of the MPS3 primary containment will continue to be monitored and maintained with the performance-based Type A test interval extended up to one test in 15 years, without undue risk to public health and safety. Therefore, the NRC staff concludes that the licensees containment inspection programs support extension of the ILRT frequency as requested in the LAR.

3.3 NRC Staff Evaluation of the Conditions and Limitations 3.3.1 Conditions and Limitations in NEI 94-01, Revision 2-A In its SE for NEI 94-01, Revision 2, dated June 25, 2008, the NRC staff concluded that the guidance in NEI 94-01, Revision 2, is acceptable for reference by licensees proposing to amend their TSs regarding containment leakage rate testing subject to six conditions and limitations.

The requirements of NEI 94-01 stayed essentially the same from the original version through Revision 2, except that the regulatory positions of RG 1.163 were incorporated, and the maximum ILRT interval was extended to 15 years. The licensees response to each of the six conditions and limitations from NEI 94-01, Revision 2-A (Section 4.1 of the SE), is listed in the table contained in the attachment to the LAR, Section 4.0, Technical Analysis. The NRC staff evaluated the licensees application to determine whether it adequately addressed these conditions.

NEI 94-01, Revision 2-A, Condition 1 Condition 1 is derived from Sections 3.1.1.1 and 4.1 of the NRC SE dated June 25, 2008, and stipulates that for calculating the Type A leakage rate, the licensee should use the definition in NEI 94-01, Revision 2-A, in lieu of the definition in ANSI/ANS-56.8-2002. In Section 4.0 of its LAR, the licensee states:

Following the NRC approval of this license amendment request, DENC will use the definition in Section 5.0 of NEI 94-01, Revision 3-A (and Revision 2-A), for calculating the Type A leakage rate when future MPS3 Type A tests are performed.

The NRC staff reviewed the definitions of performance leakage rate contained in NEI 94-01, Revisions 2 and 3-A. The staff notes that the definition in NEI 94-01 Revision 2-A, Section 5.0, remained the same in NEI 94-01, Revision 3-A. The staff finds that the licensee will use the definition found in Section 5.0 of NEI 94-01, Revision 2, for calculating the Type A leakage rate in the MPS3 Containment Leakage Rate Testing Program. Therefore, the staff concludes that this definition is acceptable and that the licensee has sufficiently addressed Condition 1.

NEI 94-01, Revision 2-A, Condition 2 Condition 2 is derived from Sections 3.1.1.3 and 4.1 of the NRC SE dated June 25, 2008, and stipulates that the licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. In Section 4.4.2, IWL Examinations, of the LAR, the licensee provided the MPS3 schedule of containment inspections.

The NRC staff reviewed LAR Section 4.4, Supplemental Inspection Requirements, and the table contained in LAR Section 4.4.2. The staff finds that the table provides the scheduled dates for 100 percent completion of each required IWE and IWL inspection and each pre-ILRT inspection. Based on its review, the staff has confirmed that the IWE/IWL 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, can be satisfied for MPS3.

Therefore, the staff concludes that the licensee can satisfy the guidance contained in NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2, and the provisions in Section 3.1.1.3 of the NRC SE dated June 25, 2008. Accordingly, the NRC staff finds that the licensee has sufficiently addressed NEI 94-01, Revision 2-A, Condition 2.

NEI 94-01, Revision 2-A, Condition 3 Condition 3 is derived from Sections 3.1.3 and 4.1 (in the enclosure, page 19) of the NRC SE dated June 25, 2008, and stipulates that the licensee addresses the areas of the containment structure potentially subjected to degradation. In LAR Section 4.0, the licensee states:

General visual examination of accessible interior and exterior surfaces of the containment system for structural problems is typically conducted in accordance with the Millstone IWE/IWL Containment lnservice Inspection Plans which implement the requirements of the ASME,Section XI, Subsections IWE and IWL, as required by 10 CFR 50.55a(g).

Previously, the Millstone IWE Program had inspected the accessible leak chase channels and plugs or caps during the general visual examination as a liner boundary. In response to NRC Information Notice 2014-07, Degradation of Leak Chase Channel System for Floor Welds of Metal Containment Shell and Containment Metallic Liner, the examination was expanded to include an inspection under E-A Containment Surfaces, Item No. E130 - Moisture Barriers.

This examination identified no deficiencies. At this time there are no primary containment surface areas that require augmented examinations in accordance with ASME Section XI, IWE-1240.

Within the Millstone IWL Program no repairs were required. The internal and external inspections confirmed the containment structure is in good material condition. No significant defects or concerns were observed on the exterior concrete and the observed indications were due to original construction. Taken together or individually, the indications do not represent a significant structural concern. The containment structure continues to retain its ability to perform as designed.

The NRC staff reviewed the information contained in LAR Section 4.4, Supplemental Inspection Requirements. The NRC SE for NEI 94-01, Revision 2, dated June 25, 2008, states, in part:

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

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 MPS3 second 10-year inservice inspection interval is the 2001 Edition with the 2003 Addenda, Subsections IWE and IWL. 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 I penetration liners.

IWL deals with Class CC reinforced concrete and post-tensioning systems.

Inaccessible Areas/Augmented Examinations The programmatic requirements for Class MC application inaccessible areas as specified in 10 CFR 50.55a(b)(2)(ix)(A) are:

(1)

The applicant or 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.

(2)

For each inaccessible area identified for evaluation, the applicant or licensee must 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.

In LAR Section 4.4, the licensee indicated that the MPS3 primary containment examinations are performed in accordance with the IWE/IWL program and satisfy the general visual examination requirements specified in 10 CFR Part 50, Appendix J, Option B, as follows: Identification and evaluation of inaccessible areas are addressed in accordance with the requirements of 10 CFR 50.55a(b)(2)(ix)(A) and (E).

LAR Section 4.4.1, IWE Examinations, states, in part:

A review was conducted for MPS3 in accordance with IWE-1241, Examination Surface Areas (1992 Edition with 1992 Addenda of ASME XI) per the initial 10-year Category E-C examination requirements. No areas were deemed susceptible to accelerated degradation and aging; therefore, augmented examinations per Category E-C were not required.

At this time, no augmented Category E-C examinations are planned for MPS3.

Bellows The MPS3 Type B and Type C testing program consists of local leak rate testing 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 the 80 MPS3 Type B 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. See Section 3.2.2, Types B and C Leak Rate Test History, below for additional detail about the monitoring of these electrical penetrations.

Bolting The licensee performs bolting examinations in accordance with the IWE/IWL 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 4.4.1, IWE Examinations, states, in part:

The only significant repair was the replacement of localized areas of the moisture barrier in October of 2005, and May of 2016.

Containment Liner Backed by Concrete The NRC staff notes that in accordance with the requirements of ASME IWE, a complete inspection of all accessible MPS3 primary containment liner surface areas is performed in each of the three periods that makes up each 10-year CISI interval.

LAR Section 4.4.1, IWE Examinations, states, in part:

Since the exterior containment concrete is fully enclosed and in a dry environment, no wicking potential exists from outside to the backside of the liner.

There are no primary containment surface areas that require augmented examination in accordance with ASME Section XI, IWE-1240. However, during each period, 100 percent of the accessible primary containment surface area is inspected. Any significant changes or potential concerns receive a detailed inspection at that time.

Containment Concrete Structure LAR Section 4.4.2, IWL Examinations, states, in part:

The 2016 examinations of the concrete exterior were conducted by a Quality Control inspector and the responsible engineer, using the approved ASME Code visual methods. During the examinations, indications noted were minor spalls, efflorescence, pop-outs, cracks, stains, nails or metal trapped within the concrete, and abandoned anchors/anchor holes. Due to the controlled environment within the enclosure building, there have been no changes in the indications.

In summary, the NRC staff finds that based on the information contained in LAR Section 4.4, the licensee has an established a CISI program that can satisfy the issues in Section 3.1.3 of the NRC SE dated June 25, 2008. Accordingly, the NRC staff concludes that the licensee has adequately addressed NEI 94-01, Revision 2-A, Condition 3.

NEI 94-01, Revision 2-A, Condition 4 Condition 4 is derived from Sections 3.1.4 and 4.1 (in the enclosure, page 19) of the NRC SE dated June 25, 2008, and stipulates that the licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable. In LAR Section 4.0, the licensee states: No major modifications to the MPS3 containment structure have been performed.

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 or ILRT, and that any plans for such major modification also include appropriate pressure testing. As stated in the licensees response to Condition 4, no major repairs or modifications have been performed to the MPS3 primary containment. Additionally, the licensee does not indicate any future major repairs or modifications. 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 related to Condition 4, as described in the NRC SE for NEI 94-01, Revision 2.

NEI 94-01, Revision 2-A, Condition 5 Condition 5 is derived from Sections 3.1.1.2 and 4.1 of the NRC SE dated June 25, 2008, and stipulates 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, then the licensee must demonstrate to the NRC staff that it is an unforeseen, emergent condition. Justification for such an extension request will be in accordance with the staff position in Regulatory Issue Summary 2008-27, Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50, dated December 8, 2008 (ADAMS Accession No. ML080020394).

NRC SE Section 3.1.1.2 for NEI 94-01, Revision 2, states:

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

The licensee states in the LAR that DENC acknowledges and accepts the NRC staff position as communicated to the nuclear industry in Regulatory Issue Summary 2008-27. The above passage from SE Section 3.1.1.2 accurately reflects the regulatory issue summary NRC staff position.

The NRC staff finds the licensee has demonstrated its understanding that any 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) would 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 NEI 94-01, Revision 2-A, Condition 5.

NEI 94-01, Revision 2-A, Condition 6 Condition 6 is derived from Section 4.1 of the NRC SE dated June 25, 2008, and stipulates that for plants licensed under 10 CFR Part 52, applicants 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-A, and EPRI TR-1009325, Revision 2, including the use of past containment ILRT data. The licensee states in LAR Section 4.0 that this is not applicable because MPS3 was not licensed under 10 CFR Part 52. The NRC staff is in agreement that NEI 94-01, Revision 2-A, Condition 6, is not applicable to MPS3.

Summary of Conditions and Limitations in NEI 94-01, Revision 2-A Based on the above evaluation of each condition of NEI 94-01, Revision 2-A, the NRC staff has determined that the licensee has adequately addressed the six conditions identified in Section 4.1 of the NRC SE for NEI 94-01, Revision 2. Therefore, the NRC staff concludes it is acceptable for DENC to adopt the six conditions and limitations of NEI 94-01, Revision 2-A, as part of the implementation documents in TS 6.8.4.f for MPS3.

3.3.2 NEI 94-01, Revision 3-A, Conditions Satisfied In its SE dated June 8, 2012, for NEI 94-01, Revision 3, the NRC staff concluded that the guidance in NEI 94-01, Revision 3, is acceptable for reference by licensees proposing to amend their TSs regarding containment leakage rate testing, subject to the following two conditions and limitations.

The licensee indicated in the LAR that MPS3 will meet the limitations and conditions of NEI 94-01 Revision 3-A, Section 4.0. Accordingly, MPS3 will be adopting, in part, the testing criteria ANSI/ANS 56.8-2002 as part of its licensing basis. As stated in Section 2.0, Purpose and Scope, of NEI 94-01 Revision 3-A, where technical guidance overlaps between NEI 94-01, Revision 3-A, and ANSI/ANS 56.8-2002, the guidance in NEI 94-01, Revision 3-A, takes precedence.

NEI 94-01, Revision 3-A, Condition 1 Section 4.0 of Enclosure 1 of the SE to the NRC letter dated June 8, 2012, for Condition 2 stipulates that:

NEI TR 94-01, Revision 3, 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 presents three separate issues that are required to be addressed:

1) The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement 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 shall be developed to restore the margin to an acceptable level.

3) Use of the allowed 9-month extension for eligible Type C valves is only authorized for nonroutine emergent conditions with exceptions as detailed in NEI 94-01, Revision 3-A, Section 10.1, Introduction.

In LAR Section 4.0, the licensee states:

Following approval of this amendment and consistent with the guidance of NEI 94-01, Rev. 3-A, DENC will assess and monitor margin between the Type B and C leakage rate summation and the regulatory limit and include this margin in a post outage report. This will include corrective actions to restore margin to an acceptable level, if required.

The NRC staff has reviewed the requirements of NEI 94-01, Revision 3, against the DENC response for Issues (1), (2), and (3) of Condition 1. The licensees response indicates that following approval of the subject amendment, DENCs actions will be consistent with the guidance in NEI 94-01, Revision 3-A. The 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 staffs SE input pertaining to Issues (1), (2), and (3). Based on the above, the NRC staff finds that DENC acknowledges all the requirements of Condition 1 and that it has established its intent for MPS3 to comply with these requirements. Therefore, the NRC staff concludes that the licensee has sufficiently addressed Condition 1 of NEI 94-01, Revision 3-A.

NEI 94-01, Revision 3-A, Condition 2 Section 4.0 of Enclosure 1 of the SE to the NRC letter dated June 8, 2012, for Condition 2 stipulates that:

The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust 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 is 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 TR 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 Types B & C total [leakage], 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 presents two separate 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, Report Requirements.

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.

In LAR Section 4.0, the licensee states:

Following approval of this amendment and consistent with the guidance of Section 11.3.2 of NEI 94-01, Rev. 3-A, DENC will estimate the amount of understatement in the Type B & C total and include determination of the acceptability in a post outage report.

The NRC staff has reviewed the requirements of NEI 94-01, Revision 3, against the DENC response for Issues (1) and (2) of Condition 2. The licensees response indicates that following approval of the subject amendment, DENCs actions will be consistent with the guidance in NEI 94-01, Revision 3-A. The staff notes that revised guidance contained in Revision 3-A, Section 11.3.2, Programmatic Controls, and Section 12.1, Report Requirements, reflects the staffs SE input pertaining to both Issues (1) and (2). Based on the above, the NRC staff finds that DENC acknowledges all the requirements of Condition 2 and that it has established its intent for MPS3 to comply with these requirements. Therefore, the NRC staff concludes that the licensee has sufficiently addressed Condition 2 of NEI 94-01, Revision 3-A.

3.3.3 Limitations and Conditions Summary 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 staffs SE incorporated in NEI 94-01, Revision 2-A. Therefore, the staff finds that the proposed changes to MPS3 TS 6.8.4.f regarding the primary containment leakage rate testing program are acceptable.

3.4 Probabilistic Risk Assessment Review 3.4.1 Risk-Informed Considerations The risk-informed considerations presented below address the fourth and fifth key principles of the staffs standards for risk-informed decisionmaking in RG 1.174, which concern the change in risk and monitoring the impact of the licensing basis change.

The licensee provided a plant-specific risk assessment for permanently extending the currently allowed containment Type A ILRT interval from 10 years to 15 years in Attachment 3 to the LAR submitted July 30, 2019.

In Section 4.6 of Attachment 1 to the LAR, the licensee states that the plant-specific risk assessment follows the guidance in NEI 94-01, Revision 2-A1; the methodology described in TR-1009325, Revision 2-A; and the NRC regulatory guidance outlined in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis. In addition, the licensee applied the methodology from the Calvert Cliffs Nuclear Power Plant request for additional information letter dated March 27, 2002 (ADAMS Accession No. ML020920100), to assess the risk from undetected containment leaks due to steel liner corrosion.

Section 9.2.3.1, General Requirements for ILRT Interval Extensions beyond Ten Years, of NEI 94-01, Revision 3 A, states that plant-specific confirmatory analyses are required when extending the Type A ILRT interval beyond 10 years. Section 9.2.3.4, Plant-Specific Confirmatory Analyses, of NEI 94-01, states that the assessment should be performed using the approach and methodology described in EPRI TR-101824312, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals. The analysis is to be performed by the licensee and retained in the plant documentation and records as part of the basis for extending the ILRT interval.

The licensee addressed each of the four conditions for the use of EPRI TR-1009325, Revision 2, which are listed in Section 4.2 of the NRC safety evaluation report. A summary of how each condition is met is provided in Sections 3.4.2.1 through 3.4.2.4 below.

3.4.2 Key Principle 4: Change in Risk is Consistent with the Commissions Policy Statement on Safety Goals Section 9.2.3.1, General Requirements for ILRT Interval Extensions beyond Ten Years, of NEI 94-01, Revision 3-A, discusses how plant-specific confirmatory analyses are required when extending the Type A ILRT interval beyond 10 years. Section 9.2.3.4, Plant-Specific Confirmatory Analyses, of NEI 94-01 states:

The assessment should be performed using the approach and methodology described in EPRI Report 1018243, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals. The analysis is to be performed by the licensee and retained in the plant documentation and records as part of the basis for extending the ILRT interval.

In its SE dated June 25, 2008, the NRC staff found the methodology in EPRI TR-1009325, Revision 2, acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT interval to 15 years, provided certain conditions are satisfied.

These conditions, which are set forth in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, stipulate that:

1) The licensee submits documentation indicating that the technical adequacy of its probabilistic risk assessment (PRA) is consistent with the requirements of RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment 1 NEI 94-01, Revision 3-A, added guidance for extending Type C LLRT surveillance intervals beyond 60 months. The guidance for extending Type A ILRT surveillance intervals beyond 10 years is the same as that in Revision 2-A.

2 EPRI TR-1018243 is also identified as EPRI TR-1009325, Revision 2-A. This report is publicly available and can be found at www.epri.com by typing 1018243 in the search field box.

Results for Risk-Informed Activities, dated March 2009 (ADAMS Accession No. ML090410014) relevant to the ILRT extension application.

2) The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small and consistent with the clarification provided in Section 3.2.4.63 of the SE for EPRI TR-1009325, Revision 2.

3) The methodology in EPRI TR-1009325, Revision 2, is acceptable, provided the average leak rate for the preexisting containment large leak accident case (i.e., accident case 3b) used by the licensee is assigned a value of 100 times the maximum allowable leakage rate (La) instead of 35 La.

4) An LAR is required in instances where containment overpressure is relied upon for emergency core cooling system (ECCS) performance. According to the clarification provided in Section 3.2.4.6 of the NRC safety evaluation report for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2, plants that rely on containment overpressure (or containment accident pressure) net positive suction head for ECCS injection must also consider CDF in the ILRT evaluation.

The licensee addressed each of the four conditions for the use of EPRI TR-1009325, Revision 2, which conditions are listed in Section 4.2 of the NRC SE. A summary of how the licensee meets each condition is provided in the sections below.

3.4.2.1 Technical Adequacy of the PRA The first condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, stipulates that the licensee submit documentation indicating that the technical adequacy of its PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension application. This RG describes an 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 decisionmaking for light-water reactors.

Internal Events The licensee addresses the MPS3 PRA technical adequacy in LAR Attachment 3, Enclosure A.

Probabilistic Risk Assessment Acceptability. As discussed in Enclosure A of Attachment 3 to the LAR, the MPS3 risk assessment performed to support the ILRT application used the current MPS3 internal events PRA model of record, which the licensee completed in 2018. The 2018 versions of the MPS3 PRA models are the most recent risk profile evaluations at MPS3 for internal events. The licensee explains its approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models. This approach includes both a proceduralized PRA maintenance and update process and the use of self-assessments and independent peer reviews.

The MPS3 PRA model for internal events received a formal industry peer review in 1999.

Subsequent to the peer review, the model had been updated several times, and findings and observations (F&Os) were addressed during each model update. All F&Os identified during the 3 Section 4.2 of the SE for EPRI TR-1009325, Revision 2, indicates that the clarification regarding small increases in risk is provided in Section 3.2.4.5; however, the clarification is actually provided in Section 3.2.4.6.

1999 Westinghouse Owners Group peer review have been addressed. A self-assessment was performed by MPS3 with support from a contractor in 2007 using guidance provided in RG 1.200, Revision 1. In June 2012, another independent contractor performed a focused PRA peer review of model upgrades incorporated since the 1999 Westinghouse Owners Group peer review to assess whether PRA upgrades, as defined by the ASME/ANS PRA standard, meet the intent of Category II Supporting Requirements. Additionally, in 2012, a focused-scope peer review was performed by the contractor. The resulting findings and resolutions and/or impact on the application were provided by MPS3 in Table A.7-1. ln the course of this review, 35 new F&Os were prepared, including 20 suggestions, 14 findings, and 1 best practice.

Lastly, an independent contractor performed a focused-scope peer review to determine compliance with Addendum A of the ASME/ANS PRA standard and RG 1.200, Revision 2. In the course of this review, a total of 114 new F&Os were generated, including 92 findings and 22 suggestions. The results of the F&Os were also provided by MPS3 in Table A.7-1. The status reflects what has been completed since the completion of the 2018 model update where most of the remaining findings were addressed. A staff review of these findings and observations found that there is no material impact on the ILRT application.

External Events With respect to external events, RG 1.174 stipulates that established acceptance guidelines are intended for comparison with a full-scope assessment of the change in the applicable risk metrics. The guidance recognizes that many PRAs are not full scope and PRA information of less than full scope may be acceptable. The methodology described in EPRI TR-1009325, Revision 2-A, which the NRC found to satisfy the key principles of risk-informed decisionmaking of RG 1.174, discusses that if the external event analysis is not of sufficient quality or detail to allow direct application of the methodology, the quality or detail will be increased, or a suitable estimate of the risk impact from the external events should be performed. This assessment can be taken from existing, previously submitted and approved analyses or another alternate method of assessing an order-of-magnitude estimate for contribution of the external event to the impact of the changed interval. Based on this, the licensee performed a very conservative first-order estimate to approximate the potential increase in large early release frequency (LERF) from the ILRT interval extension. This analysis references the currently available information for external events models and information to develop an external events multiplier to be applied to the internal events results.

The individual plant examination of external events (IPEEE) seismic and fire event models have not been updated since the original IPEEE. The insights and information from these sources have been used to estimate the effect on total external hazards such as internal fires, seismic, and others that have been evaluated in the IPEEE. The analyses include bounding evaluations for the dominant external events (fire and seismic) using the information from the MPS3 IPEEE.

Use of the IPEEE analysis to assess external hazards is considered bounding for the purpose of assessing change in risk associated with extending the ILRT interval because the IPEEE does not account for safety improvements made to the as-built, as-operated plant nor state-of-the-art PRA methods developed after the IPEEE was published. These changes, if applied to the IPEEE, would tend to decrease the calculated external hazards CDF and reduce uncertainty in the calculation.

The IPEEE fire risk analysis quantified a CDF impact by combining the frequency of fires and the probability of detection/suppression failure with the remaining safety function unavailabilities.

A systematic approach was used to identify critical fire areas where fires could fail safety functions and pose an increased risk of core damage if other safety functions are unavailable.

The CDF calculated due to fire is 4.8E-06/year, with the dominant risk being fire in the cable spreading room, switchgear rooms, control room, and charging/component cooling pump zone.

The licensee does not maintain a seismic PRA model for MPS3. The licensee estimated a mean core damage frequency of 9.1E-6/year from seismically initiated events using site-specific seismic hazards in its IPEEE. NRC Generic Issue 199 (GI-199), Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern Unites States on Existing Plants, Safety/Risk Assessment, Appendix D: Seismic Core-Damage Frequencies (ADAMS Accession No. ML100270756), provides the seismic core damage frequency estimates developed in the safety/risk assessment table of the LAR, which provides seismic CDFs using 2008 U.S.

Geological Survey seismic hazard curves. The seismic CDF (SCDF) estimate of 9.1E-6/year provided by the licensee for MPS3 is conservatively larger than the calculated simple average SCDF of 7.6E-6/year and the IPEEE weighted average SCDF of 6.5E-6/year.

As a bounding estimate for the ILRT external events risk impact assessment, the licensee chose to use a conservative ratio between the LERF and CDF. The licensee assumed that the ratio between LERF and CDF for external events was lower than the ratio for internal events, given that some internal event sequences such as interfacing system loss-of-coolant accidents and steam generator tube ruptures, contribute directly to LERF via containment bypass mechanisms. The same ratio between LERF and CDF was conservatively used by the licensee for calculation of internal and external events. The result of 4.48E-07/year was used by the licensee in its analysis to represent the seismic LERF.

The NRC staff finds that the licensees estimated SCDF is sufficient to support an order-of-magnitude MPS3 ILRT external events risk impact assessment because the licensee demonstrated that the SCDF is conservative in comparison to the GI-199 calculated simple average and the IPEEE weighted average SCDF for MPS3.

Based on the above considerations, the NRC staff finds the licensees analysis of the impact of external events acceptable for the ILRT application. Furthermore, the licensee has evaluated its internal events PRA against the current PRA standard and Revision 2 of RG 1.200. The NRC staff finds that the licensee has addressed the findings from the peer reviews and that they have no impact on the results of this application. The NRC staff concludes that the internal events PRA model used by the licensee is of sufficient technical adequacy to support the evaluation of changes to ILRT frequencies. Accordingly, the first condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, is met.

3.4.2.2 Estimated Risk Increase The second condition in Section 4.2 of the SE 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 and consistent with the guidance in RG 1.174 and the clarification provided in Section 3.2.4.5 of the NRC SE for NEI 94-01, Revision 2, and EPRI TR-1009325, Revision 2. Specifically, a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive.

In addition, a small increase in conditional containment failure probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage points. Additionally, for plants that rely on containment overpressure for net positive suction head for ECCS injection, both CDF and LERF will be considered in the ILRT evaluation and compared with the risk acceptance guidelines in RG 1.174. 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, CCFP, delta CDF, and LERF.

The licensee reported the results of the plant-specific risk assessment in Section 5 and sensitivity calculations in Section 6 of Appendix 3 to the LAR. External events are considered in Section 5.7. The reported risk impacts are based on a change in the Type A containment ILRT frequency from three tests in 10 years (the test frequency in 10 CFR Part 50, Appendix J, Option A) to one test in 15 years and also account for the risk from undetected containment leaks due to steel liner corrosion. The following conclusions can be drawn from the licensees analysis associated with extending the Type A ILRT frequency:

1) RG 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-6/year. The MPS3 design does not rely on overpressure of containment to ensure adequate net positive suction head of the ECCS pumps. Since MPS3 does not rely on containment accident pressure for ECCS net positive suction head during certain design-basis accidents, extending the ILRT interval does not impact CDF. Thus, the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small using the acceptance guidelines of RG 1.174.

RG 1.174 defines very small changes in risk as resulting increases in LERF less than 1.0E-07/year. 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 with corrosion included is estimated as 2.59E-08/year using the EPRI guidance. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of RG 1.174. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years is estimated as 1.53E-07/year using the EPRI guidance, and total estimated upper bound LERF is 9.74E-07/year. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174. The risk change resulting from a change in the Type A ILRT test interval from 3-in-10 years to 1-in-15 years bounds the 1-in-10 years to 1-in-15 years risk change.

2) The effect resulting from changing the Type A test frequency to 1-per-15 years measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 5.75E-02 person-rem/year. 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 safety evaluation report for NEI 94-01, Revision 2. Thus, the increase in the total integrated plant risk for the proposed change is considered small and supportive of the proposed change.

3) The increase in the CCFP due to the change in test frequency from 3-in-10 years to once in 15 years is 0.92 percent. NEI 94-01 states that increases in CCFP of 1.5 percent is small. This value is below the acceptance guidelines in Section 3.2.4.6 of the NRC SE for NEI 94-01, Revision 2, and supportive of the proposed change.

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

Accordingly, the second condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, and Key Principle 4 in RG 1.174 is met.

3.4.2.3 Leak Rate for the Large Preexisting Containment Leak Rate Case The third condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, stipulates that in order to make the methodology in EPRI TR-1009325, Revision 2, acceptable, the average leak rate for the preexisting containment large leak rate accident case (i.e., accident case 3b) used by the licensee shall be 100 La instead of 35 La. As noted by the licensee in Section 1.3 of to the LAR, the methodology in EPRI TR-1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the preexisting containment large leakage rate accident case (accident case 3b), and this value has been used in the MPS3 plant-specific risk assessment. Accordingly, the third condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, is met.

3.4.2.4 Applicability if Containment Overpressure is Credited for ECCS Performance The fourth condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, stipulates that in instances where containment overpressure is relied upon for ECCS performance, an LAR is required to be submitted. In Section 4.6.1 of Attachment 1 to the LAR, the licensee states that containment overpressure is not relied upon for ECCS performance at MPS3. Thus, no reliance is placed on pressure and/or temperature transients to ensure adequate net positive suction head. Accordingly, the fourth condition in Section 4.2 of the SE for EPRI TR-1009325, Revision 2, is not applicable.

NEI 94-01, Revision 3-A, documents the NRCs evaluation and acceptance of extending the Type C LLRT to 120 months, subject to the specific limitations and conditions listed in Section 4.0 of the SE. However, the licensee requested that the allowable extended interval for Type C LLRT be increased only to 75 months to be conservative, with a permissible extension (for nonroutine emergent conditions) of 9 months (84 months total). Therefore, the staff finds the proposed extension acceptable.

3.4.3 Key Principle 5: Monitor the Impact of the Proposed Change The NRC staffs SE incorporated into Revision 3-A of NEI 94-01 and Revision 2-A of NEI 94-01 contains a section entitled, The Impact of the Proposed Change Should be Monitored Using Performance Measurements Strategies, which states:

In addition to maintaining the defense-in-depth philosophy as described in Section 3.2.2 of this SE [for Key Principle 2], the applicants for TS amendments will continue to perform containment inspections during the Type A test interval in Sections 3.1.3 and 3.1.4 of this SE [as discussed below].

As documented in NUREG-1493, industry experience has shown that most ILRT failures result from leakage that is detectable by local leakage rate testing [LLRT]

(Type B and C testing). Specific testing frequencies for the local leak rate tests are reviewed prior to every refueling outage (18-month cycle). An outage scope document is issued to document the local leak rate test periodically and to ensure that all pre-maintenance and post-maintenance testing is complete. The post-outage report provides a written record of the extended testing interval changes and the reasons for the changes based on testing results and maintenance history. Based on the above measures, the LLRT program will provide continuing assurance that the most likely sources of leakage will be identified and repaired.

ANSl/ANS-56.8-2002, Section 6.4.4, also specifies surveillance acceptance criteria for Type B and Type C tests and states that: The combined [as-found]

leakage rate of all Type B and Type C tests shall be less than 0.6La when evaluated on a minimum pathway leakage rate basis, at all times when containment operability is required. It states, moreover, that: The combined leakage rate for all penetrations subject to Type B and Type C test shall be less than or equal to 0.6La as determined on an maximum pathway leakage rate basis from the as-left LLRT results. These combined leakage rate determinations shall be done with the latest leakage rate test data available, and shall be kept as a running summation of the leakage rates.

The containment components' monitoring and maintenance activities will be conducted according to the requirements of 10 CFR 50, Appendix J, and 10 CFR 50.55a.

The NRC staff finds that the above provisions are considered to be acceptable performance monitoring strategies for assuring that the risk of the proposed change will remain small.

Accordingly, Key Principle 5 is met.

3.5 Technical Evaluation Summary Based on the preceding regulatory and technical evaluations, the NRC staff concludes that the licensee has adequately implemented its primary containment leakage rate testing program consisting of ILRT and LLRT. The results of the recent ILRT and the LLRT (Type B and Type C tests) combined totals demonstrate acceptable performance and support a conclusion that the structural and leaktight integrity of the primary containment vessel is adequately managed and will continue to be periodically monitored and managed effectively. The NRC staff concludes 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 staffs SE that were incorporated into NEI 94-01, Revision 2-A. The NRC staff concludes that the risk impact for extending the ILRT intervals is consistent with the acceptance guidelines of RG 1.174.

Therefore, the NRC staff concludes that the proposed changes to MPS3 TS 6.8.4.f regarding the primary containment leakage rate testing program are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Connecticut State official was notified on June 4, 2020, of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 or changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on this finding (October 8, 2019; 84 FR 53770). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

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

Principal Contributors: B. Lee D. Hoang E. Bousquet J. Dozier Date: July 15, 2020

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