S ENERGY°McGuire DUKESteven Vice Capps Nuclear Station · 2016-03-24 · Duke Energy requests...

S DUKESteven D. CappsVice President

ENERGY°McGuire Nuclear Station

Duke EnergyMG01VP 1 12700 Hagers Ferry Road

Huntersville, NC 28078

o: 980.875.4805f: 980.875.4809

Steven.Capps@du ke-energy.com

Serial No.: MNS-16-015 10 CFR 50.90

February 18, 2016

U.S. Nuclear Regulatory CommissionATTN: Document Control DeskWashington, DC 20555-0001

MCGU IRE NUCLEAR STATION, UNIT NOS. 1 AND 2DOCKET NOS. 50-369 AND 50-370 / RENEWED LICENSE NOS. NPF-9 AND NPF-17

SUBJECT: LICENSE AMENDMENT REQUEST (LAR) FOR ONE-TIME EXTENSION OFAPPENDIX J TYPE A INTEGRATED LEAKAGE RATE TEST INTERVAL

In accordance with the provisions of 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke Energy)requests amendments, in the form of changes to the Technical Specifications (TS) to RenewedFacility Operating Licenses NPF-9 and NPF-17 for McGuire Nuclear Station (MNS) Units 1 and2. The proposed LAR would allow for a one-time extension to the 10-year frequency of theMNS Units 1 and 2 containment leakage rate test (i.e., Integrated Leakage Rate Test (ILRT) orType A test). This test is required by TS 5.5.2 "Containment Leakage Rate Testing Program."The proposed change would permit the existing ILRT frequency to be extended from 10 years to10.5 years.

The proposed change would, based on current refueling outage (RFO) projected schedules,allow Duke Energy to minimize the imPact of the ILRT on critical path outage activities by nothaving to perform the MNS Unit 1 ILRT prior to the expiration of the MNS Unit 1 10-yearinterval. Currently, the ILRT is to be performed approximately one year prior to the 10th yearanniversary of the completion of the last ILRT (October 21, 2008). If granted, this revisionwould extend the period from 10 years to 10.5 years between successive tests. In terms ofRFOs, this extension would move the performance of the next ILRT from the scheduled fall2017 End of Cycle 25 RFO to the spring 2019 End of Cycle 26 RFO.

The proposed change would provide the same benefits for MNS Unit 2 outage activities.Currently, the MNS Unit 2 ILRT is to be performed approximately one year prior to the 10th yearanniversary of the completion of the last Type A test (March 31, 2008). If granted, this revisionwould extend the period from 10 years to 10.5 years between successive tests. In terms ofRFOs, this .extension would move the performance of the next ILRT from the scheduled spring2017 End of Cycle 24 RFO to the fall 2018 End of Cycle 25 RFO.

The proposed change would also correct a couple of typographical and administrative errorsintroduced into TS 5.5.2 by the implementation of two previous TS amendments.

•lVJ •www.duke-energy.com

U.S. Nuclear Regulatory CommissionFebruary 18, 2016 of 3

Enclosure 1 provides an evaluation of the proposed change. The marked-up TechnicalSpecification, reflecting the proposed change in this submittal, is included in Attachment 1.Retyped (clean) TS pages are included in Attachment 2.

The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteriain 10 CFR 50.92(c), and it has been determined that the proposed change involves nosignificant hazards consideration. The bases for this determination are included in Enclosure 1.

This letter contains no new Regulatory Commitments and no revision to existing RegulatoryCommitments.

In accordance with Duke Energy administrative procedures that implement the QualityAssurance Program Topical Report, this proposed change has been reviewed and approved bythe Plant Operations Review Committee. A copy of this LAR is being sent to the State of NorthCarolina in accordance with 10 CER 50.91 requirements.

Duke Energy requests approval of this LAR by December 19, 2016. Once approved, theamendment will be implemented within 60 days.

Should you have any questions concerning this letter or require additional information, pleasecontact P. T. Vu of MNS Regulatory Affairs, at 980-875-4302.

I declare under penalty of perjury that the foregoing is true and correct. Executed on February18, 2016.

Sincerely,

Steven D. Capps

Enclosure

1. Evaluation of the Proposed Change

U.S. Nuclear Regulatory CommissionFebruary 18, 2016 of 3

xc: C. HaneyAdministrator, Region IIU.S. Nuclear Regulatory CommissionMarquis One Tower245 Peachtree Center Avenue NE, Suite 1200Atlanta, Ga 30303-1 257

G. E. MillerProject Manager (McGuire)U.S. Nuclear Regulatory CommissionMail Stop 0-8 G9A11555 Rockville PikeRockville, MD 20852-2738

J. ZeilerNRC Senior Resident InspectorMcGuire Nuclear Station

W. L. Cox, Ill, Section ChiefNorth Carolina Department of Health and Human ServicesRadiation Protection Section1645 Mail Service CenterRaleigh, NC 27699-1 645

Enclosure 1Page 1 of 34

Evaluation of the Proposed Change

Subject: License Amendment Request - Revise Technical Specification Section 5.5.2 forOne-Time Extension of Appendix J Type A Integrated Leakage Rate Test Interval

1.0 SUMMARY DESCRIPTION

2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

3.1 Justification for the Technical Specification Change3.1.1 Current 10 CFR 50, Appendix J Requirements3.1.2 10 CFR 50, Appendix J, Option B Licensing History3.1.3 Containment Building Description3.1.4 Plant-Specific Confirmatory Analysis

3.2 Inspections3.2.1 Primary Containment Coatings3.2.2 Inservice Inspection Program for Containment - IWE3.2.3 License Renewal3.2.4 Integrated Leakage Rate Testing (ILRT) History

3.3 Containment Leakage Rate Testing Program, Type B and Type C Testing3.4 NRC Information Notices (INs)

3.4.1 IN 2010-12, Containment Liner Corrosion3.4.2 IN 201 4-07, Degradation of Leak-Chase Channel Systems for Floor

Welds of Metal Containment Shell and Concrete Containment MetallicLiner

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements / Criteria4.2 Precedent4.3 Significant Hazards Consideration4.4 Conclusion

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Attachments: 1. Technical Specification Pages (Mark-up)2. Technical Specification Pages (Retyped)

Enc~osure IPage 2 of 34

1.0 SUMMARY DESCRIPTION

Pursuant to 10 CFR 50.90, Duke Energy requests an amendment to the McGuire NuclearStation (MNS) Unit 1, Renewed Facility Operating License NPF-9, and Unit 2, Renewed FacilityOperating License NPF-17, by incorporating the attached proposed change into the Unit 1 andUnit 2 Technical Specifications (TS). Specifically, the proposed change is a request to reviserTS 5.5.2 "Containment Leakage Rate Testing Programs' to allow one-time extension to the 10-year frequency of the MNS Unit I and Unit 2 containment leakage rate test (i.e., Integrated LeakRate Test (ILRT) or Type A test). The proposed change would permit the existing ILRTfrequency to be extended one time from 10 years'to 10.5 years.

The proposed change would, based on current refueling outage (RFO) projected schedules,allow Duke Energy to minimize the impact of the ILRT on critical path outage activities by nothaving to perform the MNS Unit I ILRT prior to the expiration of the MNS Unit 1 10-yearinterval. Currently, the ILRT is to be performed approximately one year prior to the 10th yearanniversary of the completion of the last ILRT (October 21, 2008). If granted, this revisionwould extend the period from 10 years to 10.5 years between successive tests. In terms ofRFOs, this extension would move the performance of the next ILRT from the scheduled fall2017 End of Cycle 25 RFO to the spring 2019 End of Cycle 26 RFO.

The proposed change would also provide the same benefits for MNS Unit 2 outage activities.Currently, the MNS Unit 2 ILRT is to be performed approximately one year prior to the 10th yearanniversary of the completion of the last Type A test (March 31, 2008). If granted, this revisionwould extend the period from 10 years to 10.5 years between successive tests. In terms ofRFOs, this extension would move the performance of the next ILRT from the scheduled spring2017 End of Cycle 24 RFO to the fall 2018 End of Cycle 25 RFO.

The proposed change would also correct a couple of typographical and administrative errors

introduced into TS 5.5.2 by the implementation of two previous TS amendments.

2.0 DETAILED DESCRIPTION

2.1 Current Containment Leakage Rate Testing Program

Current TS 5.5.2 specifies, "A program shall be established to implement the leakage ratetesting of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, OptionB, as modified by approved exemptions. This program shall be in accordance with theguidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-TestProgram," dated September 1995, as modified by the following exceptions:

NEI 94-01-1995, Section 9.2.3: The first Type A test performed after the May 27, 1993 (Unit1 ) and August 20, 1993 (Unit 2) Type A test shall be performed no later than plant restartafter the End Of Cycle 19 Refueling Outage (Unit 1 ) and August 19, 2008 (Unit 2); and

a. The containment visual examinations required by Regulatory Position C.3 shall beconducted 3 times every 10 years, including during each shutdown for SR 3.6.11 Type Atest, prior to initiating the Type A test."


2.2 TS Change Description

The proposed change (in bold) will revise TS 5.5.2 to state, in part:

A program shall be established to implement the leakage rate testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified byapproved exemptions. This program shall be in accordance' with the guidelines contained inRegulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," datedSeptember 1995, as modified by the following exceptions:

a. NEI 94-01-1995, Section 9.2.3: The first Type A test performed after the October 21,2008 (Unit 1) and March 31, 2008 (Unit 2) Type A test shall be performed no later thanplant restart after the End of Cycle 26 Refueling Outage (Unit 1) and no later than plantrestart after theEnd of Cycle 25 Refueling Outage (Unit 2); and

b. The containment visual examinations required by Regulatory Position C.3 shall beconducted 3 times every 10 years, including during each shutdown for SR 3.6.1.1 TypeA test, prior to initiating the Type A test. '•

The revised dates of October 21, 2008 and March 31, 2008 are the dates of last ILRTscompleted for Unit 1 and Unit 2, respectively. The revised end of cycle numbers, Cycle 26 andCycle 25, are the refueling outages when the next ILRTs will be performed for Unit 1 and Unit 2,respectively.

The proposed change would also correct a couple of typographical and administrative errors:

*NRC letter dated May 8, 2003 for TS Amendments 212/193 (Reference 14) had thecorrect designations, "a. and "b. for the two exceptions above. These amendmentsintroduced the typographical error (SR 3.6.11) into the second exception above. Itshould have been SR 3.6.1.1. The licensee provided the correct SR number in the TSmarkup but wrong SR number in the retyped TS page to the NRC.

*NRC letter dated February 13, 2008 for TS Amendment 244 (Reference 25) introducedan administrative error into TS 5.5.2. This amendment inadvertently omitted thedesignation "a." for the first exception above. Without the "a." designation for the firstexception, the designation for the second exception above was automatically changedby the word processing software to "a." during retyping of the TS page. The licenseeprovided the markup and retyped TS pages without the "a." designation to the NRC.

A markup of TS 5.5.2 is provided in Attachment 1. The retyped TS pages are provided in

Attachment 2.

3.0 TECHNICAL EVALUATION

3.1 Justification for the Technical Specification Change

The requested corrections for the typographical and administrative errors are intended torestore them to the condition intended by their original amendment applications.

The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from thecontainment, including systems and components that penetrate the containment, does not

EncLosure 1Page 4 of 34

exceed the allowable leakage values specified in the TS. 10 CFR 50, Appendix J, also ensuresthat periodic surveillance of reactor containment penetrations and isolation valves is performedso that proper maintenance and repairs are made during the service life of the containment andthe systems and components penetrating primary containment. The limitation on containmentleakage provides assurance that the containment would perform its design function following anaccident up to and including the plant design basis accident (DBA). Appendix J identifies threetypes of required tests: (1) Type A tests, intended to measure the primary containment overallintegrated leakage rate; (2) Type B tests, intended to detect local leaks and to measure leakageacross pressure-containing or leakage limiting boundaries (other than valves) for primarycontainment penetrations; and (3) Type C tests, intended to measure containment isolationvalve leakage rates. Types B and C tests identify the vast majority of potential containmentleakage paths. Type A tests identify the overall (integrated) containment leakage rate and serveto ensure continued leakage integrity of the containment structure by evaluating those structuralparts of the containment not covered by Types B and C testing.

In 1995, 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," was amended to provide a performance-based Option B for thecontainment leakage testing requirements. Option B requires that test intervals for Type A,Type B, and Type C testing be determined by using a performance-based approach.Performance-based test intervals are based on consideration of the operating history of thecomponent and resulting risk from its failure. The use of the term "performance-based" in 10CFR 50, Appendix J refers to both the performance history necessary to extend test intervals aswell as to the criteria necessary to meet the requirements of Option B.

Also in 1995, RG 1.163 was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 6)with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01,Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive,successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) testfrom three tests in 10 years to one test in 10 years. This relaxation was based on an NRC riskassessment contained in NUREG-1 493, (Reference 7) and Electric Power Research Institute(EPRI) TR-1 04285 (Reference 8), both of which showed that the risk increase associated withextending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval,provisions for extending the test interval an additional 15 months were considered in theestablishment of the intervals allowed by RG 1.163 and NEI 94-01, but this "should be usedonly in cases where refueling schedules have been changed to accommodate other factors."

In 2008, NEI 94-01, Revision 2-A, (Reference 3) was issued. This document describes anacceptable approach for implementing the optional performance-based requirements of OptionB to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of theNRC Safety Evaluation Report (SER) on NEI 94-01. The NRC SER was included in the frontmatter of this NEI report. NEI 94-01, Revision 2-A includes provisions for extending Type AILRT intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163(September 1995). It delineates a performance-based approach for determining Type A, TypeB, and Type C containment leakage rate surveillance testing frequencies. Justification forextending test intervals is based on the performance history and risk insights.


3.1 .1 Current 10 CFR 50. Appendix J Reqiuirements

Title 10 CFR Part 50, Appendix J, was revised, effective October 26, 1995, to allow licensees tochoose-containment leakage testing under either Option A, "Prescriptive Requirements," orOption B, "Performance Based Requirements." MNS has implemented the requirements of 10CFR Part 50, Appendix J, Option B for Types A, B and C testing.

RG 1.163, Section C.1, states that licensees intending to comply with 10 CFR Part 50, AppendixJ, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01Revision 0 (Reference 6) rather than using test intervals specified in American NationalStandards Institute (ANSI)/American Nuclear Society (ANS) 56.8-1994. NEI 94-01 Revision 0,Section 11.0, refers to Section 9, which states that Type A testing shall be performed during aperiod of reactor shutdown at a frequency of at least once per ten years based on acceptableperformance history. Acceptable performance history is defined as completion of twoconsecutive periodic Type A tests where the calculated performance leakage was less than 1.0La (where La is the maximum allowable leakage rate at design pressure). Elapsed time betweenthe first and last tests in a series of consecutive satisfactory tests used to determineperformance shall be at least 24 months.

Adoption of the Option B performance-based containment leakage rate testing program alteredthe frequency of measuring primary containment leakage in Type A tests but did not alter thebasic method by which Appendix J leakage testing is performed. The test frequency is basedon an evaluation of the "as found" leakage history to determine a frequency for leakage testingwhich provides assurance that leakage limits will not be exceeded. The allowed frequency forType A testing as documented in NEI 94-01 Revision 0, is based, in part, upon a genericevaluation documented in NUREG-1 493 (Reference 7). The evaluation documented inNUREG-1493 included a study of the dependence of reactor accident risks on containment leaktightness for differing types of containment types. NUREG-1493 concluded in Section 10.1.2that reducing the frequency of Type A tests (ILRT) from the original three tests per ten years toone test per twenty years was found to lead to an imperceptible increase in risk. The estimatedincrease in risk is very small because ILRTs identify only a few potential containment leakagepaths that cannot be identified by Types B and C testing, and the leaks that have been found byType A tests have been only marginally above existing requirements. Given the insensitivity ofrisk to containment leakage rate and the small fraction of leakage paths detected solely by TypeA testing, NUREG-1 493 concluded that increasing the interval between ILRTs is possible withminimal impact on public risk.

NEI 94-01, Revision 0, Section 9.1, states the following concerning the extension of Type A testintervals:

Consistent with standard scheduling practices for Technical Specifications RequiredSurveillances, intervals for recommended Type A testing given in this section may beextended by up to 15 months. This option should be used only in cases where refuelingschedules have been changed to accommodate other factors.

NEI 94-01, Revision 0, Section 9.2.2, states the following concerning Type A test intervals:

If the test interval ends while primary containment integrity is either not required or it isrequired solely for shutdown activities, the test interval may be extended indefinitely.However, a successful Type A test shall be completed prior to entering the operatingmode requiring primary containment integrity.


3.1.2 10 CFR 50. Appendix J. Option B Licensinq History

SER dated September 4, 2002 - ML022540102 (Reference 12)

The Commission issued Amendment No. 207 to Facility Operating License No. NPF-9 and

Amendment No. 188 to Facility Operating License NPF-17 for MNS, Units 1 and 2.

The amendments revised the TS to permit implementation of containment local leakage ratetesting addressed by Title 10 of the Code of Federal Regulations, Part 50, Appendix J, OptionB, and to reference Regulatory Guide 1.163, "Performance-Based Containment Leak TestProgram,"1 dated September 1995, which specifies a method acceptable to the NRC forcomplying with Option B. In addition, the TS were revised regarding soap bubble testing andleak testing of containment purge valves with resilient seals for upper and lower compartmentsand instrument rooms.

SER dated March 12, 2003 - ML030760032 (Reference 13)

The Commission issued Amendment No. 211 to Facility Operating License No. NPF-9 andAmendment No. 192 to Facility Operating License NPF-17 for MNS, Units 1 and 2.

The amendments revised the TSs to allow a one-time change in the Appendix J, Type Acontainment ILRT interval from the currently required 10-year interval to a test interval of 15years.

SER dated May 8, 2003 - ML031280431 (Reference 14)

The Commission issued Amendment No. 212 to Facility Operating License NPF-9 and

Amendment No. 193 to Facility Operating License NPF-17 for MNS, Units 1 and 2.

The amendments revised the TS to (1) modify the Surveillance Requirement to be consistentwith the design of the reactor building access openings, (2) modify the frequency of theSurveillance Requirement for visual inspections for the exposed interior and exterior surfaces ofthe reactor building, and (3) modify the administrative controls for the containment leakage ratetesting program.

SER dated February 13, 2008 - ML073400670 (Reference 25)

The Commission issued Amendment No. 244 to Facility Operating License NPF-9 for MNS,Unit 1.

The amendment revises administrative TS 5.5.2, "Containment Leak Rate Testing Program,"from the currently approved 15-year interval (since the last McGuire Unit 1 Type A test) to afrequency encompassing the end of the McGuire Unit 1 End of Cycle (EOC) 19 refueling outage(approximately 6 months beyond the previous frequency).

Enctosure 1Page 7 of 34

3.1.3 Containment Buildinqi Description

Description of the ContainmentThe Containment consists of a Containment Vessel and a separate Reactor Building enclosingan annulus. The Containment Vessel is a freestanding welded steel structure with a verticalcylinder, hemispherical dome, and a flat circular base.

The containment vessel is a freestanding welded steel structure with a vertical cylinder,hemispherical dome and a flat base. The Containment shell is anchored to the Reactor Buildingfoundation by means of anchor bolts around the circumference of the cylinder base. The baseof the Containment is 1/4 inch liner plate encased in concrete and anchored to the ReactorBuilding foundation. The base liner plate functions only as a leak-tight membrane and is notdesigned for structural capabilities. The Containment Vessel has a nominal inside diameter of115 feet, overall height of 171 feet 3 inches, nominal wall thickness of 0.75 inch, nominal domethickness of 0.6875 inch, nominal bottom thickness of 0.25 inch, and net free volume of 1.2 x106 cubic feet.

The ice condenser region is an insulated cold storage area contained within the annulus formedby the containment wall and the crane wall circumferentially over a 300-degree arc. This area is74 feet high and is maintained at 15 to 27 0F by the glycol refrigeration system. Around 2 millionpounds of borated ice is contained in an array of 1944 steel baskets (81 baskets per bay, 24bays) resting on a lower support structure and positioned laterally by horizontal lattice framesinstalled at various elevations.

The Reactor Building is a reinforced concrete structure composed of a right cylinder with ashallow dome roof and flat circular foundation slab. The cylinder has an inside radius of 63 feet6 inches and a wall thickness of 3 feet. The dome has an inside spherical radius of 87 feet andis 2 feet 3 inches thick. The foundation slab is 137 feet in diameter and 6 feet thick.

The Reactor Building houses the steel containment vessel and is designed to provideenvironmental as well as missile protection for the steel shell.

A five-foot annular space is provided between the steel containment vessel and the ReactorBuilding for control of the containment external temperatures and pressures. The annular spacealso provides a controlled air volume for filtering and provides access to penetrations for testingand inspection. Following a loss-of-coolant accident (LOCA), the annular space is kept at aslightly negative pressure to control and filter radioactive leakage, if any, from the containmentvessel and penetrations.

Containment PenetrationsSeveral penetrations are required through the containment vessel for personnel and equipmentaccess, fuel transfer and various piping systems. The containment penetrations are:

Equipment HatchThe equipment hatch is composed of a 20-foot cylindrical sleeve in the containment shell and adished head with mating, bolted flanges. The flanged joint has double compressible seals withan annular space for pressurization and testing.


Personnel LocksTwo personnel locks are provided for each unit. Each lock has double doors with aninterlocking system to prevent both doors being opened simultaneously. Remote indication isprovided to indicate the position of each door.

Double, inflatable seals are provided on each door. A top connection between the sealsprovides the capability for local leak rate testing as required. The use of double inflatable sealsallows testing of the annulus space without the use of external strongbacks or other remotedevices.

Fuel Transfer PenetrationA 20-inch fuel transfer penetration is provided for transfer of fuel to and from the fuel pool andthe Containment fuel transfer canal. The fuel transfer penetration is provided with a doublegasketed blind flange in the transfer canal and a gate valve in the fuel pool. Expansion bellowsare provided to accommodate differential movement between the connecting buildings.

Spare PenetrationsSpare penetrations are provided to accommodate future piping and electrical penetrations. Thespare penetrations consist of the penetration sleeve and head.

Penetration SleevesAll penetration sleeves are preassembled and welded into containment vessel shell plates.Each shell plate having penetration sleeves is stress relieved prior to installation into thecontainment.

Purge PenetrationsThe purge penetrations each have one interior and one exterior quick-acting tight-sealingisolation valves. During normal Plant operation, Modes 1-4, the containment purge and exhaustisolation valves are failed closed. In Modes 5, 6, and No Mode, these isolation valves willautomatically actuate closed on containment high radiation signal.

Electrical PenetrationsMedium voltage electrical penetrations for reactor coolant pump power use sealed bushings forconductor seals. The assemblies incorporate dual seals along the axis of each conductor.

Low voltage power, control and instrumentation cables enter the containment vessel throughpenetration assemblies, which have been designed to provide two leak tight barriers in serieswith each conductor.

All electrical penetrations have been designed to maintain containment integrity for DesignBasis Accident conditions including pressure, temperature and radiation. Double barriers permittesting of each assembly as required to verify that containment integrity is maintained.

Mechanical PenetrationsMechanical penetrations are treated as fabricated piping assemblies meeting the requirementsof ASME Section III, Subsections NC and which are assigned the same classification as thepiping system that includes the assembly.

The process line and flued heads making up the pressure boundary are consistent with thesystem piping materials; fabrication, inspection and analysis requirements are as required byASME Section Ill, Section NC.


Critical high temperature lines and selected engineered safety system and auxiliary lines(regardless of temperature) require the "Hot Penetration" assembly which features the exteriorguard pipe for the purpose of returning any fluid leakage to the Containment and for protectionof other penetrations in the building annular space. Other lines are treated as cold penetrationssince a leak into the annular space would not cause a personnel hazard or damage otherpenetrations in the immediate area.

Overpressure ProtectionOverpressure protection shall be provided where the potential exists to over pressurizecontainment penetration piping due to thermal expansion of the fluid trapped in the penetrationpiping. In other words, overpressure protection shall be provided to relieve the pressure buildupcaused by the heatup of a trapped volume of incompressible fluid between two positively closingvalves (due to containment temperature transient) back into containment where an open reliefpath exists. This open relief path could be the relief valve on any normally aligned component,or to the Reactor Coolant System (NC) itself.

3.1.4 Plant-Specific Confirmatory Analysis

The purpose of this analysis is to provide risk insights about extending the currently allowedcontainment Type A Integrated Leak Rate Test (ILRT) interval by 6 months for MNS Units 1 and2. The extended test interval is a one-time 6-month increase over the currently approved 10-year test interval. This translates to an extended test interval of 10.5 years. The riskassessment followed the guidelines from the following:

* NEI 94-01, Revision 3-A,* the methodology used in EPRI TR-1 04285 (Reference 8),* the NEI "Interim Guidance for Performing Risk Impact Assessments in Support of One-

Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals"from November 2001 (Reference 17),

* risk insights in support of a request for a plant's licensing basis as outlined in RegulatoryGuide (RG) 1.174 (Reference 4),

* the methodology used for Calvert Cliffs to estimate the likelihood and risk implications ofcorrosion-induced leakage of steel liners going undetected during the extended testinterval (Reference 18), and

* the methodology used in EPRI 1018243 (Revision 2-A of EPRI 1009325) (Reference16).

The findings of the MNS risk assessment confirm the general findings of previous studies thatthe risk impact associated with extending the ILRT interval from 10 years to 10.5 years is"small." The MNS plant-specific results for extending ILRT interval from the current 10 years to10.5 years are summarized below:

* Since the ILRT does not impact Core Damage Frequency (CDF), the relevant criterion isLarge Early Release Frequency (LERF). The increase in LERF resulting from a changein the Type A ILRT test interval from three in 10 years to one in 10.5 years is veryconservatively estimated to be "small."

* An additional assessment of the impact from external events was also performed. In thissensitivity case, the change in the total MNS LERF (including external events) wasconservatively estimated to be "small."


SThe change in Type A test frequency to one per 10.5 years, measured as an increase tothe total integrated plant risk for those accident sequences influenced by Type A testing,is 0.023 person-rem/year. EPRI Report No. 1009325, Revision 2-A states that a verysmall population dose is defined as an increase of < 1.0 person-rem per year, or < 1% ofthe total population dose, whichever is less restrictive for the risk impact assessment ofthe extended ILRT intervals. Moreover, the risk impact when compared to other severeaccident risks is "negligible."

* The increase in the conditional containment failure from the three in 10-year interval toone in 10.5-year interval is approximately 0.5%. EPRI Report No. 1009325, Revision 2-A, states that increases in CCFP of < 1.5 percentage points is very small. Therefore,this increase is judged to be "very small."

Therefore, increasing the ILRT interval to 10.5 years is considered to be insignificant since it

represents a "small" change to the MNS risk profile.

The NRC, in NUREG-1 493, has previously concluded that:

* Reducing the frequency of Type A tests (ILRTs) from three per 10 years to one per 20years was found to lead to an imperceptible increase in risk. The estimated increase inrisk is very small because ILRTs identify only a few potential containment leakage pathsthat cannot be identified by Type B or Type C testing, and the leaks that have beenfound by Type A tests have been only marginally above existing requirements.

* Given the insensitivity of risk to containment leakage rate and the small fraction ofleakage paths detected solely by Type A testing, increasing the interval betweenintegrated leakage-rate tests is possible with minimal impact on public risk. The impactof relaxing the ILRT frequency beyond one in 20 years has not been evaluated. Beyondtesting the performance of containment penetrations, ILRTs also test integrity of thecontainment structure.

The findings for MNS confirm these general findings on a plant-specific basis considering thesevere accidents evaluated for MNS, the MNS containment failure modes, and the localpopulation surrounding MNS.

The insights from this risk analysis support the deterministic analysis showing that there isreasonable assurance that the health and safety of the public will not be endangered byoperation in the proposed manner of this license request.

3.2 Inspections

3.2.1 Primary Containment Coatingqs

Duke Energy complies with RG 1.54, Quality Assurance Requirements for Protective CoatingsApplied to Water-Cooled Nuclear Power Plants.

The original coating materials and coating systems were specified by Engineering and appliedby the Duke Power Construction Department to all structures within the containment and theContainment Vessel. The coating systems were qualified for radiation exposure, pressure,temperature, and'water chemistry exposure during a DBA in accordance with ANSI Ni101.2.


Carboline coating materials are now used for maintenance of the existing coating systems andfor any new applications. These coating systems are specified by Engineering and applied bythe Duke Energy Maintenance Department. The Carboline coating materials have beenqualified over the existing Mobil/Valspar coatings as a mixed system and as a new coatingsystem for radiation exposure, pressure, temperature, and water chemistry exposure during aDBA in accordance with ANSI N101.2.

The elements of the McGuire Coatings Program are documented in Nuclear Operating FleetAdministrative Procedures. The McGuire Coatings Program includes periodic conditionassessments of Service Level I coatings used inside containment. As localized areas ofdegraded coatings are identified, those areas are evaluated for repair or replacement, asnecessary.

A maximum of 20,000 square feet of unqualified coatings inside Containment is considered tobe a negligible fraction of the Containment interior surfaces.

Coatings inside the Containment are classified as either qualified or unqualified for the purposesof debris generation analysis, which encompasses all of the coating systems used (i.e., epoxies,alkyd enamels, and cold galvanizing products). Unqualified coatings in Containment areassumed to fail as particulates, in accordance with NEI 04-07 guidance. The total weight ofunqualified coatings assumed to fail inside Containment is approximately 372.8 pounds, all ofwhich is assumed to transport to the strainer. Qualified coatings are assumed to fail asparticulates within a 5 D zone of influence (ZOI) based on the methodology outlined in WCAP-16568-P. The total weight of qualified coatings assumed to fail inside Containment isapproximately 167.6 pounds, all of which is assumed to transport to the strainer.

3.2.2 Inservice Inspection Program for Containment - IWE

In accordance with the requirements of Paragraph 50.55a(g), and as modified andsupplemented by paragraph 50.55a(b)(2)(ix) of 10 CER Part 50, the Inservice inspection ofClass MC metal containments at Units 1 and 2 of the MNS will be performed in accordance withthe ASME Boiler and Pressure Vessel Code, Section Xl, Division 1, 2007 Edition with the 2008Addenda (hereafter referred to as Section Xl). The examinations will be performed to the extentpracticable within the limitations of design, geometry and materials of construction of thecomponent. Examinations were scheduled for the Third Inspection Interval in accordance withASME Section Xl IWE-241 1.

Component/System Boundaries Subiect to Inspection and Examination

The boundaries of Class MC non-exempt components and their supports are shown ondrawings listed in Section 2 of the Unit Specific Containment ISl Plan.

Note: Boundaries of piping penetration subassemblies (bellows) are also shown ondrawings listed in Section 2 of the Unit Specific Containment ISl Plan. Theseboundaries include component parts classified as ASME Class NC and are indicatedsolely for convenience in performing examinations specified by the Owner and do notconstitute the boundary for the Class MC containment vessel.


Components Exempted from Examination

Vessels, parts, and appurtenances that are outside the boundaries of the containment asdefined in the Design Specifications are exempt. Mechanical process piping penetrations andsubassemblies classified as ASME Code Class 2 (NC), including bellow assemblies (exceptsurfaces of connecting welds to penetration sleeves) are exempt. However, examination ofsome piping penetration subassembly surfaces is included as Owner Specified ExaminationCategories and Requirements. Also exempted from examination are flanges and coversattached to exterior ends of containment penetration sleeves for those penetrations whereflanges and covers are attached to the interior end of the penetration sleeve.

Embedded or inaccessible portions of containment vessels, parts, and appurtenances that metthe requirements of the original Construction Code are exempt. Inaccessible portions ofcontainment vessels, parts, and appurtenances are those with surfaces that cannot be visuallyexamined by direct or remote methods on both sides, and which are embedded or otherwiseobstructed from view by structures, components, or permanent plant equipment or materials.

Portions of containment vessels, parts, and appurtenances that become embedded orinaccessible as a result of vessel repair or replacement if the conditions of IWE-1 232 and IWE-5220 are met are exempt.

Piping, pumps, and valves that are part of the containment system, or which penetrate or areattached to the containment vessel are exempt. These components shall be examined inaccordance with the rules of IWB or IWC, as appropriate to the classification defined by theDesign Specifications.

Pressure Testinq Class MC Components

Except as noted in IWE-5224, a pneumatic leakage test shall be performed followingrepair/replacement activities performed by welding or brazing, prior to returning the componentto service.

There are no periodic system pressure testing requirements for Class MC Components inSubsection IWE.

Examination Boundaries

Section 2 of the Unit Specific Containment ISI Plan contains a listing of drawings that identifyexamination areas subject to IWE examination. These drawings also show Class MCcomponent supports subject to examination, and include portions of Class NC bellows surfaces,which the Owner has elected to include in this program. Revisions to drawings are reviewed foradditions/changes to the ISI boundaries. These additions/changes are incorporated into theContainment ISI Plan as necessary. The controlled drawings are maintained in accordance withapplicable procedures and directives.

Enc~osure 1Page 13 of 34

Inspection Interval and Inspection Periods

The Second and Third Inservice Inspection Intervals for Class MC metal containments and theirsupports are shown below. Note that these intervals may not coincide with Inservice InspectionIntervals for ASME Class 1, 2, and 3 systems and components. The term "EOC" used below isan abbreviation for "End of Cycle" and the associated number indicates the sequential refuelingoutage following initial operation of the unit.

Table 3.2.2-1, Second Containment Inservice Inspection Interval

Unit 1(Notes 1, 2)

Start Date End Date07/15/2005 07/15/2008 07/15/2011 08/31/2014

1st Period 2 nd Period 3 rd PeriodOutage 1 (EOC 17) Outage 3 (EOC 19) Outage 5 (EOC 21)Outage 2 (EOC 18) Outage 4 (EOC 20) Outage 6 (EOC 22)

(1) The schedule for Interval 2 is included in this discussion as it encompasses the previousUnit 1 ILRT.

(2) The start date for Interval 2 was delayed from 07/15/04 to 07/15/05, within the 12-monthadjustment allowed by IWA-2430(d)(1). This adjustment was necessary because of thedelay in obtaining NRC approval of Relief Request Serial #03-GO-010.

Table 3.2.2-2, Third Containment Inservice Inspection Interval

Unit I(Notes 1, 2)


1St Period 2 nd Period 3 rd PeriodOutage 1 (EOC 23) Outage 3 (EOC 25) Outage 6 (EOC 28)Outage 2 (EOC 24) Outage 4 (EOC 26) Outage 7 (EOC 29)

Outage 5 (EOC 27)

(1) The 2nd Interval end date was extended from 07/15/2014 to 8/31/2014 within the 12months of 07/15/2014 and less than 11 years of 07/15/2004 as allowed by IWA-2430(d)(1), (Ref. ASME Code 1998 Edition with 2000 Addenda). Therefore, the startdate for 3rd Interval is 09/01/2014.

(2) The end date for Interval 3 is adjusted to 10 years from the original end date 07/15/2014of the 2nd Interval as allowed by IWA-2430(c)(3), (Ref. ASME Code 2007 Edition with2008 Addenda).

Encdosure 1Page 14 of 34

Table 3.2.2-3, Second Containment Inservice Inspection Interval

Unit 2(Notes 1, 2)


1st Period 2nd 3 rdPeriod Period

Outage 1 (EOC 17) Outage 3 (EOC 19) Outage 5 (EOC 21 )Outage 2 (EOC 18) Outage 4 (EOC 20) Outage 6 (EOC 22)

(1) The schedule for Interval 2 is included in this discussion as it encompasses the previousUnit 2 ILRT.

(2) The start date for Interval 2 was delayed from 07/15/04 to 07/15/05, within the 12-monthadjustment allowed by IWA-2430(d)(1). This adjustment was necessary because of thedelay in obtaining NRC approval of Relief Request Serial #03-GO-010.

Table 3.2.2-4, Third Containment Inservice Inspection Interval

Unit 2(Notes 1, 2)


1 st Period 2 nd Period 3 rd PeriodOutage 1 (EOC 23) Outage 3 (EOC 25) Outage 5 (EOC 27)Outage 2 (EOC 24) Outage 4 (EOC 26) Outage 6 (EOC 28)

(1) The 2nd Interval end date was extended from 07/15/2014 to 8/31/2014 within the 12months of 07/15/2014 and less than 11 years of 07/15/2004 as allowed by IWA-2430(d)(1), (Ref. ASME Code 1998 Edition with 2000 Addenda). Therefore, the startdate for 3rd Interval is 09/01/2014.

(2) The end date for Interval 3 is adjusted to 10 years from the original end date 07/15/2014of the 2nd Interval as allowed by IWA-2430(c)(3), (Ref. ASME Code 2007 Edition with2008 Addenda).

Examination Cate~qories and Requirements

The examination categories to be used are those listed in Table IWE-2500- I of Section XI.Specific examinations will be identified by an Item Number similar to those listed in Table IWE-2500-1 of Section XI plus an additional number to uniquely identify that examination (except asnoted below). (Example: E01 .011.001). Class MC Items to be inspected include the following:


Table 3.2.2-5, Category E-A: Containment Surfaces

Table IWE-2500-1 Item Part To Be Examined Examination Requirement(s)Comments

El1.10 Containment Vessel - Pressure Retaining Boundary

El1.11 Accessible Surface Areas General Visual,100% each Inspection Period(See Note 1)

El1.12 Wetted Surfaces of (See Note 2)Submerged Areas ______________

El1.30 Moisture Barriers General Visual,100% each Inspection Period

(1) If this examination is to be credited towards satisfying the examinations required by 10CFR 50, Appendix J, the examination shall be performed during the REQ in which aType A test is to be performed, just prior to the start of the Type A test. Duke Energy.Corporation intends to credit Item El1.11 visual exams towards satisfying therequirements of 10 CER 50, Appendix J.

(2) These examinations are applicable only to BWR containments and are not applicable atMNS.

Table 3.2.2-6, Category E-C: Containment Surfaces Requiring Augmented Examination

Table IWE-2500-1 Item IPart To Be Examined Examination Requirement(s)

_______________________ j___________________________CommentsE4.10 Containment Surface Areas

E4.1 1 Visible Surfaces VT-I Visual, 100% Each Period

(Deferral Not Permissible)

E4.12 Surface Area Grid, Minimum Volumetric, UltrasonicWall Thickness MeasurementThickness Location 100% Each Inspection Period

(Deferral Not Permissible)


Table 3.2.2-7, Category E-G: Pressure Retaining Bolting

Table IWE-2500-1 Part To Be Examined Examination Requirement(s)Item Comments

E8.10 Bolted Connections Visual VT-i,100% End of Interval DeferralPermissible

Table 3.2.2-8, Category F-A: Supports

Table IWF-2500-1 Part To Be Examined Examination Requirement(s)Item Comments

F1 .40 Supports Other Than Piping Visual VT-3,Supports (Class MC Airlock 100% End of Interval

_____________Supports) (See Notes)

(1) Although 10 CFR 50.55a does not include requirements for inservice or preserviceexamination of Class MC component supports; these supports shall be included withinthe Containment ISI Plan and shall be examined in accordance with Subsection IWF ofthe Code.

(2) The provisions of Table IWF-2500-1, footnote (3) may be used to minimizeexamination of multiple component supports on each airlock.

Owner Specified Examination Reqiuirements

Owner specified examination requirements shall be performed as specified in this Plan. Specificexamination listings and schedules are described in Part A, Section 6 of this Plan. Note that theterm "Augmented Examinations" is not used in this plan to describe examinations that areabove and beyond those required by the Code. An alternative term "Owner SpecifiedExaminations" is used to alleviate confusion with Subsection IWE, Category E-C AugmentedExaminations. Owner specified examinations may include those which are the result ofregulatory commitments, those required solely by regulation, and may include otherexaminations deemed appropriate by the Owner for inclusion in this program.

Listing of Owner-Specified Examinations:

1. VT-3 Visual Examination of Fuel Transfer Tube penetration surfaces (includingaccessible surfaces of the Fuel Transfer Tube Penetration) on the exterior of thecontainment (Annulus side). These surfaces are not readily accessible because of leadshielding and locked access ports. These areas are posted as Very High Radiation(Grave Danger); however, these conditions exist only during fuel movement. Theselocations are identified because they are not routinely accessed for general visual


examination in accordance with the Containment ISl Program. A review of similarindustry practices confirmed that many licensees consider access to these types ofareas to be sufficiently difficult as to justify exemption from the General VisualExamination requirements of the ASME Code, Subsection IWE. The operatingexperience at MNS has not identified any specific concerns with these areas that wouldwarrant changing the Containment I5l Program to require General Visual Examination ofthese areas. However, it is deemed appropriate to require that these surfaces receive aVT-3 Visual Examination once during each ten-year interval.

2. Ultrasonic Thickness Measurements on selected surfaces opposite the ice condenserareas once every ten-year interval. The number and extent of these examinations shallbe determined by Engineering. At a minimum, some examinations should be performedevery ten years opposite the Ice Condenser Floor Slab/containment vessel interface.Additional UT examinations may be warranted at locations near the Ice Condenser TopDeck Doors where condensation has been occasionally observed on the Annulus side ofthe vessel during scheduled examinations. If conditions are detected during theperformance of these examinations, a determination shall be made as to whether theconditions warrant examination of the affected surfaces under Category E-C, ItemE4. 12.

3. VT-I Visual Examination of selected surface areas along the embedment zone on theinterior side of the Containment Vessel, near elevation 725+0 (Nom.), located behindcontainment vessel thermal insulation panels. The number and location of selectedsurface areas is based on engineering judgment of the Containment ISI ProgramManager.

4. NRC Information Notice 2014-07 addresses NRC expectations for examination ofcontainment leak chase channel systems. These items shall be added to theContainment ISI Plan as Elective Examinations to be performed 100% during eachInspection Interval (instead of the 100% each Inspection Period as indicated in IN 2014-07). Duke Energy shall elect to perform a VT-3 visual examination in accordance withprocedure NDE-67. These examinations may be scheduled and performed as follows:

* 100% of the containment interior concrete floors shall be examined during eachinspection interval. Appropriately 1/3 of the floor surface areas shall be examinedduring each inspection period to determine the condition of all leak chase channelbronze caps installed in the floor within the examination area.

* The condition of the bronze caps shall be considered acceptable if there is noevidence of damage or degradation that could result in possible leakage of water intothe leakage channel systems. Evidence of boric acid in the vicinity of any bronzecap shall require evaluation by engineering to determine the acceptability of the item.

* If, during the visual examination of the leak chase channel covers, it is determinedthat there is evidence of moisture intrusion or degradation of the cover to the extentthat moisture intrusion could have occurred, then removal of the leak chase channelcover shall be necessary to assess the condition of the embedded containment linerplate within the leak chase channel. Examination of leak chase channel tubing andembedded containment liner plate using borescope would then be mandatory, andthe observed conditions shall be considered reportable per 10 CFR50.55a(b)(2)(ix)(A).


•The specific location Test No. of each bronze cap shall be identified on the

examination record.

Owner-Specified Examination Categories and Requirements

Owner specified examinations are described in the following table.

Table 3.2.2-9, Owner Specified Examinations

Item Subject to Part to Be Examined Examination Requirement /Owner Specified Requirement(s) CommentsExamination

El1.11 Accessible Surfaces General Visual None. These surfacesof Process Piping Examination do not requirePenetration Each Period examination inAssemblies (See Note 2) accordance with IWCClassified as ASME or IWE. However,Code Class NC as these surfaces form ashown on in service part of the containmentinspection drawings. primary pressure(See Note 1) boundary.

E4.12 Surface Area Grid, Volumetric None. (See Note 3)Minimum Wall (See Note 3)Thickness Location

Fl1.40 Class MC Component Visual, VT -3 None. However, AirlockSupports supports at McGuire

shall be treated as NEsupports and shall beexamined inaccordance with IWFrequirements at theOwner's discretion.

El1.30 Accessible Surfaces General Visual See Note 4.of 100% of ExaminationContainment Leak Each InspectionChase Channel Period

______________ Closures__________ _ ____________

Notes:

1. Penetration subassemblies (bellows) are classified as Code Class NC at their weldedconnection to the first outboard circumferential weld on the containment vessel sleeve.However, the Containment ISI Plan drawings indicate that pressure retaining penetrationsubassembly surfaces are included within the scope of IWE Examination from the


containment sleeve circumferential weld to the bellows assembly seal weld at the fluedhead connection. These are Owner specified examinations and need not comply withapplicable provisions of IWE or 10 CFR 50.55a(b)(2) and 10 CFR 50.55a(g)(4).

2. Visual Examination shall be performed on accessible surfaces of areas shown on theContainment 15l drawings, except that surfaces of subassemblies extending outside ofthe Reactor Shield Building need not be examined during general visual examinationsspecified for Item E1.11.

3. IWE-2500, Table IWE-2500-1, Category E-C, Item E4.12 examinations are performed todetermine the minimum wall thickness location within each grid. The examinationsspecified herein as "Owner Specified" are in addition to those required by the Code andare performed within each grid subject to E4.12 examination when the minimum wallthickness measurement initially recorded lies within 4" of the centerline of any vertical orcircumferential weld. These Owner specified examinations are not required by theCode, but have been added by the Owner to ensure that the minimum wall thicknessmeasurements within each grid are not influenced by local plate thinning that typicallycan occur adjacent to welds. These examinations have been added to be consistentwith procedure NDE-951, "Ultrasonic Thickness Measurement of Metallic ContainmentStructures".

4. A VT-3 visual examination in accordance with NDE-67, "Visual Examination (V/T-1 andVT-3) of Metal and Concrete Containment," shall be specified to satisfy theGeneral Visual Examination.

Additional Progqram Requirements

Additional programmatic requirements specified by 10 CFR 50.55a(b)(2)(ix) are describedbelow:

10 CFR 50.55a(b)(2)(ix)(A)

For Class MC applications, the licensee shall evaluate the acceptability of inaccessible areaswhen conditions exist in accessible areas that could indicate the presence of or result indegradation to such inaccessible areas. For each inaccessible area identified, the licenseeshall provide the following in the IS! Summary Report as required by IWA-6000:

*A description of the type and estimated extent of degradation, and the conditions that led

to the degradation;

*An evaluation of each area, and the result of the evaluation; and

• A description of necessary corrective actions.

10 CFR 50.55a(b)(2)(ix)(B)

When performing remotely the visual examinations required by Subsection IWE, the maximumdirect examination distance specified in Table IWA-2210-1 may be extended and the minimumillumination requirements specified in Table IWA-2210-1 may be decreased provided that the

Enclosure IPage 20 of 34

conditions or indications for which the visual examination is performed can be detected at thechosen distance and illumination.

ASME Code 2007 Edition with 2008 Addenda has no Table IWA-2210-I. Duke Energy haselected to comply with the requirements of ASME Code 2007 Edition with 2008 Addenda,Subsection IWA for performance of VT-i and VT-3 visual examinations on metal containment.

1 0 CFR 50.55a(b)(2)(ix)(J)

In general, a repair/replacement activity such as replacing a large containment penetration,cutting a large construction opening in the containment pressure boundary to replace steamgenerators, reactor vessel heads, pressurizers, or other major equipment; or other similarmodification is considered a major containment modification. When applying IWE-5000 to ClassMC pressure-retaining components, any major containment modification or repair/replacement,must be followed by a Type A test to provide assurance of both containment structural integrityand leak-tight integrity prior to returning to service, in accordance with 10 CFR 50, Appendix J,Option A or Option B on which the applicant's or licensee's Containment Leak-Rate TestingProgram is based. When applying IWE-5000, if a Type A, B, or C Test is performed, the testpressure and acceptance standard for the test must be in accordance with 1 0 CFR 50,Appendix J.

This condition modifies the requirements of IWE-5000 for major containment modification orrepair/replacement activities. These activities must be followed by a 10 CFR 50 Appendix J,type A test. This eliminates the bubble test-vacuum box technique detailed in IWE-5224following major containment repair/replacement activities.

Accessible Surface Areas

The provisions of IWE-1231(a)(3) shall be met. This paragraph describes how the total

accessible surface area is to be determined to meet this provision.

The total surface area of the containment (in square feet) was computed for the First IntervalContainment 1SI Plan. Based on the data and computations documented during 131 Interval 1, itis clear that the requirements of IWE-1 231(a) have been met during Interval 1 and theexamination surface areas were the same during Interval 2. Because the examination surfaceareas are the same during Interval 3, these computations need not be repeated in the ThirdInterval Containment ISI Plan. It is also clear that, unless major modifications are made to theContainment Vessel, the minimum required areas will remain accessible for visual examinationfrom at least one side of the vessel.

Inaccessible Surface Area

The inaccessible surface area is equal to the total surface area obstructed within the areasidentified above such that direct or remote visual examination of that area is not possible fromboth the inside and outside surface. An area need not be considered inaccessible if accesspermits examination from at least one side of the vessel. However, the location and extent of allsurface areas where access is not possible from one side shall be documented. The First andSecond Intervals Containment ISI Plan did not identify a significant number of inaccessiblesurface areas, and additional inaccessible surface areas may be documented during theinspection intervals. If additional inaccessible areas are identified during the third ISI Interval,the ISI Plan shall be updated to document this data and shall be revised as needed to


demonstrate continued compliance with IWE-1231(a)(3). Data on the location and extent ofinaccessible surface areas shall be of sufficient detail as to allow these areas to be identifiedand located on applicable inservice inspection drawings referenced in this Plan.

Requests for Relief from ASME Code and Regulatory Requirements

Currently there are no relief requests for the third interval plan.

3.2.3 License Renewal

The Containment Inservice Inspection Plan - IWE is credited in the joint Catawba and McGuireLicense Renewal Application (LIRA) with managing loss of material due to corrosion of steelsurfaces. The Containment Inservice Inspection Plan - IWE implements the requirements ofASME Code Section Xl Subsections IWE. The purpose of ASME Subsection IWE examinationis to identify and correct degradation of the accessible steel surfaces of the containment linerprior to the loss of the essentially leak tight barrier.

Section 3.5 of the LRA identified the Containment ISI Plan - IWE as an aging managementprogram for Reactor Building containment steel components. The discussion of the programand objective evidence associated with the effectiveness of the program were provided in LRAAppendix B.3.7 of the Catawba and McGuire application.

The Containment Leak Rate Testing Program is credited in the Catawba and McGuire LRA withmanaging loss of material of steel components of the Reactor Building Containment andcracking of penetration bellows. The purpose of the Containment Leak Rate Testing Program isto assure that leakage through the containment and systems and components penetratingcontainment shall not exceed allowable leakage rate values specified in the TechnicalSpecifications or associated bases and periodic surveillances of containment penetrations andisolation valves are performed.

Section 3.5 of the LRA identified the Containment Leak Rate Testing Program as an agingmanagement program for steel components of the Reactor Building Containment. Thediscussion of the program and objective evidence associated with the effectiveness of theprogram were provided in LRA Appendix B.3.8 of the Catawba and McGuire application..

3.2.4 Integrated Leakage Rate Testinq (ILRT) History

Previous Type A tests confirmed that the MNS reactor containment structure has leakage wellunder acceptance limits and represents minimal risk to increased leakage. Continued Type Band Type C testing for direct communication with containment atmosphere minimize this risk.Also, the Inservice Inspection (IWE/IWL) program and maintenance rule monitoring provideconfidence in containment integrity.

To date, five operational Type A tests have been performed on MNS Unit I and four on MNSUnit 2. There is considerable margin between these Type A test results and the TS 5.5.2 limit of0.75 La (0.225 % Weight per Day), where La is equal to 0.3% by weight of the containment airper day at the peak accident pressure. These test results demonstrate that MNS has lowleakage Containments.

Encl osure IPage 22 of 34

Table 3.2.4-1, MNS Unit I ILRT Test Results

Test Date Leakage weight % per dayApril 1983 0.1446

August 1986 0.1533May 1990 0.1965May 1993 0.1482

October 2008 0.1065

Table 3.2.4-2, MNS Unit 2 ILRT Test Results

Test Date Leakage weight % per day

May 1986 0.0837August 1989 0.1138August 1993 0.1469March 2008 0.1242

3.3 Containment Leakage Rate Testing Program, Type B and Type C Testing

MNS Type B and C testing program currently requires testing of electrical penetrations, airlocks,hatches, flanges, bellows, and containment isolation valves in accordance with 10 CFR Part 50,Appendix J, Option B. The results of the test program are used to demonstrate that propermaintenance and repairs are made on these components throughout their service life. TheType B and C testing program provides a means to protect the health and safety of plantpersonnel and the public by maintaining leakage from these components below appropriatelimits. Per TS 5.5.2, the allowable maximum pathway total for Type B and C leakage is 0.6 Lawhere La equals 140,379 sccm.

As discussed in NUREG-1 493, Type B and Type C tests can identify the vast majority of allpotential Containment leakage paths. Type B and Type C testing will continue to provide a highdegree of assurance that containment integrity is maintained.

A review of the Type B and Type C test results from 2008 through 2014 for MNS Unit 1 andfrom 2009 through 2015 for MNS Unit 2 has shown an exceptional amount of margin betweenthe actual As-Found (AF) and As-left (AL) outage summations and the regulatory requirementsas described below:

* The As-Found minimum pathway leak rate average for MNS Unit 1 shows an average of5.97% of La with a high of 7.884%La.

* The As-Left maximum pathway leak rate average for MNS Unit 1 shows an average of7.09% of La with a high of 8.416% La.


* The As-Found minimum pathway leak rate average for MNS Unit 2 shows an average of1.40% of La with a high of 2.603% La.

* The As-Left maximum pathway leak rate average for MNS Unit 2 shows an average of2.03% of La with a high of 3.188% La.

Tables 3.3-1 and 3.3-2 provide LLRT data trend summaries for MNS since 2008 for Unit 1 and2009 Unit 2 . This summary shows that there have been no As-Found failures that resulted inexceeding the Technical Specification 5.5.2 limit of 0.6 La and demonstrates a history ofsuccessful tests.

The followingperformance.

demonstrates a history of satisfactory Type B and Type C tested component

Table 3.3-1, MNS Unit 1 Type B and C LLRT Trend Summary

RFO 2008 2010 2011 2013 2014MIEOC19 MIEOC20 MIEOC21 MIEOC22 MIEOC23

(sccrn) (sccm) (sccm) (sccm) (sccm)As-FoundPathway 6049.2 8338.2 11068 7589 8871.4

Pecn f 4.309 5.940 7.884 5.406 6.320La

As-LeftPathway 9277 11155.8 11814 10196 7313.6

Pecn f 6.608 7.947 8.416 7.263 5.210La

Encdosure 1Page 24 of 34

Table 3.3-2, MNS Unit 2 Type B and C LLRT Trend Summary

RFO 2009 2011 2012 2014 2015M2EOC19 M2EOC20 M2EOC21 M2EOC22 M2EOC23

(sccm) (sccm) (sccm) (sccm) (sccm)As-FoundPathway 1357.3 1793.94 1420.3 3654.71 1612.3

Percentage 0.967 1.278 1.012 2.604 1.149of La

As-LeftPathway 4475.21 3208.69 1383 2526.24 2649.7

Percentage 3.188 2.268 0.985 1.800 1.888of La

3.4 NRC Information Notices (INs)

3.4.1 IN 201 0-1 2, "Containment Liner Corrosion"

This Information Notice was issued to alert plant operators to three events that occurred wherethe steel liner of the containment building was corroded and degraded. At Beaver Valley andBrunswick plants material had been found in the concrete which trapped moisture against theliner plate and corroded the steel. In one case, it was material intentionally placed in thebuilding, and in the other case, it was foreign material which had inadvertently been left in theform when the wall was poured. The result in both cases was that the material trapped moistureagainst the steel liner plate leading to corrosion. In the third case, Salem, an insulating materialplaced between the concrete floor and the steel liner plate adsorbed moisture and led tocorrosion of the liner plate.

Discussion:

All the referenced examples are from sites with Concrete Primary Containments with steelliners. In a Concrete Containment, the liner, in addition to helping provide a leak tight barrier,acts as the inner "form" when the concrete is poured into a containment structure. In thesecases, the liner and associated sleeves and penetrations are in direct contact with the concrete,or are in contact with intermediate materials installed during construction that communicatebetween the liner and concrete.

Duke Energy conducted a focused self-assessment of Containment Integrity in 2008. Theassessment reviewed the Containment ISI (ASME BPV Code Section XI) Program, the BoricAcid Corrosion Program, the 10 CFR 50 Appendix J Program, and the Coatings Program, aswell as Operating Experience Data for three Duke Energy Nuclear sites, Oconee, McGuire andCatawba. The McGuire specific concerns identified in the report were tracked in the CorrectiveAction Program.


The assessment found that overall the Programs were satisfying the purposes for which theyhad been established; however, there were some risks identified in the assessment that couldjustify changes to the programs. The most significant risks to the McGuire plant were listed asthe Unit 2 Fuel Transfer Tube Area and the Thermal Insulation Panels. After this assessmentwas complete, but before all the Corrective Actions and the assessment were closed, the NRCissued IN 2010-12. Due to the related nature of the issues, the response to the IN was includedinto a new Corrective Action, which contained the assessment and the responses. The newCorrective Action was to process an Engineering Change Request, which led to an EngineeringChange to provide details for an easy access inspection port through the Thermal Barrier on theSteel Containment Vessel. This EC supported the needed ISI inspection where the access tothe surface of the containment plate is not accessible.

No further actions were recommended based on evaluation of the OE provided in NRCInformation Notice 201 0-1 2.

3.4.2 IN 2014-07. Degqradation of Leak-Chase Channel Systems for Floor Welds of MetalContainment Shell and Concrete Containment Metallic Liner

The containment basemat metallic shell and liner plate seam welds of pressurized waterreactors are embedded in a 3-foot to 4-foot thick concrete floor during construction and aretypically covered by a leak-chase channel system that incorporates pressurizing testconnections. This system allows for pressure testing of the seam welds for leak-tightnessduring construction and also in service, as required. A typical basemat shell or liner weld leak-chase channel system consists of steel channel sections that are fillet welded continuously overthe entire bottom shell or liner seam welds and subdivided into zones, each zone with a testconnection.

Each test connection consists of a small carbon or stainless steel tube (less than 1-inchdiameter) that penetrates through the back of the channel and is seal-welded to the channelsteel. The tube extends up through the concrete floor slab to a small steel access (junction) boxembedded in the floor slab. The steel tube, which may be encased in a pipe, projects upthrough the bottom of the access box with a threaded coupling connection welded to the top ofthe tube, allowing for pressurization of the leak-chase channel. After the initial tests, steelthreaded plugs or caps are installed in the test tap to seal the leak-chase volume. Gasketedcover plates or countersunk plugs are attached to the top of the access box flush with thecontainment floor. In some cases, the leak-chase channels with plugged test connections mayextend vertically along the circumference of the cylindrical containment shell or liner to a certainheight above the floor.

Discussion:

During a 2EOC23 containment walkdown in response to IN 2014-07, four Steel ContainmentVessel Base Liner Plate Leak Chase Test Channel enclosure locations, #45, #89, #144 and#61, in the concrete floor were discovered to be concrete-type filled and missing the bronzecovers (caps) designed to enclose the test ports. The sealed as-found condition of these fourtest channel enclosures indicates that no leakage into the test channels below has occurred.

During a 1EOC23 containment walkdown in response to IN 2014-07, one location, #86, wasfound without a bronze cover during 1 EOC23 (AR 1686595). Similar to the Unit 2 as foundcondition, the test port was sealed with a hard, solid material obstructing the tube inner


diameter. There was a limited extent of condition performed for other test channel enclosures inthe area (approximately 10%) with no adverse findings.

An open test port could permit moisture/foreign material intrusion to the embedded containmentbase liner plate surfaces within the test channels. The sealed as-found condition of the abovelisted test channel enclosures indicated that no leakage into the test channels below hadoccurred.

Specified Safety Function

The function of the cover and seal on the test channel enclosures is to protect the steelcontainment vessel (SCV) from moisture intrusion. The following Technical Specifications areapplicable to the SCV system:

* ITS 3.6.1, Containment* ITS 5.5.2, Containment Leak Rate Testing Program

The function of the Steel Containment Vessel (SCV) Structure is to provide an essentially leak-tight barrier against the release of radioactivity and energy that may result from a Design BasisEvent. The steel containment and its penetrations establish the leakage limiting boundary of thecontainment. Maintaining the containment operable, limits the leakage of fission productradioactivity from the containment to the environment.

Testing of the Base Liner Plate welds using the test channel ports was performed during originalconstruction; afterwards, the test channel enclosures and test ports were sealed and covered toprotect the liner plate from moisture and foreign material intrusion. Foreign material intrusionalone does not introduce a degradation mechanism to the SCV. Continued containmentintegrity is currently monitored via the Containment Leak Rate Testing Program.

Unit 2 Evaluation

There is reasonable assurance that moisture intrusion affecting the integrity of the SCV has not

occurred, as noted below:

*For the four Unit 2 enclosures found without covers, the leak chase channel test tubingis sufficiently sealed by concrete and/or grout and additional coatings over two of thecovers; therefore, there is no credible evidence that moisture intrusion has occurred intothe embedded portions of the containment liner plate within the affected leak chase testchannels since original construction. Efforts were made to expose the 1/2 inch testchannel tubing for further interrogation, however the materials sealing the test channelport were unable to be removed without causing significant damage to the leak chaseenclosure.

*For the remaining enclosures with covers installed per the design drawings, the designof the leak chase enclosure cover is such that once torqued into place the cover with theseal provides an adequate barrier to moisture intrusion. The sealant material iscompressed between a metal sleeve and the bronze cover. On many of the enclosures,the epoxy floor coating covered the joint interface between the bronze cover andenclosure sleeve which provides an additional barrier.

Enclosure 1Pagle 27 of 34

* Any legacy moisture/foreign material intrusion into the test channel from originalconstruction would have become anoxic, and thus any degradation would be self-limiting.

* ILRT data shows containment is sufficiently sealed and adequate margin exists. TheMcGuire Unit 2 ILRT mass point leakage history confirms there is not degradation of thesteel containment liner.

There remains a large margin between the latest 2008 ILRT results at 0.1242% per day,and the limit of 0.225% per day. Note that the bronze caps are not removed whenperforming ILRT testing.

*Industry operating experience shows similar or more significant findings, such asmissing plugs, completely open ports, evidence of boric acid, and channels filled withwater; none of which resulted in degradation of the base liner plate that challengedoperability.

Based on the discussion above, there is reasonable assurance that the McGuire Unit 2 SCVBase Liner Plate is not degraded; and therefore, the SCV is capable of performing its specifiedsafety function. This evaluation is also applicable and bounding to McGuire Unit 1.

Extent of Condition

Unit 2All 142 test enclosure locations have been dispositioned (identified or deemed inaccessible) andthe four locations with missing bronze covers were addressed herein. It is planned to performqualified visual inspection of all accessible bronze covers during the 2EOC24 (Spring 2017)refueling outage per the guidance in NRC IN 2014-07. This inspection is intended to verify thatthe bronze covers remain adequately sealed. In response to NRC IN 2014-07, theseinspections have been added to the ISI plan.

Unit 1One location, #86, was found without a bronze cover during 1 EOC23. Similar to the Unit 2 asfound condition, the test port was sealed with a hard solid material obstructing the tube innerdiameter. There was a limited extent of condition performed for other test channel enclosures inthe area (approximately 10%) with no adverse findings. Unit 1 ILRT data was reviewed withlarge margin between the latest 2008 ILRT results at 0.1065% per day, and the limit of 0.225%per day. There is reasonable assurance that all Unit 1 test channel enclosures are coveredand/or are otherwise providing an adequate moisture barrier function. An investigation toidentify all Unit 1 test channel enclosures is planned for the 1 EOC24 (Fall 2017) refuelingoutage. This evaluation is applicable and bounding to Unit 1.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/ Criteria

The proposed change has been evaluated to determine whether applicable regulations andrequirements continue to be met. 10 CFR 50.54(o) requires primary reactor containments forwater-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR Part 50,"Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants." Appendix J


specifies containment leakage testing requirements, including the types required to ensure theleak-tight integrity of the primary reactor containment and systems and components whichpenetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria,test methodology, frequency of testing, and reporting requirements for each type of test.

RG 1.163 was developed to endorse NEI 94-01, Revision 0 with certain modificationsand additions.

The adoption of the Option B performance-based containment leakage rate testing for Type A,Type B, and Type C testing did not alter the basic method by which Appendix J leakage ratetesting is performed; however, it did alter the frequency at which Type A, Type B, and Type Ccontainment leakage tests must be performed. Under the performance-based option of 10 CFRPart 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found"leakage history to determine the frequency for leakage testing which provides assurance thatleakage limits will be maintained. The change to the Type A test frequency did not directlyresult in an increase in containment leakage.

EPRI TR-1 009325, Revision 2, provided a risk impact assessment for optimized ILRT intervalsup to 15 years, utilizing current industry performance data and risk informed guidance. NEI 94-01, Revision 2-A, Section 9.2.3.1 states that Type A ILRT intervals of up to 15 years are allowedby this guideline. The Risk Impact Assessment of Extended Integrated Leak Rate TestingIntervals, EPRI Report 1018243 (Formerly TR-1 009325, Revision 2) indicates that, in general,the risk impact associated with ILRT interval extensions for intervals up to 15 years is small.However, plant-specific confirmatory analyses are required.

The NRC staff reviewed NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2.For NEI TR 94-01, Revision 2, the NRC staff determined that it described an acceptableapproach for implementing the optional performance-based requirements of Option B to 10 CFR50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals to up to15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff found thatthe Type A testing methodology as described in ANSI!/ANS-56.8-2002, and the modified testingfrequencies recommended by NEI TR 94- 01, Revision 2, serves to ensure continued leakageintegrity of the containment structure. Type B and Type C testing ensures that individualpenetrations are essentially leak tight.

In addition, aggregate Type B and Type C leakage rates support the leakage tightness ofprimary containment by minimizing potential leakage paths.

4.2 Precedent

This license amendment request is similar in nature to the following license amendments

previously approved by the NRC to extend the Type A test frequency:

* December 29, 1994 (ML01 1080782), for Nine Mile Point Nuclear Station Unit 1,* June 2, 2003 (ML031 320686), for Vermont Yankee Nuclear Power Station,* July 20, 2009 (ML091 540158) for Arkansas Nuclear One, Unit No. 2,* August 23, 2010 (ML1 02090137) for Palisades Nuclear Plant.* October 1, 2012 (ML12250A339) for Oconee Nuclear Station Unit 1.* August 5, 2013 (ML13193A329) for Oconee Nuclear Station Units 2 and 3.


4.3 Significant Hazards Consideration

Duke Energy has evaluated whether or not a significant hazards consideration is involved withthe proposed amendment by focusing on the three standards set forth in 10 CFR 50.92,"Issuance of amendment," as discussed below. There is no significant hazards considerationinvolved with the proposed corrections to typographical and administrative errors introduced byprevious license amendments.

1. Does the proposed amendment involve a significant increase in the probability orconsequences of an accident previously evaluated?

Response: No.

The proposed amendment to the Technical Specifications (TS) involves the extension ofthe McGuire Nuclear Station (MNS) Type A containment integrated leak rate test intervalto 10.5 years. The current Type A test interval of 120 months (10 years) would beextended on a one-time basis to no longer than 10.5 years from the last Type A test.This extension is bounded by the 15 month extension, permissible only for non-routineemergent conditions, allowed in accordance with NEI 94-01 revision 0. The proposedextension also does not change the test method or procedure. The containment isdesigned to provide an essentially leak tight barrier against the uncontrolled release ofradioactivity to the environment for postulated accidents. The containment and thetesting requirements invoked to periodically demonstrate the integrity of the containmentexist to ensure the plant's ability to mitigate the consequences of an accident, and do notinvolve the prevention or identification of any precursors of an accident. The change indose risk for changing the Type A test frequency from 10 years to 10.5 years, measured,as an increase to the total integrated plant risk for those accident sequences influencedby Type A testing, is 0.023 person-remn/year. EPRI Report No. 1009325, Revision 2-Astates that a very small population dose is defined as an increase of -- 1.0 person-remnper year, or -< 1 % of the total population dose, whichever is less restrictive for the riskimpact assessment of the extended ILRT intervals. Therefore, this proposed extensiondoes not involve a significant increase in the probability of an accident previouslyevaluated.

As documented in NUREG-1 493, Performance-Based Containment Leak-Test Program,Type B and C tests have identified a very large percentage of containment leakagepaths, and the percentage of containment leakage paths that are detected only by TypeA testing is very small. The MNS Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that canbe categorized as: (1) activity based, and; (2) time based as previously discussed.Activity based failure mechanisms are defined as degradation due to system and/orcomponent modifications or maintenance. Local leak rate test requirements andadministrative controls such as configuration management and procedural requirementsfor system restoration ensure that containment integrity is not degraded by plantmodifications or maintenance activities. The design and construction requirements ofthe containment combined with the containment inspections performed in accordancewith ASME Section XI, the Maintenance Rule, and TS requirements serve to provide ahigh degree of assurance that the containment would not degrade in a manner that isdetectable only by a Type A test. Based on the above, the proposed extensions do notsignificantly increase the consequences of an accident previously evaluated.


Therefore, the proposed change does not result in a significant increase in theprobability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accidentfrom any accident previously evaluated?

Response: No.

The proposed amendment to the TS involves the extension of the MNS Type Acontainment integrated leak rate test interval from 10 years to 10.5 years. The currentType A test interval of 120 months (10 years) would be extended on a one-time basis to10.5 years from the last Type A test. The containment and the testing requirements toperiodically demonstrate the integrity of the containment exist to ensure the plant's abilityto mitigate the consequences of an accident do not involve any accident precursors orinitiators. The proposed change does not involve a physical change to the plant (i.e., nonew or different type of equipment will be installed) or a change to the manner in whichthe plant is operated or controlled.

Therefore, the proposed change does not create the possibility of a new or different kind

of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment to TS 5.5.2 involves the extension of the MNS Type Acontainment integrated leak rate test interval to 10.5 years. The current Type A testinterval of 120 months (10 years) would be extended on a one-time basis to no longerthan 10.5 years from the last Type A test. This amendment does not alter the manner inwhich safety limits, limiting safety system set points, or limiting conditions for operationare determined. The specific requirements and conditions of the TS Containment LeakRate Testing Program exist to ensure that the degree of containment structural integrityand leak tightness that is considered in the plant safety analysis is maintained. Theoverall containment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type Acontainment leak rate tests for MNS. The proposed surveillance interval extension isbounded by the 15-year ILRT interval currently authorized within NEI 94-01, Revisions 2-A and 3-A. Industry experience supports the conclusion that Type B and C testingdetects a large percentage of containment leakage paths and that the percentage ofcontainment leakage paths that are detected only by Type A testing is small. Thecontainment inspections performed in accordance with ASME Section Xl, and TS serveto provide a high degree of assurance that the containment would not degrade in amanner that is detectable only by Type A testing. The combination of these factorsensures that the margin of safety in the plant safety analysis is maintained. The design,operation, testing methods and acceptance criteria for Type A, B, and C containmentleakage tests specified in applicable codes and standards would continue to be met, withthe approval of this proposed change, since these are not affected by changes to theType A test intervals.


Therefore, the proposed change does not involve a significant reduction in a margin ofsafety.

Based on the above, Duke Energy concludes that the proposed amendment does not involve asignificant hazards consideration under the standards set forth in 10 CFR 50.92(c), and,accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion

In conclusion, based on the considerations discussed above, (1) there is reasonable assurancethat the health and safety of the public will not be endangered by operation in the proposedmanner, (2) such activities will be conducted in compliance with the Commission's regulations,and (3) the issuance of the amendment will not be inimical to the common defense and securityor to the health and safety of the public.

10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactorsto be subject to the requirements of Appendix J to 10 CFR 50, "Leakage Rate Testing ofContainment of Water Cooled Nuclear Power Plants." Appendix J specifies containmentleakage testing requirements, including the types required to ensure the leak-tightintegrity of the primary reactor containment and systems and components whichpenetrate the containment. In addition, Appendix J discusses leakage rate acceptancecriteria, test methodology, frequency of testing and reporting requirements for each typeof test.

RG 1.163 was developed to endorse NEI 94-01, Revision 0 with certain modificationsand additions.

The adoption of the Option B performance-based containment leakage rate testing forType A testing did not alter the basic method by which Appendix J leakage rate testing isperformed; however, it did alter the frequency at which Type A, Type B, and Type Ccontainment leakage tests must be performed. Under the performance-based option of10 CFR 50, Appendix J, the test frequency is based upon an evaluation that review"as-found" leakage history to determine the frequency for leakage testing which providesassurance that leakage limits will be maintained. The change to the Type A testfrequency did not directly result in an increase in containment leakage. Similarly, theproposed change to the Type A test frequency will not directly result in an increase incontainment leakage.

Based on the previous ILRT tests conducted at MNS, it is concluded that the extension of thecontainment ILRT interval from 10 to 10.5 years represents minimal risk to increased leakage.The risk is minimized by continued Type B and Type C testing performed in accordance withOption B of 10 CER Part 50, Appendix J and the overlapping inspection activities performed aspart of the following MNS inspection programs:

* Containment Inservice Inspection Program (IWE/IWL)

* Periodic Condition Assessments of Service Level I Coatings

* Containment Structural Integrity Inspection


This conclusion is supplemented by a plant-specific confirmatory analysis provided in Section3.1.4 of this submittal. The findings of the assessment confirm the general findings of previousstudies, on a plant-specific basis, that extending the ILRT interval from ten to 10.5 years isconsidered to be insignificant since it represents a "small" change to the MNS risk profile.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement withrespect to installation or use of a facility component located within the restricted area, as definedin 10 CFR 20, or would change an inspection or surveillance requirement. However, theproposed amendment does not involve (i) a significant hazards consideration, (ii) a significantchange in the types or significant increase in the amounts of any effluent that may be releasedoffsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion setforth in 10 CFR 5I.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impactstatement or environmental assessment need be prepared in connection with the proposedamendment.

6.0 REFERENCES

1. Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program,September 1995.

2. NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-BasedOption of 10 CFR Part 50, Appendix J, July 2012.

3. NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-BasedOption of 10 CFR Part 50, Appendix J, October 2008.

4. Regulatory Guide 1.174, Revision 2, An Approach for Using Probabilistic RiskAssessment in Risk-Informed Decisions on Plant-Specific Changes to the LicensingBasis, May 2011.

5. Regulatory Guide 1.200, Revision 2, An Approach for Determining the TechnicalAdequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March2009.

6. NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Optionof 10 CFR Part 50, Appendix J, July 1995.

7. NUREG-1 493, Performance-Based Containment Leak-Test Program, January 1995.

8. EPRI TR-1 04285, Risk Impact Assessment of Revised Containment Leak Rate TestingIntervals, August 1994.


9. Letter from M. J. Maxin (NRC) to J. C. Butler (NEI), Final Safety Evaluation for NuclearEnergy Institute (NEI) Topical Report (TR) 94-01, Revision 2, Industry Guideline forImplementing Performance-Based Option of 10 CFR Part 50, Appendix J, and ElectricPower Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, RiskImpact Assessment of Extended Integrated Leak Rate Testing Intervals (TAO No.MC9663), dated June 25, 2008.

10. ML15106A627, Letter from R. Hill (ASME) to J. Lubinski (NRC) dated April 13, 2015.ASME Code, Section XI Actions to Address Requirements for Examination ofContainment Leak-Chase Channels.

11. Letter from S. Bahadur (NRC) to B. Bradley (NEI), Final Safety Evaluation of NuclearEnergy Institute (NEI) Report 94-01, Revision 3, Industry Guideline for ImplementingPerformance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), datedJune 8, 2012.

12. ML0225401 02, Letter from R. Martin (NRC) to H. Barron (Duke), Issuance ofAmendment No. 207 to Facility Operating License NPF-9 and Amendment No. 188 toFacility Operating License NPF-17 - McGuire Nuclear Station, Units 1 and 2 (TAC Nos.MB3565 and MB3566) dated September 4, 2002.

13. ML030760032, Letter from R. Martin (NRC) to D. Jamil (Duke), Issuance ofAmendments RE: One-time Change in the Appendix J, Type A Containment IntegratedLeakage Rate Test Interval- McGuire Nuclear Station, Units 1 and 2 (TAC Nos. MB5307and MB5308) dated March 12, 2003.

14. ML031280431, Letter from R. Martin (NRC) to D. Jamil (Duke), McGuire Nuclear Station,Units 1 And 2 Re: Issuance Of Amendments (TAC NOS. MB6500 and MB6501) datedMay 8, 2003.

15. ML14261A051, Letter from J. Lubinski (NRC) to K. Ennis (ASME) dated March 3, 2015.NRC Information Notice 2014-07 Regarding Inspection of ContainmentLeak-Chase Channels

16. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision2-A of 1009325, EPRI, Palo Alto, CA. 1018243, October 2008.

17. Interim Guidance for Performing Risk Impact Assessments in Support of One-TimeExtensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Rev.4, Developed for NEI by EPRI and Data Systems and Solutions, November 2001.

18. Response to Request for Additional Information Concerning the License AmendmentRequest for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C. H.Cruse (Calvert Cliffs Nuclear Power Plant) to NRC (Document Control Desk), DocketNo. 50-31 7, dated March 27, 2002.

19. ML1 02090137, Letter from M. Chawla (NRC) to Vice President, Operations (Entergy)dated August 23, 2010. Palisades Nuclear Plant - Issuance of Amendment Re: One-Time Extension to the Integrated Leak Rate Test Interval (TAC NO. ME21 22).


20. ML13193A329, Letter from J. Boska (NRC) to S. Batson (Duke) dated August 5, 2013.Oconee Nuclear Station, Units 2 And 3, Issuance of Amendments Regarding Extensionof the Reactor Building Integrated Leak Rate Test (TAO NOS. ME9777 AND ME9778).

21. ML12250A339, Letter from J. Boska (NRC) to P. Gillespie (Duke) dated October 1,2012. Oconee Nuclear Station, Unit 1, Issuance of Amendment Regarding Extension ofthe Reactor Building Integrated Leak Rate Test (TAC NO. ME8407).

22. ML01 1080782, Letter from D. Brinkman (NRC) to R. Sylvia (Niagara Mohawk) datedDecember 29, 1994. Issuance of Amendment For Nine Mile Point Nuclear Station UnitNo. 1 (TAO NO. M90278).

23. ML031 320686, Letter from R. Pulsifer (NRC) to J. Thayer (Entergy) dated June 2, 2003.Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: One-TimeExtension of Appendix J Type A Integrated Leakage Rate Test Interval (TAO NO.MB6507).

24. ML091 540158, Letter from N. Kalyanam (NRC) to Vice President, Operations (Entergy)dated July 20, 2009. Arkansas Nuclear One, Unit No.2 - Issuance of Amendment Re:One-Time Extension to 10-Year Frequency of Integrated Leak Rate Test (TAO NO.MDg502).

25. ML073400670, Letter from J. Stang (NRC) to G. Peterson (McGuire) dated February 13,2008. McGuire Nuclear Station, Unit 1 Issuance Of Amendment Regarding Extension OfAppendix J, Type A Integrated Leakage Rate Test Interval (TAO No. MD4654).

Attachment 1

Technical Specification Pages

(Mark-up)

Programs and Manuals5.5

5.0 ADMINISTRATIVE CONTROLS

5.5 Programs and Manuals

The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (0DCM)

The 00CM shall contain the methodology and parameters used in thecalculation of offsite doses resulting from radioactive gaseous and liquideffluents, in the calculation of gaseous and liquid effluent monitoring alarm andtrip setpoints, and in the conduct of the radiological environmental monitoringprogram.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained.This documentation shall contain:

1. sufficient information to support the change(s) together with theappropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) do not adversely impact theaccuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after the approval of the Plant Manager orRadiation Protection Manager; and

c. Shall be submitted to the NRC in the form of a complete, legible copy ofthe entire ODCM as a part of or concurrent with the Radioactive EffluentRelease Report for the period of the report in which any change in theODCM was made. Each change shall be identified by markings in themargin of the affected pages, clearly indicating the area of the page thatwas changed, and shall indicate the date (i.e., month and year) the changewas implemented.

Containment Leakagqe Rate Testincq Progqram5.5.2

A program shall be established to implement the leakage rate testing of thecontainment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, OptionB, as modified by approved exemptions. This program shall be in accordancewith the guidelines contained in Regulatory. Guide 1.163, "Performance-BasedContainment Leak-Test Program," dated September 1995, as modified by thefollowing exceptions: March 31, 2008 October 21, 2008 tZ NEI 94-01-1995, Section 9.2.3: hefirst Type Atest performed after the *iay--27,

--199-3-(Unit 1 ) and Aug.ust 20, 1093 (Unit 2) Type A test shall be performed no laterthan plant restart after the End Of Cycle• Refueling Outage (Unit 1) andZ~utisfli£

-2008-(Unit 2), and IT

126 I n0la~terthanpla~nt restart after the End Of

L~ycie ZO r~etuelig uutageAmendment No. 2-7g/256-McGuire Units 1 and 2551 5.5-1


5.5 Programs and Manuals (continued)

•--:- The containment visual examinations required by Regulatory Position 0.3

shall be conducted 3 times every 10 years, including during each shutdownfor SR 3•6~-AType A test, prior to initiating the Type A test.

5.5.2 Cotanmnt; Leakage•u,, Rat Testing' Progqram ("""n"inucd,)

The peak calculated containment internal pressure for the design basis loss ofcoolant accident, Pa, is 14.8 psig. The containment design pressure is 15 psig.The maximum allowable containment leakage rate, La, at Pa, shall be 0.3% ofcontainment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is < 1.0 La. During the firstplant startup following testing in accordance with this program, the leakagerate acceptance criteria are _< 0.75 La for Type A tests and < 0.6 La forType B and Type C tests.

b. Airlock testing acceptance criteria for the overall airlock leakage rate is _<0.05 La when tested at > Pa. For each door, the leakage rate is _< 0.01 Lawhen tested at > 14.8 psig.

The provisions of SR 3.0.3 are applicable to the Containment Leakage RateTesting Program.

Nothing in these Technical Specifications shall be construed to modify the testingfrequencies required by 10CFR50, Appendix J.

5.5.3 Primary Coolant Sources Outside Containment

This program provides controls to minimize leakage from those portions ofsystems outside containment that could contain highly radioactive fluids during aserious transient or accident to levels as low as practicable. The systems includeContainment Spray, Safety Injection, Chemical and Volume Control, NuclearSampling, RHR, Boron Recycle, Refueling Water, Liquid Waste, and Waste Gas.The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at refueling cycle

intervals or less.

5.5.4 Deleted

5.5-2 Mc~uie Unis I nd 2 .5-2Amendment No. 21I2/I-I93

Attachment 2

Technical Specification Pages

(Retyped)


5.0 ADMINISTRATIVE CONTROLS

5.5 Programs and Manuals

The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

The ODCM shall contain the methodology and parameters used in thecalculation of offsite doses resulting from radioactive gaseous and liquideffluents, in the calculation of gaseous and liquid effluent monitoring alarm andtrip setpoints, and in the conduct of the radiological environmental monitoringprogram.

Licensee initiated changes to the ODCM:

a. Shall be documented and records of reviews performed shall be retained.

This documentation shall contain:

1. sufficient information to support the change(s) together with the

appropriate analyses or evaluations justifying the change(s), and

2. a determination that the change(s) do not adversely impact theaccuracy or reliability of effluent, dose, or setpoint calculations;

b. Shall become effective after the approval of the Plant Manager or

Radiation Protection Manager; and

c. Shall be submitted to the NRC in the form of a complete, legible copy ofthe entire ODCM as a part of or concurrent with the Radioactive EffluentRelease Report for the period of the report in which any change in theODCM was made. Each change shall be identified by markings in themargin of the affected pages, clearly indicating the area of the page thatwas changed, and shall indicate the date (i.e., month and year) the changewas implemented.

5.5.2 Containment Leakaqie Rate Testincq Procqram

A program shall be established to implement the leakage rate testing of thecontainment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, OptionB, as modified by approved exemptions. This program shall be in accordancewith the guidelines contained in Regulatory Guide 1.163, "Performance-BasedContainment Leak-Test Program," dated September 1995, as modified by thefollowing exceptions:

a. NEI 94-01-1995, Section 9.2.3: The first Type A test performed after theOctober 21, 2008 (Unit 1) and March 31, 2008 (Unit 2) Type A test shall beperformed no later than plant restart after the End Of Cycle 26 RefuelingOutage (Unit 1) and no later than plant restart after the End Of Cycle 25Refueling Outage (Unit 2), and

McGuire Units 1 and 2551AmnetNo 5.5-1


5.5 Programs and Manuals (continued)

b. The containment visual examinations required by Regulatory Position 0.3shall be conducted 3 times every 10 years, including during eachshutdown for SR 3.6.1.1 Type A test, prior to initiating the Type A test.

The peak calculated containment internal pressure for the design basis loss ofcoolant accident, Pa, is 14.8 psig. The containment design pressure is 15 psig.The maximum allowable containment leakage rate, La, at Pa, shall be 0.3% ofcontainment air weight per day.

Leakage Rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is < 1.0 La. During the firstplant startup following testing in accordance with this program, the leakagerate acceptance criteria are < 0.75 La for Type A tests and < 0.6 La forType B and Type C tests.

b. Airlock testing acceptance criteria for the overall airlock leakage rate is _<0.05 La when tested at > Pa. For each door, the leakage rate is < 0.01 Lawhen tested at > 14.8 psig.

The provisions of SR 3.0.3 are applicable to the Containment Leakage RateTesting Program.

Nothing in these Technical Specifications shall be construed to modify the testingfrequencies required by 10CFR50, Appendix J.

5.5.3 Primary Coolant Sources Outside Containment

This program provides controls to minimize leakage from those portions ofsystems outside containment that could contain highly radioactive fluids during aserious transient or accident to levels as low as practicable. The systems includeContainment Spray, Safety Injection, Chemical and Volume Control, NuclearSampling, RHR, Boron Recycle, Refueling Water, Liquid Waste, and Waste Gas.The program shall include the following:

a. Preventive maintenance and periodic visual inspection requirements; and

b. Integrated leak test requirements for each system at refueling cycle

intervals or less.

5.5.4 Deleted

McGuire Units 1 and 2552AmnetNo 5.5-2

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