ML15222A835

ENT000621 - Westinghouse, WCAP-17894-NP, Rev. 0, Component Inspection Details Supporting Aging Management of Reactor Internals at Indian Point Unit 2 (Sept. 2014)

ML15222A835
Person / Time
Site:	Indian Point
Issue date:	08/10/2015
From:	Entergy Nuclear Operations
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28133, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15222A835 (50)

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Westinghouse Non-Proprietary Cla ss 3 WCAP-17894-NP September 2014 Revision 0 Component Inspection Details Supporting Aging Management of Reactor Internals at Indian Point Unit 2

Westinghouse Non-Proprietary Class 3 This document may contain technical data subject to the export c ontrol law s of the United States

. In the event that this docume nt does contain such information, the Recipient's acc eptance of this doc ument constitutes agreement that this information in document form (or any other medium), including any attachments and exhibits hereto, shall not be exported, released or disclosed to foreign persons whether in the United States or abroad by recipient except in compliance with all U.S. export control regul ations. Recipient shall include this notic e with any reproduced or exc erpted portion of t his document or a ny document derived from, based on, incorpor ating, using or relying on th e information contained in this d ocument. *Electronically approved records are authenticated in t he electronic document management system. Westinghouse Electric Com pany LLC 1000 Westinghouse Drive Cranberry Township, PA 16066, USA © 2014 Westinghouse Electr ic Company LLC All Rights Reserved WCAP-17894-NP.doc-091514 WCAP-17894-NP Revision 0 Component Inspection Details Supporting Aging Management of Reactor Internals at Indian Point Unit 2 Nicholas R. Marino*

Reactor Internals Des ign and Analysis I Charles R. Schmidt*

Major Reactor Com ponents Design and Analysis I Ernest W. Deemer*

Reactor Internals Aging Managem ent September 2014 Approved:

Patricia C. Paesano *, Manager Reactor Inter nals Aging Manage ment Westinghouse Non-Proprietary Class 3 ii WCAP-17894-NP September 2014 Revision 0 RECORD OF REVISIONS Rev. Date Revision Description 0 See EDMS Original Issue

Westinghouse Non-Proprietary Class 3 iii WCAP-17894-NP September 2014 Revision 0 TABLE OF CONTENTS LIST OF TABLES

................................................................................................................

....................... ivLIST OF FIGURES

...............................................................................................................

....................... vLIST OF ACRONYMS

..............................................................................................................

................. viACKNOWLEDGEMENTS ..............................................................................................................

.......... vii1PURPOSE .......................................................................................................................

.............. 1-12BACKGROUND ....................................................................................................................

...... 2-13PROGRAM OWNER

.................................................................................................................

.. 3-14COMPONENT INSPECTION DETAILS OF THE INDIAN POINT UNIT 2 REACTOR INTERNALS

.....................................................................................................................

........... 4-

14.1INTRODUCTION

...........................................................................................................

4-14.2DETECTION OF AGING EFFECTS

.............................................................................

4-44.3INSPECTION RESULTS REPORTING FORMAT

.......................................................

4-74.4COMPONENTS ..............................................................................................................

4-74.4.1Primary ............................................................................................................

4-84.4.2Expansion ......................................................................................................

4-114.4.3Existing ..........................................................................................................

4-135CONCLUSION ....................................................................................................................

......... 5-16REFERENCES ....................................................................................................................

......... 6-1APPENDIX ACOMPONENT INSPECTI ON DETAILS

..................................................................

A-1 Westinghouse Non-Proprietary Class 3 iv WCAP-17894-NP September 2014 Revision 0 LIST OF T ABLES Table 4-1 IP2 Reactor Internals Co mponents Requiring Additi onal Inspections During the License Renewal Term

....................................................................................................

4-2 Westinghouse Non-Proprietary Class 3 v WCAP-17894-NP September 2014 Revision 0 LIST OF FIGURES Figure A-1Typical Westinghou se Internals

......................................................................................

A-1Figure A-2IP2 Internals

.................................................................................................................

... A-2Figure A-3IP2-CID-0010 Control R od Guide Tube Assembly

- Guide Plates (Cards) ..................

A-3Figure A-4IP2-CID-0020 Control R od Guide Tube Assembly

- Lower Flange Welds

..................

A-4Figure A-5IP2-CID-0021 Upper Internals Asse mbly - Upper Core Plate

.......................................

A-5Figure A-6IP2-CID-0022 Lower Internals Assem bly - Lower Support F orging or Castings

.........

A-6Figure A-7IP2-CID-0023 Lower Support Assem bly - Lower Support Col umn Bodies (cast)

........

A-7Figure A-8IP2-CID-002 4 Bottom-mounted Instrum entation S ystem - Bottom

-mounted Instrumentation (BMI) Col umn Bodies ..........................................................................

A-8Figure A-9IP2-CID-0030 Core Barrel Assembly

- Upper Core Barrel Flange Weld

......................

A-9Figure A-10IP2-CID-0031 Core Barrel Assembly

- Core Barrel Outlet Nozzle Welds

..................

A-10Figure A-11IP2-CID-004 0 Core Barrel Assembly

- Upper and Lo wer Core Barrel Cy linder Girth Welds

...................................................................................................................

A-11Figure A-12IP2-CID-004 1 Core Barrel Assembly

- Upper and Lo wer Core Barrel Cy linder Axial Welds

..................................................................................................................

A-12Figure A-13IP2-CID-0050 Core Barrel Assembly

- Lower Core Barrel Flange Weld

...................

A-13Figure A-14IP2-CID-0060 Baffle-former Assem bly - Baffle-edge Bolts

.......................................

A-14Figure A-15IP2-CID-0070 Baffle-former Assem bly - Baffle-former Bolts

....................................

A-15Figure A-16IP2-CID-0071 Core Barrel Assembly

- Barrel-former Bolts

.......................................

A-16Figure A-17IP2-CID-0072 Lower Support Assem bly - Lower Support Col umn Bolts

..................

A-17Figure A-18IP2-CID-0080 Baffle-former As sembly - Assembly ...................................................

A-18Figure A-19IP2-CID-009 0 Alignment and Interfacing Com ponents - Internals Hold-down Spring ........................................................................................................................

.... A-19Figure A-20IP2-CID-0100 Thermal Shield Assem bly - Thermal Shield Flexur es .........................

A-20 Westinghouse Non-Proprietary Class 3 vi WCAP-17894-NP September 2014 Revision 0 LIST OF A CRONYMS AMP Aging Manag ement Program Plan AMR Aging Manag ement Review ASME American Society of Mechanical Engineers B&PV boiler and pre ssure vess el B&W Babcock

& Wilcox BMI bottom

-mounted instrumentation BWR boiling water reactor CE Combustion Engineering CFR Code of Federal Regulations CID Component Inspection Detail EPRI Electric Power Research Institute EVT enhanced visual testing (a visual NDE method that i ncludes EVT

-1) FMECA failure mode, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC irradiation assisted stres s corrosion cracking ID identification INPO Institute of Nuclear Power Operations IP2 Indian Poi nt Nuclear Generating S tation Unit 2 ISI in-service inspection LCP lower core plate LRA License Renewal Applicati on MRP materials reliability program MSC PWROG Materials Subcommittee NDE nondestructiv e examination NEI Nuclear Ener gy Institute NRC Nuclear R egulatory Commission NSSS nuclear steam suppl y system OEM Original Equi pment Manufacturer OER Operating Experience Rep ort PWR pressurized w ater reactor PWROG Pressurized Water Reactor Owners Gro up (formerly WOG) RCS reactor coola nt system RVI reactor vess el internals SCC stress corrosion cracking SER Safety Evaluation Report SRP Standard Review Plan SSC systems, structures, and com ponents UCP upper core pl ate USP upper supp ort plate UT ultrasonic testing (a vol umetric NDE method) VT visual testing (a visual NDE method that includes VT-1 and VT-3) WOG Westinghouse Owners Group Westinghouse Non-Proprietary Class 3 vii WCAP-17894-NP September 2014 Revision 0 ACKNOWLEDGEMENTS The authors would like to thank the m embers of the Entergy Reactor Internals Aging Manage ment Program Team, including Bob Dolansky and our associates at Westinghouse for their efforts in supportin g the developm ent of this repor

t.

Westinghouse Non-Proprietary Class 3 1-1 WCAP-17894-NP September 2014 Revision 0 1 PURPOSE The management of aging degradation effects in reactor internals is required for nuclear plants considering or entering license renewal, as specifi ed in the United States Nucle ar Regulatory Commission (NRC) Standard Review P lan (SRP) [

1]. Indian Point Nuclear Generating Station Unit 2 (IP

2) will be granted an extended license by the NRC through a Safety Evaluation Report (S ER) as documented in NUREG-1930 [

2 and 16]. License Ren ewal Commitment 30 of [

2 and 16], "PWR Vessel Internals Program," committed IP2 to:

1. Participate in industry programs for investiga ting and m anaging aging effects on reactor internals.

2. Evaluate and im plement the results of the industr y programs as ap plicable to the reactor inter nals. Upon completion of t hese program s IP2 submitted an inspection pl an for reactor internals to t he NRC for review and approval.

On September 29, 201 3, IP2 entered what is called the period of "tim ely renewal" while the NRC continues it s consideration of Ente rgy's application t o renew its operating license.

An Aging Ma nagement Program Plan (AMP) [

3] was developed b y IP2 that capt ured the ind ustry guidance for additional reactor internals inspections

, based on the pr ograms sponsored by U.S. utilities through the E lectric Power Research Inst itute (EPRI) materi als reliability program (MRP) and the Pressurized Water Reacto r Owners Gro up (PWROG).

The IP2 A MP was later supplemented by

[5] to provide com pliance with MRP-227-A.

Additional re actor internals inspections of "Prima ry" and "Expansion" com ponents per the IP 2 AMP are supplemental to the "Existi ng" RVI co mponent i nspections as required per the American Society of Mechanical E ngineers (ASME) Boiler and Pressure V essel (B&PV) Code, Sec tion XI progra

m. This document contains com ponent inspectio n details (C IDs) and suppl emental information that p rovides direction on t he location, n umber, configuration, an d inspection requirements of the items that com prise the scope for the additional reactor internals inspections at IP2. As such, this doc ument suppor ts the plant-specific im plementation of additional re actor intern als inspections during the license renew al term.

Westinghouse Non-Proprietary Class 3 2-1 WCAP-17894-NP September 2014 Revision 0 2 BACKGROUND There are fiv e key drivers that affect the continue d successful operation of reactor vessel internals (RVI) at operating P WRs: License R enewal/Life Exte nsion Reliability and Maintenance Inspection In dications Issue-Specifi c Research Component-Specific Aging Managem ent and Inspection Cost Redu ction The Code of Federal Regu lations (CFR) Maintenance Rule, 10 CF R 50.65 requires monitoring of certain systems, structures, and com ponents (S SCs) agains t established goals to provi de reasonable assurance that those SS Cs are capable of fulfilling their intended functions. Li cense renewal requirements of 10 CFR Part 54, require a plant to demonstrate effect ive management of agin g, shifting the emphasis fro m identifying aging m echanisms to managing their eff ects during the license rene wal period. Together these two regulatory requirements consider all active a nd passive co mponents that are r equired for safe operation of t he plant with aging management focu sing on passive, long-lived structures and com ponents.

A plant entering or requesting license renewal is ty pically required t o define and execute an AMP for reactor internals.

For many years the U.S. nu clear power industr y has been actively engaged in eff orts to supp ort the industry goal of responding to the regulat ory requirements on m anaging agi ng degradation in r eactor internals. Various pro grams have been underway to develop gui delines for managing t he effects of aging within PWR r eactor internals. Westinghou se Owners Group (WOG) WCAP-14577 [

6] received NRC Staff review and approval and serve d as the initial ba sis for developing AMPs f or RVIs. The industr y efforts to address the conc ern were cont inued by th e EPRI MRP a nd included consideration of the three currently operating U.S. reactor designs - Westingh ouse, Combustion Engineering (CE), and Babcock &

Wilcox (B&W).

The MRP established a framework and strategy for the aging m anagement of PWR internals co mponents using proven and familiar methods for inspection, monitoring, sur veillance, and communication.

Factoring in t he accumulated industr y research data, t he following elements of an AMP were further refined [

7 and 8]: Screening criteria were developed, consi dering chemical composition, neutr on fluence exposure, temperature h istory, and re presentative stress leve ls, for determining the relative susceptibility of PWR internals co mponents to each of eight postulated aging m echanisms. PWR internals co mponents were categorized, based on the screening criteria, as f ollows: Components for which the effects fro m the pos tulated aging mechanisms are insignificant Components that are moderately susceptible to the agi ng effects Components that are significantly susceptible to t he aging effects Westinghouse Non-Proprietary Class 3 2-2 WCAP-17894-NP September 2014 Revision 0 Functionalit y assessments were perfor med based on representative PWR internals co mponents and component assem blies using irradiat ed and aged material properties, to deter mine the eff ects of the degrad ation mechanisms on component fu nctionality. Aging management strategies were developed combining the functionality assessment results with contributi ng factors to deter mine the appropriate aging management methodology, baseline examination timing, and the need and ti ming of subs equent inspections. Item s considered included com ponent accessibility

, operating experience, existi ng evaluations, and prior e xamination results.

The industr y finalized initial Inspection and Evalua tion (I&E) Guidelines for reactor internals and submitted the docum ent to the NRC with a request for a for mal SER. A supporti ng document addressing inspection requirements was also co mpleted and pr ovided to t he NRC to suppor t the I&E Guidelines review. A third docum ent, which was generate d by the industr y through the PWROG Materials Subcommittee (MSC), provides detailed engineeri ng criteria for evaluating acce ptance of inspection outcomes. This industr y developed g uidance is contained within t he following t hree documents: MRP-227-A [

9] (hereafter referred to as "the I&E Gui delines" or sim ply "MRP-227-A") provides the industr y background, li sting of gener ic reactor internals co mponents requiring inspection, type of nonde structive examination (NDE) required for each co mponent, tim ing for initial inspections, and direction f or evaluating inspection results. MRP-228 [

10] provides gui dance on the qualificati on and demonstration of the N DE techniques and other crit eria pertaining to the actua l performance of the inspections.

WCAP-1709 6-NP [11] provides direction o n engineering evaluations of inspect ion outcomes to determine acceptabilit y for continued service.

The IP2 reactor internals are integral with the reactor coolant s ystem (RCS) of a Westinghous e four-loop nuclear steam suppl y system (NSSS). IP2 reactor in ternals have a downflow baf fle-barrel regi on flow design, and a top hat design upper support plate (USP)

. An illustr ation of IP2 i nternals is provided i n Figure A-2.

As described in NUREG-1 930 [2 and 16], the applican t described the RVIs, whi ch consist of t wo basic assemblies: a n upper internals ass embly that is re moved during each refueling operation to obtain acce ss to the reactor core and a lower internals assembly that can be re moved following a co mplete core off-load. The reactor core is positioned and supported by the upper internals and lower in ternals assem blies. The individual fuel assemblies are positioned by fuel alignment pins in t he upper core plate (UCP) and the lower core plate (LCP). These pins control the orienta tion of the c ore with respect to the upper and lower internals assem blies. The lower in ternals are aligned with the uppe r internals b y the UCP alignment pins and secondarily by the head/vessel alignment pins. T he lower internals are aligned to the vessel by the lower radial support/clevis assemblies an d by the head/vessel alignment pins. T hus, the core is aligned with the vessel by a number of interfacing com ponents.

Westinghouse Non-Proprietary Class 3 2-3 WCAP-17894-NP September 2014 Revision 0 The lower internals assem bly is supporte d in the vessel by clamping to a ledge below the vessel-head mating surface and is closely gui ded at the bottom by radial support

/clevis assemblies. The bo ttom of the upper internal s assembly is closely guided by the core barrel alignment pins of t he lower internals assembly. Upper Int ernals Assembly The major sub-assem blies that constitut e the upper intern als assembl y are the: (1) UCP; (2) upper support column assemblies; (3) control rod guide tube asse mblies; and (4)

USP. During reactor operation, t he upper inter nals assem bly is preloaded against the fuel assem bly springs and the internals hold down spring by the reactor vessel h ead pressing down on the outside edge of the USP.

The upper su pport col umns and the cont rol rod guide tubes are attached to the USP. The UCP, in tur n, is attached to the upper supp ort colum ns. The USP design at IP2 is d esignated as a top hat design. The UCP is perforated to per mit coolant to pass from the core below into the upper plenum defined by the USP and UCP. The UCP positions and laterally supports the core b y fuel align ment pins extending below the plate. The UCP contac ts and preloads the fuel asse mbly springs and thus m aintains contact of the fuel assemblies wi th the LCP during reactor operation.

The upper support col umns vertically position the U CP and are designed to take the uplifti ng hydraulic flow loads and fuel spring loads on the U CP. The control rod guide tubes are bolted to the USP and pinned at the UCP. Guide tube cards are located with in the control rod guide tube assembly to guide the absorber rods. The control rod guid e tubes are also slotted in their lo wer sections to allow cool ant exiting the core to flow into the upper plenum

. The UCP alignment pins locate the UCP laterally with respect to the lower internals ass embly. The pins must laterally support t he UCP so that the plate is free to expand r adially and move axiall y during differential therm al expansion between the upper inter nals and the core barrel. The UCP alignment pins are the interfacing com ponents between the UCP and the core barrel.

Lower Int ernals Assembly The fuel assem blies are supported i nside the lower intern als assembl y on top of the LCP. The functions of the LCP are to position and support th e core and provide a metered control of reactor coolant flow into each fuel assem bly. The LCP is elevated above the lower support casting by support col umns and bolted to a ring su pport attached to the inside d iameter of the core barrel. The support c olumns transmit vertical fuel assembly loads from the LCP to the much thicker lower support casting, which provides support for the core.

The primary function of the core barrel is to supp ort the core. A large num ber of co mponents are attached to the core barrel, including the baffle/former as sembly, the core barrel outlet no zzles, the ther mal shield, the alignm ent pins that engage the UCP, the lowe r support casting, and the LCP.

The lower radial support/clevi s assemblies restrain large transverse motions of the co re barrel, but at the same time allow unrestricted radial and axial thermal expansion.

Westinghouse Non-Proprietary Class 3 2-4 WCAP-17894-NP September 2014 Revision 0 The baffle and form er assembly consists of ver tical plates called baffles and hori zontal suppor t plates called for mers. The baffle plates are bolted to the fo rmers by the baffle/for mer bolts, and the formers ar e attached to the core barrel i nside diamete r by the barrel/former bolts. Baffle plate s are secured to each other at select ed corners by edge bolts. I n addition, at IP2, corner br ackets are installed behind and bolted to the baffle plates. The baffle/for mer assembly forms the interface between the c ore and the core barrel.

The baffles provide a barrier between the core and the former region so that a hi gh concentration of fl ow in the core region can be maintained. A secondary benefit is to r educe the neutron flux on t he vessel.

Additional ne utron shielding of the reactor vessel is provided in the active core re gion by the ther mal shield attache d to the outside of the core barrel.

In the up per internals assem bly, the USP, the up per support columns, and the U CP are considered core support struct ures. In the l ower internals assem bly, the LCP, the l ower support casting, the lower support columns, the core barrel including t he core barrel fl ange, the radial support/clev is assemblies, the baffle plates, and th e former plates are classified as core su pport structur es. All RVIs ar e removable for their inspection, an d for inspection of the vessel inte rnal surface.

Based on the co mpleted evaluations, the RVI components are cat egorized within MRP-227-A as "Primary" components, "Expansion" com ponents, "E xisting Progr ams" components, or "No Additional Measures" co mponents. Descriptions of the final categories are as follows: Primary Those PWR internals that are highly susceptible to the effect s of at least one of the eight aging mechanisms were plac ed in the Prim ary group. The a ging management requirements that ar e needed to ensure functional ity of Prim ary components are described in the I&E guidelines. The Primary group also include s components that have shown a degree of tolerance to a specific aging degradation effect, but for which no hi ghly susceptible component exists or for which no hi ghly susceptible com ponent is accessible.

Expansion Those PWR internals that are highly or moderately susceptible to the effects of at least one of the eight aging m echanisms, but for which f unctionality assessment has shown a degree of tolerance to those effects, were placed in the E xpansion group. The schedule for im plementation of agin g management requirem ents for Expansio n components depends on the findings from Primary component exam inations at indivi dual plants. Existing Program s Those PWR internals that are su sceptible to the effect s of at least one of the eight aging mechanisms and for which generic and plant-sp ecific existing AMP ele ments are capable of managing those effects, were placed in the Existing Programs group.

Westinghouse Non-Proprietary Class 3 2-5 WCAP-17894-NP September 2014 Revision 0 No Additiona l Measures Pr ograms Those PWR internals for which the effects of a ll eight aging mechanisms are below the scre ening criteria were placed in the No Additiona l Measures group. Additional co mponents were placed in the No Additi onal Measures group as a result of a fa ilure mode, effects, and criticality analysis (FMECA) an d the functionality assessment. No furt her action is required by these guidelines for No Additiona l Measures co mponents agi ng management.

The categoriz ation and ana lysis used in the devel opment of MRP-22 7-A were not intended to s upersede any ASME B&PV Code Section XI requirem ents. Components that are classified as core su pport structures, as defined in AS ME B&PV Code Secti on XI IWB 2500

, Category B-N-3, have requirem ents that remain in effect and may only be altered as allowed b y 10 CFR 50.55a.

A listing of t he IP2 RVI com ponents and subcom ponents already reviewed by the NRC in the SER granting life extension that are subject to aging m anagement requirements were included as Tables 5-2, 5-3, and 5-4 of the IP2 License Renewal Application

[5]. The link between prim ary and expansio n MRP-227-A com ponents is defined in Table 5-5 of [5]. The IP2-specific MRP-227-A reactor internals components that require additional inspe ctions and the components with existing ASME Section XI inspections that are cr edited for m anaging aging in R VI are summarized by MRP-227-A classification categories an d shown in Table 4-1 of th is report.

Westinghouse Non-Proprietary Class 3 3-1 WCAP-17894-NP September 2014 Revision 0 3 PROGRAM OW NER The PWR Ve ssel Internals Program is established in accordance wit h Entergy' s "NEI 03-08 Materials Initiative Process"

[12]. The successful i mplementation and com prehensive long-ter m management of the IP2 RVI AMP will require the integration of Ente rgy organizati ons (both corporate and at IP2) and interaction with m ultiple industr y organizations includ ing, but not li mited to, the ASME, MRP, NRC, and PWROG. The responsibilities of the individual corporate and IP2 groups are delineated in appropriate site procedures. Entergy will maintain cogni zance of indu stry activities related to PWR internals inspection and aging m anagement, and will address/im plement industry guidance stemming from those activities, as appropriate u nder Nuclear Energy Institute (NEI) NE I 03-08 [13] practices.

The appropriate p ersonnel and their responsibilities are su mmarized in site procedures.

Westinghouse Non-Proprietary Class 3 4-1 WCAP-17894-NP September 2014 Revision 0 4 COMPONENT INSPECTI ON DETAILS OF TH E INDIAN POI NT UNIT 2 REACTOR INTERNALS

4.1 INTRODUCTION

The U.S. nucl ear industry, t hrough the com bined effo rts of utilities, vendors, and independent c onsultants, defined a generic guideline to assist utilities in de veloping reactor internals plant-specific agin g management program s based on inspecti on and evalua tion. The primary objectiv e for the indu stry effort and each individual plant program is to ensure the lo ng-term integrity and safe operation of the reactor internals co mponents and overall reliability of the plant by proactively m anaging aging.

The IP2 reactor internals AMP utilizes a co mbination of prevention, m itigation, and condition monitoring to manage aging in susceptible reactor internals components. Where applicable, credit is take n for existing progra ms such as water che mistry [7] and inspections prescribed by the ASME Section XI In-Service Inspe ction (ISI) Program

[4], combined with additional reactor internals inspections or evaluations as reco mmended by MRP-227-A.

The purpose of this IP2 CI D WCAP is to supp ort IP2-specific im plementation of the generic industr y additional reactor internals inspections with a focu s on components that are not already included in existing IP2 i nspection program

s. To ensure a clear understanding of the require ments, a brief description of the require d inspections is included in this section followed by descriptions of the individua l components that com prise the IP2 additi onal reactor internals MRP-227-A inspe ctions. CIDs illustrating key considerations of the re quired MRP-227-A inspecti on for the component is included for al l applicable IP2 MRP-22 7-A Primary and Expansio n components A listing of t he components applicable to IP2 ident ifying the MRP-227-A category

, required i nspection, and corresponding CID is provided in Table 4-1. The Control Rod Guide Tube Assembly: Guide Tube Support Pins (Split Pins) ar e not include d in Table 4-1 because they are managed by Original Equipment Manufacturer (OEM) reco mmendations in accordance w ith Applicant/Licensee Action Ite m 3 of the NRC Safety Evaluation for MRP-227 as described in Section 3.6 of NL-12-037 [

5].

Westinghouse Non-Proprietary Class 3 4-2 WCAP-17894-NP September 2014 Revision 0 Table 4-1 IP2 Reactor Internals Co mponents Re quiring Addi tional Inspections During the Licens e Renewal T erm Component MRP-227-A CID WCAP-17096-NP ID(2) Category Inspection Type Control Rod Guide Tube Assembly - Guide Plates (Cards)

Primary VT-3 IP2-CID-001 0 W-ID: 1 Control Rod Guide Tube Assembly - Lower Flange Welds Primary EVT-1 IP2-CID-002 0 W-ID: 2 Upper Internals Assem bly - Upper Core Pl ate Expansion EVT-1 IP2-CID-002 1 W-ID: 2.1 Lower Internals Assem bly - Lower Sup port Forging or Castings (3) Expansion EVT-1 IP2-CID-002 2 W-ID: 2.2 Lower Suppo rt Assembly

- Lower Supp ort Colum n Bodies (Cast)

Expansion EVT-1 IP2-CID-002 3 W-ID: 2.3 Bottom-mounted Instrum entation S ystem - Bottom-mounted Instrumentati on (BMI) Colum n Bodies Expansion VT-3 IP2-CID-002 4 W-ID: 2.4 Core Barrel Assembly - Upper Core Barrel Fla nge Weld Primary EVT-1 IP2-CID-003 0 W-ID: 3 Core Barrel Assembly - Core Barrel Ou tlet Nozzl e Welds Expansion EVT-1 IP2-CID-003 1 W-ID: 3.1 Core Barrel Assembly - Upper and Lower Core Barr el Cylinder Girth Welds Primary EVT-1 IP2-CID-004 0 W-ID: 4 Core Barrel Assembly - Upper and Lower Core Barr el Cylinder Axial Welds Expansion EVT-1 IP2-CID-004 1 W-ID: 4.1 Core Barrel Assembly - Lower Core Barrel Flange Weld (4) Primary EVT-1 IP2-CID-005 0 W-ID: 5 Baffle-for mer Assembly

- Baffle-edge Bolts Primary VT-3 IP2-CID-0060 W-ID: 6 Baffle-for mer Assembly

- Baffle-for mer Bolts Primary UT IP2-CID-0070 W-ID: 7 Core Barrel Assembly - Barrel-former Bolts Expansion UT IP2-CID-0071 W-ID: 7.1 Lower Suppo rt Assembly

- Lower Supp ort Colum n Bolts Expansion UT IP2-CID-007 2 W-ID: 7.2 Baffle-for mer Assembly

- Baffle-for mer Assembly Prim ary VT-3 IP2-CID-008 0 W-ID: 8 Alignment and Interfacing Components - Internals Hold-down Spring Primary Special(1) IP2-CID-009 0 W-ID: 9 Thermal Shield Assem bly - Thermal Shield Flexures Primary VT-3 IP2-CID-010 0 W-ID: 10 Core Barrel Assembly: Core Barrel Flange Existing VT-3 Not applicable Not applicable Upper Internals Assem bly: Upper Suppo rt Ring or S kirt(5) Existing VT-3 Not applicable Not applicable Westinghouse Non-Proprietary Class 3 4-3 WCAP-17894-NP September 2014 Revision 0 Table 4-1 IP2 Reactor In ternals Components Req uiring Addi tional Inspections During the Licens e Renewal T erm (cont.) Component MRP-227-A CID WCAP-17096-NP ID(2) Category Inspection Type Lower Core Plate Existing VT-3 Not applicable Not applicable Alignment and Interfacing Equipment: Clevis Insert Bolts Existing VT-3 Not applicable Not applicable Alignment and Interfacing Equipment: Upper Core Plate Align ment Pins Existing VT-3 Not applicable Not applicable Bottom Mounted Instrum entation S ystem: Flux Thimble T ubes Existing ET Not applicable Not applicable Notes: 1. Managed by plant-specific measurement program

s. 2. Confirmation of identification (ID) number pending NRC approval of WCAP-17096-NP. 3. This component is a casting at IP2. 4. At IP2 this weld is the lower core barrel to lower su pport casting weld. IP2 does not have a lower core barrel flange. 5. IP2 has a tophat design. Therefore, there is no support ring or skirt; however, the vertical sections of the tophat will be inspected.

Westinghouse Non-Proprietary Class 3 4-4 WCAP-17894-NP September 2014 Revision 0 4.2 DETECTION OF AGING EFFE CTS Inspection can be used to detect phy sical effects of degradation incl uding cracking, fracture, wear, and distortion

. The inspection technique is chosen based on th e nature and extent of the expected da mage. The recommendations supporting aging m anagement for the reactor internals, as defi ned by the industry and contained in t his report, are built around three basic inspection techniques: (1) vi sual testing (VT), (2) ultrasonic testing (UT), and (3) phy sical measurement. Three differe nt visual techniques are included:

VT-3, VT-1, and EVT-1

. Those additional reactor internals insp ections that are taken from the MRP-227-A recommendations will be applied through use of th e MRP-228 I nspection Standard [

10]. Detection of indications th at are required b y the ASME Section XI ISI Program is well-established and fi eld-proven through the application of the Section XI ISI Program

. The assumptions and process used to select the appropr iate inspection technique are described in the following su bsections. Inspection standards developed by the industry for the appl ication of the se techniques for additional reactor interna ls inspections are docum ented in MRP-228.

VT-1 Visual Examinations In MRP-227-A note that VT-1 has onl y been selected to detect distortion as evid enced by small gaps between the upper-to-lower mating surfaces of CE-wel ded core shrouds assem bled in two vert ical sections. Therefore, no add itional VT-1 i nspections ov er and above those required b y ASME Section XI ISI have been specified in MRP-227-A at IP2.

EVT-1 Enhanced Visual Exa mination for the Detectio n of Surface Breaking Flaws In the additional reactor internals inspect ions deta iled in the MRP-227-A, the EVT-1 enhanced visual examination has been identified for inspection of components where surface-br eaking flaws are a potential concern

. Any visual inspection for cracking requires a reasonable expectation that the flaw length and cr ack mouth opening displac ement meet the resolution requirements of the observ ation technique

. The EVT-1 spec ification augments the VT

-1 requirements to pro vide more rigorou s inspection standards for stress corrosion cracking (SCC) and has been demonstrated for si milar inspections in boiling water reactor (BW R) internals

. Enhanced visual exa mination (i.e., EVT-1) is also con ducted in accordance with the requirem ents descri bed for vi sual exam ination (i.e., VT-1) with additional requirements (such as camera scanning speed) currently being developed by the industr

y. Any recommendation for EVT-1 inspection will require additio nal analy sis to establi sh flaw-tolerance criteria

. The industr y is currently developing a c onsensus approach for acceptance criteria methodologies to support plant-specific additional reactor internals examinations

. Entergy has been an active participant in these initiatives and will follow the industry directive

. These acceptance criteri a methodologie s may be determined either generical ly or on a pla nt-specifi c basis because both loads and component di mensions may vary from plant to plant within a t ypical PWR design.

VT-3 Examination for General Condition Monit oring In the additional reactor internals inspect ions detaile d in the MRP-227-A, the VT-3 visual examination has been identified for inspection of component s where general condition m onitoring is requir ed. The VT-3 exam ination is inten ded to ident ify individual components with significant levels of existing Westinghouse Non-Proprietary Class 3 4-5 WCAP-17894-NP September 2014 Revision 0 degradation

. As the VT-3 exam ination is not inten ded to detect th e early stages of com ponent cracking or other incipien t degradation effects, it sho uld not be used when failur e of an indi vidual component could threaten either plant safety or operational stability

. The VT-3 exam ination m ay be appropriate for inspecting hi ghly redundant components (such as baff le-edge bolts), where a sin gle failure do es not compromise the function or integrity of the critical as sembly. The acceptan ce criteri a for visual exam inations conducted under categories B-N-2 (welded core support structures and interior attachments to reactor vessel s) and B-N-3 (rem ovable cor e support structures) are defined in IW B-3520 [14]. These criteri a are designed to provide ge neral guidelines

. The unacc eptable conditions for a VT-3 examination are: Structural distortion or displacement of parts to the extent that co mponent function m ay be impaired Loose, missing, cracked, or fractured parts, bolting, or fasteners Foreign m aterials or accu mulation of corrosion prod ucts that could interfere with control rod motion or could result in blockage of co olant flow thr ough fuel Corrosion or erosion that reduces the nominal section thickness by more than five percent Wear of mating surfaces that may lead to loss of func tion Structural degradation of i nterior attachments such that the original cross-se ctional area is reduced more than five percent The VT-3 exam ination is intended for use in situations where the degradation is readily observable. It is meant to provide an indication of condition, a nd quantitative accept ance criteri a are not generally required. In a ny particular recommendation for VT-3 visual examination, it should be possible to identify the specific conditi ons of concern. For i nstance, th e unacceptable conditi ons for wear indicate wear that might lead to loss of functi on. Guideline s for wear in a critical-alignment co mponent may be very different from the guidelines for wear in a large structural co mponent. Ultrasonic T esting Volumetric exam inations in the form of UT techni ques can be used to identif y and determ ine the length and depth of a crack in a com ponent. Although access to the surface of the com ponent is requi red to appl y the ultrasonic signals, the flaw may exist in the bulk of the material.

In this pr oposed strategy

, UT inspections h ave been reco mmended exclusively for d etection of flaws in bolts. F or the bolt inspections, any bolt with a detected flaw should be a ssumed to have failed. The size of the flaw in the bolt is not critical becau se crack growth rates are generally high, and it is assumed that the observed flaw will result in failure prior to the next i nspection opportunit

y. It has generally been observed t hrough examination performance demonstrations that UT can reliabl y (90 percent or gr eater reli ability) detect flaws that reduce the cr oss-sectional area of a bolt by 35 percent.

Westinghouse Non-Proprietary Class 3 4-6 WCAP-17894-NP September 2014 Revision 0 Failure of a single bolt does not com promise the functi on of the entire ass embly. Bolting s ystems in the reactor internals are highly redundant.

For any system of bolts, i t is possible to de monstrate multiple minimum acceptable bolting patterns.

The evaluation program must demonstrate that the rem aining bolts meet the requ irements for a m inimum bolting pattern for continued operation. Th e evaluation procedures must also de monstrate that the pattern of re maining bolts contains sufficient margin such that continuation of the bolt failure rate will not result in f ailure of the sy stem to meet the requirements for minimum acceptable bo lting pattern before the next inspection.

Establish ment of the m inimum acceptable bolting pa ttern for any system of bolts requires analy sis to demonstrate t hat the sy stem will maintain reliabilit y and integrit y in continuing to perform the intended function of the component. This analy sis is highl y plant-specific. Therefore, any recommendation for UT inspection of bolts assumes that the plant owne r will work with the designe r to establish mini mum acceptable bo lting patterns prior to t he inspection to support continued operation.

For Westinghous e-designed plants, m inimum acceptable boltin g patterns for baffle-former and barrel-form er bolts are available through the PWROG.

Entergy has been a full participant in the developm ent of the PWROG doc uments and has full acce ss and use.

Physical Measurem ent Examination Continued f unctionalit y can be confirm ed by physical measurements to evaluate the im pact caused by various degradation m echanisms such as wear or lo ss of functionality as a result of loss of prel oad or material deformation. For IP2, direct physical measurements are required only for the internals hold-down spring. An alternate option is to replace the existing internals hold-down spring w hich could eliminate the need for measurements.

Westinghouse Non-Proprietary Class 3 4-7 WCAP-17894-NP September 2014 Revision 0 4.3 INSPECTION RESULTS REPORTING FORMAT Entergy IP2, the MRP, and the PWROG approaches to aging m anagement are based on the Gen eric Aging Lessons Learned (GALL)Report approach as detailed in NUREG-1801 [

15]. This ap proach includes determining which reactor internals passive components ar e most susceptible to the aging mechanisms of concern and then determining t he proper inspection or m itigating program that provides reasonable assurance that the com ponent will continue to perform its intended fu nction thr ough the period of extended operation. The GALL-based approach wa s used at IP2 f or the initial basis of the License Renewal Application (LRA) that resulted in the NRC SER in NUREG-1930 [

2 and 16]. A key element of the MRP-227-A gui deline is the repor ting of age-r elated degradation of reactor vessel components throug h Operating Ex perience Reports (OER s). Entergy, thro ugh its participation i n PWROG and EPRI-MRP activities, will continue to benefit fro m the reporting of i nspection inf ormation and will share its own operating experience with the industr y through those groups or I nstitute of Nuclear Power Operations (INPO), as app ropriate.

The component nam ing nomenclature implemented b y the industry in PWROG WCAP-1709 6 [11] forms the basis for data recording of inspection results.

WCAP-17096-N P is currently being updated to account for changes from MRP-227 Rev. 0 to M RP-227-A, which is wh y the W-ID numbers are listed as "pending". L ocation indic ators for each component item are based on the IP2-specific design.

4.4 COMPONENTS

The Westinghouse reactor internals are part of the R CS and located inside the r eactor pressur e vessel. The reactor internals are long-lived passive structural co mponents. The i ntended funct ions are to support core cooling, enab le control rod insertion, an d maintain the integrit y of the fuel. Internals co mponents are classified as either core support stru ctures or internals structures.

All Westinghouse internals consist of two basic assemb lies: an uppe r internals assem bly that is removed during each refueling operation to obtain acces s to the reactor core

, and a lower internals ass embly that can be rem oved following a complete core off-load.

The lower internals assem bly is supported in the vessel by clamping to a ledge below the vessel head mating surfac e and is close ly guided at the bottom by radial support clevis assemblies. The upper internal s assembly is clamped at this same ledge by the reactor vessel head. The bottom of the upper intern als assembly is closely guided by the core barrel alignment pins of the lower internals assem bly. All of the assem blies and i ndividual com ponents making up the Westinghous e reactor inter nals were considered in the process of developing t he MRP-227-A requirem ents. The com plete categorization process is su mmarized in MRP-227-A and support ing basis documents. Brief descriptions of the components required to support agi ng management of reactor internals at IP2 are included in the following su bsections by MRP-227-A Prim ary, Expansion, and E xisting catego ries.

Westinghouse Non-Proprietary Class 3 4-8 WCAP-17894-NP September 2014 Revision 0 4.4.1 Primary MRP-227-A Pri mary components are those that were deter mined to be highl y susceptible to the effects of a least one of the critical degradation m echanisms affecting reactor internals or a co mponent for which no highly susceptible com ponent exists or was directly accessible. Table 4-1 contai ns a complete listing of all of the Pri mary components applicable to IP2.

Control Rod Guide Tube Assembly - Guide Plates (Cards)

The control rod guide tube assembly provides an alignment and insertion pa th for the control rods through the upper inte rnals. Guide cards provide alignment a nd an insertion path for cont rol rod assemblies, and support t he control r ods when withdrawn. The g uidance holes in th e guide cards are distorted b y wear (loss of material). The largest am ounts of wear to da te have t ypically been observ ed in the low est guide card levels.

Guidance hole wear can cause lack of alignm ent. Lack of alignm ent may impact control r od drop times. In the worst-case s cenario, control rods may jam and prevent full insertion which would caus e a technical specification co mpliance concern. WCAP-1709 6-NP Component Category ID: W-ID: 1 (pendi ng) Control Rod Guide Tube Assembly - Lower Flange Welds The control rod guide tube assembly provides alignm ent and an insertion path for control rods through the upper internal

s. The lower flange welds retain the structural alignment and stability of the com ponent. Flow in the upper head applies a bending m oment to the control rod guide tube assembly. Maximum bending stresses tend to occur near the top of t he continuous guidance section (upper flange weld location as identified in Figure A-4). Stresses may lead to form ation of SCC or fatigue cracks. Weld c racking may lead to loss of stiffness in the guide tube assemb ly and loss of support capability yielding a loss of structural stability

. Excessive deflection could im pede control assem bly insertion, resulting in a failure to perform its intended functi on. WCAP-1709 6-NP Component Category ID: W-ID: 2 (pendi ng) Core Barrel Assembly - Upper Core Barrel Flange Weld The upper co re barrel flan ge weld is an integral co mponent of t he primary core support structure. The primary concern with regard to ag ing is through-wall cracking in the weld as a r esult of stress corrosion.

Actively growing through-wall flaws would req uire attention and could cause a potential loss of core support result ing in safet y concerns. However, the core barrel is co nsidered a highl y flaw-tol erant structure, and relatively large inactive flaws are likely to be m anageable even if found

. WCAP-1709 6-NP Component Category ID: W-ID: 3 (pendi ng)

Westinghouse Non-Proprietary Class 3 4-9 WCAP-17894-NP September 2014 Revision 0 Core Barrel Assembly - Upper and Lo wer Core Bar rel Cylinder Girth Welds The core barrel cy linder girth welds are an integral com ponent of the prim ary core support str ucture. The primary concern with regard to ag ing is through-wall cracking in the weld as a r esult of stress corrosion.

Actively growing through-wall flaws would req uire attention and could cause a potential loss of core support result ing in safet y concerns. However, the core barrel is co nsidered a highl y flaw-tol erant structure, and relatively large inactive flaws are likely to be m anageable even if found

. WCAP-1709 6-NP Component Category ID: W-ID: 4 (pendi ng) Core Barrel Assembly - Lower Flange Weld The lower flange weld is an integral co mponent of th e primary core support str ucture. This weld joins th e core barrel cy linder to t he lower core su pport for ging/casting; no actual flange exists at this location. The primary concern with regard to ag ing is through-wall cracking in the weld as a r esult of stress corrosion.

Actively growing through-wall flaws would req uire attention and could cause a potential loss of core support result ing in safet y concerns. However, the co re barrel is co nsidered a highl y flaw tole rant structure and relatively large inactive flaws ar e likely to be manageable even if found

. WCAP-1709 6-NP Component Category ID: W-ID: 5 (pendi ng) Baffle-for mer Assembly

- Baffle-edge Bolts The baffle-edge bolts pro vide baffle-plate to baffle-plate attach ment along t he seams between the plates and functio n to prevent gap s forming between plates.

Structural studies have dem onstrated that baffle-edge bolts are not required to maintain the structural integrity of the baffle; therefore, baffle-edge bolts are not considered to be a safety significant component. Analysis has shown that differential therm al expansion an d swelling can cause plastic deform ation of edge bolts. The edge bo lts are in high radiation locations, and a significant potential for failure exists due to irradiation assisted st ress corrosion cracking (IASCC). Op erationally

, gaps between plates can res ult in baffle-jetting dam age to fuel asse mblies. In plants with d ownward coolant flow in t he region bet ween the baff le and the former, failure o f baffle-edge bolts is consi dered to directly contribute to baffle jetting.

Evidence of b affle-edge bolt failure would t ypically be observed thr ough broken or missing locking devices, protruding bolt he ads, or m issing bolts or bol t heads. Therefore, the primary concerns are baffle jetting, lo ose parts generation, and interference w ith fuel from broken or m issing baffle-edge bolts.

WCAP-1709 6-NP Component Category ID: W-ID: 6 (pendi ng) Baffle-for mer Assembly

- Baffle-for mer Bolts The baffle-for mer bolts att ach the baffle plates to th e baffle for mers. Documented observations of IASCC cracking of these co mponents exists in m ultiple designs in the PWR fleet wor ldwide. These highl y irradiated bol ts perform a critical safety and operation al function in the plant. Los s of a single b olt or isolated multiple failures of the baffle-former bolts ar e considered to be manageab le, but a catastrophic or Westinghouse Non-Proprietary Class 3 4-10 WCAP-17894-NP September 2014 Revision 0 clustered loss of m ultiple bolts at adjacent loca tions could cause a lack of structural stability and potentiall y raise safety and operational concerns.

WCAP-1709 6-NP Component Category ID: W-ID: 7 (pendi ng) Baffle-for mer Assembly - Assembly The baffle-for mer assembly is made up of vertical plates na med "baffles" and horizontal support plates named "formers." The baff le plates are b olted to the formers by the baffle-for mer bolts, and the formers are attached to the core barrel inside surface by the barrel-for mer bolts. Som e of the baffle plates are also bolted to eac h other at selected corners by edge bolts or brackets. The baffle-former a ssembly forms the interface bet ween the core and the core barrel. The baffle-for mer assembl y provides support,

guidance, and protection for t he reactor core, a passageway for the distribution of the reactor coolant flow to the reactor core, a nd radiation (ga mma and neutron) shielding for the reactor vessel. Void swelling and IA SCC of the integrated assem bly are the key concerns identified which could m anifest multiple degradation e ffects, such as interference with fuel assem blies, obstruction of coolant flow, loose parts g eneration, distortion and misalignment of the core, local temperature p eaks, degrada tion of control rod inser tion paths, and baffle jetting. To date, no re levant observations of these effects in the baffle-for mer assembly as a result of void sw elling or IASCC have been docum ented. WCAP-1709 6-NP Component Category ID: W-ID: 8 (pendi ng) Alignment and Interfacing Components - Internals Hold-Down Sp ring The function of the hold-down spring is to retain the reactor internals in proper alignment to the core. The primary aging degradation concern is stress rel axation as a result of long-term service. The direct result of long term stress relaxation is a loss of hol d-down forc es leading to v ibration and wear in the lower internals.

WCAP-1709 6-NP Component Category ID: W-ID: 9 (pendi ng) Thermal Shield Assem bly - Thermal Shield Flexures Thermal shield flexures are part of the th ermal shield support system. This sy stem provides prim ary attachment of the thermal s hield to the core barre l at the lower connection points of the asse mbly. The thermal shield flexures pro vide the lower structural support for the thermal shield and are required to hol d the bottom of the thermal shield con centric to the core barrel. Due to the differential deflections between the core barre l and therm al shield caused by thermal cycling resulti ng from plant operation, t he thermal shield flexures are potentially susceptible to fatigue. I ndications of wear in the com ponents (bolts, pins, and fasteners) composing the thermal sh ield-to-core ba rrel atta chment system, failure in the w elds at the base of the flexure, or failure in the weld that att aches the flexure to the therm al shield are the identified indicators of age-related c oncerns in the ass embly.

Westinghouse Non-Proprietary Class 3 4-11 WCAP-17894-NP September 2014 Revision 0 WCAP-1709 6-NP Component Category ID: W-ID: 10 (pend ing) 4.4.2 Expansion MRP-227-A Expansion co mponents are those that we re determ ined to be highly or moderately susceptible to the effects of a least one of the critical degradation mechanisms aff ecting reactor internals but which det ailed evaluations showed a degree of tol erance to those effect

s. Examination of expansion components is dependent o n the plant-sp ecific pr imary component inspection ob servations and the expansion ins pection requir ements in Table 4-6 of MRP-227-A [

9] and Table 5-3 of NL-12-0 37 [5]. Table 4-1 lists all of the Ex pansion components applic able to IP2.

Upper Internals Assem bly - Upper Core Plate The UCP is perforated to per mit coolant to pass from the core into t he upper plenum defined by the USP and the UCP. The coolant t hen exits through t he outlet nozzles in the core barrel.

The UCP positions and laterally supports the core b y fuel pins extending below the plate. The UCP contacts and preloads the fuel assembly springs and, therefore, maintai ns contact of the fuel asse mblies with the LCP during reactor operation. Ag ing concerns for this com ponent are cracking from wear and fatigue, which may compromise the ability of the UCP to pr operly align the fuel. WCAP-1709 6-NP Component Category ID: W-ID: 2.1 (pen ding) Lower Internals Assem bly - Lower Sup port Forging or Casting The lower support for ging or casting provides support for the core.

Indian Poi nt has a lower support casting. Cracking resulting in displacement of lowe r support forgings or castings would cause concerns for the operat ional and structural stability of the lower reactor internals support a ssembly. The lower support casting is welded to and supported by the core barrel, which transmits vertical loads to the vessel through the core barrel flange. Aging concerns for this component are cracking and thermal embrittlement.

WCAP-1709 6-NP Component Category ID: W-ID: 2.2 (pen ding) Lower Suppo rt Assembly

- Lower Supp ort Colum n Bodies (Cast)

The lower support col umns provide the structural lin k between the LCP and the l ower support structure.

The supports are required to keep the L CP from deforming during plant operation.

The upper sections of the lower support colum n bodies m ay experience neutr on fluences above the industry threshol ds for IASCC. The IP2 lower support col umn bodies are com posed of forged and cast materi als. The cast co mponents are considered separately because a concern exists that they m ay be more sensitive to thermal and irradiation effects.

Although stresses i n columns are primarily compressive, bending stresses or the design of the attach ment may produce localized regions of tensile stress and, therefore, have increased susceptibility to cracking.

Maintaining core stability and co re plate flatness is t he intended purpose of t his component, and loss of o ne or more may compromise this function.

Westinghouse Non-Proprietary Class 3 4-12 WCAP-17894-NP September 2014 Revision 0 WCAP-17096-NP Com ponent Category ID: W-ID: 2.3 (cast) (pending)

Bottom-mounted Instrum entation S ystem - Bottom-mounted Instrumentation (BMI) Column Bodies The BMI colum n bodies define the path for flux thimbles to be ins erted into the fuel assemblies. The plant must maintain a required num ber of functio ning flux thimbles for core mapping. Fl ux thimbles are normally withdrawn prior to refueling a nd reinserte d at the end of t he refueling activities. A key consideration is that the pri mary pressure boundar y must rem ain intact.

The BMI colum n bodies may be subje ct to fatigue due to either th ermal fatigue or flow-induc ed vibrations. Inabilit y to insert flux thim bles would be noted duri ng refueling activities. Once fl ux thimbles are inserted, t he consequences of failure of the co mponent to perfor m its intended function during the ensuing operating c ycle are ty pically considered to be mini mal. WCAP-1709 6-NP Component Category ID: W-ID: 2.4 (pen ding) Core Barrel Assembly - Core Barrel Ou tlet Nozzle W elds The core barrel outlet nozzle welds join the core ba rrel outlet n ozzles to the uppe r core barrel cy linder. The primary concern with regard to aging is thro ugh-wall cra cking in the weld as a result of stress corrosion. Activel y growing through-wall flaws woul d require attention an d could potentiall y cause jetting through the core barrel. Howeve r, the core barrel is considered a highly flaw-tolerant structure, and relatively large inactive flaws are likel y to be manageable even if found

. WCAP-1709 6-NP Component Category ID: W-ID: 3.1 (pen ding) Core Barrel Assembly - Upper and Low er Core Barr el Cylinder Axial Welds The core barrel axial welds are in place to m aintain th e cylindrical shape of the core barrel, as the core barrel cy linders originate as flat plates that are then ro lled into c ylinders. The primary concern with regard to aging is through-wall cracking in the weld as a re sult of stress corrosion. Acti vely growing through-wall flaws would req uire attention and could potentially cause jetting thro ugh the core barrel. However, the core barre l is considered a highly fla w-tolerant st ructure, and relatively large inactive flaws are likely to be manageable even if found

. WCAP-1709 6-NP Component Category ID: W-ID: 4.1 (pen ding) Core Barrel Assembly - Barrel-former Bolts The barrel-form er bolts join the barrel to the form er plates and are v ital to m aintaining t he operational and structural integrity of the integrated baffle-for mer-barrel assembly. The primary concern is bolt cracking due to IASSC and fatigue.

A loss of bolt preload due to irradiation-i nduced stress relaxation may exacerbat e fatigue issues in aging plants. The potentia l for flow-induced vibra tion due to loss of bolting constraint would contri bute to overa ll loss of function.

Loss of structural stability i s an operational and safety concern.

Westinghouse Non-Proprietary Class 3 4-13 WCAP-17894-NP September 2014 Revision 0 WCAP-1709 6-NP Component Category ID: W-ID: 7.1 (pen ding) Lower Suppo rt Assembly

- Lower Supp ort Colum n Bolts The lower support col umn bolts attach the support co lumns to the lower core su pport plate. Althoug h the bolts do not directly support the weight of the core, they help m aintain the flatness and integrity of the lower core su pport plate. Cracking from IASCC or fati gue resulting in displacement of the low er core support plate would cause concerns for the operationa l and structur al stability of the lower reactor internals support asse mbly. WCAP-1709 6-NP Component Category ID: W-ID: 7.2 (pen ding) 4.4.3 Existing MRP-227-A Existing components are those that were deter mined to be susceptible to the effects of a least one of the critical degradation m echanisms aff ecting reactor interna ls, but for which existing plant-specific AMP elements were found to be su fficient to m anage aging concerns. The MRP-227-A requirement is for IP2 to e nsure that the MRP-227-A

[9], Table 4-9, and NL 037 [5], Table 5-4, exams are included in plant-specific inspection program

s. Table 4-1 lists all of the Ex isting com ponents applicable to IP2.

Westinghouse Non-Proprietary Class 3 5-1 WCAP-17894-NP September 2014 Revision 0 5 CONCLUSION Indian Poi nt Nuclear Generating Station Unit 2 has d emonstrated a long-term commitment to aging management of reactor internals. The additiona l evaluations and a nalyses completed by the MRP industry group have pro vided clarificatio n to the level of inspection quality needed to determ ine the proper exam ination m ethod and frequen cies. It is the industry position that use of the Aging M anagement Review (AM R) produced by the LRA methodology, co mbined with any additional reactor internals inspections required b y the MRP-227-A i ndustry tables provided in MRP-227-A and the plant-specific AMP, provides reasonable assurance th at the reactor internals passive com ponents will conti nue to perform their intended fu nctions thro ugh the period of extended ope ration. ASME Section XI exam inations identifie d in the AM P for the period of exte nded operation and additional reactor internals inspections discussed in com pliance with MRP-227-A require ments as an integrated inspection pr ogram aligned with ASME Section XI 10

-year ISI exam inations will be tracked by plant-specific procedures and program

s. As discussed, th e industr y MRP-227-A gui delines also provi de for updates as experience is gained through inspection resu lts. This feedback loop will enable updates based on actual inspection experience.

The additiona l reactor internals inspections described in this docum ent, combined with the ASME Section XI ISI program inspections, existing IP2 programs, and use of OERs provide reasonable assura nce that the reactor internals will co ntinue to perform their intended functions throug h the period of exte nded operation and are in full com pliance with IP2 comm itments to manage material degradation in reactor internals.

Westinghouse Non-Proprietary Class 3 6-1 WCAP-17894-NP September 2014 Revision 0 6 REFERENCES

1. U.S. Nuclear Regulator y Commission NUREG-1800, Revision 2

, "Standard Review Plan for the Review of License Ren ewal Applications for Nuclear Power Plants," Decemb er 2010. 2. U.S. Nuclear Regulatory Commission NUREG-1930, Vols. 1 and 2

, "Safety Evaluation Report Related to the License Renewal of Indi an Point Nuclear G enerating Unit Nos. 2 and 3," Novem ber 2009. 3. Entergy Nuclear Engineering Report, IP

-RPT-11-000 36, Rev. 0

, "Indian Poi nt Energy Center Reactor V essel Internals Program

," October 3, 2011.

4. Entergy Document, SEP-ISI-IP2-001, Rev. 2, "I P2 Fourth Ten-Year Interval Inservice Inspection (ISI)/Containment Inservice Inspecti on (CII) Program Plan, Septem ber 12, 2013.

5. Entergy Letter, NL-12-037

, Rev. 0, "License Re newal Application - Revised Rea ctor Vessel Internals Program and Inspection Plan Compliant with MRP-227-A, Indian Poi nt Nuclear Generation Unit Nos. 2 an d 3, Docket Nos. 50-24 7 and 50-286, License Nos.

DPR-26 and DPR-64," Februar y 17, 2012. 6. Westinghous e Report, WCAP-14577, Rev. 1-A, "License Renewal Evaluation:

Aging Management for Reactor I nternals," March 2001.

7. Pressurized Water Reacto r Primary Water Chemist ry Guidelines, Volumes 1 and 2, Revision

6. EPRI, Palo Alto, CA: 20

07. 1014986. 8. Materials Rel iability Progr am: Screening, Categor ization, and Ranking of Reactor Internals Components f or Westinghouse and Com bustion Engineering PWR Design (MR P-191). EPRI, Palo Alto CA: 200

6. 1013234. 9. Materials Reliability Program: Pressurized Water Reac tor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto

, CA: 201

1. 1022863. 10. Materials Reliability Progr am: Inspection Standard for PWR Inte rnals - 2012 Update (MRP-228, R ev. 1).

EPRI, Palo Alto, C A: 2012. 1025147. 11. Westinghous e Report, WCAP-17096-N P, Rev. 2, "Reactor Internals Acceptance Criteria Methodolo gy and Data Requirements,"

December 2009.

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, Revision 2, "Guideline for the Managem ent of Materials I ssues," J anuary 2010. 14. ASME Boiler and Pressure Vessel Code Sec tion XI, 2001 E dition with the 2003 Addenda.

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Westinghouse Non-Proprietary Class 3 A-1 WCAP-17894-NP September 2014 Revision 0 APPENDIX A COMPONENT INSPEC TION DETAILS Figure A-1 Typical Wes tinghouse Internals

Westinghouse Non-Proprietary Class 3 A-2 WCAP-17894-NP September 2014 Revision 0 Figure A-2 IP2 Internals

Westinghouse Non-Proprietary Class 3 A-3 WCAP-17894-NP September 2014 Revision 0 Figure A-3 IP2-CID-0010 Control Ro d Guide Tube Assem bly - Guide Plates (Cards)

Westinghouse Non-Proprietary Class 3 A-4 WCAP-17894-NP September 2014 Revision 0 Figure A-4 IP2-CID-0020 Control Ro d Guide Tube Assem bly - Lower Flange Welds Westinghouse Non-Proprietary Class 3 A-5 WCAP-17894-NP September 2014 Revision 0 Figure A-5 IP2-CID-0021 Upper Internals Assembly - Upper Core Plate Westinghouse Non-Proprietary Class 3 A-6 WCAP-17894-NP September 2014 Revision 0 Figure A-6 IP2-CID-0022 Lower Internals Assembly - Lower Support Forging or Castings Westinghouse Non-Proprietary Class 3 A-7 WCAP-17894-NP September 2014 Revision 0 Figure A-7 IP2-CID-0023 Lower Support Asse mbly - Lower Support Column Bodies (cast)

Westinghouse Non-Proprietary Class 3 A-8 WCAP-17894-NP September 2014 Revision 0 Figure A-8 IP2-CID-0024 Bottom

-mounted Inst rumentation System

- Bottom

-mounted Instrumentation (BMI) Column Bodies Westinghouse Non-Proprietary Class 3 A-9 WCAP-17894-NP September 2014 Revision 0 Figure A-9 IP2-CID-0030 Core Barrel Assembly

- Upper Core Barrel Fla nge Weld Westinghouse Non-Proprietary Class 3 A-10 WCAP-17894-NP September 2014 Revision 0 Figure A-10 IP2-CID-0031 Core Barrel Assem bly - Core Barrel Outlet No zzle Welds Westinghouse Non-Proprietary Class 3 A-11 WCAP-17894-NP September 2014 Revision 0 Figure A-11 IP2-CID-0040 Core Barrel Asse mbly - Upper and Lower Core Barrel Cylinder Girth Welds Westinghouse Non-Proprietary Class 3 A-12 WCAP-17894-NP September 2014 Revision 0 Figure A-12 IP2-CID-0041 Core Barrel Asse mbly - Upper and Lower Core Barrel Cylinder Axial Welds Westinghouse Non-Proprietary Class 3 A-13 WCAP-17894-NP September 2014 Revision 0 Figure A-13 IP2-CID-0050 Core Barrel Assembly

- Lower Core Barrel Fla nge Weld Westinghouse Non-Proprietary Class 3 A-14 WCAP-17894-NP September 2014 Revision 0 Figure A-14 IP2-CID-0060 Baffle-form er Assembly - Baffle-edge Bolts Westinghouse Non-Proprietary Class 3 A-15 WCAP-17894-NP September 2014 Revision 0 Figure A-15 IP2-CID-0070 Baffle-form er Assembly - Baffle-fo rmer Bolts Westinghouse Non-Proprietary Class 3 A-16 WCAP-17894-NP September 2014 Revision 0 Figure A-16 IP2-CID-0071 Core Barrel Assembly

- Barrel-former Bolts Westinghouse Non-Proprietary Class 3 A-17 WCAP-17894-NP September 2014 Revision 0 Figure A-17 IP2-CID-0072 Lower Support Asse mbly - Lower Support Column Bolts

Westinghouse Non-Proprietary Class 3 A-18 WCAP-17894-NP September 2014 Revision 0 Figure A-18 IP2-CID-0080 Baffle-former Assembly - Assembl y

Westinghouse Non-Proprietary Class 3 A-19 WCAP-17894-NP September 2014 Revision 0 Figure A-19 IP2-CID-0090 Alignm ent and Interfacing Com ponents - Internals Hold-down Spring Westinghouse Non-Proprietary Class 3 A-20 WCAP-17894-NP September 2014 Revision 0 Figure A-20 IP2-CID-0100 Therm al Shield Assembly - Thermal Shield Flexures