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)

Text

We stingh ou se Non-Pro pri etary Cla ss 3 WCAP-17894-NP September 2014 Revis i on 0 Component Inspection Details Supporting Aging Management of Reactor Internals at Indian Point Unit 2

W e sting hou se N o n-Prop r ietary 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 do es contain such information, the Recipient's acc e ptan ce of this doc ument constitutes agreement that this information in document form (o r 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 i n co mpliance with all U.S. export control regul ations. Recipient shall include this notic e with any reproduced or exc erpted portion of t h is document or a n y document derived from, based on, incorpor ating, usi ng or relying on th e information contained in this d o cument. *Electronically approved rec o rds are aut h ent i cated in t h e el ect ro ni c d o c u m e nt m a nagem e nt sy st em. Westinghouse Electric Com p any LLC 10 0 0 West i n gh ous e D r i v e Cr anb e rr y To wn sh i p , PA 1 606 6, U S A © 201 4 W e sting hou se Electr i c Co m p an y LLC All Rig h t s Reserv ed W C AP-17894-NP.doc-0915 14 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 i gn and Analysis I Charles R. Schmidt*

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

Reactor Internals Aging Managem e nt September 2014 Approved: Patricia C. Paesano *, Manager Reactor Inter n als Aging Manage m e nt W e sting hou se N o n-Prop r ietary Class 3 ii W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 REC O RD OF REVIS I ONS Rev. Da te Revisi on Desc ription 0 See E D MS Ori g inal Issue

W e sting hou se N o n-Prop r ietary Class 3 iii W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 T A B LE OF CONTENTS LIST OF TABLES

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

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

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................. viACKNOWL EDGEMENT S ..............................................................................................................

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

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

...... 2-13PROGRAM OWNER

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

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

14.1INTRODUCTION

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4-14.2DETECTION OF AGING EFFECTS

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4-44.3INSPECTION RESULTS REPORTING FORMAT

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4-74.4COMPONE NTS ..............................................................................................................

4-74.4.1Prim ary ............................................................................................................

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

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

4-135CONCL U SION ....................................................................................................................

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

......... 6-1APPENDI X ACOMPONE NT INSPECTI ON DETAILS

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A-1 W e sting hou se N o n-Prop r ietary Class 3 iv W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 LIST OF T A BLES Table 4-1 IP2 Reactor Internals Co mponents Requiring Additi onal Inspections During the License Renewal Term

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4-2 W e sting hou se N o n-Prop r ietary Class 3 v W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 LIST OF FIGURES Figure A-1Typical Westinghou se Internals

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A-1Figure A-2IP2 Internals

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... A-2Figure A-3IP2-CID-0010 Control R o d Guide Tube Assembly - Guide Plates (Cards) ..................

A-3Figure A-4IP2-CID-0020 Control R o d Guide Tube Assembly - Lower Flange Welds

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A-4Figure A-5IP2-CID-0021 Upper Internals Asse m b ly - Upper Core Plate

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A-5Figure A-6IP2-CID-0022 Lower Internals Assem b ly - Lower Support F o rging or Castings

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A-6Figure A-7IP2-CID-0023 Lower Support Assem b ly - Lo wer Support Col u mn Bodies (cast)

........ A-7Figure A-8IP2-CID-002 4 Bottom-m o unted Instrum e ntation S y st em - Bottom-m ounted Instrumentation (BMI) Col u m n Bodies ..........................................................................

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

- U pper Core Barrel Flange Weld

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

- Core Barrel Outlet Nozzle Welds

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

- U pper and Lo wer Core Barrel Cy lin der Girth Welds

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

- U pper and Lo wer Core Barrel Cy lin der Axial Welds

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

- Lo wer Core Barrel Flange Weld

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

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

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

- Barrel-former Bolts

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A-16Figure A-17IP2-CID-0072 Lower Support Assem b ly - Lower Support Col u mn Bolts ..................

A-17Figure A-18IP2-CID-0080 Baffle-former As sem b ly - Assem b ly ...................................................

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

.... A-19Figure A-20IP2-CID-0100 Thermal Shield Assem b ly - Therm a l Shield Flexur es .........................

A-20 W e sting hou se N o n-Prop r ietary Class 3 v i W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 LIST OF A CRONYMS AMP Aging Manag e m e nt Program Plan AMR Aging Manag e m e nt Review ASME A m eri can Society of Mechanical Engineers B&PV boiler and pre ssure vess el B&W Babcock

& W ilcox BMI bottom

-m ounted instrum e n t ation BWR boiling water reactor CE Co m bustion Engineering CFR Code of Federal Regulations CID Co m ponent I n spection Detail EPRI Electric Power Research Institute EVT enhanced visual testing (a visual NDE method that i n cludes EVT

-1) FMECA failure m ode, ef fects, and criticality anal y s is GALL Generic Agin g Lessons Learned I&E Inspection and Evaluation IASCC irradiation assisted stres s c o rrosion cracking ID identification INPO Institute of Nuclear Power Operations IP2 Indian Poi n t Nuclear Generating S tation Unit 2 ISI in-service ins p ection LCP lower core plate LRA License R e ne wal Applicati o n MRP materi als reli abilit y program MSC PWROG Materials Subcommittee NDE nondestructiv e exam inatio n NEI Nuclear Ener gy Institute NRC Nuclear R e gulatory Commission NSSS nuclear steam suppl y s y ste m OEM Original Equi pment Manufacturer OER Operating Experience Rep o rt PWR pressurized w a ter rea c tor PWROG Pressurized W a ter R eacto r Owners Gro up (form erly WOG) RCS reactor coola n t s y ste m R V I reactor vess el internals SCC stress corrosion cracking SER Safety Evalua tion Report SRP S tandard Review Plan SSC sy stems, structures, and com ponents UCP upper core pl ate USP upper supp ort plate UT ultrasonic testing (a vol um etric NDE method) VT visual testing (a visual NDE method that includes VT-1 and VT-3) WOG W estinghouse Owners Group W e sting hou se N o n-Prop r ietary Class 3 v ii W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 ACKNOWL EDGEMEN T S The authors would like to thank the m e m b ers of the Entergy Reac tor Internals Aging Manage m e nt Program T e am , including Bob Dolansky and our associates at Westinghouse for their efforts in supportin g th e developm en t of this repor

t.

W e sting hou se N o n-Prop r ietary Class 3 1-1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 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 a r Regulatory Co mm is sion (NRC) Standard Review P lan (SRP) [

1]. Indian P o int Nuclear Generating Station Unit 2 (IP

2) will be granted an extended license by the NRC through a Sa fety E v aluation Report (S ER) as documented in NUREG-1930 [

2 and 16]. License Ren e wal Co mmitmen t 30 of [2 and 16], "PWR Vessel Internals Program ," committed IP2 to:

1. Participate in industry programs for investiga ting and m a naging a g ing effects on reactor internals.

2. Evaluate and im plement the results of the industr y pr ograms as ap plicable to the reactor inter n als. Upon com p letion of t h ese program s IP2 subm itted an insp ection pl an for reactor internals to t h e NRC for review and approval.

On Septem ber 29, 201 3, IP2 en tered what is called the period of "tim ely renewal" while the NRC continues it s consideration of Ente rg y's application t o renew its operating license.

An Aging Ma nagem e nt Program Plan (AMP) [

3] was developed b y IP2 that capt u red the ind u s try 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 reli abilit y program (M RP) and the Pressurized Water Reacto r Owners Gro up (PWROG).

The IP2 A M P was later supplemented by

[5] to provide com p liance 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 n sp ections as required per the American Society of Mechanical E ngineers (ASME) Boiler and Pressure V ess el (B&PV) Code, Sec tion XI progra

m. This docum ent co ntains com p o n ent inspectio n details (C IDs) and suppl em ental information that p r ovides direction on t h e location, n u m b er, configuration, an d inspection requirements of the items that com p rise the scope for the additional reactor internals inspections at IP2. As such, this doc ument suppor ts the plant-specific im plem entation of additional re actor intern als inspections during the license renew a l ter m.

W e sting hou se N o n-Prop r ietary Class 3 2-1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 2 B A CKGROUND There are fiv e key drivers that affect the continue d succes sful operation of reactor vessel internals (RVI) at operating P WRs: License R e ne wal/Life Exte nsion Reliability and Maintenance Inspection In dications Issue-Specifi c Res earch Co m ponent-S pecific Aging Managem e nt and Inspection Cost Redu ction The Code of Federal Regu lations (CFR) Maintenance Rule, 10 CF R 50.65 requi res m onitoring of certain sy stems, 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 dem o nstrate effect ive management of agin g, s h ifting t h e em phasis fro m identify ing aging m echanis ms to m a nagi ng their eff ects during the license rene w a l period. Together these two regulatory requirements consider all active a nd passive co m ponents that are r e quired for safe operation of t h e plant with aging m a nagement focu sing on passive, long-lived structures and com ponents. A plant entering or requesting license renewal is ty pi cally required t o define and execute an AMP for reactor internals.

For m a ny y ears the U.S. nu clear power industr y has been actively engaged in eff o rts to supp or t the industr y goal of responding to the regulat or y require ments on m a naging agi ng de gradation in r eactor internals. Various pro g rams have been underway to d e velop gui delines for managing t h e 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 o r 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 ope rating U.S. reactor designs - Westingh ouse, Co m bustion Engineering (CE), and Babcock &

Wilcox (B&W).

The MRP established a framework and strategy for t h e aging m a n a gem e nt of PWR internals co m ponents using pr oven and fam iliar methods for inspection, monitoring, sur v eillance, and communication.

Factoring in t h e accu m u lat e d industr y re search data, t h e following elements of an AMP were further refined [7 an d 8]: Screening criteria were developed, consi d ering chem ical co m position, neutr on fl uence exposure, tem p erature h istory, and re presentative stress leve ls, f o r deter m ining the relative susceptibility of PWR internals co m ponents to each of eight postulated aging m ech anism s. PWR internals co m ponents were categorized, based on the screening criteria, as f o llows: Co m ponents for which the effects fro m the pos tulated aging m echan ism s are insignificant Co m ponents that are m oderately susceptible to the agi ng effects Co m ponents that are significantly susceptible to t h e aging effects W e sting hou se N o n-Prop r ietary Class 3 2-2 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Functionalit y assessments were perfor m ed based on r e presentative PWR internals co m ponents and com ponent assem b lies using irradiat ed and aged materi al properties, to deter m ine the eff ects of the degrad ation m e chanism s on co m ponent fu nctio nalit y. Aging m a nag e m e nt strat e gies wer e deve loped co m b ining the functionality assess ment results with contributi ng f actors to deter m ine the appropriate aging management methodology, baseline examination tim ing, and the need and ti ming of subs equent inspections. Item s considered included com ponent accessibility , operating experience, existi ng ev aluations, and prior e x am ination results.

The industr y finalized initial Inspection and Evalua tion (I&E) Guidelines for reactor internals and subm itted the docum ent to the NRC with a request for a for m al SER. A supporti ng docum ent addressing inspection requirements was also co m p leted and pr ov ided to t h e NRC to suppor t the I&E Guidelines review. A third docum ent, which was generate d by th e industr y thr ough the PW ROG Materials Subcomm itt e e (MSC), provides detailed engineeri ng criteria for evaluating acce ptance of inspection outcom e s. This industr y de veloped g u id ance is contained within t h e following t h ree documents: MRP-227-A [

9] (hereafter referred to as "the I&E Gui d elines" or sim p ly "MRP-227-A") provides the industr y background, li sting of gener ic react or internals co m ponents requiring inspection, type of nonde structive examination (NDE) required for each co m ponent, tim ing for initial inspections, and direction f o r evaluating inspection results. MRP-228 [1 0] provides gui dance on the qualificati on and dem onstration of the N D E techniques and other crit eria pertaining to the actua l performance of the inspections.

WCAP-1709 6-NP [1 1] provides direction o n engine ering evaluations of inspect ion o u tcom es to determ ine acceptabilit y for continued service.

The IP2 reactor internals are integral with the reactor coolant s y stem (RCS) of a Westinghous e four-loop nuclear steam suppl y s y ste m (N SSS). 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 a tion of IP2 i n ternals is provided i n Figure A-2.

As described in NUREG-1 930 [2 and 16], the applican t described the RVIs, whi c h consist of t w o basic asse mblies: a n upper internals ass e m bly that is re m oved during each refueling operation to obtain acce ss to the reactor core and a lower internals asse mbly that can be re m o ved following a co m p lete c o re off-load. The reactor core is positioned and supported by the u pper internals and lower in ternals assem b lies. The indivi dual fue l assem b li es are positioned by fuel alignment pins in t h e upper core plate (UCP) and the lower core plate (LCP). These pins control the orienta tion of the c o re 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 t h e head/vessel alignment pins. T h e lower internals are aligned to the vessel by the lower radial support/clevis assemblies an d by t h e head/v essel alignment pins. T hus, the core is aligned with the vessel by a num ber of interfacing com ponents.

W e sting hou se N o n-Prop r ietary Class 3 2-3 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 The lower internals assem b ly is supporte d in the v essel by clam ping to a ledge be low the vessel-head mating surface and is closely gui ded at t h e bottom b y radial support

/clevis assemblies. The bo ttom of the upper internal s assem bly is closely g u ide d b y t h e co re barrel alignment pins of t h e lower internals asse mbly. Upper Int e r n als Asse mb ly The major sub-assem b li es that constitut e the upper intern als assembl y are the: (1) UCP; (2) upper support colum n as se m b lie s; (3) control rod guide tube asse mblies; and (4)

USP. During reactor operation, t h e upper inter n als 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 u m n s 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 n s. The USP design at IP2 is d esignated as a top hat design. The UCP is perforated to per m it coolant to pass from t h e core below into the upper plenum defined by the USP and UCP. The UCP positions and laterally supp orts the core b y fuel align m ent pins extending below the plate. The UCP contac ts and preloads the fuel asse m b ly springs and thus m a intains contact of the fuel asse mblies wi th the LCP during reactor operation.

The upper support col u m n s vertically position the U C P and are designed to take the uplifti ng hy draulic flow loads and fuel spring loads on the U C P. Th e cont rol rod guide tubes are bolted to the USP and pinned at the UCP. Guide tube cards are located with in the control rod gu ide tub e assem bly 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 e m bly. The pins m u st laterally support t h e UCP so that the plate is free to expand r a dially an d move axiall y d u ring differential therm a l expansion between the upper inter n als and the core barrel. The UCP alignment pins are the interfacing com ponents between the UCP and the core barrel.

Lower Int e r n als Asse mb ly The fuel assem b lies are supported i n side 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 m e ter e d contro l of reactor coolant flow into each fuel assem b ly. The L C P is elevated above the l o wer support casting by support col u m n s and bolted to a ring su pp ort attached to the inside d iam eter of the core barrel. The support c o lum n s transmit vertical fuel assembly loads from the LCP to the m u ch thicker lower support casting, which provides support for the core.

The prim ary f unction of the core barrel is to supp ort th e core. A large num ber of co m ponents are attached to the core barrel, including the baffle/former as se m b ly, the core barrel outlet no zzles, the ther m a l shield, the alignm ent pins that engage the UCP, the lowe r support casting, and the LCP.

The lower radial support/clevi s assem blies restrain large transverse m o tions of the co re barrel, but at the same time allow unrestricted radial and axial thermal expansion.

W e sting hou se N o n-Prop r ietary Class 3 2-4 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 The baffle and form er assem b ly consists of ver tical pl ates called baffles and hori z ontal suppor t plates called for m er s. The baffle plates are bolted to the fo rmers by the baffle/for m er b o lts, and the formers ar e attached to the core barrel i n side diamete r by the barrel/former bolts. Baffle plate s are s ecured 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 m er a s se m b ly form s the interface between the c o re 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 be nefit is to r e duce the neutron flux on t h e vessel.

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

In the up per internals assem bly , the USP, the up pe r support columns, and the U C P are considered core support struct ures. In the l o wer internals assem bly , t h e LCP, the l o wer support casting, the lower support colum n s, the core barrel including t h e core barrel fl ange, the radial support/clev is assem blies, the baffle plates, and th e form er plates are cl assified as core su pport structur es. All RVIs ar e re m ovable for their inspection, an d for inspection of the vessel inte rnal surface.

Based on the co m p leted evaluations, the RVI co m ponents are cat egorized within MRP-227-A as "Prim a ry" com ponents, "Expansion" com ponents, "E xisting Progr a m s" co m ponents, or "No Additional Measures" co m ponents. Descriptions of t h e final categories are as follows: Pri m ary Those PWR internals that are highly sus ceptible to the effect s of at least one of the eight aging mechanis ms were plac ed in the Prim ary group. The a g ing m a nage ment requirements that ar e needed to ensure functional ity of Prim ar y com ponents are described in the I&E guidelines. The Prim ary gro u p also include s co m ponents that have sho w n a degree of tolerance to a specific aging degradation effect, but for which no hi ghl y suscep tibl e co m ponent exists or for which no hi ghl y susceptible com ponent is accessible.

Expansion Those PWR internals that are highly or m oder a tely susceptible to the effects of at least one of the eight aging m echanisms, but for which f unctiona lity a sses s m ent has shown a degree of tolerance to those effects, were placed in the E xpa nsi on gr oup. The schedule for im plemen tation of agin g management requirem e nts for Expansio n com ponents depends on t h e findin g s from Primary co m ponent exam inations at indivi dual p lants. Existing Program s Those PWR internals that are su sc eptible to the effect s of at least one of the eight aging mechanis ms and for which generic and plant-sp ecific existing AMP ele m ents ar e capable of managing those effects, were placed in the Existing Programs group.

W e sting hou se N o n-Prop r ietary Class 3 2-5 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 No Additiona l Measures Pr ograms Those PWR internals for which the effects of a ll eight aging m echan is m s ar e below the scre e ning criteria were placed in the No Additiona l Measur es gr oup. Additional co m ponents were placed in the No Additi onal Measures group as a result of a fa ilure m ode, effects, and criticality anal y s i s (FMECA) an d the functionality assessment. No furt he r action is required by these guidelines for No Additiona l Measures co m ponents agi ng m a nagement.

The categoriz ation and ana ly s is used in the devel opm ent of MRP-22 7-A were not intended to s upersede any ASME B&PV Code Section XI requirem e nts. Com ponents 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 e nts that re m a in in effect and may only be altered as allowed b y 1 0 CF R 50.55a. A listing of t h e IP2 RVI com ponents and subcom ponents already r e viewed by t h e NRC in the SER granting life extension that are subject to aging m a nag e m e nt 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 co m ponents that require additional inspe c tions a nd the com ponents with existing ASME Section XI inspections that are cr edited for m a nagi ng aging in R V I are su mmarized by MR P-227-A classification categories an d shown in T a ble 4-1 of th is report.

W e sting hou se N o n-Prop r ietary Class 3 3-1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 3 PROGRAM OW NER The PWR Ve ssel Internals Program is es tablished in accordance wit h Entergy's "N EI 03-08 Materials Initiative Process"

[12]. T h e successful i m plem enta ti on and com p rehensive long-ter m management of the IP2 RVI AMP will require the integration of Ente rgy organizati ons (bot h cor porate and at IP2) and interaction with m u ltiple industr y organi zations 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 a nagem e nt, and will address/im plement industr y guidance stemming from those activities, as appropriate u nder Nuclear Energ y Instit ute (NEI) NE I 03-08 [1 3] practices.

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

W e sting hou se N o n-Prop r ietary Class 3 4-1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 4 COM P O N ENT INSPECTI ON DETAILS OF TH E INDIAN POI N T UNIT 2 REACTOR INTERNALS

4.1 INTRODUCTION

The U.S. nucl ear industry, t h rough the com b ined 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 pri m ary objectiv e for the indu stry effort and each individual plant program is to e n sure the lo ng-ter m integrity and safe operation of the reactor internals co mponents and overall reliability of the plant by pr oactively m a naging aging.

The IP2 reactor internals AMP utilizes a co m b ination of prevention, m itigation, and condition m onitoring to m a nage aging in susceptible reactor internals co m p onents. Where applicable, credit is take n for existing progra m s su ch as water che m ist ry [7] and in spections prescribed by the ASME Section XI In-Service Inspe c tion (ISI) Program

[4], co m b ined w ith additional reactor internals inspections or evaluations as reco mmend ed by MRP-227-A.

The purpose of this IP2 CI D WCAP is to supp ort IP2-specific im plem entation of the generic industr y additional reactor internals inspections with a focu s on com ponents that are not already included in existing IP2 i n spection 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 co m ponents that com p rise the IP2 additi onal reactor internals MRP-227-A inspe c tions. CIDs illustrating key co nsiderations of the re quired MRP-227-A inspecti on for t h e com ponent is included for al l applicable IP2 MRP-22 7-A Primary and Expansio n com ponents A listing of t h e com ponen ts applicable to IP2 ident if y ing the MRP-227-A category

, required i n spection, and corresponding CID is provided in T a ble 4-1. The Control Rod Guide Tube Assembly: G u ide Tube Support Pins (Split Pins) ar e not include d in Table 4-1 because they are m a naged by Origi n al Equipm ent Manufacturer (OEM) reco mmendations in accordance w ith Applicant/Licensee Action Ite m 3 of the NRC Safety Evalua tion for MRP-227 as described in Section 3.6 of NL-12-037 [

5].

W e sting hou se N o n-Prop r ietary Class 3 4-2 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Table 4-1 IP2 Reactor Internals Co mponents Re quiring Addi tional Inspections During the Licens e Renewal T e r m Component MRP-227-A CID WCAP-17096-NP ID (2) Category Inspection T y pe Control Rod Guide Tube Assembly - G u ide Plates (Cards)

Prim ary VT-3 IP2-CID-001 0 W-ID: 1 Control Rod Guide Tube Assembly - L o wer Fla nge Welds Prim ary EVT-1 IP2-CID-002 0 W-ID: 2 Upper Internals Assem b ly - Upper Core Pl ate Expansion EVT-1 IP2-CID-002 1 W-ID: 2.1 Lower Internals Assem b ly - Lower Sup port For g ing or Castings (3) Expansion EVT-1 IP2-CID-002 2 W-ID: 2.2 Lower Suppo rt Assembly - Lower Supp ort Colum n B odies (Cast)

Expansion EVT-1 IP2-CID-002 3 W-ID: 2.3 Bottom-m o u n ted Instrum e ntation S y ste m - Bottom-m ounted Instrumentati on (BMI) Colum n Bodies Expansion VT-3 IP2-CID-002 4 W-ID: 2.4 Core Barrel Assembly - U pper Core Barrel Fla nge Weld Prim ary EVT-1 IP2-CID-003 0 W-ID: 3 Core Barrel Assembly - C o re Barrel Ou tlet Nozzl e Welds Expansion EVT-1 IP2-CID-003 1 W-ID: 3.1 Core Barrel Assembly - U pper and Low e r Core Barr el Cy linder Girth Welds Prim ary EVT-1 IP2-CID-004 0 W-ID: 4 Core Barrel Assembly - U pper and Low e r Core Barr el Cy linder Axial Welds Expansion EVT-1 IP2-CID-004 1 W-ID: 4.1 Core Barrel Assembly - Lower Core Barrel Flange Weld (4) Prim ary EVT-1 IP2-CID-005 0 W-ID: 5 Baffle-for m er Assembly - Baffle-edge Bolts Pri m ary VT-3 IP2-CID-0060 W-ID: 6 Baffle-for m er Assembly - Baffle-for m er Bolts Pri m ary UT IP2-CID-0070 W-ID: 7 Core Barrel Assembly - B a rrel-for m er 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 m er Assembly - Baffle-for m er A ssembly Prim ary VT-3 IP2-CID-008 0 W-ID: 8 Alignment and Interfacing Co m ponents - Internals Hold-down Spr ing Prim ary Special (1) IP2-CID-009 0 W-ID: 9 Therm a l Shield Assem b ly - Th erm a l Shield Flexures Prim ary VT-3 IP2-CID-010 0 W-ID: 10 Core Barrel Assembly: C o re Barrel Flange Exis ting VT-3 Not applicable Not applicable Upper Internals Assem b ly: Upper Suppo rt Ring or S k i r t (5) Existing VT-3 Not applicable Not applicable W e sting hou se N o n-Prop r ietary Class 3 4-3 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Table 4-1 IP2 Reactor In ternals Components Req u iring Addi tional Inspections During the Licens e Renewal T e r m (cont.) Component MRP-227-A CID WCAP-17096-NP ID (2) Category Inspection T y pe Lower Core Plate Existing VT-3 No t applicable Not applicable Alignment and Interfacing Equipm ent: Clevis Insert Bolts Existing VT-3 Not applicable Not applicable Alignment and Interfacing Equipm ent: Upper Core Plate Align m en t Pins Existing VT-3 Not applicable Not applicable Bottom Mounted Instrum e ntation S y ste m: Flux Thimble T ubes Existing ET Not applicable Not applicable No tes: 1. Ma na ged by plant-speci fic measurem ent program

s. 2. C o nfi rm a t i on of i d e n t i fi cat i on (I D) num ber pen d i n g NR C ap pr oval o f WC AP-1 7 0 9 6-NP. 3. Th is co m p on en t is a casting at IP2. 4. At IP2 th is weld is th e l o wer co re b a rrel t o lower su ppo rt cast i ng wel d. IP 2 do es not h a ve a l o we r c o r e ba rrel fl an ge. 5. IP 2 has a t o phat desi gn. T h ere f o r e, t h er e i s n o s u p p o rt ri ng o r s k i r t; ho weve r, t h e vert i cal sect i ons of t h e t o p h at wi l l be i n s p ected.

W e sting hou se N o n-Prop r ietary Class 3 4-4 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 4.2 DET E CTI O N 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 m a ge. The reco mmendat ions supporting aging m a n a ge m e nt for the reactor internals, as defi ned by the industry and contained in t h is report, are built around three basic inspection techniques: (1) vi sual testing (VT), (2) ultrasonic testing (UT), and (3) phy s ical mea s ure m ent. Three differe nt visual techniques are included:

VT-3, VT-1, and EVT-1. T hose additional reactor internals insp ections that are taken from the MRP-227-A reco mmen d ations will be applied through use of th e MRP-228 I n spection Standard [

1 0]. Detection of indications th at are required b y t h e ASME Section XI ISI Program is well-established and fi eld-proven throug h the a pplication of the Section XI ISI Program. The assu m p ti ons and process used to select the appropr iate inspection technique are described in the following su b sections. Inspection standards developed b y the in dustr y for the appl ication of the se techniques for additional reactor interna ls inspections are docum ented in MRP-228.

VT-1 V isual Exa m inations 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 b l e d in two vert ical sections. Therefore, no add itional VT-1 i n spections ov er and above those required b y ASME Section XI ISI have been specified in MRP-227-A at IP2.

EVT-1 Enhanced V isual Exa m ination 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 exam ination has been identified for inspection of components where surface-br eaking flaws are a potential concern. Any visual inspection for cracking requires a rea s onable expectation that the flaw length and cr ack m outh op ening displac e m e nt meet the resolution requirem e nts of the observ a tion technique. The EVT-1 spec ification augments the VT

-1 re quirements to pro v ide m o re rigorou s inspection standards for stress corrosion cracking (SCC) and has been dem ons trated for si milar inspections in boiling water reactor (BW R) internals. Enhanced visual exa m inati on (i.e., EVT-1) is also con ducted in accordance with the requirem e nts descri bed for vi sual exam ination (i.e., VT-1) with additional requirem e nts (such as camera scanning speed) currently being dev e loped b y t h e industr y. Any reco mmendat ion for EVT-1 inspection will require additio nal analy s is to establi sh flaw-tolerance criteria. The industr y is currently d e veloping a c onsensus approach for acceptance criteria m e thodolo g ies to support plant-specific additional reactor internals examinations. Ente rgy has been an active participant in these initiatives and will follow the industry directive. These acceptance criteri a methodologie s m a y be deter m ined ei ther generical ly or on a pla n t-specifi c basis because both loads and co m ponent di mensions may var y fro m plant to plant within a t ypical PWR design.

VT-3 Exam ination for General Condition Monit o ring 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 e d. The VT-3 exam in ation is inten d ed to ident i fy indi vi dual c o m ponents with significant levels of existing W e sting hou se N o n-Prop r ietary Class 3 4-5 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 degradation. As the VT-3 exam ination is not inten d ed to detect th e early stages of com ponent cracking or other incipien t degradation effects, it sho u ld n o t be use d when failur e of an indi vi dual com ponent could threaten either plant safety or operational stability. The VT-3 exam ination m a y b e appropriate for inspecting hi ghl y red unda nt com ponents (such as baff le-edge bolts), where a sin g le failure do es not co m p ro m i se t h e function or integrity of the critical as s e m b ly. The acceptan ce criteri a for visual exam i n ations c onducted 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 dis p lacement of parts to the extent that co m ponent function m a y be im pa i r e d Loose, m i ssing, cracked, or fractured parts, bolting, or fasteners Foreign m a te rials or accu m u l a tion of corrosion prod ucts that could interfere with control rod m o tion or could result in b lockage of co olant flow thr ough fuel Corrosion or erosion that reduces the nominal section thickness by m o r e than five percent Wear of m a ti ng surfaces that m a y lead to loss of func tion Structural degradation of i n terior attachments such that the original cross-se ctional area is reduced m o re 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 a nce criteri a are not generally required. In a n y particular reco mmendation for VT-3 vi sual exa m ination, it should be possible to identify the specific conditi ons of c oncern. For i n stance, 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 m ponent m a y be very different from the guidelines for wear in a large structural co m ponent. Ultrasonic T e sting 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 p onent is requi red to appl y the ultrasonic signals, the flaw m a y exist in the bulk of the material.

In this pr oposed strategy , UT inspections h a ve been reco mmended exclusively for d e tection of flaws in bolts. F o r the bolt ins p ections, any bolt wit h a detected flaw should be a ssu m e d 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 n spection opportunit

y. It ha s generally be en observed t h roug h exam ination performance dem onstrations 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.

W e sting hou se N o n-Prop r ietary Class 3 4-6 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Failure of a single bolt doe s not com p romise the functi on of the entire ass e m bly. Bolting s y stems in the reactor internals are highly redundant.

For any s y stem of bolts, i t is possible to de m onstr ate m u ltiple mini m u m acceptable bolting patterns.

The evaluation program m u st dem onstrat e that the rem a i n ing bolts meet the requ irem ents for a m ini m u m bolting pattern f o r contin ued operation. Th e evaluation procedures m u st also de m onstr ate tha t the pattern of re m a ining bo lts contains sufficient m a r g in such that continuation of the bolt failure rate will not result in f a ilure of the sy stem to m eet the requirements for m i nim u m acceptable bo lting pattern before the next inspection.

Establish m en t of the m inim u m acceptable bolting pa ttern for any s y stem of bolt s requires analy s is to dem onstrate t h at the sy ste m will m a intain reliabilit y and integrit y in continuing to perform the intended function of th e co m ponent. This analy s is is highl y pla n t-specific. Therefore, any reco mmendation for UT inspection of bolts assumes that the plant owne r will work with the designe r to establish mini m u m acceptable bo lting patterns prior to t h e insp ection to support continued operation.

For Westinghous e-designed plants, m inim u m 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 en t of the PWROG doc uments and has full acce ss and use.

Phy s ical Measurem ent Examination Continued f u nctionalit y can be confirm e d b y ph y s ical mea s ure m ents to evaluate the im pa ct cau sed by various degradation m echanism s such as wear or lo ss of functionality as a result of loss of prel oad or materi al defor m ation. For IP2, direct physical m easur e ments ar e required only for the internals hold-down spring. An alternate option is to replace the existing internals hold-down spring w h ich could eliminate the ne e d for m eas urem e nts.

W e sting hou se N o n-Prop r ietary Class 3 4-7 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 4.3 INSPECTI ON RESUL T S RE PORTING FORMAT Entergy IP2, the MRP, and the PWROG approaches to aging m a nag e m e nt are based on the Gen e ric Aging Lessons Learned (GALL)Report approach as detailed in NUREG-1801 [

1 5]. This ap proach includes determining which reactor internals passive com ponents ar e m o st susceptible to the aging mechanisms of concern and then determining t h e pr oper inspection or m itigating program that provides reasonable assurance that the com ponen t 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 o r the initial basis of the License Renewal Application (LRA) that resulted in the NRC SER in NUREG-1930 [

2 and 1 6]. A key elem en t of the MRP-227-A gui deli ne is the repor ting of age-r e lated degradation of reactor vessel co m ponents 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 n spection inf o rm atio n and will share its own operating experience with the industr y t h rough those groups or I n stitute of Nuclear Power Operations (INPO), as app ropriate.

The com ponent nam ing no menclature im plemented b y the in dustr y in PWROG WCAP-1709 6 [1 1] form s 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 R P-227-A, which is wh y t h e W-ID numbers are listed as "pending". L o cation indic a tors for each com pone nt item ar e based on the IP2-specific design.

4.4 COMPONENTS

The Westinghouse reactor internals are part of the R C S and located inside the r eactor pressur e vessel. The reactor internals are long-lived passive structural co m p onents. The i n tended funct ions are to support core cooling, enab le control rod insertion, an d m a intain the integrit y of the fuel. Internals co m ponents are classifi ed 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 rem oved during each refueling operation to obtain acces s to the reactor core , and a lower internals ass e m bly that can be rem oved following a co m p lete co re off-load.

T h e lower internals assem bly is supported in the vessel by clam ping to a ledge below the vessel head ma ting surfac e and is close ly guided at the bottom by radial support clevis assemblies. The upper internal s assembly is clam ped at this same ledge by the reactor vessel head. The bottom of the upper intern als assembly is c l osely guided by the core barrel alignment pins of the lower internals assem bly. All of the assem b lies and i ndividual com ponents making up the Westinghous e reactor inter n als were considered in the process of developing t h e MR P-227-A requirem e n ts. The com p l e te cat egorization process is su mmarized in MRP-227-A and support ing basis documents. Brief descriptions of the co m ponents required to support agi ng management of reactor internals at IP2 are included in the following su b sections by MRP-227-A Prim ary , Exp a nsion, and E x isting catego r ies.

W e sting hou se N o n-Prop r ietary Class 3 4-8 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 4.4.1 Primary MRP-227-A Pri m ary components are those that were deter m ined to be highl y s u sceptible to the effects of a least one of the critical degradation m e chanism s a ffecting reactor internals or a co m ponent for which no highl y susceptible com ponent exists or was directly accessible. Table 4-1 contai ns a co m p lete listing of all of the Pri m ary com ponents applicable to IP2.

Control Rod Guide Tube Asse mbly - G u ide Plates (Cards)

The control rod guide tube asse mbly 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 h e control r ods when withdrawn. The g u idan ce 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 y pically been observ e d in the low est guide card levels.

Guidance hole wear can ca use lack of alignm ent. Lack of alignm ent may im pact control r od dr o p tim es. In the worst-case s cenario, control rods may jam and prevent full insertion which would caus e a technical specification co m p liance c oncern. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 1 (pendi ng) Control Rod Guide Tube Assembly - L o wer Flange Welds The control rod guide tube asse mbly provides alignm e n t 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 o ment to the control rod guide tube asse mbly. Ma xim u m bending stresses tend to occur near the top of t h e c ont inuous guida nce section (upper flange weld location as identified in Figure A-4). Stresses ma y lead to form ation of SCC or fatigue cracks. Weld c r acking may lead to loss of stiffness in the guide tube assemb ly and loss of support capability y i elding a loss of structural stability

. Excessive deflection could im pede control assem b ly insertion, resulting in a failure to perform its in tended functi on. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 2 (pendi ng) Core Barrel Assembly - U pper Core Barrel Flange Weld The upper co re barrel flan ge weld is an integral co m p onent of t h e p r im ary core support structu r e. The primary conc ern with regard to ag ing is through-wall cracking in the weld as a r e sult of stress corrosion.

Actively gr o w ing thro ugh-wall flaws would req u ire 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 lar g e inactive flaws are likely to be m a nageable even if found

. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 3 (pendi ng)

W e sting hou se N o n-Prop r ietary Class 3 4-9 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Core Barrel Assembly - Upper and Lo we r Core Bar rel Cy linder Girth Welds The core barrel cy lin der gi rth welds are an integral com ponent of the prim ary co re support str u cture. The primary conc ern with regard to ag ing is through-wall cracking in the weld as a r e sult of stress corrosion.

Actively gr o w ing thro ugh-wall flaws would req u ire 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 lar g e inactive flaws are likely to be m a nageable even if found

. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 4 (pendi ng) Core Barrel Assembly - Lower Flange Weld The lower flange weld is an integral co m ponent of th e prim ary co re support str u cture. This weld joins th e core barrel cy linder to t h e lower core su pport for g ing/cas ting; no actual flange exists at this location. The primary conc ern with regard to ag ing is through-wall cracking in the weld as a r e sult of stress corrosion.

Actively gr o w ing thro ugh-wall flaws would req u ire 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 larg e inactive flaws ar e likely to be m a nageable even if found

. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 5 (pendi ng) Baffle-for m er Assembly - Baffle-edge Bolts The baffle-edge bolts pro v i d e baffle-plate to baffle-plate attach men t along t h e seam s between the plates and functio n t o prevent gap s form ing between plates.

Structural studies have dem onstr ated that baffle-edge bolts are not required to m a intain the structural integrity of the baffle; therefore, baffle-edge bolts are not considered to be a safety signifi cant co m pone nt. Anal y s is h as shown that differential therm a l expansion an d swelling can cause plastic deform ation of edge bol ts. 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 u lt in baffle-jetting dam a g e to fuel asse m b lie s. In plants with d o wnward coolant flow in t h e region bet w een the baff le and the former, failure o f baffle-edge bolts is consi d ered to directly contribute to baffle jetting.

Evidence of b a ffle-edge bolt failure would t y picall y be observed thr ough broken or m i ssing lo cking devices, protruding bolt he ads, or m is s ing 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 Com p o n ent Category ID: W-ID: 6 (pendi ng) Baffle-for m er Assembly - Baffle-for m er Bolts The baffle-for m er bolts att ach the baffle plates to th e baffle for m er s. Docu m e nte d observations of IASCC cracking of these co m ponents exists in m u ltiple d esi gns in the PWR fleet wor ldwide. These highl y irradiated bol ts perform a critical safety and operation a l function in the plant. Los s of a single b o lt or isolated m u ltiple failures of the baffle-former bolts ar e considered to be manageab le, but a catastrophic or W e sting hou se N o n-Prop r ietary Class 3 4-10 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 clustered loss of m u ltiple bolts at adjacent loca tions could cause a lack of structural stability a nd potentiall y rai se safety and operational concerns.

WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 7 (pendi ng) Baffle-for m er Asse mbly - Asse mbly The baffle-for m er a sse mbl y is made up of vertical plates na m e d "b affles" and horizontal support plates na m e d "for mers." The baff le plates are b o lted to the for m ers by the baffle-for m er bolts, and the for m ers are atta ched to the core barrel inside surface by the barrel-for m er bolts. Som e of t h e baffle plates are also bolted to eac h other at selected corners by edge bolts or brackets. The baffle-former a sse mbly forms the interface bet w een the core and the core barrel. The baffle-for m er assembl y provides support, guidance, and protection for t h e 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 S CC of the integrated assem bly are the key concerns identified which could m a n ifest m u ltiple degradation e ffects, such as interference with fuel assem b lies, obstruction of coolant flow, loose parts g e neration, distortion and misalignment of the core, local te m p erature p eaks, degrada tion of control rod inser tio n paths, and b a ffle jetting. To date, no re levant observations of these effects in the baffle-for m er a ssem b ly as a result of void sw elling or IASCC have been docum ented. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 8 (pendi ng) Alignment and Interfacing Co m ponents - 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 a xation as a result of long-term ser v ice. The direct result of long term stress relaxation is a loss of hol d-down forc e s leading to v ibration and wear in the lower internals.

WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 9 (pendi ng) Therm a l Shield Assem b ly - Therm a l Shield Flexures Therm a l shield flexures are part of the th erm a l sh ield support s y stem. This sy stem provides prim ary attach ment of the thermal s h ield to the core barre l at the lower connection points of the asse m b ly. The therm a l shield flexures pro v ide the lower structural support for the t h erm a l shield and are required to hol d the bottom of the thermal shield con cent ric to the core barrel. Due to the differential deflections between the core barre l and therm a l shield caused by therm a l cy cling resulti ng from plant operation, t h e therm a l shield flexures are potentially susceptible to fatigue. I ndications of wear in the com ponents (bolts, pins, and fasteners) com posing the thermal sh ield-to-core ba rrel atta ch ment sy stem , fa ilure in the w e lds at the base of the flexure, or failure in the weld that att aches the flexure to the therm a l shield are the identified indicators of age-related c oncerns in the ass e m bly.

W e sting hou se N o n-Prop r ietary Class 3 4-11 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 10 (pend ing) 4.4.2 Expansion MRP-227-A Expansion co m ponents are those that we re determ ined to be hi ghl y or m oderately susceptible to the effects of a least one of the critical degradation mechanisms aff ecting reactor internals but which det a iled evaluations showed a degree of tol e rance to those effect

s. Examination of expansion co m ponents is dependent o n the plant-sp ecific pr im ary com ponent inspection ob servations and the expansion ins p ection requir e m e nts in Table 4-6 of M R P-227-A [9] and Table 5-3 of NL-12-0 37 [5]. Table 4-1 lists all of the Ex pansi on com ponents applic able to IP2.

Upper Internals Assem b ly - Upper Core Plate The UCP is perforated to per m it coolant to pass from t h e core into t h e upper plenum defined by the USP and the UCP. The coolant t h en exits through t h e outle t nozzles in the core barrel.

The UCP positions and laterally supp orts the core b y fuel pi ns extending be low the plate. The UCP contacts and preloads the fuel asse mbly springs and, therefore, maintai n s contact of the fuel asse m b lie s with the LCP during reactor operation. Ag ing concerns for this com p onent are cracking from wear and fatigue, which may co m p ro m ise t h e abilit y of t h e UCP to pr operl y align t h e fuel. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 2.1 (pen ding) Lower Internals Assem b ly - Lower Sup port For g ing or Casting The lower support for g ing or casting provides support for the core.

Indian Poi n t has a lower support casting. Cracking resulting in displacement of lowe r support forgi n gs or castings would cause concerns for the operat ional and structural stability of the l o wer reactor internals support a ssem b ly. The lower support casting is welded to and supported by t h e core barrel, which transm its vert ical loads to the vessel throug h the c o re barrel flange. Aging concerns for th is co m ponent are cracking and thermal em brittlement.

WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 2.2 (pen ding) Lower Suppo rt Assembly - Lower Supp ort Colum n Bodies (Cast)

The lower support col u m n s provide the structural lin k between the LCP and the l o wer support structure.

The supports are required to keep the L C P from defo r m ing during plant operation.

The upper sections of the lower support colum n bodies m a y experience neutr on fluences above the industry threshol ds for IASCC. The IP2 lower support col u m n bodies are com posed of forged and cast materi als. The cast co mponents are considered separately because a concern exists that they m a y be m o r e sensitive to therm a l and irradiation effects. Although stresses i n colum n s ar e primarily com p ressive, bending stress es 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 h e intended purpose of t h i s co m ponent, and loss of o n e or m o re may com p rom i se this function.

W e sting hou se N o n-Prop r ietary Class 3 4-12 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 WCAP-17096-NP Com p o n ent Category ID: W-ID: 2.3 (cast) (pending)

Bottom-m o u n ted Instrum e ntation S y ste m - Bottom-m ounted Instrumentation (BMI) Column Bodies The BMI colum n bodies define the path for flux thimbl es to be ins e rted into the fuel assemblies. The plant m u st maintain a required num ber of functio ni n g flux t h im bles for core mapping. Fl ux thim bles are norm a lly withdrawn prior to refueling a nd reinserte d at the end of t h e refueling activities. A key consideration is that the pri m ary pressure boundar y must rem a in intact.

The BMI colum n bodies may be subje c t to fatigue d u e to either th erm a l fatigue or flow-induc ed vibrations. Inabilit y to i n sert flux thim bl es would be noted duri ng r e fueling activities. Once fl ux thim bles are inserted, t h e consequences of failure of the co m ponent to perfor m it s in tended function during the ensuing operating c y cle are ty pically considered to be mini m a l. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 2.4 (pen ding) Core Barrel Assembly - C o re Barrel Ou tlet Nozzle W e lds The core barrel outlet nozzle welds join the core ba rrel outlet n o zzles to the uppe r core barrel cy li nder. The prim a r y concern with regard to aging is thro ugh-wall cra c king in the weld as a result of stress corrosion. Activel y growin g thro ugh-wall flaws woul d require attention an d cou ld potentiall y cause jetting through the core barrel. Howeve r, the core barrel is considered a highl y fl aw-tolerant structure, and relatively larg e inactive flaws are likel y to be m a nageable even if found

. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 3.1 (pen ding) Core Barrel Assembly - U pper and Low e r Core Barr el Cy linder Axial Welds The core barrel axial welds are in place to m a intain th e cy lindrical shape of the core barrel, as the core barrel cy linders originate as flat plates that are then ro lled into c y li nders. The primary concern with regard to aging is thr ough-wall cracking in the weld as a re sult of stress corrosion. Acti vely growing through-wall flaws would req u ire attention and could potentia l ly cause jetting thro ugh t h e core barrel. However, the core barre l is considered a highly fla w-tolerant st r u cture, and relatively large inactive flaws are likely to be m a nageable even if found

. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 4.1 (pen ding) Core Barrel Assembly - B a rrel-for m er Bolts The barrel-form er bolts join the barrel to the form er plates and are v ital to m a intaining t h e oper a tional and structural integrity of the integrated baffle-for m e r-barr e l ass e m bly. The prim a ry concern is bolt cracking due to IASSC and fatigue.

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

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

W e sting hou se N o n-Prop r ietary Class 3 4-13 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 7.1 (pen ding) Lower Suppo rt Assembly - Lower Supp ort Colum n Bolts The lower support col u m n bolts attach the support co lu m n s to the lower core su pport plate. Althoug h the bolts do not directly support the weight of the core, they help m a inta in the flatness and integrity of the lower core su pport plate. Cracking from IASCC or fati gue resulting in displacement of the low e r core support plate would cause concerns for the operationa l and structur al stability of the lower reactor internals support asse m b ly. WCAP-1709 6-NP Com p o n ent Category ID: W-ID: 7.2 (pen ding) 4.4.3 Existing MRP-227-A Existing components are those that were deter m ined to be susceptible to the effects of a least one of the critical degradation m ech anisms aff ecting reactor interna ls, but for which existing plant-specific AMP elements were found to be su fficient to m a nage aging concerns. The MRP-227-A requirem e nt is for IP2 to e n sure 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.

W e sting hou se N o n-Prop r ietary Class 3 5-1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 5 CONCLUSION Indian Poi n t Nuclear Generating Station Unit 2 has d e m onstrated a long-term co mmitment to aging management of reactor internals. The additiona l evaluations and a n aly ses com p leted by the MRP industr y grou p have pro v id ed clarificatio n to the level of inspection qualit y neede d to determ in e the proper exam ination m e thod and frequen c ies. It is the industr y positi on that use of the Aging M a nagem e nt Review (AM R) produced by the LRA methodology, co m b ined wi th any addit ional reactor internals inspections required b y the MRP-227-A i ndustr y tabl es provi ded 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 p lianc e with MRP-227-A require ments as an integrated inspection pr ogram aligned with ASME Section XI 10

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

s. As discussed, th e industr y MRP-227-A gui deli nes 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, com b ine d with the ASME Section XI ISI program inspe c tions, existing IP2 program s, an d use of OERs provide reasonable assura nce that the reactor internals will co ntinue to perform thei r intended functions throug h the period of exte nded operation and are in full com p lian ce wit h IP2 comm it me nts to m a n a ge m a t e rial degradation in reactor internals.

W e sting hou se N o n-Prop r ietary Class 3 6-1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 6 REFERENCES

1. U.S. Nuclear Regulator y Co mm ission NUREG-1800, Revision 2

, "Standard Review Plan for the Review of License Ren e wal Applications for Nucl ear Power Plants," Decemb er 2010. 2. U.S. Nuclear Regulatory Commission NUREG-1930, Vols. 1 and 2 , "Safety Eval uation Report Related to the License Renewal of Indi an Point Nuclear G e nerating Unit Nos. 2 and 3," Novem b er 2009. 3. Enterg y Nuclear Engineering Report, IP-RPT-11-000 36, Rev. 0 , "Indian Poi n t E n erg y Center Reactor V ess el Internals Program

," Oct ober 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 P lan, Septem ber 12, 2013.

5. Enterg y Letter, NL-12-037

, Rev. 0, "License Re new a l Application - Revised Rea c tor Vess el Internals Program and Inspection Plan Co m p liant with MRP-227-A, Indian Poi n t Nuclear Generation Unit Nos. 2 an d 3, Docket Nos. 50-24 7 and 50-2 86, License Nos.

DPR-26 and DPR-64," Februar y 17 , 20 12. 6. Westinghous e Report, WCAP-14577, Rev. 1-A, "License Renewal Evaluation:

Aging Managem e nt for Reactor I n ternals," March 2001.

7. Pressuri zed Water Reacto r Prima ry Water Chemist r y Guidelines, Volumes 1 and 2, Revision

6. EPRI, Palo Alto, CA: 20

07. 1 014 986. 8. Materials Rel iability Progr am: Screening, Categor ization, and Ranking of Reactor Internals Components f o r Westinghouse and Com bus tion E ngi neering PWR Design (MR P-191). EPRI , Palo Alto CA: 200

6. 1 013 234. 9. Materials Reliability Progr a m: Pressur i ze d Water Reac tor Intern als Inspection and Evaluation Gui d eli n es (MRP-227-A). EPRI, Palo Alto , CA: 201 1. 1 022 863. 10. Materials Reliability Progr am: Inspection St andard f o r PWR Inte rnals - 2012 Update (M RP-228, R ev. 1). EPRI, Palo Alto, C A: 2012. 10 2 5147. 11. Westinghous e Report, WCAP-17096-N P, Rev. 2, "Reactor Internals Acceptance Criteria Methodolo g y and Data Requirements,"

December 2009.

12. Ente rg y N u cle a r Ma nagem e nt Ma nua l, E N-DC-2 0 2 , R e v. 6 , "N EI 0 3-0 8 Ma teria ls Ini tia ti ve Process ," N o vem b er 21, 2013. 13. Nuclear Energy Institute G u ideline, NEI 03-08 , Revision 2, "Guideline for the Managem e nt of Materials I ssues," J a nua ry 2010. 14. ASME Boiler and Pressure Vessel Code Sec tion XI, 2001 E d ition with the 2003 Addenda.

15. U.S. Nuclear Regulatory Commission NUREG-1801, Revision 2, V o lum e s 1 and 2, "Generic Aging Lessons Learned (GALL) Repo rt," Decem ber 2010. 16. U.S. Nuclear Regulator y Co mm ission NUREG-1930, Sup p lement 1, "Safety E v aluation Report Related to the License Renewal of Indi an Point Nuclear G e nerating Unit Nos. 2 and 3," August 2 011.

W e sting hou se N o n-Prop r ietary Class 3 A-1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 APPENDIX A COMPONENT INSPEC TION DETAILS Figure A-1 Typical Wes tinghouse Internals

W e sting hou se N o n-Prop r ietary Class 3 A-2 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-2 IP2 Internals

W e sting hou se N o n-Prop r ietary Class 3 A-3 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-3 IP2-CID-0010 Control Ro d Guide Tube Assem bly - Guide Plates (Cards)

W e sting hou se N o n-Prop r ietary Class 3 A-4 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-4 IP2-CID-0020 Control Ro d Guide Tube Assem bly - Lower Flange Welds W e sting hou se N o n-Prop r ietary Class 3 A-5 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-5 IP2-CID-0021 Upper Internals Assembly - Upper Core Plate W e sting hou se N o n-Prop r ietary Class 3 A-6 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-6 IP2-CID-0022 Lower Internals Assembly - Lower Support Forging or Castings W e sting hou se N o n-Prop r ietary Class 3 A-7 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-7 IP2-CID-0023 Lower Support Asse mbly - Lower Support Column Bodies (cast)

W e sting hou se N o n-Prop r ietary Class 3 A-8 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-8 IP2-CID-0024 Bottom-m o unted Inst rum e ntation System

- Bottom-m ounted Instru menta tion (BMI) Column Bodies W e sting hou se N o n-Prop r ietary Class 3 A-9 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-9 IP2-CID-0030 Core Barrel Assembly

- Upper Core Barrel Fla n ge Weld W e sting hou se N o n-Prop r ietary Class 3 A-1 0 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-10 IP2-CID-0031 Core Barrel Assem bly - Core Barrel Outlet No zzle Welds W e sting hou se N o n-Prop r ietary Class 3 A-1 1 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-11 IP2-CID-0040 Core Barrel Asse mbly - Upper and Lower Core Barrel Cylinder Girth Welds W e sting hou se N o n-Prop r ietary Class 3 A-1 2 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-12 IP2-CID-0041 Core Barrel Asse mbly - Upper and Lower Core Barrel Cylinder Axial Welds W e sting hou se N o n-Prop r ietary Class 3 A-1 3 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-13 IP2-CID-0050 Core Barrel Assembly

- Lower Core Barrel Fla nge Weld W e sting hou se N o n-Prop r ietary Class 3 A-1 4 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-14 IP2-CID-0060 Baffle-form er Assem bly - Baffle-edge Bolts W e sting hou se N o n-Prop r ietary Class 3 A-1 5 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-15 IP2-CID-0070 Baffle-form er Assem bly - Baffle-fo rm er Bolts W e sting hou se N o n-Prop r ietary Class 3 A-1 6 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-16 IP2-CID-0071 Core Barrel Assembly

- Barrel-former Bolts W e sting hou se N o n-Prop r ietary Class 3 A-1 7 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-17 IP2-CID-0072 Lower Support Asse mbly - Lower Support Column Bolts

W e sting hou se N o n-Prop r ietary Class 3 A-1 8 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-18 IP2-CID-0080 Baffle-former Assembl y - Assembl y

W e sting hou se N o n-Prop r ietary Class 3 A-1 9 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-19 IP2-CID-0090 Alignm ent and Interfacing Com ponents - Internals Hold-down Spring W e sting hou se N o n-Prop r ietary Class 3 A-2 0 W C A P-1 789 4-N P Sep t em b e r 2014 R e vi si on 0 Figure A-20 IP2-CID-0100 Therm al Shield Assembly - Thermal Shield Flexures