Latest revision as of 21:24, 1 March 2020

Text

Official Exhibit - NYS000313-00-BD01 - Notification of Entergy'S Submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3
	ML12335A466
Person / Time
Site:	Indian Point
Issue date:	07/15/2010
From:	Bessette P, Sutton K; Entergy Nuclear Operations, Morgan, Morgan, Lewis & Bockius, LLP
To:	Lathrop K, Lawrence Mcdade, Richard Wardwell; US Atomic Energy Commission (AEC)
	SECY RAS
References
	RAS 21618, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, RAS E-374
	Download: ML12335A466 (99)
	v • d • e

United States Nuclear Regulatory Commission Official Hearing Exhibit Entergy Nuclear Operations, Inc.

In the Matter of:

(Indian Point Nuclear Generating Units 2 and 3) c.\.t.p.R REGU{.q"" ASLBP #: 07-858-03-LR-BD01 l~~'

Docket #: 05000247 l 05000286

< Exhibit #: NYS000313-00-BD01 Identified: 10/15/2012

~

0

'"" Admitted: 10/15/2012 Withdrawn:

NYS000313

'"~ i ' Submitted: December 22, 2011 y~ 0- Rejected: Stricken:

" ****. .- Other:

A~ ~-31tf

~

I i I \i

"\ ! )' ;

Morgan, Lewis & Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004 Morgan Lewis Tel: 202.739.3000 COUNSELORS AT LAW Fax: 202.739.3001 www:morganlewis,com Kathryn M. Sutton DOCKETED Partner USNRC 202.739.5738 ksutton@morganlewis*.com June 15, 2010 (4:45 p.m.)

PaulM. Bessette OFFICE OF SECRETARY Partner RULEMAKINGS AND 202,739.5796 ADJUDICATIONS STAFF pbessette@morganlewis.com July 15, 2010 Lawrence G. McDade, Chainnan Dr. Richard E. Wardwell Dr. Kaye D. Lathrop Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

__.__________->>~cket~Entfll'gy-lY.ll£~fl'--Qp_erations,-.m...f....(lruJian P9mJ N.!tclear Gener.atigg ______ ._____ ...._.__ _

Units 2 and 3), Docket Nos. SO-247-LR and SO-286-LR RE: Notification of Entergy's Submittal ofthe Reactor Vessel Internals Program for Indian Point Units 2 and 3

Dear Administrative Judges:

Entergy Nuclear Operations, Inc. ("Entergy") is providing this notice to the Atomic Safety and Licensing Board ("Board") and the parties regarding Entergy's submittal'ofthe Reactor Vessel Internals Program for Indian Point Units 2 and 3 to the U.S. Nuclear Regulatory Commission ("NRC") on July 14,2010. See NL-I0-063, Letter from Fred Dacimo, Entergy, to NRC Document Control Desk, "Amendment 9 to License Renewal Application (LRA) - Reactor

. Vessel Internals Program" (July 14, 2010). A copy of NL-l 0-063 is attached for your reference.

Counsel is providing this notification insofar as the Reactor Vessel Internals Program may be relevant and material to admitted contention NYS-25. '

i OAGI0001229_00001

Morgan Lewis Lawrence G. McDade, Chairman COUNSELORS AT LAW Dr. Richard E. W;:rrdwell*

Dr. Kaye D. Lathrop July 15,2010 Page 2 athryn M. Sutton, Esq.

Paul M. Bessette, Esq.

Counsel for Entergy Nuclear Operations,* Inc.

CBM/als Attachment cc: Service List OAGI0001229_00002

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY' AND LICENSING BOARD In the Matter of ) Docket Nos. 50-247-LR and

) 50-286-LR ENTERGY NUCLEAR OPERATIONS, INC. )

)

(Indian Point Nuclear Generating Units 2 and 3) )

~-----------) July 15, 2010 CERTIFICATE OF SERVICE I hereby certify that copies of the letter entitled "Notification of Entergy's Submittal of the Reactor Vessel Internals Program for Indian Point Units 2 and 3," dated July 15, 2010, were served this 15th day of July, 2010 upon the persons listed below, by first class mail and e-mail as shown below.

Administrative Judge Administrative Judge Lawrence G. McDade, Chair Kaye D. Lathrop

. Atomic Safety and Licensing Board Panel Atomic Safety and Licensing Board Panel Mail Stop: T -3 F23 190 Cedar Lane E.

U.S. Nuclear Regulatory Commission Ridgway, CO 81432

-~*-~~*---Washington,DG-20555-0001-*~----*-*------~-(E-mail-;.- kd12@nrc~govj------~-----~~~~---~~--*~~**-*

(E-mail: 19m1@nrc.gov)

Administrative Judge Office of the Secretary*

Richard E. Wardwell Attn: Rulemaking and Adjudications Staff Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Mail Stop: T-3 F23 Washington, D.C. 20555-0001

. U;S. NucJearRegtllatory CotIlll1ission .(E-mail:. hearingdocket@nrc.gov)

Washington, DC 20555-0001 (E-mail: rew@nrc.gov)

Office of Commission Appellate Adjudication Josh Kirstein, Law Clerk U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Panel

. Mail Stop: 0-16G4 . Mail Stop: T-3 F23 Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission (E-mail: ocaamail@nrc.gov) Washington, DC 20555-0001 (E-mail: losh.Kirstein@nrc.gov)

OAGI0001229 00003

Page 2 Sherwin E. Turk, Esq. Greg Spicer, Esq.

Beth N. Mizuno, Esq. Office of the Westchester County Attorney David E. Roth, Esq. 148 Martine Avenue, 6th Floor Brian G. Harris, Esq. White Plains, NY 10601 Andrea Z. Jones, Esq. (E-mail: gssl@westchestergov.com )

Office of the General Counsel Mail Stop:, 0-15 D21 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (E-mail: set@nrc.gov)

(E-mail: bnml@nrc.gov)

(E-mail: david.roth@nrc.gov)

(E-mail: brian.harris@nrc.gov)

(E-mail: andrea.jones@nrc.gov)

Manna Jo Greene Thomas F. Wood, Esq.

Environmental Director Daniel Riesel, Esq.

Hudson River Sloop Clearwater, Inc. Ms. Jessica Steinberg, J.D.

724 Wolcott Avenue Sive, Paget & Riesel, P .C.

Beacon, NY 12508 460 Park Avenue (E-mail: mannajo@c1earwater.org) New York, NY 10022 (E-mail: driesel@sprlaw.com)

(E-mail: jsteinberg@sprlaw.com)

Stephen C. Filler, Board Member John Louis Parker, Esq.

--~----'Hudson-RiverSloop*~learwater,.-Inc-.--~-------Regicinal-Attorney 303 South Broadway, Suite 222 Office of General Counsel, Region 3 Tarrytown, NY 10591 NYS Dept. of Environmental Conservation (E-mail: sfiller@nylawline.com) 21 S. Putt Corners Road New Paltz, New York 12561-1620 (E-mail: jlparker@gw.dec.state.ny.us)

Ross Gould, Member . Michael J. Delaney, V.P. - Energy Hudson River Sloop Clearwater, Inc. New York City Economic Development 10 Park Avenue, #5L Corp.

New York, NY 10016 110 William Street (E-mail: rgouldesq@gmail.com) New York, NY 10038 (E-mail: mdelaney@nycedc.com)

OAGI0001229 00004

Page 3 Phillip Musegaas, Esq. Daniel E. O'Neill, Mayor Deborah Brancato, Esq. James Siermarco, M.S.

Riverkeeper, Inc. Liaison to Indian Point 828 South Broadway Village of Buchanan Tarrytown, NY 10591 Municipal Building (E-mail: phillip@riverkeeper.org) 236 Tate Avenue (E-mail: dbrancato@riverkeeper.org) Buchanan, NY 10511-1298 (E-mail: vob@bestweb.net)

Robert D. Snook, Esq. MylanL. Denerstein, Esq.

Assistant Attorney General Executive Deputy Attorney General, Office of the Attorney General Social Justice State of Connecticut Office of the Attorney General

--- ---- S5-ElmStreet -- -----of-the State of-New-York---

P.O. Box 120 120 Broadway, 25 th Floor Hartford, CT 06141-0120 New York, New York 10271 (E-mail: Robert.Snook@po.state.ct.us) (E-mail: Mylan.Denerstein@oag.state.ny.us )

Andrew M. Cuomo, Esq. - Janice A. Dean Attorney General of the State of New York Office of the Attorney General John J. Sipos, Esq. of the State of New York Charlie Donaldson Esq. Assistant Attorney General Assistants Attorney General 120 Broadway, 26th Floor The Capitol New York, New York 10271

- ----~-----Albany;-NY--1-2224-034_1---------*---------------------(E-mail;.--Janice.-I>ean@oag.state.-ny.us}--.:.------------- -- - -

(E-mail: iohn.sipos@oag.state.ny.us) -

Joan Leary Matthews, Esq.

Senior Attorney for Special Projects Office of the General Counsel New York State Department of Environmental :Conservation 625 Broadway, 14th Floor Albany, NY 12207 (E-mail: jlmatthe@gw.dec.state.ny.us)

OAGI0001229 00005

Page 4

Original and 2 copies provided to the Office of the Secretary.

. ~

Paul M. Bessette, Esq.

Counsel for Entergy Nuclear Operations, Inc.

DB 1/65220145.1 OAGI0001229 00006

Entergy Nuclear Northeast Indian Point Energy Center

--- Entergx 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511~0249 Tel (914) 788-2055 Fred Dacimo Vice President License Renewal NL-10-063 July 14, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Amendment 9 to License Renewal Application (LRA) -

Reactor Vessel Internals Program Indian Point Nuclear Generating Unit Nos. 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 J

REFERENCES:

.1. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application" (NL-07-039)

2. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Boundary Drawings (NL-07-040)

3. Entergy Letter dated April 23, 2007, F. R. Dacimo to Document Control Desk, "License Renewal Application Environmental Report References (N L-07 -041 )

. 4. Entergy Letter dated October 11,2007, F. R, Dacimo to Document Control Desk, "License Renewal Application (LRA)" (NL-07 -124)

5. Entergy Letter November 14, 2007, F. R, Dacimo to Document Control Desk, "Supplement to License Renewal Application (LRA)

Environmental Report References" (NL-07-133)

Dear Sir or Madam:

In the referenced letters, Entergy ~uclear Operations, Inc. applied for renewal of the Indian Point Energy Center operating license. This letter contains Amendment 9 to the License Renewal Application (LRA) regarding the Reactor Vessel Internals Program.

If you have any questions, or require additional information, please contact Mr. Robert Walpole at 914-734-6710.

OAGI0001229 00007

NL-10-063 Docket Nos. 50-247 & 50-286 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on 2 "/~-Q"Olo Sincerely, FRD/dmt

Attachment:

1. Amendment 9 to License Renewal Application -

Reactor Vessel Internals Program cc: Mr. Samuel J. Collins, Regiona:l Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel, Special Counsel Mr. John Boska, NRR Senior Project Manager Ms. Kimberly Green, Project Manager-NRC Resident Inspector's Office Mr. Paul Eddy, New York State Department of Public Service Mr. Francis J. Murray, President and CEO, NYSERDA OAGI0001229 00008

ATTACHMENT 1 TO NL-10-063 Amendment 9 to License Renewal Application -

Reactor Vessel Internals Program ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286 LICENSE NOS. DPR-26 AND DPR-64 OAGI0001229 00009

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 1 of 90 INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 LICENSE RENEWAL APPLICATION (LRA)

AMMENDMENT 9 The LRA is revised as described below. (underline - added, strikethrough - deleted) 2.3.1.2 Reactor Vessel Internals The reactor vessel internals for each unit are described in the reactor coolant system description (Unit 2, Reactor Vessel Internals; Unit 3, Reactor Vessel Internals).

For both units, the lower core support structure, the upper core support structure, and the incore instrumentation support structure are the three major parts of the reactor internals.

Lower Core Support Structure The major member of the reactor vessel internals is the lower core support structure consisting of the following components included in this evaluation.

core baffle/former assembly: bolts core baffle/former assembly: plates core barrel assembly: bolts, screws core barrel assembly: axial flexure plates (thermal shield flexures), flange, ring, shell, thermal shield. lower core barrel flange weld. upper core barrel flange weld core barrel assembly: outlet nozzles lower internals assembly: clevis insert bolt lower internals assembly: clevis insert lower internals assembly: intermediate diffuser plate lower internals assembly: fuel alignment pin lower internals assembly: lower core plate lower internals assembly: lower core support plate column sleeves lower internals assembly: lower core support column bolt lower internals assembly, lower core support column castings: column cap, lower core support lower internals assembly: radial key lower internals assembly: secondary core support (energy absorbing device) specimen guides (not subject to aging management review) specimen plugs (installed in IP2 only; not subject to aging management review)

OAGI0001229 00010

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 2 of 90 The lower core support structure is supported at its upper flange from a ledge in the reactor vessel. Within the core barrel are a core baffle and a lower core plate, both of which are attached to the core barrel wall. The lower core support structure provides passageways for the coolant flow. The lower core plate at the bottom of the core below the baffle plates provides support and orientation for the fuel assemblies. Fuel alignment pins (two for each assembly) are also inserted into this plate. Columns are placed between the lower core plate and core support casting in order to provide stiffness and to transmit the core load to the core support casting. Adequate coolant distribution is obtained through the use of the lower core plate and a diffuser plate.

Upper Core Support Structure The "top hat with deep beam features" upper core support structure consists of the following components included in this evaluation.

upper internals assembly, rod control cluster assembly (RCCA) guide tube assembly: bolts upper internals assembly, RCCA guide tub~ assembly: guide tube (including lower flange weld), guide plates upper internals assembly, RCCA guide tube assembly: support pin upper internals assembly: core plate alignment pin upper internals assembly: head/vessel alignment pin upper internals assembly: hold~down spring upper internals assembly: support column upper internals assembly, mixing devices: support column orifice .base, support column mixer upper internals assembly: upper core plate, fuel alignment pin upper internals assembly: support assembly (including ring), upper support plate upper internals assembly: upper support column bolt The support columns establish the spacing between the upper support assembly and the upper core plate and are fastened at top and bottom to these plates and beams.

The RCCA guide tube assemblies shield and guide the control rod drive shafts and control rods.

They are fastened to the upper support and are guided by pins in the upper core plate for proper orientation and support. Additional guidance for the control rod drive shafts is provided by the control rod shroud tube which is attached to the upper support plate and guide tube.

In-Core Instrumentation Support Structure The in-core instrumentation support structures consist of the following components inch:Jded in this evaluation.

thermocouple conduit flux thimble guide tube bottom mounted instrumentation column OAGI0001229 00011

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 3 of 90 An upper system (thermocouple conduit) is used to convey and support thermocouples penetrating the vessel through the head, and a lower system (flux thimble guide tube) is used to convey and support flux thimbles penetrating the vessel through the bottom.

The upper system utilizes the reactor vessel head penetrations. Instrumentation port columns are slip-connected to in-line columns that are in turn fastened to the upper support plate. These port columns protrude through the head penetrations. The thermocouples are carried through the$e port columns and the upper support plate at positions above their readout locations. The thermocouple conduits are supported from the columns of the upper core support system.

Table 2.3.1-2-IP2 and Table 2.3.1-2-IP3 list the mechanical components subject to aging management review and component intended functions for the reactor vessel internals.

Table 3.1.2-2-IP2 and Table 3.1.2-2-IP3 provide the results of the aging management review for the reactor vessel internals.

OAGI0001229 00012

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 4 of 90 Table 2.3.1-4-IP2 Reactor Vessel Internals Components Subject to Aging Management Review Component Type Intende'd Function

- ,, ',~;

LbJvelCore"Support St~ucture

" , ' , ",J' ':

,:>:<,;',' i" ",' "

Core baffle/former assembly Structural support

'obolts Core baffle/former assembly Structural support opiates Flow distribution Shielding Core barrel assembly Structural support obolts and screws GOFe BaFFel assemBly StFUctuFaI support oaxial flexuFO plates Flew distFiButien oflaHge ,Sllieldin§ ofiR9 osHeij

°tlle FA'la I sl=lield Core barrel assembly Structural SUI2I20rt o axial flexure I2lates (thermal shield flexures}

Core barrel assembly Structural SUI2I20rt o flange Core barrel assembly Structural SUI2I20rt o ring Flow distribution o shell Shielding o thermal shield Core barrel assembly Structural SUI2I20rt o lower core barrel flange weld o ul::mer core barrel flange weld OAGI0001229 00013

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 90 Core barrel assembly Flow distribution ooutlet nozzles Lower internals assembly Structural support oclevis insert bolt

°clevis insert ofuel alignment pin olower core support plate column sleeves olower core support plate column bolt oradial key Lower internals assembly Flow, distribution ointermediate diffuser plate Lower internals assembly Structural support olower core plate Flow distribution olower core support castings ocolumn cap olower core support osecondary core support Upper Core Support Structure-Upper Internals Assembly RGGA §l;liEle tl;lse asseFRsly StFl;lstl;lral Sl;lpport oeett o§l;liEle tl;lse oSl;lpport piR RCCA guide tube assembly Structural support obolt RCCA guide tube assembly Structural support oguide tube (including lower flange welds}

RCCA guide tube assembly Structural support oguide plates RCCA guide tube assembly Structural support osupport pin OAGI0001229 00014

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 6 of 90 Core plate alignment pin Structural support Head / vessel alignment pin Structural support Hold-down spring Structural support Mixing devices Structural support

-support column orifice base Flow distribution

-support column mixer Support column Structural support Upper core plate, fuel alignment Structural support pin Flow distribution Upper support plate, support Structural support assembly (including ring)

Upper support column bolt Structural support Bottom mounted instrumentation Structural support column Flux thimble guide tube Structural support Thermocouple conduit Structural support OAGI0001229 00015

NL-10-063 Attachment 1 .

Docket Nos. 50-247 & 50-286 Page 7 of 90 3.1.2.1.2 Reactor Vessel Internals Materials Reactor vessel internals components are constructed of the following materials.

cast austenitic stainless steel

nickel alloy stainless steel Environment Reactor vessel internals components are exposed to the following environments.

neutron fluence treated borated water treated borated water> 140°F treated borated water> 482°F Aging Effects Requiring Management The following aging effects associated with the reactor vessel internals require management.

change in dimensions

cracking

cracking - fatigue

loss of material

loss of material - wear

loss of preload reduction of fracture toughness Aging Management Programs The following aging management programs manage the aging effects for reactor vessel internals components.

Inservice Inspection

Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS)

Reactor Vessel Internals Water Chemistry Control - Primary and Secondary OAGI0001229 00016

NL-10-063 Attachment 1 DocketNos. 50-247 & 50-286 Page 8 of 90 3.1.2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swelling Loss of fracture toughness due to neutron irradiation embrittlement and change in dimensions (void swelling) oould ocour in stainless steel and nickel alloy reactor vessel internals components exposed to reactor coolant and neutron flux will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. MRP-227. The RVI Program will use nondestructive examinations ruDE) and other inspection methods to manage aging effects for reactor vessel

,internals. To manage .loss of fraoture toughness in vessel internals oomponents, WEG will (1) partioipate in the industry programs for investigating and managing

~ing e#ects on reactor internals; (2) evaluate and implement the results of the iRcJustry programs as applioable to the reaotor internals; and (3) upon oompletion of tAese programs, but not less than 24 months before entering the period of extended eperation, submit an inspeotion plan for reastor internals to the NRG for review and approval. This oommitmont is inoluded in the UF5AR 5upplement, Appendix A, aeotions A.2.1.41 and A.3.1.41.

3.1.2.2.9 Loss of Preload due to Stress Relaxation Loss of preload due to thermal stress relaxation (creep) would onlY,be a concern in very high temperature applications (> 700°F) as stated in the ASME Code,Section II, Part 0, Table 4. No IPEC internals components operate at > 700°F. Therefore, loss of preload due to thermal stress relaxation (creep) is not an applicable aging effect for the reactor vessel internals components. However, irradiation-enhanced creep

{irradiation creep) or irradiation enhanced stress relaxation (lSR) is an athermal Qfocess that depends on the neutron fluence and stress; and, on void swelling if Qfesent. Nevertheless Therefore, loss of preload of stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. to the extent that industry

!eveloped reastor vessel internals aging _management programs address these

~ing e#eots. The IPEG oommitment to these RVI programs is inoluded in UF5AR Supplement, Appendix /\, Sections /\.2.1.41 and /\.3.1.41.

3.1.2.2.15 Changes in Dimensions due to Void Swelling Changes in dimensions due to void swelling Gould oGGlir in stainless steel and nickel alloy reactor internal components exposed to reactor coolant will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. To manage GRanges in dimensions of vessel internals oomponents, IPEG will (1) partioipate in tAe industry programs for investigating and managing aging e#oots on reastor imornals; (2) evaluate and implement the results of tho industry programs as OAGI0001229 00017

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 9 of 90 applicable to the reactor internals; and (3) upon completion of thoso programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval. This commitment is included in the UFSAR Supplement, Appendix A, Sectiens A.2.1.41 and A.3.1.41.

3.1.2.2,17 Cracking due to Stress Corrosion Cracking. Primary Water Stress Corrosion Cracking. and Irradiation-Assisted Stress Corrosion Cracking Cracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), and irradiation-assisted stress corrosion cracking (IASCC) ooti-IG GGGI:H= in PWR stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals (RVI) Program. The RVI Program will implement the EPRI Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. MRP-227. The RVI Program will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. To manage cracking in vessel internals components, IPEC maintains the Water Chemistry Centrol Primary and Secondary Program and will (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicablo to the reactor intornals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval. The IPEC commitment to these RVI programs is included in UFSAR Supplement, Appendix A, Sections A.2.1.41 and A.3.1.41.

OAGI0001229 00018

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 10 of 90 Table 3.1.1 Summary of Aging Management Programs for the Reactor Coolant System Evaluated in Chapter IV of NUREG-1801 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Aging Further Item Aging Effect!

Component Management Evaluation Discussion Number Mechanism Programs Recommended 3.1.1-22 Stainless steel and Loss of FSAR supplement No, but licensee GORsisteRt 1,!,litt:l fl.IYR~G ~ 8Q~. Loss of nickel alloy reactor fracture commitment to (1) commitment to -fracture toughness of stainless steel vessel internals toughness due participate in be confirmed and nickel alloy reactor vessel components exposed to to neutron industry RVI aging internals components will be managed reactor coolant and irradiation programs (2) by the Reactor Vessel Internals neutron flux embrittlement, implement Program. agiRg maRagemoRt -

void swelling applicable results I3FogFams. +t:le sommitmeRt to tt:lese (3) submit for NRC RVll3FogFams is iRsIl:lEleEliR YFSAR approval> 24 Sl:ll3l3leA=leRt, Al3l3eREli* .A., SestioRs months before the A2.~.4~ aREl A.3.~.4~. /

extended period an See Section 3.1.2.2.6.

RVI inspection plan based on industry recommendation ..

o G>

o o

o

~

1<0 o

o o

<0

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 11 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Aging Further Item Aging Effect!

Component Management Evaluation Discussion Number Mechanism Programs Recommended 3.1.1-27 Stainless steel and Loss of FSAR supplement No, but licensee Loss ef ~Felea8 8l:1e te stFess nickel alloy reactor preload due to commitment to (1) commitment to Fela*atieR JSFeef3) is a seRseFR feF vessel internals screws, stress participate in be confirmed a~~lisatieRs at teFR~eFatl:lFeS t:li~t:leF bolts, tie rods, and hold- relaxation industry RVI aging tt:laR tt:lese ef IPeG FeasteF lJessel aR8 down springs programs (2) iRtemals SeFR~eReRts. Tt:lemfem, less implement ef ~Felea8 8l:1e te stFess Fela*atieR applicable results (sFee~) is Ret aR a~~lisal3le a~iR~

(3) submit for NRC effest feF tt:le FeastsF '/essel iRtemals approval> 24 SeFR~eReRts. NelJertt:leless, less of months before the preload of stainless steel and nickel extended period an alloy reactor vessel internals RVI inspection plan components will be managed by the based on industry Reactor Vessel Internals Program.

recommendation. seRsisteRt ',yitt:l iR8l:1stF]! 8elJele~e8 FeasteF ..tessel iRtemals a~iR~

maRa~eFReRt ~Fe~FaFRS Tt:le seFRFRitFReRt te tt:lese RVI ~Fe~Fams is iRsIl:l8e8 iR YFS,o,R Sl:l~~lemeRt,

,o,~~eR8i* A, SestieRs ,0..2.1.41* aR8 A3.1.41.

See Section 3.1.2.2.9.

o G>

o o

o

~

1<0 o

o o

~

o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 12 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Aging Further Item Aging Effectl Component Management Evaluation Discussion Number Mechanism Programs Recommended 3.1.1-30 Stainless steel reactor Cracking due Water Chemistry No, but licensee G9RsisteRt !/-,itR IloHdReG ~ gQ~.

vessel internals to stress and FSAR commitment Cracking _of stainless steel reactor components (e.g., corrosion supplement needs to be vessel internals components will be Upper internals cracking, commitment to (1) confirmed managed by the Water Chemistry assembly, RCCA guide irradiation- participate in Control - Primary and Secondary tube assemblies, assisted stress industry RVI aging Program and either the Reactor Vessel Baffle/former assembly, corrosion programs (2) Internals Program or the Inservice Lower internal cracking' implement Insl2ection Program. 91' etAeF RVI assembly, shroud applicable results a§iR§ maRa§elfleRt J'}F9§Fams. +Re assemblies, Plenum (3) submit for NRC s91f11f1itmeRt t9 tRese 9tReF RVI cover and plenum approval> 24 J'}F9§F8mS is iRsII:IEleEl iR bl~S.A,R cylinder, Upper grid months before the SI:IJ'}J'}lemeRt, AJ'}J'}eREli* A, Sesti9Rs assembly, Control rod* extended period an A2.~.4~ aREl .0..3.1.41.

guide tube (CRGT) RVI inspection plan See Section 3.1.2.2.12.

assembly, Core support based on industry shield assembly, Core recommendation.

barrel assembly, Lower grid assembly, Flow distributor assembly, Thermal shield, Instrumentation support structures) o G>

o o

o

~

1<0 o

o o

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Aging Further Item Aging Effectl Component Management Evaluation Discussion Number Mechanism Programs Recommended 3.1.1-33 Stainless steel and Changes in FSAR supplement No, but licensee GSRsisteRt witR ~61REG ~ gQ~.

nickel alloy reactor dimensions commitment to (1) commitment to Changes In dimensions of stainless vessel internals due to void participate in be confirmed steel and nickel alloy reactor vessel components swelling industry RVI aging internals components will be managed programs (2) by the Reactor Vessel Internals implement Program. RVI agiRg maRagemeRt applicable results I3FsgFams. +Re 6smmitmeRt ts tRese (3) submit for NRC RVI I3FsgFams is iRsIl:lEleEi iR 61pSAR approval> 24 Sl:ll3l3lemeRt, A1313eRElix A, SestisRs months before the A.~.~.4~ aREI,A,.d.~.4~.

extended period an See Section 3.1.2.2.15.

RVI inspection plan based on industry recommendation.

o G>

o o

o

~

1<0 o

o o

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 90 Table 3.1.1 : Reactor Coolant System, NUREG-1801 Vol. 1 Aging Further Item Aging Effect!

Component .Management Evaluation Discussion Number Mechanism Programs Recommended 3.1.1-37 Stainless steel and Cracking due Water Chemistry No, but licensee GSRsisteRt witR NIdR~G ~ gQ~.

nickel alloy reactor to stress and FSAR commitment Cracking _of stainless steel and nickel vessel internals corrosion supplement needs to be alloy reactor vessel internals components (e.g., cracking, commitment to (1) confirmed components will be managed by the Upper internals primary water participate in Water Chemistry Control - Primary assembly, RCCA guide stress industry RVI aging and Secondary Program and either tube assemblies, Lower corrosion programs (2) the Reactor Vessel Internals Program internal assembly, CEA cracking, implement or the Inservice Ins(2ection Program.

shroud assemblies, irradiation- applicable results sy stReF RVI a)iR) FRaRa)eFReRt Core shroud assembly, assisted stress (3) submit for NRC I3FS)FaFRs. +Re SSFRFRitFReRt ts tRese Core support shield corrosion approval> 24 stReF RVI I3FS)FaFRS is iRsl~EleEl iR assembly, Core barrel cracking months before the Id~SAR S~l3l3leFReRt, Al3l3eREli* A, assembly, Lower grid extended period an SestisRs A.2.~.41 aREl A.3.1.41.

assembly, Flow RVI inspection plan See Section 3.1.2.2.17.

distributor assembly) based on industry recommendation.

o G>

o o

o

~

1<0 o

o o

~

w

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Aging Further Item- Aging Effect!

Component Management Evaluation Discussion Number Mechanism Programs Recommended 3.1.1-63 Steel reactor vessel Loss of Inservice Inspection No The Inservice Inspection Program and flange, stainless steel material due to (IWB, IWC, and the Reactor Vessel Internals Program and nickel alloy reactor wear IWD) manages loss of-material due to wear vessel internals of the steel reactor vessel flange and exposed to reactor stainless steel and nickel alloy reactor coolant (e.g., upper and vessel internals components.

lower internals assembly, CEA shroud assembly, core support barrel, upper grid assembly, core support shield assembly, lower grid assembly) o G>

o o

o

~

1<0 o

o o

~

.j>..

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50~286 Page 16 of 90 NOTES FOR TABLES 3.1.2 1 IP2 THROUGH 3.1.24 IP3 Generic Notes A. Consistent with NUREG-1801 item for component, material, environment, aging effect and aging management program.

AMP is consistent with NUREG-1801 AMP.

B. Consistent with NUREG-1801 item for component, material, environment, aging effect and aging management program.

AMP has exceptions to NUREG-1801 AMP.

C. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program. AMP is consistent with NUREG-1801 AMP.

D. Component is different, but consistent with NUREG-1801 item for material, environment, aging effect and aging management program. AMP has exceptions to NUREG-1801 AMP.

E. Consistent with NUREG-1801 material, environment, and aging effect but a different aging management program is credited.

F. Material not in NUREG-1801 for this component.

G. Environment not in NUREG-1801 for this component and material.

H. Aging effect not in NUREG-1801 for this component, material and environment combination.

I. Aging effect in NUREG-1801 for this component, material and environment combination is not applicable.

J. Neither the component nor the material and environment combination is evaluated iii NUREG-1801.

Plant-Specific Notes 101. This component. material. environment and aging effect combination is considered in the Reactor Vessel Internals Program.

As documented in MRP-227, the basis for the RVIProgram, this combination warrants no additional aging management.

NUREG 1801,Section XI.M16 states: "No further aging management revie'.'.' is necessary if the applicant provides a commitment in the FSAR supplemont to (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before eritering the period of extended operation, submit an inspection_plan forreactor internals to the NRG for review and approval." IPEG commitment can be found in o Appendix A (UFSAR supplement) of the license renev,al application.

G>

o o

o

~

1<0 o

o o

~

(J'1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 17 of 90 102. This item is considered a match to NUREG-1801 even though the environments are different because the aging effect of cracking due to fatigue is independent of the environment.

103. These components are subject to cracking due to fatigue as identified in the generic entry in the first line of this table.

104. The One-Time Inspection Program will verify effectiveness of the Water Chemistry Control - Primary and Secondary Program.

105. The original inconel guide tube support pins (split pins) were replaced in both units with X-750 pins. The IP3 X-750 split pins, in service since 1987, were replaced in 2009 with stainless steel pins. The IP2 X-750 pins, installed in 1995, remain in service. Future pin replacements will be based on the pin design, industry experience, manufacturer recommendations and plant specific considerations.

o G>

o o

o

~

1<0 o

o o

~

(j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 18 of 90 Table 3.1.2-2-IP2 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management .1801 Vol. Notes Type. Function Item

. Management Programs 2 Item Reactor vessel Structural Stainless Treated borated Cracking - TLAA- metal IV.B2-31 3.1.1-5 A internals support steel, water fatigue fatigue (R-53) .

components cAss, nickel alloy

~ ~--'0/J+/-,"L-::'-'::' E_

':;"',: .," /; .'

,)~~;:[t"*~*

0 Lower Core Support slfi1~ti.Jie" . .. .

"0. . ~~ '9 .:.";: c.." " "" ""

0" ,

0','- "

Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1- ~A, baffle/former support steel water> 140°F dimensions Internals R\4 (R-126) 33 101 assembly s9A=1FRitFReAt

bolts Cracking Water Chemistry IV.B2-10 3.1.1- ~A, Control - Primary (R-125) 30 .w.t and Secondary Reactor Vessel Internals R\4 s9A=1FRitFReAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 19 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Loss of Reactor Vessel IV.B2-5 3.1.1- £A, preload Internals R\4 (R-129) 27 101 S9A=1A=1itffieAt Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1- £A, water> 140°F fracture Internals R\4 (R-128) 22 101 Neutron fluence . toughness S9 ffi ffi itffi e At Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-1 3.1.1- £ A-;

baffle/former support steel water> 140°F dimensions Internals R\4 (R-124) 33 W4-assembly Flow S9A=1ffiitffi9At

plates distribution Shielding Cracking Water Chemistry IV.B2-2 3.1.1- £ A-;

Control - Primary (R-123) 30 W4-and Secondary Reactor Vessel Internals R\4 S9ffiffiitffieAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-3 3.1.1- £A, water > 140°F fracture Internals R\4 (R-127) 22 101 Neutron fluence toughness S9A=1 ffi itffi eAt o

G>

o o

o

~

1<0 o

o o

~

00

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 20 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item .

Core barrel Structural Stainless Treated borat~d Change in Reactor Vessel IV.B2-4 3.1.1- sG, assembly support steel water> 140°F dimensions Internals R¥J. (R-126) 33 101

bolts_and s9mFAitFA9At screws Cracking Water Chemistry IV.B2-10 . 3.1.1- sA-,

Control - Primary (R-125) 30 W4-and Secondary Reactor Vessel Internals R¥J.

s9mFAitm9At Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-5 3.1.1- sA, preload Internals R¥J. (R-129) 27 101 s9FAFAitFA9At Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1- sA, water> 140°F fracture Internals R¥J. (R-128) 22 101 Neutron fluence toughness s9mFAitFA9At o

G>

o o

o

~

1<0 o

o o

~

<0

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 21 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- f.A, assembly support steel water> 140°F dimensions Internals R¥J. (R-121) 33 101

axial flexure ~ seFRFRitFReAt plates distribution (thermal Shielding Cracking Water Chemistry IV.B2-8 3.1.1- f..6r, Control - Primary (R-120) 30 -U}4.

shield flexures) and Secondary Reactor Vessel Internals R¥J.

sOFRFRitFRent Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-26 3.1.1- ..f.

material- Internals (R-142} 63 wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- f.A, water> 140°F fracture Internals R¥J. (R-122) 22 101 Neutron fluence toughness sornFRitrnent o

G>

o o

o

~

1<0 o

o o

w o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 22 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material' Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- ~A, assembly support steel water> 140°F dimensions Internals R\4 (R-121 ) 33 101

flange ~ S91f11f1 itlfl e At ElistFil:H:lti9A ShielEliAg Cracking Water Chemistry IV.B2-8 3.1.1.- ~A-,

Control - Primary (R-120) 30 +Q4-and Secondary Reactor Vessel Internals R\4 s91f11f1itlfleAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

- Loss of Inservice IV.B2-34 3.1.1- ~

material- Inspection (R-115l 63 wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- ~A, water> 140°F fracture Internals R\4 (R-122) 22 101 Neutron fluence toughness s9FAlflitlfleAt o

G>

o o

o

~

1<0 o

o o

w

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 23 of 90 Table 3.~ .2-2-IP2: Reactor Vessel Internals -

Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- fA, assembly support steel water> 140°F dimensions Internals R¥l- (R-121 ) 33 101

ring Flow S91fllflitlfleAt

shell 'distribution

thermal shield Shielding Cracking Water Chemistry IV.B2-8 3.1.1- fA, Control- Primary (R-120) 30 101 and Secondary Reactor Vessel Internals R¥l-s9FRIflitlfleAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- fA, water> 140°F fracture Internals R¥l- (R-122) 22 101 Neutron fluence toughness s9FRIflitFReAt o

G>

o o

o

~

1<0 o

o o

w

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 24 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- ~

assembly sUI2I2 0rt steel water> 140°F dimensions Internals (R-121) 33 101

lower core barrel flange Cracking Water Chemistry IV.B2-8 3.1.1- J; weld Control....; Primary (R-120) 30

ul2l2er core and Secondary barrel flange Reactor Vessel weld Internals Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- ~

water> 140°F fracture Internals (R-122) 22 101 Neutron fluence toughness o

G>

o o

o

~

1<0 o

o o

w w

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 25 of 90 Table 3.1.2*2*IP2: Reactor Vessel Internals Aging Effect Aging NUREG*

Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Flow Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- £A, assembly distribution steel water> 140°F dimensions Internals RW (R-121 ) 33 101

outlet nozzles s9Ff11f1itlfleRt Cracking Water Chemistry IV.B2-8 3.1.1- £ A-,

Control - Primary (R-120) :30 4.Q.1.

and Secondary Reactor Vessel Internals RW s9Ff11f1itlfleRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o

w

.j>..

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page'26 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management , 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- f.A, assembly support water dimensions Internals RVJ. (R-134) 33 101

clevis insert 68Ff1Ff1itFfl8Rt bolt Cracking Water Chemistry IV.B2-16 3.1.1- £A, Control - Primary (R-133) 37 101 and Secondary Reactor Vessel Internals RVJ.

68Ff1Ff1itFfl8Rt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-14 3.1.1- f.A, preload Internals RVJ. (R-137) 27 101 68FRFflitFR8Rt Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- £A, water fracture Internals RVJ. (R-135) 22 101 Neutron fluence toughness 68FRFflitFfl8Rt o

G>

o o

o

~

1<0 o

o o

w (J1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 27 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management* Programs 2 Item Lower internals Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- sA, assembly support water dimensions Internals RVJ. (R-131) 33 101

clevis insert 69A'!A'!itA'!eRt Cracking Water Chemistry IV.B2-20 3.1.1:.. sA, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals RVJ.

- 69 A'! A'! itA'! e Rt Loss of* Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

,~

Loss of Inservice . IV.B2-26 3.1.1- E material- Inspection (R-142) 63 wear r

)

o G>

o o

o

~

1<0 o

o o

w (j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 28 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Flow Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- £G, assembly distribution steel water> 140°F dimensions Internals RVJ. (R-131) 33 101

intermediate seA'lR'litR'l9At diffuser plate Cracking Water Chemistry IV.B2-20 3.1.1- £G, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals RVJ.

seR'lR'litR'l9At Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o

w

NL-10-063

.Attachment1 Docket Nos. 50-247 & 50-286 Page 29 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item

(

Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- ~A.

assembly support steel water> 140°F dimensions Internals RV.J. (R-134) 33 101

fuel alignment S9A=1A=1itA=leRt pin Cracking Water Chemistry IV.B2-16 3.1.1- ~A.

Control - Primary (R-133) 37 101 and Secondary Reactor Vessel Internals RV.J.

s9A=1A=1itA=leRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- ~A, water> 140°F fracture Internals RV.J. (R-135) 22 101 Neutron fluence toughness s9A=1A=1itA=leRt o

G>

o o

o

~

1<0 o

o o

w 00

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 30 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- ~A, assembly support steel water> 140°F dimensions Internals RVJ. (R-131) 33 101

lower core Flow 69 A'! A'! itA'! e Rt plate distribution Cracking Water Chemistry IV.B2-20 3.1.1- ~A--;-

Control - Primary (R-130) 37 W-t and Secondary R~JI 69A'!A'!itrneRt Inservice InsQection Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary . (RP-24) 83 and Secondary Loss of Inservice IV.B2-26 3.1.1- J; material- InsQection (R-142) 63 wear Treated borated Reduction of Reactor Vessel IV.B2-18 3.1.1- ~A, water> 140°F fracture Internals RVJ. (R-132) 22 101 Neutron fluence toughness 69 A'! A'! itA'! e Rt o

G>

o o

o

~

1<0 o

o o

W

<0

NL-10-063 Attachment 1 DrQcket Nos. 50-247 & 50-286 Page 31 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Structural CASS Treated borated Change in Reactor Vessel IV.B2-23 3.1.1- £A, assembly support water> 482°F dimensions Internals RV+ (R-139) 33 .. 101

lower core Flow S9FRFflitFfleAt support distribution castings Cracking Water Chemistry IV.B2-24 3.1.1- £ A-,

- column cap Control - Primary (R-138) 30 w.t

-lower core and Secondary support Reactor Vessel column Internals RV+

bodies S9 FfI Ffl itFfl e At Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Thermal Aging and IV.B2-21 3.1.1- A water> 482°F fracture Neutron Irradiation (R-140) 80 Neutron fluence toughness Embrittlement of I

Cast Austenitic Stainless Steel (CASS) o G>

o o

o

~

)<0 o

o o

.j>..

o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 32 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect* Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- £A, assembly support steel water> 140°F dimensions Internals R\A (R-134 ) 33 101

lower core s9FRFRitFReAt support plate Cracking ~ Water Chemistry IV.B2-16 3.1.1- £A, column bolt Control - Primary (R-133) 37 W4-and Secondary Reactor Vessel Internals R\A S9FRFRitFReAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-25 3.1.1- £A, preload Internals R\A (R-13E3) 27 101 s9FRFRitFReAt Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- £A, water> 140°F fracture Internals R\A (R-135) 22 101 Neutron fluence toughness s9mFRitmeAt o

G>

o o

o

~

1<0 o

o o.j>o.

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 33 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component. Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-23 3.1.1- -EA-,

assembly support steel water> 140°F dimensions Internals R¥J. (R-139) 33 101

lower core S9A'1A'1itA'leRt support plate column Cracking Water Chemistry IV.B2-24 3.1.1- ~A, sleeves Control - Primary (R-138) 30 101 and Secondary Reactor Vessel Internals R¥J.

s9A'1A'1itA'leRt Loss of . Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Reactor Vessel IV.B2-22 3.1.1- ~A,

- water> 140°F fracture Internals R¥J. (R-141) 22 101 Neutron fluence toughness s9A'1A'1itA'leRt o

G>

o o

o

~

1<0 o

o o

.j>..

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 34 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- ~A, assembly support steel water> 140°F dimensions Internals~ (R-131)* 33 101

radial key S9A=1A=1ibfleAt Cracking Water Chemistry IV.B2-20 3.1.1- ~A, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals~

s9A=1A=1itA=leAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-26 3.1.1- E material- Inspection (R-14?) 63 wear o

G>

o o

o

~

1<0 o

o o

.j>.

w

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 35 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- ~G, assembly support steel water> 140°F dimensions Internals R¥J. (R-131) 33 101

secondary . Flow G9Ff! Ffl itFfl8Rt core support distribution Cracking Water Chemistry IV.B2-20 3.1.1- ~G, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals R¥J.

G9FflFflitFf!eRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and SecondarY o

G>

o o

o

~

1<0 o

o o

.j>..

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 36 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material ' Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item

~Pktr!i~f!£flrRe~rtstruc~~;e,*~JJ~per Internals Assembly: ': 1.;~' ,:.,

,~

',.,,' oCJ\

,1,: :: '

' i; .. ,'(

'

'ii.; .' .: ",

RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-27 3.1.1- gA, tube assembly support steel water> 140°F dimensions Internals R¥J. (R-119) 33 101

bolt s9FRFRitFR8Rt Cracking c Water Chemistry IV.B2-28 3.1.1- gA, Control - Primary (R-118) , 37 101 and Secondary Reactor Vessel Internals R¥J.

s9FRFRitFR8Rt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary, (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1- gG, preload Internals R¥J. (R-114) 27 101 s9FRFRitFR8Rt o

G>

o o

o

~

1<0 o

o o

.j>..

(J1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 37 of 90' Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Functi9n Item Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1- £A, tube assembly support steel water> 140°F dimensions Internals R¥J. (R-117) 33 101

guide tube S9Ff1FRitFfl9Rt (including lower flange Cracking Water Chemistry IV.B2-30 3.1.1- £A, welds) Control - Primary (R-116) 30 W4-and Secondary Reactor Vessel Internals R¥J.

69Ff1FRitFfl9Rt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o

.j>..

(j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 38 of 90 Table 3.1.2*2*IP2: Reactor Vessel Internals r Aging Effect Aging NUREG*

Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item M~nagement Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1- ~

tube assembly SUt;mO rt steel _ water> 140°F dimensions Internals (R-117) 33

guide I2lates Cracking Water Chemistry IV.B2-30 3.1.1- ~

Control - Primary (R-116} 30 and Secondary Reactor Vessel Internals Loss of Water Chemistry IV.B2-32 3.1.1- A I material Control - Primary (RP-24} 83 and Secondary Loss of Reactor Vessel IV.B2-34 3.1.1- £ material- Internals (R-115} 63 wear o

G>

o o

o

~

1<0 o

o o

.j>..

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 39 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals

. Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item RCCAguide Structural Nickel alloy, Treated borated Change in Reactor Vessel IV.B2-27 3.1.1- £A, tube assembly support Stainless water dimensions Internals RVl (R-119) 33 101

support pin steel GOA=lA=litA=leAt Cracking Water Chemistry IV.B2-28 3.1.1- E, 105 Control - Primary (R-118) 37 A, ~G~

and Secondary Reactor Vessel Internals RVl cOA=lA=litA=leAt Loss of Water Chemistry IV.B2-32 .3.1.1- A material .. Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o

.j>..

00

NL-10-063 Attachment 1-Docket Nos. 50-247 & 50-286 Page 40 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core plate Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1- ~A, alignment pin support steel water> 140°F dimensions Internals RVJ. (R-113) 33 101 S9FRFRitFReRt Cracking Water Chemistry IV.B2-40 3.1.1- ~A, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals RVJ.

s9FRFRitFReAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-34 3.1.1- E material- Inspection (R115) 63 wear o

G>

o o

o

~

1<0 o

o o.j>o.

<0

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 41 of 90 Table 3.1.2-2-IP2: ReactorVessellnternals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs* 2 Item Head / vessel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- £G, alignment pin support steel water> 140°F dimensions Internals R\4 (R-107) 33 101 S9Ff1Ff1itFfleAt Cracking Water Chemistry IV.B2-42 3.1.1- £G, Control - Primary (R-106) 30 101

- and Secondary Reactor Vessel Internals R\4

- s9Ff1Ff1itmeAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Inservice IV.B2-34 3.1.1- E material- Inspection (R115) 63 wear o

G>

o o

o

~

1<0 o

o o(J'1 o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 42 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Hold-down Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- £A, spring support steel water> 140°F dimensions Internals RVl- (R-107) 33 101 69Ff1Ff1itFfleAt Cracking Water Chemistry IV.B2-42 3.1.1- £A, Control - Primary (R-106) 30 101 and Secondary Reactor Vessel Internals RVl-69R=1Ff1 itFfl eAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1- £A, preload InternalsRVl- (R-114) 27 ~

69R=1Ff1itFfleRt o

G>

o o

o

~

1<0 o

o o(J'1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 43 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Mixing devices Structural CASS Treated borated Change in Reactor Vessel IV.B2-35 3.1.1- ~G,

support support water> 482°F dimensions Internals R¥J. (R-110) 33 101 column orifice Flow seFRFRitFReAt base distribution

support Cracking Water Chemistry IV.B2-36 3.1.1- ~G, column mixer Control - Primary (R-109) 30 101 and Secondary .,-

Reactor Vessel Internals R¥J.

seFflFRitFReAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Treated borated Reduction of Thermal Aging and IV.B2-37 3.1.1- A water> 482°F fracture Neutron Irradiation (R-111) 80 Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS) o G>

o o

o

~

1<0 o

o o(J'1

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 44 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Support Structural Stainless T reatedborated Change in Reactor Vessel IV.B2-35 3.1.1- gA, column support steel water> 140°F dimensions Internals R\4 (R-110) 33 101 s9FRFRitFR9Rt Cracking Water Chemistry IV.B2-36 3.1.1- gA, Control - Primary (R-109) 30 101 and Secondary Reactor Vessel Internals RVJ.

s9FRFRitFR9Rt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP:-24) 83 and Secondary Upper core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1- gA, plate, fuel support steel water> 140°F dimensions Internals RVJ. (R-117) 33 101 alignment pin Flow s9FRFRitFReRt distribution Cracking Water Chemistry IV.B2-40 3.1.1- gA, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals RVJ.

s91flFRitFReRt o

G>

o o

o

~

1<0 o

o o(J'1 w

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 45 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1a01 Vol. Notes Type Function Item Management Programs 2 Item Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- ~A, plate, support support steel water> 140°F dimensions Internals RVJ. (R-107) 33 101 assembly 69Ff1Ff1itFfleAt (including ring)

Cracking Water Chemistry IV.B2-42 3.1.1- ~A, Control - Primary (R-106) 30 4.Q4.

and Secondary Inservice Inspection RVJ.

69Ff1Ff1itFfleAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o(J1

.j>..

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 46 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1- .!; A, .

column bolt support steel water> 140°F dimensions Internals RVJ. (R-113 33 101 6emmitmeRt Cracking Water Chemistry IV.B2-40 3~ 1.1- -EA, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals RVJ.

, 6emmitmeRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1- .!; A, preload Internals RVJ. (R-114) 27 101

. 6emmitmeRt o

G>

o o

o

~

1<0 o

o o(J1 (J1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 47 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item

, ' ; ,..*:i::--.

"nc~r~I~!f:tfJ:f17ent~tiori ~u~~~rfStr{jctu,:e:}if~~;~; ,.. ' . '

~t;;

..:;_~'f1~';':: ' ....' '.~~/ .

.:1~niE ;, "~:~

Bottom Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- ~A, mounted support steel water> 140°F dimensions Internals RV+ (R-144) 33 101 instrumentation 69R=1R=1itR=l9Rt column Cracking Water Chemistry IV.B2-12 3.1.1- ~A, Control - Primary (R-143) 30 ~

i and Secondary Reactor Vessel Internals RV+

69R=1R=1itR=leRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o(J'1 (j)

NL-10-063*

Attachment 1 Docket Nos. 50-247 & 50-286 Page 48 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals Aging Effect Aging NUREG~

Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Flux thimble Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- -EA, guide tube support steel water> 140°F dimensions . Internals RV\. (R-144) 33 101 S9A=1A=1itA=leAt Cracking Water Chemistry IV.B2-12 3.1.1- fA, Control - Primary (R-143) 30 101 and Secondary Reactor Vessel Internals RV\.

s9A=1A=1itmeAt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Thermocouple Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- EG, conduit support steel water> 140°F dimensions Internals RV\. (R-144) 33 101 S9A=1A=1 itA=l eAt Cracking Water Chemistry IV.B2-12 3.1.1- fG, Control - Primary (R-143) 30 101 and Secondary Reactor Vessel Internals RV\.

s9A=1A=1itmeAt o

G>

o o

o

~

1<0 o

o o(J1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 49 of 90 Table 3.1.2-2-IP2: Reactor Vessel Internals

- Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary o

G>

o o

o

~

1<0 o

o o(J'1 00

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 50 of 90 Table 3.1.2*2*IP3 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2*2*IP3: Reactor Vessel Internals Aging Effect Aging NUREG*

Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Reactor Structural Stainless Treated borated Cracking - TLAA...:. metal IV.B2-31 3.1.1-5 A vessel support steel, water fatigue fatigue (R-53) internals CASS, components nickel alloy

""y~ {',,',

,__ :00;, , , ' / , . " "

L9vv,~r; (fore SUPPOrtSl(L!ct'ure

,;',. "",,~L,"~'

,'." :r"'" ,~ ,.;~~",~n

,.:'S!;~I;

'>.i~~~!

Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1-33 -EA' ,

baffle/former support steel water> 140°F dimensions Internals RVl (R-126) 101 assembly 69FflFflitrneAt

bolts Cracking Water Chemistry IV.B2-1D 3.1.1-30 !;A-,

Control - Primary (R-125) w.t and Secondary Reactor Vessel Internals RVl 69rnFflitrneAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o(J'1

<0

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 51 of 90 Table *3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG~

Component Intended Table 1 Material Environment Requiring Management 1801 Vol:- Notes Type Function Item Management Programs 2 Item Loss of Reactor Vessel IV.B2-5 3.1.1-:27 ~A, preload Internals RVl- (R-129) 101 S9R'lR'litR'leAt Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1-22 ~A, water> 140°F fracture Internals RVl- (R-128) 101 Neutron fluence toughness S9 R'l R'l itR'l e At Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-1 3.1.1-33 ~A--;-

baffle/former support steel water> 140°F dimensions Internals RVl- (R-124 ) -W assembly Flow S9 R'l R'l itR'l 9 At

plates distribution Shielding Cracking Water Chemistry. IV.B2-2 3.1.1-30 ~A--;-

Control - Primary (R-123) -W and Secondary Reactor Vessel Internals RVl-S9R'lR'litR'leAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-3 3.1.1-22 ~A, water> 140° F fracture Internals RVl- (R-127) 101 Neutron fluence toughness s9R'lR'litR'leAt o

G>

o o

o

~

1<0 o

o o

(j) o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 52 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring. Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1-33 £G, assembly support steel water> 140°F dimensions Internals R¥J. (R-126) 101

bolts_and S9 A=! A=! itA=! e Rt screws Cracking Water Chemistry IV.B2-10 3.1.1-30 £A, Control - Primary (R-125) w.t and Secondary Reactor Vessel Internals R¥J.

s9Ff1Ff1itFfleRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-5 3.1.1-27 £A, preload Internals R¥J. (R-129) 101 S9 FfI FfI itFfl e Rt Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1-22 £A, I. water> 140°F fracture Internals R¥J. (R-128) 101 Neutron fluence toughness s9A=1A=1itFfleRt o

G>

o o

o

~

1<0 o

o o

(j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 53 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 J;A, assembly support steel water> 140°F dimensions Internals RVJ. (R-121) 101

axial flexure ~ G9R'1R'1 itR'leAt plates distributi9A

. (thermal ShieldiAg Cracking Water Chemistry IV.B2-8 3.1.1-30 J;A, shield Control - Primary (R-120) W4-flexures)

. and Secondary Reactor Vessel Internals RVJ.

G9R'1R'1itFReAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary . (RP-24) and Secondary Loss of Reactor Vessel IV.B2-26 3.1.1-63 £ material- Internals (R-142) wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 J;A, water> 140°F fracture Internals RVJ. (R-122) 101 Neutron fluence toughness G9R'1R'1 itR'leAt o

G>

o o

o

~

1<0 o

o o

(j)

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 54 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural - Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 gA, assembly support steel water> 140°F dimensions Internals R\4 (R-121) 101

flange ~ S9Ff1Ff1 itFfleAt distributi9A ShieldiAg Cracking Water Chemistry IV.B2-8 3.1.1-30 gA, Control - Primary (R-120) +G+

and Secondary Reactor Vessel Internals R\4

- S9Ff1Ff1 itFfleAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-34 3.1.1-63 g material- Inspection (R-115) wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 -EA-,

water> 140°F fracture Internals R\4 (R-122) 101 Neutron fluence toughness S9Ff1Ff1 itFA.9At o

G>

o o

o

~

1<0 o

o o

(j) w

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 55 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 £A, assembly support steel water> 140°F dimensions Internals R¥J. (R-121) 101

ring Flow S9FRFRitFReRt

shell distribution

thermal Shielding Cracking Water Chemistry IV.B2-8 3.1.1-30 £A, shield Control - Primary (R-120) 101 and Secondary Reactor Vessel Internals R¥J.

s9FRFRitFReRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 £A, water> 140°F fracture Internals R¥J. (R-122) 101 Neutron fluence toughne~s e9Ff1Ff1itFfleRt o

G>

o o

o

~

1<0 o

o o

(j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 56 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG'-

Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 ..s assembly sUQQort steel water> 140°F dimensions Internals (R-121) 101

lower core barrel flange Cracking Water ChemistrY IV.B2-8 3.1.1-30 ..f weld Control - PrimarY (R-120)

uQQer core and SecondarY barrel flange Reactor Vessel weld Internals Loss of Water ChemistrY IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and SecondarY Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 ..s water> 140°F fracture Internals (R-122) 101 Neutron fluence toughness o

G>

o o

o

~

1<0 o

o o

(j)

(J'1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 57 of 90 Table 3.1.2*2*IP3: Reactor Vessel Internals Aging Effect Aging NUREG*

Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core barrel Flow Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 ~A.

assembly distribution steel water> 140°F dimensions Internals -R¥J. (R-121 ) 101

outlet 69Ff1Ff1itFfleRt nozzles Cracking Water Chemistry IV.B2-8 3.1.1-30 ~A-;

Control - Primary (R-120) 4-Q4.

and Secondary Reactor Vessel Internals -R¥J.

69Ff1Ff1 itmeRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

(j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 58 of 90 Table 3.1.2*2*IP3: Reactor Vessel Internals Aging Effect Aging NUREG*

Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-15 3.1.1-33 gA, internals support water dimensions Internals fW.l. (R-134) 101 assembly 69R=lR=litR=leAt

clevis insert bolt Cracking Water Chemistry IV.B2-16 3.1.1-37 gA, Control - Primary (R-133) 101 and Secondary Reactor Vessel Internals fW.l.

. 69Ff1R=litR=leAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-14 3.1.1-27 gA, preload Internals fW.l. (R-137) 101 69Ff1R=litR=leAt Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 gA, water fracture Internals fW.l. (R-135) 101 Neutron fluence toughness 69R=lR=litR=leAt o

G>

o o

o

~

1<0 o

o o

(j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 59 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 _~A, internals support water dimensions Internals R¥l-- (R-131) 101 assembly 69Ff1FRitFReAt

clevis insert Cracking Water Chemistry IV.B2-20 3.1.1-37 ~A, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals R¥l-69FRFRitFR9At Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-26 3.1.1-63 E material- Inspection (R-142) wear o

G>

o o

o

~

1<0 o

o o

(j) 00

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 60 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring .Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower Flow Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 £G, internals distribution steel water> 140°F dimensions Internals R¥J. (R-131) 101 assembly S9ffiffiitffieRt

intermediate diffuser plate Cracking Water Chemistry IV.B2-20 3.1.1-37 £G, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals R¥J.

s9ffiffiitffi9Rt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

(j)

<0

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 61 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management, , Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1-33 f.A, internals support steel water> 140°F dimensions Internals RW (R-134) 101 assembly s9FRFRitFReAt

fuel alignment Cracking Water Chemistry IV.B2-16 3.1.1-37 -EA, pin Control - Primary (R-133) 101 and Secondary Reactor Vessel Internals RW s9R=1FRitFReAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 f.A, water> 140°F .fracture Internals RW (R-135) 101 Neutron fluence toughness s9FRFRitFR9At o

G>

o o

o

~

1<0 o

o o

o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 62 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 £,A, internals support steel water> 140°F dimensions Internals R¥J. (R-131) 101 assembly Flow s9FRFRitmeAt

lower core distribution plate Cracking Water Chemistry IV.B2-20 3.1.1-37 £,A-,

Control - Primary (R-130) W4-and Secondary R}.ll S9FRFRitFR9At Inservice InsQection Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-26 3.1.1-63 ~

material- InsQection (R-142}

wear Treated borated Reduction of Reactor Vessel IV.B2-18 3.1.1-22 £,A, water> 140°F fracture Internals R¥J. (R-132) 101 Neutron fluence toughness s9mFRitFReAt o

G>

o o

o

~

1<0 o

o o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 63 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function" Item Management Programs 2 Item Lower Structural CASS Treated borated Change in Reactor Vessel IV.B2-23 3.1.1-33 ~A, internals support water> 482°F dimensions Internals RVJ. (R-139) 101 assembly Flow S9AUflitm9Rt

lower core distribution support Cracking Water Chemistry IV.B2-24 3.1.1-30 ~A-;

castings Control - Primary (R-138) 4.Q4.

- column cap and Secondary

-lower core Reactor Vessel support Internals RVJ.

column , s9mmitmeRt bodies Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Thermal Aging and IV.B2-21 3.1.1-80 A water> 482°F fracture Neutron Irradiation (R-140)

Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS) o G>

o o

o

~

1<0 o

o o

~

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 64 of 90 Table 3.1.2-2-IP3: ReactorVessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function . Management Programs 2 Item Item Lower Structural Stainless* Treated borated Change in Reactor Vessel IV.B2-15 *3.1.1-33 !;A, internals support steel water> 140°F dimensions Internals R¥J. (R-134) 101 assembly semmitmeRt .

lower core support plate Cracking Water Chemistry IV.B2-16 3.1.1-37 !;A-;

column bolt Control - Primary (R-133) -W and Secondary .

Reactor Vessel Internals R¥J.

semmitmeRt Loss of Water Chemistry IV.B2-32 3.1:1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-25 3.1.1-27 £A, preload Internals R¥J. (R~136) 101 semmitmeRt Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 £A, wate-r> 140°F fracture Internals R¥J. (R-13'5) 101 Neutron fluence toughness semmitmeRt o

G>

o o

o

~

1<0 o

o o

w

NL-10-063 Attachment 1 .

Docket Nos. 50-247 & 50-286 Page 65 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-23 3.1.1-33 £A, internals support steel water> 140°F dimensions Internals R¥J. (R.,139) 101 assembly S9FRFRitFR9Rt

lower core Cracking Water Chemistry IV.B2-24 3.1.1-30 £A, support plate column Control - Primary (R-138) 101*

sleeves and Secondary Reactor Vessel Internals R¥J.

s9FRFRitFR9Rt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Reactor Vessel IV.B2-22 3.1.1-22 £A, water> 140°F fracture Internals R¥J. (R-141) 101 Neutron fluence toughness S9FRFRitFR9Rt o

G>

o o

o

~

1<0 o

o o

.j>..

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 66 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 .

Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 £A, internals support steel water> 140°F dimensions Internals R\4 (R-131) 101 assembly G9FRmitmeAt .

radial key Cracking Water Chemistry IV.B2-20 3.1.1-37 £A, Control - Primary (R-130) 101 and Secondary

, Reactor Vessel Internals R\4 139 FfI FfI itFfl e At Loss of .Water Chemistry _ IV.B2-32 3.1.1-83 A material Control - Primary (RP-24)-

and Secondary Loss of Inservice IV.B2-26 3.1.1-63 E material- InspeCtion (R-142) 0 wear o

G>

o o

o

~

1<0 o

o o

(J'1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 67 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes.

Type Function Item Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 -EG ,

internals support steel . water> 140°F dimensions Internals R¥J. (R-131) 101 assembly Flow S9Ff1 FflitFfl eRt

secondary distribution core support Cracking Water Chemistry IV.B2-20 3.1.1-37 gG, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals R¥J.

seFflffiitffieRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

(j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 68 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-27 3.1.1-33 £A, tube assembly support steel water> 140°F dimensions Internals R¥!- (R-119) 101

bolt commitment Cracking Water Chemistry IV.B2-28 3.1.1-37 £A, Control - Primary (R-118) 101 and Secondary Reactor Vessel Internals R¥!- .

commitment Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1-27 £G, preload InternalsR¥!- (R-114) 101 commitment o

G>

o o

o

~

1<0 o

o o

. NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 69 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol.. Notes Type Function Item Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1-33 .!;A, tube assembly- support steel water> 140°F dimensions Internals R¥l (R-117) 101 I

guide tube G9mmitmeRt (including lower flange Cracking Water Chemistry IV.B2-30 3.1.1-30 .!;A, welds) Control - Primary (R-116) W4-and Secondary Reactor Vessel Internals R¥l G9A=1mitmeRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

00

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286

. Page 70 of 90 Table 3.1.2-2~IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1-33 .£ tube assembly SU(;!(;!ort steel water> 140°F dimensions Internals (R-117)

guide (;!Iates Cracking Water Chemist!Y IV.B2-30 3.1.1-30 .£ Control - Prima!y (R-116) and Seconda!y Reactor Vessel Internals Loss of Water Chemist!Y IV.B2-32 3.1.1-83 A material Control - Prima!y (RP-24) and Seconda!y

.. Loss of Reactor Vessel IV.B2-34 3.1.1-63 £ material- Internals (R-115) wear o

G>

o o

o

~

1<0 o

o o

<0

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286

. Page 71 of 90 Table 3.1.-2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item RCCA guide Structural Nickel alloy, Treated borated Change in Reactor Vessel IV.B2-27 3.1.1-33 ~A, tube assembly support Stainless water dimensions Internals RW (R-119) 101

support pin steel cOffiffiitffieRt Cracking Water Chemistry IV.B2-28 3.1.1-37 ~

Control - Primary (R-118) 105 A, and Secondary W4-ReaCtor Vessel Internals RW cOffiffiitffi9Rt Loss of Water Chemistry IV.B2-3? 3.1.1-83 A

.- material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

00 o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 72 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Core plate Structural Stainless . Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 gA, alignment pin support steel water> 140°F dimensions Internals RVJ. (R-113) 101 I 68FfHflitFfleAt Cracking Water Chemistry IV.B2-40 3.1.1-37 gA, Control - Primary (R-112) 101 and Secondary Reactor Vessel Internals RVJ.

68 FfI FfI itFfl eAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-34 3.1.1-63 E material- Inspection (R115) wear o

G>

o o

o

~

1<0 o

o o

00

NL-10-063 -

Attachment 1 Docket Nos. 50-247 & 50-286 Page 73 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals

,-- Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Head / vessel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 . -EG, alignment pin support steel water> 140°F dimensions Internals R\4 (R-107) 101 69R'1R'1itR'leAt Cracking Water Chemistry IV.B2-42 3.1.1-30 gG, Control - Primary (R-106) 101 and Secondary I Reactor Vessel Internals R\4 -

69R'1FflitFfleAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material -Control - Primary (RP-24) and Secondary Loss of Inservice IV.B2-34 3.1.1-63 E material- Inspection (R115) wear o

G>

o o

o

~

1<0 o

o o

00

~

. NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 74 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Hold~down Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 f.A, spring support steel water> 140°F dimensions Internals R\4 (R-107) 101 G9rnmitmeRt .

Cracking Water Chemistry IV.B2-42 3.1.1-30 f.A, Control -:- Primary (R-106) 101 and Secondary Reactor Vessel Internals R\4 G9rnmitm9Rt Loss of Water Chemistry IV.B2-32 3.1.1-83 A.

material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1-27 f.A-;

preload Internals R\4 (R-114) W4-G9FBmitm9Rt o

G>

o o

o

~

1<0 o

o o

00 w

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 75 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function **Item Management Programs 2 Item Mixing Structural CASS Treated borated Change in Reactor Vessel IV.B2-35 3.1.1-33 ~G, devices support water> 482°F dimensions Internals RVt (R-110) 101

support 'Flow G9A=lA=litA=leRt column distribution orifice base Cracking Water Chemistry IV.B2-36 3.1.1-30 ~G,

support Control - Primary (R-109) 101 column and Secondary mixer Reactor Vessel Internals RVt G9A=lA=l itA=leRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Treated borated Reduction of Thermal Aging and IV.B2-37 3.1,1-80 A water> 482°F fracture Neutron Irradiation (R-111 )

Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS) o G>

o o

o

~

1<0 o

o o

00

.j>.

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 76 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Support StruCtural Stainless Treated borated Change in Reactor Vessel IV.B2-35 3.1.1-33 -EA, column support steel water> 140°F dimensions Internals RVJ. (R-110) 101 S9FflFflitFReAt Cracking Water Chemistry N.B2-36 3.1.1-30 £A, Control - Primary (R-109) 101 and Secondary Reactor Vessel Internals RVJ.

s9FflFflitFR9At Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Upper core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 £A, plate, fuel support steel,- water> 140°F dimensions Internals RVJ. (R-117) 101 alignment pin Flow s9FRFflitFfl9At distribution Cracking Water Chemistry IV.B2-40 3.1.1.,37 £A, Control - Primary . (R-112) 101 and Secondary Reactor Vessel Internals RVJ.

s9FRFflitFfleAt o

G>

o o

o

~

1<0 o

o o

00 (J'1

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 77 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effec.t Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item

. Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 !;A, plate, support support steel water> 140°F dimensions Internals R¥J. (R-107) 101 assembly s9FRFRitFReRt (including ring} Cracking Water Chemistry IV.B2-42 3.1.1-30 !;A--;-

Control - Primary (R-106) W+

and Secondary Inservice Insgection R¥J.

s9FRFRitmeRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

00 (j)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 78 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item

-, Management Programs 2 Item Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 £A, column bolt support steel water> 140°F dimensions Internals R\4 (R-113 101 e9FRFRitFReAt

- Cracking Water Chemistry IV.B2-40 3.1.1-37 £A, Control - Primary (R-112) 101 and Secondary Reactor Vessel Internals R\4 s9FRFRitFReAt Loss of ' Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-38 3.1.1-27 £A, preload Internals R\4 (R-114) 101 s9FRFRitFR9At o

G>

o o

o

~

1<0 o

o o

00

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 79 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Bottom Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1-33 £A, mounted support steel water> 140°F dimensions Internals RVJ. (R-144) 101 instrumentatio G9FRFRitFR9At n column Cracking .Water Chemistry IV.B2-12 3.1.1-'30 £ A-,

Control - Primary . (R-143) W4-and Secondary Reactor Vessel Internals RVJ.

G9FnFRitFR9At Loss of Water Chemistry

IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

00 00

NL-10-063 Attachment 1 DocketNos. 50-247 & 50-286 Page 80 of 90 Table 3.1.2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Managemen~ Programs 2 Item Flux thimble Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1,1-33 £A, guide tube support steel water> 140°F dimensions Internals R¥J. (R-144) 101 6sFRmitmeRt Cracking Water Chemistry IV.B2-12 3.1.1-30 £A, Control - Primary (R-143) 101 and Secondary Reactor Vessel Internals R¥J.

6smmitmeRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Thermocouple Structural Stainless Treated borated Change in Reactor Vessel . IV.B2-11 3.1.1-33 £G, conduit support steel water> 140°F dimensions Internals R¥J. (R-144) 101

'- s8FRFRitFReAt Cracking Water Chemistry IV.B2-12 3.1.1-30 £G, Control - Primary (R-143) 101

. and Secondary Reactor Vessel Internals R¥J.

s8FRmitFReAt o

G>

o o

o

~

1<0 o

o o

00

<0

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 81 of 90 Table 3.1.'2-2-IP3: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Item Management Programs 2 Item Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary o

G>

o o

o

~

1<0 o

o o

<0 o

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 82 of 90 A.2.1.41 Reactor Vesse.1 Internals Aging Management Activities The Reactor Vessel Internals (RVI) Program is a new plant specific program to manage aging effects of reactor vessel internals using the guidance from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals are presented in MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The MRP also developed inspection requirements specific to the inspection methods delineated in MRP-227, as well as requirements for gualification of the nondestructive examination *(NDE).systems used to perform those inspections. These inspection requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."

MRP-227 and MRP-228 provide the basis of the IPEe Reactor Vessel Internals (RVI) Program.

Revisions to MRP-227 and MRP-228, including any changes resulting from the NRe review of the documents (issued as MRP-227-A and MRP-228-A) will be incorporated into the IPEe RVI Program. The RVI Program will monitor the effects of aging degradation mechanisms on the intended function of the internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with MRP-227 inspection re*quirements and evaluation acceptance ci-iteria.

The IPEe RVI Program will be implemented and maintained in accordance with the guidance in NEI 03-08 [Addenda], Addendum A. "ReS Materials Degradation Management Program Guidelines." Any deviations from mandatory, needed, or good practice implementation requirements established in MRP-227 or MRP-228, will be dispositioned in accordance with the NEI 03-08 implementation protocol. The RVI Program will be implemented prior to the period of extended operation. To manage loss of fracture toughness, cracking, change in dimensions fvoid swelling), and loss of preload in vessel internals Gomponents, the site will (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval. . .

OAGI0001229 00091

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 83 of 90 A.3.1.41 Reactor Vessel Internals Aging Management Activities The Reactor Vessel Internals (RVI) Program is a new plant specific program to manage aging effects of reactor vessel internals using the guidance from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals are presented in MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines." The MRP also developed inspection requirements specific to the inspection methods delineated in MRP-227, as well as requirements for gualification of the nondestructive examination (NDE) systems used to perform those*

inspections. These inspection requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."

MRP-227 and MRP-228 proVide the basis of the IPEe Reactor Vessel Internals (RVI) Program.

Revisions to MRP-227 and MRP-228, including any changes resulting from the NRe review of the documents (issued as MRP-227-A and MRP-228-A) will be incorporated into thelPEe RVI Program. The RVI Program will monitor the effects of aging degradation mechanisms on the intended function of the internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with MRP-227 inspection requirements and evaluation acceptance criteria. .

The IPEe RVI Program will be implemented and maintained in accordance with the guidance* in NEI 03-08 [Addenda], Addendum A. "ReS Materials Degradation Management Program Guidelines." Any deviations from mandatory, needed, or good practice implementation requirements established in MRP-227 or MRP-228, will be dispositioned in accordance with the NEI 03-08 implementation protocol. The RVI Program will be implemented prior to the period of extended operation. To manage loss of fracture toughness, cracking, change in dimensions (void swolling), and loss of preload in vossol intornals compononts, tho sito will (1) participate in the industry programs for investigating and managing aging offocts on roactor intornals; (2) evaluate and implement the results of the industry programs as applicable to the reactor intornals; and (3) upon complotion of these programs, but not less than 24 months bofore entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

OAGI0001229 00092

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 84 of 90 Section B.1.42 of the LRA is completely new.

8.1.42 Reactor Vessel Internals Program Program Description TheReactor Vessel Internals Program is a new plant-specific program. Revision 1 of NUREG-1801 includes no aging management program description for PWR reactor vessel internals.

NUREG-1801,Section XI.M16, PWR Vessel Internals, instead defers to the guidance provided in Chapter IV line items as appropriate. The Chapter IV line item guidance recommends actions to:

" ... (1) participate in the industry programs for investigating and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval."

The industry programs for investigating and managing aging effects on reactor internals are part of the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP developed inspection and evaluation (I&E) guidelines for managing the effects of aging on pressurized water reactor vessel internals. These guidelines are presented in MRP-227, "Materials Reliability Program: Pressurized Water Reactor Internals" Inspection and Evaluation Guidelines." The I&E guidelines include:

summary descriptions of PWR internals and functions;

summary of the categorization and aging management strategy development of potentially susceptible locations, based on the safety and economic consequences of aging degradation;

direction for methods, extent, and frequency of one-time, periodic, and conditional examinations and other aging management methodologies;

acceptance criteria for the one-time, periodic, and conditional examinations and other aging management methodologies; and

methods for evaluation of aging effects that exceed the examination acceptance criteria.

The MRP also developed inspection procedure requirements specific to the inspection methods delineated in MRP-227, as well as requirements for qualification of the nondestructive examination (NDE) systems used to perform those inspections. These inspection procedure requirements are presented in MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals."

MRP-227 and MRP-228 provide the basis of the IPEC Reactor Vessel Internals (RVI) Program.

Revisions to MRP-227 and MRP-228, including any changes resulting from the NRC review of the documents (issued as MRP-227-A and MRP-228-A), will be incorporated into the IPEC RVI Program.

OAGI0001229 00093

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 85 of 90 The RVI Program will monitor the effects of aging on the intended function of the internals through periodic and conditional examinations. The RVI Program will detect and evaluate cracking, loss of material, reduction of fracture toughness, loss of preload and dimensional changes of vessel internals components in accordance with'MRP-227 inspection recommendations and evaluation acceptance criteria.

IPEG will implement and maintain the RVI Program in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RGS Materials Degradatiol) Management Program Guidelines,"

Any deviations from mandatory, needed, or good practice implementation activities established in MRP-227 or MRP-228, will be managed in accordance with the NEI 03-08 implementation protocol.

Evaluation

1. Scope of Program MRP-227 guidelines are applicable to reactor internal structural components. The scope does not include consumable items such as ftJel assemblies and reactivity control assemblies which are periodically replaced based on neutron flux exposure. The scope does not include welded attachments to the reactor vessel which are'considered part of the vessel, or nuclear instrumentation (flux thimble tubes) which forms part of the reactor coolant pressure boundary. Other programs manage the effects of aging on these components.

MRP-227 separates PWR internals components into four groups depending on (1) their susceptibility t6 and tolerance of aging effects, and (2) the existence of programs that*

manage the effects of aging. These groupings include:

Primary - those internals components that are highly susceptible to the effects of at least one aging mechanism (identified in Table 4-3 of MRP-227);

. Expansion - those internals components that are highly or moderately susceptible to the effects of at least one aging mechanism, but for which functionality assessment has shown a degree of tolerance to those effects (identified in Table 4-6 of MRP-227);

Existing Programs - those internals components that are susceptible to the effects of at least one aging mechanism and for which generic and plant-specific existing AMP elements are capable of managing those effects (identified in Table 4-9 of MRP-227); and

No Additional Measures - those internals components for which the effects of aging mechanisms are below the MRP-227 screening criteria (internals components not included in Tables 4-3, 4-6 or 4-9 of MRP-227).

The categorization of internals components for Westinghouse PWRs, ~s presented in MRP-227, applies to IPEG Unit 2 and Unit 3 vessel internals. The component inspections identified in MRP-227, Tables 4-3 and 4-6 for primary and expansion group components, define the scope of the IPEG RVI Program inspections. Those components subject to aging -

management by existing programs, as delineated in MRP-227, Table 4-9, are included in OAGI0001229_00094

NL-10~063 Attachment 1 Docket Nos. 50:"247 & 50~286 Page 86 of 90 the scope of those programs, and are not part of the RVI Program inspections.

Components that are not included in Tables 4-3, 4-6 or 4-9 are considered to be within the scope of the program, but require no specific inspections.

2. Preventive Actions The Reactor Vessel Internals Program is a condition monitoring program that does not include preventive actions. However, primary water chemistry is maintained in accordance

with EPRI guidelines by the Water Chemistry Control- Primary and Secondary Program, which minimizes the potential for stress corrosion cracking (SCC) and irradiation assisted stress corrosion cracking (IASCC).

Plant operations also influence aging of the vessel internals. The general assumptions about plant operations used in the development of the MRP-227 guidelines are applicable to the IPEC units. The units are base loaded and implemented low leakage core loading patterns within the first 30 years of operation. IPEC has implemented no design changes to reactor vessel internals beyond those identified in general industry guidance or recommended by Westinghouse.

3. Parameters Monitored or Inspected The RVI Program will monitor the effects of aging on the intended function of the internals through periodic and conditional examinations and other aging management methods, as required. As described in MRP-227, the program contains elements that will monitor and inspect for the parameters that indicate the progress of each of these effects. The program will use NDE techniques to detect loss of material through wear, identify distortion of components, and locate cracks.

Visual examinations (VT-3) will be used to detect wear. Visual examinations (VT-3) will also*

detect distortion or cracking through indications such as gaps or displacement along component joints and broken or damaged bolt locking systems. Direct measurements of spring height will be used to detect distortion of the internals hold down spring. Visual examinations (EVT-1) will be used to detect crack-like surface flaws of components and welds. Volumetric (ultrasonic) examinations will be used to locate cracking of bolting.

(MRP-227, Tables 4-3 and 4-6)

OAGI0001229 00095

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 87 of 90

4. Detection of Aging Effects The RVI Program will detect cracking, loss of material reduction of fracture toughness, loss of preload and dimensional changes (distortion) of vessel internals components in accordance with MRP-227. The NDE systems (Le .. the combinations of equipment.

procedure, and personnel) used to detect these aging effects will be qualified in accordance with MRP-228. The RVI Program will. conduct inspections of primary group components as follows (MRP-227, Table 4-3):

Periodic visual examinations (VT-3) will detect loss of material due to wear from control rod guide tube guide plates and thermal shield flexure plates.

Periodic visual examinations (VT-3) of the baffle former assembly plates and edge bolts will detect symptoms of distortion due to void swelling or cracking from IASCC.

These symptoms include abnormal interactions with fuel assemblies, gaps or displacement along component joints, broken or damaged bolt locking systems, and failed or missing bolts.

Direct measurements of spring height will detect distortion of the internals hold down spring due to a loss of stiffness. Measurementswill be taken periodically, as needed to determine the life of the spring.

Periodic visual examinations (EVT-1) will detect crack-like surface flaws of the control rod guide tube assembly lower flange welds and the upper core barrel to flange weld.

Volumetric (UT) examinations will locate cracking of baffle former bolting. Baseline and subsequent measurements will be used to confirm the stability of the bolting pattern.

Indications from EVT-1 or UT inspections may result in additional inspections of expansion group components, as determined by expansion criteria delineated in MRP-227, Table 5-3.

The relationships between primary group component inspection findings and additional inspections of expansion group components are as follows.

Indications from the EVT-1 inspections of the control rod guide tube assembly lower flange welds may result in* EVT-1 inspections of the lower support column bodies and VT-3 inspections of bottom mounted instrumentation column bodies to detect cracking.

Indications from the EVT-1 inspection of the upper core barrel to flange weld may result in EVT-1 inspections of the remaining core barrel welds

Indications from the UT inspections of baffle former bolting may result in UT inspections of the lower support column bolts and the barrel former bolts for cracking. .

OAGI0001229 00096

NL-10-063.

Attachment 1 Docket Nos. 50-247 & 50-286 Page 88 of 90

5. Monitoring and Trending The RVI program uses the inspection guidelines for PWR internals in MRP-227.

Inspections in accordance with these guidelines will provide timely detection of aging effects. In addition to the inspections of primary group components, expansion group components have been defined should the scope of examinatioA and re-examination need to be expanded beyond the primary group. Records of inspection results are maintained allowing for comparison with subsequent inspection results.

IPEC will share inspection results with the industry in accordance with the good practice recommendations of MRP-227. The IPEC-specific results will be incorporated into an overall industry report that will track industry progress and will aid in evaluation of potentially significant issues, identification of fleet trends, and determination of any needed revisions to MRP-227 guidelines.

6. Acceptance Criteria The RVI Program acceptance criteria are from Section 5 of MRP-227. Table 5-3 of MRP-227 provides the acceptance criteria for inspections of the primary and expansion group components. The criteria for expanding the examinations from the primary group components to include the expansion group components are a/so delineated in MRP-227, Table 5-3. The examination acceptance criteria include: (i) specific, descriptive relevant conditions for the visual (VT-3) examinations; (ii) requirements for recording and dispositioning surface breaking indications that are detected and sized for length by the visual (EVT-1) examinations; and (iii) requirements for system-level assessment of bolted assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits.

7. Corrective Action Conditions adverse to quality; such as failures, malfunctions, deviations, defective material

. and equipment, and nonconformances; are promptly identified and corrected. In the case of significant conditions adverse to quality, measures are implemented to ensure that the cause of the nonconformance is determined and that corrective action is taken to preclude recurrence. In addition, the cause of the significant condition adverse to quality and the corrective action implemented is documented and reported to appropriate levels of management.' The Entergy (10 CFR Part 50, Appendix B) Quality Assurance Program, including relevant corrective action controls, applies to the RVI Program.

Any detected condition that does not satisfy the examination acceptance criteria must be processed through the corrective action program. Example methods for analytical disposition of unacceptable conditions are discussed or referenced in Section 6 of MRP-227. These methods or other demonstrated and verified alternative methods may be used.

The alternative of component repair and replacement of PWR internals is subject to the applicable requirements of the ASME Code Section XI.

OAGI0001229 00097

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 89 of 90

8. Confirmation Process This attribute is discussed in Section B.0.3.

9. Administrative Controls This attribute is discussed in Section B.0.3.

10. Operating Experience Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. However, PWR internals aging degradation has been observed in European PWRs, specifically with regard to cracking of baffle-former bolting. For this reason, the U.S. PWR owners and operators created a program to inspect the baffle-former bolting to determine whether similar aging degradation might be expected to occur in U.S. plants. A benefit of this decision was the experience gained with the UT examination techniques used in the inspections. .

In addition, the industry began laboratory testing projects to gather the materials data necessary to support future inspections and evaluations. Other confirmed or suspected material degradation concerns that the industry has identified for PWR components are.

wear in thimble tubes, potential wear in control rod guide tube guide plates, and cracking in some high-strength bolting. The industry has addressed the last concern primarily through replacement of high-strength bolting with bolt material that is less susceptible to cracking and by improved control of pre-load.

The RVI Program established in accordance with the MRP-227 guidelines is a new program.

Accordingly, there is no direct programmatic history for IPEG. However, program inspections will use qualified techniques similar to those successfully used at IPEG and throughoutthe industry for ASME Section XI Gode inspections. Internals inspections (VT-3) have been conducted at IPEG in accordance withASME Section XI Gode requirements, .

with no indications of component degradation. IPEG has appropriately respo'nded to industry operating experience for reactor vessel internals. For example, guide tube support pins (split pins) have been replaced in both units on the basis of industry experience. As with other U.S. commercial PWR plants, cracking of baffle former bolts is recognized as a potential issue for the IPEG units. As a result, IPEG has monitored industry developments and recommendations regarding these components.

Develop'ment of the MRP-227 guidelines is based upon industry operating experience, research data, and vendor evaluations. Reactor vessel internals aging degradation incidents in both U.S. and foreign plants were considered in the development of the MRP-227 guidelines. As implemented, this program will account for applicable future operating experience during the period of extended operation.

OAGI0001229 00098

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 90 of 90 Conclusion The RVI Program will be effective at managing aging effects since it will incorporate proven monitoring techniques, acceptance criteria, corrective actions, and administrative controls in accordance with MRP-227 and MRP-228 guidelines and current IPEe programs. The RVI Program will provide reasonable assurance that the effects of aging are managed such that applicable components will continue to perform their intended functions consistent with the current licensing basis through the period of extended operation.

OAGI0001229 00099

ML12335A466: Difference between revisions

Latest revision as of 21:24, 1 March 2020

Contents

Text

Dear Administrative Judges:

SUBJECT:

REFERENCES:

Dear Sir or Madam:

Attachment:

Navigation menu

ML12335A466: Difference between revisions

Latest revision as of 21:24, 1 March 2020

Text

Dear Administrative Judges:

SUBJECT:

REFERENCES:

Dear Sir or Madam:

Attachment:

Navigation menu

Search