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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 In the Matter of

Entergy Nuclear Operations, Inc. (Indian Point Nuclear Generating Units 2 and 3)

ASLBP #:07-858-03-LR-BD01 Docket #:05000247 l 05000286 Exhibit #:

Identified:

Admitted: Withdrawn:

Rejected: Stricken: Other: NYS000313-00-BD01 10/15/2012 10/15/2012 NYS000313 Submitted: December 22, 2011 c.\.t.p.R REGU{.q"" < 0 '" '" '" i 0-" .-****. i "\

! )' ; Morgan, Lewis & Bockius LLP 1111 Pennsylvania Avenue, NW Washington, DC 20004 Tel: 202.739.3000 Fax: 202.739.3001 www:morganlewis,com Kathryn M. Sutton Partner 202.739.5738 ksutton@morganlewis*.com PaulM. Bessette Partner 202,739.5796 pbessette@morganlewis.com July 15, 2010 I I 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

\i Morgan Lewis COUNSELORS AT LAW DOCKETED USNRC June 15, 2010 (4:45 p.m.) OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF __ . __________ 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 Lawrence G. McDade, Chairman Dr. Richard E. W;:rrdwell*

Dr. Kaye D. Lathrop July 15,2010 Page 2 CBM/als Attachment cc: Service List Morgan Lewis COUNSELORS AT LAW athryn M. Sutton, Esq. Paul M. Bessette, Esq. Counsel for Entergy Nuclear Operations,*

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

Administrative Judge Richard E. Wardwell Atomic Safety and Licensing Board Panel Mail Stop: T-3 F23 . U;S. NucJearRegtllatory Co tIlll1 ission Washington, DC 20555-0001 (E-mail: rew@nrc.gov)

Office of Commission Appellate Adjudication U.S. Nuclear Regulatory Commission . Mail Stop: 0-16G4 . Washington, DC 20555-0001 (E-mail: ocaamail@nrc.gov)

Office of the Secretary*

Attn: Rulemaking and Adjudications Staff U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 . (E-mail:.

hearingdocket@nrc.gov)

Josh Kirstein, Law Clerk Atomic Safety and Licensing Board Panel Mail Stop: T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 (E-mail: losh.Kirstein@nrc.gov)

OAGI0001229 00003 Page 2 Sherwin E. Turk, Esq. Beth N. Mizuno, Esq. David E. Roth, Esq. Brian G. Harris, Esq. Andrea Z. Jones, Esq. 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 Environmental Director Hudson River Sloop Clearwater, Inc. 724 Wolcott Avenue Beacon, NY 12508 (E-mail: mannajo@c1earwater.org)

Greg Spicer, Esq. Office of the Westchester County Attorney 148 Martine Avenue, 6th Floor White Plains, NY 10601 (E-mail: gssl@westchestergov.com ) Thomas F. Wood, Esq. Daniel Riesel, Esq. Ms. Jessica Steinberg, J.D. Sive, Paget & Riesel, P .C. 460 Park Avenue New York, NY 10022 (E-mail: driesel@sprlaw.com) (E-mail: jsteinberg@sprlaw.com)

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

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 Ross Gould, Member Hudson River Sloop Clearwater, Inc. 10 Park Avenue, #5L New York, NY 10016 (E-mail: rgouldesq@gmail.com)

New Paltz, New York 12561-1620 (E-mail: jlparker@gw.dec.state.ny.us ) . Michael J. Delaney, V.P. -Energy New York City Economic Development Corp. 110 William Street New York, NY 10038 (E-mail: mdelaney@nycedc.com)

OAGI0001229 00004 Page 3 Phillip Musegaas, Esq. Deborah Brancato, Esq. Riverkeeper, Inc. 828 South Broadway Tarrytown, NY 10591 (E-mail: phillip@riverkeeper.org) (E-mail: dbrancato@riverkeeper.org)

Robert D. Snook, Esq. Assistant Attorney General Office of the Attorney General State of Connecticut

S5-ElmStreet P.O. Box 120 Hartford, CT 06141-0120 (E-mail: Robert.Snook@po.state.ct.us ) Daniel E. O'Neill, Mayor James Siermarco, M.S. Liaison to Indian Point Village of Buchanan Municipal Building 236 Tate Avenue Buchanan, NY 10511-1298 (E-mail: vob@bestweb.net)

MylanL. Denerstein, Esq. Executive Deputy Attorney General, Social Justice Office of the Attorney General -------of-the State of-New-York---

120 Broadway, 25 th Floor New York, New York 10271 (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 -

(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

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

SUBJECT:

REFERENCES:

Dear Sir or Madam:

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 .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 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)

In the referenced letters, Entergy 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 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 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:

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

-" ,

",', ",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::m er core barrel flange weld OAGI0001229 00013 Core barrel assembly ooutlet nozzles Lower internals assembly 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 ointermediate diffuser plate Lower internals assembly olower core plate olower core support castings ocolumn cap olower core support osecondary core support Flow distribution Structural support Flow distribution , Structural support Flow distribution NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 90 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 Core plate alignment pin Head / vessel alignment pin Hold-down spring Mixing devices -support column orifice base -support column mixer Support column Upper core plate, fuel alignment pin Upper support plate, support assembly (including ring) Upper support column bolt Bottom mounted instrumentation column Flux thimble guide tube Thermocouple conduit Structural support Structural support Structural support Structural support Flow distribution Structural support Structural support Flow distribution Structural support Structural support Structural support Structural support Structural support NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 6 of 90 OAGI0001229 00015 3.1.2.1.2 Reactor Vessel Internals Materials NL-10-063 Attachment 1 . Docket Nos. 50-247 & 50-286 Page 7 of 90 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 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 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 o >> G> o o o ...... 1<0 o o o <0 Table 3.1.1 NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 10 of 90 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 Item Aging Effect! Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-22 Stainless steel and Loss of FSAR supplement No, but licensee GORsisteRt 1,!,litt:l 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 aREl

/ 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 o Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect! Aging Number Component Mechanism Management Programs 3.1.1-27 Stainless steel and Loss of FSAR supplement nickel alloy reactor preload due to commitment to (1) vessel internals screws, stress participate in bolts, tie rods, and hold-relaxation industry RVI aging down springs programs (2) implement applicable results (3) submit for NRC approval>

24 months before the extended period an RVI inspection plan based on industry recommendation.

Further Evaluation Recommended No, but licensee commitment to be confirmed NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 11 of 90 Discussion Loss ef 8l:1e te stFess Fela*atieR JSFeef3) is a seRseFR feF at tt:laR tt:lese ef IPeG FeasteF lJessel aR8 iRtemals Tt:lemfem, less ef 8l:1e te stFess Fela*atieR is Ret aR effest feF tt:le FeastsF '/essel iRtemals NelJertt:leless, less of preload of stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals Program. seRsisteRt

',yitt:l iR8l:1stF]!

FeasteF .. tessel iRtemals Tt:le seFRFRitFReRt te tt:lese RVI is iRsIl:l8e8 iR YFS,o,R 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 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effectl Aging Number Component Mechanism Management Programs 3.1.1-30 Stainless steel reactor Cracking due Water Chemistry vessel internals to stress and FSAR components (e.g., corrosion supplement Upper internals cracking, commitment to (1) assembly, RCCA guide irradiation-participate in tube assemblies, assisted stress industry RVI aging Baffle/former assembly, corrosion programs (2) Lower internal cracking' implement assembly, shroud applicable results assemblies, Plenum (3) submit for NRC cover and plenum approval>

24 cylinder, Upper grid months before the assembly, Control rod* extended period an guide tube (CRGT) RVI inspection plan assembly, Core support based on industry shield assembly, Core recommendation.

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

Further Evaluation Recommended No, but licensee commitment needs to be confirmed NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 12 of 90 Discussion G9RsisteRt

!/-,itR IloHdReG Cracking _of stainless steel reactor vessel internals components will be managed by the Water Chemistry Control -Primary and Secondary Program and either the Reactor Vessel Internals Program or the Inservice Insl2ection Program. 91' etAeF RVI a§iR§ maRa§elfleRt J'}F9§Fams.

+Re s91f11f1itmeRt t9 tRese 9tReF RVI J'}F9§F8mS is iRsII:IEleEl iR SI:IJ'}J'}lemeRt, AJ'}J'}eREli*

A, Sesti9Rs aREl .0..3.1.41.

See Section 3.1.2.2.12.

o >> G> o o o ...... 1<0 o o o Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effectl Aging Number Component Mechanism Management Programs 3.1.1-33 Stainless steel and Changes in FSAR supplement nickel alloy reactor dimensions commitment to (1) vessel internals due to void participate in components swelling industry RVI aging programs (2) implement applicable results (3) submit for NRC approval>

24 months before the extended period an RVI inspection plan based on industry recommendation.

Further Evaluation Recommended No, but licensee commitment to be confirmed NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 90 Discussion GSRsisteRt witR Changes In dimensions of stainless steel and nickel alloy reactor vessel internals components will be managed by the Reactor Vessel Internals Program. RVI agiRg maRagemeRt I3FsgFams.

+Re 6smmitmeRt ts tRese RVI I3FsgFams is iRsIl:lEleEi iR 61pSAR Sl:ll3l3lemeRt, A1313eRElix A, SestisRs See Section 3.1.2.2.15.

o >> G> o o o ...... 1<0 o o o w Table 3.1.1 : Reactor Coolant System, NUREG-1801 Vol. 1 Item Aging Effect! Aging Number Component Mechanism . Management Programs 3.1.1-37 Stainless steel and Cracking due Water Chemistry nickel alloy reactor to stress and FSAR vessel internals corrosion supplement components (e.g., cracking, commitment to (1) Upper internals primary water participate in assembly, RCCA guide stress industry RVI aging tube assemblies, Lower corrosion programs (2) internal assembly, CEA cracking, implement shroud assemblies, irradiation-applicable results Core shroud assembly, assisted stress (3) submit for NRC Core support shield corrosion approval>

24 assembly, Core barrel cracking months before the assembly, Lower grid extended period an assembly, Flow RVI inspection plan distributor assembly) based on industry recommendation.

Further Evaluation Recommended No, but licensee commitment needs to be confirmed NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 90 Discussion GSRsisteRt witR Cracking _of stainless steel and nickel alloy reactor vessel internals components will be managed by the Water Chemistry Control -Primary and Secondary Program and either the Reactor Vessel Internals Program or the Inservice Ins(2ection Program. sy stReF RVI a*)iR*) FRaRa*)eFReRt I3FS*)FaFRs.

+Re SSFRFRitFReRt ts tRese stReF RVI I3FS*)FaFRS is iR Al3l3eREli*

A, SestisRs aREl A.3.1.41.

See Section 3.1.2.2.17.

o >> G> o o o ...... 1<0 o o o .j>.. Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Item-Aging Effect! Aging Number Component Mechanism Management Programs 3.1.1-63 Steel reactor vessel Loss of Inservice Inspection flange, stainless steel material due to (IWB, IWC, and and nickel alloy reactor wear IWD) vessel internals exposed to reactor coolant (e.g., upper and lower internals assembly, CEA shroud assembly, core support barrel, upper grid assembly, core support shield assembly, lower grid assembly)

Further Evaluation Recommended No NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 90 Discussion The Inservice Inspection Program and the Reactor Vessel Internals Program manages loss of-material due to wear of the steel reactor vessel flange and stainless steel and nickel alloy reactor vessel internals components.

o >> G> o o o ...... 1<0 o o o (J'1 NOTES FOR TABLES 3.1.2 1 IP2 THROUGH 3.1.24 IP3 Generic Notes NL-10-063 Attachment 1 Docket Nos. 50-247 &

Page 16 of 90 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 Appendix A (UFSAR supplement) of the license renev,al application.

o >> G> o o o ...... 1<0 o o o (j) 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 ...... Table 3.1.2-2-IP2 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Aging Effect Aging Material Environment Requiring Management Type. Function . Management Programs Reactor vessel Structural Stainless Treated borated Cracking -TLAA-metal internals support steel, water fatigue fatigue components cAss, nickel alloy E_ ':;"',: .," /; Lower Core Support . .. " "0. . '9 .:.";:

c .." " "" "" 0" , " '" , 0','-Core Structural Stainless Treated borated Change in Reactor Vessel baffle/former support steel water> 140°F dimensions Internals R\4 assembly s9A=1FRitFReAt

bolts Cracking Water Chemistry Control -Primary and Secondary Reactor Vessel Internals R\4 s9A=1FRitFReAt Loss of Water Chemistry material Control -Primary and Secondary NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 18 of 90 NUREG-Table 1 .1801 Vol. Notes 2 Item Item IV.B2-31 3.1.1-5 A (R-53) .. .'

0 " . IV.B2-4 3.1.1-(R-126) 33 101 IV.B2-10 3.1.1-(R-125) 30 .w.t IV.B2-32 3.1.1-A (RP-24) 83 o >> G> o o o ...... 1<0 o o o 00 Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Treated borated water> 140°F Neutron fluence . Core Structural Stainless Treated borated baffle/former support steel water> 140°F assembly Flow

plates distribution Shielding Treated borated water > 140°F Neutron fluence Aging Effect Requiring Management Loss of preload Reduction of fracture toughness Change in dimensions Cracking Loss of material Reduction of fracture toughness Aging Management Programs Reactor Vessel Internals R\4 S9A=1A=1itffieAt Reactor Vessel Internals R\4 S9 ffi ffi itffi e At Reactor Vessel Internals R\4 S9A=1ffiitffi9At Water Chemistry Control -Primary and Secondary Reactor Vessel Internals R\4 S9ffiffiitffieAt Water Chemistry Control -Primary and Secondary Reactor Vessel Internals R\4 S9A=1 ffi itffi eAt NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 19 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-5 3.1.1-£A, (R-129) 27 101 IV.B2-6 3.1.1-£A, (R-128) 22 101 IV.B2-1 3.1.1-£ A-; (R-124) 33 W4-IV.B2-2 3.1.1-£ A-; (R-123) 30 W4-IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-3 3.1.1-£A, (R-127) 22 101 o >> G> o o o ...... 1<0 o o o <0 Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Core barrel Structural Stainless Treated assembly support steel water> 140°F

bolts_and screws -Treated borated water> 140°F Neutron fluence Aging Effect Aging Requiring Management Management Programs Change in Reactor Vessel dimensions Internals R¥J. s9mFAitFA9At Cracking Water Chemistry Control -Primary and Secondary Reactor Vessel Internals R¥J. s9mFAitm9At Loss of Water Chemistry material Control -Primary and Secondary Loss of Reactor Vessel preload Internals R¥J. s9FAFAitFA9At Reduction of Reactor Vessel fracture Internals R¥J. toughness s9mFAitFA9At NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 20 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item . IV.B2-4 3.1.1-sG, (R-126) 33 101 IV.B2-10 . 3.1.1-sA-, (R-125) 30 W4-IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-5 3.1.1-sA, (R-129) 27 101 IV.B2-6 3.1.1-sA, (R-128) 22 101 o >> G> o o o ...... 1<0 o o o w o Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Core barrel Structural Stainless Treated borated assembly support steel water> 140°F

axial flexure plates distribution (thermal Shielding shield flexures)

Treated borated water> 140°F Neutron fluence Aging Effect Requiring Management Change in dimensions Cracking Loss of material Loss of material-wear Reduction of fracture toughness Aging Management Programs Reactor Vessel Internals R¥J. seFRFRitFReAt Water Chemistry Control -Primary and Secondary Reactor Vessel Internals R¥J. sOFRFRitFRent Water Chemistry Control -Primary and Secondary Reactor Vessel Internals Reactor Vessel Internals R¥J. sornFRitrnent NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 21 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-7 3.1.1-f.A, (R-121) 33 101 IV.B2-8 3.1.1-f..6r, (R-120) 30 -U}4. IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-26 3.1.1-..f. (R-142} 63 IV.B2-9 3.1.1-f.A, (R-122) 22 101 o >> G> o o o ...... 1<0 o o o w Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material' Environment Core barrel Structural Stainless Treated borated assembly support steel water> 140°F

flange ElistFil:H:lti9A ShielEliAg

-Treated borated water> 140°F Neutron fluence Aging Effect Requiring Management Change in dimensions Cracking Loss of material Loss of material-wear Reduction of fracture toughness Aging Management Programs Reactor Vessel Internals R\4 S91f11f1 itlfl e At Water Chemistry Control -Primary and Secondary Reactor Vessel Internals R\4 s91f11f1itlfleAt Water Chemistry Control -Primary and Secondary Inservice Inspection Reactor Vessel Internals R\4 s9FAlflitlfleAt NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 22 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-7 3.1.1-(R-121 ) 33 101 IV.B2-8 3.1.1.-(R-120) 30 +Q4-IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-34 3.1.1-(R-115l 63 IV.B2-9 3.1.1-(R-122) 22 101 o >> G> o o o ...... 1<0 o o o w Table .2-2-IP2:

Reactor Vessel Internals

-Component Intended Type Function Material Environment Core barrel Structural Stainless Treated borated assembly support steel water> 140°F

ring Flow

shell 'distribution

thermal shield Shielding Treated borated water> 140°F Neutron fluence Aging Effect Requiring Management Change in dimensions Cracking Loss of material Reduction of fracture toughness Aging Management Programs Reactor Vessel Internals R¥l-S91fllflitlfleAt Water Chemistry Control-Primary and Secondary Reactor Vessel Internals R¥l-s9FRIflitlfleAt Water Chemistry Control -Primary and Secondary Reactor Vessel Internals R¥l-s9FRIflitFReAt NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 23 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-7 3.1.1-fA, (R-121 ) 33 101 IV.B2-8 3.1.1-fA, (R-120) 30 101 IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-9 3.1.1-fA, (R-122) 22 101 o >> G> o o o ...... 1<0 o o o w w Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Core barrel Structural Stainless Treated borated assembly sUI2I2 0rt steel water> 140°F

lower core barrel flange weld

ul2l2er core barrel flange weld Treated borated water> 140°F Neutron fluence Aging Effect Requiring Management Change in dimensions Cracking Loss of material Reduction of fracture toughness Aging Management Programs Reactor Vessel Internals Water Chemistry Control....;

Primary and Secondary Reactor Vessel Internals Water Chemistry Control -Primary and Secondary Reactor Vessel Internals NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 24 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-7 3.1.1-(R-121) 33 101 IV.B2-8 3.1.1-J; (R-120) 30 IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-9 3.1.1-(R-122) 22 101 o >> G> o o o ...... 1<0 o o o w .j>.. Table 3.1.2*2*IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Core barrel Flow Stainless Treated borated assembly distribution steel water> 140°F

outlet nozzles Aging Effect Requiring Management Change in dimensions Cracking Loss of material Aging Management Programs Reactor Vessel Internals RW s9Ff11f1itlfleRt Water Chemistry Control -Primary and Secondary Reactor Vessel Internals RW s9Ff11f1itlfleRt Water Chemistry Control -Primary and Secondary NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 25 of 90 NUREG* Table 1 1801 Vol. Item Notes 2 Item IV.B2-7 3.1.1-£A, (R-121 ) 33 101 IV.B2-8 3.1.1-£ A-, (R-120) :30 4.Q.1. IV.B2-32 3.1.1-A (RP-24) 83 o >> G> o o o ...... 1<0 o o o w (J1 Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Lower internals Structural Nickel alloy Treated borated assembly support water

clevis insert bolt Treated borated water Neutron fluence Aging Effect Aging Requiring Management Management Programs Change in Reactor Vessel dimensions Internals RVJ. 68Ff1Ff1itFfl8Rt Cracking Water Chemistry Control -Primary and Secondary Reactor Vessel Internals RVJ. 68Ff1Ff1itFfl8Rt Loss of Water Chemistry material Control -Primary and Secondary Loss of Reactor Vessel preload Internals RVJ. 68FRFflitFR8Rt Reduction of Reactor Vessel fracture Internals RVJ. toughness 68FRFflitFfl8Rt , NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page'26 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-15 3.1.1-f.A, (R-134) 33 101 IV.B2-16 3.1.1-£A, (R-133) 37 101 IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-14 3.1.1-f.A, (R-137) 27 101 IV.B2-17 3.1.1-£A, (R-135) 22 101 o >> G> o o o ..... 1<0 o o o w (j) Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Lower internals Structural Nickel alloy Treated borated assembly support water

clevis insert Aging Effect Requiring Management*

Change in dimensions Cracking -Loss of* material Loss of material-wear ) Aging Management Programs Reactor Vessel Internals RVJ. 69A'!A'!itA'!eRt Water Chemistry Control -Primary and Secondary Reactor Vessel Internals RVJ. 69 A'! A'! itA'! e Rt Water Chemistry Control -Primary and Secondary Inservice . Inspection NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 27 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-19 3.1.1-sA, (R-131) 33 101 IV.B2-20 3.1.1:.. sA, (R-130) 37 101 IV.B2-32 3.1.1-A (RP-24) 83 IV.B2-26 3.1.1-E (R-142) 63 r o >> G> o o o ...... 1<0 o o o w ...... Table 3.1.2-2-IP2:

Reactor Vessel Internals Component Intended Type Function Material Environment Lower internals Flow Stainless Treated borated assembly distribution steel water> 140°F

intermediate diffuser plate Aging Effect Requiring Management Change in dimensions Cracking Loss of material Aging Management Programs Reactor Vessel Internals RVJ. seA'lR'litR'l9At Water Chemistry Control -Primary and Secondary Reactor Vessel Internals RVJ. seR'lR'litR'l9At Water Chemistry Control -Primary and Secondary NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 28 of 90 NUREG-Table 1 1801 Vol. Item Notes 2 Item IV.B2-19 3.1.1-£G, (R-131) 33 101 IV.B2-20 3.1.1-£G, (R-130) 37 101 IV.B2-32 3.1.1-A (RP-24) 83 o >> G> o o o ...... 1<0 o o o w 00 Table 3.1.2-2-IP2: