NL-10-063, Amendment 9 to License Renewal Application, Reactor Vessel Internals Program

Amendment 9 to License Renewal Application, Reactor Vessel Internals Program

ML102010102
Person / Time
Site:	Indian Point
Issue date:	07/14/2010
From:	Dacimo F Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-10-063
Download: ML102010102 (93)

Text

Enter-gy Nuclear Northeast Indian Point Energy Center cEntergy 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 788-2055 Fred Dacimno 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

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 (NL-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 Nuclear 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.

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 Sincerely, FRD/dmt

Attachment:

1. Amendment 9 to License Renewal Application -

Reactor Vessel Internals Program cc: Mr. Samuel J. Collins, Regional 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

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

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)

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 tube 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 InstrumentationSupport Structure The in-core instrumentation support structures consist of the following components included in this evaluation.

thermocouple conduit flux thimble guide tube bottom mounted instrumentation column

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 these 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 2.3.1-2-1P2 and Table 2.3.1-2-1P3 list the mechanical components subject to aging management review and component intended functions for the reactor vessel internals.

Table 3.1.2-2-1P2 and Table 3.1.2-2-1P3 provide the results of the aging management review for the reactor vessel internals.

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 4 of 90 Table 2.3.1-4-1P2 Reactor Vessel Internals Components Subject to Aging Management Review Component Type Intended Function Lower Core Support Structur Core baffle/former assembly Structural support

-bolts Core baffle/former assembly Structural support

-plates Flow distribution Shielding Core barrel assembly Structural support

-bolts and screws Core barrel assembly Structural support

-axial floxuropao Flew dictribution 4 lange Core barrel assembly Structural support

- axial flexure Plates (thermal shield flexures)

Core barrel assembly Structural support

• flanme Core barrel assembly Structural support

• ring Flow distribution "shell Shielding

• thermal shield Core barrel assembly Structural support

• lower core barrel flange weld

• upper core barrel flanqe weld

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 5 of 90 Core barrel assembly Flow distribution

• outlet nozzles Lower internals assembly Structural support

'clevis insert bolt

'clevis insert

-fuel alignment pin

-lower core support plate column sleeves

-lower core support plate column bolt

-radial key Lower internals assembly Flow distribution

-intermediate diffuser plate Lower internals assembly Structural support

-lower core plate Flow distribution

-lower core support castings

-column cap

-lower core support

-secondary core support Upper Core Support Structure-UpperInternalsAssembly RCCA guide tube assembly Structural support

-bolt RCCA guide tube assembly Structural support

• quide tube (including lower flange welds)

RCCA guide tube assembly Structural support

'guide plates RCCA guide tube assembly Structural support

-support pin

NL-1 0-063 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 IncreInstrumentationSuort Structure Bottom mounted instrumentation Structural support column Flux thimble guide tube Structural support Thermocouple conduit Structural support

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 > 4820 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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 8 of 90 3.1 .2.2.6 Loss of Fracture Toughness due to Neutron Irradiation Embrittlement and Void Swellinq Loss of fracture toughness due to neutron irradiation embrittlement and change in dimensions (void swelling) Go'ud oeeur 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 (NDE) and other inspection methods to manage aging effects for reactor vessel internals. To manage lo66 Of fracture toughness in ve...l internals .. om.poncnts,

'PEC will (1) pa.ticipat in the indust.' programs"- for ine;tigating and managing aging effects On raet i (2) evaluate and implement the results of the Mindustr'; programns as applicable to the rcactor internals; and (3)upon comFpletion of those Pr*oFams, but noRt I6e thaRn 24 mon9ths bhfore enteFrng the per~iod Of Wotnfdod opeFation, submit an inspe*ionR plan for reaGtO* *n*tr* als to the NRC for* erviw and approal. ThiS comm;;itmet* i includI in the iFSAR Supplement, Appendix A, Section)s A.2.4.4 and A.3.1.4 1.

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 (> 7000 F) as stated in the ASME Code,Section II, Part D, Table 4. No IPEC internals components operate at > 7000 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 (ISR) is an athermal process that depends on the neutron fluence and stress; and, on void swelling if present. Ne-ertheless 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 Proaram will use nondestructive examinations (NDE) and other inspection methods to manage aging effects for reactor vessel internals. to the extent that industr+Y developed roactorF Yessel internals aging management programs address these aging e#fects. The IPEC commitment to these RVI programsR is-iclde , in UFSAR Supplement, Appendix A, SeGtions A.2.1 .41 and A.3.14 .1 3.1.2.2.15 Changes in Dimensions due to Void Swelling Changes in dimensions due to void swelling Gc'-!d occUr 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-maaage changes in dimensionsof vessel git*Ra*,l GWm*pnents, IPEr will (1) paticipate in the i.dust. ' programs for ietgn and managing agig . O reactor

.efectS inRternals:; (2)evaluate and ipentthe results of the industry programns as

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 9 of 90 applicablo to the rcactOr internals; and (3) upon completion of theso programAS, but not less than 24 months bofo-ro onterFing the poriod of extended oporation, submit an inspeG*tin plan for reaGtr i*nRter**l* to the NRl for ro,;w and approa.*l This.

coFmmitmnenti included in the UFSAR Supplement, Appendix A, Sections A.2.1 .1 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) Geuld eGGuF 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 intemrn's Gomponents, 'PEC maiRtaiRs the Wator Chemist,', Cnrol, Primar, and Seondar' PFogram and will (1)participato inthe industry programs foinestgain and managing aging offects on reactor internals; (2)evaluate and implement tho rosults of the industrY programsl as applicablo to the reactor intFRnals; and (3) upon com:pletion of these proganram, but noet less than 24 months bofore entering tho poriod of extended operation, submit anrnspection plan for reactor interna*ls t th NIlC for review and approval. The IPEC cmm-itmRnt to these RVI prorams ;s,,iludod InlFSAR Supploment, App^ndix A, Secti*os A.2.1.1.1 an A.3.1.4 1.

NL-1 0-063 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 Effectl Aging Further Item Component Mechanism Management Evaluation Discussion Number Programs Recommended 3.1.1-22 Stainless steel and Loss of FSAR supplement No, but licensee ConsistOnt With N.UR EG 1801. 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. agi'g manag.m.n.

void swelling applicable results programs. Tho* c.mmitn..A,+t to, thos, (3) submit for NRC R'.' program. is incudod. in UFSAR approval > 24 Supp.oMont, Appendix A, Sect'ion months before the A.2.1.41 and A.3.1.41.

extended period an See Section 3.1.2.2.6.

RVI inspection plan based on industry recommendation.

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 Effectl Aging Further Item Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-27 Stainless steel and Loss of FSAR supplement No, but licensee Loss of prolad duo to stross nickel alloy reactor preload due to commitment to (1) commitment to relaxation (croop) is a con.... for vessel internals screws, stress participate in be confirmed applications*, at temperaturFe higho, bolts, tie rods, and hold- relaxation industry RVI aging than thoso of 'PEC ractOr '-ossol and down springs programs (2) MintoralS c.ompon.nts. ThoroForo, los implement of pel.*ad duo to stross rolaxation applicable results (croop) is not an applicable aging (3) submit for NRC ff ot for the *oact.. r Voss...l in;ton.a.S approval > 24 coponmients. l,-oss of

,ev.orFthelo-,

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. cAAGonsistet With inutydeveloped react*o Wess1 internals aging management programs The commitment to these RVI pogramsi incldted in FSAR Supp12emet, Appendix Setin A2 1 4 and See Section 3.1.2.2.9.

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 Effectl Aging Further Item Mechanism Management Evaluation Discussion Number Component Programs Recommended 3.1.1-30 Stainless steel reactor Cracking due Water Chemistry No, but licensee COncistont with .NUREG 1801.

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 Inspection Progqram. by etheF-RV4 assembly, shroud applicable results aging manag.m.nt program..... Th assemblies, Plenum (3) submit for NRC commitmont to thoso othor RVI cover and plenum approval > 24 progr.m. is ;inudod in UFSAR cylinder, Upper grid months before the Supple. mon, Appendix A, Soeti*ns assembly, Control rod extended period an A.2.1.41 ard A.3.1.4 !.

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)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 13 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1 801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-33 Stainless steel and Changes in FSAR supplement No, but licensee Consistent'with NUREG 1801.

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 aging managemnt applicable results programs. Tho  ;÷mmrm÷ to tho,'

(3) submit for NRC RVI progra.m. i;- incudh in UFSAR approval > 24 Supplement, Appendix A, Sections months before the A.2.1.'41 -;ndA.3.4.4 1.

extended period an See Section 3.1.2.2.15.

RVI inspection plan based on industry recommendation.

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 14 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1 801 Vol. 1 Item Aging Effect/ Aging Further Number Component Mechanism Management Evaluation Discussion Programs Recommended 3.1.1-37 Stainless steel and Cracking due Water Chemistry No, but licensee Ct-,, nt,

,th i;A NU RE*G 1801 .

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 Inspection Program.

shroud assemblies, irradiation- applicable results by other RVI aging managG*.ent Core shroud assembly, assisted stress (3) submit for NRC programs. The cmien to thoso, Core support shield corrosion approval > 24 other RVI programs is include in assembly, Core barrel cracking months before the UF=FSAR Supplement, Appendix A, assembly, Lower grid extended period an SectioRs A.2.1.41 and A.3.1.41.

assembly, Flow RVI inspection plan See Section 3.1.2.2.17.

distributor assembly) based on industry recommendation.

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 15 of 90 Table 3.1.1: Reactor Coolant System, NUREG-1801 Vol. 1 Aging Effect/ Aging Further Item Mechanism Management Evaluation Discussion Number Component 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)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 16 of 90 NOTES FOR TABLES 3.1.2 1 lP2 THROUGH 3.1.2 4 11P3 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-1 801 AMP.

C. Component is different, but consistent with NUREG-1 801 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-1 801 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-1 801 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 in NUREG-1801.

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

As documented in MRP-227, the basis for the RVI Program, this combination warrants no additional aging managqement.

NURE 181, Sctinsttos:"Nofu~or aingmangomont roVioW is nocossar,' ifthe applicant providos a X.M1 cmitmont ithFSRupplomont to (1)participato in the industry programs for in;Vostigating and mAngg agingofctS onracrrl a; (2) evaluate ard imprmnt theo resultS f tho indu6ty programs as applicabl, to the roctor trnWAl and (3)upon comAplotion Of thoso programs, but no~t l6ss than 221 months bofo-ro- ontoering the poriod of oxtondod oporation, submit an inspection plan for F;reactor itrasto the N'RC for ro;v iow and approval." IPECG commFitmont can be foundi Appendix A (UJFSARI supplomont) Of the liconAso r-enoa4 application-.

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.

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 18 of 90 Table 3.1.2-2-1P2 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function 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 Lower Core Su .ort Structure .

Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1- E A, baffle/former support steel water > 140°F dimensions Internals 4W (R-126) 33 101 assembly GGtA bolts__ __

Cracking Water Chemistry IV.B2-10 3.1.1- EA-,

Control - Primary (R-125) 30 1-and Secondary Reactor Vessel Internals R.M4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 19 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Loss of Reactor Vessel IV.B2-5 3.1.1- E A, preload Internals PA4 (R-129) 27 101 GGFRM taeRt Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1- E A, water > 140°F fracture Internals PV- (R-128) 22 101 Neutron fluence toughness GOmMitmeRt Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-1 3.1.1- E A-,

baffle/former support steel water > 140°F dimensions Internals PA4 (R-124) 33 1-assembly Flow GOitMI t

• plates distribution Shielding Cracking Water Chemistry IV.B2-2 3.1.1- E A-Control - Primary (R-123) 30 1-and Secondary Reactor Vessel Internals PA4 AAM A-AtR 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- E A, water > 140°F fracture Internals PA4 (R-127) 22 101 Neutron fluence toughness GGtFAGRt

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 20 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Type Fntin Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1- E G, assembly support steel water > 140°F dimensions Internals PMV (R-126) 33 101 bolts-and eGwGM i.tMRt screws Cracking Water Chemistry IV.B2-10 3.1.1- EA-Control - Primary (R-125) 30 1-and Secondary Reactor Vessel Internals RV4 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- E A, preload Internals PM4 (R-129) 27 101 Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1- E A, water > 140°F fracture Internals PM4 (R-128) 22 101 Neutron fluence toughness GOM4WtMet

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 21 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E A, assembly support steel water > 140°F dimensions Internals AkV (R-121) 33 101

- axial flexure F4ew GOM.M.t.

plates ,dibuti,

,+,I4 (thermal ,$htebdiGR Cracking Water Chemistry IV.B2-8 3.1.1- EA-shield Control - Primary (R-120) 30 40 flexures) and Secondary Reactor Vessel Internals PV4 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- E material - Internals (R-142) 63 wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- E A, water > 140°F fracture Internals PA4 (R-122) 22 101 Neutron fluence toughness emitUmei

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 22 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E A, assembly support steel water > 140°F dimensions Internals R-V (R-121) 33 101 flange GOM __w dotouen

=nScdCracking Water Chemistry IV.B2-8 3.1.1- E A-,

Control - Primary (R-120) 30 1-and Secondary Reactor Vessel Internals PMV Gr-eR.m 4.eA 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 (R-1 15) 63 wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1- E A, water > 140°F fracture Internals PMV (R-122) 22 101 Neutron fluence toughness eeRnitment

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 23 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Function Material Environment Requiring Management 1801 Vol. Item Type a yp FunctioManagement Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E A, assembly support steel water > 140°F dimensions Internals RVI (R-121) 33 101

" ring Flow i t.

• shell distribution

• thermal shield Shielding Cracking Water Chemistry IV.B2-8 3.1.1- EA Control - Primary (R-120) 30 101 and Secondary Reactor Vessel Internals PAI 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- E A, water > 140°F fracture Internals PMV (R-122) 22 101 Neutron fluence toughness GGwAt4#i9Rt

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 24 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E assembly suDDort steel water > 140°F dimensions Internals (R-121) 33 101

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

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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 25 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG-Typoen Incten Material Environment Requiring Management 1801 Vol Table 1 Notes Type Function Management Programs 2 Item Item Core barrel Flow Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1- E A, assembly distribution steel water > 140°F dimensions Internals R-W (R-121) 33 101 outlet nozzles commitment Cracking Water Chemistry IV.B2-8 3.1.1- E A-Control - Primary (R-120) 30 and Secondary Reactor Vessel Internals V-I Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 26 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Lower internals Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- E A, assembly support water dimensions Internals PVI (R-134) 33 101 clevis insert #Ritment bolt Cracking Water Chemistry IV.B2-16 3.1.1- E A, Control - Primary (R-133) 37 101 and Secondary Reactor Vessel Internals PMV Geon 4FRA1 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- E A, preload Internals -V4 (R-137) 27 101 Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- E A, water fracture Internals P,4 (R-135) 22 101 Neutron fluence toughness Otnent

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 27 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Typet Fnctone Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Lower internals Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E A, assembly support water dimensions Internals PM (R-131) 33 101 clevis insert GGcAtm~nt Cracking Water Chemistry IV.B2-20 3.1.1- E A, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals PM GGRomtmnet 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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 28 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Type Fntin Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Lower internals Flow Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E G, assembly distribution steel water > 140°F dimensions Internals PM (R-131) 33 101 intermediate GOM diffuser plate Cracking Water Chemistry IV.B2-20 3.1.1- E G, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals PA4 cGRmRtmcGRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 29 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Function Material Environment Requiring Management 1801 Vol. Item Notes Type Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- E A, assembly support steel water> 140°F dimensions Internals PM (R-134) 33 101 fuel alignment comm4ment pin Cracking Water Chemistry IV.B2-16 3.1.1- E A, Control - Primary (R-133) 37 101 and Secondary Reactor Vessel Internals PV4 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- E A, water > 140'F fracture Internals PA4 (R-135) 22 101 Neutron fluence toughness GOMMit!eRt

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 30 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Typet Fnctone Material Environment Requiring Management 1801 Vol. te Notes Type Function Management Programs 2 Item Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.1B2-19 3.1.1- E A, assembly support steel water > 140°F dimensions Internals R-V (R-131) 33 101 lower core Flow GOtMent plate distribution Cracking Water Chemistry IV.B2-20 3.1.1- E A-Control - Primary (R-130) 37 401-and Secondary RV' cOMMitmont Inservice Inspection 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 Treated borated Reduction of Reactor Vessel IV.B2-18 3.1.1- E A, water > 140°F fracture Internals PMV (R-132) 22 101 Neutron fluence toughness GOFFft~eRt

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 31 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Lower internals Structural CASS Treated borated Change in Reactor Vessel IV.B2-23 3.1.1- E A, assembly support water > 482 0 F dimensions Internals RV4 (R-139) 33 101

• lower core Flow eeGG.AitM t support distribution castings Cracking Water Chemistry IV.B2-24 3.1.1- E

- column cap Control - Primary (R-138) 30 04-

- lower core and Secondary support Reactor Vessel column Internals PMV bodies _______

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 0 F fracture Neutron Irradiation (R-140) 80 Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 32 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Inten Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-15 3.1.1- E A, assembly support steel water > 140°F dimensions Internals MVI (R-134) 33 101 slower core rM.M.at.mAA support plate column bolt Cracking Water Chemistry IV.B2-16 3.1.1- E A-Control - Primary (R-133) 37 4" and Secondary Reactor Vessel Internals PM4 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- E A, preload Internals PM4 (R-136) 27 101

.....t..

.

Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1- E A, water > 140°F fracture Internals R-V (R-135) 22 101 Neutron fluence toughness t

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 33 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Environment Requiring Management 1801 Vol. Item Type Function Material Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-23 3.1.1- E A, assembly support steel water > 140°F dimensions Internals PV-\ (R-139) 33 101 elower core GGRmFA4RSt support column plate Cracking Water Chemistry IV.B2-24 3.1.1- E A, sleeves Control - Primary (R- 138) 30 101 and Secondary Reactor Vessel Internals PVM GGMRnRtmeRt 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- E A, water > 140°F fracture Internals -V4 (R-141) 22 101 Neutron fluence toughness GGFsim4ReRt

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 34 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol Item Type Function Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E A, assembly support steel water > 140°F dimensions Internals PM (R-131) 33 101 radial key Cracking Water Chemistry IV.B2-20 3.1.1- E A, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals R-V GGFMmt eR 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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 35 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Lower internals Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1- E G, assembly support steel water > 140°F dimensions Internals RV4 (R-131) 33 101 secondary Flow Geimwtmeft core support distribution Cracking Water Chemistry IV.B2-20 3.1.1- EG, Control - Primary (R-130) 37 101 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 36 of 90 Table 3.11.2-2-1132: Reactor Vessel Internals Aging NUREG- Table 1 Notes Management 1801 Vol.

Item Programs .2 Item

_b-i-Upper Core Support Structure - Upper Internals Assembly RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.1B2-27 3.1.1- EA, tube assembly support steel water > 140'F dimensions Internals PMV (R-1 19) 33 101

- bolt A= A-At4 Cracking Water Chemistry IV.B2-28 3.1.1- E A, Control - Primary (R-118) 37 101 and Secondary Reactor Vessel Internals PV4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.1B2-38 3.1.1 1E0, preload Internals PV4 (R- 114) 27 101 GeRmRtR4Gm

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 37 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol Item Type Function Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1- E A, tube assembly support steel water > 140°F dimensions Internals PA4 (R-1 17) 33 101 guide tube GGM (including lower flange Cracking Water Chemistry IV.1B2-30 3.1.1- E A, weldsa Control - Primary (R-1 16) 30 1-and Secondary Reactor Vessel Internals RVI Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 38 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Environment Requiring Management 1801 Vol. Item Type Function Material Management Programs 2 Item RCCA guide Structural Stainless Treated borated Changqe in Reactor Vessel IV.B2-29 3.1.1- E tube assembly support steel water > 140°F dimensions Internals (R-1 17) 33 guide plates Cracking Water Chemistry IV.B2-30 3.1.1- E Control - Primary (R-1 16) 30 and Secondary Reactor Vessel Internals Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary Loss of Reactor Vessel IV.B2-34 3.1.1- E material - Internals (R-115) 63 wear

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 39 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Inten Management Programs 2 Item RCCA guide Structural NiGkel-aley, Treated borated Change in Reactor Vessel IV.B2-27 3.1.1- E A, tube assembly support Stainless water dimensions Internals FM (R-119) 33 101 support pin steel Cracking Water Chemistry IV.B2-28 3.1.1- E, 105 Control - Primary (R-118) 37 A,4 and Secondary Reactor Vessel Internals PVM GeMMitmeRt Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 40 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Core plate Structural Stainless Treated borated Change in Reactor Vessel IV.1B2-39 3.1.1- E A, alignment pin support steel water > 140°F dimensions Internals PA4 (R-1 13) 33 101 GOR44M R Cracking Water Chemistry IV.B2-40 3.1.1- E A, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals PM 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 (R1i15) 63 wear

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 41 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Head / vessel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- E G, alignment pin support steel water > 140°F dimensions Internals VI, (R-107) 33 101 GGRR#Apt Cracking Water Chemistry IV.B2-42 3.1.1- E G, Control - Primary (R-106) 30 101 and Secondary Reactor Vessel Internals 4 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 (R1 15) 63 wear

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 42 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Hold-down Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1- E A, spring support steel water > 140°F dimensions Internals P,4 (R-107) 33 101 GGM~t~~

Cracking Water Chemistry IV.B2-42 3.1.1- E A, Control - Primary (R-106) 30 101 and Secondary Reactor Vessel Internals P-V4 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- E A-preload Internals MV4 (R-114) 27 1 GGte4m*mPt

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 43 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Function Material Environment Requiring Management 1801 Vol. Item Type Management Programs 2 Item Mixing devices Structural CASS Treated borated Change in Reactor Vessel IV.B2-35 3.1.1- E G,

• support support water > 482 0 F dimensions Internals RV- (R- 110) 33 101 column orifice Flow GGtFeA t base bsupport m distribution Cracking Water Chemistry IV.B2-36 3.1.1- E G, column mixer Control - Primary (R- 109) 30 101 and Secondary Reactor Vessel Internals R-V4 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 > 4820F fracture Neutron Irradiation (R-1 11) 80 Neutron fluence toughness Embrittlement of Cast Austenitic Stainless Steel (CASS)

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 44 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table I Notes Environment Requiring Management 1801 Vol. Item Type Function Material Management Programs 2 Item Support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-35 3.1.1- E A, column support steel water > 140°F dimensions Internals RVP (R- 110) 33 101 Cracking Water Chemistry IV.B2-36 3.1.1- E A, Control - Primary (R-109) 30 101 and Secondary Reactor Vessel Internals -A4 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- E A, plate, fuel support steel water > 140°F dimensions Internals RV4 (R-117) 33 101 alignment pin Flow r M. 0tmeA distribution Cracking Water Chemistry IV.B2-40 3.1.1- E A, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals RVt GO*, tRW4=#

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 45 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Environment Requiring Management 1801 Vol. Notes Type Function Material Management Programs 2 Item 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- E A, plate, support support steel water > 140°F dimensions Internals PMVI (R-107) 33 101 assembly _OMmntme~t (including rincq) Cracking Water Chemistry IV.B2-42 3.1.1- E A-Control - Primary (R-106) 30 and Secondary Inservice Inspection RV4 OMviMien4 Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 46 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1- E A, column bolt support steel water > 140°F dimensions Internals P,4 (R-1 13 33 101 GeOAm; t Cracking Water Chemistry IV.B2-40 3.1.1- E A, Control - Primary (R-112) 37 101 and Secondary Reactor Vessel Internals PA4 ZemGGRifept 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- E A, preload Internals PA4 (R-1 14) 27 101 GGRR4Stet

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 47 of 90 Bottom Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- EA, mounted support steel water > 140°F dimensions Internals -V4 (R- 144) 33 101 instrumentation GGR;Fn4RnA~4 column Cracking Water Chemistry IV.B2-12 3.1.1- E A-,

Control - Primary (R-143) 30 40 and Secondary Reactor Vessel Internals P-4 G*OMM4.MSR Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 48 of 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Type Fntin Material Environment Requiring Management 1801 Vol. Ite Type Function Management Programs 2 Item Item Flux thimble Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1- E A, guide tube support steel water > 140°F dimensions Internals PM4 (R-144) 33 101 GGcM tMARt Cracking Water Chemistry IV.B2-12 3.1.1- E A, Control - Primary (R-143) 30 101 and Secondary Reactor Vessel Internals PMV ZeRmt9eAt 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- E G, conduit support steel water > 140°F dimensions Internals PM (R-144) 33 101 Cracking Water Chemistry IV.B2-12 3.1.1- E G, Control - Primary (R-143) 30 101 and Secondary Reactor Vessel Internals PM

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 49 oi 90 Table 3.1.2-2-1P2: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Environment Requiring Management 1801 Vol. tem Notes Type Function Material Management Programs 2 Item Item Loss of Water Chemistry IV.B2-32 3.1.1- A material Control - Primary (RP-24) 83 and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 50 of 90 Table 3.1.2-2-1P3 Reactor Vessel Internals Summary of Aging Management Review Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Type Inten Material Environment Requiring Management 1801 Vol. Item Notes Type Function 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 Core Structural Stainless Treated borated Change in Reactor Vessel [V.132-4 3.1.1-33 E A, baffle/former support steel water > 140°F dimensions Internals R-V (R-126) 101 assembly aAAitAet bolts Cracking Water Chemistry IV.B2-10 3.1.1-30 E A-Control - Primary (R-125) 1-and Secondary Reactor Vessel Internals -V4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 51 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Loss of Reactor Vessel IV.B2-5 3.1.1-27 E A, preload Internals PRV4 (R- 129) 101 Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1-22 E A, water > 140°F fracture Internals R-V (R-128) 101 Neutron fluence toughness Gemmfitre Core Structural Stainless Treated borated Change in Reactor Vessel IV.B2-1 3.1.1-33 E A-,

baffle/former support steel water > 140°F dimensions Internals V-I (R-124) assembly Flow GeMMit.IAnt

• plates distribution Shielding Cracking Water Chemistry IV.B2-2 3.1.1-30 E A-Control - Primary (R-123) 1-and Secondary Reactor Vessel Internals PMV GemmtmenI 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 E A, water > 140°F fracture Internals PMV (R-127) 101 Neutron fluence toughness Gommtment

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 52 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Ie Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-4 3.1.1-33 E G, assembly support steel water > 140°F dimensions Internals PM (R-126) 101 bolts-and oe itFAeAt

.

screws Cracking Water Chemistry IV.B2-10 3.1.1-30 EA-,

Control - Primary (R-125) and Secondary Reactor Vessel Internals PMVI GGRMiteR4 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 E A, preload Internals PR-V (R-129) 101 Treated borated Reduction of Reactor Vessel IV.B2-6 3.1.1-22 E A, water > 140°F fracture Internals P-4 (R-128) 101 Neutron fluence toughness commwtment

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 53 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. tem Notes Type Function Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly support steel water > 140°F dimensions Internals PV-I (R-121) 101 axial flexure Flow Gammitme4 plates (thermal *;oetihb,÷i,,

&hie*buee Cracking Water Chemistry IV.B82-8 3.1.1-30 E A-shield Control - Primary (R-120) 4-I flexures) and Secondary Reactor Vessel Internals PV-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 E material - Internals (R- 142) wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 E A, water> 140°F fracture Internals:RVJ (R-122) 101.

Neutron fluence toughness GGMmFet

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 54 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly support steel water > 140°F dimensions Internals MV4 (R-121) 101 flange Flew GeeMMtm .

Cracking Water Chemistry IV.B2-8 3.1.1-30 E A-Control - Primary (R-120) and Secondary Reactor Vessel Internals MVI 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 (R-1 15) wear Treated borated Reduction of Reactor Vessel IV.B2-9 3.1.1-22 E A, water > 140°F fracture Internals P1-4 (R-122) 101 Neutron fluence toughness GGmitRIent

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 55 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly support steel water > 140°F dimensions Internals PA4 (R-121) 101

• ring Flow GOMM tMAnt

• shell distribution

• thermal Shielding Cracking WaterChemistry IV.B2-8 3.1.1-30 EA shield Control - Primary (R-120) 101 and Secondary Reactor Vessel Internals P-V-GGR4RRit~n9t 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 E A, water > 140°F fracture Internals PMV (R-122) 101 Neutron fluence toughness GGfMM.MeRt

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 56 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Typet Fnctone Material Environment Requiring Management 1801 Vol. te Notes Type Function Management Programs 2 Item Item Core barrel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E, assembly support steel water > 140°F dimensions Internals (R-121) 101

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

" upper 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 _E water > 140°F fracture Internals (R-122) 101 Neutron fluence toughness

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 57 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Environment Requiring Management 1801 Vol. Notes Type Function Material Management Programs 2 Item Item Core barrel Flow Stainless Treated borated Change in Reactor Vessel IV.B2-7 3.1.1-33 E A, assembly distribution steel water > 140°F dimensions Internals RVI (R-121) 101

- outlet eenntmneAt nozzles Cracking Water Chemistry IV.B2-8 3.1.1-30 EA-Control - Primary (R-120) 1-and Secondary Reactor Vessel Internals R-VM GOMMitMeat Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 58 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals, Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-15 3.1.1-33 E A, internals support water dimensions Internals PA4 (R-134) 101 assembly GGeMmRtme

• clevis bolt insert Cracking Water Chemistry IV.B2-16 3.1.1-37 E A, Control - Primary (R-133) 101 and Secondary Reactor Vessel Internals PA4 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 E A, preload Internals RV4 (R-137) 101 Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 E A, water fracture Internals PM4 (R-135) 101 Neutron fluence toughness

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 59 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Component Fnctiond Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Lower Structural Nickel alloy Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E A, internals support water dimensions Internals R-VJ (R-131) 101 assembly clevis insert Cracking Water Chemistry IV.82-20 3.1.1-37 E A, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals R-VM 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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 60 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Environment Requiring Management 1801 Vol. Item Type Function Material Management Programs 2 Item Lower Flow Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E G, internals distribution steel water > 140°F dimensions Internals -V4 (R-131) 101 assembly _em _M__

• intermediate diffuser plate Cracking Water Chemistry IV.B2-20 3.1.1-37 E G, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals P1M4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 61 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- 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 E A, internals support steel water > 140OF dimensions Internals -V4 (R- 134) 101 assembly GFA. t AGR fuel alignment Cracking Water Chemistry IV.B2-16 3.1.1-37 E A, pin Control - Primary (R-133) 101 and Secondary Reactor Vessel Internals PR-4 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 E A, water > 140°F fracture Internals PA4 (R-135) 101 Neutron fluence toughness GG4M.iti.t

N L-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 62 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E A, internals support steel water > 140'F dimensions Internals -V4 (R-131) 101 assembly Flow

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

Control - Primary (R-130) 40 and Secondary RI comm*itmontIIV I Inservice Inspection 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 Treated borated Reduction of Reactor Vessel IV.B2-18 3.1.1-22 E A, water > 140°F fracture Internals -V4 (R-132) 101 Neutron fluence toughness GeMmnitment

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 63 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Ie Lower Structural CASS Treated borated Change in Reactor Vessel IV.B2-23 3.1.1-33 E A, internals support water > 482 0 F dimensions Internals R-V (R-139) 101 assembly Flow slower core distribution Cracking Water Chemistry IV.B2-24 3.1.1-30 E A-,

support castings Control - Primary (R-138) 40 - column cap and Secondary

- lower core Reactor Vessel support Internals R-V column 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 > 4820 F fracture Neutron Irradiation (R-140)

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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 64 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- 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 E A, internals support steel water > 140°F dimensions Internals PMV (R-134) 101 assembly

• lower core support plate Cracking Water Chemistry IV.B2-16 3.1.1-37 EA-column bolt Control - Primary (R-133) and Secondary Reactor Vessel Internals R-V4 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 E A, preload Internals PMV (R-136) 101 Treated borated Reduction of Reactor Vessel IV.B2-17 3.1.1-22 E A, water > 140°F fracture Internals R-V4 (R-135) 101 Neutron fluence toughness GeOMMOieRt

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 65 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-23 3.1.1-33 E A, internals support steel water > 140'F dimensions Internals PM (R-139) 101 assembly

• lower core support plate Cracking Water Chemistry IV.B2-24 3.1.1-30 E A, column Control - Primary (R-138) 101 sleeves and Secondary Reactor Vessel Internals FM 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 E A, water > 140°F fracture Internals RPM (R-141) 101 Neutron fluence toughness GGFMimSR4

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 66 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Environment Requiring Management 1801 Vol. Item Type Function Material Management Programs 2 Item Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E A, I internals support steel water > 140°F dimensions Internals -V4 (R-131) 101 assembly radial key Cracking Water Chemistry IV.B2-20 3.1.1-37 E A, Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals P-V4 eGRRiteRt 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

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 67 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function 2 Item Management Programs Lower Structural Stainless Treated borated Change in Reactor Vessel IV.B2-19 3.1.1-33 E G, internals support steel water > 140°F dimensions Internals R-V (R-131) 101 assembly Flow secondary distribution core support Cracking Water Chemistry IV.B2-20 3.1.1-37 EG Control - Primary (R-130) 101 and Secondary Reactor Vessel Internals RV4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 68 of 90 Table 3.1 .2-2-1133: Reactor Vessel Internals Aging Effect Aging NUREG- Table 1 Component Intended Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2IItem t RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.1B2-27 3.1.1-33 EA, tube assembly support steel water > 140OF dimensions Internals RV4 (R-1 19) 101

• bolt Gomm.4--tA Cracking Water Chemistry IV.B2-28 3.1.1-37 E A, Control - Primary (R-118) 101 and Secondary Reactor Vessel Internals RMV GGeR4tmeAt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.1B2-38 3.1.1-27 EG, preload Internals PM4 101 (R- 114)

GeM R4tmet

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 69 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1-33 E A, tube assembly support steel water > 140°F dimensions Internals -V4 (R-117) 101

- guide tube GGR4RtRn9Rt (including lower flange Cracking Water Chemistry IV.1B2-30 3.1.1-30 E A, welds). Control - Primary (R-116) 401-and Secondary Reactor Vessel Internals FMV Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 70 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Ie RCCA guide Structural Stainless Treated borated Change in Reactor Vessel IV.B2-29 3.1.1-33 E tube assembly sumport steel water > 140°F dimensions Internals (R-1 17)

- guide plates Cracking Water Chemistry IV.B2-30 3.1.1-30 E Control - Primary (R-1 16) and Secondary Reactor Vessel Internals Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.B2-34 3.1.1-63 E material - Internals (R-1 15) wear

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 71 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Function Material Environment Requiring Management 1801 Vol. Notes Type Management Programs 2 Item Item RCCA guide Structural Nikel-alley, Treated borated Change in Reactor Vessel IV.B2-27 3.1.1-33 E A, tube assembly support Stainless water dimensions Internals PMV (R-1 19) 101 support pin steel eemmitme At Cracking Water Chemistry IV.B2-28 3.1.1-37 E Control - Primary (R-1 18) 105 A-and Secondary 40---

Reactor Vessel Internals PM4 Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 72 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Notes Material Environment Requiring Management 1801 Vol. Item Type Function 2 Item Management Programs Core plate Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 E A, alignment pin support steel water > 140°F dimensions Internals PA4 (R-113) 101 Cracking Water Chemistry IV.B2-40 3.1.1-37 E A, Control - Primary (R-1 12) 101 and Secondary Reactor Vessel Internals PA4 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 (R1 15) wear

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 73 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Head /vessel Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 E G, alignment pin support steel water > 140°F dimensions Internals PM (R-107) 101 GGR~Mo4 Cracking Water Chemistry IV.B2-42 3.1.1-30 E G, Control - Primary (R-106) 101 and Secondary Reactor Vessel Internals PM 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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 74 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 mponen Fnctind Material Environment Requiring Management 1801 Vol. te Notes Type Function Management Programs 2 Item Item Hold-down Structural Stainless Treated borated Change in Reactor Vessel IV.B2-41 3.1.1-33 E A, spring support steel water > 140°F dimensions Internals PIM (R-107) 101 Cracking Water Chemistry IV.B2-42 3.1.1-30 EA, Control - Primary (R-106) 101 and Secondary Reactor Vessel Internals R-V GGR*4RR Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary Loss of Reactor Vessel IV.1B2-38 3.1.1-27 EA-preload Internals PM. (R- 114) 40 GGRmmRtmemt

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 75 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Mixing Structural CASS Treated borated Change in Reactor Vessel IV.B2-35 3.1.1-33 E G, devices support water > 482 0 F dimensions Internals PA4 (R- 110) 101

" support Flow AneMeMMit column distribution orifice base Cracking Water Chemistry IV.B2-36 3.1.1-30 EG

• support Control - Primary (R-109) 101 column and Secondary mixer Reactor Vessel Internals PA4 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 0 F fracture Neutron Irradiation (R-1 11)

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

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 76 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-35 3.1.1-33 E A, column support steel water > 140'F dimensions Internals PA4 (R- 110) 101 Cracking Water Chemistry IV.B2-36 3.1.1-30 E A, Control - Primary (R-109) 101 and Secondary Reactor Vessel Internals RV-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 E A, plate, fuel support steel water > 140°F dimensions Internals P-V-M (R-1 17) 101 alignment pin Flow GeMMR.tM eAt distribution Cracking Water Chemistry IV.B2-40 3.1.1-37 E A, Control - Primary (R-112) 101 and Secondary Reactor Vessel Internals PMV GGR4RRtW4.

NL-1 0-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 77 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item 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 E A, plate, support support steel water > 140°F dimensions Internals -V4 (R-107) 101 assembly ccM itM It (including____

ring) Cracking Water Chemistry IV.1B2-42 3.1.1-30 E A-Control - Primary (R-106) 1-and Secondary Inservice Inspection -V4 GOeM, *GRt Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 78 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Notes Type Function Management Programs 2 Item Item Upper support Structural Stainless Treated borated Change in Reactor Vessel IV.B2-39 3.1.1-33 E A, column bolt support steel water > 140°F dimensions Internals PM (R-113 101 Cracking Water Chemistry IV.B2-40 3.1.1-37 E A, Control - Primary (R-112) 101 and Secondary Reactor Vessel Internals PA4 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 E A, preload InternalsRW (R-114) 101

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 79 of 90 Table 3.1 .2-2-1133: Reactor Vessel Internals Aging Effect Aging NUREG-Component Intended Table 1 Notes Material Environment Requiring Management 1801 Vol.

Type Function Item Management Programs 2 Item

.1.

Bottom Structural Stainless Treated borated Change in Reactor Vessel IV.B2-11 3.1.1-33 EA, mounted support steel water > 140°F dimensions Internals RV4 (R-144) 101 instrumentatio ee~mmitFeB n column Cracking Water Chemistry IV.B2-12 3.1.1-30 EA-,

Control - Primary (R-143) 40 and Secondary Reactor Vessel Internals -V4 GGs nitnet Loss of Water Chemistry IV.1B2-32 3.1.1-83 1 A material Control - Primary (RP-24) and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 80 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component Intended Aging Effect Aging NUREG- Table 1 Material Environment Requiring Management 1801 Vol. Item Notes Type Function Management Programs 2 Item Flux thimble Structural Stainless Treated borated Change in Reactor Vessel IV.B2-111 3.1.1-33 E A, guide tube support steel water > 140°F dimensions Internals PA4 (R-144) 101 eGGm ne04 Cracking Water Chemistry IV.B2-12 3.1.1-30 E A, Control - Primary (R-143) 101 and Secondary Reactor Vessel Internals PVI 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 E G, conduit support steel water > 140°F dimensions Internals PM (R-144) 101 GORMM.tRtR Cracking Water Chemistry IV.B2-12 3.1.1-30 E G, Control - Primary (R-143) 101 and Secondary Reactor Vessel Internals PA4 GGi n m nt

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 81 of 90 Table 3.1.2-2-1P3: Reactor Vessel Internals Component IntendedTal1 Aging Effect Aging NUREG- Table 1 Type Function Material Environment Requiring Management 1801 Vol. Notes Management Programs 2 Item Item Loss of Water Chemistry IV.B2-32 3.1.1-83 A material Control - Primary (RP-24) and Secondary

NL-10-063 Attachment 1 Docket Nos. 50-247 & 50-286 Page 82 of 90 A.2.1.41 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 quidance from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP). The MRP inspection and evaluation (M&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 qualification 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 IPEC Reactor Vessel Internals (RVI) Proqram.

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. The RVI Program will monitor the effects of aging degradation mechanisms on the intended function of the internals throuqh 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 IPEC RVI Program will be implemented and maintained in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation Management Program Guidelines." Any deviations from mandatory, needed, or good practice implementation reguirements 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 maFnago o O of. fracturo toughness, crac.ing, chang. in d...m.'OnSn 196 IYei IIl"g)

I~ vesse If IIeea IGRI GGVIII eF*tI IOe the wIIilli ()Ip *1 GIpI te4*II the in~dustry programns for invoctigating and managin aig effects on rcactorneras (2) evaluate and im~plement the roSUltS of the nutyporm as applicablo to the roacto internals; and (3) upon completion of those programsR, but not loss thanR 24 monGths _bo~foo onterfing the period of oxtondod operation, submiwt an inspection plan for reactor internals to the NIRC for review and approval.

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 Progqram (MRP). The MRP inspection and evaluation (M&E)

Quidelines 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 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. 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 IPEC RVI Program will be implemented and maintained in accordance with the guidance in NEI 03-08 [Addendal, Addendum A, "RCS 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- manaFgo les of fracture t.ughnoss, cracking, chanRg *I;dimensions (void sWcllg), andiloS Of prclead in vessel intcralS coMpononts, the site will (1) patiGipate in the induetry programse for investigating andRmanaging agin~g effects on rcactor OntoFralc"; (2) evaluate and im~plem:ent the results of the ind~ustry programs a6 appliabOG-4h to the reactor aintornals; and (3) upon complction Of those program~s, but not less than 24 mRonthS before entcring the period of cextended operation, submiet an inspection plan for reactorF internal to the NIRC fo-r review and approval.

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.

B.1.42 Reactor Vessel Internals Program Program Description The Reactor 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.

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.

IPEC will implement and maintain'the RVI Program in accordance with the guidance in NEI 03-08 [Addenda], Addendum A, "RCS Materials Degradation 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 fuel 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 to 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, as presented in MRP-227, applies to IPEC 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 IPEC RVI Program inspections. Those components subject to aging management by existing programs, as delineated in MRP-227, Table 4-9, are included in

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)

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 (i.e., 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. Measurements will 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.

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 examination 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 also 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 X1.

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 IPEC. However, program inspections will use qualified techniques similar to those successfully used at IPEC and throughout the industry for ASME Section Xl Code inspections. Internals inspections (VT-3) have been conducted at IPEC in accordance with ASME Section XI Code requirements, with no indications of component degradation. IPEC has appropriately responded 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 IPEC units. As a result, IPEC has monitored industry developments and recommendations regarding these components.

Development 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.

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 IPEC 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.