AEP-NRC-2012-82, Transmittal of Reactor Vessel Internals Aging Management Program

Transmittal of Reactor Vessel Internals Aging Management Program

ML12284A320
Person / Time
Site:	Cook
Issue date:	10/01/2012
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2012-82
Download: ML12284A320 (75)

Text

INDIANA MICHIGAN Indiana Michigan Power POWERý One Cook Place Bridgman, MI 49106 A unit of American Electric Power IndianaMichiganPower.com October 1, 2012 AEP-NRC-2012-82 10 CFR 50.4 Docket Nos.: 50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 TRANSMITTAL OF REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM

Reference:

1) Letter from P-T Kuo, U. S. Nuclear Regulatory Commission (NRC), to M. K. Nazar, Indiana Michigan Power Company (I&M), "Safety Evaluation Report Related to the License Renewal of the Donald C. Cook Nuclear Plant, Units 1 and 2," dated May 29, 2005. Agencywide Documents Access and Management System (ADAMS)

Accession No. ML051510015.

2) Regulatory Issue Summary 2011-07, "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," dated July 21, 2011.

ADAMS Accession No. ML111990086.

3) Letter from J. P. Gebbie, Indiana Michigan Power Company, to U. S. Nuclear Regulatory Commission, "Donald C. Cook Nuclear Plant Units 1 and 2, Docket No. 50-315 and 50-316, Revision to Regulatory Commitments Associated with Application for Renewed Operating Licenses," AEP-NRC-2011-38, dated September 1, 2011. ADAMS Accession No. ML11256A017.

Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP), Units 1 and 2, is submitting the Reactor Vessel Internals (RVI) Aging Management Program (AMP). By Reference 1, the NRC published Safety Evaluation Report (SER) Related to the License Renewal of the CNP. The SER contained a list of commitments made by I&M, specifically, I&M committed to submit the RVIs Plates, Forgings, Welds, and Bolting Program for NRC Staff review and approval three years prior to the period of extended operations. I&M also committed to implement the Cast Austenitic Stainless Steel (CASS) Evaluation Program prior to the period of extended operation. By Reference 3, I&M submitted a revision to the commitment regarding RVIs Plates, Forgings, Welds, and Bolting Program consistent with the guidance contained in Reference 2. The guidance contained in Reference 2 allowed CNP to modify the commitments to reflect a requirement to submit the AMP for Unit 1 no later than October 1, 2012. The due date for Unit 2 remains December 23, 2014.

U.S. Nuclear Regulatory Commission AEP-NRC-2012-82 Page 2 to this letter provides the CNP RVI AMP and satisfies the commitments for submitting the Unit 1 and Unit 2 RVIs Plates, Forgings, Welds, and Bolting Program and implementing the CASS Evaluation Program. Enclosure 2 contains a list of commitments made in the RVI AMP.

Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President DMB/kmh

Enclosures:

1. Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program

2. List of Regulatory Commitments c: C. A. Casto, NRC Region III J. T. King - MPSC S. M. Krawec - AEP Ft Wayne MDEQ- RMD/RPS NRC Resident Inspector T.J. Wengart, NRC Washington DC

ENCLOSURE, 1 TO AEP-NRC-2012-82 DONALD C. COOK NUCLEAR PLANT REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program

~ AMERICAN ELECTRIC POWER Donald C. Cook Nuclear Plant Units 1 and 2 Reactor Vessel Internals Aging Management Program Page 1 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program REVISION APPROVAL SHEET TITLE: Reactor Vessel Internals Aging Management Program Donald C. Cook Nuclear Plant, Units I and 2 PROGRAM ACCEPTANCE Prepared: __ _ _ _ __ _ _ _ _/&3-?

Kevin A. Kalchik Engineer II Reviewed:

Alex M. Olp f Engineer II Approved:

Dona_*ld S. Natfito-n cZ"# ~' -d#"

Manager Design Engineering The Revision Approval Sheet will be signed and the following Revision Control Sheet shall be completed to provide a record of the revision history each time this document is revised. The signatures above apply only to the changes made in the revision noted. Signatures for superseded revisions are retrievable through Donald C. Cook Nuclear Plant archives.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program REVISION CONTROL SHEET Major changes shall be outlined within the table below. Minor editorial and formatting revisions are not required to be logged.

REVISION [ DATE REVISION

SUMMARY

0 9/3/2012 Initial issuance.

Prepared: K. Kalchik Reviewed: A. Olp Approved: D. Naughton Page 3 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program TABLE OF CONTENTS 1.0 P U R PO SE ........................................................................................................................... 8 2.0 DESRIPTION OF CNP REACTOR INTERNALS .................................................. 9 2.1 Unit 1 OperatingExperience and Records Search................................................ 13 2.1.1 ControlRod Guide Tube and Split Pin Replacement..................................... 13 2.1.2 Barrel-FormerBolt Inspection and PartialReplacement.............................. 13 2.1.3 Clevis Insert Bolt Degradation........................................................................ 13 2.2 Unit 2 OperatingExperience and Records Search................................................ 14 2.2.1 Modificationfrom 15XJ5 to 1 7X1 7 Design .................................................... 14 2.2.2 ControlRod Guide Tube Split Pin Replacement ........................................... 14 2.2.3 ControlRod Guide Tube Cap Screw Modification......................................... 14 2.2.4 Baffle-FormerBolt PartialReplacement......................................................... 14 3.0 EXISTING PROGRAMS AND ACTIVITIES ........................................................ 15 3.1 ASME Section XI ISI Program.............................................................................. 15 3.2 PrimaryStrategic Water ChemistryPlan............................................................... 15 3.3 Thimble Tube Multifrequency Eddy CurrentInspection ....................................... 15 3.4 Industry Involvement .............................................................................................. 15 4.0 INSPECTION & EVALUATION GUIDELINES .................................................... 16 4.1 Guideline Background............................................................................................ 17 4.2 MRP-22 7-A Applicability to CNP........................................................................... 17 4.3 NEI 03-08 Guidancein MRP-22 7-A ........................................................................ 18 4.3.1 NEI 03-08 Mandatory..................................................................................... 18 4.3.2 NEI 03-08 Needed ............................................................................................ 18 4.4 Safety Evaluation Report Conditionsand Limitations........................................... 20 4.4.1 Topical Report Conditions............................................................................... 20 4.4.2 Applicant/LicenseeAction Items .................................................................... 20 5.0 PROGRAM ATTRIBUTE EVALUATION ............................................................. 26 5.1 Scope of Program..................................................................................................... 28 5.2 PreventiveA ctions ................................................................................................... 29 5.3 ParametersMonitored/Inspected............................................................................ 29 5.4 Detection of Aging Effects ....................................................................................... 31 5.5 Monitoringand Trending....................................................................................... 32 5.6 A cceptance Criteria.................................................................................................. 33 5.7 CorrectiveA ctions .................................................................................................... 34 5.8 ConfirmationProcess .............................................................................................. 35 5.9 Administrative Controls.......................................................................................... 35 5.10 OperatingExperience .............................................................................................. 35 6.0 RE FE RE N C E S ................................................................................................................. 37 APPENDIX A: PRIMARY INSPECTION COMPONENTS ........................................... 40 APPENDIX B: EXPANSION INSPECTION COMPONENTS ......................................... 45 APPENDIX C: EXISTING PROGRAMS COMPONENTS ............................................. 49 APPENDIX D: ACCEPTANCE AND EXPANSION CRITERIA ..................................... 51 APPENDIX E: REACTOR COMPONENT ILLUSTRATIONS ...................................... 58 Page 4 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program LIST OF FIGURES Figure 2-1 Overview of Typical Westinghouse Internals ......................................................... 9 Figure E- 1 Typical Westinghouse Control Rod Guide Card .................................................. 59 Figure E-2 Lower Section of Control Rod Guide Tube Assembly ......................................... 60 Figure E-3 M ajor Core Barrel W elds ...................................................................................... 61 Figure E-4 Bolting Systems Used in Westinghouse Core Baffles ......................................... 62 Figure E-5 Core Baffle/Barrel Structure ................................................................................. 63 Figure E-6 Bolting in a Typical Westinghouse Baffle-Former Structure ................................ 64 Figure E-7 Vertical Displacement between the Baffle Plates and Bracket at the Bottom of the Baffle-Former-Barrel Assembly (exaggerated) ............................................................... 65 Figure E-8 Schematic Cross-Sections of the Westinghouse Hold-down Springs ................... 66 Figure E-9 Typical Therm al Shield Flexure ............................................................................ 66 Figure E-I10 Lower Core Support Structure ............................................................................ 67 Figure E-1 1 Lower Core Support Structure - Core Support Plate Cross-Section ................... 68 Figure E-12 Typical Core Support Column ............................................................................. 68 Figure E- 13 Examples of Bottom-Mounted Instrumentation (BMI) Column Designs ........... 69 Page 5 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program LIST OF ACRONYMS AND ABBREVIATIONS AMP Aging Management Program AMR Aging Management Review ARDM Age Related Degradation Mechanism ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox BMI Bottom Mounted Instrumentation BWR Boiling Water Reactor CAP Corrective Actions Program CASS Cast Austenitic Stainless Steel CE Combustion Engineering CFR Code of Federal Regulations CLB Current Licensing Basis CRDM Control Rod Drive Mechanism CRGT Control Rod Guide Tube CNP Donald C. Cook Nuclear Plant EC Engineering Change EFPY Effective Full Power Year EPRI Electric Power Research Institute ET Electromagnetic Testing (eddy current)

EVT Enhanced Visual Testing FMECA Failure Mode, Effects, and Criticality Analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation I&M Indiana & Michigan Power IASCC Irradiation Assisted Stress Corrosion Cracking IGSCC Inter-Granular Stress Corrosion Cracking INPO institute of Nuclear Power Operations ISI In-Service inspection ISR Irradiation-enhanced Stress Relaxation JCO Justification for Continued Operation LRA License Renewal Application LRSS Lower Radial Support System MRP Materials Reliability Program NDE Non-Destructive Examination NEI Nuclear Energy Institute NOS Nuclear Oversight Section NRC U.S. Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OE Operating Experience OEM Original Equipment Manufacturer PH Precipitation Hardenable PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group Page 6 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program PWSCC Primary Water Stress Corrosion Cracking QA Quality Assurance QAPD Quality Assurance Program Description RAI Request for Additional Information RCS Reactor Coolant System RFO Refueling Outage RI-FG Reactor Internals Focus Group RFO Re-Fueling Outage RV- Reactor Vessel RVI Reactor Vessel Internals SCC Stress Corrosion Cracking SE Safety Evaluation SRP Standard Review Plan SS Stainless Steel UCP Upper Core Plate UFSAR Updated Final Safety Analysis Report UT Ultrasonic Testing VT Visual Testing WCAP Westinghouse Commercial Atomic Power WOG Westinghouse Owners Group Page 7 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 1.0 PURPOSE The purpose of the Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program is to manage the effects of aging on reactor vessel internals for the remainder of the operating license of the plant. The program implements guidance from the Electric Power Research Institute provided in MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor internals Inspection and Evaluation Guidelines" (MRP-227-A), and MRP-228, "Materials Reliability Program: Inspection Standard for PWR Internals" (MRP-228) to manage aging effects on the reactor vessel internal components. Existing programs are credited including the American Society of Mechanical Engineers Boiler & Pressure Vessel Code Section XI In-Service Inspection Program, Primary Strategic Water Chemistry Plan, and Incore Instrumentation Thimble Tube Multifrequency Eddy Current Inspections.

The program meets Donald C. Cook Nuclear Plant Nuclear Regulatory Commission commitments for license renewal as described in Appendix A of NUREG- 1831, "Safety Evaluation Report Related to the License Renewal of the Donald C. Cook Nuclear Plant, Units 1 and 2" (NUREG-1831). Specifically, NUREG-1831 Appendix A Items 19, 20, and 36, as modified by AEP-NRC-2011-38 in accordance with NRC RIS 11-07 "NRC Regulatory Issue Summary 2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management." These commitments require that a Reactor Vessel Internals Aging Management Program be developed in accordance with industry guidance and provided to the Nuclear Regulatory Commission for review and approval.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 2.0 DESRIPTION OF CNP REACTOR INTERNALS Donald C. Cook Nuclear Plant Units 1 and 2 are four-loop Westinghouse reactors in the downflow configuration. A schematic view of typical Westinghouse internals from MRP-227-A is shown in Figure 2-1. Illustrations of components can be found in Appendix E.

- ROD TRAVEL HOUSING

- INSTRUMENTATION PORTS

- THERMAL SLEEVE

- LIFTING LUG

- CLOSURE HEAD ASSEMBLY HOLD-DOWN SPRING

- CONTROL ROD GUIDE TUBE

- CONTROL ROD DRIVE SHAFT

- INLET NOZZLE CONTROL ROD CLUSTER (WITHDRAWI ACCESS PORT INSTRUMENTATION REACTOR VESSEL THIMBLE GUIDES 41 RADIAL SUPPORT CORE SUPPORT LOWER CORE PLATE Figure 2-1 Overview of Typical Westinghouse Internals Page 9 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Intemals Aging Management Program The following summary of Westinghouse PWR internals is an excerpt from MRP-227-A:

All Westinghouse internals consist of two basic assemblies: an upper internalsassembly that is removed during each refueling operation to obtain access to the reactorcore and a lower internals assembly that can be removedfollowing a complete core off-load.

The lower internalsassembly is supportedin the vessel by clamping to a ledge below the vessel-head mating surface and is closely guided at the bottom by radialsupport/clevis assemblies. The upper internals assembly is clamped at this same ledge by the reactor vessel head. The bottom of the upper internals assembly is closely guided by the core barrelalignmentpins of the lower internals assembly.

UpperInternalsAssembly The majorsub-assemblies that comprise the upper internalsassembly are the: (1) upper coreplate (UCP)andfuel alignmentpins; (2) upper support column assemblies; (3) control rod guide tube assemblies andflow downcomers; (4) upperplenum; and (5) upper supportplate assembly.

Duringreactor operation,the upper internals assembly is preloadedagainstthefuel assembly springs and the internals holddown springs by the reactorvessel headpressing down on the outside edge of the upper supportplate (USP). The USP acts as the divider between the upperplenum and the reactor vessel head and as a relatively stiff basefor the rest of the upper internals. The upper supportcolumns and the guide tubes are attachedto the USP. The UCP, in turn, is attachedto the upper support columns. The USP assemblies are designatedas one of three different designs: (1) a deep beam design, (2) a top hat design, or (3) an invertedtop hat design.

The UCP is perforatedto permit coolant to passfrom the core below into the upper plenum defined by the USP and the UCP. The coolant then exits through the outlet nozzles in the core barrel. The UCP positions and laterallysupports the core by fuel alignmentpins extending below the plate. The UCP contacts andpreloads thefuel assembly springs and thus maintains contact of the fuel assemblies with the lower core plate (LCP)during reactoroperation.

The upper support columns verticallyposition the UCP and are designedto take the uplifting hydraulicflow loads andfuel spring loads on the UCP. The guide tubes are bolted to the USP andpinnedat the UCPso they can be easily removed if replacement is desired. The guide tubes are designed to guide the control rods in and out of thefuel assemblies to control power generation. The guide tubes are also slotted in their lower sections to allow coolant exitingfrom the core toflow into the upperplenum.

The upper instrumentationcolumns are bolted to the USP. These columns supportthe thermocouple guide tubes that lead the thermocouplesfrom the reactorhead into the upperplenum to just above the UCP.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program The UCPalignmentpins locate the UCP laterally with respect to the lower internals assembly. The pins must laterally support the UCP so that the plate is free to expand radially and move axially during differentialthermal expansions between the upper internals and the core barrel.The UCP alignmentpins are the interfacing components between the UCPand the core barrel.The UCP alignmentpins are shrunk-fit and welded into the core barreland the core barrelbearingpad. The gap sizes between the alignmentpins and the matching inserts are customized.

The USP, the upper support columns, and the UCP are typically consideredcore support structures.

Lower InternalsAssembly The reactorcore is positionedand supported by the lower internals and upper internals assemblies. The individualfuel assemblies are positionedby fuel alignmentpins in the LCP and in the UCP. These pins control the orientationof the core with respect to the lower internals and upper internals assemblies. The lower internalsare alignedwith the upper internals by the UCPalignmentpins andsecondarily by the head/vessel alignment pins. The lower internalsare orientatedto the vessel by the lower radialkeys and by the head/vessel alignmentpins. Thus, the core is aligned with the vessel by a number of interfacingcomponents.

The lower internals assembly is supported in the vessel by clamping to a ledge below the vessel-head mating surface and closely guided at the bottom by radialsupport/clevis assemblies. The upper internalsassembly is clamped at this same ledge by the reactor vessel head. The bottom of the upper internals assembly is closely guidedby the core barrelalignmentpins of the lower internalsassembly.

The fuel assemblies are supportedinside the lower internals assembly on top of the LCP.

The LCP is elevated above the lower supportforging by support columns and bolted to a ring supportattachedto the inside diameter of the core barrel.The support columns transmitverticalfuel assembly loadsfrom the LCP to the much thicker lower support forging. The lower supportforging is welded to and supportedby the core barrel,which transmitsvertical loads to the vessel through the core barrelflange.

The functions of the LCP are to position and support the core andprovide a metered control of reactorcoolantflow into eachfuel assembly. The LCP is located near the bottom of the lower support assembly, inside the core barrel,andabove the lower supportforging.

The function of the lower supportforging or casting is to provide supportfor the core.

The lower supportforging is attachedwith afull-penetrationweld to the lower end of the core barrel.In this position it can provide uninterruptedsupport to the core. The core sits directly on the LCP, which is supported by the lower support columns that are attachedto and extend above the lower supportforging. Some four-loopplants employ a Page 11 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program cast lower support instead of aforging. The functions, loads, andsupportinghardware are the same except for dimensions.

The primaryfunction of the core barrelis to support the core. A large number of components are attachedto either the core barrelor the core barrelflange, including the baffle/former assembly, the outlet nozzles, the neutron panel assemblies or thermal shield, the alignmentpins that engage the UCP and the LCP, the lower supportforging, and the LCP. The radialkeys restrain large transverse motions of the core barrelbut at the same time allow unrestrictedradialand axial thermal expansions.

The baffle andformer assembly is made up of verticalplates called baffles and horizontal supportplates calledformers. The baffle plates are bolted to the formers by the baffle/former bolts, and the formers are attachedto the core barrelinside diameter by the barrel/formerbolts. The baffle/former assemblyforms the interface between the core and the core barrel.The baffles provide a barrierbetween the core and the former region so that a high concentration offlow in the core region can be maintained A secondary benefit, although not a requirementof the baffles, is to reduce the neutronflux on the vessel.

Baffle plates are secured to each other at selected corners by edge bolts. In addition, in some installations,corner brackets are installedbehind and bolted to the baffle plates.

The function of the core barreloutlet nozzles is to direct the reactorcoolant, after it leaves the core, radially outward through the reactorvessel outlet nozzles. The core barreloutlet nozzles are located in the upperportion of the core barreldirectly below the flange and are attachedto the core barrel,each in line with a vessel outlet nozzle.

Additional neutron shielding of the reactorvessel is provided in the active core region by neutron panels or thermal shields that are attachedto the outside of the core barrel.

Specimen guides that contain specimensfor determining the irradiationeffects of the vessel during the life of the plant are attachedto the neutron panels/thermalshields.

The flux thimble is a long, slender stainlesssteel tube that passesfrom an external seal table, through the bottom mounted nozzle penetration,through the lower internals assembly, andfinally extends to the top of the fuel assembly. The flux thimble provides a pathfor the neutronflux detector into the core and is subjected to reactorcoolant pressureand temperatureon the outside surface and to atmospheric conditions on the inside. The flux thimble pathfrom the seal table to the bottom mounted nozzles is defined by flux thimble guide tubes, which arepart of the primarypressure boundary and not considered to be part of the internals. The bottom-mounted instrumentation(BMJ) columns provide a pathfor theflux thimbles from the bottom of the vessel into the core.

The BMI columns align the flux thimble path with instrumentationthimbles in thefuel assembly.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program The LCP and thefuel alignmentpins, the lower supportforging or casting,the lower support columns, the core barrel,the core barrelflange, the radialsupport keys, the baffle plates, and theformer plates are typically classifiedas core supportstructures.

2.1 Unit 1 OperatingExperience and Records Search A review of plant records was performed to locate unit specific RVI OE and design changes.

Some examples of the results for Unit 1 are discussed in the following sections.

2.1.1 ControlRod Guide Tube and Split Pin Replacement The original split pins were fabricated from alloy X-750. These pins were replaced with an improved stress design fabricated from alloy X-750 in 1985 (U 1C9). Installation was expedited through replacement of control rod guide tube (CRGT) assemblies using spares available from a CNP Unit 2 modification discussed in Section 2.2.1.

2.1.2 Barrel-FormerBolt Inspection and PartialReplacement A barrel-former bolt was discovered on the lower core plate after defueling in 1994 (U1C14).

Inspection was performed in 1995 (U 1C 15) to determine the origin of the retrieved bolt. All barrel-former bolts were visually inspected and a sample of bolts was mechanically agitated to determine if additional bolts were loose. The two horizontally adjacent bolts to the vacant location were loose. A total of three bolts were replaced with oversized bolts. Replacement efforts required three holes to be machined into the core barrel for tool access. This work was completed in 1997 (U1C16).

2.1.3 Clevis InsertBolt Degradation Indications were discovered in the clevis bolts of the lower radial support system while performing the ASME 10-year ISI in 2010 (U1C23). The plant is currently operating based on a JCO provided by the OEM. Development of a repair methodology and associated tooling is under development.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 2.2 Unit 2 OperatingExperience and Records Search A review of plant records was performed to locate unit specific RVI OE and design changes.

Some examples of the results for Unit 2 are listed in the following sections.

2.2.1 Modificationfrom 15X15 to 17X1 7 Design Modifications were made on Unit 2 to accept 17X 17 fuel assemblies. Appropriate changes were made to the RVIs at the manufacturer's shop prior to operation. Spare parts generated from the modification, including the 15X15 CRGT assemblies, were retained by I&M. These spare CRGT assemblies were later installed in Unit I as described in Section 2.1.1.

2.2.2 ControlRod Guide Tube Split Pin Replacement The original Unit 2 split pins were fabricated from alloy X-750. A small number of these pins failed and were retrieved from two steam generators in 1985. A JCO was provided by the OEM to operate for the remainder of the fuel cycle with this known degraded condition. The original pins were replaced during the following RFO with an improved stress design fabricated from alloy X-750. This work was completed in 1986 (U2C6).

2.2.3 ControlRod Guide Tube Cap Screw Modification Each CRGT in Unit 2 has four hold down socket head cap screws fastening it to the support plate. A number of bolts and threaded holes were damaged during the Unit 2 CRGT split pin replacement campaign. Two CRGT hold down socket head cap screws were broken during untorquing, leaving the threaded portion of the cap screws in the tapped support plate holes.

Also, the threads of two tapped holes were damaged during the split pin replacement effort.

Each damaged location was on a different CRGT assembly. These bolt locations were abandoned as supported by analysis from the OEM. Three high strength bolts were installed at the remaining available locations on these four CRGT assemblies. This work was completed in 1986 (U2C6).

2.2.4 Baffle-FormerBolt PartialReplacement CNP Unit 2 original baffle-former bolts are internal hex with a cross tack welded lock bar. A number of baffle-former bolts were discovered on the lower core plate in 2010 (U2C 19). Visual inspection revealed 18 failed bolts ranging from broken or missing lock bars to broken or missing bolt heads in a local area on the large south baffle plate. Bolts with visual indications were replaced. Replacement was expanded to bolts in adjacent rows and columns in the plate to bound the edge of the local degradation. Bolt samples were removed from the other three large baffle plates and inspected to ensure degradation was not occurring at symmetric locations. A total of 52 bolts were replaced with two locations left vacant.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 3.0 EXISTING PROGRAMS AND ACTIVITIES There are a number of programs that support the CNP RVI AMP. These existing and ongoing programs are an integral part of the aging management strategy at CNP.

3.1 ASME Section XI ISI Program The ASME Section XI program monitors for aging effects in reactor vessel internals through periodic inspections. Results are dispositioned in accordance with the appropriate acceptance criteria provided in the code. This program has been effective at identifying and managing aging in the reactor vessel internals.

3.2 PrimaryStrategic Water ChemistryPlan The Primary Strategic Water Chemistry Plan is used to control water chemistry to minimize or eliminate material degradation due to contaminants. This program monitors chemistry and maintains concentrations within the system-specific tolerance. The program follows the guidance provided in the EPRI PWR Water Chemistry Guidelines. This program has been effective for controlling water chemistry to minimize material degradation.

3.3 Thimble Tube Multifrequency Eddy CurrentInspection The Incore Instrumentation Thimble Tube Multifrequency Eddy Current Inspection program periodically inspects for thimble tube wear in accordance with NRC Bulletin 88-09. This program has been effective in identifying loss of material due to wear prior to leakage. This has allowed pre-emptive corrective actions to maintain proactive management of the thimble tubes.

3.4 Industry Involvement CNP participates in industry activities including the Pressurized Water Reactor owner's Group and the Electric Power Research Institute Materials Reliability Program. Participation includes attending meetings, providing input to work products, and implementing work products as applicable and appropriate for CNP.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 4.0 INSPECTION & EVALUATION GUIDELINES I&M's strategy for managing aging effects of reactor vessel internals at CNP includes performing augmented inspections as described in guidance provided by the EPRI MRP. The MRP inspection & evaluation guidelines for managing the effects of aging on PWR internals are documented in MRP-227-A. These guidelines do not reduce, alter, or otherwise affect current ASME B&PV Code Section XI or plant-specific licensing inservice inspection requirements.

The MRP developed the companion document MRP-228 which contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE systems used to perform those inspections.

All PWR internals were placed into four functional groups. The following is an excerpt from MRP-227-A, Section 3.3.1.

"Primary:those PWR internals that are highly susceptible to the effects of at least one of the eight aging mechanisms were placed in the Primarygroup. The aging management requirements that are needed to ensurefunctionality of Primarycomponents are described in these I&E guidelines. The Primarygroup also includes components which have shown a degree of tolerance to a specific aging degradationeffect, butfor which no highly susceptible component exists orfor which no highly susceptible component is accessible.

Expansion: those PWR internals that are highly or moderately susceptible to the effects of at least one of the eight agingmechanisms, butfor which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group.

The schedule for implementation of agingmanagement requirementsfor Expansion components will depend on thefindingsfrom the examinationsof the Primary components at individualplants.

" Existing Programs:those PWR internals that are susceptible to the effects of at least one of the eight aging mechanisms andfor which generic andplant-specific existing AMP elements are capable of managing those effects, were placed in the Existing Programsgroup.

"No Additional Measures: those PWR internalsfor which the effects of all eight aging mechanisms are below the screening criteriawere placed in the No Additional Measuresgroup. Additional components were placed in the No Additional Measures group as a result of FMECA andthe functionality assessment. No further action is requiredby these guidelinesfor managing the aging of the No Additional Measures components.

The categorizationand analysisprocesses describedherein are not intended to supersede any ASME B&PV Code Section X1 [2] requirements.Any components that are classified as core supportstructures as defined in ASME B&PV Code Section XI JWB 2500 IWA 9000, and listed in Table IWB 2500-1. Category B-N-3 [2] have requirementsthat remain in effect and may only be alteredas allowed by 10CFR50.55a[4].

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 4.1 Guideline Background The first EPRI MRP guidance for PWR RVI AMPs was published in December 2008 as MRP-227, Revision 0. EPRI submitted the report for NRC staff review and approval in January 2009.

The NRC issued the final SE, Revision 0, for MRP-227, Revision 0 in June 2011. Revision I to the SE on MRP-227, Revision 0 was issued in December 2011 and included in MRP-227-A published in December 2011.

4.2 . MRP-227-A Applicability to CNP There are three general assumptions used in the MRP-227-A.

4.2.1 GeneralAssumption 1 The following is the first general assumption from Section 2.4 of MRP-227-A:

30 years of operation with high leakage core loadingpatterns (freshfuel assemblies loaded in peripherallocations)followed by implementation of a low-leakagefuel management strategyfor the remaining 30 years of operation.

Both CNP units changed from a high leakage to a low leakage core pattern prior to 30 years of operation. It is more conservative to operate with a low leakage core pattern than a high leakage core pattern. Therefore, both CNP units are bounded by this assumption.

4.2.2 GeneralAssumption 2 The following is the second general assumption from Section 2.4 of MRP-227-A:

Base load operation, i.e., typically operates at fixedpower levels anddoes not usually vary power on a calendaror load demand schedule.

Both CNP units are base load plants which operate at a fixed power level and do not vary power on a calendar or load demand schedule. Therefore, both CNP units are bounded by this assumption.

4.2.3 GeneralAssumption 3 The following is the third general assumption from Section 2.4 of MRP-227-A:

No design changes beyond those identified in general industry guidance or recommended by the original vendors.

All U.S. PWR operating plants met this assumption as of May 2007 for the three designs identified in MRP-227-A. Each CNP unit is discussed individually in the following sections.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 4.2.3.1 Unit 1 Applicability No modifications have been made to CNP Unit 1 RVIs since May 2007. Therefore CNP Unit 1 is bounded by this assumption.

Degraded bolts and a degraded dowel pin were discovered in the LRSS clevis inserts in the RV during the 2010 RFO (U1C23). The unit is operating with these degraded bolts and degraded dowel pin on an interim analysis while I&M prepares for a repair. Section 2.1.3 discusses this topic in further detail. I&M has engaged the OEM for analysis in support of a repair. CNP Unit I will continue to be bounded by this assumption following repair.

4.2.3.2 Unit 2 Applicability Baffle-former bolt degradation was observed and repaired during the 2010 RFO (U2C 19).

Section 2.2.4 discusses this topic in further detail. Repair was performed by the OEM. No other modifications have been made to the CNP Unit 2 RVIs. Therefore, CNP Unit 2 is bounded by this assumption.

4.3 NEI 03-08 Guidancein MRP-227-A There are one "Mandatory", five "Needed", and zero "Good Practice" elements identified in MRP-227-A under the NEI-03-08 implementation protocol. These elements are discussed in the following sections.

4.3.1 NEI 03-08 Mandatory The following is the NEI 03-08 "mandatory" element from Section 7.2 of MRP-227-A:

Each commercial U.S. PWR unit shall develop and document a programfor management of aging of reactorinternalcomponents within thirty-six months following issuance of MRP-22 7-Rev. 0 (that is, no later than December31, 2011).

This mandatory element requires that a program for management of aging of reactor internal components is developed by December 31, 2011. WCAP-17300, "Reactor Vessel Internals Program Plan for Aging Management of Reactor Internals at D.C. Cook Nuclear Plant Unit 1" and WCAP- 17301, "Reactor Vessel Internals Program Plan for Aging Management of Reactor Internals at D.C. Cook Nuclear Plant Unit 2" were completed in February 2011. These documents are superseded by this document. This element is fulfilled for both CNP units.

4.3.2 NEI 03-08 Needed There are five NEI 03-08 "needed" elements in MRP-227-A. These are addressed in the following sections.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 4.3.2.1 Needed Element 1 The following is the NEI 03-08 "needed" element from Section 7.3 of MRP-227-A:

Each commercial U.S. PWR unit shall implement Tables 4-1 through 4-9 and Tables 5-1 through 5-3for the applicable design within twenty-four monthsfollowing issuance of MRP-22 7-A.

MRP-227-A was issued in December 2011 making implementation of the applicable tables needed by December 2013. The applicable Westinghouse tables contained in MRP-227-A are Table 4-3 for primary components, Table 4-6 for expansion components, Table 4-9 for existing programs, and Table 5-3 for acceptance and expansion criteria. These tables are included as Appendix A, Appendix B, Appendix C, and Appendix D respectively. The CNP RVI AMP implements these tables. This element is fulfilled for both CNP units.

4.3.2.2 Needed Element 2 The following is the NEI 03-08 "needed" element from Section 7.4 of MRP-227-A:

Examinationsspecified in these guidelines shall be conducted in accordancewith the Inspection Standard(MRP-228).

MRP-228 is the companion document to MRP-227-A. Internals examinations conducted as specified in MRP-227-A will be in accordance with MRP-228. This element is fulfilled for both CNP units.

4.3.2.3 Needed Element 3 The following is the NEI 03-08 "needed" element from Section 7.5 of MRP-227-A:

Examinationresults that do not meet the examination acceptance criteriadefined in Section 5 of these guidelinesshall be recordedand entered in the plant correctiveaction program and dispositioned.

Conditions are documented and dispositioned in accordance with the Corrective Action Program.

This element is fulfilled for both CNP units.

4.3.2.4 Needed Element 4 The following is the NEI 03-08 "needed" element from Section 7.6 of MRP-227-A:

Each commercial U.S. PWR unit shallprovide a summary reportof all inspections and monitoring,items requiringevaluation, and new repairsto the MAP ProgramManager within 120 days of the completion of an outage during which PWR internals within the scope of A4RP-227 are examined.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program I&M will provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs within the scope of MRP-227-A to the MRP Program Manager within 120 days of the completion of an outage during which CNP PWR internals within the scope of MRP-227-A are examined. This element is fulfilled for both CNP units.

4.3.2.5 Needed Element 5 The following is the NEI 03-08 "needed" element from Section 7.7 of MRP-227-A:

If an engineeringevaluation is used to disposition an examination result that does not meet the examinationacceptance criteriain Section 5, this engineeringevaluationshall be conducted in accordancewith a NRC-approved evaluation methodology.

Inspection results from MRP-227-A inspections that do not meet the acceptance criteria will be dispositioned in accordance with an NRC approved methodology, or the methodology will be submitted for NRC approval prior to implementation. This element is fulfilled for both CNP units.

4.4 Safety EvaluationReport Conditions and Limitations There are a number of conditions and limitations as described in "Revision 1 to the Safety Evaluation by the Office of Nuclear Reactor Regulation Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision

0) Project No. 669." (MRP-227 SE). These include seven Topical Report Conditions and eight Applicant/Licensee Action Items.

4.4.1 Topical Report Conditions The topical report conditions contained in the MRP-227 SE were incorporated into MRP-227-A.

The CNP RVI AMP is consistent with MRP-227-A. Therefore all topical report conditions are fulfilled for both CNP units.

4.4.2 Applicant/LicenseeAction Items The Applicant/Licensee Action Items contained in Revision 1 to the SE on MRP-227, Revision 0 are discussed in the following sections.

4.4.2.1 Applicant/LicenseeAction Item 1 (MRP-227 SE Sections 3.2.5.1 and 4.2.1)

"Applicability of FMECA and Functionality Analysis Assumptions" from Section 4.2.1 of the MRP-227 SE:

As addressedin Section 3.2.5.1 of this SE, each applicant/licenseeis responsiblefor assessingits plant's design and operatinghistory anddemonstratingthat the approved version of MRP-227 is applicable to the facility. Each applicant/licenseeshall refer, in particular,to the assumptions regardingplant design and operatinghistory made in the Page 20 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program FMECA andfunctionality analysesfor reactors of their design (i.e., Westinghouse, CE, or B& W) which support MRP-227 and describe the process usedfor determiningplant-specific differences in the design of their R VI components or plant operatingconditions, which result in different component inspection categories. The applicant/licenseeshall submit this evaluationfor NRC review and approvalas partof its applicationto implement the approved version of MRP-22 7.

This Action Item addresses the applicability of the FMECA and functionality analysis assumptions made in the development of MRP-227-A to individual facilities. I&M is participating in PWROG project PA-MSC-0938, "Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0" to address this item. The results of the evaluation will be provided to the NRC prior to the period of extended operation for each unit.

4.4.2.2 Applicant/Licensee Action Item 2 (MRP-227 SE Sections 3.2.5.2 and 4.2.2)

"PWR Vessel Internal Components Within the Scope of License Renewal" from Section 4.2.2 of the MRP-227 SE:

As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressedin 10 CFR 54.4, each applicant/licenseeis responsiblefor identifying which R VI components are within the scope of LR for itsfacility. Applicants/licenseesshall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the R VI components that are within the scope of LR for theirfacilities in accordancewith 10 CFR 54.4. If the tables do not identify all the R VI components that are within the scope of LR for itsfacility, the applicantor licensee shall identify the missing component(s) andpropose any necessary modifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specificAMP. The AMP shallprovide assurancethat the effects of aging on the missing component(s) will be managedfor the periodof extended operation.

This Action Item requires the licensee to verify that all the RVI components within the scope for license renewal at that facility have been considered in applicable documents in development of MRP-227-A. I&M is participating in PWROG project PA-MSC-0938, "Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0" to address this item. The results of the evaluation will be provided to the NRC prior to the period of extended operation for each unit.

4.4.2.3 Applicant/LicenseeAction Item 3 (MRP-22 7 SE Sections 3.2.5.3 and 4.2.3)

"Evaluation of the Adequacy of Plant-Specific Existing Programs" from Section 4.2.3 of the MRP-227 SE:

As addressedin Section 3.2.5.3 in this SE, applicants/licenseesof CE and Westinghouse are requiredto perform plant-specific analysis either to justify the acceptabilityof an applicant's/licensee'sexisting programs,or to identify changes to the programsthat Page 21 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program should be implemented to manage the aging of these componentsfor the periodof extended operation. The results of this plant-specific analyses anda description of the plant-specificprogramsbeing reliedon to manage aging of these components shall be submitted as part of the applicant's/licensee'sAMP application.The CE and Westinghouse components identifiedfor this type ofplant-specific evaluationinclude: CE thermalshieldpositioningpins and CE in-core instrumentationthimble tubes (Section 4.3.2 in MRP-227), and Westinghouse guide tube supportpins (splitpins) (Section 4.3.3 in MRP-22 7).

CNP Unit 1 and Unit 2 both have X-750 split pins. Project requests have been initiated to investigate split pin replacement for each unit. I&M will provide the NRC with the strategy for managing split pins prior to the period of extended operation for each unit.

4.4.2.4 Applicant/Licensee Action Item 4 (MRP-227 SE Sections 3.2.5.4 and 4.2.4)

"B&W Core Support Structure Upper Flange Stress Relief' from Section 4.2.4 of the MRP-227 SE:

As discussed in Section 3.2.5.4 of this SE, the B& W applicants/licenseesshall confirm that the core supportstructureupperflange weld was stress relieved during the original fabricationof the Reactor Pressure Vessel in order to confirm the applicabilityof MRP-227, as approvedby the NRC, to theirfacility. If the upperflange weld has not been stress relieved, then this component shall be inspected as a "Primary"inspection category component. If necessary, the examinationmethods andfrequencyfor non-stress relieved B& W core supportstructureupperflange welds shall be consistent with the recommendations in MRP-227, as approvedby the NRC, for the Westinghouse and CE upper core support barrelwelds. The examination coveragefor this B& Wflange weld shall conform to the staff's imposed criteriaas describedin Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee'sresolution of this plant-specific action item shall be submitted to the NRCfor review and approval.

This item is specific to the Babcock & Wilcox designed plant and it is not applicable to CNP.

No action is required.

4.4.2.5 Applicant/LicenseeAction Item 5 (MRP-22 7 SE Sections 3.3.5 and 4.2.5)

"Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components" from Section 4.2.5 of the MRP-227 SE:

As addressedin Section 3.3.5 in this SE, applicants/licenseesshall identify plant-specific acceptance criteriato be applied when performing the physical measurements required by the NRC-approved version of MRP-22 7for loss of compressibilityfor Westinghouse hold down springs, andfor distortion in the gap between the top and bottom core shroud segments in CE units with core barrelshrouds assembled in two vertical sections. The applicant/licenseeshall include its proposedacceptance criteriaand an explanationof how the proposedacceptance criteriaare consistent with the plants' licensing basis and Page 22 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program the need to maintain the functionality of the component being inspectedunder all licensing basis conditions of operationduring the period of extended operation as part of their submittal to apply the approvedversion of MRP-22 7.

CNP Unit 1 and Unit 2 both have 304 SS hold down springs. MRP-227-A guidance includes physical measurement of 304 SS hold down springs. This action item requires acceptance criteria to be provided to the NRC. CNP plant specific acceptance criteria will be developed and submitted to the NRC prior to the first required physical measurement. The hold down springs will be replaced if acceptance criteria are not developed in lieu of performing the first required physical measurement.

4.4.2.6 Applicant/LicenseeAction Item 6 (MRP-227 SE Sections 3.3.6 and 4.2.6)

"Evaluation of Inaccessible B&W Components" from Section 4.2.6 of the MRP-227 SE:

As addressedin Section 3.3.6 in this SE, MRP-227 does not propose to inspect the following inaccessiblecomponents: the B& W core barrelcylinders (including vertical andcircumferentialseam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices, B& W core barrel-to-formerbolts and their locking devices, andB& W core barrelassembly internal baffle-to-baffle bolts. The MRP also identified that although the B& W core barrelassembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectableusing currently available examination techniques.

Applicants/licensees shalljustify the acceptabilityof these components for continued operation through the period of extended operation by performing an evaluation, or by proposinga scheduled replacement of the components. As part of their applicationto implement the approvedversion of MRP-22 7, applicants/licenseesshallprovide their justificationfor the continuedoperability of each of the inaccessiblecomponents and,if necessary,provide theirplanfor the replacementof the componentsfor NRC review and approval.

This item is specific to the Babcock & Wilcox designed plant and it is not applicable to CNP.

No action is required.

4.4.2.7 Applicant/LicenseeAction Item 7 (MRP-227 SE Sections 3.3.7 and 4.2.7)

Section 4.2.7, "Plant-Specific Evaluation of CASS Materials" from the MRP-227 SE:

As discussed in Section 3.3.7 of this SE, the applicants/licenseesof B& W, CE, and Westinghouse reactors are requiredto develop plant-specific analyses to be appliedfor theirfacilities to demonstrate that B& WIMI guide tube assembly spiders and CRGT spacercastings, CE lower supportcolumns, and Westinghouse lower supportcolumn bodies will maintain theirfunctionality during the periodof extended operation orfor additionalR VI components that may befabricatedfrom CASS, martensiticstainless steel or precipitationhardenedstainless steel materials.These analyses shall also consider the Page 23 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program possible loss offracture toughness in these components due to thermaland irradiation embrittlement, and may also need to consider limitations on accessibilityfor inspection and the resolution/sensitivityof the inspection techniques. The requirementmay not apply to components that were previously evaluatedas not requiringaging management during development of MRP-22 7. That is, the requirement would apply to componentsfabricated from susceptible materialsfor which an individual licensee has determined aging management is required,for example during their review performed in accordancewith Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain thefunctionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licenseeshall include the plant-specificanalysis as part of their submittal to apply the approved version of MRP-22 7.

A plant specific evaluation of RVI CASS materials is required in this Action Item. I&M is participating in PWROG project PA-MSC-0938, "Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0" to address this item. The results of the evaluation will be provided to the NRC prior to the period of extended operation for each unit.

4.4.2.8 Applicant/LicenseeAction Item 8 (MRP-227 SE Sections 3.5.1 and4.2.8)

Section 3.5.1, "Submittal of Information for Staff Review and Approval" from the MRP-227 SE:

In addition to the implementation of MRP-227 in accordancewith NEI 03-08, applicants/licenseeswhose licensing basis contains a commitment to submit a PWR R VI AMP and/or inspectionprogram shall also make a submittalfor NRC review and approvalto credit their implementation of MRP-227, as amended by this SE. An applicant's/licensee'sapplicationto implement MRP-22 7, as amended by this SE shall include thefollowing items (1) and (2). Applicants who submit applicationsfor LR after the issuance of this SE shall, in accordancewith the NUREG-1801, Revision 2, submit the informationprovided in thefollowing items (1) through (5)for staffreview and approval.

This Action Item includes five parts. However, parts 3-5 are only applicable to licensees who submit license renewal applications after the issuance of the MRP-227 SE. I&M submitted CNP LRA in October 2003, the MRP-227 SE was issued in December 2011. Therefore, parts 3-5 are not applicable to CNP. Action Item 8, parts 1 and 2 are discussed in the following sections.

4.4.2.8.1 GALL Revision 2 Requirement Section 3.5.1, Part I from the MRP-227 SE:

An AMP for the facility that addressesthe 10 program elements as defined in NUREG-1801, Revision 2, AMP XI. M16A.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program The CNP RVI AMP addresses the ten program elements as defined in NUREG- 1801, Revision 2, AMP XI.M16A. This item is addressed in detail in Section 5.0 of this document. Therefore, part 1 of this Action Item is fulfilled for both units.

4.4.2.8.2 RVI AMP Submittal Requirements Section 3.5.1, Part 2 from the MRP-227 SE:

To ensure the MRP-22 7 programand the plant-specific action items will be carriedout by applicants/licensees,applicants/licenseesare to submit an inspectionplan which addressesthe identifiedplant-specificaction itemsfor staff review and approval consistent with the licensing basisfor the plant. If an applicant/licenseeplans to implement an AMP which deviatesfrom the guidanceprovided in MRP-227, as approved by the NRC, the applicant/licenseeshall identify where theirprogram deviatesfrom the recommendationsof MRP-227, as approved by the NRC, and shallprovide ajustification for any deviation which includes a considerationof how the deviation affects both "Primary"and "Expansion" inspection category components.

The CNP RVI AMP addresses the plant-specific action items. The CNP RVI AMP does not deviate from MRP-227-A. Therefore, part 2 of this Action Item is fulfilled for both units upon submittal of the CNP RVI AMP to the NRC for review and approval.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 5.0 PROGRAM ATTRIBUTE EVALUATION The CNP RVI AMP addresses the 10 program elements as defined in NUREG-1801, Revision 2,Section XI.M16A. This is in accordance with Applicant/Licensee Action Item 8, part 1 from the MRP-227 SE.

The following is the program description from NUREG-1801, Revision 2,Section XI.M16A:

ProgramDescription This program relies on implementation of the Electric Power Research Institute (EPRI)

Report No. 1016596 (MRP-227) and EPRI Report No. 1016609 (MRP-228) to manage the aging effects on the reactorvessel internal (R VI) components.

This program is used to manage the effects of age-relateddegradationmechanisms that are applicable in general to the PWR RVI components at the facility. These aging effects include (a) variousforms of cracking, including stress corrosion cracking (SCC), which also encompasses primary water stress corrosioncracking (PWSCC), irradiation assistedstress corrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; (b) loss of materialinduced by wear; (c) loss offracture toughness due to either thermal aging or neutron irradiationembrittlement; (d) changes in dimension due to void swelling; and (e) loss ofpreloaddue to thermal and irradiation-enhancedstress relaxation or creep.

The program applies the guidance in MRP-227for inspecting, evaluating, and, if applicable, dispositioningnon-conformingR VI components at the facility. The program conforms to the definition of a sampling-basedcondition monitoringprogram,as defined by the Branch TechnicalPositionRSLB-1, with periodic examinations and other inspections of highly-affected internals locations. These examinationsprovide reasonableassurancethat the effects of age-relateddegradationmechanisms will be managedduring the periodof extended operation. The program includes expanding periodic examinations andother inspections if the extent of the degradationeffects exceeds the expected levels.

The MRP-227 guidancefor selecting R VI components for inclusion in the inspection sample is based on a four step rankingprocess. Through this process, the reactor internalsfor all three PWR designs were assignedto one of the followingfour groups:

Primary,Expansion,Existing Programs,andNo Additional Measures components.

Definitions of each group areprovided in GALL ChapterIX.B.

The result of thisfour-step sample selectionprocess is a set of PrimaryInternals Component locationsfor each of the three plant designs that are expected to show the leading indicationsof the degradation effects, with anotherset of Expansion Internals Component locations that are specified to expand the sample should the indications be more severe than anticipated.The degradationeffects in a third set of internals locations are deemed to be adequatelymanagedby Existing Programs,such as ASME Code,Section XI, 1I Examination Category B-N-3 examinations of core support structures.A Page 26 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program fourth set of internals locations are deemed to require no additionalmeasures.As a result, the program typically identifies 5 to 15% of the R VI locations as Primary Component locationsfor inspections, with another 7 to 10% of the R VI locations to be inspected as Expansion Components, as warrantedby the evaluation of the inspection results. Another 5 to 15% of the internalslocations are covered by Existing Programs, with the remainderrequiringno additionalmeasures. This process thus uses appropriate componentfunctionality criteria,age relateddegradationsusceptibility criteria,and failure consequence criteriato identify the components that will be inspected under the program in a manner that conforms to the sampling criteriafor sampling-based condition monitoringprograms in Section A. 1.2.3.4 of NRC Branch PositionRLSB-i.

Consequently, the sample selection process is adequate to assure that the intended function(s) of the PWR reactorinternalcomponents are maintainedduring the periodof extended operation.

The program's use of visual examination methods in MRP-227for detection of relevant conditions (andthe absence of relevant conditions as a visual examination acceptance criterion) is consistent with the ASME Code,Section XI rulesfor visual examination.

However, the program's adoption of the MRP-22 7 guidancefor visual examinations goes beyond the ASME Code,Section XI visual examination criteriabecause additional guidance is incorporatedinto MRP-22 7 to clarify how the particularvisual examination methods will be used to detect relevant conditions and describes in more detail how the visual techniques relate to the specific R VI components and how to detect their applicable age-relateddegradationeffects.

The technical basisfor detecting relevant conditions using volumetric ultrasonic testing (UT) inspection techniques can be found in MRP-228, where the review of existing bolting UT examination technicaljustifications has demonstratedthe indicationdetection capability of at least two vendors, and where vendor technicaljustificationis a requirementprior to any additionalbolting examinations.Specifically, the capability of program's UT volumetric methods to detect loss of integrity ofPWR internals bolts, pins, andfasteners, such as baffle-former bolting in B& W and Westinghouse units, has been well demonstratedby operatingexperience. In addition,the program 's adoption of the MRP-22 7 guidanceandprocess incorporatesthe UT criteria in MRP-228, which calls for the technicaljustifications that are neededfor volumetric examination method demonstrations,requiredby the ASME Code,Section V.

The program also includesfuture industry operatingexperience as incorporatedin periodic revisions to MRP-22 7. The program thusprovides reasonableassurancefor the long-term integrity and safe operation of reactorinternals in all commercial operating U.S, PWR nuclearpower plants.

Age-related degradationin the reactorinternals is managedthrough an integrated program. Specific features of the integratedprogram are listed in the following ten program elements. Degradationdue to changes in materialproperties (e.g., loss of fracture toughness) was considered in the determinationof inspection recommendations and is managed by the requirement to use appropriatelydegradedproperties in the Page 27 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program evaluation of identifieddefects. The integratedprogram is implemented by the applicant through an inspectionplan that is submitted to the NRCfor review and approvalwith the applicationfor license renewal.

11 Refer to the GALL Report, Chapter1, for applicabilityof variouseditions of the ASME Code, Section X1.

5.1 Scope of Program The following is NUREG-1801, Revision 2,Section XI.M16A, Evaluation and Technical Basis Element 1:

The scope of the program includes all R VI components at the Donald C. Cook Nuclear Plant Unitl and Unit 2, which are built to a Westinghouse NSSS design. The scope of the program applies the methodology andguidance in the most recently NRC endorsed version of MRP-22 7, which provides augmented inspectionandflaw evaluation methodologyfor assuringthefunctional integrity of safety-related internalsin commercial operating U.S. PWR nuclearpower plants designed by B& W, CE, and Westinghouse. The scope of components consideredfor inspection under MRP-227 guidance includes core support structures (typically denotedas Examination Category B-N-3 by the ASME Code,Section XI), those R V1 components that serve an intended license renewal safetyfunction pursuant to criteriain 10 CFR 54.4(a)(1), andother R VI components whose failure could prevent satisfactory accomplishment of any of the functions identifiedin 10 CFR 54.4 (a)(1) (i), (ii), or (iii). The scope of the programdoes not include consumable items, such asfuel assemblies, reactivity control assemblies, and nuclear instrumentation,because these components are not typically within the scope of the components that are requiredto be subject to an aging management review (AMR),

as defined by the criteriaset in 10 CFR 54.21(a)(1). The scope of the program also does not include welded attachments to the internalsurface of the reactor vessel because these components are considered to be ASME Code Class I appurtenancesto the reactor vessel and are adequatelymanaged in accordancewith an applicant'sAMP that corresponds to GALL AMP XI.M1, "ASME Code, Section XlInservice Inspection, Subsections IWB, IWC, andIWD. "

The scope of the program includes the response bases to applicable license renewal applicantaction items (LRAAIs) on the MRP-227 methodology, and any additional programs,actions, or activities that are discussed in these LRAAI responses and credited for aging management of the applicant'sR VI components. The LRAAIs are identified in the staff's safety evaluation on MRP-22 7 and include applicable action items on meeting those assumptions thatformed the basis of the MRP's augmentedinspection andflaw evaluation methodology (as discussed in Section 2.4 of MRP-22 7), andNSSS vendor-specific orplant-specific LRAA~s as well. The responses to the LRAAIs on MRP-22 7 are provided in Appendix C of the LRA.

The guidance in MRP-22 7 specifies applicabilitylimitations to base-loadedplants and thefuel loadingmanagement assumptions upon which the functionality analyses were Page 28 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program based. These limitations andassumptions require a determinationof applicabilityby the applicantfor each reactorand are covered in Section 2.4 of MRP-227.

A description of CNP RVIs is provided in Section 2.0. The scope of the CNP RVI AMP applies the methodology and guidance in MRP-227-A. The program does not consider consumable items or welded attachments to the internal surface of the reactor vessel. The applicable licensee action items from the MRP-227 SE are addressed in Section 4.4. The licensee action item responses are in this document rather than the LRA because the LRA was issued prior to the MRP-227 SER. Determination of the applicability of CNP RVIs to the applicability limitations identified in MRP-227-A is addressed in Section 4.2.

The CNP RVI AMP scope is consistent with NUREG-1801, Revision 2,Section XI.Ml6A.

5.2 Preventive Actions The following is NUREG- 1801, Revision 2,Section XI.M l6A, Evaluation and Technical Basis Element 2:

The guidance in MRP-22 7 relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive agingmechanisms (e.g., loss of material induced by general,pitting corrosion, crevice corrosion, or stress corrosion cracking or any of itsforms [SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintainedin accordancewith the Water ChemistryProgram. The program description, evaluation,and technical basis of water chemistry arepresentedin GALL AMP XI.M2, "Water Chemistry."

The CNP Primary Water Chemistry Plan is consistent with NUREG-1801, Revision 0,Section XI.M2. Further details on this program can be found in Section 3.2.

The CNP RVI AMP preventive actions are consistent with NUREG-1801, Revision 2,Section XI.M16.

5.3 ParametersMonitored/Inspected The following is NUREG-1801, Revision 2,Section XI.Ml6A, Evaluation and Technical Basis Element 3:

The program manages thefollowing age-relateddegradation effects and mechanisms that are applicable in general to the R VI components at the facility: (a) cracking induced by SCC, P WSCC, IASCC, orfatigue/cyclical loading; (b) loss of materialinduced by wear; (c) loss offracture toughness induced by either thermalaging or neutron irradiationembrittlement; (d) changes in dimension due to void swelling and irradiation growth, distortion,or deflection; and (e) loss ofpreloadcaused by thermal and irradiationenhanced stress relaxationor creep. For the management of cracking, the programmonitorsfor evidence of surface breaking linear discontinuitiesif a visual inspection technique is used as the non-destructionexamination (NDE) method, orfor Page 29 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program relevantflaw presentationsignals if a volumetric UT method is used as the NDE method.

For the management of loss of material,the program monitorsfor gross or abnormal surface conditions that may be indicative of loss of materialoccurringin the components.

For the management of loss ofpreload, the program monitorsfor gross surface conditions that may be indicative of loosening in applicable bolted,fastened,keyed, or pinned connections. The program does not directly monitorfor loss offracture toughness that is induced by thermal aging or neutron irradiationembrittlement, or by void swelling and irradiationgrowth; instead, the impact of loss offracture toughness on component integrity is indirectlymanaged by using visual or volumetric examination techniques to monitorfor cracking in the components and by applying applicable reducedfracture toughnessproperties in theflaw evaluationsif cracking is detected in the components and is extensive enough to warranta supplementalflaw growth orflaw tolerance evaluation under the MRP-22 7 guidance or ASME Code, Section X1 requirements. The program uses physical measurements to monitorfor any dimensional changes due to void swelling, irradiationgrowth, distortion,or deflection.

Specifically, the program implements the parametersmonitored/inspectedcriteriafor Westinghouse designed PrimaryComponents in Table 4-3 of MRP-227. Additionally, the program implements the parametersmonitored/inspectedcriteriafor Westinghouse designed Expansion Components in Table 4-6 of MRP-227. The parameters monitored/inspectedfor Existing Program Componentsfollow the basesfor referenced Existing Programs,such as the requirementsfor ASME Code Class R V1 components in ASME Code,Section XI, Table IWB-2500-1, Examination CategoriesB-N-3, as implemented through the applicant'sASME Code,Section XI program, or the recommendedprogramfor inspecting Westinghouse designedflux thimble tubes in GALL AMP XI. M3 7, "Flux Thimble Tube Inspection. "No inspections, exceptfor those specified in ASME Code,Section XI, are requiredfor components that are identifiedas requiring "No Additional Measures, " in accordancewith the analyses reportedin MRP-227.

The CNP RVI AMP manages age-related degradation effects including SCC, IASCC, wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling and irradiation growth, and thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep.

These effects are monitored using visual examination, surface examination, volumetric examination, and physical measurements. The program implements Table 4-3 and Table 4-6 from MRP-227-A which are included as Appendix A and Appendix B, respectively, in this document. The program credits the ASME Section XI ISI program described in Section 3.1 which is consistent with NUREG-1801, Revision 0, XI.Ml. The program also credits the thimble tube inspection program which is consistent with the intent of the 10 GALL elements.

The CNP RVI AMP parameters monitored/inspected are consistent with NUREG- 1801, Revision 2,Section XI.M16A.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 5.4 Detection ofAging Effects The following is NUREG-1801, Revision 2,Section XI.Ml6A, Evaluation and Technical Basis Element 4:

The detection of aging effects is covered in two places: (a) the guidance in Section 4 of MRP-22 7 provides an introductorydiscussion andjustificationof the examination methods selectedfor detecting the aging effects of interest; and (b) standardsfor examination methods, procedures, andpersonnelare provided in a companion document, MRP-228. In all cases, well-established methods were selected. These methods include volumetric UT examination methods for detectingflaws in bolting,physical measurementsfor detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) examinationsfor detecting effects rangingfrom general conditions to detection and sizing of surface-breakingdiscontinuities.Surface examinations may also be used as an alternative to visual examinationsfor detection andsizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, andfatigue is monitored/inspectedby either VT-I or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). The VT-3 visual methods may be appliedfor the detection of cracking only when the flaw tolerance of the component or affected assembly, as evaluatedfor reduced fracture toughness properties, is known and has been shown to be tolerantof easily detected largeflaws, even under reducedfracture toughness conditions.In addition, VT-3 examinations are used to monitor/inspectfor loss of materialinduced by wear andfor general aging conditions,such as gross distortioncaused by void swelling and irradiationgrowth or by gross effects of loss ofpreloadcaused by thermal and irradiation-enhancedstress relaxationand creep.

In addition,the program adopts the recommended guidance in MRP-227for defining the Expansion criteriathat need to be appliedto inspections of Primary Components and Existing Requirement Components andfor expandingthe examinations to include additionalExpansion Components. As a result, inspectionsperformed on the R VI components areperformed consistent with the inspectionfrequency and sampling bases for Primary Components, Existing Requirement Components, and Expansion Components in MRP-227, which have been demonstratedto be in conformance with the inspection criteria,sampling basis criteria,andsample Expansion criteriain Section A. 1.2.3.4 of NRC Branch PositionRLSB- 1.

Specifically, the program implements the parametersmonitored/inspectedcriteriaand basesfor inspecting the relevantparameter conditionsfor Westinghouse designed Primary Components in Table 4-3 of MRP-227 andfor Westinghouse designed Expansion Components in Table 4-6 ofMRP-227.

The program is supplemented by the following plant specific Primary Component and Expansion Component inspectionsfor the program (as applicable):there are no supplemental Primaryor Expansion componentsfor the CNPprogram.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program In addition, in some cases (as defined in MRP-227), physical measurements are used as supplemental techniques to managefor the gross effects of wear, loss ofpreloaddue to stress relaxation, orfor changes in dimension due to void swelling, deflection or distortion. The physical measurements methods appliedin accordance with this program include the measurement of the 304 SS hold down springs in accordancewith MRP-22 7 Table 4-3.

The detection of aging effects credited for augmented inspections in the CNP RVI AMP are based on guidance in MRP-227-A and MRP-228. The program implements the guidance of MRP-227-A Table 4-3 and Table 4-6 which are included as Appendix A and Appendix B, respectively. This includes the measurement of the 304 SS hold down springs in each unit in accordance with MRP-227-A Table 4-3.

The CNP RVI AMP detection of aging effects is consistent with NUREG-1801, Revision 2,Section XI.M16A.

5.5 Monitoringand Trending The following is NUREG-1801, Revision 2,Section XI.M16A, Evaluation and Technical Basis Element 5:

The methods for monitoring,recording,evaluating,and trending the data that resultfrom the program's inspections are given in Section 6 of MRP-22 7 and its subsections. The evaluation methods include recommendationsforflaw depth sizing andfor crack growth determinationsas wellfor performing applicable limit load, linearelastic and elastic plasticfractureanalyses of relevantflaw indications. The examinations and re-examinations requiredby the MRP-227 guidance, together with the requirements specified in MRP-228for inspection methodologies, inspectionprocedures, and inspectionpersonnel,provide timely detection, reporting,and corrective actions with respect to the effects of the age-relateddegradationmechanisms within the scope of the program. The extent of the examinations, beginning with the sample of susceptible PWR internals component locations identified as Primary Component locations, with the potentialfor inclusion of Expansion Component locations if the effects are greaterthan anticipated,plus the continuation of the Existing Programsactivities, such as the ASME Code, Section XL, Examination CategoryB-N-3 examinationsfor core support structures, provides a high degree of confidence in the totalprogram.

The CNP RVI AMP implements MRP-227-A guidance for monitoring, recording, evaluating, and trending data that result from inspections. Inspections methodologies, inspection procedures, and inspection personnel guidance provided in MRP-228 will be followed. The program also credits monitoring performed by the ASME Section XI program.

The CNP RVI AMP monitoring and trending is consistent with NUREG-1801, Revision 2,Section XI.M 16A.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 5.6 Acceptance Criteria The following is NUREG-1801, Revision 2,Section XI.Ml6A, Evaluation and Technical Basis Element 6:

Section 5 of MRP-22 7 provides specific examinationacceptance criteriafor the Primary andExpansion Component examinations.For components addressed-by examinations referencedto ASME Code,Section XI, the IWB-3500 acceptance criteriaapply. For other components covered by Existing Programs,the examination acceptancecriteria are describedwithin the ExistingProgram reference document.

The guidance in MRP-22 7 contains three types of examination acceptance criteria:

" For visual examination (andsurface examinationas an alternative to visual examination), the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition,there are requirements to record and dispositionsurface breakingindications that are detected and sizedfor length by VT-1/E VT-1 examinations;

" For volumetric examination, the examination acceptance criterionis the capability for reliable detection of indications in bolting, as demonstratedin the examination Technical Justification;in addition, there are requirementsfor system-level assessment of bolted orpinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and Forphysical measurements, the examination acceptance criterionfor the acceptable tolerance in the measured differential heightfrom the top of the plenum rib pads to the vessel seating surface in B& Wplants are given in Table 5-1 of MRP-227. The acceptance criterionfor physical measurements performed on the height limits of the Westinghouse-designedhold-down springs will be developedprior to the first physical measurement.

The CNP RVI AMP applies examination acceptance criteria provided in MRP-227-A, Section 5.

In addition, WCAP-17096, Revision 2 "Reactor Internals Acceptance Criteria Methodology and Data Requirements" (WCAP-17096) has been developed by the PWROG and submitted by EPRI to the NRC for review and approval. I&M will evaluate degraded components using the supplemental guidance in WCAP-17096 as applicable. Application of guidance in WCAP-17096 will include any conditions or limitations resulting from the NRC review currently in-progress.

Acceptance criteria for the hold-down springs will be developed prior to the first physical measurement. Components inspected by the ASME Section XI program will be subject to acceptance criteria in the ASME Code as described in that program. The thimble tube inspection program contains acceptance criteria for inspection results.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program The CNP RVI AMP acceptance criteria are consistent with NUREG-1801, Revision 2,Section XI.M16A.

5.7 CorrectiveActions The following is NUREG-1801, Revision 2,Section XI.M 16A, Evaluation and Technical Basis Element 7:

Corrective actionsfollowing the detection of unacceptableconditions arefundamentally providedfor in each plant's correctiveaction program.Any detected conditions that do not satisfy the examination acceptance criteriaare requiredto be dispositionedthrough the plant correctiveaction program, which may require repair,replacement, or analyticalevaluationfor continuedservice until the next inspection. The dispositionwill ensure that design basisfunctions of the reactorinternals components will continue to be fulfilledfor all licensing basis loads andevents. Examples of methodologies that can be used to analyticallydisposition unacceptable conditions arefound in the ASME Code,Section XI or in Section 6 of MRP-22 7. Section 6 of MRP-22 7 describes the options that are availablefor dispositionof detected conditions that exceed the examination acceptance criteriaof Section 5 of the report. These include engineering evaluation methods, as well as supplementary examinations to further characterizethe detected condition, or the alternativeof component repairand replacementprocedures. The latter are subject to the requirementsof the ASME Code,Section XI. The implementation of the guidance in MRP-227,plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-relatedcomponents addressedin accordancewith the corrective actions of 10 CFR Part50, Appendix B or its equivalent, as applicable.

Other alternativecorrective action bases may be used to disposition relevant conditions if they have been previously approvedor endorsed by the NRC. Examples ofpreviously NRC-endorsed alternative correctiveactions bases include those correctiveactions bases for Westinghouse-design R VI components that are defined in Tables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghouse Report No. WCAP-14577-Rev. 1-A, orfor B& W-designedR VI components in B& W Report No. BA W-2248. Westinghouse Report No.

WCAP-14577-Rev. 1-A was endorsedfor use in an NRC SE to the Westinghouse Owners Group, dated February 10, 2001. B&W Report No. BA W-2248 was endorsedfor use in an SE to Framatome Technologies on behalfof the B& W Owners Group, dated December 9, 1999. Alternative corrective action bases not approved or endorsedby the NRC will be submittedfor NRC approvalprior to their implementation.

Corrective actions are recorded and dispositioned in accordance with the CNP Corrective Actions Program (CAP) and Quality Assurance Program Description. The CAP includes procedure guidance for action initiation, condition action and closure, conduct of evaluations, conduct of effectiveness reviews, and conduct of causal evaluations.

The CNP RVI AMP corrective actions are consistent with NUREG-1801, Revision 2,Section XI.M 16A.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 5.8 ConfirmationProcess The following is NUREG-1801, Revision 2,Section XI.M 16A, Evaluation and Technical Basis Element 8:

Site quality assuranceprocedures, review and approvalprocesses, and administrative controls are implemented in accordancewith the requirementsof 10 CFR Part50, Appendix B, or their equivalent, as applicable. It is expected that the implementation of the guidance in MRP-227 will provide an acceptable level of qualityfor inspection,flaw evaluation,and other elements of aging management of the PWR internals that are addressedin accordancewith the 10 CFR Part50, Appendix B, or their equivalent (as applicable), confirmationprocess, andadministrativecontrols.

The CNP Corrective Action Program and Quality Assurance Program Description have been developed in accordance with 10 CFR Part 50, Appendix B. The CNP RVI AMP implements the guidance in MRP-227-A.

The CNP RVI AMP confirmation process is consistent with NUREG-1801, Revision 2,Section XI.M16A.

5.9 Administrative Controls The following is NUREG-1801, Revision 2,Section XI.M 16A, Evaluation and Technical Basis Element 9:

The administrativecontrolsfor such programs, including their implementing procedures and review and approvalprocesses, are under existing site 10 CFR 50 Appendix B Quality Assurance Programs,or their equivalent, as applicable. Such a program is thus expected to be establishedwith a sufficient level of documentation and administrative controls to ensure effective long-term implementation.

The CNP Quality Assurance Program Description has been developed in accordance with 10 CFR Part 50, Appendix B. The Quality Assurance Program Description ensures proper administrative controls on the CNP RVI AMP.

The CNP RVI AMP administrative controls are consistent with NUREG-1801, Revision 2,Section XI.M16A.

5.10 OperatingExperience The following is NUREG-l1801, Revision 2,Section XI.M 16A, Evaluation and Technical Basis Element 10:

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Relatively few incidents ofPWR internals agingdegradation have been reportedin operating U.S. commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-22 7-A. The applicantis expected to review subsequent operating experiencefor impact on its program or to participatein industry initiatives thatperform thisfunction.

The applicationof the MRP-227 guidance will establish a considerableamount of operatingexperience over the next few years. Section 7 of MRP-22 7 describes the reportingrequirementsfor these applications,and the planfor evaluating the accumulatedadditionaloperatingexperience.

I&M will continue to be engaged with industry groups for sharing and reviewing OE in accordance with the Operating Experience Program. In addition, I&M will maintain industry involvement as described in Section 3.4. I&M will follow reporting requirements provided in MRP-227-A.

The CNP RVI AMP operating experience is consistent with in NUREG-1801, Revision 2,Section XI.M16A.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program

6.0 REFERENCES

6.1 American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components,"

2004 Edition, American Society of Mechanical Engineers, New York, NY.

6.2 CNP Document 01-DCP-0125, "Reactor Vessel Core Barrel - Replace Missing Bolts at Locations A-4, A-5 and A-6," March 1997 6.3 CNP Document 1-MOD-55520, "Replace Unit 1 Reactor Vessel Closure Head (1-OME-1)," July 2006.

6.4 CNP Document 2-MOD-55516, "Replace Unit 2 Reactor Vessel Closure Head (2-OME-1)," June 2007.

6.5 CNP Document 2-OHP-SP-045, "Unit 2 Cycle V-VI Refueling Procedure," May 1986 6.6 CNP Document 12-EHP-6040-PER-324, Revision 6, "Incore Instrumentation Thimble Tube Multifrequency Eddy Current Inspection," December 2011.

6.7 CNP Document "Cook Nuclear Plant Primary Strategic Water Chemistry Plan,"

Revision 7, November 2011.

6.8 CNP Document AEP-NRC-2011-38, "Revision to Regulatory Commitments Associated with Application for Renewed Operating Licenses," September 2011.

6.9 CNP Document AR 2010-1804, "Rx Vessel Core Support Lug Bolting Anomalies,"

Originated March 2010.

6.10 CNP Document AR 2010-10940, "Debris Found in 1-OME-1 on the Core Plate,"

Originated October 2010.

6.11 CNP Document Contract-6223, "Control Rod Guide Tube Support Pin Replacement," June 1985.

6.12 CNP Document "D. C. Cook Nuclear Plant Updated Final Safety Analysis Report,"

Revision 24, March 2012.

6.13 CNP Document EC-0000050972, Revision 2, "Replace Reactor Vessel Baffle Bolts," November 2010.

6.14 CNP Document EC-0000051640, "RX Vessel Lower Radial Support System (LRSS) Clevis Replacement Bolting for Unit 1," July 2012.

6.15 CNP Document GT 00846697, "License Renewal Implementation (Y10) for the RVI Program," Originated February 2009.

6.16 CNP Document GT 2012-1808, "Unit 1 CRGT Split Pin Replacement," Originated February 2012.

6.17 CNP Document GT 2012-1809, "Unit 2 CRGT Split Pin Replacement," Originated February 2012.

6.18 CNP Document ISI PROGRAM 4 INTERVAL, Revision 2, "ISI Program Plan Fourth Ten-Year Inservice Inspection interval Donald C. Cook Nuclear Plant, Units 1 & 2," May 2011.

6.19 CNP Document "License Renewal Application: Donald C. Cook Nuclear Plant,"

October 2003.

6.20 CNP Document LRP-EAMP-01, Revision 3, "Evaluation of Aging Management Programs for License Renewal," November 2005.

6.21 CNP Document PMI-7030, Revision 40, "Corrective Action Program," May 2012.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 6.22 CNP Document QAPD, Revision 22, "Donald C. Cook Nuclear Plant Quality Assurance Program Description," May 2012.

6.23 CNP Document RFC-01-2858, "Work to support Control Rod Guide Tube Replacement," June 1985.

6.24 CNP Document RFC-DC-01-2353, "Removal of Eight Part Length Rods, Installation of Anti-Rotation Devices on Each Part Length Rod CRDM, Install Eight Thimble Plug Devices in Place of the Part Length Rod," November 1978.

6.25 CNP Document RFC-DC-02-988, "Modify Unit 2 Fuel Assemblies from a 15X15 to a 17X17 Fuel Rod Array," June 1976.

6.26 CNP Document RFC-DC-02-2355, "Installation of Permanent Anti-Rotational Devices for Part Length CRDM Lead Screws," December 1978.

6.27 CNP Document RFC-DC-02-2924, "Control Rod Guide Tube Cap Screw Modifications," May 1986 6.28 MaterialsReliability Program:Screening, Categorization,andRanking of Reactor Internals Componentsfor Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

6.29 Materials ReliabilityProgram:PressurizedWater ReactorInternalsInspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

6.30 Materials Reliability Program:Inspection Standardfor PWR Internals (MRP-228).

EPRI, Palo Alto, CA: 2009. 1016609.

6.31 NEI Document NEI 03-08, Revision 2, "Guideline for the Management of Materials Issues," Nuclear Energy Institute, Washington, DC, January 2010.

6.32 NRC Document ML11308A770, "Revision I to the Final Safety Evaluation of the Electric Power Research institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, 'Pressurized Water Reactor (PWR)

Internals Inspection and Evaluation Guidelines' (TAC NO. ME0680," December 2011.

6.33 NRC Document NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report," December 2010.

6.34 NRC Document NUREG-1831, "Safety Evaluation Report Related to the License Renewal of the Donald C. Cook Nuclear Plant, Units 1 and 2," Docket Nos. 50-315 and 50-316. Indiana Michigan Power Company, July 2005.

6.35 NRC Document NRCB 88-09, "Thimble Tube Thinning in Westinghouse Reactors," July 1988.

6.36 NRC Document RIS 11-07 "License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," July 2011.

6.37 Westinghouse Document "

Attachment:

Description of Additional Guide Tube Repairs for D.C. Cook No. 1" 6.38 Westinghouse Document WCAP-1 1000, "D. C. Cook Unit 2 Estimated Operability with Failed Control Rod Guide Tube Support Pins," January 1985.

6.39 Westinghouse Document WCAP-14577, Revision 1-A, "License Renewal Evaluation: Aging Management for Reactor Internals," March, 2001.

6.40 Westinghouse Document WCAP- 17096-NP, Revision 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," December 2009 Page 38 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 6.41 Westinghouse Document WCAP-17300, Revision 0, "Reactor Vessel Internals Program Plan for Aging Management of Reactor Internals at D.C. Cook Nuclear Plant Unit 1," February 2011.

6.42 Westinghouse Document WCAP-17301, Revision 0, "Reactor Vessel Internals Program Plan for Aging Management of Reactor Internals at D.C. Cook Nuclear Plant Unit 2," February 2011.

Page 39 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program APPENDIX A: PRIMARY INSPECTION COMPONENTS The following is Table 4-3 "Westinghouse plants Primary components" from MRP-227-A. The CNP RVI AMP implements the guidance provided in this table.

Page 40 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Item Applicability Effect (Mechanism) Expansion (Note 1)Link Examination(Note 1)

Method/Frequency Examination Coverage Control Rod Guide Tube All plants Loss of Material None Visual (VT-3) examination no 20% examination of the Assembly (Wear) later than 2 refueling outages number of CRGT Guide plates (cards) from the beginning of the assemblies, with all guide license renewal period, and cards within each selected no earlier than two refueling CRGT assembly examined.

outages prior to the start of the license renewal period. See Figure 4-20 Subsequent examinations are required on a ten-year interval.

Control Rod Guide Tube All plants Cracking (SCC, Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible)

Assembly Fatigue) instrumentation examination to determine the CRGT lower flange weld Lower flange welds Aging (BMI) column bodies, presence of crack-like surfaces and adjacent base Management (IE Lower support surface flaws in flange welds metal on the individual and TE) column bodies (cast) no later than 2 refueling periphery CRGT Upper core plate outages from the beginning assemblies.

Lower support of the license renewal period (Note 2) forging/casting and subsequent examination on a ten-year interval. See Figure 4-21 Core Barrel Assembly All plants Cracking (SCC) Lower support Periodic enhanced visual 100% of one side of the Upper core barrel flange column bodies (non (EVT-1) examination, no later accessible surfaces of the weld cast) than 2 refueling outages from selected weld and adjacent Core barrel outlet the beginning of the license base metal (Note 4).

nozzle welds renewal period and subsequent examination on a See Figure 4-22 ten-year interval.

Core Barrel Assembly All plants Cracking (SCC, Upper and lower Periodic enhanced visual 100% of one side of the Upper and lower core IASCC, Fatigue) core barrel cylinder (EVT-1) examination, no later accessible surfaces of the barrel cylinder girth welds axial welds than 2 refueling outages from selected weld and adjacent the beginning of the license base metal (Note 4).

renewal period and subsequent examination on a See Figure 4-22 ten-year interval. II Page 41 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Examination (Note 1)

Item Applicability Effect (Mechanism) (Note 1)Link Expansion Method/Frequency Examination Coverage Core Barrel Assembly All plants Cracking (SCC, None Periodic enhanced visual 100% of one side of the Lower core barrel flange Fatigue) (EVT-1) examination, no later accessible surfaces of the weld (Note 5) than 2 refueling outages from selected weld and adjacent the beginning of the license base metal (Note 4).

renewal period and subsequent examination on a See Figure 4-22 ten-year interval.

Baffle-Former Assembly All plants with Cracking (IASCC, None Visual (VT-3) examination, Bolts and locking devices Baffle-edge bolts baffle-edge Fatigue) that with baseline examination on high fluence seams.

bolts results in between 20 and 40 EFPY 100% of components

" Lost or broken and subsequent accessible from core side locking devices examinations on a ten-year (Note 3).

" Failed or interval.

missing bolts See Figure 4-23

" Protrusion of bolt heads Aging Management (IE and ISR)

(Note 6)

Baffle-Former Assembly All plants Cracking (IASCC, Lower support Baseline volumetric (UT) 100% of accessible bolts Baffle-former bolts Fatigue) column bolts, examination between 25 and (Note 3). Heads accessible Aging Barrel-former bolts 35 EFPY, with subsequent from the core side. UT Management (IE examination on a ten-year accessibility may be and ISR) interval, affected by complexity of (Note 6) head and locking device designs.

See Figures 4-23 and 4-24 Page 42 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Item Applicability Effect (Mechanism) Expansion Examination ItmAplcbliyEfet(eca m Link (Note 1) Method/Frequency (Note 1) Examination Coverage Baffle-Former Assembly All plants Distortion (Void None Visual (VT-3) examination to Core side surface as Assembly Swelling), or check for evidence of indicated.

(Includes: Baffle plates, baffle Cracking (IASCC) distortion, with baseline edge bolts and indirect effects that results in examination between 20 and See Figures 4-24, 4-25, 4-of void swelling in former

• Abnormal 40 EFPY and subsequent 26 and 4-27 plates) interaction with fuel examinations on a ten-year assemblies interval.

" Gaps along high fluence baffle joint

" Vertical displacement of baffle plates near high fluence joint

" Broken or damaged edge bolt locking systems along high fluence baffle joint Alignment and Interfacing All plants with Distortion (Loss of None Direct measurement of spring Measurements should be Components 304 stainless Load) height within three cycles of taken at several points Internals hold down spring steel hold down the beginning of the license around the circumference springs Note: This renewal period. If the first set of the spring, with a mechanism was not of measurements is not statistically adequate strictly identified in sufficient to determine life, number of measurements the original list of spring height measurements at each point to minimize age-related must be taken during the next uncertainty.

degradation two outages, in order to mechanisms [7]. extrapolate the expected See Figure 4-28 spring height to 60 years. I _I Page 43 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Thermal Shield Assembly All plants with Cracking None Visual (VT-3) no later than 2 100% of thermal shield Thermal shield flexures thermal shields (Fatigue) refueling outages from the flexures.

or Loss of beginning of the license Material (Wear) renewal period. Subsequent See Figures 4-29 and 4-that results in examinations on a ten-year 36 thermal shield interval.

flexures excessive wear, fracture, or complete separation Notes to Table 4-3:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.

2. A minimum of 75% of the total identified sample population must be examined.

3. A minimum of 75% of the total population (examined + unexamined), including coverage consistent with the Expansion criteria inTable 5-3, must be examined for inspection credit.

4. A minimum of 75% of the total weld length (examined + unexamined), including coverage consistent with the Expansion criteria in Table 5-3, must be examined from either the inner or outer diameter for inspection credit.

5. The lower core barrel flange weld may be alternatively designated as the core barrel-to-support plate weld insome Westinghouse plant designs.

6. Void swelling effects on this component is managed through management of void swelling on the entire baffle-former assembly.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program APPENDIX B: EXPANSION INSPECTION COMPONENTS The following is Table 4-6 "Westinghouse plants Expansion components" from MRP-227-A.

The CNP RVI AMP implements the guidance provided in this table.

Page 45 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Effect Primary Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Upper Internals Assembly All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessible Upper core plate (Fatigue, Wear) flange weld examination, surfaces (Note 2).

Re-inspection every 10 years following initial inspection.

Lower Internals Assembly All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessible Lower support forging or Aging flange weld examination, surfaces (Note 2).

castings Management (TE Re-inspection every 10 in Casting) years following initial See Figure 4-33.

inspection.

Core Barrel Assembly All plants Cracking Baffle-former Volumetric (UT) 100% of accessible bolts.

Barrel-former bolts (IASCC, Fatigue) bolts examination. Accessibility may be Aging Re-inspection every 10 limited by presence of Management (IE, years following initial thermal shields or neutron Void Swelling inspection, pads (Note 2).

and ISR)

See Figure 4-23.

Lower Support Assembly All plants Cracking Baffle-former Volumetric (UT) 100% of accessible bolts Lower support column bolts (IASCC, Fatigue) bolts examination, or as supported by plant-Aging Re-inspection every 10 specific justification (Note Management (IE years following initial 2).

and ISR) inspection.

See Figures 4-32 and 4-33.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Effect Primary Link Examination Item Applicability (Mechanism) (Note 1) MethodlFrequency Examination Coverage (Note 1)

Core Barrel Assembly All plants Cracking (SCC, Upper core Enhanced visual (EVT-1) 100% of one side of the Core barrel outlet nozzle Fatigue) barrel flange examination, accessible surfaces of the welds Aging weld Re-inspection every 10 selected weld and Management (IE years following initial adjacent base metal (Note of lower inspection. 2) sections)

See Figure 4-22 Core Barrel Assembly All plants Cracking (SCC, Upper and Enhanced visual (EVT-1) 100% of one side of the Upper and lower core barrel IASCC) lower core examination, accessible surfaces of the cylinder axial welds Aging barrel cylinder e-inspection every 10 years selected weld and Management girth welds following initial inspection, adjacent base metal (Note (IE) 2).

See Figure 4-22 Lower Support Assembly All plants Cracking Upper core Enhanced visual (EVT-1) 100% of accessible Lower support column bodies (IASCC) barrel flange examination, surfaces (Note 2).

(non cast) Aging weld Re-inspection every 10 Management years following initial See Figure 4-34.

(IE) inspection.

Lower Support Assembly All plants Cracking Control rod Visual (EVT-1) examination. 100% of accessible Lower support column bodies (IASCC) guide tube Re-inspection every 10 support columns (Note 2).

(cast) including the (CRGT) lower years following initial detection of flanges inspection. See Figure 4-34.

fractured support columns Aging Management (IE)

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Effect Primary Link Examination Item Applicability Method/Frequency Examination Coverage (Mechanism)

Item Applicability (Note 1) (Note 1)

Bottom Mounted All plants Cracking Control rod Visual (VT-3) examination of 100% of BMI column Instrumentation System (Fatigue) guide tube BMI column bodies as bodies for which difficulty Bottom-mounted including the (CRGT) lower indicated by difficulty of is detected during flux instrumentation (BMI) column detection of flanges insertion/withdrawal of flux thimble bodies completely thimbles. insertion/withdrawal.

fractured column Re-inspection every 10 bodies years following initial See Figure 4-35.

Aging inspection.

Management Flux thimble (IE) insertion/withdrawal to be monitored at each inspection interval.

Notes to Table 4-6:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 5-3.

2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program APPENDIX C: EXISTING PROGRAMS COMPONENTS The following is Table 4-9 "Westinghouse plants Existing Programs components" from MRP-227-A. The CNP RVI AMP is consistent with the information provided in this table.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Core barrel flange (Wear) Section Xl examination to specified frequency.

determine general condition for excessive wear.

Upper Internals Assembly All plants Cracking (SCC, ASME Code Visual (VT-3) All accessible surfaces at Upper support ring or skirt Fatigue) Section Xl examination, specified frequency.

Lower Internals Assembly All plants Cracking (IASCC, ASME Code Visual (VT-3) All accessible surfaces at Lower core plate Fatigue) Section XI examination of the lower specified frequency.

XL lower core plate (Note 1) Aging core plates to detect Management (IE) evidence of distortion and/or loss of bolt integrity.

Lower Internals Assembly All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Lower core plate (Wear) Section Xl examination, specified frequency.

XL lower core plate (Note 1)

Bottom Mounted All plants Loss of material NUREG-1801 Surface (ET) Eddy current surface Instrumentation System (Wear) Rev. 1 examination, examination as defined in Flux thimble tubes plant response to IEB 88-09.

Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components (Wear) Section Xl examination, specified frequency.

Clevis insert bolts (Note 2)

Alignment and Interfacing All plants Loss of material ASME Code Visual (VT-3) All accessible surfaces at Components (Wear) Section Xl examination, specified frequency.

Upper core plate alignment pins Notes to Table 4-9:

1. XL = 'Extra Long" referring to Westinghouse plants with 14-foot cores.

2. Bolt was screened in because of stress relaxation and associated cracking; however, wear of the clevis/insert is the issue.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program APPENDIX D: ACCEPTANCE AND EXPANSION CRITERIA The following is Table 5-3 "Westinghouse plants examination acceptance and expansion criteria" from MRP-227-A. The CNP RVI AMP implements the guidance provided in this table.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Examination ExamiationAdditional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Additinal Eaitio (Note 1) Acceptance Criteria Control Rod Guide All plants Visual (VT-3) None N/A N/A Tube Assembly examination.

Guide plates (cards)

The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

Control Rod Guide All plants Enhanced visual (EVT-1) a. Bottom-mounted a. Confirmation of surface- a. For BMI column bodies, Tube Assembly examination, instrumentation breaking indications in two or the specific relevant Lower flange welds (BMI) column bodies more CRGT lower flange condition for the VT-3 welds, combined with flux examination is completely The specific relevant thimble insertion/withdrawal fractured column bodies.

condition is a detectable b. Lower support difficulty, shall require visual crack-like surface column bodies (VT-3) examination of BMI indication. (cast), upper core column bodies by the b. For cast lower support plate and lower completion of the next column bodies, upper core support forging or refueling outage. plate and lower support casting forging/castings, the specific relevant condition

b. Confirmation of surface- is a detectable crack-like breaking indications in two or surface indication.

more CRGT lower flange welds shall require EVT-1 examination of cast lower support column bodies, upper core plate and lower support forging/castings within three fuel cycles following the initial observation.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1)

Core Barrel Assembly All plants Periodic enhanced visual a. Core barrel outlet a. The confirmed detection and a and b. The specific Upper core barrel flange (EVT-1) examination, nozzle welds sizing of a surface-breaking relevant condition for the weld b. Lower support indication with a length greater expansion core barrel column bodies (non than two inches in the upper outlet nozzle weld and The specific relevant lower support column body condition is a detectable cast) core barrel flange weld shall require that the EVT-1 examination is a detectable crack-like surface examination be expanded to crack-like surface indication. include the core barrel outlet indication.

nozzle welds by the completion of the next refueling outage.

b. If extensive cracking in the core barrel outlet nozzle welds is detected, EVT-1 examination shall be expanded to include the upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles following the initial observation.

Core Barrel Assembly All plants Periodic enhanced visual None None None Lower core barrel flange (EVT-1) examination.

weld (Note 2)

The specific relevant condition is a detectable crack-like surface indication.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1)

Core Barrel Assembly All plants Periodic enhanced visual Upper core barrel The confirmed detection and The specific relevant Upper core barrel cylinder (EVT-1) examination, cylinder axial welds sizing of a surface-breaking condition for the expansion girth welds indication with a length greater upper core barrel cylinder than two inches in the upper axial weld examination is a The specific relevant core barrel cylinder girth welds detectable crack-like condition is a detectable shall require that the EVT-1 surface indication.

crack-like surface examination be expanded to indication, include the upper core barrel cylinder axial welds by the completion of the next refueling outage.

Core Barrel Assembly All plants Periodic enhanced visual Lower core barrel The confirmed detection and The specific relevant Lower core barrel cylinder (EVT-1) examination, cylinder axial welds sizing of a surface-breaking condition for the expansion girth welds indication with a length greater lower core barrel cylinder than two inches in the lower axial weld examination is a The specific relevant core barrel cylinder girth welds detectable crack-like condition is a detectable shall require that the EVT-1 surface indication.

crack-like surface examination be expanded to indication. include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1)

Baffle-Former All plants Visual (VT-3) None N/A N/A Assembly with baffle- examination.

Baffle-edge bolts edge bolts The specific relevant conditions are missing or broken locking devices, failed or missing bolts, and protrusion of bolt heads.

Baffle-Former All plants Volumetric (UT) a. Lower support a. Confirmation that more than a and b. The examination Assembly examination, column bolts 5% of the baffle-former bolts acceptance criteria for the Baffle-former bolts actually examined on the four UT of the lower support baffle plates at the largest column bolts and the The examination b. Barrel-former bolts distance from the core barrel-former bolts shall be acceptance criteria for (presumed to be the lowest established as part of the the UT of the baffle- dose locations) contain examination technical former bolts shall be unacceptable indications shall justification.

established as part of require UT examination of the the examination lower support column bolts technical justification. within the next three fuel cycles.

b. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require UT examination of the barrel-former bolts.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Additinal Eaitio (Note 1) Acceptance Criteria Baffle-Former All plants Visual (VT-3) None N/A N/A Assembly examination.

Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near high fluence joints, and broken or damaged edge bolt locking systems along high fluence baffle plate joints.

Alignment and All plants Direct physical None N/A N/A Interfacing Components with 304 measurement of spring Internals hold down stainless height.

spring steel hold down springs The examination acceptance criterion for this measurement is that the remaining compressible height of the spring shall provide hold-down forces within the plant-specific design tolerance.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1)

Thermal Shield All plants Visual (VT-3) None N/A N/A Assembly with thermal examination.

Thermal shield flexures shields The specific relevant conditions for thermal shield flexures are excessive wear, fracture, or complete separation.

Notes to Table 5-3:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

2. The lower core barrel flange weld may alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program APPENDIX E: REACTOR COMPONENT ILLUSTRATIONS The following reactor component figures have been reproduced from MRP-227-A.

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Donald C. Cook Nuclear Plant Reactor Vessel Intemals Aging Management Program Figure E-1 Typical Westinghouse Control Rod Guide Card Page 59 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Upper Guide Tube Upper Support Plate i

Lower Guide tube 4-LP Sheaths and C-Tubes a

Figure E-2 Lower Section of Control Rod Guide Tube Assembly Page 60 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Flange Weld Axial Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Axial Weld Lower Barrel Circumferential Weld Lower Barrel Axial Weld Thermal Shield Flexure Core Barrel to Support Plate Weld Figure E-3 Major Core Barrel Welds Page 61 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program

't 0)

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a) 0

ýEECO CO CO Figure E-4 Bolting Systems Used in Westinghouse Core Baffles Page 62 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program INTERNALS SUPPORT LEDGE-THERMAL SHIELD BAFFLE FORMER LOWER

.CORE PLATE CORE SUPPORT COLUMN DIFFUSER PLATE CORE SUPPORT FORGING Figure E-5 Core Baffle/Barrel Structure Page 63 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program BAFFLE TO FORME BOLT(LONO& U(oR1 CORKER EDGE BRACI~r BAFFLE TO FOOME BOLT Figure E-6 Bolting in a Typical Westinghouse Baffle-Former Structure Page 64 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Vertical Displacement Figure E-7 Vertical Displacement between the Baffle Plates and Bracket at the Bottom of the Baffle-Former-Barrel Assembly (exaggerated)

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Figure E-8 Schematic Cross-Sections of the Westinghouse Hold-down Springs W Id 7-

/

Figure E-9 Typical Thermal Shield Flexure Page 66 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Plate Lower Core Support Structure Core Support Plate (Forging)

Figure E-1O Lower Core Support Structure Page 67 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program LOWER CORE PLATE DIFFUSER PLATE SCORE SUPPORT PLATE/FORGING BOTTOM MOUNTED INSTRUMENTATION COLUMN Figure E-11 Lower Core Support Structure - Core Support Plate Cross-Section Figure E-12 Typical Core Support Column Page 68 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program

/

I Figure E-13 Examples of Bottom-Mounted Instrumentation (BMI) Column Designs Page 69 of 69

ENCLOSURE 2 TO AEP-NRC-2012-82 DONALD C. COOK NUCLEAR PLANT LIST OF REGULATORY COMMITMENTS to AEP-NRC-2012-82 Page 1 List of Regulatory Commitments The following table identifies those actions committed to by Indiana Michigan Power Company in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

SCHEDULED COMMITMENT COMPLETION DATE This Action Item addresses the applicability of the FMECA and functionality analysis assumptions made in the development of MRP-227-A to individual facilities. I&M is participating in PWROG project PA-MSC-0938, "Support for Applicant Action Items 1, 2, Unit 1: October 25, 2014 and 7 from the Final Safety Evaluation on MRP-227, Revision 0" to address this item.

Unit 2: December 23, 2017 The results of the evaluation will be provided to the NRC prior to the period of extended operation for each unit.

This Action Item requires the licensee to verify that all the RVI components within the scope for license renewal at that facility have been considered in applicable documents in development of MRP-227-A. I&M is participating in PWROG project PA-MSC- Unit 1: October 25, 2014 0938, "Support for Applicant Action Items 1, 2, and 7 from the Final Safety Evaluation on MRP-227, Revision 0" to address this item.

Unit 2: December 23, 2017 The results of the evaluation will be provided to the NRC prior to the period of extended operation for each unit.

CNP Unit 1 and Unit 2 both have X-750 split pins. Project requests have been initiated to investigate split pin replacement for each Unit 1: October 25, 2014 unit.

I&M will provide the NRC with the strategy for managing split pins Unit 2: December 23, 2017 prior to the period of extended operation for each unit.

CNP Unit 1 and Unit 2 both have 304 SS hold down springs. MRP-227-A guidance includes physical measurement of 304 SS hold uirio thesirst down springs. This action item requires acceptance criteria to be required physical provided to the NRC. CNP plant specific acceptance criteria will be developed and submitted to the NRC prior to the first required physical measurement. The hold down springs will be replaced if Unit 2 Prior to the first acceptance criteria are not developed in lieu of performing the first required physical required physical measurement. measurement.

to AEP-NRC-2012-82 Page 2 A plant specific evaluation of RVI CASS materials is required in this Action Item. I&M is participating in PWROG project PA-MSC-0938, "Support for Applicant Action Items 1, 2, and 7 from the Final Unit 1: October 25, 2014 Safety Evaluation on MRP-227, Revision 0" to address this item The results of the evaluation will be provided to the NRC prior to Unit 2: December 23, 2017 the period of extended operation for each unit.