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 POWERý A unit of American Electric Power Indiana Michigan Power One Cook Place Bridgman, MI 49106 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 Page 2 AEP-NRC-2012-82 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.

2.

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

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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:

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Kevin A. Kalchik Engineer II Reviewed:

Alex M. Olp f Engineer II Approved:

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

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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 P O SE...........................................................................................................................

8 2.0 DESRIPTION OF CNP REACTOR INTERNALS..................................................

9 2.1 Unit 1 Operating Experience and Records Search................................................

13 2.1.1 Control Rod Guide Tube and Split Pin Replacement.....................................

13 2.1.2 Barrel-Former Bolt Inspection and Partial Replacement.............................. 13 2.1.3 Clevis Insert Bolt Degradation........................................................................

13 2.2 Unit 2 Operating Experience and Records Search................................................

14 2.2.1 Modification from 15XJ5 to 1 7X1 7 Design....................................................

14 2.2.2 Control Rod Guide Tube Split Pin Replacement...........................................

14 2.2.3 Control Rod Guide Tube Cap Screw Modification.........................................

14 2.2.4 Baffle-Former Bolt Partial Replacement.........................................................

14 3.0 EXISTING PROGRAMS AND ACTIVITIES........................................................

15 3.1 ASME Section XI ISI Program..............................................................................

15 3.2 Primary Strategic Water Chemistry Plan...............................................................

15 3.3 Thimble Tube Multifrequency Eddy Current Inspection.......................................

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 Guidance in 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 Conditions and Limitations...........................................

20 4.4.1 Topical Report Conditions...............................................................................

20 4.4.2 Applicant/Licensee Action Items....................................................................

20 5.0 PROGRAM ATTRIBUTE EVALUATION.............................................................

26 5.1 Scope of Program.....................................................................................................

28 5.2 Preventive A ctions...................................................................................................

29 5.3 Parameters Monitored/Inspected............................................................................

29 5.4 Detection of Aging Effects.......................................................................................

31 5.5 Monitoring and Trending.......................................................................................

32 5.6 A cceptance Criteria..................................................................................................

33 5.7 Corrective A ctions....................................................................................................

34 5.8 Confirmation Process..............................................................................................

35 5.9 Administrative Controls..........................................................................................

35 5.10 Operating Experience..............................................................................................

35 6.0 R E F E R E 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 INSTRUMENTATION THIMBLE GUIDES RADIAL SUPPORT CORE SUPPORT ACCESS PORT 41 REACTOR VESSEL 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 internals assembly that is removed during each refueling operation to obtain access to the reactor core and a lower internals assembly that can be removed following a complete core off-load.

The lower internals assembly is supported in the vessel by clamping to a ledge below the vessel-head mating surface and is closely guided at the bottom by radial support/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 barrel alignment pins of the lower internals assembly.

Upper Internals Assembly The major sub-assemblies that comprise the upper internals assembly are the: (1) upper core plate (UCP) and fuel alignment pins; (2) upper support column assemblies; (3) control rod guide tube assemblies and flow downcomers; (4) upper plenum; and (5) upper support plate assembly.

During reactor operation, the upper internals assembly is preloaded against the fuel assembly springs and the internals holddown springs by the reactor vessel head pressing down on the outside edge of the upper support plate (USP). The USP acts as the divider between the upper plenum and the reactor vessel head and as a relatively stiff base for the rest of the upper internals. The upper support columns and the guide tubes are attached to the USP. The UCP, in turn, is attached to the upper support columns. The USP assemblies are designated as one of three different designs: (1) a deep beam design, (2) a top hat design, or (3) an inverted top hat design.

The UCP is perforated to permit coolant to pass from 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 laterally supports the core by fuel alignment pins extending below the plate. The UCP contacts and preloads the fuel assembly springs and thus maintains contact of the fuel assemblies with the lower core plate (LCP) during reactor operation.

The upper support columns vertically position the UCP and are designed to take the uplifting hydraulic flow loads and fuel spring loads on the UCP. The guide tubes are bolted to the USP and pinned at the UCP so they can be easily removed if replacement is desired. The guide tubes are designed to guide the control rods in and out of the fuel assemblies to control power generation. The guide tubes are also slotted in their lower sections to allow coolant exiting from the core toflow into the upper plenum.

The upper instrumentation columns are bolted to the USP. These columns support the thermocouple guide tubes that lead the thermocouples from the reactor head into the upper plenum to just above the UCP.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program The UCP alignment pins 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 differential thermal expansions between the upper internals and the core barrel. The UCP alignment pins are the interfacing components between the UCP and the core barrel. The UCP alignment pins are shrunk-fit and welded into the core barrel and the core barrel bearing pad. The gap sizes between the alignment pins and the matching inserts are customized.

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

Lower Internals Assembly The reactor core is positioned and supported by the lower internals and upper internals assemblies. The individual fuel assemblies are positioned by fuel alignment pins in the LCP and in the UCP. These pins control the orientation of the core with respect to the lower internals and upper internals assemblies. The lower internals are aligned with the upper internals by the UCP alignment pins and secondarily by the head/vessel alignment pins. The lower internals are orientated to the vessel by the lower radial keys and by the head/vessel alignment pins. Thus, the core is aligned with the vessel by a number of interfacing components.

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 radial support/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 barrel alignment pins of the lower internals assembly.

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

The LCP is elevated above the lower support forging by support columns and bolted to a ring support attached to the inside diameter of the core barrel. The support columns transmit vertical fuel assembly loads from the LCP to the much thicker lower support forging. The lower support forging is welded to and supported by the core barrel, which transmits vertical loads to the vessel through the core barrel flange.

The functions of the LCP are to position and support the core and provide a metered control of reactor coolant flow into each fuel assembly. The LCP is located near the bottom of the lower support assembly, inside the core barrel, and above the lower support forging.

The function of the lower support forging or casting is to provide support for the core.

The lower support forging is attached with a full-penetration weld to the lower end of the core barrel. In this position it can provide uninterrupted support to the core. The core sits directly on the LCP, which is supported by the lower support columns that are attached to and extend above the lower support forging. Some four-loop plants employ a Page 11 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program cast lower support instead of a forging. The functions, loads, and supporting hardware are the same except for dimensions.

The primary function of the core barrel is to support the core. A large number of components are attached to either the core barrel or the core barrel flange, including the baffle/former assembly, the outlet nozzles, the neutron panel assemblies or thermal shield, the alignment pins that engage the UCP and the LCP, the lower support forging, and the LCP. The radial keys restrain large transverse motions of the core barrel but at the same time allow unrestricted radial and axial thermal expansions.

The baffle and former assembly is made up of vertical plates called baffles and horizontal support plates called formers. The baffle plates are bolted to the formers by the baffle/former bolts, and the formers are attached to the core barrel inside diameter by the barrel/former bolts. The baffle/former assembly forms the interface between the core and the core barrel. The baffles provide a barrier between 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 requirement of the baffles, is to reduce the neutron flux on the vessel.

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

The function of the core barrel outlet nozzles is to direct the reactor coolant, after it leaves the core, radially outward through the reactor vessel outlet nozzles. The core barrel outlet nozzles are located in the upper portion of the core barrel directly below the flange and are attached to the core barrel, each in line with a vessel outlet nozzle.

Additional neutron shielding of the reactor vessel is provided in the active core region by neutron panels or thermal shields that are attached to the outside of the core barrel.

Specimen guides that contain specimens for determining the irradiation effects of the vessel during the life of the plant are attached to the neutron panels/thermal shields.

The flux thimble is a long, slender stainless steel tube that passes from an external seal table, through the bottom mounted nozzle penetration, through the lower internals assembly, and finally extends to the top of the fuel assembly. The flux thimble provides a path for the neutron flux detector into the core and is subjected to reactor coolant pressure and temperature on the outside surface and to atmospheric conditions on the inside. The flux thimble path from the seal table to the bottom mounted nozzles is defined by flux thimble guide tubes, which are part of the primary pressure boundary and not considered to be part of the internals. The bottom-mounted instrumentation (BMJ) columns provide a path for the flux thimbles from the bottom of the vessel into the core.

The BMI columns align the flux thimble path with instrumentation thimbles in the fuel assembly.

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Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program The LCP and the fuel alignment pins, the lower support forging or casting, the lower support columns, the core barrel, the core barrel flange, the radial support keys, the baffle plates, and the former plates are typically classified as core support structures.

2.1 Unit 1 Operating Experience 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 Control Rod 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 1 C9). 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-Former Bolt Inspection and Partial Replacement A barrel-former bolt was discovered on the lower core plate after defueling in 1994 (U1C14).

Inspection was performed in 1995 (U 1 C 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 Insert Bolt 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 Operating Experience 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 Modification from 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 Control Rod 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 Control Rod 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-Former Bolt Partial Replacement 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 Primary Strategic Water Chemistry Plan 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 Current Inspection 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 Primary group. The aging management requirements that are needed to ensure functionality of Primary components are described in these I&E guidelines. The Primary group also includes components which have shown a degree of tolerance to a specific aging degradation effect, but for which no highly susceptible component exists or for 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 aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group.

The schedule for implementation of aging management requirements for Expansion components will depend on thefindingsfrom the examinations of the Primary components at individual plants.

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

" No Additional Measures: those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the No Additional Measures group. Additional components were placed in the No Additional Measures group as a result of FMECA and the functionality assessment. No further action is required by these guidelines for managing the aging of the No Additional Measures components.

The categorization and analysis processes described herein are not intended to supersede any ASME B&P V Code Section X1 [2] requirements. Any components that are classified as core support structures 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 requirements that remain in effect and may only be altered as 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 General Assumption 1 The following is the first general assumption from Section 2.4 of MRP-227-A:

30 years of operation with high leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for 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 General Assumption 2 The following is the second general assumption from Section 2.4 of MRP-227-A:

Base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or 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 General Assumption 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 Guidance in 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 program for management of aging of reactor internal components within thirty-six months following issuance of MRP-22 7-Rev. 0 (that is, no later than December 31, 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 months following 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:

Examinations specified in these guidelines shall be conducted in accordance with 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:

Examination results that do not meet the examination acceptance criteria defined in Section 5 of these guidelines shall be recorded and entered in the plant corrective action 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 shall provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MAP Program Manager 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 engineering evaluation is used to disposition an examination result that does not meet the examination acceptance criteria in Section 5, this engineering evaluation shall be conducted in accordance with 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 Evaluation Report 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/Licensee Action 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/Licensee Action 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 addressed in Section 3.2.5.1 of this SE, each applicant/licensee is responsible for assessing its plant's design and operating history and demonstrating that the approved version of MRP-22 7 is applicable to the facility. Each applicant/licensee shall refer, in particular, to the assumptions regarding plant design and operating history made in the Page 20 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program FMECA and functionality analyses for reactors of their design (i.e., Westinghouse, CE, or B& W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their R VI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to 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-22 7 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 addressed in 10 CFR 54.4, each applicant/licensee is responsible for identifying which R VI components are within the scope of LR for its facility. Applicants/licensees shall 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 their facilities in accordance with 10 CFR 54.4. If the tables do not identify all the R VI components that are within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) and propose any necessary modifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specific AMP. The AMP shall provide assurance that the effects of aging on the missing component(s) will be managed for the period of 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/Licensee Action 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 addressed in Section 3.2.5.3 in this SE, applicants/licensees of CE and Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an applicant 's/licensee's existing programs, or to identify changes to the programs that Page 21 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program should be implemented to manage the aging of these components for the period of extended operation. The results of this plant-specific analyses and a description of the plant-specific programs being relied on to manage aging of these components shall be submitted as part of the applicant's/licensee's AMP application. The CE and Westinghouse components identified for this type ofplant-specific evaluation include: CE thermal shield positioning pins and CE in-core instrumentation thimble tubes (Section 4.3.2 in MRP-227), and Westinghouse guide tube support pins (split pins) (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/licensees shall confirm that the core support structure upper flange weld was stress relieved during the original fabrication of the Reactor Pressure Vessel in order to confirm the applicability of MRP-227, as approved by the NRC, to their facility. If the upper flange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component. If necessary, the examination methods andfrequency for non-stress relieved B& W core support structure upper flange welds shall be consistent with the recommendations in MRP-227, as approved by the NRC, for the Westinghouse and CE upper core support barrel welds. The examination coverage for this B& Wflange weld shall conform to the staff's imposed criteria as described in Sections 3.3.1 and 4.3.1 of this SE. The applicant 's/licensee's resolution of this plant-specific action item shall be submitted to the NRC for 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/Licensee Action 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 addressed in Section 3.3.5 in this SE, applicants/licensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-22 7for loss of compressibility for Westinghouse hold down springs, and for distortion in the gap between the top and bottom core shroud segments in CE units with core barrel shrouds assembled in two vertical sections. The applicant/licensee shall include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are 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 inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version 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/Licensee Action 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 addressed in Section 3.3.6 in this SE, MRP-227 does not propose to inspect the following inaccessible components: the B& W core barrel cylinders (including vertical and circumferential seam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices, B& W core barrel-to-former bolts and their locking devices, and B& W core barrel assembly internal baffle-to-baffle bolts. The MRP also identified that although the B& W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques.

Applicants/licensees shall justify the acceptability of these components for continued operation through the period of extended operation by performing an evaluation, or by proposing a scheduled replacement of the components. As part of their application to implement the approved version of MRP-22 7, applicants/licensees shall provide their justification for the continued operability of each of the inaccessible components and, if necessary, provide their plan for the replacement of the components for 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/Licensee Action 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/licensees of B& W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B& WIMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operation or for additional R VI components that may befabricatedfrom CASS, martensitic stainless steel or precipitation hardened stainless 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 thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiring aging management during development of MRP-22 7. That is, the requirement would apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licensee shall include the plant-specific analysis 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/Licensee Action Item 8 (MRP-227 SE Sections 3.5.1 and 4.2.8)

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 accordance with NEI 03-08, applicants/licensees whose licensing basis contains a commitment to submit a PWR R VI AMP and/or inspection program shall also make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by this SE. An applicant 's/licensee's application to implement MRP-22 7, as amended by this SE shall include the following items (1) and (2). Applicants who submit applications for LR after the issuance of this SE shall, in accordance with the NUREG-1801, Revision 2, submit the information provided in the following items (1) through (5)for staff review 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 addresses the 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 program and the plant-specific action items will be carried out by applicants/licensees, applicants/licensees are to submit an inspection plan which addresses the identifiedplant-specific action items for staff review and approval consistent with the licensing basis for the plant. If an applicant/licensee plans to implement an AMP which deviates from the guidance provided in MRP-227, as approved by the NRC, the applicant/licensee shall identify where their program deviates from the recommendations of MRP-227, as approved by the NRC, and shall provide a justification for any deviation which includes a consideration of 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:

Program Description 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 reactor vessel internal (R VI) components.

This program is used to manage the effects of age-related degradation mechanisms that are applicable in general to the PWR RVI components at the facility. These aging effects include (a) various forms of cracking, including stress corrosion cracking (SCC), which also encompasses primary water stress corrosion cracking (PWSCC), irradiation assisted stress corrosion cracking (IASCC), or cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss offracture toughness due to either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling; and (e) loss ofpreload due to thermal and irradiation-enhanced stress relaxation or creep.

The program applies the guidance in MRP-22 7for inspecting, evaluating, and, if applicable, dispositioning non-conforming R VI components at the facility. The program conforms to the definition of a sampling-based condition monitoring program, as defined by the Branch Technical Position RSLB-1, with periodic examinations and other inspections of highly-affected internals locations. These examinations provide reasonable assurance that the effects of age-related degradation mechanisms will be managed during the period of extended operation. The program includes expanding periodic examinations and other inspections if the extent of the degradation effects exceeds the expected levels.

The MRP-227 guidance for selecting R VI components for inclusion in the inspection sample is based on a four step ranking process. Through this process, the reactor internals for all three PWR designs were assigned to one of the following four groups:

Primary, Expansion, Existing Programs, and No Additional Measures components.

Definitions of each group are provided in GALL Chapter IX.B.

The result of this four-step sample selection process is a set of Primary Internals Component locations for each of the three plant designs that are expected to show the leading indications of the degradation effects, with another set of Expansion Internals Component locations that are specified to expand the sample should the indications be more severe than anticipated. The degradation effects in a third set of internals locations are deemed to be adequately managed by 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 additional measures. As a result, the program typically identifies 5 to 15% of the R VI locations as Primary Component locations for inspections, with another 7 to 10% of the R VI locations to be inspected as Expansion Components, as warranted by the evaluation of the inspection results. Another 5 to 15% of the internals locations are covered by Existing Programs, with the remainder requiring no additional measures. This process thus uses appropriate component functionality criteria, age related degradation susceptibility criteria, and failure consequence criteria to identify the components that will be inspected under the program in a manner that conforms to the sampling criteria for sampling-based condition monitoring programs in Section A. 1.2.3.4 of NRC Branch Position RLSB-i.

Consequently, the sample selection process is adequate to assure that the intended function(s) of the PWR reactor internal components are maintained during the period of extended operation.

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

However, the program 's adoption of the MRP-22 7 guidance for visual examinations goes beyond the ASME Code,Section XI visual examination criteria because additional guidance is incorporated into MRP-22 7 to clarify how the particular visual 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-related degradation effects.

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 technical justifications has demonstrated the indication detection capability of at least two vendors, and where vendor technical justification is a requirement prior to any additional bolting examinations. Specifically, the capability of program's UT volumetric methods to detect loss of integrity ofPWR internals bolts, pins, and fasteners, such as baffle-former bolting in B& W and Westinghouse units, has been well demonstrated by operating experience. In addition, the program 's adoption of the MRP-22 7 guidance and process incorporates the UT criteria in MRP-228, which calls for the technical justifications that are neededfor volumetric examination method demonstrations, required by the ASME Code,Section V.

The program also includes future industry operating experience as incorporated in periodic revisions to MRP-22 7. The program thus provides reasonable assurance for the long-term integrity and safe operation of reactor internals in all commercial operating U.S, PWR nuclear power plants.

Age-related degradation in the reactor internals is managed through an integrated program. Specific features of the integrated program are listed in the following ten program elements. Degradation due to changes in material properties (e.g., loss of fracture toughness) was considered in the determination of inspection recommendations and is managed by the requirement to use appropriately degraded properties in the Page 27 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program evaluation of identified defects. The integrated program is implemented by the applicant through an inspection plan that is submitted to the NRC for review and approval with the application for license renewal.

11 Refer to the GALL Report, Chapter 1, for applicability of various editions 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 and guidance in the most recently NRC endorsed version of MRP-22 7, which provides augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants designed by B& W, CE, and Westinghouse. The scope of components considered for inspection under MRP-227 guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code,Section XI), those R V1 components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other R VI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4 (a) (1) (i), (ii), or (iii). The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review (AMR),

as defined by the criteria set in 10 CFR 54.21(a)(1). The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class I appurtenances to the reactor vessel and are adequately managed in accordance with an applicant's AMP that corresponds to GALL AMP XI.M1, "ASME Code, Section Xl Inservice Inspection, Subsections IWB, IWC, and IWD. "

The scope of the program includes the response bases to applicable license renewal applicant action 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's R 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 that formed the basis of the MRP's augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 of MRP-22 7), and NSSS vendor-specific or plant-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 applicability limitations to base-loaded plants and the fuel loading management 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 and assumptions require a determination of applicability by the applicant for each reactor and 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 aging mechanisms (e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program. The program description, evaluation, and technical basis of water chemistry are presented in 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 Parameters Monitored/Inspected The following is NUREG-1801, Revision 2,Section XI.Ml6A, Evaluation and Technical Basis Element 3:

The program manages the following age-related degradation effects and mechanisms that are applicable in general to the R VI components at the facility: (a) cracking induced by SCC, P WSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss offracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling and irradiation growth, distortion, or deflection; and (e) loss ofpreload caused by thermal and irradiation enhanced stress relaxation or creep. For the management of cracking, the program monitors for evidence of surface breaking linear discontinuities if a visual inspection technique is used as the non-destruction examination (NDE) method, or for Page 29 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program relevant flaw presentation signals if a volumetric UT method is used as the NDE method.

For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components.

For the management of loss ofpreload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss offracture toughness that is induced by thermal aging or neutron irradiation embrittlement, or by void swelling and irradiation growth; instead, the impact of loss offracture toughness on component integrity is indirectly managed by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable reduced fracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation under the MRP-22 7 guidance or ASME Code, Section X1 requirements. The program uses physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.

Specifically, the program implements the parameters monitored/inspected criteria for Westinghouse designed Primary Components in Table 4-3 of MRP-227. Additionally, the program implements the parameters monitored/inspected criteria for Westinghouse designed Expansion Components in Table 4-6 of MRP-227. The parameters monitored/inspected for Existing Program Components follow the bases for referenced Existing Programs, such as the requirements for ASME Code Class R V1 components in ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-N-3, as implemented through the applicant's ASME Code,Section XI program, or the recommended program for inspecting Westinghouse designed flux thimble tubes in GALL AMP XI. M3 7, "Flux Thimble Tube Inspection. "No inspections, except for those specified in ASME Code,Section XI, are required for components that are identified as requiring "No Additional Measures, " in accordance with the analyses reported in 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.

Page 30 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program 5.4 Detection of Aging 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 introductory discussion and justification of the examination methods selected for detecting the aging effects of interest; and (b) standards for examination methods, procedures, and personnel are provided in a companion document, MRP-228. In all cases, well-established methods were selected. These methods include volumetric UT examination methods for detecting flaws in bolting, physical measurements for detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by 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 applied for the detection of cracking only when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss ofpreload caused by thermal and irradiation-enhanced stress relaxation and creep.

In addition, the program adopts the recommended guidance in MRP-227 for defining the Expansion criteria that need to be applied to inspections of Primary Components and Existing Requirement Components and for expanding the examinations to include additional Expansion Components. As a result, inspections performed on the R VI components are performed consistent with the inspection frequency and sampling bases for Primary Components, Existing Requirement Components, and Expansion Components in MRP-227, which have been demonstrated to be in conformance with the inspection criteria, sampling basis criteria, and sample Expansion criteria in Section A. 1.2.3.4 of NRC Branch Position RLSB-1.

Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for Westinghouse designed Primary Components in Table 4-3 of MRP-227 and for Westinghouse designed Expansion Components in Table 4-6 ofMRP-227.

The program is supplemented by the following plant specific Primary Component and Expansion Component inspections for the program (as applicable): there are no supplemental Primary or Expansion components for the CNP program.

Page 31 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program In addition, in some cases (as defined in MRP-22 7), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss ofpreload due to stress relaxation, or for changes in dimension due to void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program include the measurement of the 304 SS hold down springs in accordance with 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 Monitoring and 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 result from the program's inspections are given in Section 6 of MRP-22 7 and its subsections. The evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as wellfor performing applicable limit load, linear elastic and elastic plastic fracture analyses of relevant flaw indications. The examinations and re-examinations required by the MRP-227 guidance, together with the requirements specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects of the age-related degradation mechanisms 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 potential for inclusion of Expansion Component locations if the effects are greater than anticipated, plus the continuation of the Existing Programs activities, such as the ASME Code, Section XL, Examination Category B-N-3 examinations for core support structures, provides a high degree of confidence in the total program.

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.

Page'32 of 69

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 examination acceptance criteria for the Primary and Expansion Component examinations. For components addressed-by examinations referenced to ASME Code,Section XI, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.

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

" For visual examination (and surface examination as 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 disposition surface breaking indications that are detected and sized for length by VT-1/E VT-1 examinations;

" For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment of bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and For physical measurements, the examination acceptance criterion for the acceptable tolerance in the measured differential height from 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 criterion for physical measurements performed on the height limits of the Westinghouse-designed hold-down springs will be developed prior 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.

Page 33 of 69

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

Corrective actions following the detection of unacceptable conditions are fundamentally provided for in each plant's corrective action program. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found 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 available for disposition of detected conditions that exceed the examination acceptance criteria of Section 5 of the report. These include engineering evaluation methods, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures. The latter are subject to the requirements of 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-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Other alternative corrective action bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Examples ofpreviously NRC-endorsed alternative corrective actions bases include those corrective actions 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, or for B& W-designed R VI components in B& W Report No. BA W-2248. Westinghouse Report No.

WCAP-14577-Rev. 1-A was endorsed for use in an NRC SE to the Westinghouse Owners Group, dated February 10, 2001. B&W Report No. BA W-2248 was endorsed for use in an SE to Framatome Technologies on behalf of the B& W Owners Group, dated December 9, 1999. Alternative corrective action bases not approved or endorsed by the NRC will be submitted for NRC approval prior 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.

Page 34 of 69

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

Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, 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 quality for inspection, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with the 10 CFR Part 50, Appendix B, or their equivalent (as applicable), confirmation process, and administrative controls.

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 administrative controls for such programs, including their implementing procedures and review and approval processes, 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 established with 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 Operating Experience The following is NUREG-l1801, Revision 2,Section XI.M 16A, Evaluation and Technical Basis Element 10:

Page 35 of 69

Donald C. Cook Nuclear Plant Reactor Vessel Internals Aging Management Program Relatively few incidents ofPWR internals aging degradation have been reported in operating U.S. commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-22 7-A. The applicant is expected to review subsequent operating experience for impact on its program or to participate in industry initiatives that perform this function.

The application of the MRP-227 guidance will establish a considerable amount of operating experience over the next few years. Section 7 of MRP-22 7 describes the reporting requirements for these applications, and the plan for evaluating the accumulated additional operating experience.

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.

Page 36 of 69

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.

Page 37 of 69

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 Materials Reliability Program: Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

6.29 Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

6.30 Materials Reliability Program: Inspection Standard for 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 Effect Expansion Link Examination Item Applicability (Mechanism)

(Note 1)

Method/Frequency (Note 1)

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 Effect Expansion Link Examination Item Applicability (Mechanism)

(Note 1)

Method/Frequency (Note 1)

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 in Table 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 in some Westinghouse plant designs.

6.

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

Page 44 of 69

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 Method/Frequency Examination Coverage Item Applicability (Mechanism)

(Note 1)

(Note 1)

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.

Page 46 of 69

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)

Page 47 of 69

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

(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 (I E) 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).

Page 48 of 69

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.

Page 49 of 69

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.

Page 50 of 69

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.

Page 52 of 69

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 The specific relevant column bodies (non than two inches in the upper outlet nozzle weld and condition is a detectable cast) core barrel flange weld shall lower support column body 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.

Page 53 of 69

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.

Page 54 of 69

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.

Page 55 of 69

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.

Page 56 of 69

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.

Page 57 of 69

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.

Page 58 of 69

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 i

4-LP Upper Support Plate Lower Guide tube 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 Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Circumferential Weld Core Barrel to Support Plate Weld Axial Weld Lower Barrel Axial Weld Lower Barrel Axial Weld Thermal Shield Flexure Figure E-3 Major Core Barrel Welds Page 61 of 69

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

't 0)

_0W a) 0

ýEE CO 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 DIFFUSER PLATE CORE SUPPORT COLUMN 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)

Page 65 of 69

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.

COMMITMENT SCHEDULED 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 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.

Unit 1: October 25, 2014 Unit 2: December 23, 2017