ML14030A134

WCAP-17789-NP, Rev. 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals.

ML14030A134
Person / Time
Site:	Beaver Valley
Issue date:	01/27/2014
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
References
L-13-398 WCAP-17789-NP, Rev. 1
Download: ML14030A134 (103)

Text

Enclosure A L-13-398 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Beaver Valley Power Station Unit 1, Revision 1 (102 pages follow)

Westinghouse Non-Proprietary Class 3 WCAP-17789-NP January 2()14 Revision 1 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Beaver Valley Power Station Unit 1 Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17789-NP Revision 1 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Beaver Valley Power Station Unit 1 Stephen M. Parker*

Reactor Internals Aging Management Daniel B. Denis*

Materials Center of Excellence Joshua K. McKinley*

Materials Center of Excellence January 2014 Approved: Patricia C. Paesano*, Manager Reactor Internals Aging Management

• Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066

© 2014 Westinghouse Electric Company LLC All Rights Reserved

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii TABLE OF CONTENTS LIST OF TABLES ........................................................................................................................................ v LIST OF FIGURES ..................................................................................................................................... vi LIST OF ACRONYM S ............................................................................................................................... vii ACKNOW LEDGEM ENTS ......................................................................................................................... ix P U R PO S E ..................................................................................................................................... 1-1 2 BACKGROUND .......................................................................................................................... 2-1 3 SITE PW R VESSEL INTERNALS PROGRAM OW NER ......................................................... 3-1 3.1 Site Vice President ........................................................................................................... 3-1 3.2 Director Site Engineering ................................................................................................ 3-1 3.3 M anager Site Technical Services Engineering ................................................................ 3-1 3.4 M anager Site Design Engineering ................................................................................... 3-1 3.5 M anager Site Chemistry .................................................................................................. 3-2 3.6 Site PW R Vessel Internals Program Owner .................................................................... 3-2 3.7 Fleet PW R Vessel Internals Program Owner ................................................................... 3-3 3.8 Outage M anagement ........................................................................................................ 3-3 4 DESCRIPTION OF THE BEAVER VALLEY UNIT 1 REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS ............................................. 4-1 4.1 Existing Beaver Valley Unit I Programs ......................................................................... 4-3 4.1.1 ASME Section XI Inservice Inspection Subsections IWB, IWC, and IWD P ro g ram ........................................................................................................... 4-4 4.1.2 Flux Thimble Tube Inspection Program .......................................................... 4-4 4.1.3 Primary W ater Chemistry Program ................................................................. 4-4 4.2 Supporting Beaver Valley Unit 1 Programs and Aging Management Supportive Plant Enhancements .................................................................................................................. 4-5 4.2.1 Reactor Internals Aging M anagement Review Process ................................... 4-5 4.2.2 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) .................................................................................... 4-5 4.2.3 Control Rod Guide Tube Support Pin Replacement Project ........................... 4-6 4.3 Industry Programs ............................................................................................................ 4-6 4.3.1 W CAP-14577, Aging M anagement for Reactor Internals ............................... 4-6 4.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines ........... 4-6 4.3.3 W CAP-1745 1-P, Reactor Internals Guide Tube W ear .................................. 4-10 4.3.4 Ongoing Industry Programs ........................................................................... 4-11 4.4 Summary ........................................................................................................................ 4-11 5 BEAVER VALLEY REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES .............................................................................................................................. 5-1 WCAP- I7789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv 5.1 GALL Revision 2 Element 1: Scope of Program ........................................................... 5-1 5.2 GALL Revision 2 Element 2: Preventive Actions .......................................................... 5-3 5.3 GALL Revision 2 Element 3: Parameters Monitored or Inspected ................................ 5-4 5.4 GALL Revision 2 Element 4: Detection of Aging Effects ............................................. 5-5 5.5 GALL Revision 2 Element 5: Monitoring and Trending .............................................. 5-10 5.6 GALL Revision 2 Element 6: Acceptance Criteria ...................................................... 5-1 1 5.7 GALL Revision 2 Element 7: Corrective Actions ........................................................ 5-12 5.8 GALL Revision 2 Element 8: Confirmation Process .................................................... 5-14 5.9 GALL Revision 2 Element 9: Administrative Controls ................................................ 5-14 5.10 GALL Revision 2 Element 10: Operating Experience ................................................. 5-15 6 D EM O N STR A TION .................................................................................................................... 6-1 6.1 Demonstration of Topical Report Conditions Compliance to SE on MRP-227, R ev isio n 0 ........................................................................................................................ 6-2 6.2 Demonstration of Applicant/Licensee Action Item Compliance to SE on MRP-227, Rev ision 0 ........................................................................................................................ 6 -3 6.2.1 SE Applicant/Licensee Action Item I: Applicability of FMECA and Functionality A nalysis A ssum ptions ............................................................... 6-3 6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal ............................................................. 6-5 6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Program s ............................................................................. 6-6 6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper F lange Stress Relief ......................................................................................... 6-7 6.2.5 SE Applicant/Licensee Action Item 5: Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components

......................................................................................................................... 6 -7 6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W C o m po nents ..................................................................................................... 6-8 6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS M ate ria ls .......................................................................................................... 6 -8 6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff R eview and A pproval .................................................................................... 6-11 7 PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE .......................... 7-1 8 IM PLEM EN TIN G D O CU M EN TS .............................................................................................. 8-1 9 REF ER E N C E S ............................................................................................................................. 9-1 APPENDIX A ILLUSTRATIONS .................................................................................................... A-1 APPENDIX B BEAVER VALLEY UNIT I LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

TABLE ................................................................................ B-1 APPENDIX C MRP-227-A AUGMENTED INSPECTIONS ......................................................... C-1 WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 V LIST OF TABLES Table 6-1 Topical Report Condition Compliance to SE on MRP-227 ............................................. 6-2 Table 6-2 Summary of BV Unit I CASS Components and their Susceptibility to TE .................. 6-10 Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary .7-1 Table B-I Beaver Valley Unit I LRA Aging Management Review Summary .......................... B-1 Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-D esigned Internals .................................................................................................... C -1 Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-D esigned Internals .......................................................................................................... C -7 Table C-3 MRP-227-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals ....................................... C-10 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for W estinghouse-D esigned Internals ................................................................................ C- 12 WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vi LIST OF FIGURES Figure A-i Illustration of Typical Westinghouse Internals Assembly .............................................. A-I Figure A-2 Typical Westinghouse Control Rod Guide Card ................................................................... A-2 Figure A-3 Lower Section of Control Rod Guide Tube Assembly .......................................................... A-3 Figure A-4 M ajor C ore B arrel W elds ...................................................................................................... A -4 Figure A-5 Bolting Systems used in Westinghouse Core Baffles ........................................................... A-5 Figure A-6 Core Baffl e/Barrel Structure ................................................................................................. A -6 Figure A-7 Bolting in a Typical Westinghouse Baffle-Former Structure ................................................ A-7 Figure A-8 Vertical Displacement between the Baffle Plates and Bracket at the Bottom of the Baffle-Form er-B arrel A ssem bly .................................................................................................... A -8 Figure A-9 Schematic Cross-Sections of the Westinghouse Hold Down Springs ................................... A-9 Figure A -10 Typical Therm al Shield Flexure .......................................................................................... A -9 Figure A-11 Lower Core Support Structure .................................................................................... A-10 Figure A-12 Lower Core Support Structure - Core Support Plate Cross-Section ................................. A-Il Figure A-13 Typical Core Support Column .................................................................................... A- 1I Figure A-14 Examples of BMI Column Designs ................................................................................... A-12 WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii LIST OF ACRONYMS AMP Aging Management Program Plan AMR Aging Management Review ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox BMI bottom-mounted instrumentation BV Beaver Valley BVPS Beaver Valley Power Station BWR boiling water reactor CASS cast austenitic stainless steel CE Combustion Engineering CFR Code of Federal Regulations CLB current licensing basis CRGT control rod guide tube ECP Engineering Change Package EFPY effective full-power years EPRI Electric Power Research Institute ET electromagnetic testing (eddy current)

EVT enhanced visual testing (a visual NDE method that includes EVT-1)

FENOC FirstEnergy Nuclear Operating Company FMECA failure modes, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC irradiation-assisted stress corrosion cracking INPO Institute of Nuclear Power Operations ISI inservice inspection ISR irradiation-enhanced stress relaxation LRA License Renewal Application LRAAI license renewal applicant action items MRP Materials Reliability Program NDE nondestructive 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 OER Operating Experience Report PH precipitation-hardenable (heat treatment)

PWR pressurized water reactor PWROG Pressurized Water Reactor Owners Group (formerly WOG)

PWSCC primary water stress corrosion cracking WCAP- 17789-NP January 2014 Revision I

viii WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF ACRONYMS (cont.)

QA quality assurance RCC rod cluster control RCS Reactor Coolant System RIS Regulatory Issue Summary RO refueling outage RV reactor vessel RVI reactor vessel internals SCC stress corrosion cracking SE Safety Evaluation SER Safety Evaluation Report SRP Standard Review Plan SS stainless steel TE thermal embrittlement UFSAR Updated Final Safety Analysis Report UT ultrasonic testing (a volumetric NDE method)

VT visual testing (a visual NDE method that includes VT- I and VT-3)

WANO World Association of Nuclear Operators WOG Westinghouse Owners Group XL extra-long Westinghouse fuel Trademark Statement:

INCONEL is a registered trademark of Special Metals, a Precision Castparts Corp. company.

WCAP-17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix ACKNOWLEDGEMENTS The authors would like to thank Wesley Williams and Zach Warchol of FirstEnergy Nuclear Operating Company and our associates at Westinghouse for their efforts in supporting development of this WCAP.

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1 1 PURPOSE The purpose of this report is to document the Beaver Valley Power Station (BVPS) Unit 1, hereafter Beaver Valley (BV) Unit 1, Reactor Vessel Internals (RVI) Aging Management Program Plan (AMP).

The purpose of the AMP is to manage the effects of aging on reactor vessel internals through the license renewal period. BV Unit I enters the license renewal period on January 29, 2016. This document provides a description of the program as it relates to the management of aging effects identified in various regulatory and updated industry-generated documents in addition to the program documented in BV Unit 1 operating procedure NOP-CC-5004 [1] in support of license renewal program evaluations. This AMP is supported by existing BV Unit I documents and procedures and, as needed by industry experience or directive in the future, will be updated or supported by additional documents to provide clear and concise direction for the effective management of aging degradation in reactor internals components. These actions provide assurance that operations at BV Unit I will continue to be conducted in accordance with the current licensing basis (CLB) for the reactor vessel internals by fulfilling License Renewal commitments [2], United States (U.S.) Nuclear Regulatory Commission (NRC) expectations in the Regulatory Issue Summary (RIS) [3], American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI Inservice Inspection (ISI) programs [4], and industry requirements [5]. This AMP fully captures the intent of the additional industry guidance for reactor internals augmented inspections, based on the programs sponsored by U.S. utilities through the Electric Power Research Institute (EPRI) managed Materials Reliability Program (MRP) and the Pressurized Water Reactor Owners Group (PWROG).

The main objectives for the BV Unit I RVI AMP are to:

• Demonstrate that the effects of aging on the RVI will be adequately managed for the period of extended operation in accordance with 10 CFR 54 [6].

• Summarize the role of existing BV Unit 1 AMPs in the RVI AMP.

• Define and implement the industry-defined (EPRI/MRP and PWROG) pressurized water reactor (PWR) RVI requirements and guidance for managing aging of reactor internals.

• Provide an inspection plan summary for the BV Unit 1 reactor internals.

BV Unit I License Renewal Commitment 18 [2], "PWR Vessel Internals Program," commits BV Unit I to:

1. Participatein the industry programsapplicable to BVPS Unit 1 for investigatingand managing aging effects on reactor internals;

2. Evaluate and implement the results of the industryprograms as applicable to the B VPS Unit I reactor internals;and, WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-2

3. Upon completion of these programs, but not less than 24 months before entering the period of extended operations,submit an inspectionplanfor the BVPS Unit I reactor internals to the NRC for review and approval.

Augmented inspections, based on required program enhancements resulting from industry programs, will become part of the BV Unit I ASME B&PV Code,Section XI program [4]. Corrective actions for augmented inspections will be developed using repair and replacement procedures equivalent to those requirements in ASME B&PV Code,Section XI, or as determined independently by FirstEnergy Nuclear Operating Company (FENOC), or in cooperation with the industry, to be equivalent or more rigorous than currently defined procedures.

This AMP for the BV Unit 1 reactor internals demonstrates that the program adequately manages the effects of aging for reactor internals components and establishes the basis for providing reasonable assurance that the internals components will continue to perform their intended function through the BV Unit I license renewal period of extended operation. This WCAP supports the BV Unit I License Renewal Commitment 18 which includes a submission to the U.S. Nuclear Regulatory Commission (NRC) of an inspection plan for the PWR Vessel Internals Program, as it would be implemented from the participation of BV Unit I in industry initiatives, 24 months prior to entering the period of extended operation. The implementation schedule for this commitment requires submission to the NRC no later than January 29, 2014.

The development and implementation of this program meets the guidelines provided in the RIS [3].

January 2014 WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-1 2 BACKGROUND The management of aging degradation effects in reactor internals is required for nuclear plants considering or entering license renewal, as specified in the NRC Standard Review Plan [7]. The U.S.

nuclear power industry has been actively engaged in recent years in a significant effort to support the industry goal of responding to these requirements. Various programs have been underway within the industry over the past decade to develop guidelines for managing the effects of aging within PWR reactor internals. In 1997, the WOG issued WCAP-14577 [8], "License Renewal Evaluation: Aging Management for Reactor Internals," which was reissued as Revision 1-A in 2001 after receiving NRC Staff review and approval. Later, an effort was engaged by the EPRI MRP to address the PWR internals aging management issue for the three currently operating U.S. reactor designs - Westinghouse, Combustion Engineering (CE), and Babcock & Wilcox (B&W).

The MRP first established a framework and strategy for the aging management of PWR internals components using proven and familiar methods for inspection, monitoring, surveillance, and communication. Based upon that framework and strategy, and on the accumulated industry research data, the following elements of an Aging Management Program were further developed [8, 9]:

Screening criteria were developed, considering chemical composition, neutron fluence exposure, temperature history, and representative stress levels, for determining the relative susceptibility of PWR internals components to each of eight postulated aging mechanisms (further discussed in Section 4 of this Program).

PWR internals components were categorized, based on the screening criteria, into categories that ranged from:

- Components for which the effects from the postulated aging mechanisms are insignificant,

- Components that are moderately susceptible to the aging effects, and

- Components that are significantly susceptible to the aging effects.

Functionality assessments were performed based on representative plant designs of PWR internals components and assemblies of components using irradiated and aged material properties, to determine the effects of the degradation mechanisms on component functionality.

Aging management strategies were developed combining the results of functionality assessment with several contributing factors to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections. Items considered included component accessibility, operating experience (OE), existing evaluations, and prior examination results.

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-2 The industry guidance is contained within two separate EPRI MRP documents:

MRP-227-A [5], "PWR Internals Inspection and Evaluation Guidelines," (hereafter referred to as the "I&E Guidelines" or simply "MRP-227-A") provides the industry background, listing of reactor internals components requiring inspection, type of NDE required for each component, timing for initial inspections, and criteria for evaluating inspection results. MRP-227-A provides a standardized approach to PWR internals aging management for each unique reactor design (Westinghouse, B&W, and CE).

MRP-228 [10], "Inspection Standard for PWR Internals," provides guidance on the qualification/demonstration of the NDE techniques and other criteria pertaining to the actual performance of the inspections.

The PWROG has also developed WCAP-17096-NP, Revision 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements" for the MRP-227 inspections, where feasible [11]. This document has been submitted to the NRC for review and approval, and will be updated to incorporate changes from MRP-227-A [5]. Final reports are to be developed and available for industry use in support of planned license renewal inspection commitments. In some cases, individual plants will develop plant-specific acceptance criteria for some internals components where a generic approach is not practical.

The BV Unit I reactor internals are integral with the reactor coolant system (RCS) of a Westinghouse three-loop nuclear steam supply system (NSSS), a typical illustration of which is provided in Figure A-1.

As described in NUREG-1929 [2], the BV Unit 1 consist of three major assemblies: the lower core support structure (also known as the "lower internals"), the upper core support structure (also known as the "upper internals"), and the in-core instrumentation support structure (includes components that are part of the "upper internals" or the "lower internals"). These assemblies provide a number of functions, such as: core support; aligning, guiding and limiting movement of core components; directing coolant flow; and, providing shielding.

The lower core support structure assembly consists of the core barrel, the core baffle, the lower core plate and support columns, the thermal shield or neutron shield pads, and the core support welded to the core barrel. A ledge in the reactor vessel supports the lower core support structure at its upper flange and a radial support system attached to the vessel wall restrains its lower end from transverse motion. Within the core barrel, an axial baffle and a lower core plate are attached to the core barrel wall and form the enclosure periphery of the assembled core. The lower core support structure and core barrel control and provide passageways for coolant flow. The lower core plate, positioned at the bottom level of the core below the baffle plates, supports and orients the fuel assemblies.

Unit I uses a one-piece thermal shield fixed to the core barrel at the top with rigid bolted connections.

Rectangular specimen guides, welded to the outside of the thermal shield for insertion and irradiation of material samples during reactor operation, extend to the top of the thermal shield.

The upper core support assembly consists of the upper support assembly and the upper core plate, between which, are support columns and rod cluster control (RCC) guide tube assemblies. The support columns establishing the spacing between the upper support assembly and the upper core plate are WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 2-3 fastened at the top and bottom to these plates. They transmit mechanical loadings between the upper support and upper core plate and serve as thermocouple passageways.

The RCC guide tube assemblies that shield and guide the control rod drive shafts and control rods are fastened to the upper support and oriented and supported by pins in the upper core plate. The upper guide tube attached to the upper support plate and guide tube also guides the control rod drive shafts.

The in-core instrumentation support structures consist of an upper system (components of which are parts of the "upper internals") to support thermocouples penetrating the vessel through the head and a lower system (components of which are part of the "lower internals") to support flux thimbles penetrating the vessel through the bottom.

The upper system has instrumentation port columns, slip-connected to in-line columns fastened, in turn, to the upper support plate. The thermocouples, conveyed through these port columns and the upper support plate at positions, are above their readout locations.

The lower in-core instrumentation support system uses reactor vessel bottom-mounted instrumentation columns (flux thimble guide tubes) which guide and protect the retractable, cold-worked stainless steel flux thimbles that are pushed upward into the reactor core. The thimbles, closed at the leading ends, are the pressure boundary between the reactor pressurized water and the containment atmosphere. All reactor vessel internals are removable for their inspection, and for inspection of the vessel internal surface.

BV Unit 1 was granted a license for extended operation by the NRC through the issuance of a safety evaluation report (SER) in NUREG-1929 [2]. In the SER, the NRC concluded that the BV Unit 1 License Renewal Application (LRA) adequately identified the RV internals components within the scope of license renewal, as required by 10 CFR 54.4(a), and those subject to an AMR, as required by 10 CFR 54.21(a)(1) [6] and; therefore, is acceptable. A listing of the BV Unit I reactor vessel internals components and subcomponents, already reviewed by the NRC in the SER that are subject to AMP requirements, is included in Table B-I.

In accordance with 10 CFR Part 54 [6], frequently referred to as the License Renewal Rule, BV Unit 1 has developed a program to direct the performance of aging management reviews of mechanical structures and components [12]. The U.S. industry, as noted through the efforts of the MRP and PWROG, has further investigated the components and subcomponents that require aging management to support continued reliable function. As designated by the protocols of NEI 03-08 [13], "Guidelines for the Management of Materials Issues", each plant will be required to use MRP-227-A and MRP-228 to develop and implement an AMP for reactor internals no later than three years after the initial industry issuance of MRP-227, Revision 0. MRP-227, Revision 0 was issued in December 2008, and plant AMPs must therefore be completed by December 2011, or sooner, if required by plant-specific License Renewal Commitments. According to [3], BV Unit 1 is a Category B plant that is expected to submit their RVI AMP based on the guidance of MRP-227-A, consistent with their commitments. Per the LRA [2], BV Unit 1 has a commitment to submit their AMP for approval by the NRC no later than January 29, 2014.

The information contained in this AMP fully complies with the requirements and guidance of the referenced documents. The AMP will manage aging effects of the RVI so that the intended functions will be maintained consistent with the current licensing basis for the period of extended operation.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 SITE PWR VESSEL INTERNALS PROGRAM OWNER The PWR Vessel Internals Program [1] manages the effects of age-related degradation mechanisms of reactor vessel internals. The successful implementation and comprehensive long-term management of the BV Unit I RVI AMP will require the integration of FirstEnergy organizations, corporately and at Beaver Valley, and interaction with multiple industry organizations including, but not limited to, the ASME, MRP, NRC, and PWROG. The responsibilities of the individual FirstEnergy corporate and Beaver Valley groups are provided in the following paragraphs. FENOC will maintain cognizance of industry activities related to PWR internals inspection and aging management, and will address/implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices.

The overall responsibility for administration of the RVI AMP is the Site Manager of Technical Services Engineering.

Additional responsibilities and the appropriate responsible personnel, as described in [1], are discussed in the following subsections.

3.1 SITE VICE PRESIDENT Has responsibility for ensuring that sufficient financial and manpower resources are made available to effectively and efficiently implement the PWR RVI Program at the site.

3.2 DIRECTOR SITE ENGINEERING Has responsibility for and sponsorship of the site PWR RVI Program which includes examination, repair, mitigation, reporting, and results trending.

3.3 MANAGER SITE TECHNICAL SERVICES ENGINEERING Is responsible for the development, implementation, and maintenance of the Site PWR RVI Program.

3.4 MANAGER SITE DESIGN ENGINEERING Maintains overall design authority for PWR RVI and its associated components.

Maintains overall design authority for safety analyses related to the PWR RVI and its associated components.

Ensures development and completion of Engineering Change Packages (ECPs) that may be required for the implementation of mitigation and/or replacement activities.

Provides support for the completion of assessments, evaluations and analyses for the PWR RVI and its associated components/materials as requested or assigned.

• Ensures the design drawings related to the PWR RVI are maintained.

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-2 3.5 MANAGER SITE CHEMISTRY Ensures primary system Chemistry controls are adequately implemented, maintained and comply with regulatory requirements and appropriate industry guidelines.

Ensures that Chemistry program changes required by regulation, fostered by industry guidance documents or identified as prudent for maintaining RCS integrity and reliability are evaluated and implemented as appropriate in a timely manner.

3.6 SITE PWR VESSEL INTERNALS PROGRAM OWNER Ensures coordination of the PWR RVI Program activities among other departments and/or interfacing/affected site programs.

Ensures that examination, repair and assessment activities of affected components/materials comply with regulatory commitments and Industry guidance provided by the EPRI MRP, PWR Owners Groups, Institute of Nuclear Power Operations (INPO), World Association of Nuclear Operators (WANO), and/or other appropriate Industry organizations Ensures preparation and review of PWR RVI related program documents, license amendment requests, relief requests, required reports, and other documents submitted to the NRC.

Ensures Industry experience regarding PWR RVI and PWR RVI materials/components are reviewed and that any site program revisions are implemented in a timely manner.

Coordinates the generation and maintenance of the site-specific PWR RVI Inspection/Implementation Plan. Maintenance of the Inspection/Implementation Plan includes periodic reviews to ensure that the plan reflects current industry experience and data, including advancements in mitigation capabilities and strategies.

Ensures that the examinations detailed in the site PWR RVI Program inspection plan are performed at the prescribed times and frequency.

Evaluates the effects of changes to other interfacing site programs on the PWR RVI Program.

Interfacing programs/groups may include but are not limited to:

o Inservice Inspection (ISI) Program o Reactor Engineering o Nuclear Fuels and Core Design Performs and coordinates strategic planning for PWR RVI Program Components. Strategic planning includes, but may not be limited to, planning for examinations and mitigation activities necessary to lessen the detrimental consequences associated with PWR RVI degradation.

Ensures dissemination of appropriate PWR RVI Program experience and information to other Site, FENOC and industry groups.

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-3 Completes the program health report in accordance with NOP-ER-2101 [141 for the PWR RVI Program.

3.7 FLEET PWR VESSEL INTERNALS PROGRAM OWNER

• Has overall responsibility for and sponsorship of the FENOC PWR Vessel Internals Program.

a Facilitates communication and coordination between the FENOC PWR sites regarding PWR Vessel Internals issues.

0 Facilitates communication of industry information and experience regarding PWR Vessel Internals issues to and from the FENOC PWR sites.

• Initiates and coordinates the review and evaluation of industry guidance documents related to PWR vessel internals issues.

• Provides an interface to the EPRI Materials Reliability Program (MRP) in accordance with reference NOBP-SS-7000, EPRI Committee and User Group Member Expectations [15], and NOP-CC-5001, Materials Degradation Management Program [16].

• Provides oversight of the site PWR vessel internals programs to ensure their effectiveness. This includes coordination of periodic program self-assessments.

3.8 OUTAGE MANAGEMENT Ensures outage schedules needed to support PWR Vessel Internals Program activities, including implementation of ECPs, are complete and maintained.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-1 4 DESCRIPTION OF THE BEAVER VALLEY UNIT 1 REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS The U.S. nuclear industry, through the combined efforts of utilities, vendors, and independent consultants, has defined a generic guideline to assist utilities in developing reactor internals plant-specific aging management programs based on inspection and evaluation. The intent of this program is to ensure the long-term integrity and safe operation of the reactor internals components. FENOC has developed this AMP in conformance with the 10 Generic Aging Lessons Learned (GALL) [17] attributes and MRP-227-A [5].

This reactor internals AMP utilizes a combination of prevention, mitigation, and condition monitoring.

Where applicable, credit is taken for existing programs such as water chemistry [18], inspections prescribed by the ASME Section XI Inservice Inspection Program [4], thimble tube inspections [19], and mitigation projects such as support pin replacement [20], combined with augmented inspections or evaluations as recommended by MRP-227-A.

Aging degradation mechanisms that impact internals have been identified and documented in BV Unit 1 Aging Management Reviews [21 ] prepared using the business practice document [12] in support of the license renewal effort. The overall outcome of the reviews and the additional work performed by the industry, as summarized in MRP-227-A, is to provide appropriate augmented inspections for reactor internals components to provide early detection of the degradation mechanisms of concern. Therefore, this AMP is consistent with the existing BV Unit 1 AMR methodology and the additional industry work summarized in MRP-227-A. All sources are consistent and address concerns about component degradation resulting from the following eight material aging degradation mechanisms identified as affecting reactor internals:

Stress Corrosion Cracking (SCC)

Stress corrosion cracking (SCC) refers to local, nonductile cracking of a material due to a combination of tensile stress, environment, and metallurgical properties. The actual mechanism that causes SCC involves a complex interaction of environmental and metallurgical factors. The aging effect is cracking.

• Irradiation-Assisted Stress Corrosion Cracking Irradiation-assisted stress corrosion cracking (IASCC) is a unique form of SCC that occurs only in highly irradiated components. The aging effect is cracking.

• Wear Wear is caused by the relative motion between adjacent surfaces, with the extent determined by the relative properties of the adjacent materials and their surface condition. The aging effect is loss of material.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-2 Fatigue Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads and temperatures. After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack. Corrosion fatigue is included in the degradation description.

Low-cycle fatigue is defined as cyclic loads that cause significant plastic strain in the highly stressed regions, where the number of applied cycles is increased to the point where the crack eventually initiates. When the cyclic loads are such that significant plastic deformation does not occur in the highly stressed regions, but the loads are of such increased frequency that a fatigue crack eventually initiates, the damage accumulated is said to have been caused by high-cycle fatigue. The aging effects of low-cycle fatigue and high-cycle fatigue are additive.

Fatigue crack initiation and growth resistance are governed by a number of material, structural, and environmental factors such as stress range, loading frequency, surface condition, and presence of deleterious chemical species. Cracks typically initiate at local geometric stress concentrations such as notches, surface defects, and structural discontinuities. The aging effect is cracking.

• Thermal Aging Embrittlement Thermal aging embrittlement is the exposure of delta ferrite within cast austenitic stainless steel (CASS), martensitic stainless steel, and precipitation-hardenable (PH) stainless steel to high inservice temperatures, which can result in an increase in tensile strength, a decrease in ductility, and a loss of fracture toughness. Some degree of thermal aging embrittlement can also occur at normal operating temperatures for CASS, martensitic stainless steel, and PH stainless steel internals. CASS components have a duplex microstructure and are particularly susceptible to this mechanism. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.

Irradiation Embrittlement Irradiation embrittlement is also referred to as neutron embrittlement. When exposed to high-energy neutrons, the mechanical properties of stainless steel and nickel-based alloys can be changed. Such changes in mechanical properties include increasing yield strength, increasing ultimate strength, decreasing ductility, and a loss of fracture toughness. The irradiation embrittlement aging mechanism is a function of both temperature and neutron fluence. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-3 Void Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation of microscopic cavities in the material. These cavities result from the nucleation and growth of clusters of irradiation-produced vacancies. Helium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material. Void swelling may produce dimensional changes that exceed the tolerances on a component. Strain gradients produced by differential swelling in the system may produce significant stresses. Severe swelling

(>5 percent by volume) has been correlated with extremely low fracture toughness values. Also included in this mechanism is irradiation growth of anisotropic materials, which is known to cause significant dimensional changes within in-core instrumentation tubes that are fabricated from zirconium alloys. While the initial aging effect is dimensional change and distortion, severe void swelling may result in cracking under stress.

Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation-Enhanced Creep The loss of preload aging effect can be caused by the aging mechanisms of stress relaxation or creep. Thermal stress relaxation (or primary creep) is defined as the unloading of preloaded components due to long-term exposure to elevated temperatures, as seen in PWR internals. Stress relaxation occurs under conditions of constant strain where part of the elastic strain is replaced with plastic strain. Available data show that thermal stress relaxation appears to reach saturation in a short time (< 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />) at PWR internals temperatures.

Creep (or more precisely, secondary creep) is a slow, time- and temperature-dependent, plastic deformation of materials that can occur at stress levels below the yield strength (elastic limit).

Creep occurs at elevated temperatures where continuous deformation takes place under constant stress. Secondary creep in austenitic stainless steels is associated with temperatures higher than those relevant to PWR internals even after taking into account gamma heating. However, irradiation-enhanced creep (or more simply, irradiation creep) or irradiation-enhanced stress relaxation (ISR) is an athermal process that depends on the neutron fluence and stress, and it can also be affected by void swelling should it occur. The aging effect is a loss of mechanical closure integrity (or preload) that can lead to unanticipated loading that, in turn, may eventually cause subsequent degradation by fatigue or wear and result in cracking.

The BV Unit 1 RVI AMP is focused on meeting the requirements of the 10 elements of an aging management program as described in NUREG-1801, GALL Report Section XI.M16A for PWR Vessel Internals. In the BV Unit I RVI AMP, this is demonstrated through application of existing BV AMR methodology that credits inspections prescribed by the ASME Section XI Inservice Inspection Program, existing BV programs, and additional augmented inspections based on MRP-227-A recommendations. A description of the applicable existing BV programs and compliance with the elements of the GALL is contained in the following subsections.

4.1 EXISTING BEAVER VALLEY UNIT 1 PROGRAMS The overall strategy of FENOC for managing aging in reactor internals components is supported by the following existing programs [23]:

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-4

• ASME Section XI Inservice Inspection Subsections IWB, IWC, and IWD Program

• Flux Thimble Tube Inspection Program

• Primary Water Chemistry Program These are established programs that support the aging management of RCS components in addition to the RVI components. Although affiliated with and supporting the RVI AMP, they will be managed under the existing programs.

Brief descriptions of the programs are included in the following subsections.

4.1.1 ASME Section XI Inservice Inspection Subsections IWB, IWC, and IWD Program The BV Unit I ASME Section X1 Inservice Inspection, Subsections IWB, IWC, and IWD Program [4] is in accordance with ASME Section XI 2001 Edition with the 2003 Addenda [22] and is subject to the limitations and modifications of 10 CFR 50.55a. The program provides for condition monitoring of Class 1, 2, and 3 pressure-retaining components, including welds, pump casings, valve bodies, integral attachments, and pressure-retaining bolting. The program is updated as required by 10 CFR 50.55a.

The BV Unit I ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program is augmented by the Primary Water Chemistry Program [18] where applicable.

4.1.2 Flux Thimble Tube Inspection Program The BV Unit 1 Flux Thimble Tube Inspection Program [19] serves to identify loss of material due to wear prior to leakage by monitoring for and predicting unacceptable levels of wall thinning in the Movable Incore Detector System Flux Thimble Tubes, which serve as a Reactor Coolant System (RCS) pressure boundary. The program implements the recommendations of NRC IE Bulletin 88-09, Thimble Tube Thinning in Westinghouse Reactors [24].

The main attribute of the program is periodic nondestructive examination (NDE) of the flux thimble tubes which provides actual values of existing tube wall thinning. This information provides the basis for an extrapolation to determine when tube wall thinning will progress to an unacceptable value. Based on this prediction, preemptive actions are taken to reposition, replace or isolate the affected thimble tube prior to a pressure boundary failure.

4.1.3 Primary Water Chemistry Program The main objective of the Primary Water Chemistry Program [18] is to mitigate damage caused by corrosion and stress corrosion cracking. The Primary Water Chemistry Program relies on monitoring and control of water chemistry based on EPRI TR- 1014986, PWR Primary Water Chemistry Guidelines [25].

The One-Time Inspection Program will be used to verify the effectiveness of the Primary Water Chemistry Program for the circumstances identified in NUREG-1 801 that require augmentation of the Primary Water Chemistry Program.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-5 4.2 SUPPORTING BEAVER VALLEY UNIT 1 PROGRAMS AND AGING MANAGEMENT SUPPORTIVE PLANT ENHANCEMENTS 4.2.1 Reactor Internals Aging Management Review Process A comprehensive review of aging management of reactor internals was performed according to the requirements of the License Renewal Rule [6] as directed by BV business practice BVBP-LRP-0003, "Mechanical Screening, and Aging Management Review" [12]. License Renewal Project Document LRBV-MAMR-06B [21] documents the results of the aging management review performed in support of BV Unit I license renewal for reactor internals. The BV Unit 1 LRA was approved by the NRC in NUREG-1929 [2]. RVI components specifically noted as requiring aging management, as identified in the NUREG, are summarized in Appendix B Table B-1 of this AMP.

The AMR supported the LRA as follows:

1. Identified applicable aging effects requiringmanagement

2. Associated aging managementprograms to manage those aging effects

3. Identified enhancements or modifications to existing programs,new aging management programs, or any other actions requiredto support the conclusions reached in the review Aging management reviews were performed for each BV Unit 1 system that contained long-lived, passive components requiring aging management review, in accordance with BV business practice BVBP-LRP-0003 [12]. This review is not repeated here, but the results are fully incorporated into the BV Unit 1 RVI AMP.

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

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

Program is a new program that BV Unit I will implement prior to the period of extended operation. RVIs will be inspected in accordance with ASME Code Section XI, Subsection IWB, Category B-N-3. This inspection will be augmented to detect the effects of loss of fracture toughness due to thermal aging and neutron irradiation embrittlement of CASS components. The program will include identification of the limiting susceptible components from the standpoint of thermal aging susceptibility, neutron fluence, and cracking. For each identified component, aging management will be accomplished through either a supplemental examination or a component-specific evaluation, including a mechanical loading assessment. BV Unit 1 will participate in the EPRI Materials Reliability Program established to investigate the impacts of aging on PWR vessel internal components. The results of this project will provide additional basis for the inspections and evaluations performed under this program. Refer to Appendix B, Section B.2.40 of the LRA [23] for more information regarding this program.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-6 4.2.3 Control Rod Guide Tube Support Pin Replacement Project The control rod guide tube support pins are used to align the bottom of the control rod guide tube assembly into the top of the core plate. In general, SCC prevention is aided by adherence to strict primary water chemistry limits that effectively prevent SCC and greatly reduce the probability of IASCC. The limits imposed by the Primary Water Chemistry Program at BV Unit 1 are consistent with the latest EPRI guidelines as described in Section 4.1.

Since 1990, ultrasonic testing has indicated that SCC has occurred in certain second generation alloy X-750 (Grade 688) support pins in various plants with greater than 55,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation. Prior to replacement, numerous support pins at other plants using alloy X-750 material with the same heat treatment as that of the BV Unit I pins failed during removal or during operation between 110,900 and 149,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation. The alloy X-750 support pins previously in Unit 1 were installed in August, 1983, and had operated at the time of replacement for approximately 155,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

In response to the industry concern for SCC of the alloy X-750 material, FENOC replaced all of the upper internals guide tube support pins at BV Unit 1 (September/October 2007) with Westinghouse-supplied cold worked Type 316 Stainless Steel support pins to mitigate the possibility of continued SCC of these components. Detailed descriptions of the replacement are contained within FENOC Engineering Change Package, ECP 06-0291-001 [20].

4.3 INDUSTRY PROGRAMS 4.3.1 WCAP-14577, Aging Management for Reactor Internals The Westinghouse Owners Group (WOG, now PWROG) topical report WCAP-14577 [8] contains a technical evaluation of aging degradation mechanisms and aging effects for Westinghouse RVI components. The WOG sent the report to the NRC staff to demonstrate that WOG member plant owners that subscribed to the WCAP could adequately manage effects of aging on RVI during the period of extended operation, using approved aging management methodologies of the WCAP to develop plant-specific aging management programs.

The AMR for the BV Unit I internals, documented in [21] was completed in accordance with the requirements of WCAP-14577 [8].

4.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines MRP-227-A, as discussed in Section 2, was developed by a team of industry experts including utility representatives, NSSS vendors, independent consultants, and international committee representatives who reviewed available data and industry experience on materials aging. The objective of the group was to develop a consistent, systematic approach for identifying and prioritizing inspection and evaluation requirements for reactor internals. The following subsections briefly describe the industry process.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-7 4.3.2.1 MRP-227-A, RVI Component Categorizations MRP-227-A used a screening and ranking process to aid in the identification of required inspections for specific RVI components. MRP-227-A credited existing component inspections, when they were deemed adequate, as a result of detailed expert panel assessments conducted in conjunction with the development of the industry document. Through the elements of the process, the reactor internals for all currently licensed and operating PWR designs in the United States were evaluated in the MRP program; and appropriate inspection, evaluation, and implementation requirements for reactor internals were defined.

Based on the completed evaluations, the RVI components are categorized within MRP-227-A as "Primary" components, "Expansion" components, "Existing Programs" components, or "No Additional Measures" components, as described as follows:

• 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 the I&E guidelines. The Primary group also includes components that 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 depends on the findings from 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 and for which generic and plant-specific existing AMP elements are capable of managing those effects, were placed in the Existing Programs group.

No Additional Measures Programs 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 a failure mode, effects, and criticality analysis (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 used in the development of MRP-227-A are not intended to supersede any ASME B&PV Code Section XI [22] requirements. Any components that are classified as core WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-8 support structures, as defined in ASME B&PV Code Section XI IWB-2500, Category B-N-3, have requirements that remain in effect and may only be altered as allowed by 10 CFR 50.55a.

4.3.2.2 NEI 03-08 Guidance within MRP-227-A The industry program requirements of MRP-227-A are classified in accordance with the requirements of the NEI 03-08 protocols. The MRP-227-A guideline includes Mandatory and Needed elements as follows:

0 Mandatory There is one Mandatory element:

1. Each commercial U.S. PWR unit shall develop and document a programfor management of aging of reactorinternals components within thirty-six monthsfollowing issuance of MRP-227-Rev. 0 (that is, no later than December 31, 2011).

BV Unit I Applicability: MRP-227, Revision 0 was officially issued by the industry in December 2008. An AMP must therefore be developed by December 2011. To fulfill this requirement and the license renewal commitments provided in Section 1, FENOC developed NOP-CC-5004, Revision 0, "Pressurized Water Reactor Vessel Internals Program" [1]. This program was effective prior to December 2011 to meet this requirement.

According to the NRC Regulatory Issue Summary (RIS) [31, BV Unit I qualifies as a Category B plant because they have a renewed license with a commitment to submit an AMP/inspection plan based on MRP-227 but that have not yet been required to do so by their commitment. This AMP fulfills the license renewal commitment to submit an implementation schedule for BV Unit 1 in accordance with MRP-227-A [5] to the NRC no later than January 29, 2014.

• Needed There are five Needed elements, with the fifth element being conditional based on examination results:

1. Each commercial US. PWR unit shall implement MRP-227-A, Tables 4-1 through 4-9 and Tables 5-1 through 5-3for the applicable design within twenty-four months following issuance of MRP-227-A.

BV Unit 1 Applicability: MRP-227-A augmented inspections have been appropriately incorporated into this AMP for the license renewal period. The applicable Westinghouse tables contained in MRP-227-A, Table 4-3 (Primary), Table 4-6 (Expansion), Table 4-9 (Existing), and Table 5-3 (Examination Acceptance and Expansion Criteria) and are attached herein as Appendix C, Tables C-1, C-2, C-3, and C-4 respectively.

2. Examinationsspecified in the MRP-227-A guidelines shall be conducted in accordance with Inspection Standard,MRP-228 [10].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-9 BV Unit 1 Applicability: Inspection standards will be in accordance with the requirements of MRP-228 [10]. These inspection standards will be used for augmented inspection at BV Unit 1 as applicable where required by MRP-227-A directives.

3. Examinationresults that do not meet the examination acceptancecriteriadefined in Section 5 of the MRP-22 7-A guidelines shall be recordedand entered in the plant corrective actionprogram and dispositioned.

BV Unit I Applicability: BV Unit I will comply with this requirement.

4. Each commercial U.S. PWR unit shallprovide a summary report of all inspectionsand monitoring, items requiringevaluation, and new repairsto the MRP ProgramManagerwithin 120 days of the completion of an outage duringwhich PWR internals within the scope of MRP-227-A are examined BV Unit 1 Applicability: As discussed in Section 4.3.4, FENOC will participate in future industry efforts and will adhere to industry directives for reporting, response, and follow-up.

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

BV Unit 1 Applicability: BV Unit I will evaluate any examination results that do not meet the examination acceptance criteria in Section 5 of MRP-227-A in accordance with an NRC-approved methodology.

4.3.2.3 GALL AMP Development Guidance It should be noted that Section XI.M16A of NUREG-1 801, Revision 2 [17] includes a description of the attributes that make up an acceptable AMP. These attributes are consistent with the BV Unit 1 Aging Management Review process. Evaluation of the BV Unit 1 RVI AMP against GALL attribute elements is provided in Section 5 of this AMP.

As part of License Renewal, BV Unit 1 agreed to participate in the industry programs applicable to BVPS for investigating and managing aging effects on reactor internals. The industry efforts have defined the required inspections and examination techniques for those components critical to aging management of RVI, The results of the industry recommended inspections, as published in MRP-227-A, serve as the basis for identifying any augmented inspections that are required to complete the BV Unit 1 RVI AMP.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-10 4.3.2.4 MRP-227-A Applicability to BV Unit 1 The applicability of MRP-227-A to BV Unit I requires compliance with the following MRP-227-A assumptions:

30 years of operation with high-leakage core loadingpatterns (freshfuel assemblies loadedin peripherallocations)followed by implementation of a low-leakagefuel management strategyfor the remaining30 years of operation.

BV Unit 1 Applicability: According to the BV RVI Program [1], Unit I had approximately 17 years of operation with fresh fuel assemblies at peripheral locations. The history of the Unit I core designs were reviewed and verified to fall within the assumptions of MRP-227 [1].

According to the LRA [23], BV Unit I currently employs a standard L4P low-leakage core loading pattern. No change to the low leakage core design philosophy is anticipated for the extended plant operating license.

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

BV Unit 1 Applicability: BV Unit 1 operates as a base load unit [1].

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

BV Unit 1 Applicability: MRP-227-A states that the recommendations are applicable to all U.S.

PWR operating plants as of May 2007 for the three designs considered. FENOC has not made any modifications to the Unit 1 internals beyond those identified in general industry guidance or recommended by the original vendor since May 2007. Therefore, there are no differences in component inspection categories [1].

Based on the plant-specific applicability, as stated, the MRP-227-A work is representative for Beaver Valley Unit 1.

4.3.3 WCAP-17451-P, Reactor Internals Guide Tube Wear The PWROG recently funded a program to develop a tool to facilitate prediction of continued operation of reactor upper internals guide tubes from a guide card and lower guide tube continuous guidance wear standpoint, as well as to establish an initial inspection schedule based on the various guide tube designs for the utilities participating in this program. A technical basis document was created for this program, WCAP-17451 -P, Revision 1, "Reactor Internals Guide Tube Wear- Westinghouse Domestic Fleet Operational Projections" [26] which developed a guide plate (card) initial inspection schedule for Westinghouse NSSS designed plants. The intent of this industry guidance is to replace the current guide plate (card) inspection requirements within the next revision to MRP-227 [5].

Beaver Valley Unit 1 is a three loop plant with a 17x 17 standard guide tube design. According to Section 5.4 of the WCAP [26], the generic initial guide card and continuous guidance inspection measurement WCAP-17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-11 EFPY range for this guide tube design is 30 to 34 effective full-power years (EFPY). Beaver Valley Unit 1 was evaluated as a part of this technical basis and therefore, an alternative initial inspection measurement can be performed during an outage within a time range from 30 to 38 EFPY.

4.3.4 Ongoing Industry Programs The U.S. industry, through both the EPRI/MRP and the PWROG, continues to sponsor activities related to RVI aging management, including planned development of a standard NRC submittal template, development of a plant-specific implementation program template for currently licensed U.S. PWR plants, and development of acceptance criteria and inspection disposition processes. FENOC will maintain cognizance of industry activities related to PWR internals inspection and aging management.

FENOC will also address/implement industry guidance, stemming from those activities, as appropriate under NEI 03-08 practices.

4.4

SUMMARY

It should be noted that the FENOC BV Unit 1, the MRP, and the PWROG approaches to aging management are based on the GALL approach to aging management strategies. This approach includes a determination of which reactor internals passive components are most susceptible to the aging mechanisms of concern and then determination of the proper inspection or mitigating program that provides reasonable assurance that the component will continue to perform its intended function through the period of extended operation. The GALL-based approach was used at Beaver Valley for the initial basis of the LRA that resulted in the NRC SER in NUREG-1929 [2].

The approach used to develop the BV Unit I AMP is fully compliant with regulatory directives and approved documents. The additional evaluations and analysis completed by the MRP industry group have provided clarification to the level of inspection quality needed to determine the proper examination method and frequencies. The tables provided in MRP-227-A and included as Appendix C of this AMP provide the level of examination required for each of the components evaluated.

It is the Beaver Valley position that use of the AMR produced by the LRA methodology, combined with any additional augmented inspections required by the MRP-227-A industry tables provided in Appendix C, provides reasonable assurance that the reactor internals passive components will continue to perform their intended functions through the period of extended operation.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5 BEAVER VALLEY REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES The BV Unit 1 RVI AMP is credited for aging management of RVI components for the following eight aging degradation mechanisms and their associated effects:

• Stress corrosion cracking

• Irradiation-assisted stress corrosion cracking

• Wear

• Fatigue

• Thermal aging embrittlement

• Irradiation embrittlement

• Void swelling and irradiation growth

• Thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep The attributes of the BV Unit I RVI AMP and compliance with NUREG-1801 (GALL Report),Section XI.M I6A, "PWR Vessel Internals" [ 17] are described in this section. The GALL identifies 10 attributes for successful component aging management. The framework for assessing the effectiveness of the projected program is established by the use of the 10 elements of the GALL.

FENOC fully utilized the GALL process contained in NUREG- 1801 [17] in performing the aging management review of the reactor internals in the license renewal process. However, FENOC made a commitment (see NUREG-1929 [2]) to incorporate the following: (1) BV Unit 1 will continue to participate in the industry programs applicable to BV Unit I for investigating and managing aging effects on reactor internals, (2) evaluate and implement the results of the industry programs as applicable to the BV Unit 1 reactor internals; and, (3) upon completion of these programs, but not less than 24 months before entering the period of extended operations, submit an inspection plan for the BV Unit 1 reactor internals to the NRC for review and approval.

This AMP is consistent with that process and includes consideration of the augmented inspections identified in MRP-227-A and fully meets the requirements of the commitment and GALL, Revision 2.

Specific details of the BV Unit I reactor internals AMP are summarized in the following subsections.

5.1 GALL REVISION 2 ELEMENT 1: SCOPE OF PROGRAM GALL Report AMP Element Description "The scope of the program includes all RVI components at the Beaver Valley Unit 1 Nuclear Plant,which is built to a Westinghouse NSSS design. The scope of the programapplies the methodology and guidance in the most recently NRC-endorsed version of MRP-22 7, which provides augmented inspection andflaw evaluation methodologyfor assuringthe functional integrity of safety-related internals in commercialoperating U.S. PWR nuclearpower plants designed by B& W, CE, and Westinghouse. The scope of components consideredfor inspection under MRP-227 guidance includes core supportstructures (typically denoted as Examination CategoryB-N-3 by the ASME Code,Section XI), those RVI components that serve an intended license renewal safety function pursuantto criteriain 10 CFR 54.4(a)(1), and other R VI WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-2 components whose failure could prevent satisfactoryaccomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(1), (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 requiredto 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 internalsurface of the reactorvessel because these components are consideredto be ASME Code Class I appurtenancesto the reactor vessel and are adequately managed in accordance with an applicant'sAMP that correspondsto GALL AMP XI.M1, 'ASME Code, Section XA Inservice Inspection, Subsections IWB, IWC, and IWD. '

The scope of the program includes the response bases to applicablelicense renewal applicant action items (LRAAIs) on the MRP-227 methodology, and any additionalprograms, actions, or activitiesthat are discussedin these LRAAI responses and creditedfor aging management of the applicant'sRVI components. The LRAAIs are identified in the staffs safety evaluation on MRP-227 and include applicable action items on meeting those assumptions thatformed the basis of the MRP's augmented inspection andflaw evaluation methodology (as discussed in Section 2.4 of MRP-227), and NSSS vendor-specific orplant-specificLRAAIs as well. The responses to the LRAAIs on MRP-227 are provided in Appendix C of the LRA.

The guidance of MRP-22 7 specifies applicabilitylimitations to base-loadedplants and the fuel loadingmanagement assumptions upon which the functionality analyses were based These limitations and assumptions requirea determinationof applicabilityby the applicantfor each reactorand are covered in Section 2.4 of MRP-22 7" [17].

BV Unit 1 Program Scope The BV Unit 1 RVI consist of three major assemblies: (1) the lower core support structure (also known as the "lower internals"), (2) the upper core support structure (also known as the "upper internals"), and (3) the in-core instrumentation support structure (includes components that are part of the "upper internals" or the "lower internals"). Additional RVI details are provided in Section 3.2.2.2 of the BV Unit I Updated Final Safety Analysis Report (UFSAR).

The BV Unit 1 RVI subcomponents that required aging management review are indicated in the previously submitted Table 2.3.1-2 of the BV Unit I LRA [23]. The information in this table is included as part of the table in Appendix B. The table lists the subcomponents of the RVI that required aging management review along with each subcomponent intended function(s).

The BV Unit 1 Reactor Internals AMR was conducted and documented in LRBV-MAMR-06B [21]. The table summarizing the results of that review was also documented in Table 3.1.2-2 of the BV Unit 1 LRA

[23].This table is included in Appendix B of this AMP. The table identifies the aging effects that require management for the components requiring AMR. A column in the tables lists the program/activity that is credited to address the component and aging effect during the period of extended operation. The NRC has reviewed and approved the aging management strategy presented in the Appendix B tables as documented in the SER on license renewal [2].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-3 The results of the industry research provided by MRP-227-A, summarized in the tables of Appendix C, provide the basis for the required augmented inspections, inspection techniques to permit detection and characterizing of the aging effects (cracks, loss of material, loss of preload, etc.) of interest, prescribed frequency of inspection, and examination acceptance criteria. The information provided in MRP-227-A is rooted in the GALL methodology. The basic assumptions of MRP-227-A, Section 2.4 are met by BV Unit I and are addressed in subsection 4.3.2.4 of this AMP. The Topical Report Conditions and Applicant/Licensee Action Items provided by the NRC in the Safety Evaluation (SE) on MRP-227, Revision 0 [5] are met by Beaver Valley and demonstration of compliance is addressed in Section 6.1 for the Topical Report Conditions and in Section 6.2 for the Applicant/Licensee Action Items. The BV Unit 1 RVI AMP scope is additionally based on previously established and approved GALL Report approaches through application of the MRP-227-A [5] methodologies to determine those components that require aging management.

Conclusion This element complies with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [ 17] and Commitment 18 in the Beaver Valley SER.

5.2 GALL REVISION 2 ELEMENT 2: PREVENTIVE ACTIONS GALL Report AMP Element Description "The guidance in MRP-227 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 crackingor any of its forms

[SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitoredand maintainedin 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"' [17].

BV Unit 1 Preventive Action The BV Unit 1 RVI AMP includes the Primary Water Chemistry Program [18] as an existing program that complies with the requirements of this element. A description and applicability to the BV Unit 1 RVI AMP is provided in the following subsection.

Primary Water Chemistry Program To mitigate aging effects on component surfaces that are exposed to water as process fluid, chemistry programs are used to control water chemistry for impurities (e.g., dissolved oxygen, chloride, fluoride, and sulfate) that accelerate corrosion. This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below the system-specific limits. The BV Unit I PWR Primary Water Chemistry Program [ 18] is based on the current, approved revisions of EPRI PWR Primary Water Chemistry Guidelines.

The limits of known detrimental contaminants imposed by the chemistry monitoring program are consistent with the EPRI PWR Primary Water Chemistry Guidelines [25].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 54 Conclusion This element complies with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [17] and Commitment 18 in the BV Unit 1 SER.

5.3 GALL REVISION 2 ELEMENT 3: PARAMETERS MONITORED OR INSPECTED GALL Report AMP Element Description "The program manages the following age-relateddegradationeffects and mechanisms that are applicablein general to the R VI components at the facility: (a)cracking induced by SCC, PWSCC, IASCC, orfatigue/cyclical loading; (b) loss of materialinduced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiationembrittlement; (d) changes in dimension due to void swelling and irradiationgrowth, distortion, or deflection; and (e)loss ofpreloadcaused by thermal and irradiation-enhancedstress relaxationor creep. For the management of cracking, the program monitors the evidence of surface breaking linear discontinuitiesif a visual inspection technique is used as the non-destructionexamination (NDE) method, orfor relevantflaw presentationsignals ifa volumetric UT method is used as the NDE method. For the management of loss of material,the program monitorsfor gross or abnormal surface conditions that may be indicative of loss of material occurringin the components. For the management of loss ofpreload,the program monitorsfor gross surface conditions that may be indicative of looseningin applicable bolted,fastened, keyed, or pinned connections. The program does not directly monitorfor loss offracture toughness that is induced by thermalaging or neutron irradiationembrittlement, or by void swelling and irradiationgrowth; instead,the impact of loss offracture toughness on component integrity is indirectly managed by using visual or volumetric examination techniques to monitorfor cracking in the components and by applying applicablereducedfracture toughness properties in theflaw evaluations if cracking is detected in the components and is extensive enough to warranta supplementalflaw growth orflaw tolerance evaluation under MRP-227 guidance or ASME Code,Section XI requirements. The program uses physical measurements to monitorfor any dimensionalchanges due to void swelling, irradiationgrowth, distortion, or deflection.

Specifically, the program implements the parametersmonitored/inspectedcriteriafor Westinghouse designed Primary Components in Table 4-3 ofMRP-227. Additionally, the program implements the parametersmonitored/inspectedcriteriafor Westinghouse designed Expansion Components in Table 4-6 of MRP-227. The parametersmonitored/inspectedfor Existing Program Components follow the basesfor referenced Existingprograms,such as the requirementsfor ASME Code Class RVI components in ASME Code,Section XI, Table IWB-2500-1, ExaminationCategoriesB-N-3, as implemented through the applicant'sASME Code,Section XI program, or the recommended programfor inspecting Westinghouse-designedflux thimble tubes in GALL AMP XIM3 7, "Flux Thimble Tube Inspection." No inspections, except for those specified in ASME Code, Section X7, are requiredfor components that are identified as requiring "No AdditionalMeasure, " in accordancewith the analyses reportedin MRP-227"

[17].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-5 BV Unit 1 Parameters Monitored or Inspected The BV Unit 1 AMP monitors, inspects, and/or tests for the effects of the eight aging degradation mechanisms on the intended function of the BV Unit 1 PWR internals components through inspection and condition monitoring activities in accordance with the augmented requirements defined under industry directives as contained in MRP-227-A and ASME Section XI [22].

This AMP implements the requirements for the Primary Component inspections from Table 4-3 of MRP-227-A (included in Appendix C of this AMP as Table C-1), the Expansion Component inspections from Table 4-6 of MRP-227-A (included in Appendix C of this AMP as Table C-2), and the Existing Component inspections from Table 4-9 of MRP-227-A (included in Appendix C of this AMP as Table C-3). These tables contain requirements to monitor and inspect the RVI through the period of extended operation to address the effects of the eight aging degradation mechanisms. It is noted in Appendix C, Table C-1 that the PWROG has recently developed initial examination period requirements for guide plate (card) wear for Westinghouse NSSS designed plants [26] that replace the current requirements in MRP-227-A [5].

For license renewal, the ASME Section XI Program [4] consists of periodic volumetric, surface, and/or visual examination of components for assessment, signs of degradation, and corrective actions. The requirements of MRP-227-A only augment and do not replace or modify the requirements of ASME Section XI. This program is consistent with the corresponding program described in the GALL Report

[17].

Appendices B and C of this AMP provide a detailed listing of the components and subcomponents and the parameters monitored, inspected, and/or tested.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-I1801,Section XI.M16A [17] and Commitment 18 in the BV Unit I SER.

5.4 GALL REVISION 2 ELEMENT 4: DETECTION OF AGING EFFECTS GALL Report AMP Element Description "The detection of agingeffects is covered in two places. (a) the guidance in Section 4 of MRP-227 provides an introductorydiscussion andjustification of the examination methods selectedfor detecting the aging effects of interest; and (b) standardsfor examination methods, procedures, andpersonnel 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 measurementsfor detecting changes in dimension, and various visual (VT-3, VT-i, andEVT-1) examinationsfor detecting effects rangingfrom general conditions to detection and sizing of surface-breakingdiscontinuities. Surface examinations may also be used as an alternative to visual examinationsfor detection and sizing of surface-breaking discontinuities.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-6 Cracking caused by SCC, IASCC, andfatigue is monitored/inspectedby either VT-1 or EVT-1 examination (for internalsother than bolting) or by volumetric UT examination (bolting). The VT-3 visual methods may be appliedfor the detection of cracking only when theflaw tolerance of the component or affected assembly, as evaluatedfor reducedfracture toughness properties,is known and has been shown to be tolerantof easily detected largeflaws, even under reduced fracture toughness conditions. In addition, VT-3 examinationsare used to monitor/inspectfor loss of materialinduced by wear andfor general agingconditions, such as gross distortion caused by void swelling and irradiationgrowth or by gross effects of loss ofpreloadcaused by thermal and irradiation-enhancedstress relaxationand creep.

In addition, the program adopts the recommended guidance in MRP-227for defining the Expansion criteriathat needed to be applied to inspections of PrimaryComponents and Existing Requirement Components andfor expanding the examinations to include additionalExpansion Components. As a result, inspectionsperformedon the RVI components areperformed consistent with the inspectionfrequency and sampling basesfor PrimaryComponents, Existing Requirement Components, and Expansion Components in MRP-227, which have been demonstratedto be in conformance with the inspection criteria,sampling basis criteria,and sample Expansion criteriain Section A. 1.2.3.4 ofNRC Branch PositionRLSB-1.

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

The program is supplemented by the following plant-specific Primary Component and Expansion Component inspectionsfor the program (as applicable): for BV Unit 1, no additionalPrimaryor Expansion components are relevant to the scope of aging managementfor the RVI.

In addition, in some cases (as defined in MRP-227), physical measurements are used as supplemental techniques to managefor the gross effects of wear, loss ofpreloaddue to stress relaxation, orfor changes in dimension due to void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program include thatfor the hold down spring. The hold down spring at BV Unit 1 isfabricatedfrom Type 304 SS that requires inspection by physical measurement" [17].

BV Unit 1 Detection of Aging Effects Detection of indications that are required by the ASME Section XI ISI Program [4] is well established and field-proven through the application of the Section XI ISI Program. Those augmented inspections that are taken from the MRP-227-A recommendations will be applied through use of the MRP-228 inspection standard. This AMP implements the augmented inspection requirements of Table 4-3, Table 4-6, and Table 4-9 from MRP-227-A for the Primary, Expansion, and Existing Components, respectively.

These are included in Appendix C of this AMP for reference. These tables include the inspection frequency and sampling basis. For the Expansion Components of MRP-227-A, this AMP implements the expansion requirements of Table 5-3 of MRP-227-A (included in Appendix C of this AMP as Table C-4).

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-7 Inspection can be used to detect physical effects of degradation including cracking, fracture, wear, and distortion. The choice of an inspection technique depends on the nature and extent of the expected damage. The recommendations supporting aging management for the reactor internals, as contained in this report, are built around three basic inspection techniques: (1) visual, (2) ultrasonic, and (3) physical measurement The three different visual techniques include VT-3, VT-i, and EVT-1. The assumptions and process used to select the appropriate inspection technique are described in the following subsections.

Inspection standards developed by the industry for the application of these techniques for augmented reactor internals inspections are documented in MRP-228 [10].

VT-i Visual Examinations The acceptance criteria for visual examinations conducted under categories B-N-2 (welded core support structures and interior attachments to reactor vessels) and B-N-3 (removable core support structures) are defined in IWB-3520 [22]. VT-I visual examination is intended to identify crack-like surface flaws.

Unacceptable conditions for a VT-I examination are:

Crack-like surface flaws on the welds joining the attachment to the vessel wall that exceed the allowable linear flaw standards of IWB-3510 [22]

Structural degradation of attachment welds such that the original cross-sectional area is reduced by more than 10 percent These requirements are defined to ensure the integrity of attachment welds on the ferritic pressure vessel.

Although the IWB-3520 criteria do not directly apply to austenitic stainless steel internals, the clear intent is to ensure that the structure will meet minimum flaw tolerance fracture requirements. In the MRP-227-A recommendations, VT-I examinations have been identified for components requiring close visual examinations with some estimate of the scale of deformation or wear. Note that in MRP-227-A, VT-I has only been selected to detect distortion as evidenced by small gaps between the upper-to-lower mating surfaces of CE-welded core shrouds assembled in two vertical sections. Therefore, no additional VT-I inspections over and above those required by ASME Section XI ISI have been specified.

EVT-I Enhanced Visual Examination for the Detection of Surface Breaking Flaws In the augmented inspections detailed in the MRP-227-A for reactor internals, the EVT- 1 enhanced visual examination has been identified for inspection of components where surface-breaking flaws are a potential concern. Any visual inspection for cracking requires a reasonable expectation that the flaw length and crack mouth opening displacement meet the resolution requirements of the observation technique. The EVT-I specification augments the VT-I requirements to provide more rigorous inspection standards for stress corrosion cracking and has been demonstrated for similar inspections in boiling water reactor (BWR) internals. Enhanced visual examination (i.e., EVT-1) is also conducted in accordance with the requirements described for visual examination (i.e., VT-i) with additional requirements (such as camera scanning speed). Any recommendation for EVT-1 inspection will require additional analysis to establish flaw-tolerance criteria, which must take into account potential embrittlement due to thermal aging or neutron irradiation. The industry, through the PWROG, has developed an approach for acceptance criteria methodologies to support plant-specific augmented examinations. This work is summarized in WCAP-17096-NP, "Reactor Internals Acceptance Criteria WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-8 Methodology and Data Requirements" [11 ]. The acceptance criteria developed using these methodologies may be created on either a generic or plant-specific basis because both loads and component dimensions may vary from plant-to-plant within a typical PWR design.

VT-3 Examination for General Condition Monitoring In the augmented inspections detailed in the MRP-227-A for reactor internals, the VT-3 visual examination has been identified for inspection of components where general condition monitoring is required. The VT-3 examination is intended to identify individual components with significant levels of existing degradation. As the VT-3 examination is not intended to detect the early stages of component cracking or other incipient degradation effects, it should not be used when failure of an individual component could threaten either plant safety or operational stability. The VT-3 examination may be appropriate for inspecting highly redundant components (such as baffle-edge bolts), where a single failure does not compromise the function or integrity of the critical assembly.

The acceptance criteria for visual examinations conducted under categories B-N-2 (welded core support structures and interior attachments to reactor vessels) and B-N-3 (removable core support structures) are defined in IWB-3520. These criteria are designed to provide general guidelines. The unacceptable conditions for a VT-3 examination are:

• Structural distortion or displacement of parts to the extent that component function may be impaired;

• Loose, missing, cracked, or fractured parts, bolting, or fasteners;

• Foreign materials or accumulation of corrosion products that could interfere with control rod motion or could result in blockage of coolant flow through fuel;

• Corrosion or erosion that reduces the nominal section thickness by more than 5 percent;

• Wear of mating surfaces that may lead to loss of function;

• Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5 percent.

The VT-3 examination is intended for use in situations where the degradation is readily observable. It is meant to provide an indication of condition, and quantitative acceptance criteria are not generally required. In any particular recommendation for VT-3 visual examination, it should be possible to identify the specific conditions of concern. For instance, the unacceptable conditions for wear indicate wear that might lead to loss of function. Guidelines for wear in a critical-alignment component may be very different from the guidelines for wear in a large structural component.

Surface Examination In order to further characterize discontinuities on the surface of components, surface examination can supplement either visual (VT-3) or (VT-1/EVT-1) examinations specified in these guidelines. This WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-9 supplemental examination may thus be used to reject or accept relevant indications. A surface examination is an examination that indicates the presence of surface discontinuities, and the ASMIE B&PV Code [22] lists magnetic particle, liquid penetrant, eddy current, and ultrasonic examination methods as surface examination alternatives. Here, only the electromagnetic testing (ET), also called eddy current surface examination method, is covered.

When selected for use as a supplemental examination to examinations performed in these guidelines, an ET examination is conducted in accordance with the requirements of the inspection standard [10].

ET examination is widely used for heat exchanger tubing inspections. Eddy currents are induced in the inspected object by electromagnetic coils, with disruptions in the eddy current flow caused by surface or near-surface anomalies detected by suitable instrumentation. Industry experience with ET examination is relatively robust, especially in the aerospace and petroleum refinery industries. The experience base for PWR nuclear systems is moderately robust, in particular for examination of steam generator, flux thimble, and heat exchanger tubing.

Ultrasonic Testing Volumetric examinations in the form of ultrasonic testing (UT) techniques can be used to identify and determine the length and depth of a crack in a component. Although access to the surface of the component is required to apply the ultrasonic signals, the flaw may exist in the bulk of the material. In this proposed strategy, UT inspections have been recommended exclusively for detection of flaws in bolts. For the bolt inspections, any bolt with a detected flaw should be assumed to have failed. The size of the flaw in the bolt is not critical because crack growth rates are generally high, and it is assumed that the observed flaw will result in failure prior to the next inspection opportunity. It has generally been observed through examination performance demonstrations that UT can reliably (90 percent or greater reliability) detect flaws that reduce the cross-sectional area of a bolt by 35 percent.

Failure of a single bolt does not compromise the function of the entire assembly. Bolting systems in the reactor internals are highly redundant. For any system of bolts, it is possible to demonstrate multiple acceptable bolting patterns. The evaluation program must demonstrate that the remaining bolts meet the requirements for an acceptable bolting pattern for continued operation. The evaluation procedures must also demonstrate that the pattern of remaining bolts contains sufficient margin such that continuation of the bolt failure rate will not result in failure of the system to meet the requirements for an acceptable bolting pattern before the next inspection.

Establishment of the acceptable bolting pattern for any system of bolts requires analysis to demonstrate that the system will maintain reliability and integrity in continuing to perform the intended function of the component. This analysis is highly plant-specific. Therefore, any recommendation for UT inspection of bolts assumes that the plant owner will work with the designer to establish acceptable bolting patterns prior to the inspection to support continued operation.

Physical Measurement Examination Continued functionality can be confirmed by physical measurements to evaluate the impact caused by various degradation mechanisms such as wear or loss of functionality as a result of loss of preload or WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-10 material deformation. For BV Unit 1, direct physical measurements are required only for the hold down spring.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [17] and Commitment 18 in the BV Unit 1 SER.

5.5 GALL REVISION 2 ELEMENT 5: MONITORING AND TRENDING GALL Report AMP Element Description "The methods for monitoring,recording,evaluating,and trending the data that resultfrom the program's inspections are given in Section 6 of MRP-227 and its subsections. The evaluation methods include recommendationsforflaw depth sizing andfor crackgrowth determinationsas well for performing applicable limit load, linear elastic and elastic-plasticfracture analyses of relevantflaw indications. The examinations and re-examinationsrequiredby the MRP-227 guidance, together with the requirements specified in MRP-228for inspection methodologies, inspectionprocedures,and inspectionpersonnel,provide timely detection, reporting, and corrective actions with respect to the effects of the age-relateddegradationmechanisms within the scope of the program. The extent of the examinations, beginning with the sample of susceptible PWR internalscomponent locations identifiedas Primary Component locations, with the potentialfor inclusion of Expansion Component locations if the effects are greater than anticipated,plus the continuationof the Existing Programsactivities,such as the ASME Code, Section X7, Examination Category B-N-3 examinationsfor core supportstructures,provides a high degree of confidence in the total programs"[17].

BV Unit 1 Monitoring and Trending Operating experience with PWR reactor internals has been generally proactive. Flux thimble wear and control rod guide tube split pin cracking issues were identified by the industry and continue to be actively managed. The extremely low frequency of failure in reactor internals makes monitoring and trending based on OE somewhat impractical. The majority of the materials aging degradation models used to develop the MRP-227-A guidelines are based on test data from reactor internals components removed from service. The data are used to identify trends in materials degradation and forecast potential component degradation. The industry continues to share both material test data and OE through the auspices of the MRP and PWROG. FENOC has in the past and will continue to maintain cognizance of industry activities and shared information related to PWR internals inspection and aging management.

Inspections credited in Appendix B are based on utilizing the BV Unit 1 10-year ISI program and the augmented inspections derived from MRP-227-A and repeated here in Appendix C. The MiRP-227-A inspections only augment and do not replace the existing ASME Section X1 ISI requirements. These inspections, where practical, are scheduled to be conducted in conjunction with typical 10-year ISI examinations.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-11 Appendix C, Tables C-1, C-2, and C-3 identify the augmented Primary and Expansion inspection and monitoring recommendations, and the Existing programs credited for inspection and aging management.

As discussed in MRP-227-A, inspection of the "Primary" components provides reasonable assurance for demonstrating component current capacity to perform the intended functions. It is noted in Appendix C, Table C-i that the PWROG has recently developed initial examination period requirements for guide plate (card) wear for Westinghouse NSSS designed plants [26] that replace the current requirements in MIRP-227-A [5]. Table C-4 in Appendix C identifies the MRP-227-A expansion criteria from the Primary components. If these expansion criteria are met for a component, the associated Expansion component is to be inspected to manage the aging degradation.

Reporting requirements are included as part of the MRP-227-A guidelines. Consistent reporting of inspection results across all PWR designs will enable the industry to monitor reactor internals degradation on an ongoing industry basis as the period of extended operation moves forward. Reporting of examination results will allow the industry to monitor and trend results and take appropriate preemptive action through update of the MRP guidelines.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.M16A [17] and Commitment 18 in the BV Unit I SER.

5.6 GALL REVISION 2 ELEMENT 6: ACCEPTANCE CRITERIA GALL Report AMP Element Description "Section 5 ofMRP-227 provides specific examination acceptance criteriafor the Primary and Expansion Component examinations. For components addressedby examinations referenced to ASME Code,Section XI, the IWB-3500 acceptance criteriaapply. For other components covered by Existing Programs,the examination acceptancecriteriaare described within the Existing Programreference document.

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

• For visual examination (andsurface examination as an alternative to visual examination),

the examination acceptancecriterion 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 sizedfor length by VT-i/EVT-i examinations;

• For volumetric examination, the examination acceptancecriterion is the capabilityfor reliabledetection of indicationsin bolting, as demonstrated in the examination Technical Justification,;in addition,there are requirementsfor system-level assessment of bolted or pinned assemblies with unacceptablevolumetric (UT) examination indicationsthat exceed specified limits; and

" Forphysical measurements, the examination acceptance criterionfor the acceptable tolerance in the measureddifferential heightfrom the top of the plenum rib pads to the vessel WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-12 seatingsurface in B& Wplants are given in Table 5-1 of MRP-22 7. The acceptancecriterion for physical measurements performedon the height limits of the Westinghouse-designedhold-down springs are requiredfor 304 SS hold down springs. BV Unit 1 has a 304 SS hol down spring, therefore, BV Unit I is requiredto produce acceptance criteriafor the physical measurements on the hold down spring" [17].

BV Unit 1 Acceptance Criteria Those recordable indications that are the result of inspections required by the existing BV Unit 1 ISI program scope are evaluated in accordance with the applicable requirements of the ASME Code through the existing Corrective Action Program [27].

Inspection acceptance and expansion criteria are provided in Appendix C, Table C-4. These criteria will be reviewed periodically as the industry continues to develop and refine the information and will be included in updates to BV Unit 1 procedures to enable the examiner to identify examination acceptance criteria considering state-of-the-art information and techniques. FENOC has a commitment to develop acceptance criteria for the hold down spring physical measurements that will be consistent with the licensing basis for BV Unit 1 [5].

Augmented inspections, as defined by the MRP-227-A requirements included in this AMP as Appendix C, Table C-I, Table C-2, and Table C-3, that result in recordable relevant conditions will be entered into the plant Corrective Action Program and addressed by appropriate actions that may include enhanced inspection, repair, replacement, mitigation actions, or analytical evaluations. An example of an analytical evaluation is using an acceptable bolting WCAP approach such as those commonly used to support continued component or assembly functionality. Additional analysis to establish acceptable bolting pattern evaluation criteria for the baffle-former bolt assembly is also considered in determining the acceptance of inspection results to support continued component or assembly functionality.

The industry, through various cooperative efforts, is working to construct a consensus set of tools in line with accepted and proven methodologies to support this element. One of these tools is the PWROG document WCAP- 17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," [11], which details acceptance criteria methodology for the MRP-227 Primary and Expansion components. Status is monitored through direct FENOC cognizance of industry (including PWROG) activities related to PWR internals inspection and aging management.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.M1 6A [17] and Commitment 18 in the BV Unit 1 SER.

5.7 GALL REVISION 2 ELEMENT 7: CORRECTIVE ACTIONS GALL Report AMP Element Description "Corrective actionsfollowing the detection of unacceptable conditions arefundamentally providedfor in each plant's corrective actionprogram. Any detected conditions that do not WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-13 satisfy the examination acceptance criteriaare requiredto be dispositionedthrough the plant corrective actionprogram, which may require repair,replacement, or analyticalevaluationfor continuedservice until the next inspection. The disposition will ensure that design basisfunctions of the reactor internalscomponents will continue to be fulfilledfor all licensing basis loads and events. Examples of methodologies that can be used to analyticallydisposition unacceptable conditions arefound in the ASME Code,Section XI or in Section 6 ofMRP-227. Section 6 of MRP-227 describes the options that are availablefor disposition of detected conditions that exceed the examination acceptance criteriaof Section 5 of the report.These include engineering evaluation methods, as well as supplementary examinationsto further characterizethe detected condition, or the alternative of component repairand replacementprocedures. The latter are subject to the requirements of the ASME Code, Section X7. The implementation of the guidance in MRP-227, plus the implementation of any ASME Code requirements,provides an acceptable level of aging management of safety-relatedcomponents addressedin accordance with the corrective actions of 10 CFR Part50, Appendix B or its equivalent, as applicable.

Other alternativecorrective action bases may be used to disposition relevant conditions if they have been previously approvedor endorsed by the NRC. Examples ofpreviously NRC-endorsed alternative corrective actions bases include those corrective actions basesfor Westinghouse-design RVI 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-145 77-Rev. ]-A, or for B& W-designed RVI components in B& W Report No. BA W-2248. Westinghouse Report No. WCAP-14577-Rev. 1-A was endorsedfor use in an NRC SE to the Westinghouse Owners Group, dated February10, 2001. B& W Report No.

BA W-2248 was endorsedfor 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 submittedfor NRC approvalprior to their implementation" [17].

BV Unit 1 Corrective Action The existing Beaver Valley procedure for corrective actions, the "Corrective Action Program" [27] and the ASME Section XI ISI program [4], will be credited for this element. These procedures establish the BV Unit 1 repair and replacement requirements of ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" [22]. These requirements include the identification of a repair cycle, repair plan, and verification of acceptability for replacements. BV Unit 1 is committed to performing corrective actions for augmented inspections using repair and replacement procedures equivalent to those requirements in ASME B&PV Code,Section XI and MRP-227-A, Section 6 [5].

Conclusion This element complies with the corresponding aging management attribute in NUREG-I 801,Section XI.M16A [17] and Commitment 18 in the BV Unit I SER.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-14 5.8 GALL REVISION 2 ELEMENT 8: CONFIRMATION PROCESS GALL Report AMP Element Description "Site quality assuranceprocedures,review and approvalprocesses, and administrativecontrols are implemented in accordancewith 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 qualityfor inspection,flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with the 10 CFR Part50, Appendix B, or their equivalent (as applicable),confirmationprocess, and administrative controls" [171.

BV Unit 1 Confirmation Process BV Unit 1 has an established 10 CFR Part 50, Appendix B, Program [28] that addresses the elements of corrective actions, confirmation process, and administrative controls. The BV Unit I Program includes non-safety-related structures, systems, and components. Quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.MI6A [17] and Commitment 18 in the BV Unit 1 SER.

5.9 GALL REVISION 2 ELEMENT 9: ADMINISTRATIVE CONTROLS GALL Report AMP Element Description "The administrativecontrolsfor such programs,including their implementingprocedures and review and approvalprocesses, are under existing site 10 CFR 50 Appendix B Quality Assurance Programs,or their equivalent, as applicable.Such a program is thus expected to be established with a sufficient level of documentation and administrative controls to ensure effective long-term implementation" [ 17].

BV Unit 1 Administrative Controls BV Unit I has an established 10 CFR Part 50, Appendix B Program [28] that addresses the elements of corrective actions, confirmation process, and administrative controls. The BV Unit 1 program includes non-safety-related structures, systems, and components. Quality assurance (QA) procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-15 Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-I 801,Section XI.M16A [17] and Commitment 18 in the BV Unit I SER.

5.10 GALL REVISION 2 ELEMENT 10: OPERATING EXPERIENCE GALL Report AMP Element Description "Relativelyfew incidents ofPWR internalsaging degradationhave been reportedin operating US. commercialPWR plants. A summary of observationsto date is provided in Appendix A of MRP-22 7-A. The applicant is expected to review subsequent operatingexperiencefor impact on its program or to participatein industry initiatives thatperform this function.

The application of the MRP-227 guidance will establish a considerableamount of operating experience over the next few years. Section 7 ofMRP-227 describes the reportingrequirements for these applications,and the planfor evaluatingthe accumulatedadditionaloperating experience" [17].

BV Unit 1 Operating Experience Extensive industry and BV Unit 1 OE has been reviewed during the development of the RVI AMP. The experience reviewed includes NRC Information Notices 84-18, "Stress Corrosion Cracking in PWR Systems" [29] and 98-11, "Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants"

[30]. Most of the industry OE reviewed has involved cracking of austenitic stainless steel baffle-former bolts or SCC of high-strength internals bolting. SCC of control rod guide tube support pins has also been reported.

Early plant OE related to hot functional testing and reactor internals is documented in plant historical records. Inspections performed as part of the 10-year ISI program have been conducted as designated by existing commitments and would be expected to discover overall general internals structure degradation.

To date, very little degradation has been observed industry-wide.

Industry OE is routinely reviewed by FENOC engineers using Institute of Nuclear Power Operations (INPO) OE, the Nuclear Network, and other information sources as directed under the applicable procedure [31 ], for the determination of additional actions and lessons learned. These insights, as applicable, can be incorporated into the plant systems quarterly health reports and further evaluated for incorporation into plant programs.

A review of industry and plant-specific experience with RVI reveals that the U.S. industry, including FENOC and BV Unit 1, has responded proactively to industry issues relative to reactor internals degradation. Two examples that demonstrate this proactive response is the replacement of the Unit 1 control rod guide tube split pins in 2007 and the replacement of flux thimble tubes in 2000, which are briefly described in the following paragraphs.

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• BV Unit I Control Rod Guide Tubes Support Pins In response to the industry concern for SCC of the alloy X-750 material, FENOC replaced all of the upper internals guide tube support pins at BV Unit 1 (September - October 2007) with Westinghouse supplied cold worked Type 316 SS support pins to mitigate the possibility of continued SCC of these components.

Detailed descriptions of the replacement are contained within FENOC Engineering Change Package, ECP 06-0291-001 [20].

• BV Unit 1 Flux Thimble Tubes During the Unit I Cycle 13 Refueling Outage (February - April 2000), a proactive decision was made to replace 18 flux thimble tubes at Unit 1 which were either inoperable or showed the greatest amount of tube wall thinning. This action was taken to ensure that the Technical Specification minimum number of operable flux thimble tubes would be satisfied [23].

A key element of the MRP-227-A guideline is the reporting of age-related degradation of RVI components. FENOC, through its participation in PWROG and EPRI-MRP activities, will continue to benefit from the reporting of inspection information and will share its own OE with the industry through the reporting requirements of Section 7 of MRP-227-A. The collected information from MRP-227-A augmented inspections will benefit the industry in its continued response to RVI aging degradation.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.M16A [17] and Commitment 18 in the BV Unit I SER.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-1 6 DEMONSTRATION Beaver Valley Unit 1 has demonstrated a long-term commitment to aging management of reactor internals. This AMP is based on an established history of programs to identify and monitor potential aging degradation in the reactor internals. Programs and activities undertaken in the course of fulfilling that commitment include:

The examinations required by ASME Section XI for the BV Unit 1 reactor vessel internals have been performed during each 10-year interval since plant operations commenced.

• As documented in Beaver Valley operational procedures, reports are continuously reviewed by Beaver Valley personnel for applicable issues that indicate operating procedures or programs require updates based on new OE.

• Review of Nuclear Oversight Section (NOS) audit reports, NRC inspection reports, and INPO evaluations indicate no unacceptable issues related to RVI inspections.

• The Primary Water Chemistry Program at Beaver Valley has been effective in maintaining oxygen, halogens, and sulfate at levels sufficiently low to prevent SCC of the reactor vessel internals.

• Replacement control rod guide tube support pins for BV Unit 1 in 2007 were fabricated from cold worked Type 316 SS materials to increase resistance to SCC (versus original pins) [20].

• Flux thimble tubes were proactively replaced to ensure that the Technical Specification minimum number would be satisfied [23].

a Beaver Valley has participated in the PWROG program to develop initial examination period requirements for guide plate (card) wear for Westinghouse NSSS designed plants [26].

FENOC has actively participated in past and ongoing EPRI and PWROG RVI activities. FENOC will continue to maintain cognizance of industry activities related to PWR internals inspection and aging management; and will address/implement industry guidance, stemming from those activities, as appropriate under NEI 03-08 practices.

This AMP fulfills the approved license renewal methodology requirement to identify the most susceptible components and to inspect those components with an indication detection level commensurate with the expected degradation mechanism indication. Augmented inspections, derived from the information contained in MRP-227-A, the industry I&E Guidelines, have been utilized in this AMP to build on existing plant programs. This approach is expected to encourage detection of a degradation mechanism at its first appearance consistent with the ASME approach to inspections. This approach provides reasonable assurance that the internals components will continue to perform their intended function through the period of extended operation.

Typical ASME Section XI examinations identified in the AMP for the period of extended operation are to be performed at BV Unit I in Fall 2016, refueling outage 25 (RO-25). The augmented inspections WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-2 discussed in compliance with MRP-227-A requirements have been integrated in the implementation schedule, which is shown in Section 7. Integration of the required inspections will be tracked to completion. As discussed, the industry MRP-227-A guidelines also provide for updates as experience is gained through inspection results. This feedback loop will enable updates based on actual inspection experience.

The augmented inspections described in this document, as summarized in Appendix C, combined with the ASME Section XI ISI program inspections, existing Beaver Valley programs, and use of Operating Experience Reports (OERs), provide reasonable assurance that the reactor internals will continue to perform their intended functions through the period of extended operation.

Table 6-1 lists the seven topical report conditions and Section 6.2 lists the eight applicant action items that came out of the NRC review of MRP-227, as listed in [5], as well as their compliance within this AMP.

6.1 DEMONSTRATION OF TOPICAL REPORT CONDITIONS COMPLIANCE TO SE ON MRP-227, REVISION 0 Table 6-1 Topical Report Condition Compliance to SE on MRP-227 Topical Condition Applicable/Not Applicable Compliance in AMP

1. High consequence components in Applicable The upper core plate and the lower support the "No Additional Measures" forging or casting components are added to Inspection Category Table C-2 as "Expansion Components" linked to the "Primary Component," the CRGT lower flange weld.

2. Inspection of components subject to Applicable The upper and lower core barrel cylinder irradiation-assisted stress corrosion girth welds and the lower core barrel flange cracking weld are moved from Table C-2 "Expansion Components" to Table C-I "Primary Components."

3. Inspection of high consequence Not Not applicable for BV Unit 1 components subject to multiple Applicable degradation mechanisms

4. Imposition of minimum Applicable Notes 2 through 4 were added to Table C-1, examination coverage criteria for as well as Note 2 to Table C-2 to reflect this "Expansion" inspection category condition.

components

5. Examination frequencies for baffle- Applicable In Table C-I for the baffle-former bolts, the former bolts and core shroud bolts inspection frequency was changed from 10 to 15 additional effective full-power years (EFPY) to subsequent examination on a ten-year interval.

6. Periodicity of the re-examination of Applicable "Re-inspection every 10 years following "Expansion" inspection category initial inspection" was added to every components component under the Examination Method/Frequency column in Table C-2.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3 Table 6-1 Topical Report Condition Compliance to SE on MRP-227 (cont.)

Topical Condition Applicable/Not Applicable Compliance in AMP

7. Updating of MRP-227, Revision 0, Applicable Section 5 is updated to reflect XI.MI6A Appendix A from GALL Revision 2 [17].

6.2 DEMONSTRATION OF APPLICANT/LICENSEE ACTION ITEM COMPLIANCE TO SE ON MRP-227, REVISION 0 6.2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions "As addressedin Section 3.2.5.1 of this SE, each applicant/licenseeis responsiblefor assessing its plant's design and operatinghistory and demonstratingthat the approved version of MRP-22 7 is applicable to the facility. Each applicant/licenseeshall refer, in particular,to the assumptions regardingplant design and operatinghistory made in the FMECA andfunctionality analysesfor reactors of their design (i.e., Westinghouse, CE, or B& W) which support MRP-22 7 anddescribe the process usedfor determiningplant-specific differences in the design of their RVI components or plant operatingconditions, which result in different component inspection categories. The applicant/licenseeshall submit this evaluationfor NRC review and approval as part of its applicationto implement the approved version ofMRP-227. This is Applicant/Licensee Action Item 1" [5].

BV Unit 1 Compliance The process used to verify that BV Unit I is reasonably represented by the generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MRP-227-A [5] is as follows:

1. Identification of typical Westinghouse PWR internal components (MRP-191, Table 4-4

[9]).

2. Identification of BV Unit 1 PWR internals components.

3. Comparison of the typical Westinghouse PWR internals components to the BV Unit I PWR internals components.

a. Confirmation that no additional items were identified by this comparison (primarily supports Applicant/Licensee Action Item 2).

b. Confirmation that the materials identified for BV Unit I are consistent with those materials identified in MRP- 191, Table 4-4.

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c. Confirmation that the BV Unit 1 internals are the same as, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

4. Confirmation that the BV Unit 1 operating history is consistent with the assumptions in MRP-227-A regarding core loading patterns.

5. Confirmation that the BV Unit 1 RVI materials operated at temperatures within the original design basis parameters.

6. Determination of stress values based on design basis documents.

7. Confirmation that any changes to the BV Unit 1 RVI components do not impact the application of the MRP-227-A generic aging management strategy.

BV Unit 1 reactor internals components are reasonably represented by the design and operating history assumptions regarding neutron fluence, temperature, materials, and stress values in the MRP-191 generic FMECA and the MRP-232 functionality analyses based on the following:

I. BV Unit 1 operating history is consistent with the assumptions in MRP-227-A with regard to neutron fluence.

a. The FMECA and functionality analyses for MRP-227-A were based on the assumption of 30 years of operation with high-leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy. BV Unit 1 had approximately 17 years of operation with fresh fuel assemblies at peripheral locations (high-leakage core loading pattern). The low-leakage loading pattern has been applied to all subsequent core designs through current operation. No change to the low-leakage core design philosophy is anticipated for the extended plant operating license [1,23]. By operating with a high-leakage core design for less than 30 years, FENOC has taken a conservative approach. Therefore, BV Unit 1 meets the fluence and fuel management assumptions in MRP-191 and requirements for MRP-227-A application.

b. BV Unit I has operated under base load conditions for the majority of the life of the plant [1]. Therefore, BV Unit I satisfies the assumptions in MRP documents regarding operational parameters affecting fluence.

2. The BV Unit I reactor coolant system operates between Thot and Tcold [1,32], which are not less than approximately 547°F for T,0od and not higher than 606°F for Thot. The design temperature for the reactor vessel is 650'F. BV Unit 1 operating history is within original design basis parameters and therefore consistent with the assumptions used to develop the MRP-227-A aging management strategy with regard to temperature operational parameters.

3. BV Unit I internals components and materials are comparable to the typical Westinghouse PWR internals components (MRP-191, Table 4-4).

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a. No additional components were identified for BV Unit I by this comparison [23].

b. Materials identified for BV Unit 1 are consistent or nearly equivalent with those materials identified in MRP-191, Table 4-4 for Westinghouse-designed plants.

Where differences exist, there is no impact on the BV Unit I RVI program or the component is already credited as being managed under an alternate BV Unit 1 aging management program.

c. BV Unit I internals are the same as, or equivalent to the typical Westinghouse PWR internals regarding design and fabrication.

4. Modifications to the BV Unit 1 reactor internals made over the lifetime of the plant are those specifically directed by Westinghouse, the Original Equipment Manufacturer (OEM) [1]. The design has been maintained over the lifetime of the plant as specified by the OEM, operational parameters are compliant with MRP-227-A requirements with regard to fluence and temperature, and the components and materials are the same as those considered in MRP-191. Therefore, the BV Unit 1 stress values are represented by the assumptions in MRP-191, MRP-232, and MRP-227-A, confirming the applicability of the generic FMECA.

Conclusion BV Unit I complies with Applicant/Licensee Action Item I of the NRC SE on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal "As discussed in Section 3.2.5.2 of this SE, consistent with the requirementsaddressedin 10 CFR 54.4, each applicant/licenseeis responsiblefor identifying which RVI components are within the scope of LR for its facility. Applicants/licenseesshall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the RVI components that are within the scope of LR for theirfacilities in accordancewith 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for its facility, the applicantor licensee shall identify the missing component(s) andpropose any necessary modifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specificAMP. The AMP shallprovide assurancethat the effects of aging on the missing component(s) will be managedfor the periodof extended operation. This issue is Applicant/Licensee Action Item 2" [5].

BV Unit 1 Compliance This action item requires comparison of the RVI components that are within the scope of license renewal for BV Unit I to those components contained in MRP-191, Table 4-4. A detailed tabulation of the BV Unit I RVI components was completed and compared favorably to the typical Westinghouse PWR WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-6 internals components in MRP-191. All components required to be included in the BV Unit 1 program [1, 23] are consistent with those contained in MRP-191.

Several components have different materials than specified in MRP- 191, but these have no effect on the recommended MRP aging; therefore, no modifications to the program detailed in MRP-227-A need to be proposed.

This supports the requirement that the AMP shall provide assurance that the effects of aging on the BV Unit 1 RVI components within the scope of license renewal, but not included in the generic Westinghouse-designed PWR RVI components from Table 4-4 of MRP-191, will be managed for the period of extended operation.

The generic scoping and screening of the RVI as summarized in MRP-191 and MRP-232 to support the inspection sampling approach for aging management of reactor internals specified in MRP-227-A is applicable to BV Unit I with no modifications.

Conclusion BV Unit 1 complies with Applicant/Licensee Action Item 2 of the NRC SE on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs "As addressedin Section 3.2.5.3 in this SE, applicants/licenseesof CE and Westinghouse are requiredto perform plant-specific analysis either tojustify the acceptabilityof an applicant's/licensee'sexisting programs, or to identify changes to the programs that should be implemented to manage the agingof these componentsfor the period of extended operation. The results of this plant-specific analyses and a description of the plant-specificprograms 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 identifledfor this type ofplant-specific evaluation include: CE thermal shieldpositioningpins and CE in-core instrumentationthimble tubes (Section 4.3.2 in MRP-227), and Westinghouse guide tube support pins (splitpins) (Section 4.3.3 in MRP-227). This is Applicant/Licensee Action Item 3" [5].

BV Unit 1 Compliance BV Unit 1 is compliant with the requirements in Table 4-9 of MRP-227-A as applicable to Unit 1, as shown in Appendix C, Table C-3. This is detailed in the plant-specific Beaver Valley program documents for ASME Section XI [1, 4] and the plant-specific flux thimble program [19].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-7 Conclusion BV Unit I complies with Applicant/Licensee Action Item 3 of the NRC SE on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief "As discussed in Section 3.2.5.4 of this SE, the B& W applicants/licenseesshall confirm that the core support structure upperflange weld was stress relieved duringthe originalfabricationof the ReactorPressure Vessel in order to confirm the applicabilityof MRP-227, as approvedby the NRC, to theirfacility. 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 andfrequencyfor non-stress relieved B& W core supportstructure upper flange welds shall be consistent with the recommendations in MRP-227, as approvedby the NRC, for the Westinghouse and CE upper core support barrelwelds. The examination coveragefor this B& W flange weld shall conform to the staff's imposed criteriaas described in Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee'sresolution of this plant-specific action item shall be submitted to the NRC for review and approval.This is Applicant/Licensee Action Item 4" [5].

BV Unit 1 Compliance This Applicant/Licensee Action Item is not applicable to BV Unit I since it only applies to B&W plants.

Conclusion Applicant/Licensee Action Item 4 of the NRC SE on MRP-227, Revision 0 is not applicable to BV Unit I.

6.2.5 SE Applicant/Licensee Action Item 5: Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components "As addressedin Section 3.3.5 in this SE, applicants/licenseesshall identify plant-specific acceptance criteriato be appliedwhen performing the physical measurements requiredby the NRC-approved version ofMRP-227 for loss of compressibilityfor Westinghouse hold down springs, andfor distortion in the gap between the top and bottom core shroudsegments in CE units with core barrelshroudsassembled in two vertical sections. The applicant/licenseeshall include its proposed acceptancecriteriaand an explanationof how the proposedacceptance criteriaare consistent with the plants' licensing basis and 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 partof their submittal to apply the approvedversion of MRP-227. This is Applicant/Licensee Action Item 5" [5].

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-8 BV Unit 1 Compliance See Table 7-1. BV Unit 1 utilizes a Type 304 SS hold down spring; therefore, FENOC is planning to perform inspections/physical measurements on the BV Unit 1 hold-down spring according to MRP-227-A. FENOC has a commitment to develop acceptance criteria for the hold down spring physical measurements that will be consistent with the licensing basis for BV Unit 1 [5].

Conclusion BV Unit 1 complies with Applicant/Licensee Action Item 5 of the NRC SE on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components "As addressed in Section 3.3.6 in this SE, MRP-227 does notpropose to inspect the following inaccessiblecomponents: the B& W core barrelcylinders (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-formerbolts and their locking devices, and B& W core barrelassembly internalbaffle-to-baffle bolts. The MRP also identified that although the B& W core barrel assembly internalbaffle-to-baffle bolts are accessible, the bolts are non-inspectableusing currently available examination techniques.

Applicants/licensees shalljustify the acceptabilityof these componentsfor continued operation through the period of extended operationby performing an evaluation, or by proposinga scheduled replacement of the components. As part of their applicationto implement the approved version of MRP-227, applicants/licenseesshallprovide theirjustificationfor the continued operability of each of the inaccessiblecomponents and, if necessary,provide theirplanfor the replacement of the componentsfor NRC review and approval. This is Applicant/Licensee Action Item 6" [5].

BV Unit 1 Compliance This Applicant/Licensee Action Item is not applicable to BV Unit 1 since it only applies to B&W plants.

Conclusion Applicant/Licensee Action Item 6 of the NRC SE on MRP-227, Revision 0 is not applicable to BV Unit 1.

6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS Materials "As discussedin Section 3.3. 7 of this SE, the applicants/licenseesof B& W, CE, and Westinghouse reactorsare requiredto develop plant-specific analyses to be appliedfor their facilities to demonstrate that B& W IMI guide tube assembly spiders and CRGT spacer castings, WCAP- 17789-NP January 2014 Revision 1

WESTfNGHOUSE NON-PROPRIETARY CLASS 3 6-9 CE lower support columns, and Westinghouse lower supportcolumn bodies will maintain their functionality during the period of extended operation orfor additionalRVI components that may be fabricatedfrom CASS, martensiticstainless steel or precipitationhardenedstainless steel materials. These analyses shall also consider the possible loss offracture toughness in these components due to thermal and irradiationembrittlement, and may also need to consider limitations on accessibilityfor inspection and the resolution/sensitivityof the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiringaging management during development of MRP-227. That is, the requirementwould apply to componentsfabricatedfrom susceptible materialsfor which an individual licensee has determined aging management is required,for example duringtheir 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/licenseeshall include the plant-specificanalysis as part of their submittal to apply the approvedversion of MRP-227. This is Applicant/Licensee Action Item 7" [5].

BV Unit 1 Compliance Applicant/Licensee Action Item 7 from the staffs final SE on MRP-227, Revision 0 [5] states:

"For CASS, if the application of applicable screening criteria for the component's material demonstrates that the components are not susceptible to either thermal embrittlement or irradiation embrittlement, or the synergistic effects of thermal embrittlement and irradiation embrittlement combined, then no other evaluation would be necessary. For assessment of CASS materials,the licensee or applicantfor license renewal may apply the criteria in the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast A ustenitic Stainless Steel Components" (NRC ADAMS Accession No. ML003717179) as the basis for determining whether the CASS materialsare susceptible to the thermal aging mechanism [5]."

The Beaver Valley Unit 1 reactor vessel (RV) internals CASS components and the assessment of their susceptibility to thermal embrittlement (TE) are summarized in Table 6-2.

Based on the criteria of [331, the BV Unit I CASS mixer bases on upper support columns, CASS bases for upper support columns, and CASS lower support columns, are not susceptible to TE.

Conclusive confirmation of material composition under TE susceptibility thresholds was not demonstrated for the CASS stand-alone mixers, the supports, gussets, and clamps on the upper instrumentation columns, the intermediate flanges in the control rod guide tubes, nor for the bottom mounted instrumentation (BMI) cruciforms; thus, it is conservatively assumed that they are potentially susceptible to TE. The susceptibility of the mixers, intermediate flanges, and BMI cruciforms to TE was considered in the development of MRP-227-A [5]. The BV Unit I supports, gussets, and clamps on the upper instrumentation columns are CASS. However, in MRP-191, the upper instrumentation conduit and supports, gussets, and clamps were screened as wrought material (304 SS). These CASS parts were evaluated under the guidelines of the MRP-191 FMECA in support of Applicant/Licensee Action Items 1 and 2.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-10 Irradiation may also cause a material to become embrittled. The stand-alone mixers, mixer bases and bases for the upper support columns, lower support columns, and BMI cruciforms screened in at the MRP-191 irradiation screening level [9]; thus, for these components, susceptibility to irradiation embrittlement (IE) was considered in the development of MRP-227-A [5]. The intermediate flanges and the supports, gussets, and clamps on the upper instrumentation columns screened below the MRP-191 irradiation screening level; thus, they are not susceptible to IE. Of these CASS components, the lower support columns are subject to an expansion inspection under MRP-227-A.

No martensitic stainless steel, or martensitic precipitation hardened stainless steel materials were identified in the BV Unit I RV internals.

Conclusion The BV Unit 1 CASS RV internal components meet the requirements for application of MRP-227-A. The results of this CASS evaluation do not conflict with the MRP-227-A strategy for aging management of RVIs. It is concluded that continued application of the strategy of MRP-227-A will meet the requirement for managing age-related degradation of the BV Unit 1 CASS RV internal components.

Table 6-2 Summary of BV Unit I CASS Components and their Susceptibility to TE Susceptibility to TE Molybdenum (Based on the NRC CASS Component Content Casting Ferrite Content Criteria 1331)

Control rod guide tube Low 0.5 max Static >20% Potentially susceptible to intermediate flanges (1) TE (2)

Upper instrumentation Low 0.5 max Static >20% Potentially susceptible to supports, brackets and TE (2) clamps Flow mixer devices, Low 0.5 max Static >20% Potentially susceptible to with and without TE (2) thermocouple Upper support column, Low 0.5 max Static <20% Not susceptible to TE flow mixer base Upper support column Low 0.5 max Static <20% Not susceptible to TE Bases Lower support column Low 0.5 max Static _<20% Not susceptible to TE bodies Bottom-mounted Low 0.5 max Static >20% Potentially susceptible to TE (2) instrumentation (BMI),

standard cruciforms Notes:

I. Intermediate flanges may have alternate material CASS.

2. Where insufficient data are available to assess the ferrite content, the ferrite content is assumed >20% and the material is listed as potentially susceptible to TE.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-11 6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff Review and Approval "As addressedin Section 3.5.1 in this SE, applicants/licenseesshall make a submittalfor NRC review and approval to credit their implementation of MRP-227, as amended by this SE, as an AMP for the RVI components at theirfacility. This submittalshall include the information identified in Section 3.5.1 of this SE. This is Applicant/Licensee Action Item 8" [5].

BV Unit 1 Compliance BV Unit 1, per the RIS [3], is considered a Category B plant that is expected to submit their RVI AMP based on the guidance of MRP-227-A, consistent with their commitments. Per the LRA [2], BV Unit 1 has a commitment to submit their AMP for approval by the NRC no later than January 29, 2014.

Conclusion BV Unit 1 complies with Applicant/Licensee Action Item 8 of the NRC SE on MRP-227, Revision 0, and therefore meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7 PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE The requirements of MRP-227-A are based on an 18-month refueling cycle and consider both EFPY and cumulative operation. The information contained in Table 7-1 is based on this information and includes a description of the past inspections, as well as the latest scope of inspections pertaining to the reactor internals AMP. Should a change occur in plant operational practices or operating experience result in changes to the projections, appropriate updates will be performed on affected plant documentation in accordance with approved procedures.

Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary Refueling Project Estimated Outage Month/Year EFPY AMP-Related Scope(') Inspection Method and Criteria Comments 21 Spring 2012 25.3 Not applicable Not applicable Not applicable 22 Fall 2013 26.7 Not applicable Not applicable Not applicable 23 Spring 2015 28.1 Not applicable Not applicable Not applicable 24 Fall 2016 29.5 ASME Code Section XI ASME Code Section XI Extended period of operation begins 10-Year ISI at midnight on January 29, 2016 25 Spring 2018 30.9 Initial MRP-227-A augmented MRP-227-A inspections in BV Unit I plans to begin extended inspections for control rod accordance with MRP-228 operation during Cycle 24. BV has guide tube lower flange welds, specifications the option to perform these upper and lower core barrel inspections until RO-25. The flange welds, upper and lower inspection window for these core barrel cylinder girth components is plus or minus two welds, and thermal shield refueling cycles from the beginning flexures completed during or of extended operation.

before this outage 26 Fall 2019 32.3 Initial MRP-227-A augmented MRP-227-A inspections in The inspection window for the hold inspections for hold down accordance with MRP-228 down spring is plus or minus three spring completed during or specifications refueling cycles from the beginning before this outage of extended operation.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-2 Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary (cont.)

Refueling Project Estimated Outage Month/Year EFPY AMP-Related Scope(') Inspection Method and Criteria Comments 27 Spring 2021 33.7 Initial MRP-227-A augmented MRP-227-A inspections in The inspection window for baffle-inspections for baffle-former accordance with MRP-228 former bolts is between 25 and 35 bolts completed during or specifications EFPY. FENOC has the option to before this outage perform these inspections until RO-28.

28 Fall 2022 35.1 Not applicable Not applicable Not applicable 29 Spring 2024 36.5 Not applicable Not applicable Not applicable 30 Fall 2025 37.9 ASME Code Section XI ASME Code Section XI The inspection window for 17x17 10-Year ISI MRP-227-A inspections in standard guide tubes in Initial MRP-227-A augmented accordance with MRP-228 Westinghouse three-loop plants is inspections for guide plates specifications. 30 to 34 EFPY. As Beaver Valley (cards) completed during or Unit 1 was a participating plant for before this outage this analysis, an additional four EFPY can be applied to the initial inspection measurement schedule.

Therefore, the initial inspection must be performed before Beaver Valley Unit 1 reaches 38 EFPY. See WCAP- 17451-P [26] for additional information regarding the inspection schedule and requirements.

31 Spring 2027 39.3 Initial MRP-227-A augmented MRP-227-A inspections in The inspection window for these inspections for baffle-edge accordance with MRP-228 components is between 20 and 40 bolts and baffle-former specifications EFPY assembly completed during or before this outage WCAP 1 789-P Jauary201 WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-3 Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary (cont.)

Refueling Project Estimated Outage Month/Year EFPY AMP-Related Scope0') Inspection Method and Criteria Comments 32 Fall 2028 40.7 Subsequent MRP-227-A MRP-227-A inspections in The inspection window for these augmented inspections for accordance with MRP-228 components is 10 years after the control rod guide tube lower specifications initial inspection.

flange welds, upper and lower core barrel flange welds, upper and lower core barrel cylinder girth welds, and thermal shield flexures completed during or before this outage 33 Spring 2030 42.1 Not applicable Not applicable Not applicable 34 Fall 2031 43.5 Subsequent MRP-227-A MRP-227-A inspections in The inspection window for these augmented inspections for accordance with MRP-228 components is 10 years after the baffle-former bolts completed specifications initial inspection.

during or before this outage 35 Spring 2033 44.9 Not applicable Not applicable Not applicable 36 Fall 2034 46.3 Not applicable Not applicable Not applicable 37 Spring 2036 47.7 Not applicable Not applicable Not applicable N/A 48. Not applicable Not applicable Renewed Operating License expires January 29, 2036 Note:

1. Future refueling outage plans are subject to change due to considerations to coordinate and optimize outage refueling activities.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 8-1 8 IMPLEMENTING DOCUMENTS As noted within this AMP document, the BV Unit 1 PWR Vessel Internals Program is documented in NOP-CC-5004 [1]. The BV Unit 1 AMP also references the Primary Water Chemistry Program and the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program. MRP-227-A augmented examinations (Appendix C) recommended as a result of industry programs will be included in the existing ASME Section XI program.

FENOC documents associated with the existing Beaver Valley programs and considered to be implementing documents of the PWR Vessel Internals Program are:

• BVPM-CHEM-000 1, Primary Systems Strategic Water Chemistry Plan [ 18]

• ISIE-ECP-3, Flux Thimble Tube Examination Program [19]

• NOP-CC-5710, ASME Section XI Inservice Inspection (ISI) Program [4]

The RVI AMP relies on the Primary Water Chemistry Program for maintaining high water purity to reduce susceptibility to cracking due to SCC. Additional procedures may be updated or created as OE for augmented examinations is accumulated.

Based on this information, the AMP for BV Unit I RVI provides reasonable assurance that the aging effects will be managed such that the components within the scope of license renewal will continue to perform their intended functions consistent with the CLB for the period of extended operation.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 9 REFERENCES

1. Beaver Valley Nuclear Operating Procedure, NOP-CC-5004, Rev. 2, "Pressurized Water Reactor Vessel Internals Program," November 27, 2012.

2. U.S. Nuclear Regulatory Commission, NUREG-1929, "Safety Evaluation Report Related to the License Renewal of Beaver Valley Power Station, Units 1 and 2," Docket Nos. 50-334 and 50-412, FirstEnergy Nuclear Operating Company, October 2009.

3. U.S. Nuclear Regulatory Commission Document, MLI 11990086, "NRC Regulatory Issue Summary 2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management," July 21, 2011.

4. Beaver Valley Nuclear Operating Procedure, NOP-CC-5710, Rev. 1, "ASME Section XI Inservice Inspection (ISI) Program," November 30, 2012.

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

6. U.S. Nuclear Regulatory Commission, Code of Federal Regulations, 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants." Washington D.C., Federal Register, Volume 77, No. 39907, dated May 8, 1995 and last updated on July 6, 2012.

7. U.S. Nuclear Regulatory Commission Document, NUREG-1 800, Rev. 2, "Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants (SRP-LR)," December 2010.

8. Westinghouse Report, WCAP-14577, Rev. 1-A, "License Renewal Evaluation: Aging Management for Reactor Internals," March 2001.

9. MaterialsReliability Program: Screening, Categorizationand Ranking of ReactorInternals Componentsfor Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

10. MaterialsReliabilityProgram. Inspection StandardforPWR Internals - 2012 Update (MRP-228, Rev 1). EPRI, Palo Alto, CA: 2012. 1025147.

11. Westinghouse Report, WCAP-17096-NP, Rev. 2, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," December 2009.

12. Beaver Valley Business Practice, BVBP-LRP-0003, Rev. 7, "Mechanical Screening and Aging Management Review," July 12, 2007.

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

14. Beaver Valley Nuclear Operating Procedure, NOP-ER-2101, Rev. 8, "Engineering Program Management," July 11, 2013.

15. Beaver Valley Nuclear Operating Business Practice, NOBP-SS-7000, Revision 2, "EPRI Committee and User Group Member Expectations," May 18, 2006.

16. Beaver Valley Nuclear Operating Procedure, NOP-CC-5001, Revision 3, "Materials Degredation Management Program (MDMP)," July 8, 2013.

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-2

17. U.S. Nuclear Regulatory Commission Document, NUREG-1801, Rev. 2, "Generic Aging Lessons Learned (GALL) Report," December 2010.

18. Beaver Valley Program Manual, BVPM-CHEM-0001, Revision 0, "Primary Systems Strategic Water Chemistry Plan," April 22, 2013.

19. Beaver Valley Procedure, ISIE-ECP-3, Revision 7, "Flux Thimble Tube Examination Program,"

September 25, 2012.

20. FirstEnergy Engineering Change Package, ECP 06-0291-001, Revision 1, "Split Pin Modification,"

September 27, 2007.

21. Beaver Valley Power Station License Renewal Project Document, LRBV-MAMR-06B, Revision 7, "Aging Management Review of Reactor Vessel Internals," October 6, 2008.

22. ASME Boiler and Pressure Vessel Code Section XI, 2001 Edition with the 2003 Addenda.

23. FENOC Report, "Beaver Valley Power Station License Renewal Application," August 2007 (NRC ADAMS Accession Numbers ML072430916, ML072470493, and ML072470523).

24. U.S. NRC Bulletin 88-09, "Thimble Tube Thinning in Westinghouse Reactors," July 26, 1988.

25. Pressurized Water Reactor Primary Water Chemistry Guidelines, Revision 6, EPRI, Palo Alto, CA:

2007. 1014986.

26. Westinghouse Report, WCAP- 17451 -P, Rev. 1, "Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections," October 2013.

27. Beaver Valley Nuclear Operating Procedure, NOP-LP-2001, Revision 32, "Corrective Action Program," June 27, 2013.

28. FENOC Program Manual, FENOCQAP, Revision 18, "Quality Assurance Program Manual,"

November 26, 2012.

29. U.S. Nuclear Regulatory Commission Information Notice 84-18, "Stress Corrosion Cracking in Pressurized Water Reactor Systems," March 7, 1984.

30. U.S. Nuclear Regulatory Commission Information Notice 98-11, "Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants," March 25, 1998.

31. Beaver Valley Nuclear Operating Procedure, NOP-LP-2100, Revision 6, "Operating Experience Program," December 11, 2012.

32. Westinghouse Letter, PCWG-07-46, Rev. 0, "Beaver Valley Units I & 2 (DLW/DMW): Approval of Category IV PCWG Parameters to Support the Extended Power Uprate," August 27, 2007.

33. U.S. Nuclear Regulatory Commission Letter, "License Renewal Issue No. 98-0030, ThermalAging Embrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000 (NRC ADAMS Accession No. ML003717179).

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-I WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-i APPENDIX A ILLUSTRATIONS Figure A-1 Illustration of Typical Westinghouse Internals Assembly WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 Figure A-2 Typical Westinghouse Control Rod Guide Card WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3 Upper Guide Tube Upper Sup lport Plate IC Lower Guide tube Sheaths and C-Tubes Figure A-3 Lower Section of Control Rod Guide Tube Assembly WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-4 Flange Weld - Axial Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel

-, Axlal Weld Lower Barrel Circumferential Weld Lower Barrel Axial Weld Thermal Shield Flexure Core Barrel to Support Plate Weld Figure A-4 Major Core Barrel Welds WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-5

@0000 00000 40 000800 4100000 00000 J)

(D0 cu Figure A-5 Bolting Systems used in Westinghouse Core Baffles WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-6 INTERNALS SUPPORT LEDGE-THERMAL SHIELD-BAFFLE -

FORMER-LOWER CORE PLATE CORE SUPPORT COLUMN DIFFUSER PLATE CORE SUPPORT FORGING Figure A-6 Core Baffle/Barrel Structure WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7 BAFFLETO FOAMER DOLTLO4G &smIRi)

COMNER EDGE BRACKET BAFFLE TO FORMER BMLT Figure A-7 Bolting in a Typical Westinghouse Baffle-Former Structure WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-8 Figure A-8 Vertical Displacement between the Baffle Plates and Bracket at the Bottom of the Baffle-Former-Barrel Assembly WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-9 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-9 TOP SUPPORT PLATE Figure A-9 Schematic Cross-Sections of the Westinghouse Hold Down Springs W Id Figure A-10 Typical Thermal Shield Flexure WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-b Lower Core Plate Lower Core Support Structure Core Support Plate (Forging)

Figure A-II Lower Core Support Structure WCAP- 17789-NP January 2014 Revision I

WESTfNGHOUSE NON-PROPRIETARY CLASS 3 A-I WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-I II T/ T / TI/LOWER , CORE PLATE K, ..... ... ...

wz!z 2= 6VM~z ww2MDIFFUSER PLATE CORE SUPPORT PLATE/FORGING CORE SUPPORT COLUMN BOTTOM MOUNTED INSTRUMENTATION COLUMN Figure A-12 Lower Core Support Structure - Core Support Plate Cross-Section Figure A-13 Typical Core Support Column WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-12 Fu A Figure A-14 Examples of BMI Column Designs WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-1 APPENDIX B BEAVER VALLEY UNIT 1 LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

TABLE The content and numerical identifiers in Table B-I of Appendix B are extracted from Table 3.1.2-2 of the license renewal application approved by the NRC [23] and Attachment 1 of [2 1].

Table B-I Beaver Valley Unit I LRA Aging Management Review Summary Aging Effect Requiring Aging Management Component Type (I) Management Program(2) Comments

1. Core baffle/former assembly Change in dimensions PWR Vessel Internals (bolt) (B.2.33)

2. Core baffle/former assembly Cracking PWR Vessel Internals (bolt) (B.2.33)

3. Core baffle/former assembly Cracking Water Chemistry (B.2.42)

(bolt)

4. Core baffle/former assembly Cumulative fatigue TLAA (bolt) damage

5. Core baffle/former assembly Loss of fracture PWR Vessel Internals (bolt) toughness (B.2.33)

6. Core baffle/former assembly Loss of material Water Chemistry (B.2.42)

(bolt)

7. Core baffle/former assembly Loss of preload PWR Vessel Internals (bolt) (B.2.33)

8. Core baffle/former assembly Change in dimensions PWR Vessel Internals (plates) (B.2.33)

9. Core baffle/former assembly Cracking PWR Vessel Internals (plates) (B.2.33)

10. Core baffle/former assembly Cracking Water Chemistry (B.2.42)

(plates)

11. Core baffle/former assembly Cumulative fatigue TLAA (plates) damage

12. Core baffle/former assembly Loss of fracture PWR Vessel Internals (plates) toughness (B.2.33)

13. Core baffle/former assembly Loss of material Water Chemistry (B.2.42)

(plates)

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-2 Table B-I Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type () Management Program(2) Comments

14. Core barrel (shell, ring, Change in dimensions PWR Vessel Internals flange, nozzle, thermal (B.2.33) shield/pad)

15. Core barrel (shell, ring, Cracking PWR Vessel Internals flange, nozzle, thermal (B.2.33) shield/pad)

16. Core barrel (shell, ring, Cracking Water Chemistry (B.2.42) flange, nozzle, thermal shield/pad)

17. Core barrel (shell, ring, Cumulative fatigue TLAA flange, nozzle, thermal damage shield/pad)

18. Core barrel (shell, ring, Loss of fracture PWR Vessel Internals flange, nozzle, thermal toughness (B.2.33) shield/pad)

19. Core barrel (shell, ring, Loss of material Water Chemistry (B.2.42) flange, nozzle, thermal shield/pad)

20. Core barrel assembly (bolt) Change in dimensions PWR Vessel Internals (B.2.33)

21. Core barrel assembly (bolt) Cracking PWR Vessel Internals (B.2.33)

22. Core barrel assembly (bolt) Cracking Water Chemistry (B.2.42)

23. Core barrel assembly (bolt) Cumulative fatigue TLAA damage

24. Core barrel assembly (bolt) Loss of fracture PWR Vessel Internals toughness (B.2.33)

25. Core barrel assembly (bolt) Loss of material Water Chemistry (B.2.42)

26. Core barrel assembly (bolt) Loss of preload PWR Vessel Internals (B.2.33)

27. Instrumentation support Change in dimensions PWR Vessel Internals structure (flux thimble guide (B.2.33) tube)

28. Instrumentation support Cracking PWR Vessel Internals structure (flux thimble guide (B.2.33) tube)

WCAP-17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-3 Table B-i Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type () Management Program(2 ) Comments

29. Instrumentation support Cracking Water Chemistry (B.2.42) structure (flux thimble guide tube)

30. Instrumentation support Cumulative fatigue TLAA structure (flux thimble guide damage tube)

31. Instrumentation support Loss of material Water Chemistry (B.2.42) structure (flux thimble guide tube)

32. Instrumentation support Change in dimensions PWR Vessel Internals structure (thermocouple conduit) (B.2.33)

33. Instrumentation support Cracking PWR Vessel Internals structure (thermocouple conduit) (B.2.33)

34. Instrumentation support Cracking Water Chemistry (B.2.42) structure (thermocouple conduit)

35. Instrumentation support Cumulative fatigue TLAA structure (thermocouple conduit) damage

36. Instrumentation support Loss of material Water Chemistry (B.2.42) structure (thermocouple conduit)

37. Lower internals assembly Change in dimensions PWR Vessel Internals (clevis insert bolt) (B.2.33)

38. Lower internals assembly Cracking Water Chemistry (B.2.42)

(clevis insert bolt)

39. Lower internals assembly Cracking PWR Vessel Internals (clevis insert bolt) (B.2.33)

40. Lower internals assembly Cumulative fatigue TLAA (clevis insert bolt) damage

41. Lower internals assembly Loss of fracture PWR Vessel Internals (clevis insert bolt) toughness (B.2.33)

42. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(clevis insert bolt)

43. Lower internals assembly Loss of preload PWR Vessel Internals (clevis insert bolt) (B.2.33)

44. Lower internals assembly Change in dimensions PWR Vessel Internals (clevis insert) (B.2.33)

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-4 Table B-I Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type () Management Program(2) Comments

45. Lower internals assembly Cracking Water Chemistry (B.2.42)

(clevis insert)

46. Lower internals assembly Cracking PWR Vessel Internals (clevis insert) (B.2.33)

47. Lower internals assembly Cumulative fatigue TLAA (clevis insert) damage

48. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(clevis insert)

49. Lower internals assembly Loss of material ASME Section XI (clevis insert) Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.2)

50. Lower internals assembly Change in dimensions PWR Vessel Internals (Core support forging and lower (B.2.33) support column)

51. Lower internals assembly Cracking PWR Vessel Internals (Core support forging and lower (B.2.33) support column)

52. Lower internals assembly Cracking Water Chemistry (B.2.42)

(Core support forging and lower support column)

53. Lower internals assembly Cumulative fatigue TLAA (Core support forging and lower damage support column)

54. Lower internals assembly Loss of fracture PWR Vessel Internals (Core support forging and lower toughness (B.2.33) support column)

55. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(Core support forging and lower support column)

56. Lower internals assembly Change in dimensions PWR Vessel Internals (fuel alignment pin) (B.2.33)

57. Lower internals assembly Cracking Water Chemistry (B.2.42)

(fuel alignment pin)

58. Lower internals assembly Cracking PWR Vessel Internals (fuel alignment pin) (B.2.33)

59. Lower internals assembly Cumulative fatigue TLAA (fuel alignment pin) damage WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-5 Table B-1 Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type () Management Program(2) Comments

60. Lower internals assembly Loss of fracture PWR Vessel Internals (fuel alignment pin) toughness (B.2.33)

61. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(fuel alignment pin)

62. Lower internals assembly Change in dimensions PWR Vessel Internals (lower core plate) (B.2.33)

63. Lower internals assembly Cracking PWR Vessel Internals (lower core plate) (B.2.33)

64. Lower internals assembly Cracking Water Chemistry (B.2.42)

(lower core plate)

65. Lower internals assembly Cumulative fatigue TLAA (lower core plate) damage

66. Lower internals assembly Loss of fracture PWR Vessel Internals (lower core plate) toughness (B.2.33)

67. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(lower core plate)

68. Lower internals assembly Change in dimensions PWR Vessel Internals (lower support column bolt) (B.2.33)

69. Lower internals assembly Cracking PWR Vessel Internals (lower support column bolt) (B.2.33)

70. Lower internals assembly Cracking Water Chemistry (B.2.42)

(lower support column bolt)

71. Lower internals assembly Cumulative fatigue TLAA (lower support column bolt) damage

72. Lower internals assembly Loss of fracture PWR Vessel Internals (lower support column bolt) toughness (B.2.33)

73. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(lower support column bolt)

74. Lower internals assembly Loss of preload PWR Vessel Internals (lower support column bolt) (B.2.33)

75. Lower internals assembly Change in dimensions PWR Vessel Internals (radial key) (B.2.33)

76. Lower internals assembly Cracking PWR Vessel Internals (radial key) (B.2.33)

77. Lower internals assembly Cracking Water Chemistry (B.2.42)

(radial key)

WCAP-17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-6 Table B-1 Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type () Management Program(2) Comments

78. Lower internals assembly Cumulative fatigue TLAA (radial key) damage

79. Lower internals assembly Loss of material ASME Section XI (radial key) Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.2)

80. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(radial key)

81. Lower internals assembly Change in dimensions PWR Vessel Internals (secondary core support, (B.2.33) head/vessel alignment pin, head cooling spray nozzle)

82. Lower internals assembly Cracking PWR Vessel Internals (secondary core support, (B.2.33) head/vessel alignment pin, head cooling spray nozzle)

83. Lower internals assembly Cracking Water Chemistry (B.2.42)

(secondary core support, head/vessel alignment pin, head cooling spray nozzle)

84. Lower internals assembly Cumulative fatigue TLAA (secondary core support, damage head/vessel alignment pin, head cooling spray nozzle)

85. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(secondary core support, head/vessel alignment pin, head cooling spray nozzle)

86. Lower internals assembly Change in dimensions PWR Vessel Internals (Unit 1 diffuser plate) (B.2.33)

87. Lower internals assembly Cracking PWR Vessel Internals (Unit 1 diffuser plate) (B.2.33)

88. Lower internals assembly Cracking Water Chemistry (B.2.42)

(Unit 1 diffuser plate)

89. Lower internals assembly Cumulative fatigue TLAA (Unit I diffuser plate) damage

90. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(Unit I diffuser plate)

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-7 Table B-1 Beaver Valley Unit 1 LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type () Management Program(2) Comments

91. Lower internals assembly Change in dimensions PWR Vessel Internals (Unit I lower support column (B.2.33) casting)

92. Lower internals assembly Cracking PWR Vessel Internals (Unit 1 lower support column (B.2.33) casting)

93. Lower internals assembly Cracking Water Chemistry (B.2.42)

(Unit 1 lower support column casting)

94. Lower internals assembly Cumulative fatigue TLAA (Unit I lower support column damage casting)

95. Lower internals assembly Loss of fracture PWR Vessel Internals (Unit 1 lower support column toughness (B.2.33) (3) casting)

96. Lower internals assembly Loss of material Water Chemistry (B.2.42)

(Unit I lower support column casting)

97. RCCA guide tube assembly Change in dimensions PWR Vessel Internals (bolt) (B.2.33)

98. RCCA guide tube assembly Cracking PWR Vessel Internals (bolt) (B.2.33)

99. RCCA guide tube assembly Cracking Water Chemistry (B.2.42)

(bolt) 100. RCCA guide tube assembly Cumulative fatigue TLAA (bolt) damage 101. RCCA guide tube assembly Loss of material PWR Vessel Internals (bolt) (B.2.33) 102. RCCA guide tube assembly Loss of preload PWR Vessel Internals (bolt) (B.2.33) 103. RCCA guide tube assembly Change in dimensions PWR Vessel Internals (guide tube) (B.2.33) 104. RCCA guide tube assembly Cracking PWR Vessel Internals (guide tube) (B.2.33) 105. RCCA guide tube assembly Cracking Water Chemistry (B.2.42)

(guide tube)

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-8 Table B-I Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type () Management Program(2) Comments 106. RCCA guide tube assembly Cumulative fatigue TLAA (guide tube) damage 107. RCCA guide tube assembly Loss of material Water Chemistry (B.2.42)

(guide tube) 108. RCCA guide tube assembly Change in dimensions PWR Vessel Internals (support pin) (B.2.33) 109. RCCA guide tube assembly Cracking PWR Vessel Internals (support pin) (B.2.33) 110. RCCA guide tube assembly Cracking Water Chemistry (B.2.42)

(support pin)

11. RCCA guide tube assembly Cumulative fatigue TLAA (support pin) damage 112. RCCA guide tube assembly Loss of material Water Chemistry (B.2.42)

(support pin) 113. Upper internals assembly Change in dimensions PWR Vessel Internals (Core plate alignment pin) (B.2.33) 114. Upper internals assembly Cracking Water Chemistry (B.2.42)

(Core plate alignment pin) 115. Upper internals assembly Cracking PWR Vessel Internals (Core plate alignment pin) (B.2.33) 116. Upper internals assembly Cumulative fatigue TLAA (Core plate alignment pin) damage 117. Upper internals assembly Loss of material ASMEE Section XI (Core plate alignment pin) Inservice Inspection, Subsections IWB, IWC, and IWD (B.2.2) 118. Upper internals assembly Loss of material Water Chemistry (B.2.42)

(Core plate alignment pin) 119. Upper internals assembly Change in dimensions PWR Vessel Internals (fuel alignment pin) (B.2.33) 120. Upper internals assembly Cracking PWR Vessel Internals (fuel alignment pin) (B.2.33) 121. Upper internals assembly Cracking Water Chemistry (B.2.42)

(fuel alignment pin) 122. Upper internals assembly Cumulative fatigue TLAA (fuel alignment pin) damage WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-9 Table B-I Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type ) Management Program(2) Comments 123. Upper internals assembly Loss of material Water Chemistry (B.2.42)

(fuel alignment pin) 124. Upper internals assembly Change in dimensions PWR Vessel Internals (hold-down spring) (B.2.33) 125. Upper internals assembly Cracking PWR Vessel Internals (hold-down spring) (B.2.33) 126. Upper internals assembly Cracking Water Chemistry (B.2.42)

(hold-down spring) 127. Upper internals assembly Cumulative fatigue TLAA (hold-down spring) damage 128. Upper internals assembly Loss of material Water Chemistry (B.2.42)

(hold-down spring) 129. Upper internals assembly Loss of preload PWR Vessel Internals (hold-down spring) (B.2.33) 130. Upper internals assembly Change in dimensions PWR Vessel Internals (support column mixer base) (B.2.33) 131. Upper internals assembly Cracking Water Chemistry (B.2.42)

(support column mixer base) 132. Upper internals assembly Cracking PWR Vessel Internals (support column mixer base) (B.2.33) 133. Upper internals assembly Cumulative fatigue TLAA (support column mixer base) damage 134. Upper internals assembly Loss of fracture PWR Vessel Internals (support column mixer base) toughness (B.2.33) (3) 135. Upper internals assembly Loss of material Water Chemistry (B.2.42)

(support column mixer base) 136. Upper internals assembly Change in dimensions PWR Vessel Internals (support column) (B.2.33) 137. Upper internals assembly Cracking PWR Vessel Internals (support column) (B.2.33) 138. Upper internals assembly Cracking Water Chemistry (B.2.42)

(support column) 139. Upper internals assembly Cumulative fatigue TLAA (support column) damage 140. Upper internals assembly Loss of material Water Chemistry (B.2.42)

(support column)

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-10 Table B-1 Beaver Valley Unit I LRA Aging Management Review Summary (cont.)

Aging Effect Requiring Aging Management Component Type (1) Management Program(2 ) Comments 141. Upper internals assembly Change in dimensions PWR Vessel Internals (upper core plate, upper support (13.2.33) plate and support assembly) 142. Upper internals assembly Cracking PWR Vessel Internals (upper core plate, upper support (13.2.33) plate and support assembly) 143. Upper internals assembly Cracking Water Chemistry (B.2.42)

(upper core plate, upper support plate and support assembly) 144. Upper internals assembly Cumulative fatigue TLAA (upper core plate, upper support damage plate and support assembly) 145. Upper internals assembly Loss of material Water Chemistry (13.2.42)

(upper core plate, upper support plate and support assembly) 146. Upper internals assembly Change in dimensions PWR Vessel Internals (upper support column bolt) (13.2.33) 147. Upper internals assembly Cracking PWR Vessel Internals (upper support column bolt) (13.2.33) 148. Upper internals assembly Cracking Water Chemistry (B.2.42)

(upper support column bolt) 149. Upper internals assembly Cumulative fatigue TLAA (upper support column bolt) damage 150. Upper internals assembly Loss of material Water Chemistry (B.2.42)

(upper support column bolt) 151. Upper internals assembly Loss of preload PWR Vessel Internals (upper support column bolt) (13.2.33)

Notes:

1. The numbers contained in this column reflect the identical numbers in the Beaver Valley LRA table referenced [23].

2. Information in parentheses are the Appendix B section numbers in the Beaver Valley LRA [23].

3. The aging management program referenced for this component and aging effect in Table 3.1.2-2 of [23] is "Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) (B.2.40)". This has been revised to reference the "PWR Vessel Internals" program according to BV internal documentation [21 ].

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-1 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-I APPENDIX C MRP-227-A AUGMENTED INSPECTIONS Table C-1 MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Control Rod Guide All plants Loss of Material None Refer to WCAP-17451-P, Refer to WCAP-17451-P, Tube Assembly (Wear) Revision 1 [26] Revision 1 [26]

Guide plates (cards) (Note 7) See Figure A-2 (Note 7)

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

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

Upper core plate, beginning of the license (Note 2)

Lower support renewal period and See Figure A-3 forging/casting subsequent examination on a ten-year interval.

Core Barrel Assembly All plants Cracking (SCC) Lower support Periodic enhanced visual 100% of one side of the Upper core barrel flange column bodies (EVT-1) examination, no accessible surfaces of the weld (non-cast) later than 2 refueling selected weld and adjacent Core barrel outlet outages from the beginning base metal (Note 4).

nozzle welds of the license renewal period See Figure A-4 and subsequent examination on a ten-year interval.

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-2 Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

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

of the license renewal period See Figure A-4 and subsequent examination on a ten-year interval.

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 accessible surfaces of the weld (Note 5) later than 2 refueling selected weld and adjacent outages from the beginning base metal (Note 4).

of the license renewal period and subsequent examinations on a ten-year interval.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-3 Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Aibil Effect Expansion Link Examination Item Applcability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Baffle-Former All plants Cracking None Visual (VT-3) examination, Bolts and locking devices Assembly with baffle- (IASCC, with baseline examination on high-fluence seams.

Baffle-edge bolts edge bolts Fatigue) that between 20 and 40 EFPY 100% of components results in and subsequent accessible from core side

" Lost or broken examinations on a ten-year (Note 3).

locking interval. See Figures A-5, A-6, and devices A-7

" Failed or missing bolts

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

(Note 6)

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

See Figures A-5 and A-6 WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 CA Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Baffle-Former All plants Distortion (Void None Visual (VT-3) examination Core side surface, as Assembly Swelling), or to check for evidence of indicated.

Assembly Cracking distortion, with baseline See Figure A-8 (Includes: Baffle plates, (IASCC) that examination between 20 and baffle edge bolts and results in: 40 EFPY and subsequent indirect effects of void o Abnormal examinations on a ten-year swelling in former plates) interaction interval.

with fuel assemblies

" 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 joints WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-5 Table C-I MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Alignment and All plants Distortion (Loss None Direct measurement of Measurements should be Interfacing with 304 of Load) spring height within three taken at several points Components stainless Note: This cycles of the beginning of around the circumference Intemals hold down steel hold mechanism was the license renewal period. If of the spring, with a spring down not strictly the first set of measurements statistically adequate springs identified in the is not sufficient to determine number of measurements at NOTE: original list of life, spring height each point to minimize BV Unit 1 age-related measurements must be taken uncertainty.

hold down degradation during the next two outages, See Figure A-9 spring is mechanisms. in order to extrapolate the 304 SS expected spring height to 60 years.

WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-6 Table C-1 MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Effect Expansion Link Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Thermal Shield All plants Cracking None Visual (VT-3) no later than 2 100% of thermal shield Assembly with thermal (Fatigue) or refueling outages from the flexures.

Thermal shield flexures shields Loss of Material beginning of the license See Figures A-4 and A-10 (Wear) that renewal period. Subsequent results in examinations on a ten-year thermal shield interval.

flexures excessive wear, fracture, or complete separation Notes:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table C-4.

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 C-4, 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 C-4, 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.

7. Per WCAP-1745 I-P, Revision 1 [26], initial examination period requirements for guide plate (card) wear have been developed to replace the requirements in MRP-227-A

[5].

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-7 Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Item Applicability Effect Primary Link Examination Item____Applicability____ (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Upper Internals All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessible Assembly (Fatigue, Wear) flange weld examination, surfaces (Note 2).

Upper Core Plate Re-inspection every 10 years following initial inspection.

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

Lower support forging or BV Unit 1 Management Re-inspection every 10 years See Figure A-12.

castings has a lower (TE in Casting) following initial inspection.

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

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

(IE, Void See Figure A-7 Swelling and ISR)

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

Management See Figures A- 11, A-12 (1E and ISR) and A-13 WCAP- 17789-NP January 2014 Revision 1

WESTfNGHOUSE NON-PROPRIETARY CLASS 3 C-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-8 Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Item Applicability Effect (Mechanism) Primary (Note Link

1) Examination(Note 1)

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

(IE of lower See Figure A-4 sections)

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

(IE) See Figure A-4 Lower Support All plants Cracking Upper core barrel Enhanced visual (EVT-1) 100% of accessible Assembly NOTE: (IASCC) flange weld examination, surfaces (Note 2).

Lower support column Not Aging Re-inspection every 10 years See Figures A-11, A-12, bodies applicable Management following initial inspection, and A- 13 (non cast) to BV (IE)

Unit I Lower Support All plants Cracking Control rod guide Visual (EVT-1) 100% of accessible Assembly (IASCC) tube (CRGT) examination, support columns (Note 2).

Lower support column including the lower flanges Re-inspection every 10 years See Figures A-11, A- 12, bodies detection of following initial inspection. and A-13 (cast) fractured support columns Aging Management (IE)

WCAP- 17789-NP January 2014 Revision 1

WESTI`NGHOUSE NON-PROPRIETARY CLASS 3 C-9 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-9 Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Examination Item d/Frequency (Note 1)

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

fractured Re-inspection every 10 years See Figures A-12 and A-column bodies following initial inspection. 14 Aging Flux thimble Management insertion/withdrawal to be (LE) monitored at each inspection interval.

Notes:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table C-4.

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).

WCAP- I7789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-i 0 Table C-3 MRP-227-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Core barrel flange (Wear) Section XI to determine general specified frequency.

condition for excessive wear.

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

Upper support ring or skirt Lower Internals All plants Cracking ASME Code Visual (VT-3) examination All accessible surfaces at Assembly (IASCC, Section XI of the lower core plates to specified frequency.

Lower core plate Fatigue) detect evidence of distortion XL lower core plate Aging and/or loss of bolt integrity.

(Note 1) Management (IE)

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

Lower core plate XL lower core plate (Note 1)

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

Januaiy 2014 WCAP- 17789-NP January 2014 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-11 Table C-3 MIRP-227-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals (cont.)

Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage Alignment and All plants Loss of material ASME Code Visual (VT-3) examination. All accessible surfaces at Interfacing (Wear) Section XI specified frequency.

Components (Note 2)

Clevis insert bolts Alignment and All plants Loss of material ASME Code Visual (VT-3) examination. All accessible surfaces at Interfacing (Wear) Section XI specified frequency.

Components Upper core plate alignment pins Notes:

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.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-12 Table C-4 MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (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.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-13 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link (s) Expansion Criteria Additinal Eaitio (Note 1) Acceptance Criteria Control Rod Guide All plants Enhanced visual a. Bottom- a. Confirmation of a. For BMI column Tube Assembly (EVT-1) mounted surface-breaking bodies, the specific Lower flange welds examination instrumentation indications in two or more relevant condition for The specific (BMI) column CRGT lower flange welds, the VT-3 examination is relevant condition bodies combined with flux completely fractured is a detectable b. Lower support thimble column bodies.

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

b. Confirmation of crack-like surface surface-breaking indication.

indications in two or 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.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-14 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-14 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1) Acceptance__riteria Core Barrel Assembly All plants Periodic enhanced a. Core barrel a. The confirmed detection a and b. The specific Upper core barrel flange visual (EVT-1) outlet nozzle and sizing of a surface- relevant condition for weld examination, welds breaking indication with a the expansion core

b. Lower support length greater than two barrel outlet nozzle weld The specific column bodies inches in the upper core and lower support relevant condition (non cast) barrel flange weld shall column body is a detectable require that the EVT- I examination is a crack-like surface examination be expanded detectable crack-like indicationk to include the core outlet surface 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 follow the initial observation.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-15 C-i 5 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1) Acceptance Criteria Core Barrel Assembly All plants Periodic enhanced None None None Lower core barrel flange visual (EVT-1) weld (Note 2) examination.

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

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

indication. I examination be expanded to include the upper core barrel cylinder axial welds by the completion of the next refueling outage.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-16 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

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

indication. 1 examination be expanded to include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

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.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-17 Table C4 MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (NoteI) A1)eptance__riteria Baffle-Former All plants Volumetric (UT) a. Lower support a. Confirmation that more a and b. The Assembly examination, column bolts than 5% of the baffle- examination acceptance Baffle-former bolts The examination former bolts actually criteria for the UT of the acceptance criteria b. Barrel-former examined on the four lower support column for the UT of the b el- r baffle plates at the largest bolts and the barrel-baffle-former bolts distance from the core former bolts shall be shall be established (presumed to be the lowest established as part of the as part of the dose locations) contain examination technical examination unacceptable indications justification.

technical shall require UT justification. examination of the lower support column bolts 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.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-18 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination ExamiationAdditional 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.

WCAP- 17789-NP January 2014 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-19 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-i 9 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (NoteI) Acceptance)Criteria Alignment and All plants Direct physical None N/A N/A Interfacing with 304 measurement or Components stainless spring height.

Internals hold down steel hold The examination spring down acceptance springs criterion for this NOTE: measurement is BV Unit 1 that the remaining hold down compressible spring is height of the spring 304 SS shall provide hold-down forces within the plant-specific design tolerance.

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:

I. The examination acceptance criterion for visual examination is the absence of the specified relevance 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.

WCAP- 17789-NP January 2014 Revision I