beaver valley power station unit 1, wcap-17789-np, rev. 1 ... · 3.4 m anager site design...

Enclosure AL-13-398

PWR Vessel Internals Program Plan for Aging Management of Reactor Internals atBeaver Valley Power Station Unit 1, Revision 1

(102 pages follow)

Westinghouse Non-Proprietary Class 3

WCAP-17789-NP January 2(Revision 1

PWR Vessel InternalsProgram Plan for AgingManagement ofReactor Internals atBeaver Valley Power StationUnit 1

Westinghouse

)14

WESTINGHOUSE NON-PROPRIETARY CLASS 3

WCAP-17789-NPRevision 1

PWR Vessel Internals Program Plan for Aging Managementof 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 LLC1000 Westinghouse Drive

Cranberry Township, PA 16066

© 2014 Westinghouse Electric Company LLCAll 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 P O 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 ................................................................................... 3-13.5 M anager Site Chemistry .................................................................................................. 3-23.6 Site PW R Vessel Internals Program Owner .................................................................... 3-2

3.7 Fleet PW R Vessel Internals Program Owner ................................................................... 3-33.8 Outage M anagement ........................................................................................................ 3-3

4 DESCRIPTION OF THE BEAVER VALLEY UNIT 1 REACTOR INTERNALS AGINGMANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS ............................................. 4-14.1 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 .......................................................... 4-44.1.3 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 ................................... 4-54.2.2 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

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 iv

5.1 GALL Revision 2 Element 1: Scope of Program ........................................................... 5-15.2 GALL Revision 2 Element 2: Preventive Actions .......................................................... 5-35.3 GALL Revision 2 Element 3: Parameters Monitored or Inspected ................................ 5-45.4 GALL Revision 2 Element 4: Detection of Aging Effects ............................................. 5-55.5 GALL Revision 2 Element 5: Monitoring and Trending .............................................. 5-105.6 GALL Revision 2 Element 6: Acceptance Criteria ...................................................... 5-1 15.7 GALL Revision 2 Element 7: Corrective Actions ........................................................ 5-125.8 GALL Revision 2 Element 8: Confirmation Process .................................................... 5-145.9 GALL Revision 2 Element 9: Administrative Controls ................................................ 5-145.10 GALL Revision 2 Element 10: Operating Experience ................................................. 5-15

6 D EM O N STR A TIO N .................................................................................................................... 6-16.1 Demonstration of Topical Report Conditions Compliance to SE on MRP-227,

R ev isio n 0 ........................................................................................................................ 6 -26.2 Demonstration of Applicant/Licensee Action Item Compliance to SE on MRP-227,

R ev isio n 0 ........................................................................................................................ 6 -36.2.1 SE Applicant/Licensee Action Item I: Applicability of FMECA and

Functionality A nalysis A ssum ptions ............................................................... 6-36.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components

within the Scope of License Renewal ............................................................. 6-56.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-

Specific Existing Program s ............................................................................. 6-66.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper

F lange Stress R elief ......................................................................................... 6-76.2.5 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&WC o m po nents ..................................................................................................... 6-8

6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASSM ate ria ls .......................................................................................................... 6 -8

6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for StaffR 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 R E F E R E N C E S ............................................................................................................................. 9-1

APPENDIX A ILLUSTRATIONS .................................................................................................... A-1

APPENDIX B BEAVER VALLEY UNIT I LICENSE RENEWAL AGING MANAGEMENTREVIEW SUMMARY TABLE ................................................................................ B-1

APPENDIX C MRP-227-A AUGMENTED INSPECTIONS ......................................................... C-1

WCAP- 17789-NP January 2014Revision 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


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


WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii

LIST OF ACRONYMS

AMP Aging Management Program PlanAMR Aging Management ReviewASME American Society of Mechanical EngineersB&PV Boiler and Pressure VesselB&W Babcock & WilcoxBMI bottom-mounted instrumentationBV Beaver ValleyBVPS Beaver Valley Power StationBWR boiling water reactorCASS cast austenitic stainless steelCE Combustion EngineeringCFR Code of Federal RegulationsCLB current licensing basisCRGT control rod guide tubeECP Engineering Change PackageEFPY effective full-power yearsEPRI Electric Power Research InstituteET electromagnetic testing (eddy current)EVT enhanced visual testing (a visual NDE method that includes EVT-1)FENOC FirstEnergy Nuclear Operating CompanyFMECA failure modes, effects, and criticality analysisGALL Generic Aging Lessons LearnedI&E Inspection and EvaluationIASCC irradiation-assisted stress corrosion crackingINPO Institute of Nuclear Power OperationsISI inservice inspectionISR irradiation-enhanced stress relaxationLRA License Renewal ApplicationLRAAI license renewal applicant action itemsMRP Materials Reliability ProgramNDE nondestructive examinationNEI Nuclear Energy InstituteNOS Nuclear Oversight SectionNRC U.S. Nuclear Regulatory CommissionNSSS nuclear steam supply systemOE Operating ExperienceOEM Original Equipment ManufacturerOER Operating Experience ReportPH precipitation-hardenable (heat treatment)PWR pressurized water reactorPWROG Pressurized Water Reactor Owners Group (formerly WOG)PWSCC primary water stress corrosion cracking


WESTINGHOUSE NON-PROPRIETARY CLASS 3 viii

LIST OF ACRONYMS (cont.)

QA quality assuranceRCC rod cluster controlRCS Reactor Coolant SystemRIS Regulatory Issue SummaryRO refueling outageRV reactor vesselRVI reactor vessel internalsSCC stress corrosion crackingSE Safety EvaluationSER Safety Evaluation ReportSRP Standard Review PlanSS stainless steelTE thermal embrittlementUFSAR Updated Final Safety Analysis ReportUT 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 OperatorsWOG Westinghouse Owners GroupXL extra-long Westinghouse fuel

Trademark Statement:

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

WCAP-17789-NP January 2014Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix

ACKNOWLEDGEMENTS

The authors would like to thank Wesley Williams and Zach Warchol of FirstEnergy Nuclear OperatingCompany and our associates at Westinghouse for their efforts in supporting development of this WCAP.


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, hereafterBeaver 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 documentprovides 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 BVUnit 1 operating procedure NOP-CC-5004 [1] in support of license renewal program evaluations. ThisAMP is supported by existing BV Unit I documents and procedures and, as needed by industryexperience 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 internalscomponents. 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 andPressure 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 reactorinternals 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 PressurizedWater 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. Participate in the industry programs applicable to BVPS Unit 1 for investigating and managing

aging effects on reactor internals;

2. Evaluate and implement the results of the industry programs as applicable to the B VPS Unit Ireactor 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 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 NuclearOperating Company (FENOC), or in cooperation with the industry, to be equivalent or more rigorous thancurrently 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 reasonableassurance 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].

WCAP- 17789-NP January 2014



2 BACKGROUND

The management of aging degradation effects in reactor internals is required for nuclear plantsconsidering 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 theindustry goal of responding to these requirements. Various programs have been underway within theindustry over the past decade to develop guidelines for managing the effects of aging within PWR reactorinternals. In 1997, the WOG issued WCAP-14577 [8], "License Renewal Evaluation: AgingManagement for Reactor Internals," which was reissued as Revision 1-A in 2001 after receiving NRCStaff review and approval. Later, an effort was engaged by the EPRI MRP to address the PWR internalsaging 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 internalscomponents using proven and familiar methods for inspection, monitoring, surveillance, andcommunication. 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 ofPWR internals components to each of eight postulated aging mechanisms (further discussed inSection 4 of this Program).

PWR internals components were categorized, based on the screening criteria, into categories thatranged 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 PWRinternals components and assemblies of components using irradiated and aged materialproperties, to determine the effects of the degradation mechanisms on component functionality.

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



The industry guidance is contained within two separate EPRI MRP documents:

MRP-227-A [5], "PWR Internals Inspection and Evaluation Guidelines," (hereafter referred to asthe "I&E Guidelines" or simply "MRP-227-A") provides the industry background, listing ofreactor internals components requiring inspection, type of NDE required for each component,timing for initial inspections, and criteria for evaluating inspection results. MRP-227-A providesa 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 thequalification/demonstration of the NDE techniques and other criteria pertaining to the actualperformance of the inspections.

The PWROG has also developed WCAP-17096-NP, Revision 2, "Reactor Internals Acceptance CriteriaMethodology and Data Requirements" for the MRP-227 inspections, where feasible [11]. This documenthas been submitted to the NRC for review and approval, and will be updated to incorporate changes fromMRP-227-A [5]. Final reports are to be developed and available for industry use in support of plannedlicense renewal inspection commitments. In some cases, individual plants will develop plant-specificacceptance 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 Westinghousethree-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 coresupport structure (also known as the "lower internals"), the upper core support structure (also known asthe "upper internals"), and the in-core instrumentation support structure (includes components that arepart 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 coolantflow; and, providing shielding.

The lower core support structure assembly consists of the core barrel, the core baffle, the lower core plateand support columns, the thermal shield or neutron shield pads, and the core support welded to the corebarrel. A ledge in the reactor vessel supports the lower core support structure at its upper flange and aradial support system attached to the vessel wall restrains its lower end from transverse motion. Withinthe core barrel, an axial baffle and a lower core plate are attached to the core barrel wall and form theenclosure periphery of the assembled core. The lower core support structure and core barrel control andprovide passageways for coolant flow. The lower core plate, positioned at the bottom level of the corebelow 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 ofmaterial 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 supportcolumns establishing the spacing between the upper support assembly and the upper core plate are



fastened at the top and bottom to these plates. They transmit mechanical loadings between the uppersupport 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 arefastened to the upper support and oriented and supported by pins in the upper core plate. The upper guidetube 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 partsof the "upper internals") to support thermocouples penetrating the vessel through the head and a lowersystem (components of which are part of the "lower internals") to support flux thimbles penetrating thevessel 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 uppersupport plate at positions, are above their readout locations.

The lower in-core instrumentation support system uses reactor vessel bottom-mounted instrumentationcolumns (flux thimble guide tubes) which guide and protect the retractable, cold-worked stainless steelflux thimbles that are pushed upward into the reactor core. The thimbles, closed at the leading ends, arethe pressure boundary between the reactor pressurized water and the containment atmosphere. All reactorvessel 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 safetyevaluation report (SER) in NUREG-1929 [2]. In the SER, the NRC concluded that the BV Unit 1License Renewal Application (LRA) adequately identified the RV internals components within the scopeof license renewal, as required by 10 CFR 54.4(a), and those subject to an AMR, as required by 10 CFR54.21(a)(1) [6] and; therefore, is acceptable. A listing of the BV Unit I reactor vessel internalscomponents and subcomponents, already reviewed by the NRC in the SER that are subject to AMPrequirements, is included in Table B-I.

In accordance with 10 CFR Part 54 [6], frequently referred to as the License Renewal Rule, BV Unit 1has developed a program to direct the performance of aging management reviews of mechanicalstructures and components [12]. The U.S. industry, as noted through the efforts of the MRP andPWROG, has further investigated the components and subcomponents that require aging management tosupport continued reliable function. As designated by the protocols of NEI 03-08 [13], "Guidelines forthe Management of Materials Issues", each plant will be required to use MRP-227-A and MRP-228 todevelop and implement an AMP for reactor internals no later than three years after the initial industryissuance of MRP-227, Revision 0. MRP-227, Revision 0 was issued in December 2008, and plant AMPsmust therefore be completed by December 2011, or sooner, if required by plant-specific License RenewalCommitments. According to [3], BV Unit 1 is a Category B plant that is expected to submit their RVIAMP based on the guidance of MRP-227-A, consistent with their commitments. Per the LRA [2], BVUnit 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 thereferenced documents. The AMP will manage aging effects of the RVI so that the intended functions willbe maintained consistent with the current licensing basis for the period of extended operation.

WCAP- 17789-NP January 2014Revision 1


3 SITE PWR VESSEL INTERNALS PROGRAM OWNER

The PWR Vessel Internals Program [1] manages the effects of age-related degradation mechanisms ofreactor vessel internals. The successful implementation and comprehensive long-term management of theBV Unit I RVI AMP will require the integration of FirstEnergy organizations, corporately and at BeaverValley, 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 BeaverValley groups are provided in the following paragraphs. FENOC will maintain cognizance of industryactivities related to PWR internals inspection and aging management, and will address/implementindustry 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 ServicesEngineering.

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

3.1 SITE VICE PRESIDENT

Has responsibility for ensuring that sufficient financial and manpower resources are madeavailable 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 includesexamination, 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 RVIProgram.

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 associatedcomponents.

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

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

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



3.5 MANAGER SITE CHEMISTRY

Ensures primary system Chemistry controls are adequately implemented, maintained and complywith regulatory requirements and appropriate industry guidelines.

Ensures that Chemistry program changes required by regulation, fostered by industry guidancedocuments or identified as prudent for maintaining RCS integrity and reliability are evaluated andimplemented 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/orinterfacing/affected site programs.

Ensures that examination, repair and assessment activities of affected components/materialscomply with regulatory commitments and Industry guidance provided by the EPRI MRP, PWROwners Groups, Institute of Nuclear Power Operations (INPO), World Association of NuclearOperators (WANO), and/or other appropriate Industry organizations

Ensures preparation and review of PWR RVI related program documents, license amendmentrequests, relief requests, required reports, and other documents submitted to the NRC.

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

Coordinates the generation and maintenance of the site-specific PWR RVIInspection/Implementation Plan. Maintenance of the Inspection/Implementation Plan includesperiodic reviews to ensure that the plan reflects current industry experience and data, includingadvancements in mitigation capabilities and strategies.

Ensures that the examinations detailed in the site PWR RVI Program inspection plan areperformed 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) Programo Reactor Engineeringo Nuclear Fuels and Core Design

Performs and coordinates strategic planning for PWR RVI Program Components. Strategicplanning includes, but may not be limited to, planning for examinations and mitigation activitiesnecessary to lessen the detrimental consequences associated with PWR RVI degradation.

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



Completes the program health report in accordance with NOP-ER-2101 [141 for the PWR RVIProgram.

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 PWRVessel Internals issues.

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

* Initiates and coordinates the review and evaluation of industry guidance documents related toPWR vessel internals issues.

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

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

3.8 OUTAGE MANAGEMENT

Ensures outage schedules needed to support PWR Vessel Internals Program activities, includingimplementation of ECPs, are complete and maintained.



4 DESCRIPTION OF THE BEAVER VALLEY UNIT 1 REACTORINTERNALS AGING MANAGEMENT PROGRAMS ANDINDUSTRY 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 agingmanagement programs based on inspection and evaluation. The intent of this program is to ensure thelong-term integrity and safe operation of the reactor internals components. FENOC has developed thisAMP in conformance with the 10 Generic Aging Lessons Learned (GALL) [17] attributes andMRP-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], inspectionsprescribed by the ASME Section XI Inservice Inspection Program [4], thimble tube inspections [19], andmitigation projects such as support pin replacement [20], combined with augmented inspections orevaluations as recommended by MRP-227-A.

Aging degradation mechanisms that impact internals have been identified and documented in BV Unit 1Aging Management Reviews [21 ] prepared using the business practice document [12] in support of thelicense renewal effort. The overall outcome of the reviews and the additional work performed by theindustry, as summarized in MRP-227-A, is to provide appropriate augmented inspections for reactorinternals 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 worksummarized in MRP-227-A. All sources are consistent and address concerns about componentdegradation resulting from the following eight material aging degradation mechanisms identified asaffecting reactor internals:

Stress Corrosion Cracking (SCC)

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

* Irradiation-Assisted Stress Corrosion Cracking

Irradiation-assisted stress corrosion cracking (IASCC) is a unique form of SCC that occurs onlyin highly irradiated components. The aging effect is cracking.

* Wear

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



Fatigue

Fatigue is defined as the structural deterioration that can occur as the result of repeatedstress/strain cycles caused by fluctuating loads and temperatures. After repeated cyclic loading ofsufficient magnitude, microstructural damage can accumulate, leading to macroscopic crackinitiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loadingcan lead to growth of the initiated crack. Corrosion fatigue is included in the degradationdescription.

Low-cycle fatigue is defined as cyclic loads that cause significant plastic strain in the highlystressed regions, where the number of applied cycles is increased to the point where the crackeventually initiates. When the cyclic loads are such that significant plastic deformation does notoccur in the highly stressed regions, but the loads are of such increased frequency that a fatiguecrack eventually initiates, the damage accumulated is said to have been caused by high-cyclefatigue. 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, andpresence of deleterious chemical species. Cracks typically initiate at local geometric stressconcentrations such as notches, surface defects, and structural discontinuities. The aging effect iscracking.

* 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 highinservice 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 atnormal operating temperatures for CASS, martensitic stainless steel, and PH stainless steelinternals. CASS components have a duplex microstructure and are particularly susceptible to thismechanism. While the initial aging effect is loss of ductility and toughness, unstable crackextension is the eventual aging effect if a crack is present and the local applied stress intensityexceeds 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 bechanged. Such changes in mechanical properties include increasing yield strength, increasingultimate strength, decreasing ductility, and a loss of fracture toughness. The irradiationembrittlement aging mechanism is a function of both temperature and neutron fluence. While theinitial aging effect is loss of ductility and toughness, unstable crack extension is the eventualaging effect if a crack is present and the local applied stress intensity exceeds the reduced fracturetoughness.



Void Swelling and Irradiation Growth

Void swelling is defined as a gradual increase in the volume of a component caused by formationof microscopic cavities in the material. These cavities result from the nucleation and growth ofclusters of irradiation-produced vacancies. Helium produced by nuclear transmutations can havea significant impact on the nucleation and growth of cavities in the material. Void swelling mayproduce dimensional changes that exceed the tolerances on a component. Strain gradientsproduced 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. Alsoincluded in this mechanism is irradiation growth of anisotropic materials, which is known tocause significant dimensional changes within in-core instrumentation tubes that are fabricatedfrom zirconium alloys. While the initial aging effect is dimensional change and distortion, severevoid 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 orcreep. Thermal stress relaxation (or primary creep) is defined as the unloading of preloadedcomponents due to long-term exposure to elevated temperatures, as seen in PWR internals. Stressrelaxation occurs under conditions of constant strain where part of the elastic strain is replacedwith plastic strain. Available data show that thermal stress relaxation appears to reach saturationin a short time (< 100 hours) at PWR internals temperatures.

Creep (or more precisely, secondary creep) is a slow, time- and temperature-dependent, plasticdeformation 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 constantstress. Secondary creep in austenitic stainless steels is associated with temperatures higher thanthose relevant to PWR internals even after taking into account gamma heating. However,irradiation-enhanced creep (or more simply, irradiation creep) or irradiation-enhanced stressrelaxation (ISR) is an athermal process that depends on the neutron fluence and stress, and it canalso be affected by void swelling should it occur. The aging effect is a loss of mechanical closureintegrity (or preload) that can lead to unanticipated loading that, in turn, may eventually causesubsequent 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 agingmanagement program as described in NUREG-1801, GALL Report Section XI.M16A for PWR VesselInternals. In the BV Unit I RVI AMP, this is demonstrated through application of existing BV AMRmethodology 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. Adescription of the applicable existing BV programs and compliance with the elements of the GALL iscontained 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 thefollowing existing programs [23]:



* 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 theRVI components. Although affiliated with and supporting the RVI AMP, they will be managed under theexisting 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] isin accordance with ASME Section XI 2001 Edition with the 2003 Addenda [22] and is subject to thelimitations and modifications of 10 CFR 50.55a. The program provides for condition monitoring of Class1, 2, and 3 pressure-retaining components, including welds, pump casings, valve bodies, integralattachments, 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 isaugmented 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 wearprior to leakage by monitoring for and predicting unacceptable levels of wall thinning in the MovableIncore Detector System Flux Thimble Tubes, which serve as a Reactor Coolant System (RCS) pressureboundary. The program implements the recommendations of NRC IE Bulletin 88-09, Thimble TubeThinning in Westinghouse Reactors [24].

The main attribute of the program is periodic nondestructive examination (NDE) of the flux thimble tubeswhich provides actual values of existing tube wall thinning. This information provides the basis for anextrapolation to determine when tube wall thinning will progress to an unacceptable value. Based on thisprediction, preemptive actions are taken to reposition, replace or isolate the affected thimble tube prior toa 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 bycorrosion and stress corrosion cracking. The Primary Water Chemistry Program relies on monitoring andcontrol 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 WaterChemistry Program for the circumstances identified in NUREG-1 801 that require augmentation of thePrimary Water Chemistry Program.



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 therequirements 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 DocumentLRBV-MAMR-06B [21] documents the results of the aging management review performed in support ofBV Unit I license renewal for reactor internals. The BV Unit 1 LRA was approved by the NRC inNUREG-1929 [2]. RVI components specifically noted as requiring aging management, as identified inthe NUREG, are summarized in Appendix B Table B-1 of this AMP.

The AMR supported the LRA as follows:

1. Identified applicable aging effects requiring management

2. Associated aging management programs to manage those aging effects

3. Identified enhancements or modifications to existing programs, new aging managementprograms, or any other actions required to support the conclusions reached in the review

Aging management reviews were performed for each BV Unit 1 system that contained long-lived, passivecomponents 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 RVIAMP.

4.2.2 Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic StainlessSteel (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. RVIswill be inspected in accordance with ASME Code Section XI, Subsection IWB, Category B-N-3. Thisinspection will be augmented to detect the effects of loss of fracture toughness due to thermal aging andneutron irradiation embrittlement of CASS components. The program will include identification of thelimiting susceptible components from the standpoint of thermal aging susceptibility, neutron fluence, andcracking. For each identified component, aging management will be accomplished through either asupplemental examination or a component-specific evaluation, including a mechanical loadingassessment. BV Unit 1 will participate in the EPRI Materials Reliability Program established toinvestigate the impacts of aging on PWR vessel internal components. The results of this project willprovide additional basis for the inspections and evaluations performed under this program. Refer toAppendix B, Section B.2.40 of the LRA [23] for more information regarding this program.



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 tubeassembly into the top of the core plate. In general, SCC prevention is aided by adherence to strict primarywater chemistry limits that effectively prevent SCC and greatly reduce the probability of IASCC. Thelimits imposed by the Primary Water Chemistry Program at BV Unit 1 are consistent with the latestEPRI 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 hours of operation. Prior toreplacement, numerous support pins at other plants using alloy X-750 material with the same heattreatment as that of the BV Unit I pins failed during removal or during operation between 110,900 and149,000 hours 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 hours.

In response to the industry concern for SCC of the alloy X-750 material, FENOC replaced all of the upperinternals guide tube support pins at BV Unit 1 (September/October 2007) with Westinghouse-suppliedcold worked Type 316 Stainless Steel support pins to mitigate the possibility of continued SCC of thesecomponents. Detailed descriptions of the replacement are contained within FENOC Engineering ChangePackage, 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 atechnical evaluation of aging degradation mechanisms and aging effects for Westinghouse RVIcomponents. The WOG sent the report to the NRC staff to demonstrate that WOG member plant ownersthat subscribed to the WCAP could adequately manage effects of aging on RVI during the period ofextended operation, using approved aging management methodologies of the WCAP to developplant-specific aging management programs.

The AMR for the BV Unit I internals, documented in [21] was completed in accordance with therequirements 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 utilityrepresentatives, NSSS vendors, independent consultants, and international committee representatives whoreviewed available data and industry experience on materials aging. The objective of the group was todevelop a consistent, systematic approach for identifying and prioritizing inspection and evaluationrequirements for reactor internals. The following subsections briefly describe the industry process.



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 forspecific RVI components. MRP-227-A credited existing component inspections, when they were deemedadequate, as a result of detailed expert panel assessments conducted in conjunction with the developmentof the industry document. Through the elements of the process, the reactor internals for all currentlylicensed and operating PWR designs in the United States were evaluated in the MRP program; andappropriate 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 AdditionalMeasures" components, as described as follows:

* Primary

Those PWR internals that are highly susceptible to the effects of at least one of the eight agingmechanisms were placed in the Primary group. The aging management requirements that areneeded to ensure functionality of Primary components are described in the I&E guidelines. ThePrimary group also includes components that have shown a degree of tolerance to a specificaging degradation effect, but for which no highly susceptible component exists or for which nohighly susceptible component is accessible.

* Expansion

Those PWR internals that are highly or moderately susceptible to the effects of at least one of theeight aging mechanisms, but for which functionality assessment has shown a degree of toleranceto those effects, were placed in the Expansion group. The schedule for implementation of agingmanagement requirements for Expansion components depends on the findings from theexaminations 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 agingmechanisms and for which generic and plant-specific existing AMP elements are capable ofmanaging 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 screeningcriteria were placed in the No Additional Measures group. Additional components were placed inthe 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 formanaging the aging of the No Additional Measures components.

The categorization and analysis used in the development of MRP-227-A are not intended to supersedeany ASME B&PV Code Section XI [22] requirements. Any components that are classified as core



support structures, as defined in ASME B&PV Code Section XI IWB-2500, Category B-N-3, haverequirements 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 ofthe NEI 03-08 protocols. The MRP-227-A guideline includes Mandatory and Needed elements asfollows:

0 Mandatory

There is one Mandatory element:

1. Each commercial U.S. PWR unit shall develop and document a program for management ofaging of reactor internals components within thirty-six months following 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 inDecember 2008. An AMP must therefore be developed by December 2011. To fulfill thisrequirement and the license renewal commitments provided in Section 1, FENOC developedNOP-CC-5004, Revision 0, "Pressurized Water Reactor Vessel Internals Program" [1]. Thisprogram 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 Bplant because they have a renewed license with a commitment to submit an AMP/inspection planbased on MRP-227 but that have not yet been required to do so by their commitment. This AMPfulfills the license renewal commitment to submit an implementation schedule for BV Unit 1 inaccordance 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 andTables 5-1 through 5-3for the applicable design within twenty-four months following issuance ofMRP-227-A.

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

2. Examinations specified in the MRP-227-A guidelines shall be conducted in accordance withInspection Standard, MRP-228 [10].



BV Unit 1 Applicability: Inspection standards will be in accordance with the requirements ofMRP-228 [10]. These inspection standards will be used for augmented inspection at BV Unit 1as applicable where required by MRP-227-A directives.

3. Examination results that do not meet the examination acceptance criteria defined in Section 5 ofthe MRP-22 7-A guidelines shall be recorded and entered in the plant corrective action programand dispositioned.

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

4. Each commercial U.S. PWR unit shall provide a summary report of all inspections andmonitoring, items requiring evaluation, and new repairs to the MRP Program Manager within120 days of the completion of an outage during which PWR internals within the scope ofMRP-227-A are examined

BV Unit 1 Applicability: As discussed in Section 4.3.4, FENOC will participate in futureindustry 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 theexamination acceptance criteria in Section 5, this engineering evaluation shall be conducted inaccordance with a NRC-approved evaluation methodology.

BV Unit 1 Applicability: BV Unit I will evaluate any examination results that do not meet theexamination 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 theattributes that make up an acceptable AMP. These attributes are consistent with the BV Unit 1 AgingManagement Review process. Evaluation of the BV Unit 1 RVI AMP against GALL attribute elements isprovided in Section 5 of this AMP.

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



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-Aassumptions:

30 years of operation with high-leakage core loading patterns (fresh fuel assemblies loaded inperipheral locations) followed by implementation of a low-leakage fuel management strategy forthe remaining 30 years of operation.

BV Unit 1 Applicability: According to the BV RVI Program [1], Unit I had approximately 17years of operation with fresh fuel assemblies at peripheral locations. The history of the Unit Icore 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 coreloading pattern. No change to the low leakage core design philosophy is anticipated for theextended plant operating license.

Base load operation, i.e., typically operates at fixed power levels and does not usually vary poweron a calendar or 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 theoriginal vendors.

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 madeany modifications to the Unit 1 internals beyond those identified in general industry guidance orrecommended by the original vendor since May 2007. Therefore, there are no differences incomponent inspection categories [1].

Based on the plant-specific applicability, as stated, the MRP-227-A work is representative for BeaverValley 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 operationof reactor upper internals guide tubes from a guide card and lower guide tube continuous guidance wearstandpoint, as well as to establish an initial inspection schedule based on the various guide tube designsfor 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 FleetOperational Projections" [26] which developed a guide plate (card) initial inspection schedule forWestinghouse NSSS designed plants. The intent of this industry guidance is to replace the current guideplate (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 Section5.4 of the WCAP [26], the generic initial guide card and continuous guidance inspection measurement

WCAP-17789-NP January 2014Revision 1


EFPY range for this guide tube design is 30 to 34 effective full-power years (EFPY). Beaver ValleyUnit 1 was evaluated as a part of this technical basis and therefore, an alternative initial inspectionmeasurement 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 relatedto 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. PWRplants, and development of acceptance criteria and inspection disposition processes. FENOC willmaintain cognizance of industry activities related to PWR internals inspection and aging management.FENOC will also address/implement industry guidance, stemming from those activities, as appropriateunder NEI 03-08 practices.

4.4 SUMMARY

It should be noted that the FENOC BV Unit 1, the MRP, and the PWROG approaches to agingmanagement are based on the GALL approach to aging management strategies. This approach includes adetermination of which reactor internals passive components are most susceptible to the agingmechanisms of concern and then determination of the proper inspection or mitigating program thatprovides reasonable assurance that the component will continue to perform its intended function throughthe period of extended operation. The GALL-based approach was used at Beaver Valley for the initialbasis 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 andapproved documents. The additional evaluations and analysis completed by the MRP industry grouphave provided clarification to the level of inspection quality needed to determine the proper examinationmethod and frequencies. The tables provided in MRP-227-A and included as Appendix C of this AMPprovide 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 withany additional augmented inspections required by the MRP-227-A industry tables provided in AppendixC, provides reasonable assurance that the reactor internals passive components will continue to performtheir intended functions through the period of extended operation.



5 BEAVER VALLEY REACTOR INTERNALS AGINGMANAGEMENT PROGRAM ATTRIBUTES

The BV Unit 1 RVI AMP is credited for aging management of RVI components for the following eightaging 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), SectionXI.M I 6A, "PWR Vessel Internals" [ 17] are described in this section. The GALL identifies 10 attributesfor successful component aging management. The framework for assessing the effectiveness of theprojected 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 agingmanagement review of the reactor internals in the license renewal process. However, FENOC made acommitment (see NUREG-1929 [2]) to incorporate the following: (1) BV Unit 1 will continue toparticipate in the industry programs applicable to BV Unit I for investigating and managing aging effectson reactor internals, (2) evaluate and implement the results of the industry programs as applicable to theBV Unit 1 reactor internals; and, (3) upon completion of these programs, but not less than 24 monthsbefore entering the period of extended operations, submit an inspection plan for the BV Unit 1 reactorinternals to the NRC for review and approval.

This AMP is consistent with that process and includes consideration of the augmented inspectionsidentified 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 NuclearPlant, which is built to a Westinghouse NSSS design. The scope of the program applies themethodology and guidance in the most recently NRC-endorsed version of MRP-22 7, whichprovides augmented inspection and flaw evaluation methodology for assuring the functionalintegrity of safety-related internals in commercial operating U.S. PWR nuclear power plantsdesigned by B& W, CE, and Westinghouse. The scope of components considered for inspectionunder MRP-227 guidance includes core support structures (typically denoted as ExaminationCategory B-N-3 by the ASME Code, Section XI), those RVI components that serve an intendedlicense renewal safety function pursuant to criteria in 10 CFR 54.4(a) (1), and other R VI



components whose failure could prevent satisfactory accomplishment of any of the functionsidentified in 10 CFR 54.4(a) (1)(1), (ii), or (iii). The scope of the program does not includeconsumable items, such as fuel assemblies, reactivity control assemblies, and nuclearinstrumentation, because these components are not typically within the scope of the componentsthat are required to be subject to an aging management review (AMR), as defined by the criteriaset in 10 CFR 54.21(a)(1). The scope of the program also does not include welded attachmentsto the internal surface of the reactor vessel because these components are considered to be ASMECode Class I appurtenances to the reactor vessel and are adequately managed in accordancewith an applicant's AMP that corresponds to GALL AMP XI.M1, 'ASME Code, Section XAInservice Inspection, Subsections IWB, IWC, and IWD. '

The scope of the program includes the response bases to applicable license renewal applicantaction items (LRAAIs) on the MRP-227 methodology, and any additional programs, actions, oractivities that are discussed in these LRAAI responses and credited for aging management of theapplicant's RVI components. The LRAAIs are identified in the staffs safety evaluation on MRP-227 and include applicable action items on meeting those assumptions that formed the basis ofthe MRP's augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 ofMRP-227), and NSSS vendor-specific orplant-specific LRAAIs as well. The responses to theLRAAIs on MRP-227 are provided in Appendix C of the LRA.

The guidance of MRP-22 7 specifies applicability limitations to base-loaded plants and the fuelloading management assumptions upon which the functionality analyses were based Theselimitations and assumptions require a determination of applicability by the applicant for eachreactor and 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 knownas 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 "upperinternals" or the "lower internals"). Additional RVI details are provided in Section 3.2.2.2 of the BV UnitI Updated Final Safety Analysis Report (UFSAR).

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

The BV Unit 1 Reactor Internals AMR was conducted and documented in LRBV-MAMR-06B [21]. Thetable 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 requiremanagement for the components requiring AMR. A column in the tables lists the program/activity that iscredited to address the component and aging effect during the period of extended operation. The NRChas reviewed and approved the aging management strategy presented in the Appendix B tables asdocumented in the SER on license renewal [2].

WCAP-17789-NP January 2014Revision 1


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 andcharacterizing of the aging effects (cracks, loss of material, loss of preload, etc.) of interest, prescribedfrequency of inspection, and examination acceptance criteria. The information provided in MRP-227-A isrooted in the GALL methodology. The basic assumptions of MRP-227-A, Section 2.4 are met by BVUnit I and are addressed in subsection 4.3.2.4 of this AMP. The Topical Report Conditions andApplicant/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 forthe Topical Report Conditions and in Section 6.2 for the Applicant/Licensee Action Items. The BV Unit 1RVI AMP scope is additionally based on previously established and approved GALL Report approachesthrough application of the MRP-227-A [5] methodologies to determine those components that requireaging 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


"The guidance in MRP-227 relies on PWR water chemistry control to prevent or mitigate agingeffects that can be induced by corrosive aging mechanisms (e.g., loss of material induced bygeneral, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms[SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintained inaccordance with the Water Chemistry Program. The program description, evaluation, andtechnical 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 programthat complies with the requirements of this element. A description and applicability to the BV Unit 1 RVIAMP 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, chemistryprograms 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 chemistryto keep peak levels of various contaminants below the system-specific limits. The BV Unit I PWRPrimary Water Chemistry Program [ 18] is based on the current, approved revisions of EPRI PWRPrimary Water Chemistry Guidelines.

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


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 ORINSPECTED


"The program manages the following age-related degradation effects and mechanisms that areapplicable in general to the R VI components at the facility: (a) cracking induced by SCC,PWSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss offracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d)changes in dimension due to void swelling and irradiation growth, distortion, or deflection; and(e)loss ofpreload caused by thermal and irradiation-enhanced stress relaxation or creep. Forthe management of cracking, the program monitors the evidence of surface breaking lineardiscontinuities if a visual inspection technique is used as the non-destruction examination (NDE)method, or for relevant flaw presentation signals ifa volumetric UT method is used as the NDEmethod. For the management of loss of material, the program monitors for gross or abnormalsurface conditions that may be indicative of loss of material occurring in the components. Forthe management of loss ofpreload, the program monitors for gross surface conditions that maybe indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. Theprogram does not directly monitor for loss offracture toughness that is induced by thermal agingor neutron irradiation embrittlement, or by void swelling and irradiation growth; instead, theimpact of loss offracture toughness on component integrity is indirectly managed by using visualor volumetric examination techniques to monitor for cracking in the components and by applyingapplicable reduced fracture toughness properties in the flaw evaluations if cracking is detected inthe components and is extensive enough to warrant a supplemental flaw growth or flaw toleranceevaluation under MRP-227 guidance or ASME Code, Section XI requirements. The programuses physical measurements to monitor for any dimensional changes due to void swelling,irradiation growth, distortion, or deflection.

Specifically, the program implements the parameters monitored/inspected criteria forWestinghouse designed Primary Components in Table 4-3 ofMRP-227. Additionally, theprogram implements the parameters monitored/inspected criteria for Westinghouse designedExpansion Components in Table 4-6 of MRP-227. The parameters monitored/inspected forExisting Program Components follow the bases for referenced Existing programs, such as therequirements for ASME Code Class RVI components in ASME Code, Section XI, Table IWB-2500-1, Examination Categories B-N-3, as implemented through the applicant's ASME Code,Section XI program, or the recommended program for inspecting Westinghouse-designed fluxthimble tubes in GALL AMP XIM3 7, "Flux Thimble Tube Inspection." No inspections, exceptfor those specified in ASME Code, Section X7, are required for components that are identified asrequiring "No Additional Measure, " in accordance with the analyses reported in MRP-22 7"[17].



BV Unit 1 Parameters Monitored or Inspected

The BV Unit 1 AMP monitors, inspects, and/or tests for the effects of the eight aging degradationmechanisms 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 industrydirectives 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 ofMRP-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 ExistingComponent inspections from Table 4-9 of MRP-227-A (included in Appendix C of this AMP asTable 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 inAppendix 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 currentrequirements in MRP-227-A [5].

For license renewal, the ASME Section XI Program [4] consists of periodic volumetric, surface, and/orvisual examination of components for assessment, signs of degradation, and corrective actions. Therequirements 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 andthe 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


"The detection of aging effects is covered in two places. (a) the guidance in Section 4 of MRP-227 provides an introductory discussion and justification of the examination methods selected fordetecting the aging effects of interest; and (b) standards for examination methods, procedures,and personnel are provided in a companion document, MRP-228. In all cases, well-establishedmethods were selected. These methods include volumetric UT examination methods for detectingflaws in bolting, physical measurements for detecting changes in dimension, and various visual

(VT-3, VT-i, and EVT-1) examinations for detecting effects ranging from general conditions to

detection and sizing of surface-breaking discontinuities. Surface examinations may also be usedas an alternative to visual examinations for detection and sizing of surface-breaking

discontinuities.



Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1examination (for internals other than bolting) or by volumetric UT examination (bolting). TheVT-3 visual methods may be applied for the detection of cracking only when the flaw tolerance ofthe component or affected assembly, as evaluated for reduced fracture toughness properties, isknown and has been shown to be tolerant of easily detected large flaws, even under reducedfracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspect forloss of material induced by wear and for general aging conditions, such as gross distortioncaused by void swelling and irradiation growth or by gross effects of loss ofpreload caused bythermal and irradiation-enhanced stress relaxation and creep.

In addition, the program adopts the recommended guidance in MRP-227 for defining theExpansion criteria that needed to be applied to inspections of Primary Components and ExistingRequirement Components and for expanding the examinations to include additional ExpansionComponents. As a result, inspections performed on the RVI components are performed consistentwith the inspection frequency and sampling bases for Primary Components, ExistingRequirement Components, and Expansion Components in MRP-227, which have beendemonstrated to be in conformance with the inspection criteria, sampling basis criteria, andsample Expansion criteria in Section A. 1.2.3.4 of NRC Branch Position RLSB-1.

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

The program is supplemented by the following plant-specific Primary Component and ExpansionComponent inspections for the program (as applicable): for BV Unit 1, no additional Primary orExpansion components are relevant to the scope of aging management for the RVI.

In addition, in some cases (as defined in MRP-227), physical measurements are used assupplemental techniques to manage for the gross effects of wear, loss ofpreload due to stressrelaxation, or for changes in dimension due to void swelling, deflection or distortion. Thephysical measurements methods applied in accordance with this program include that for thehold down spring. The hold down spring at BV Unit 1 isfabricatedfrom Type 304 SS thatrequires 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 establishedand field-proven through the application of the Section XI ISI Program. Those augmented inspectionsthat are taken from the MRP-227-A recommendations will be applied through use of the MRP-228inspection standard. This AMP implements the augmented inspection requirements of Table 4-3, Table4-6, 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 inspectionfrequency and sampling basis. For the Expansion Components of MRP-227-A, this AMP implements theexpansion requirements of Table 5-3 of MRP-227-A (included in Appendix C of this AMP as Table C-4).



Inspection can be used to detect physical effects of degradation including cracking, fracture, wear, anddistortion. The choice of an inspection technique depends on the nature and extent of the expecteddamage. The recommendations supporting aging management for the reactor internals, as contained inthis report, are built around three basic inspection techniques: (1) visual, (2) ultrasonic, and (3) physicalmeasurement The three different visual techniques include VT-3, VT-i, and EVT-1. The assumptionsand 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 augmentedreactor 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 supportstructures and interior attachments to reactor vessels) and B-N-3 (removable core support structures) aredefined 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 theallowable linear flaw standards of IWB-3510 [22]

Structural degradation of attachment welds such that the original cross-sectional area is reducedby 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 intentis 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 visualexaminations with some estimate of the scale of deformation or wear. Note that in MRP-227-A, VT-Ihas only been selected to detect distortion as evidenced by small gaps between the upper-to-lower matingsurfaces of CE-welded core shrouds assembled in two vertical sections. Therefore, no additional VT-Iinspections 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 visualexamination has been identified for inspection of components where surface-breaking flaws are apotential concern. Any visual inspection for cracking requires a reasonable expectation that the flawlength and crack mouth opening displacement meet the resolution requirements of the observationtechnique. The EVT-I specification augments the VT-I requirements to provide more rigorousinspection standards for stress corrosion cracking and has been demonstrated for similar inspections inboiling water reactor (BWR) internals. Enhanced visual examination (i.e., EVT-1) is also conducted inaccordance with the requirements described for visual examination (i.e., VT-i) with additionalrequirements (such as camera scanning speed). Any recommendation for EVT-1 inspection will requireadditional analysis to establish flaw-tolerance criteria, which must take into account potentialembrittlement due to thermal aging or neutron irradiation. The industry, through the PWROG, hasdeveloped an approach for acceptance criteria methodologies to support plant-specific augmentedexaminations. This work is summarized in WCAP-17096-NP, "Reactor Internals Acceptance Criteria



Methodology and Data Requirements" [11 ]. The acceptance criteria developed using these methodologiesmay be created on either a generic or plant-specific basis because both loads and component dimensionsmay 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 visualexamination has been identified for inspection of components where general condition monitoring isrequired. The VT-3 examination is intended to identify individual components with significant levels ofexisting degradation. As the VT-3 examination is not intended to detect the early stages of componentcracking or other incipient degradation effects, it should not be used when failure of an individualcomponent could threaten either plant safety or operational stability. The VT-3 examination may beappropriate for inspecting highly redundant components (such as baffle-edge bolts), where a single failuredoes not compromise the function or integrity of the critical assembly.

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

* Structural distortion or displacement of parts to the extent that component function may beimpaired;

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

* Foreign materials or accumulation of corrosion products that could interfere with control rodmotion 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 isreduced more than 5 percent.

The VT-3 examination is intended for use in situations where the degradation is readily observable. It ismeant to provide an indication of condition, and quantitative acceptance criteria are not generallyrequired. In any particular recommendation for VT-3 visual examination, it should be possible to identifythe specific conditions of concern. For instance, the unacceptable conditions for wear indicate wear thatmight lead to loss of function. Guidelines for wear in a critical-alignment component may be verydifferent 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 cansupplement either visual (VT-3) or (VT-1/EVT-1) examinations specified in these guidelines. This



supplemental examination may thus be used to reject or accept relevant indications. A surfaceexamination is an examination that indicates the presence of surface discontinuities, and the ASMIEB&PV Code [22] lists magnetic particle, liquid penetrant, eddy current, and ultrasonic examinationmethods as surface examination alternatives. Here, only the electromagnetic testing (ET), also called eddycurrent surface examination method, is covered.

When selected for use as a supplemental examination to examinations performed in these guidelines, anET 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 theinspected object by electromagnetic coils, with disruptions in the eddy current flow caused by surface ornear-surface anomalies detected by suitable instrumentation. Industry experience with ET examination isrelatively robust, especially in the aerospace and petroleum refinery industries. The experience base forPWR 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 anddetermine the length and depth of a crack in a component. Although access to the surface of thecomponent is required to apply the ultrasonic signals, the flaw may exist in the bulk of the material. Inthis proposed strategy, UT inspections have been recommended exclusively for detection of flaws inbolts. For the bolt inspections, any bolt with a detected flaw should be assumed to have failed. The sizeof the flaw in the bolt is not critical because crack growth rates are generally high, and it is assumed thatthe observed flaw will result in failure prior to the next inspection opportunity. It has generally beenobserved through examination performance demonstrations that UT can reliably (90 percent or greaterreliability) 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 thereactor internals are highly redundant. For any system of bolts, it is possible to demonstrate multipleacceptable bolting patterns. The evaluation program must demonstrate that the remaining bolts meet therequirements for an acceptable bolting pattern for continued operation. The evaluation procedures mustalso demonstrate that the pattern of remaining bolts contains sufficient margin such that continuation ofthe bolt failure rate will not result in failure of the system to meet the requirements for an acceptablebolting pattern before the next inspection.

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

Physical Measurement Examination

Continued functionality can be confirmed by physical measurements to evaluate the impact caused byvarious degradation mechanisms such as wear or loss of functionality as a result of loss of preload or



material deformation. For BV Unit 1, direct physical measurements are required only for the hold downspring.

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


"The methods for monitoring, recording, evaluating, and trending the data that result from theprogram's inspections are given in Section 6 of MRP-227 and its subsections. The evaluationmethods include recommendations for flaw depth sizing and for crack growth determinations aswell for performing applicable limit load, linear elastic and elastic-plastic fracture analyses ofrelevant flaw indications. The examinations and re-examinations required by the MRP-227guidance, together with the requirements specified in MRP-228 for inspection methodologies,inspection procedures, and inspection personnel, provide timely detection, reporting, andcorrective actions with respect to the effects of the age-related degradation mechanisms withinthe scope of the program. The extent of the examinations, beginning with the sample ofsusceptible PWR internals component locations identified as Primary Component locations, withthe potential for inclusion of Expansion Component locations if the effects are greater thananticipated, plus the continuation of the Existing Programs activities, such as the ASME Code,Section X7, Examination Category B-N-3 examinations for core support structures, provides ahigh 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 andcontrol rod guide tube split pin cracking issues were identified by the industry and continue to be activelymanaged. The extremely low frequency of failure in reactor internals makes monitoring and trendingbased on OE somewhat impractical. The majority of the materials aging degradation models used todevelop the MRP-227-A guidelines are based on test data from reactor internals components removedfrom service. The data are used to identify trends in materials degradation and forecast potentialcomponent degradation. The industry continues to share both material test data and OE through theauspices of the MRP and PWROG. FENOC has in the past and will continue to maintain cognizance ofindustry 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 theaugmented inspections derived from MRP-227-A and repeated here in Appendix C. The MiRP-227-Ainspections only augment and do not replace the existing ASME Section X1 ISI requirements. Theseinspections, where practical, are scheduled to be conducted in conjunction with typical 10-year ISIexaminations.



Appendix C, Tables C-1, C-2, and C-3 identify the augmented Primary and Expansion inspection andmonitoring 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 fordemonstrating 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 thePrimary 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 ofinspection results across all PWR designs will enable the industry to monitor reactor internals degradationon an ongoing industry basis as the period of extended operation moves forward. Reporting ofexamination 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


"Section 5 ofMRP-227 provides specific examination acceptance criteria for the Primary andExpansion Component examinations. For components addressed by examinations referenced toASME Code, Section XI, the IWB-3500 acceptance criteria apply. For other components covered

by Existing Programs, the examination acceptance criteria are described within the ExistingProgram reference document.

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

* For visual examination (and surface examination as an alternative to visual examination),

the examination acceptance criterion is the absence of any of the specific, descriptive

relevant conditions; in addition, there are requirements to record and disposition surfacebreaking indications that are detected and sized for length by VT-i/EVT-i examinations;

* For volumetric examination, the examination acceptance criterion is the capability forreliable detection of indications in bolting, as demonstrated in the examination Technical

Justification,; in addition, there are requirements for system-level assessment of bolted or

pinned assemblies with unacceptable volumetric (UT) examination indications that exceed

specified limits; and

" For physical measurements, the examination acceptance criterion for the acceptabletolerance in the measured differential height from the top of the plenum rib pads to the vessel



seating surface in B& Wplants are given in Table 5-1 of MRP-22 7. The acceptance criterionfor physical measurements performed on the height limits of the Westinghouse-designed hold-down springs are required for 304 SS hold down springs. BV Unit 1 has a 304 SS hol downspring, therefore, BV Unit I is required to produce acceptance criteria for the physicalmeasurements 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 ISIprogram scope are evaluated in accordance with the applicable requirements of the ASME Code throughthe existing Corrective Action Program [27].

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

Augmented inspections, as defined by the MRP-227-A requirements included in this AMP asAppendix C, Table C-I, Table C-2, and Table C-3, that result in recordable relevant conditions will beentered into the plant Corrective Action Program and addressed by appropriate actions that may includeenhanced inspection, repair, replacement, mitigation actions, or analytical evaluations. An example of ananalytical evaluation is using an acceptable bolting WCAP approach such as those commonly used tosupport continued component or assembly functionality. Additional analysis to establish acceptablebolting pattern evaluation criteria for the baffle-former bolt assembly is also considered in determiningthe 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 linewith accepted and proven methodologies to support this element. One of these tools is the PWROGdocument WCAP- 17096-NP, "Reactor Internals Acceptance Criteria Methodology and DataRequirements," [11], which details acceptance criteria methodology for the MRP-227 Primary andExpansion components. Status is monitored through direct FENOC cognizance of industry (includingPWROG) 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


"Corrective actions following the detection of unacceptable conditions are fundamentallyprovided for in each plant's corrective action program. Any detected conditions that do not



satisfy the examination acceptance criteria are required to be dispositioned through the plantcorrective action program, which may require repair, replacement, or analytical evaluation for

continued service until the next inspection. The disposition will ensure that design basis functionsof the reactor internals components will continue to be fulfilled for all licensing basis loads and

events. Examples of methodologies that can be used to analytically disposition unacceptable

conditions are found in the ASME Code, Section XI or in Section 6 ofMRP-227. Section 6 ofMRP-227 describes the options that are available for disposition of detected conditions that

exceed the examination acceptance criteria of Section 5 of the report. These include engineering

evaluation methods, as well as supplementary examinations to further characterize the detectedcondition, or the alternative of component repair and replacement procedures. The latter aresubject to the requirements of the ASME Code, Section X7. The implementation of the guidance inMRP-227, plus the implementation of any ASME Code requirements, provides an acceptablelevel of aging management of safety-related components addressed in accordance with the

corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Other alternative corrective action bases may be used to disposition relevant conditions if theyhave been previously approved or endorsed by the NRC. Examples ofpreviously NRC-endorsed

alternative corrective actions bases include those corrective actions bases for 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 ofWestinghouse Report No. WCAP-145 77-Rev. ]-A, or for B& W-designed RVI components in B& WReport No. BA W-2248. Westinghouse Report No. WCAP-14577-Rev. 1-A was endorsed for use inan NRC SE to the Westinghouse Owners Group, dated February 10, 2001. B& W Report No.BA W-2248 was endorsed for use in an SE to Framatome Technologies on behalf of the B& WOwners Group, dated December 9, 1999. Alternative corrective action bases not approved orendorsed by the NRC will be submitted for NRC approval prior to their implementation" [17].

BV Unit 1 Corrective Action

The existing Beaver Valley procedure for corrective actions, the "Corrective Action Program" [27] andthe ASME Section XI ISI program [4], will be credited for this element. These procedures establish theBV Unit 1 repair and replacement requirements of ASME Code Section XI, "Rules for InserviceInspection of Nuclear Power Plant Components" [22]. These requirements include the identification of arepair cycle, repair plan, and verification of acceptability for replacements. BV Unit 1 is committed toperforming 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.



5.8 GALL REVISION 2 ELEMENT 8: CONFIRMATION PROCESS


"Site quality assurance procedures, review and approval processes, and administrative controlsare implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, or theirequivalent, as applicable. It is expected that the implementation of the guidance in MRP-227 willprovide an acceptable level of quality for inspection, flaw evaluation, and other elements of agingmanagement of the PWR internals that are addressed in accordance with the 10 CFR Part 50,Appendix B, or their equivalent (as applicable), confirmation process, and administrativecontrols" [171.

BV Unit 1 Confirmation Process

BV Unit 1 has an established 10 CFR Part 50, Appendix B, Program [28] that addresses the elements ofcorrective actions, confirmation process, and administrative controls. The BV Unit I Program includesnon-safety-related structures, systems, and components. Quality assurance (QA) procedures, review andapproval processes, and administrative controls are implemented in accordance with the requirements of10 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


"The administrative controls for such programs, including their implementing procedures andreview and approval processes, are under existing site 10 CFR 50 Appendix B Quality AssurancePrograms, or their equivalent, as applicable. Such a program is thus expected to be establishedwith a sufficient level of documentation and administrative controls to ensure effective long-termimplementation " [ 17].

BV Unit 1 Administrative Controls

BV Unit I has an established 10 CFR Part 50, Appendix B Program [28] that addresses the elements ofcorrective actions, confirmation process, and administrative controls. The BV Unit 1 program includesnon-safety-related structures, systems, and components. Quality assurance (QA) procedures, review andapproval processes, and administrative controls are implemented in accordance with the requirements of10 CFR 50, Appendix B.



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


"Relatively few incidents ofPWR internals aging degradation have been reported in operatingUS. commercial PWR plants. A summary of observations to date is provided in Appendix A ofMRP-22 7-A. The applicant is expected to review subsequent operating experience for impact onits program or to participate in industry initiatives that perform this function.

The application of the MRP-227 guidance will establish a considerable amount of operatingexperience over the next few years. Section 7 ofMRP-227 describes the reporting requirementsfor these applications, and the plan for evaluating the accumulated additional operatingexperience" [17].

BV Unit 1 Operating Experience

Extensive industry and BV Unit 1 OE has been reviewed during the development of the RVI AMP. Theexperience reviewed includes NRC Information Notices 84-18, "Stress Corrosion Cracking in PWRSystems" [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-formerbolts or SCC of high-strength internals bolting. SCC of control rod guide tube support pins has also beenreported.

Early plant OE related to hot functional testing and reactor internals is documented in plant historicalrecords. Inspections performed as part of the 10-year ISI program have been conducted as designated byexisting 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 applicableprocedure [31 ], for the determination of additional actions and lessons learned. These insights, asapplicable, can be incorporated into the plant systems quarterly health reports and further evaluated forincorporation into plant programs.

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



* 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 upperinternals guide tube support pins at BV Unit 1 (September - October 2007) with Westinghouse suppliedcold 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, ECP06-0291-001 [20].

* BV Unit 1 Flux Thimble Tubes

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

A key element of the MRP-227-A guideline is the reporting of age-related degradation of RVIcomponents. FENOC, through its participation in PWROG and EPRI-MRP activities, will continue tobenefit from the reporting of inspection information and will share its own OE with the industry throughthe reporting requirements of Section 7 of MRP-227-A. The collected information from MRP-227-Aaugmented 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.



6 DEMONSTRATION

Beaver Valley Unit 1 has demonstrated a long-term commitment to aging management of reactorinternals. This AMP is based on an established history of programs to identify and monitor potentialaging degradation in the reactor internals. Programs and activities undertaken in the course of fulfillingthat commitment include:

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

* As documented in Beaver Valley operational procedures, reports are continuously reviewed byBeaver Valley personnel for applicable issues that indicate operating procedures or programsrequire updates based on new OE.

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

* The Primary Water Chemistry Program at Beaver Valley has been effective in maintainingoxygen, halogens, and sulfate at levels sufficiently low to prevent SCC of the reactor vesselinternals.

* Replacement control rod guide tube support pins for BV Unit 1 in 2007 were fabricated from coldworked 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 minimumnumber would be satisfied [23].

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

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

This AMP fulfills the approved license renewal methodology requirement to identify the most susceptiblecomponents and to inspect those components with an indication detection level commensurate with theexpected degradation mechanism indication. Augmented inspections, derived from the informationcontained in MRP-227-A, the industry I&E Guidelines, have been utilized in this AMP to build onexisting plant programs. This approach is expected to encourage detection of a degradation mechanism atits first appearance consistent with the ASME approach to inspections. This approach providesreasonable assurance that the internals components will continue to perform their intended functionthrough the period of extended operation.

Typical ASME Section XI examinations identified in the AMP for the period of extended operation are tobe performed at BV Unit I in Fall 2016, refueling outage 25 (RO-25). The augmented inspections



discussed in compliance with MRP-227-A requirements have been integrated in the implementationschedule, which is shown in Section 7. Integration of the required inspections will be tracked tocompletion. As discussed, the industry MRP-227-A guidelines also provide for updates as experience isgained through inspection results. This feedback loop will enable updates based on actual inspectionexperience.

The augmented inspections described in this document, as summarized in Appendix C, combined with theASME Section XI ISI program inspections, existing Beaver Valley programs, and use of OperatingExperience Reports (OERs), provide reasonable assurance that the reactor internals will continue toperform 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 itemsthat came out of the NRC review of MRP-227, as listed in [5], as well as their compliance within thisAMP.

6.1 DEMONSTRATION OF TOPICAL REPORT CONDITIONS COMPLIANCE TOSE ON MRP-227, REVISION 0

Table 6-1 Topical Report Condition Compliance to SE on MRP-227

Topical Condition Applicable/Not Compliance in AMPApplicable

1. High consequence components in Applicable The upper core plate and the lower supportthe "No Additional Measures" forging or casting components are added toInspection Category Table C-2 as "Expansion Components"

linked to the "Primary Component," theCRGT lower flange weld.

2. Inspection of components subject to Applicable The upper and lower core barrel cylinderirradiation-assisted stress corrosion girth welds and the lower core barrel flangecracking weld are moved from Table C-2 "Expansion

Components" to Table C-I "PrimaryComponents."

3. Inspection of high consequence Not Not applicable for BV Unit 1components subject to multiple Applicabledegradation 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, theformer 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 everycomponents component under the Examination

Method/Frequency column in Table C-2.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-3

Table 6-1 Topical Report Condition Compliance to SE on MRP-227

(cont.)

Topical Condition Applicable/Not Compliance in AMPApplicable

7. Updating of MRP-227, Revision 0, Applicable Section 5 is updated to reflect XI.MI6AAppendix A from GALL Revision 2 [17].

6.2 DEMONSTRATION OF APPLICANT/LICENSEE ACTION ITEMCOMPLIANCE TO SE ON MRP-227, REVISION 0

6.2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and FunctionalityAnalysis Assumptions

"As addressed in Section 3.2.5.1 of this SE, each applicant/licensee is responsible for assessingits plant's design and operating history and demonstrating that the approved version of MRP-22 7is applicable to the facility. Each applicant/licensee shall refer, in particular, to the assumptionsregarding plant design and operating history made in the FMECA and functionality analyses forreactors of their design (i.e., Westinghouse, CE, or B& W) which support MRP-22 7 and describethe process used for determining plant-specific differences in the design of their RVI componentsor plant operating conditions, which result in different component inspection categories. Theapplicant/licensee shall submit this evaluation for NRC review and approval as part of itsapplication to implement the approved version ofMRP-227. This is Applicant/Licensee ActionItem 1" [5].

BV Unit 1 Compliance

The process used to verify that BV Unit I is reasonably represented by the generic industry programassumptions with regard to neutron fluence, temperature, materials, and stress values used in thedevelopment 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 IPWR 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 thosematerials identified in MRP- 191, Table 4-4.



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

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

5. Confirmation that the BV Unit 1 RVI materials operated at temperatures within theoriginal 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 theapplication of the MRP-227-A generic aging management strategy.

BV Unit 1 reactor internals components are reasonably represented by the design and operating historyassumptions regarding neutron fluence, temperature, materials, and stress values in the MRP-191 genericFMECA 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 withregard to neutron fluence.

a. The FMECA and functionality analyses for MRP-227-A were based on theassumption of 30 years of operation with high-leakage core loading patterns followedby 30 years of low-leakage core fuel management strategy. BV Unit 1 hadapproximately 17 years of operation with fresh fuel assemblies at peripheral locations(high-leakage core loading pattern). The low-leakage loading pattern has beenapplied to all subsequent core designs through current operation. No change to thelow-leakage core design philosophy is anticipated for the extended plant operatinglicense [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 fluenceand fuel management assumptions in MRP-191 and requirements for MRP-227-Aapplication.

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

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

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



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 thosematerials 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 thecomponent is already credited as being managed under an alternate BV Unit 1aging management program.

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

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

Conclusion

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

6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within theScope of License Renewal

"As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressed in 10 CFR54.4, each applicant/licensee is responsible for identifying which RVI components are within thescope of LR for its facility. Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tablescontain all of the RVI components that are within the scope of LR for their facilities inaccordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are withinthe scope of LR for its facility, the applicant or licensee shall identify the missing component(s)and propose any necessary modifications to the program defined in MRP-227, as modified by thisSE, when submitting its plant-specific AMP. The AMP shall provide assurance that the effects ofaging on the missing component(s) will be managed for the period of extended operation. Thisissue is Applicant/Licensee Action Item 2" [5].


This action item requires comparison of the RVI components that are within the scope of license renewalfor BV Unit I to those components contained in MRP-191, Table 4-4. A detailed tabulation of the BVUnit I RVI components was completed and compared favorably to the typical Westinghouse PWR



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 therecommended MRP aging; therefore, no modifications to the program detailed in MRP-227-A need to beproposed.

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

The generic scoping and screening of the RVI as summarized in MRP-191 and MRP-232 to support theinspection sampling approach for aging management of reactor internals specified in MRP-227-A isapplicable 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, andtherefore meets the requirement for application of MRP-227-A as a strategy for managing age-relatedmaterial degradation in reactor internals components.

6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-SpecificExisting Programs

"As addressed in Section 3.2.5.3 in this SE, applicants/licensees of CE and Westinghouse arerequired to perform plant-specific analysis either tojustify the acceptability of anapplicant 's/licensee's existing programs, or to identify changes to the programs that should beimplemented to manage the aging of these components for the period of extended operation. Theresults of this plant-specific analyses and a description of the plant-specific programs beingrelied on to manage aging of these components shall be submitted as part of theapplicant's/licensee 's AMP application. The CE and Westinghouse components identifled for thistype ofplant-specific evaluation include: CE thermal shield positioning pins and CE in-coreinstrumentation thimble tubes (Section 4.3.2 in MRP-227), and Westinghouse guide tube supportpins (split pins) (Section 4.3.3 in MRP-227). This is Applicant/Licensee Action Item 3" [5].


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



Conclusion

BV Unit I complies with Applicant/Licensee Action Item 3 of the NRC SE on MRP-227, Revision 0, andtherefore meets the requirement for application of MRP-227-A as a strategy for managing age-relatedmaterial degradation in reactor internals components.

6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper FlangeStress Relief

"As discussed in Section 3.2.5.4 of this SE, the B& W applicants/licensees shall confirm that thecore support structure upper flange weld was stress relieved during the original fabrication ofthe Reactor Pressure Vessel in order to confirm the applicability of MRP-227, as approved by theNRC, to their facility. If the upper flange weld has not been stress relieved, then this componentshall be inspected as a "Primary" inspection category component. If necessary, the examinationmethods and frequency for non-stress relieved B& W core support structure upper flange weldsshall be consistent with the recommendations in MRP-227, as approved by the NRC, for theWestinghouse and CE upper core support barrel welds. The examination coverage for this B& W

flange weld shall conform to the staff's imposed criteria as described in Sections 3.3.1 and 4.3.1of this SE. The applicant 's/licensee's resolution of this plant-specific action item shall besubmitted to the NRC for review and approval. This is Applicant/Licensee Action Item 4" [5].


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 UnitI.

6.2.5 SE Applicant/Licensee Action Item 5: Application of Physical Measurements aspart of I&E Guidelines for B&W, CE, and Westinghouse RVI Components

"As addressed in Section 3.3.5 in this SE, applicants/licensees shall identify plant-specificacceptance criteria to be applied when performing the physical measurements required by theNRC-approved version ofMRP-227 for loss of compressibility for Westinghouse hold downsprings, and for distortion in the gap between the top and bottom core shroud segments in CEunits with core barrel shrouds assembled in two vertical sections. The applicant/licensee shallinclude its proposed acceptance criteria and an explanation of how the proposed acceptancecriteria are consistent with the plants' licensing basis and the need to maintain the functionalityof the component being inspected under all licensing basis conditions of operation during theperiod of extended operation as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 5" [5].




See Table 7-1. BV Unit 1 utilizes a Type 304 SS hold down spring; therefore, FENOC is planning toperform 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 physicalmeasurements that will be consistent with the licensing basis for BV Unit 1 [5].

Conclusion


6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&WComponents

"As addressed in Section 3.3.6 in this SE, MRP-227 does not propose to inspect the followinginaccessible components: the B& W core barrel cylinders (including vertical and circumferentialseam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices,B& W core barrel-to-former bolts and their locking devices, and B& W core barrel assemblyinternal baffle-to-baffle bolts. The MRP also identified that although the B& W core barrelassembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable usingcurrently available examination techniques.

Applicants/licensees shall justify the acceptability of these components for continued operationthrough the period of extended operation by performing an evaluation, or by proposing ascheduled replacement of the components. As part of their application to implement the approvedversion of MRP-227, applicants/licensees shall provide theirjustification for the continuedoperability of each of the inaccessible components and, if necessary, provide their plan for thereplacement of the components for NRC review and approval. This is Applicant/Licensee ActionItem 6" [5].


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 Unit1.

6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS Materials

"As discussed in Section 3.3. 7 of this SE, the applicants/licensees of B& W, CE, andWestinghouse reactors are required to develop plant-specific analyses to be applied for theirfacilities to demonstrate that B& W IMI guide tube assembly spiders and CRGT spacer castings,


WESTfNGHOUSE NON-PROPRIETARY CLASS 3 6-9

CE lower support columns, and Westinghouse lower support column bodies will maintain theirfunctionality during the period of extended operation or for additional RVI components that may

be fabricatedfrom CASS, martensitic stainless steel or precipitation hardened stainless steelmaterials. These analyses shall also consider the possible loss offracture toughness in these

components due to thermal and irradiation embrittlement, and may also need to considerlimitations on accessibility for inspection and the resolution/sensitivity of the inspectiontechniques. The requirement may not apply to components that were previously evaluated as notrequiring aging management during development of MRP-227. That is, the requirement would

apply to components fabricated from susceptible materials for which an individual licensee has

determined aging management is required, for example during their review performed inaccordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent

with the plant's licensing basis and the need to maintain the functionality of the components

being evaluated under all licensing basis conditions of operation. The applicant/licensee shallinclude the plant-specific analysis as part of their submittal to apply the approved version ofMRP-227. This is Applicant/Licensee Action Item 7" [5].


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 orirradiation embrittlement, or the synergistic effects of thermal embrittlement and irradiation

embrittlement combined, then no other evaluation would be necessary. For assessment of CASSmaterials, the licensee or applicant for license renewal may apply the criteria in the NRC letter ofMay 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of CastA ustenitic Stainless Steel Components" (NRC ADAMS Accession No. ML003 717179) as the basis

for determining whether the CASS materials are 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 basesfor upper support columns, and CASS lower support columns, are not susceptible to TE.

Conclusive confirmation of material composition under TE susceptibility thresholds was notdemonstrated for the CASS stand-alone mixers, the supports, gussets, and clamps on the upperinstrumentation columns, the intermediate flanges in the control rod guide tubes, nor for the bottommounted instrumentation (BMI) cruciforms; thus, it is conservatively assumed that they are potentiallysusceptible to TE. The susceptibility of the mixers, intermediate flanges, and BMI cruciforms to TE wasconsidered in the development of MRP-227-A [5]. The BV Unit I supports, gussets, and clamps on theupper instrumentation columns are CASS. However, in MRP-191, the upper instrumentation conduit andsupports, gussets, and clamps were screened as wrought material (304 SS). These CASS parts wereevaluated under the guidelines of the MRP-191 FMECA in support of Applicant/Licensee Action Items 1and 2.



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

No martensitic stainless steel, or martensitic precipitation hardened stainless steel materials wereidentified 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. Theresults of this CASS evaluation do not conflict with the MRP-227-A strategy for aging management ofRVIs. It is concluded that continued application of the strategy of MRP-227-A will meet the requirementfor 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 TEMolybdenum (Based on the NRC

CASS Component Content Casting Ferrite Content Criteria 1331)Control rod guide tube Low 0.5 max Static >20% Potentially susceptible tointermediate flanges (1) TE (2)

Upper instrumentation Low 0.5 max Static >20% Potentially susceptible tosupports, brackets and TE (2)

clamps

Flow mixer devices, Low 0.5 max Static >20% Potentially susceptible towith and without TE (2)

thermocouple

Upper support column, Low 0.5 max Static <20% Not susceptible to TEflow mixer base

Upper support column Low 0.5 max Static <20% Not susceptible to TEBases

Lower support column Low 0.5 max Static _<20% Not susceptible to TEbodies

Bottom-mounted Low 0.5 max Static >20% Potentially susceptible toinstrumentation (BMI), TE (2)

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.



6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff Reviewand Approval

"As addressed in Section 3.5.1 in this SE, applicants/licensees shall make a submittal for NRCreview and approval to credit their implementation of MRP-227, as amended by this SE, as anAMP for the RVI components at their facility. This submittal shall include the informationidentified in Section 3.5.1 of this SE. This is Applicant/Licensee Action Item 8" [5].


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

Conclusion




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 informationcontained in Table 7-1 is based on this information and includes a description of the past inspections, as well as the latest scope of inspectionspertaining to the reactor internals AMP. Should a change occur in plant operational practices or operating experience result in changes to theprojections, 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 EstimatedOutage 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


24 Fall 2016 29.5 ASME Code Section XI ASME Code Section XI Extended period of operation begins1 0-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 extendedinspections for control rod accordance with MRP-228 operation during Cycle 24. BV hasguide tube lower flange welds, specifications the option to perform theseupper and lower core barrel inspections until RO-25. Theflange welds, upper and lower inspection window for thesecore barrel cylinder girth components is plus or minus twowelds, and thermal shield refueling cycles from the beginningflexures 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 holdinspections for hold down accordance with MRP-228 down spring is plus or minus threespring completed during or specifications refueling cycles from the beginningbefore this outage of extended operation.



Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary(cont.)

Refueling Project EstimatedOutage 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 35bolts completed during or specifications EFPY. FENOC has the option tobefore this outage perform these inspections until RO-

28.



30 Fall 2025 37.9 ASME Code Section XI ASME Code Section XI The inspection window for 17x1710-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 fourEFPY can be applied to the initialinspection measurement schedule.Therefore, the initial inspectionmust be performed before BeaverValley Unit 1 reaches 38 EFPY. SeeWCAP- 17451-P [26] for additionalinformation 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 theseinspections for baffle-edge accordance with MRP-228 components is between 20 and 40bolts and baffle-former specifications EFPYassembly completed during orbefore this outage

WCAP 1 789-P Jauary201



Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary(cont.)

Refueling Project EstimatedOutage 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 theseaugmented inspections for accordance with MRP-228 components is 10 years after thecontrol rod guide tube lower specifications initial inspection.flange welds, upper and lowercore barrel flange welds, upperand lower core barrel cylindergirth welds, and thermal shieldflexures completed during orbefore this outage


34 Fall 2031 43.5 Subsequent MRP-227-A MRP-227-A inspections in The inspection window for theseaugmented inspections for accordance with MRP-228 components is 10 years after thebaffle-former bolts completed specifications initial inspection.during or before this outage




N/A 48. Not applicable Not applicable Renewed Operating License expiresJanuary 29, 2036

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



8 IMPLEMENTING DOCUMENTS

As noted within this AMP document, the BV Unit 1 PWR Vessel Internals Program is documented inNOP-CC-5004 [1]. The BV Unit 1 AMP also references the Primary Water Chemistry Program and theASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD Program. MRP-227-Aaugmented examinations (Appendix C) recommended as a result of industry programs will be included inthe existing ASME Section XI program.

FENOC documents associated with the existing Beaver Valley programs and considered to beimplementing 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 toreduce susceptibility to cracking due to SCC. Additional procedures may be updated or created as OE foraugmented examinations is accumulated.

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



9 REFERENCES

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

2. U.S. Nuclear Regulatory Commission, NUREG-1929, "Safety Evaluation Report Related to theLicense 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 Summary2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals AgingManagement," July 21, 2011.

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

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

6. U.S. Nuclear Regulatory Commission, Code of Federal Regulations, 10 CFR Part 54, "Requirementsfor 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 forthe 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 Managementfor Reactor Internals," March 2001.

9. Materials Reliability Program: Screening, Categorization and Ranking of Reactor InternalsComponentsfor Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, PaloAlto, CA: 2006. 1013234.

10. Materials Reliability Program. Inspection Standardfor PWR Internals - 2012 Update (MRP-228,Rev 1). EPRI, Palo Alto, CA: 2012. 1025147.

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

12. Beaver Valley Business Practice, BVBP-LRP-0003, Rev. 7, "Mechanical Screening and AgingManagement 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 ProgramManagement," July 11, 2013.

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

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



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

18. Beaver Valley Program Manual, BVPM-CHEM-0001, Revision 0, "Primary Systems Strategic WaterChemistry 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 (NRCADAMS 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 - WestinghouseDomestic Fleet Operational Projections," October 2013.

27. Beaver Valley Nuclear Operating Procedure, NOP-LP-2001, Revision 32, "Corrective ActionProgram," 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 inPressurized Water Reactor Systems," March 7, 1984.

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

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

32. Westinghouse Letter, PCWG-07-46, Rev. 0, "Beaver Valley Units I & 2 (DLW/DMW): Approval ofCategory 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, Thermal AgingEmbrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000 (NRC ADAMSAccession No. ML003717179).


WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-IWESTINGHOUSE NON-PROPRIETARY CLASS 3 A-i

APPENDIX AILLUSTRATIONS

Figure A-1 Illustration of Typical Westinghouse Internals Assembly


WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2

Figure A-2 Typical Westinghouse Control Rod Guide Card


WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3

UpperGuide Tube

LowerGuide tube

Sheaths andC-Tubes

Upper SupPlate

IC

lport

Figure A-3 Lower Section of Control Rod Guide Tube Assembly



Flange Weld

Upper Core Barrel toLower Core BarrelCircumferential Weld

Lower BarrelCircumferential Weld

Core Barrel toSupport Plate Weld

- Axial Weld

Lower Barrel-, Axlal Weld

Lower BarrelAxial Weld

Thermal ShieldFlexure

Figure A-4 Major Core Barrel Welds



@0000

00000

000800

4100000

00000

40

J)

(D0

cu

Figure A-5 Bolting Systems used in Westinghouse Core Baffles



INTERNALSSUPPORT LEDGE-

THERMAL SHIELD-

BAFFLE -

FORMER-

LOWERCORE PLATE

DIFFUSER PLATE

CORE SUPPORT

COLUMN

CORE SUPPORTFORGING

Figure A-6 Core Baffle/Barrel Structure



BAFFLETO FOAMERDOLTLO4G &smIRi)

COMNER EDGE BRACKETBAFFLE TO FORMER BMLT

Figure A-7 Bolting in a Typical Westinghouse Baffle-Former Structure



Figure A-8 Vertical Displacement between the Baffle Plates and Bracket at the Bottom of theBaffle-Former-Barrel Assembly



TOP SUPPORT PLATE

Figure A-9 Schematic Cross-Sections of the Westinghouse Hold Down Springs

W Id

Figure A-10 Typical Thermal Shield Flexure


WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-10WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-b

Lower Core Plate

Lower CoreSupport Structure

Core SupportPlate (Forging)

Figure A-II Lower Core Support Structure


WESTfNGHOUSE NON-PROPRIETARY CLASS 3 A-I IWESTINGHOUSE NON-PROPRIETARY CLASS 3 A-I I

T / T / , TI/LOWER CORE PLATE

K, ..... ... wz! z ... 2= 6VM~z ww2MDIFFUSER PLATE

CORE SUPPORTPLATE/FORGING

CORE

SUPPORT

COLUMN BOTTOM MOUNTEDINSTRUMENTATIONCOLUMN

Figure A-12 Lower Core Support Structure - Core Support Plate Cross-Section

Figure A-13 Typical Core Support Column



Fu A

Figure A-14 Examples of BMI Column Designs


WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-1

APPENDIX BBEAVER 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 thelicense 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 ManagementComponent 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)



Table B-I Beaver Valley Unit I LRA Aging Management Review Summary(cont.)

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

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

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

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

17. Core barrel (shell, ring, Cumulative fatigue TLAAflange, nozzle, thermal damageshield/pad)

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

19. Core barrel (shell, ring, Loss of material Water Chemistry (B.2.42)flange, nozzle, thermalshield/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 TLAAdamage

24. Core barrel assembly (bolt) Loss of fracture PWR Vessel Internalstoughness (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 Internalsstructure (flux thimble guide (B.2.33)tube)

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



Table B-i Beaver Valley Unit I LRA Aging Management Review Summary(cont.)

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

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

30. Instrumentation support Cumulative fatigue TLAAstructure (flux thimble guide damagetube)

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

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

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

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

35. Instrumentation support Cumulative fatigue TLAAstructure (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)





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 lowersupport column)

53. Lower internals assembly Cumulative fatigue TLAA(Core support forging and lower damagesupport 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 lowersupport 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



Table B-1 Beaver Valley Unit I LRA Aging Management Review Summary(cont.)


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)


WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-6WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-6



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

79. Lower internals assembly Loss of material ASME Section XI(radial key) Inservice Inspection,


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, headcooling spray nozzle)

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

83. Lower internals assembly Cracking Water Chemistry (B.2.42)(secondary core support,head/vessel alignment pin, headcooling spray nozzle)

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

85. Lower internals assembly Loss of material Water Chemistry (B.2.42)(secondary core support,head/vessel alignment pin, headcooling 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)



Table B-1 Beaver Valley Unit 1 LRA Aging Management Review Summary(cont.)


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 columncasting)

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

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 columncasting)

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)





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,


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




Aging Effect Requiring Aging ManagementComponent 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)




Aging Effect Requiring Aging ManagementComponent 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 supportplate and support assembly)

144. Upper internals assembly Cumulative fatigue TLAA(upper core plate, upper support damageplate and support assembly)

145. Upper internals assembly Loss of material Water Chemistry (13.2.42)(upper core plate, upper supportplate 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 "ThermalAging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) (B.2.40)". This has beenrevised to reference the "PWR Vessel Internals" program according to BV internal documentation [21 ].


WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-1WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-I

APPENDIX CMRP-227-A AUGMENTED INSPECTIONS

Table C-1 MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals

Effect Expansion Link ExaminationItem 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-3forging/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-4and subsequent examinationon a ten-year interval.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-2WESTINGHOUSE 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-4and subsequent examinationon 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 adjacentoutages from the beginning base metal (Note 4).of the license renewal periodand subsequentexaminations on a ten-yearinterval.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-3


Aibil Effect Expansion Link ExaminationItem Applcability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage

Baffle-Former All plants Cracking None Visual (VT-3) examination, Bolts and locking devicesAssembly with baffle- (IASCC, with baseline examination on high-fluence seams.

Baffle-edge bolts edge bolts Fatigue) that between 20 and 40 EFPY 100% of componentsresults 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, anddevices A-7

" Failed ormissing bolts

" Protrusion ofbolt heads

AgingManagement(IE and ISR)(Note 6)

Baffle-Former All plants Cracking Lower support Baseline volumetric (UT) 100% of accessible boltsAssembly (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


WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-4WESTINGHOUSE NON-PROPRIETARY CLASS 3 CA



Baffle-Former All plants Distortion (Void None Visual (VT-3) examination Core side surface, asAssembly 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 fuelassemblies

" Gaps alonghigh fluencebaffle joint

" Verticaldisplacementof baffle platesnear highfluence joint

" Broken ordamaged edgebolt lockingsystems alonghigh fluencebaffle joints





Alignment and All plants Distortion (Loss None Direct measurement of Measurements should beInterfacing with 304 of Load) spring height within three taken at several pointsComponents 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 adequatesprings 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 60years.



Table C-1 MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals(cont.)


Thermal Shield All plants Cracking None Visual (VT-3) no later than 2 100% of thermal shieldAssembly 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. Subsequentresults in examinations on a ten-yearthermal shield interval.flexuresexcessive wear,fracture, orcompleteseparation

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 forinspection 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 fromeither 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].



Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals

Item Applicability Effect Primary Link ExaminationItem____Applicability____ (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage

Upper Internals All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessibleAssembly (Fatigue, Wear) flange weld examination, surfaces (Note 2).

Upper Core Plate Re-inspection every 10 yearsfollowing initial inspection.

Lower Internals All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessibleAssembly 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.

supportforging

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

Barrel-former bolts (IASCC, bolts examination. Accessibility may beFatigue) 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-7Swelling andISR)

Lower Support All plants Cracking Baffle-former Volumetric (UT) 100% of accessible boltsAssembly (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


WESTfNGHOUSE NON-PROPRIETARY CLASS 3 C-8WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-8

Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals(cont.)

Effect Primary Link ExaminationItem Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) 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-4sections)

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 accessibleAssembly 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 accessibleAssembly (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

AgingManagement(IE)


WESTI`NGHOUSE NON-PROPRIETARY CLASS 3 C-9WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-9

Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals(cont.)

Item

Bottom MountedInstrumentation System

Bottom-mountedinstrumentation (BMI)column bodies

Examinationd/Frequency (Note 1)

Examination Coverage

Cracking(Fatigue)including thedetection ofcompletelyfracturedcolumn bodies

Control rod guidetube (CRGT)lower flanges

Visual (VT-3) examinationof BMI column bodies asindicated by difficulty ofinsertion/withdrawal of fluxthimbles.

Re-inspection every 10 yearsfollowing initial inspection.

Flux thimbleinsertion/withdrawal to bemonitored at each inspectioninterval.

100% of BMI columnbodies for which difficultyis detected during fluxthimbleinsertion/withdrawal.

See Figures A-12 and A-14

AgingManagement(LE)

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 examinationis required (including both the accessible and inaccessible portions).

WCAP- I 7789-NP January 2014Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-10WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-i 0

Table C-3 MRP-227-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-DesignedInternals

EffectItem 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 excessivewear.

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

Upper support ring orskirt

Lower Internals All plants Cracking ASME Code Visual (VT-3) examination All accessible surfaces atAssembly (IASCC, Section XI of the lower core plates to specified frequency.

Lower core plate Fatigue) detect evidence of distortionXL 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 atAssembly (Wear) Section XI specified frequency.

Lower core plateXL lower core plate(Note 1)

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

09.

WCAP- 17789-NP Januaiy 2014



Table C-3 MIRP-227-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-DesignedInternals(cont.)

EffectItem Applicability (Mechanism) Reference Examination Method Examination Coverage

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

Upper core platealignment 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.



Table C-4 MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals

Examination Additional ExaminationItem 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/ATube Assembly Examination

Guide plates (cards) The specificrelevant conditionis wear that couldlead to loss ofcontrol rodalignment andimpede controlassembly insertion.



Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals(cont.)

Examination Additional ExaminationItem Applicability Acceptance Criteria Expansion Link (s) Expansion Criteria Additinal Eaitio


Control Rod Guide All plants Enhanced visual a. Bottom- a. Confirmation of a. For BMI columnTube 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 supportindication. (cast), upper core difficulty, shall require column bodies, upper

plate and lower visual (VT-3) examination core plate and lowersupport forging or of BMI column bodies by support forging/castings,casting the completion of the next the specific relevant

refueling outage. condition is a detectableb. Confirmation of crack-like surfacesurface-breaking indication.indications in two or moreCRGT lower flange welds

shall require EVT-1examination of cast lowersupport column bodies,upper core plate and lowersupport forging/castingswithin three fuel cyclesfollowing the initialobservation.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-14WESTINGHOUSE NON-PROPRIETARY CLASS 3 C- 14



(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 acrack-like surface examination be expanded detectable crack-likeindicationk to include the core outlet surface indication.

nozzle welds by the

completion of the nextrefueling outage.

b. If extensive cracking inthe core barrel outletnozzle welds is detected,EVT-1 examination shallbe expanded to include theupper six inches of theaccessible surfaces of thenon-cast lower supportcolumn bodies withinthree fuel cycles followthe initial observation.






Core Barrel Assembly All plants Periodic enhanced None None None

Lower core barrel flange visual (EVT-1)

weld (Note 2) examination.The specificrelevant conditionis a detectablecrack-like surfaceindication.

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 beexpanded to include theupper core barrel cylinderaxial welds by thecompletion of the nextrefueling outage.




Examination Additional ExaminationItem Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Additinal Eaitio


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 axialrelevant condition inches in the lower core weld examination is ais a detectable barrel cylinder girth welds detectable crack-like

crack-like surface shall require that the EVT- surface indication.

indication. 1 examination beexpanded to include thelower core barrel cylinderaxial welds by thecompletion of the nextrefueling outage.

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

Baffle-edge bolts edge bolts The specificrelevant conditionsare missing orbroken lockingdevices, failed ormissing bolts, andprotrusion of boltheads.



Table C4 MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals(cont.)


(NoteI) A1)eptance__riteria

Baffle-Former All plants Volumetric (UT) a. Lower support a. Confirmation that more a and b. TheAssembly 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 lowersupport column boltswithin the next three fuelcycles.

b. Confirmation that morethan 5% of the lowersupport column boltsactually examined containunacceptable indicationsshall require UTexamination of the barrel-former bolts.




Examination ExamiationAdditional ExaminationItem Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Additinal Eaitio


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

Assembly

The specificrelevant conditionsare evidence ofabnormalinteraction withfuel assemblies,gaps along highfluence shroudplate joints, verticaldisplacement ofshroud plates nearhigh fluence joints,and broken ordamaged edge boltlocking systemsalong high fluencebaffle plate joints.





(NoteI) Acceptance)Criteria

Alignment and All plants Direct physical None N/A N/AInterfacing with 304 measurement orComponents stainless spring height.

Internals hold down steel hold The examinationspring down acceptance

springs criterion for thisNOTE: measurement is

BV Unit 1 that the remaining

hold down compressiblespring is height of the spring

304 SS shall provide hold-down forces withinthe plant-specificdesign tolerance.

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

Thermal shield flexures shields The specificrelevant conditionsfor thermal shieldflexures areexcessive wear,fracture, orcompleteseparation.

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.


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