wcap-17077-np, rev 1, 'pwr vessel internals program plan for … · 2012-10-15 ·...

Enclosure to Serial: RNP-RA/12-0100103 Pages (including cover page)

ENCLOSUREPWR Vessel Internals Program Plan for Aging

Management of Reactor Internals atRobinson Nuclear Plant

Westinghouse Non-Proprietary Class 3

WCAP-17077-NPRevision 1

PWR Vessel InternalsProgram Plan for AgingManagement ofReactor Internals atRobinson Nuclear Plant

Westinghouse

August 2012

WESTINGHOUSE NON-PROPRIETARY CLASS 3

WCAP-17077-NPRevision 1

PWR Vessel Internals Program Plan for Aging Managementof Reactor Internals at Robinson Nuclear Plant

Karli N. Szweda*Reactor Internals Aging Management

Joshua K. McKinley*Materials Center of Excellence I

Richard A. Basel*Reactor Internals Aging Management

Cheryl L. Boggess*Reactor Internals Aging Management

August 2012

Approved: Patricia C. Paesano*, ManagerReactor Internals Aging Management

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

Westinghouse Electric Company LLC1000 Westinghouse Drive

Cranberry Township, PA 16066

© 2012 Westinghouse Electric Company LLCAll Rights Reserved

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii

TABLE OF CONTENTS

L IST O F T A B L E S ........................................................................................................................................ v

L IST O F FIG U R E S .................................................................................................................................... vii

L IST O F A C R O N Y M S ................................................................................................................................ ix

ACKNOWLEDGEMENTS ......................................................................................................................... xi

P U R P O S E ..................................................................................................................................... 1-1

2 B A C K G R O U N D .......................................................................................................................... 2-1

3 PRO G RA M O W N E R ................................................................................................................... 3-13.1 Corporate Nuclear Engineering Services Chief Engineering Section .............................. 3-13.2 R N P Engineering ............................................................................................................. 3-13.3 RNP Environmental and Chemistry ................................................................................. 3-1

4 DESCRIPTION OF THE H. B. ROBINSON REACTOR INTERNALS AGINGMANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS ............................................. 4-14.1 Existing RN P Program s ................................................................................................... 4-3

4.1.1 Primary Water Chemistry Program ................................................................. 4-44.1.2 ASME Section XI Inservice Inspection Program ............................................ 4-44.1.3 Flux Thimble Eddy Current Program .............................................................. 4-4

4.2 Supporting RNP Programs and Aging Management Supportive Plant Enhancements ... 4-54.2.1 Reactor Internals Aging Management Review Process ................................... 4-54.2.2 Flux Thim ble Tubes ........................................................................................ 4-54.2.3 Reactor Head and Control Rod Drive Mechanism Replacement .................... 4-64.2.4 Control Rod Guide Tube Split Pin Replacement Project ................................ 4-6

4.3 Industry Program s ............................................................................................................ 4-64.3.1 WCAP-14577, Aging Management for Reactor Internals ............................... 4-64.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines ........... 4-74.3.3 Ongoing Industry Programs ........................................................................... 4-10

4.4 Sum m ary ........................................................................................................................ 4-11

5 H.B. ROBINSON REACTOR INTERNALS AGING MANAGEMENT PROGRAMA T T R IB U T E S .............................................................................................................................. 5-15.1 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 ...................................................... 5-115.7 GALL Revision 2 Element 7: Corrective Actions ........................................................ 5-12

WCAP- 17077-NP August 2012Revision I

iv WESTINGHOUSE NON-PROPRIETARY CLASS

5.8 GALL Revision 2 Element 8:, Confirmation Process .................................................... 5-135.9 GALL Revision 2 Element 9: Administrative Controls ................................................ 5-145.10 GALL Revision 2 Element 10; Operating Experience ................................................. 5-14

6 DEMONSTRATION ..... '...... r .......................... ........................ :6-1

6.1 Demonstration of Topical Report Conditions Compliance to SE onMRP-227, Revision 0 ...................... v ...................... 6-2

6.2 Demonstration of Applicant/Licensee Action Item Compliance to SE on MRP-227,R ev ision 0 ........................................ ........................................................................ ... 6-36.2.1 SE Applicant/Licensee Action Item 1: ADplicability of FMECA and

Functionality Analysis Assumptions ............................................................... 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

Flange Stress R elief ......................................................................................... 6-66.2.5 SE Applicant/Licensee Action Item 5: Application of Physical Measurements

as part of I&E Guidelines for B&W, CE, and Westingliouse RVI Components......................................................................................................................... 6 -7

6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&WC om ponents ..................................................................................................... 6-7

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

6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for StaffReview and A pproval ...................................................................................... 6-9

7 PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE ............................... 7-1

8 IM PLEM ENTING DOCUM ENTS .............................................................................................. 8-1

9 R E FE R EN C E S ............................................................................................................................. 9-1

APPENDIX A ILLUSTRA TIONS ......................................................................................... A-I

APPENDIX B ROBINSON LICENSE RENEWAL AGING MANAGEMENT REVIEWSUM M A RY TA BLES ................................................................................... B-I

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

WCAP- I 7077-NP August 2012Revision I

.WESTINGHOUSE NON-PROMETARY CLASS 3 v

.. LIST OF TABLES

Table 6-1 Topical Report Conditions Compliance to SE on' MRP-227 ........................................ 6-2

Table 7-1 Aging Management Program Enhancement and Inspection ImplementationS um m ary ......................................................................................................................... 7-1

Table B-1 LRA Aging Management Review Summary Table 3.1'1 Robinson LRA ......... B-1

Table B-2 LRA Aging Management Review Summary Table 3.1-2 Robinson LRA ..................... B-2

Table C-I MRP-227-A Primary Inspection' and Monitoring Recommendations forW estinghouse-Designed Internals ..................................... .............................. C-1

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

,Westinghouse-Designed Internals .................... ..................................... C-6

Table C-3 MRP-227-A Existing Inspection and Aging Management Programs Credited inRecommendations for Westinghouse-DeSigned Internals". .............................................. C-9

Table. C4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations forW estinghouse-D esigned Internals ........................................................................... C-11

• . .,..z

I

.J "'• .9 .

WCAP-1 7077-NP August 2012Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 vii

LIST OF FIGURES

Figure A-i Illustration of a Typical Westinghouse Internals ............................................................... A-1

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 Core Barrel W elds ...................................................................................................... A -4

Figure A-5 Bolting Systems used in Westinghouse Core Baffles ........................................................... A-5

Figure A-6 Core Baffle/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-Barrel A ssem bly ................................................................................................ A -8

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

Figure A-10 Typical Thermal Shield Flexure .......................................................................................... A-9

Figure A-I I Lower Core Support Structure ..................................................................................... A-10

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

Figure A- 13 Typical Core Support Column .......................................................................................... A-I l

Figure A-14 Examples of BMI Column Designs ................................................................................... A-12


3.; --

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ix

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 instrumentationBWR boiling water reactorCASS cast austenitic stainless steelCE Combustion EngineeringCES Corporate Chief Engineering SectionCFR Code of Federal RegulationsCLB current licensing basisCRGT control rod guide tubeEFPY effective full-power yearsEPRI Electric Power Research InstituteET electromagnetic testing (eddy current)EVT enhanced visual testing (a visual NDE method that includes EVT-1)FMECA failure mode, effects, and criticality analysisGALL Generic Aging Lessons LearnedI&E Inspection and EvaluationIASCC irradiation-assisted stress corrosion crackingIGSCC intergranular stress corrosion crackingINPO Institute of Nuclear Power OperationsISI inservice inspectionISR irradiation-enhanced stress relaxationLRA License Renewal ApplicationMRP Materials Reliability ProgramNDE nondestructive examinationNEI Nuclear Energy InstituteNES Nuclear Engineering & ServicesNOS Nuclear Oversight SectionNRC U.S. Nuclear Regulatory CommissionNSSS nuclear steam supply systemOBE operating basis earthquakeOE 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 crackingQA quality assuranceRCS reactor coolant systemRI-FG Reactor Internals Focus Group


X WESTINGHOUSE. NON-PROPRIETARY CLASS 3.

LIST OF ACRONYMS-(cont.)

RIS Regulatory Issue SummaryRO refueling outage .

RNP H. B. Robinson Steam Electric Plant, Unit Number 2, or Robinson Nuclear PlantRV reactor vesselRVI reactor vessel internalsSCC stress corrosion crackingSER Safety Evaluation ReportSRP Standard Review PlanSS stainless steelTE thermal embrittlementUT ultrasonic testing (a volumetric NDE method)VT visual testing (a visual NDE method that includes VT-I and VT-3)WOG Westinghouse Owners GroupXL Extra-long Westinghouse Fuel

Trademark Statement:

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

WCAP-17077-NP August 2012Revision I

. WESTINGHOUSE NON-PROPRIETARY CLASS 3 xi

ACKNOWLEDGEMENTS

The authors would like to thank the members of the Progres's Enetgy Aging Management Program Teamled by David Martrano and Anthony James and our associates at Westinghouse, including Dr. RandyLott, for their efforts in supporting development of this WCAP.


- rff

'1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 1-1

1 PURPOSE

The purpose of this report is to document the H.B. Robinson Steam Electric Plant, Unit Number 2,hereafter Robinson Nuclear Plant (RNP), Reactor Vessel Internals (RVI) Aging Management ProgramPlan (AMP). The purpose of the AMP is to manage the effects of aging on reactor vessel internalsthrough the license renewal period. RNP entered the license renewal period on August 1, 2010. Thisdocument provides a description of the program as it relates to the management of aging effects identifiedin various regulatory and updated industry-generated documents in addition to the program documentedin RNP calculation RNP-L/LR-0614 [8] in support of license renewal program evaluations. This AMP issupported by existing RNP documents and procedures and, as needed by industry experience or directivein the future, will be updated or supported by additional documents to provide clear and concise directionfor the effective management of aging degradation in reactor internals components. These actionsprovide assurance that operations at RNP will continue to be conducted in accordance with the currentlicensing basis (CLB) for the reactor vessel internals by fulfilling License Renewal commitments [2],American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code SectionXI Inservice Inspection (ISI) programs [17], and industry requirements [I I]. This AMP fully captures theintent of the additional industry guidance for reactor internals augmented inspections, based on theprograms sponsored by U.S. utilities through the Electric Power Research Institute (EPRI) managedMaterials Reliability Program (MRP) and the Pressurized Water Reactor Owners Group (PWROG).

The main objectives for the RNP RVI AMP are to:

* Demonstrate that the effects of aging on the RVI will be adequately managed for the period ofextended operation in accordance with 10 CFR 54 [1].

* Summarize the role of existing RNP 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 RNP reactor internals.

RNP License Renewal Commitment 33 [2], "Pressurized Water Reactor Vessel Internals Program,"commits RNP to:

I. Participate in industry programs to investigate aging effects and determine the appropriate AMPactivities to address baffle and former assembly issues, and to address changes in dimensions dueto void swelling.

2. Evaluate the results of completed research projects from the Westinghouse Owners Group(formerly WOG, now PWROG) and the EPRI MRP, and factor them into the PWR VesselInternals Program as appropriate.


1-2 WESTINGHOUSE NON-- PROPRIETARY CLASS 3

3. Implement an augmented inspection during the license renewal term.

Augmented inspections, based on required program enhancements resulting from industry programs, willbecome part of the RNP ASME B&PV Code, Section XI program [.17]. Corrective actions for augmentedinspections willbe deyeloped using repair and replacement procedures equivalent to those requirementsin ASME B&PV Code, Section XI, or as determined independently by Progress Energy, or in cooperationwith the industry, to be equivalent or more rigorous than currently defined procedures.

This AMP for the RNP reactor internals demonstrates that the program adequately manages the effects ofaging for reactor internals components and establishes the basis for providing reasonable assurance thatthe internals components will continue to perform their intended function through the RNP licenserenewal period of extended operation. Revision 0 of this WCAP supported the RNP license renewalcommitment to submit to the U.S. Nuclear Regulatory Commission (NRC) an inspection plan for thePWR Vessel .Internals Program, as 'it would be implemented, from RNP's participation in industryinitiatives, 24 months priorto the augmented inspection. Since MRP-227-A was released, the NRCpubl.ished.aRegulatory Issue Summary (RIS) that implemented new guidelines as to when plants mustsubmit an inspection plan. According to the NRC RIS [44], because; RNP has submitted their AMP forapproval by the NRC, RNP may withdraw their submittal and provide new and revised commitments tothe NRC to resubmit their AMP in accordance with MIRP-227-A [ 11] no later than October 1, 2012.Therefore, Revision 1 of this AMP incorporates the changes from MRP-227-A and the RIS.

The development and implementation of this program meets the guidelines provided, in the RIS [44].

WCAP- I 7077-NP August 2012Revision 1

WESTINGHOUSE' NON-PROPRIETARY CLASS 3 2-1

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 [3]. The U.S.nuclear power industry has been actively engaged in recentyears in a significant effort to support theindustry goal of responding to these requirements' Various programs have been underwiy witfiin theindustry over the past decade to develop guidelines for managing the effects of aging within PWR reactor'internals. In 1997, the WOG issued WCAP-14577 [26], "License Renewal Evaluation: AgingManagement for Reactor Internals," which was reissued as Revision I-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 [26, 32]:

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 Prcigram).

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, existing evaluations, and prior examination results.


2-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3

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

* MRP-227-A [11], "PWR Internals Inspection and Evaluation Guidelines," (hereafter referred to.as the "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 [27], "Inspection Standard for Reactor Internals Components," provides guidance onthe qualification/demonstration of the NDE techniques and other criteria pertaining to the actual,performance of the inspections..

The PWROG1 has also developed "Reactor Internals Acceptance Criteria Methodology and DataRequirements" for the MRP-227 inspections, where feasible [45]. This document has been submitted tothe NRC for review and approval, and will be.updated to incorporate changes from MRP-227-A [II].Final reports are to be developed and be available for industry use in support of planned license renewalinspection commitments. In some cases, individual plants will develop plant-specific acceptance criteriafor some internals components where a generic approach is not practical.

The RNP 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 provoided in Figure A-1.

As described in NUREG- 1785 [2], the RNP internals are designed to support, align, and guide the corecomponents and to support and guidein-coreinstrumentation. The RVI consist of two basic . , .. •assemblies -.-;an upper internals assembly that is .removed dur. each.refueling operation to obtain naccessto the reactor core, and a lower internals assembly that can be removed, if desired, following. complete..core unload.

The lower internals assembly is supported in the vessel by resting on a ledge in the vessel head-matingsurface and is closely guided at the bottom by radial support/clevis assemblies. The upper internalsassembly is clamped at this same ledge by the reactor vessel head. The bottom of the upper internalsassembly is closely guided by the core barrel alignment pins of the lower internals assembly.

The lower internals comprise the core barrel, thermal shield, core baffle assembly, lower core plate,intermediate diffuser plate, bottom support plate, and supporting structures. The upper internals package(upper core support structure) is a rigid member composed of the top support plate and deep beamsections, support columns, control rod guide tube assemblies, and the upper core plate. Upon upperinternals assembly installation, the last three parts are physically located inside the core barrel.

The in-core instrumentation includes in-core flux guide thimbles to permit the insertion of movabledetectors for measurement of the neutron flux distribution within the reactor core. Movable miniatureneutron flux detectors are available to scan the active length of selected fuel assemblies to provide remotereading of the relative three-dimensional flux distribution. The thimbles are inserted into the reactor corethrough guide tubes, or conduits, extending from the bottom of the reactor vessel (RV) through theconcrete shield area and then up to a thimble seal table. Since the movable detector thimbles are closed at


WESTINGHou'st NON-pROPRIETARY CLASS 3 2m3

the leading (reactor) end, they are dryinside. The thimbles thus serve as a pressure barrier between thereactor coolant pressure and the atmosphere. Mechanical seals between the retractable thimbles and theconduits are provided at the seal table.

RNP was granted a license for extended operation by the NRC through the issuance of a SER inNUREG-1785 [2]. In the SER, the NRC concluded that the RNP License Renewal Application (LRA)adequately identified the RVI system structures and components that aresubject to an AMR, as requiredby 10 CFR 54.21(a)(1) [I ]. A listing of the RNP reactor vessel internals components and subcomponents,already reviewed by the NRC in the SER that are subject to AMP requirements, is included inTables B-I and B-2.

In accordance with 10 CFR Part 54 [1], frequently referred to as the License Renewal Rule, RNP hasdeveloped a procedure to direct the performance of aging management reviews of mechanical structuresand components [9]. The U.S. industry, as noted through the efforts of the MRP and PWROG, hasfurther investigated the components and subcomponents that require aging management to support -

continued reliable function. As designated by the protocols of NEI 03-08 [12], "Guidelines for the,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 [44], because RNP has. submitted their AMP forapproval by the NRC, RNPmay withdraw their submittal and provide new and revised commitments to the NRC to resubmit theirAMP in accordance with MRP-227-A [11] no later than October 1, 2012.

The information contained in this AMP fully complies;withWthe requirements and guidance. of the'...:.referenced documents. The-AMPwill manage agifig effects Of thRVI-so that the intended functionswillbe maintained consistent with the -cufrent' licensing bdis for the: pei'od:of extended operation.

WCAP-1I7077-NP August 2012Revision I


3 PROGRAM OWNER

The PWR Vessel Internals Program is part of the "Reactor Coolant System Material IntegrityManagement Program" [7]. The successful implementation and comprehensive long-term management ofthe RNP RVI AMP will require the integration of Progress Energy organizations, corporately and at RNP,and interaction with multiple industry organizations including, but not limited to, the ASME, MRP, NRC,and PWROG. The responsibilities of the individual Progress Energy corporate and RNP groups areprovided in the following paragraphs. Progress Energy will maintain cognizance of industry activitiesrelated to PWR internals inspection and aging management, and will address/implement industryguidance stemming from those activities, as appropriate under NEI 03-08 practices.

The overall responsibility for administration of the RVI AMP is RNP Senior Management.

Additional responsibilities and the appropriate responsible personnel are discussed in the followingsubsections.

3.1 CORPORATE NUCLEAR ENGINEERING SERVICES CHIEF ENGINEERINGSECTION

The Nuclear Engineering Services (NES) Chief Engineering Section (CES) is responsible for providinggovernance and oversight of the implementation of the PWR Vessel Internals Program, including:

Ensuring that appropriate programs are established and maintained to support inspection andmitigation activities for the reactor internals.

* Providing oversight of plant implementation of inspection and mitigation activities.

Maintaining cognizance of industry activities related to PWR internals inspection and agingmanagement.

3.2 RNP ENGINEERING

RNP Engineering is responsible for the overall development and implementation of the PWR VesselInternals Aging Management Program, including:

Planning, control, and implementation of the RVI AMP mitigation, inspection, and repairactivities, as approved by site senior management.

Review and approval of vendor programs involved in various reactor internals activities.

Ensuring that required inspections and supporting activities are implemented in the timesspecified.

3.3 RNP ENVIRONMENTAL AND CHEMISTRY

RNP Environmental and Chemistry is responsible for:

WCAP- 17077-NP August 2012Revision 1


Maintaining primary water chemistry inlaccordance with: approved RNP procedures andspecifications. 9-.

Participation in industry activities addressing water chemistry issues as they relate to minimizingthe potential initiation and growth of primary water stress corrosion cracking (PWSCC) innickel-baded alloys and intergranularsiress corrosion cracking (IGSCC) in austenitic stainlesssteel. components. "

2'.


WESTINdHOUSENON-PROPRIETARY CLASS 3 ._47i

4 DESCRIPTION OF THE H. B. ROBINSON REACTOR INTERNALSAGING MANAGEMENT PROGRAMS AND INDUSTRYPROGRAMS

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. As noted in Section 3 of this AMP, the PWRVessel Internals Program is a part of the "Progress Energy Reactor Coolant System Material IntegrityManagement Program" [7]. The intent of this program is to ensure the long-term integrity and safeoperation of the reactor internals components. RNP has developed this AMP in conformance with the 10Generic Aging Lessons Learned (GALL) [43] attributes and MRP-227-A.

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 [ 15, 19], inspectionsprescribed by the ASME Section XI Inservice Inspection Program [17], thimble tube inspections [18], andmitigation projects such as split pin replacement [25] and [46], combined with augmented inspections orevaluations as recommended by MRP-227-A.

Aging degradation mechanisms that impact internals have been identified and documented in RNP AgingManagement Reviews [5, 6] prepared using the corporate procedural guidance document [9] in support ofthe License Renewal effort. The overall outcome of the reviews and the additional work performed bythe industry, 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 RNP 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.


4-.2 WESTINGHOUSE NON-PROPRIETARY CLASS 3

Fatigue -

Fatigue is defined as the structural deterioration that can.occur as the result of repeatedstress/strain cycles causedby 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 orthermal 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 crackeyentually 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) and. precipitation-hardenable (PH) stainless steel to high inservice temperatures, whichcan result in an increase in tensile, strength,, a decrease in ductility, and a loss of fracturetoughness. Some degree of thermal aging embrittlement can also occur at no .rma! operatingtemperatures for CASS and PH stainless steel internals. CASS components have a duplexmicrostructure and are particularly susceptible to t.his mechanism.. While the initial aging effect is

.loss of ductility and toughness, unstable crack extension is the eyentual aging effect if a crack ispresent and the local applied stress intensity exceeds the reduced fracture toughness.

Irradiation Embrittlement

Irradiation embrittlement is also referred to as neutron embrittlement. When exposed to high-energy neutrons, the mechanical properties of stainless steel and nickel-based alloys can be .

• changed. Such changes in mechanical properties include increasing yield strength, 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.


WESTINGHOUISE N.ON-ROPRIETARY CLASS. 3 .4-3•

* Void Swelling and Irradiation Growth

Void swelling is defined as a gradual increase in the volume of a component caused by formationof microscopic cavities in the material These cavities :result from the nucleation and growth ofclusters of irradiation-produced vacancies. •-lelium 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 ifi-c6re instrumentation tubes that ar6 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 reaich 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 stres's§e-•els below'the ,ield strength (elaýtic limit).Creep occurs at elevated temperatures where: ontinuous defdrmaitiori takes place undeit constantstrain. Secbndary creep iii austenitic stainless stelels is: associated with temlierature• higher thanthose relevant to PWR internals even after itking into account gamma heating. However,irradiation-enhanced creep (or more simply; irradiation .creep) or irradiation-enhanced stress

.rlaxation :(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 unahticipatted loading that, in turn, may eventually causesubsequent degradation by fatigue or wear and result in cracking.

The RNP RVI AMP is focused on meeting the requirements of the 10 elements of an aging managementprogram as described in NUREG-1801, GALL Report Section XI.M16A for PWR Vessel Internals. Inthe RNP RVI AMP, this is demonstrated through application of existing RNP AMR methodology thatcredits inspections prescribed by the ASME Section XIllnservice Inspection Program, existing RNPprograms, and additional augmented inspections based on MRP-227-A recommendations.'A descriptionof the' applicable existing RNP programs and compliance With the elements of the GALL is contained inthe following subsections. .

4.1 EXISTING RNP PROGRAMS

RNP's overall strategy for managing aging in reactor internals components is supported by the followingexisting programs:


4-4 WESTINGHOUSE NON-PROMETARY CLASS 3

0 Primary Water Chemistry Program.0 . ASME SectionXI Inservice Inspection Program* Flux Thimble Eddy Current 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 Primary Water Chemistry Program

The RNP Primary Water Chemistry Program [ 16] is used to mitigate aging effects on component surfacesthat are exposed to water as process fluid. Chemistry programs are used to control water chemistry forimpurities that accelerate corrosion andconaminants that maycause cracking due to SCC. This programrelies on monitoring and control of water chemistry to keep peak levels of various contaminants below thesystem-specific limits. The RNP Primary Water Chemistry Program is based on the current revision ofEPRI PWR Primary Water Chemistry Guidelines [19]. Later revisions of the guidelines will be usedwhen issued. The limits imposed by the RNP program meet the intent of the industry standard foraddressing primary water chemistry [19].

The evaluation. of this program against the 10 attributes in the GALL for Program XI.M2 [43] in supportof the RNP LRA remains applicable.

4.1.2 ASME Section XI Inservice Inspection Program

The RNP ASME Code Section XI, ISI Program [17] is implemented.to monitor for aging effects such ascracking, loss of preload due to stress relaxation or irradiation creep, loss of material, and reduction offracture toughness due to thermal embrittlement. For RNP, inspections conducted under the reactorinternals AMPs will be controlled as a combination of ASME Section XI ISI exams on core supportstructures and augmented exams performed under that ISI Program for the remaining reactor internalscomponents addressed within MiRP-227-A. The RNP Section XI, 10-year ISI examinations supportingthe license renewal period are currently scheduled for fall 2011, RO-27.

The evaluation of this program against the 10 attributes in the GALL for Program XI.M1 [43] in supportof the RNP LRA remains applicable.

4.1.3 Flux Thimble Eddy Current Program

Flux thimble tubes are long, slender, stainless steel tubes that are seal welded at one end with flux thimbletube plugs, which pass through the vessel penetration, through the lower internals assembly, and finally.extend to the top of the fuel assembly. The bottom-mounted instrumentation (BMI) column assembliesprovide a path for the flux thimbles into the core from the bottom of the vessel and protect the fluxthimbles during operation of the reactor. The flux thimble, provides a path for the neutron flux detectorinto the core and is subject to reactor coolant pressure on the outside and containment pressure on theinside.


WESTINGHOUSE NON-PROPRIETARY CLASS 3' 4-5

The RNP thimble tube inspection program is an existing plant-specific program that satisfies NRCBulletin 88-09 requirements that a tube wear inspection procedure be established [18] and maintained forWestinghouse-supplied reactors that use bottom-mounted flux'thimble tube in'strumentAtion. Theprogram follows Branch Technical Position RLSB-1, Aging Management Review - Generic, which isincluded in Appendix A of NUREG-1800: The program oincl udes eddy current testing requirements for-"thimble tubes and triteria for determining sample size, inspection'frequency, flaw evaluation, andcorrective action in accordance with NRC Bulletin 88-09 [20].

The evaluation of this program against the 10 attributes in the GALL for -Program X1'.M37 [43]' in supportof the RNP LRA remains applicable.

4.2 SUPPORTING RNP 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 [1] as directed by corporate procedure EGR-NGGC-0504,"Mechanical System Aging Management Review for License Renewal" [9]. The corporate proceduredirects the use of calculations as the vehicle to implement the license renewal aging management reviewand to document the results. Calculation RNP-L/LR-0354A [5] and RNP-L/LR-0354B [6] document theresults of the aging management review performed in support of RNP license renewal for reactorinternals. The RNP LRA was approved by the NRC in NUREG-1785 [2]. RVI components specifically',noted as requiring aging management, as identified in the NUREG, are summarized in Appendix BTables B- I and B-2 of this AMP. . : • .. . .

The calculations supported the' LRA, as follows'!• . ' ., .. . ". , ' . . . . . . .. •

1. Ideniifiedapplicable aging effects requiring managemeni

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 calculation

Aging management reviews were performed for each RNP system that contained long-lived, passivecomponents requiring aging management review, in accordance with the screening process of RNPEGR-NGGC-0503, "Mechanical Component Screening for LicenSe Renewal" [21 ]. This'review is notrepeated here, but the results are fully incorporated into the RNP RVI AMP.

4.2.2 Flux Thimble Tubes

A comprehensive e'valuation was performed to evaluate the expected wear for the BMI combinationthimble tubes at RNP, and the results are documented in WCAP-12202 [23], which wasprovided inWestinghouse Letter, CPL-89-570, "Carolina Power & Light Company, H. B. Robinson Unit 2, BMIThimble Wear Program Reports," May 25, 1989 [47]. As a result of the evaluation and subsequent


4-6 WESTINGHOUSE. NON-PROPRIETARY CLASS 3

operation, a plant modification was performed. The modification consisted of a cut-and-cap operation tosecure thimbles N-5 and N-12 into their respective conduits at the seal table as.a result of failure to.achieve full insertion of the components. Following the initial modification, the remaining portions ofthimbles N,5 and N-I12 were removed through the core in a subsequent refueling outage with the finalconfiguration being two completely removed thimbles.

4.2.3 Reactor Head and Control Rod DriveMechanism Replacement,

The problems associated with degradation of Alloy 600 materials and PWSCC are well known anddocumented in the industry. As a result of the identification of the existence of the concern at RNP, aprogram was undertaken to replace the undesirable material with one more resistant to the effects ofPWSCC .for the reactor head. A side benefit noted to the selected component replacement option was theintroduction of design features to facilitate refueling outage activities. Detailed descriptions of all facetsof the replacement are retained in the plant records.

4.2.4 Control Rod Guide Tube Split 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 RNP are consistent with the latestEPRI guidelines as described in Section 4. 1.

The original RNP support pins were fabricated from INCONEL® Alloy X-750 that was hot rolled,.solution treated or annealed, and age hardened atvyarious temperatures and times depending on heat,manufacturer, and fabrication date. Support pins made of this material with the associated heat treatmentswere shown to be susceptible to IGSCC and likely to fail during the lifetime of a nuclear power plant.Westinghouse developed an improved support pin design and fabrication technique that significantlyreduced the susceptibility to IGSCC while maintaining the fatigue and wear requirements necessary tosupport continued uninterrupted service.

In response to the industry concern, split pins were replaced at RNP during the 1990 RO-.I 3. A secondreplacement occtqrred in spring. 2010 RO-26, which replaced the existing X-750 support pins with theWestinghouse-designed cold-worked Type 316 SS support pins to mitigate the concern for potential SCCinherent with the Alloy X-750 material [46]. Detailed descriptions of all facets of the replacement areretained in the plant records.

4.3 INDUSTRY PROGRAMS

4.3.1 WCAP-14577, Aging Management for Reactor Internals

The Westinghouse Owner's Group (WOG, now PWROG) topical report WCAP-14577 [26] 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 of


ME!MNGHOUSENON-PROPRIETARY CLASS 3- 477

extended operation, using approved aging management rhetho6dologies of the WCAP to developplant-specific aging management programs; -

The aging management review for the RNP internals,' documented in [5, 6] was'completed in accordance'with the requirements of WCAP-14577 [26].

4.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines

NMRP-227-A, as discussed in Section 2, Was developed by a team of-industry experts includingutilityrepresentatives, NSSS vendors, independent consultants, and international committee representatives whoreviewed available ddta;and industry experience on materials aging. The objective'of the group was todevelop a consistent, systematic approach for identifying and prioritizing inspection and evaluatibnrequirements 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'ategorized Within MRP-227-A as."Primary" components, "Expansion" comionibhitý, "Existiiig Programs" components, or "No Additionhal-Measures" components, as described asf6llows: , ,, ..:

Primar. - - " " .

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 guidelines4 . ThePrimary group also includes components that have shown a degree of tolerance toa 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 P.WR 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 requirements. Anycomponents that are classified as core supportstructures, as defined in ASME B&PV Code Section XI IWB.2500, Category B-N-3, have requirementsthat 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. guideline includes Mandatory andNeeded elements as follows:

* 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-22 7-Rev. 0 (that is, no later than December 31, 2011). ,

RNP Applicability: MRP-227, Revision 0 was officially issued by the industry, in December'2008. An AMP must therefore be developed by December 2011. Progress Energy developed thisAMP for RNP to meet its license renewal commitment that pre-dates the implementation datecontained in MRP-227, Revision 0.



According to the NRC Regulatory Issue Summary (RIS) [44], because RNP has submitted theirAMP for approval by the NRC, RNP may withdraw their submittal and provide new and revisedcommitments to the NRC to resubmit their AMP in accordance with MRP-227-A [11] no laterthan October 1, 2012.

* Needed

There are four Needed elements:

1. Each commercial U.S. P WR unit shall implement MRP-22 7-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.

RNP Applicability: MRP-227-A augmented inspections have been appropriately incorporatedinto the RNP ISI Program 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-I, 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 [2.7].

RNP Applicability: -Inspection standards will be in.accordance with the requirements of.MRP-228 [27]. These inspection standards. will beused for augmented inspectioniatRNP asapplicable 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 enteredlin the plant corrective action programand dispositioned.

RNP Applicability: RNP 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-22 7-A are examined.

RNP Applicability: As discussed in Section 4, Progress Energy will participate in future industryefforts and will adhere to industry directives for reporting, response, and follow-up.

4.3.2.3 GALL AMP Development Guidance

It should be noted that Section XI.M16A of NUREG-1801, Revision 2 [43] includes a description of theattributes that make up an acceptable AMP. These attributes are consistent with the RNP AgingManagement Review process. Evaluation of the RNP RVI AMP against GALL attribute elements isprovided in Section 5 of this program plan.

WCAP-17077-NP * August 2012Revision I


As part of License Renewal, RNP agreed to participate in industry activities associated with thedevelopment of the standard Industry Guideline for Inspection and Evaluation of Reactor Internals, Theindustry efforts have defined the required inspections and examination techniques for those componentscritical to aging management of RVI. The results of the industry recommended inspections, as publishedin MRP-227-A, serve as the basis for identifying any augmented inspections that are required to complete

the RNP RVI AMP.

4.3.2.4 MRP-227-A Applicability to RNP

The applicability of MRP 7227-A to RNP 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.

RNP Applicability: RNP fuel management program changed from a high- to a low-leakage coreloading pattern prior to 30 years of operation.

Base load operation, i.e., typically operates atfixed power levels and does not usually vary poweron a calendar or load demand schedule.

RNP Applicability: RNP operates as a base load unit.

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

RNP Applicability: MRP-227-A states that the recommendations are applicable to all U.S. PWRoperating plants as of May 2007 for the three designs considered. RNP has not made anymodifications to reactor internals components since May 2007 that would have an impact on theapplicability of MRP-227-A. RNP replaced control rod guide tube (CRGT) support pins with anupgraded material in spring 2010. The modification has no impact on the applicability ofMRP-227-A and is an example of the RNP proactive approach to managing aging reactorinternals.

Based on the RNP applicability, as stated, the MRP-227-A work is representative for Robinson NuclearPlant.

4.3.3 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. Progress Energy willmaintain cognizance of industry activities related to PWR internals inspection and aging management;


WESTINGHQUSE-NON-'PROPRJETARY CLASS 3 4-11

and will address/implement industry guidance, stemming from those activities, as appropriate underNEI 03-08 practices.

4.4 SUMMARY i4

It should be noted that the Progress Energy RNP, the MRP, and the PWROG approaches to aging.management 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 RNP for the initial basis of theLRA that resulted in the NRC SER in NUREG-1785 [2].

The approach used to develop RNP AMPs is fully compliant with regulatory directives and approveddocuments. The additional evaluations and analysis completed by the MRP industry group have providedclarification to the level of inspection quality needed to determine the proper examination method andfrequencies. The tables provided in MRP-227-A and included as Appendix C of this AMP provide thelevel of examination required for each of the components evaluated.

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



5 H.B. ROBINSON REACTOR INTERNALS AGING MANAGEMENTPROGRAM ATTRIBUTES

The RNP RVI AMP is credited for aging management of RVI components for the following eight agingdegradation mechanisms and their associated effects:

* Stress corrosion cracking0 Irradiation-assisted stress corrosion cracking* Wear* Fatigue* Thermal aging embrittlement0 Irradiation embrittlement* Void swelling and irradiation growth0 Thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep

The attributes of the RNP Reactor Internals AMP and compliance with NUREG-1801 (GALL Report),Section XI.M16A, "PWR Vessel Internals" [43] are described in this section. The GALL identifies 10attributes for successful component aging management. The framework for assessing the effectiveness ofthe projected program is established by the use of the 10 elements of the GALL.

RNP fully utilized the GALL process contained in NUREG- 1801 [4] in performing the agingmanagement review of the reactor internals in the license renewal process. However, RNP made acommitment (see NUREG-1785 [2]) to incorporate the following: (1) RNP will continue to participate inindustry programs to investigate aging effects and determine the appropriate AMP activities to addressbaffle and former assembly issues, and to address change in dimensions due to void swelling, (2) as WOGand EPRI MRP research projects are completed, RNP will evaluate the results and factor them into thePWR Vessel Internals Program as appropriate, and (3) RNP will implement an augmented inspectionduring the license renewal term. Augmented inspections, based on required program enhancementsresulting from industry programs, will become part of the ASME B&PV Code, Section XI program.

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 RNP 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 H.B. Robinson Unit 2 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. P WR nuclear power plantsdesigned by B& W, CE, and Westinghouse. The scope of components considered for inspectionunder MRP-22 7 guidance includes core support structures (typically denoted as ExaminationCategory B-N-3 by the ASME Code, Section XI), those RVI components that serve an intended


5-2 WESTTNGHOUSF-.)NON-PROPRIETARY CLAS$ I

license renewal safety function pursuant to criteria in 10,CFR 54.4(a) (1), and other R VI.

components whose failure could prevent satisfactory accomplishment of any of the functionsidentified in 10 CFR 54.4(a) (1)(i) , (ii), or (iii). The scope of the program does not include

consumable items, such as fuel assemblies, reactivity control assemblies, and 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 10CFR 54.21(a)(1), The scope of the program also does not include welded attachments.to the internal surface of the reactor vessel because these components are considered to be ASME

Code Class I appurtenances to the reactor vessel and are adequately managed in accordance

with an applicant's AMP that corresponds to GALL AMP XI.M1, "ASME Code, Section XIInservice Inspection, Subsections IWB, IWC, and IWD."

The scope of the program includes the response bases to applicable license renewal applicant

action items (LRAAIs) on the MRP-22 7 methodology, and any additional programs, actions, oractivities that are discussed in these LRAAI responses and credited for aging management of the

applicant'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 of

the MRP's augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 ofMRP-22 7), and NSSS vendor-specific or plant-specific LRAAIs as well. The responses to the

LRAAIs on MRP-227 are provided in Appendix C of the LRA.

The guidance of MRP-22 7 specifies applicability limitations to base-loaded plants and the fuel

loading management assumptions upon which the functionality analyses were -based. These•.

limitations and assumptions require a determination of applicability by the applicant for eachreactor and are covered in Section 2.4 ofMRP-22 7 "'[43.]..

RNP Program Scope , ,1 -, ... .

The RNP RVI consist of two basic assemblies: (1) an upper internals assembly that is removed duringeach refueling operation to obtain access -to the reactorcore, and (2) a lower internals assembly that can

be removed, if desired, following a complete core unload. Additional RVI details are provided inSections 3.9.5 and 4.1 of the RNP UFSAR.

The RNP RyI subcomponents that required aging management review are indicated in the previouslysubmitted Table 2.3-1 of the RNP Application for Renewal Operating License [37]. The portion of this

table associated with the internals is included as part.of the tables in Appendix B. The table lists the.subcomponents of the RVI that required aging management review along with each subcomponentpassive function(s) and reference(s) to the corresponding AMR table(s) in Section 3 ,of the RNP LicenseRenewal Application.

The RNP, Reactor Internals Aging Management Review, was conducted and documented! in the RNP

Aging Management Review Calculations [5, 6]. The table summarizing the results of that review is alsoincluded in the tables of Appendix B. The tables identify those aging effects that require management for

those components requiring AMR.. A column in the tables lists the program/activity that is credited toaddress the component and aging effect during the period of extended operation. The NRC has reviewed


WESTINGHOUSE NOMPROPRIETARY CLASS 3. 5,-3

and approved the aging management strategy presented in the Appendix B tables as documented in theSER on license renewal [2].

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, lossof 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 metbyRobinson Unit 2 and are addressed in subsection 4.312.4 of this AMP. The Topical Report Conditionsand Applicant/Licensee Action Items provided by the NRC in the SE on MRP-227, Revision 0 [11] aremet by RNP and demonstration of compliance is addressed in Section 6.1 for the Topical ReportConditions and in Section 6.2 for the Applicant/Licensee Action Items. The RNP RVI AMP scope isadditionally based on previously established and approved GALL Report approaches through applicationof the WCAP-14577 [26] methodologies to determine those components that require aging management.

Conclusion

This element complies with the corresponding aging managementattribute in NUREG-1801,Section XI.MI6A [43] and Commitment 33 in the RNP 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 formns[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 pl'esented in GALL AMP XLM2, 'Water Chemistry"' [43].

RNP Preventive Action

The RNP reactor internals AMP includes the Primary Water Chemistry Program [ 15, 16] as an existihgprogram that complies with the requirements of this element. A description and applicability to the RNPreactor internals AMP is provided in the following subsection.

Primary Water Chemistry Program,

To mitigate aging effects on component surfaces that are exposed to water as process fluid, 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 RNP PWR Primary'Water Chemistry Program [15, 16] is based on the current, approved revisions of EPRI PWR PrimaryWater Chemistry Guidelines.


5-4 WESTINGHOUSE NON-PROPRIETARIY CLASS 3

This program is consistent with the corresponding program described in the GALL Report [22].

The limits of known detrimental contaminants imposed by the chemistry monitoring program areconsistent with the EPRI PWR Primary Water Chemistry Guidelines [19].

Conclusion

This element complies with thecorresponding aging management attribute in NUREG-1801,Section XI.M16A [43] and Commitment 33 in the RNP 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 compqnents at the facility: (a) cracking induced by SCC,PWSCC, L4SCC, 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 if a 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 !oss of material occurring in the components. Forthe management ofloss ofpreload, the program monitors for gross surface.conditions. that may.be 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 aging

or neutron irradiation embrittlement, or by void swelling and irradiation growth; instead, theimpact of loss offracture toughness on component integrity is indirectly manqged by using visualor volumetric examination techniques to monitor for cracking in the components and by applying

applicable reduced fracture toughness properties in the flaw evaluations if cracking is detected inthe components and is extensive enough to warrant a supplementalflaw growth orflaw 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 flux


WESTINGHOUSE NON-PROPRIETARY CLASS 3 - -5

thimble tubes in GALL AMP XIM3 7, "Flux Thimble Tube Inspection. " No inspeetions, exceptfor those specified in ASME Code, Section XI, are required for components that are identified asrequie'ing "No Additional Measure;' ",in accordance with the analyses reported in MRP-22 7"

[43].

RNP Parameters Monitored or Inspected

The RNP AMP monitors, inspects, and/or'tests for the effects of the eight aging degradation mechanismson the intended function of the RNP PWR internals components through inspection and conditionmonitoring activities in accordance with the augmented requirements defined under industry directives ascontained in MRP-227-: and ASME Seciion XI. ".

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 Ex-pansion 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 inspeci the RVI through the period ofextended operation to address the effects of the eight aging degradation mechanisms.

For license renewal, the ASME Section X1 Program consists of periodic volumetric, surface, and/or visualexamination 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 ASMESection X1. This program is consistent with the corresponding program described in the GALL Report[22].

Appendices B and C of this AMP provide a detailed listirig'obfthe components and subcomponents andthe parameters' nionitored, inspected, and/or tested.

Conclusion

This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,

Section XI.MI6A [43] and Commitment 33 in the RNP SER.

5.4 GALL REVISION 2ELEMENT 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 for

detecting the aging effects of interest; and (b) standards for examination methods, procedures,and personnel are provided in'a companion document, MRP:228. In all cases, well-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-1, and EVT-1) examinations for detecting effects ranging from general conditibns to

detection and sizing of surface-breaking discontinuities. Surface examinations may also be used



as an alternative to visual examinationsfor detection and sizing of surface-breakingdiscontinuities.

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 reduced.fracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspect for-loss of material induced by wear and for general aging conditions, such as gross 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-22 7for 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 of MRP-22 7 and for Westinghousle. designed Expansion Components in. Table 4-6 ofMRP-22 7.

The program is supplemented.by the following plant-specifc Primary Component and ExpansionComponent inspections for the program (as applicable).: for RNP, no additional.Primary or.Expansion components are relevant to the scope of aging management for the R VI.:

-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. The. ...physical measurements methods applied in accordance with. this program include no specific.component at RNP. The hold down spring at RNP isfabricatedfrom Type 304 SS that doesrequire inspection by physical measurement; however, RNP intends to replace the hold downspring in fall 2013. Therefore, RNP will not be required to perform physical measurements of the

;hold down spring per MRP-22 7-A '. [43].

RNP Detection of Aging Effects.

Detection of indications that are required by the ASME Section XI ISI Program is well established andfield-proven through the application of the Section XI ISI Program. Those augmented inspections that aretaken from the MRP-227-A recommendations will be applied through use of the MRP-228 InspectionStandard. This AMP implements the augmented inspection requirements of Table 4-3, Table 4-6, and


WESTMHOUSENONý,PROPRIETARY CLASS 3 5-7

Table 4-9 from MRP-227-A forthe Primary, Ex'pansion,'and Existing Components, respectively. Theseare included in Appendix C of this AMP for reference. These tables include the inspection frequency andsampling bases. For the Expansion Components of MRP-227-A, this AMP implements the expansionrequirements 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. Three different visual techniques are includeVT-3, VT-1, 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 augmrentedreactor internals inspections are documented in MRP-228.

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 [41]. 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

•-Structural degradation of attachment welds such that the original cross-sectional area: is reducedby more than 10 percent

These reqdii'ements are defined to ensure the integrity of attachment Welds on'the ferritic pressure vessel.Although the IWB-3520 criteria do not'directlyapply 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. In MRP-227-A note that VT-I hasonly 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- I 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-I) is also:conducted inaccordance with the requirements described for visual examination (i.e., VT-I) with additional


5-8 WESTINGHOUSE NON-, PROPRIETARY CLASS 3

requirements (such as camera scanning speed) currently being developed by the industry. Anyrecommendation for EVT-1 inspection will require additional analysis to establish flaw-tolerance criteria,which must take into account potential embrittlement due to thermal aging or neutron irradiation. Theindustry, through the *PWROG, has developed an approach for acceptance criteria methodologies tosupport plant-specific augmented examinations. This work is summarized in WCAP- 17096-NP, "ReactorInternals Acceptance Criteria Methodology and.Data Requirements" [45]. The acceptance criteriadeveloped using these methodologies maybe created on either a generic or plant-specific basis becauseboth loads and component dimensions may vary from plant-to-plant within a typical PWR design.

VT-3 Examination for General Condition Monitoring

In the augmented inspections detailed in the MRP-227-A for reactor internals, the VT-3 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 component-cracking 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 support

structures 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: ..

* Structuraldistortion or displacement of-parts.to the extent that. component function may be,.impaired;

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

Foreign materials or accumulation of corrosion products that could interfere with control 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 is .

reduced 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.


WESTIN614OUSE NON PROPRIETARY CLASS I 5-9

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 recommendedexclusively for detection of flaws, in.bolts. 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 itsis 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 multipleminimum acceptable bolting patterns. The evaluation program must demonstrate that the~remaining bolts,meet the requirements for a minimum bolting pattern for continued operation..• The evaluation procedures.must also demonstrate that the pattern of remaining bolts contains sufficient margin such that continuationof the bolt failure rate will not result in failure of the system to meet the requirements for minimum.acceptable bolting pattern before the next inspection.

Establishment of the minimum acceptable bolting pattern for any system of bolts requires analysis to,demonstrate that the system will maintain reliability and integrity in continuing to perform the intendedfunction of the component. This analysis is highly plant-specific. Therefore, any recommendation forUT inspection of bolts assumes that the plant owner will work with the designer to establish minimum'acceptable bolting patterns prior to the inspection to support continued operation. For Westinghousedesigned plants, minimum acceptable bolting patterns for baffle-fornier: and barrel-former bolts areavailable through the PWROG. Progress Energy has been a full participant in the development of the,PWROG documents and has full access and use.

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 ormaterial deformation. For RNP, direct physical measurements are required only for:the hold downspring; however, RNP is planning to replace the current hold down spring in fall 2013. Therefore, RNPwill not be required to perform a physical measurement of the hold down spring.

Conclusion .

This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.MI6A [43] and Commitment 33 in the RNP 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 aswellfor performing applicable limit load, linear elastic and elastic-plastic fracture analyses ofrelevant flaw. indications. The examinations and re-examinations required by the MRP-22 7guidance, 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 within

the 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 than

* anticipated, plus the continuation of the Existing Programs activities, such as the ASME Code,Section XI, Examination Category. B-N-3 examinations for core support structures, provides ahigh degree of confidence in the total program" [43.].,

RNP 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 frequenicy of failure in reactor internals makes monitoring and trendingbased on operating experience somewhat impractical, The majority of the materials aging degradationmodels used to develop the MRP-227-A guidelines are basedon test data from reactor internalscomponents removed from service. The data is used to identify trends in materials degradation andforecast potential component degradation. The industry continues to share both material test data andoperating experience through the auspices. of the MRP and PWROG. Progress Energy has in the past andwill continue to maintain cognizance of industry activities and shared information related to PWRinternals inspection and aging management.

Inspections credited in Appendix B are based on utilizing the RNP 10-year ISI program and theaugmented inspections derived from MRP-227-A and repeatedhere in Appendix C. The MRP-227-Ainspections only augment and do not replace the existing ASME Section XI ISI requirements. Theseinspections, where practical, are scheduledto be conducted in conjunction with typical 10-year ISIexaminations.

Appendix C, Tables C-1, C-2, and C-3 identify the augmented Primary andExpansion 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. Table C-4 in Appendix Cidentifies the MRP-227-A expansion criteria from the Primary components. If these expansion criteriaare met for a component, the associated Expansion component is to be inspected to manage the agingdegradation.


WEST.fNGHOUSE NON-PROPRiETARY CLASS 3 5-11.

Reporting requirements are included aspaht of thieMRP-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 preemptiveaction through update of the MRP guidelines.

Conclusion

This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,

Section XI.M16A [43] and Commitment 33 in the RNP 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 coveredby 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:

. "0 'For visual examination (and surface examihation'as an alternative to visual examination),the examination acceptance criterion is the absehbe& of any of the specif descriptiverelevant conditions; inhaddition, there are reqidir-mneht&'to-record and disposition surface'

breaking indications that are detected' and sized for length by VT-1/E VT-1 examinations, '

% For volumetric examination, the examination acceptance criterion is the capability for

reliable detection of indications in bolting, as demonstrated in the examination Technical:Justification; in addition, there are requirements for system-level assessment of bolted or

pinned assemblies with unacceptable volumetric (UT) examination indications that exceedspecified limits, and

Forphysical measurements, the excamination acceptance criterion for the acceptable 'tolerance in the measured differential height from the top of the plenum rib pads to the vesselseating 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. RNP is planning to replace the• hold down spring in Fall 2013; therefore, RNP is not required to produce acceptance criteriafor the physical measurements on the hold down spring" •[43]. *


5-.12 WESTINGHOUSE NON-PROPRIETARY CLASS 3

RNP Acceptance Criteria

Those recordable indications that are the result of inspections required by the existing RNP ISI programscope are evaluated in accordance with the, applicable requirements of the ASME Code through theexisting Corrective Action Program [28]..

Inspection acceptance andexpansion 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 RNP procedures to enable the examiner to identify examination acceptance criteriaconsidering state-of-the-art information and techniques .

Augmented inspections, as defined by the MRP-227-A requirements included in this AMP as AppendixC, Table C-1,.Table C-2, and Table C-3, that result in recordable relevant conditions will be entered intothe plant Corrective Action Program and addressed by appropriate actions that may include enhancedinspection, repair, replacement,. mitigation actions, or analytical evaluations. An example of an analyticalevaluation, is using a minimum bolting WCAP approach such as those commonly used to supportcontinued component or assembly functionality. Additional analysis to establish acceptable boltingpattern evaluation criteria for the baffle-former bolt assembly, as contained in various industry documents[38], is also considered in determining the acceptance of inspection results to support continuedcomponent 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 WCAP4 17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data, -Requirements,'" [45], which details acceptance criteria methodology for the MRP-227 Primary andExpansion'components. Status is monitored through.direct Progress Energy cognizance of industry(including PWROG) activities related to PWR internals inspection.and aging management.

Conclusion

This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.MI6A [43] and Commitment 33 in the RNP 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. notsatisfy the examination acceptance criteria are required to be dispositioned through the plantcorrective action program, which may require. repair, replacement, or analytical evaluation forcontinued 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 andevents. Examples of methodologies that can be used to analytically disposition unacceptableconditions are found in the ASME Code, Section XI or in Section 6 of MRP-22 7. 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 detected

• condition, or the alternative of component repair and replacement procedures. The latter are.subject to the requirements of the ASME Code, Section XI. The implementation of the guidance inMRP-22 7, 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 1O CFR Part 50, Appendix B or its equivalent, as applicable.- •. | : '.

Other alternative corrective action bases may be used to disposition relevant conditions if they.have been previously approved or endorsed by the NRC. Examples ofpreviously NRC-endorsedalternative 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 of..Westinghouse Report No. WCAP-14577-Rev. ]-A, or for B& W-designed RVI components in B& W

Report No. BA W-2248. Westinghouse Report No. WCAP-14577-Rev. 1-A was endorsed for use inan NRC SE to the Westinghouse Owners Group, dated February 10; 2001. B& WReport No.BA W-2248 was endorsed for use in an SE to Framatome Technologies on behalf of the B& W,Owners Group, dated December 9, 1999. Alternative corrective action bases not approved orendorsed by the NRC will be submitted for NRC approval prior to their implementation " [43].

RNP Corrective Action

The existing RNP .procedure for corrective actions, the "Condition Evaluation and Corrective ActionProcess," [30] and the ASME Section XI ISI program [17], will be credited for this element. Theseprocedures establish the RNP repair. and replacement requirements of ASME Code'Section XI, "Rules for'Inservice Inspection of Nuclear Power Plant Components." These requirements'include the identification',of a repair cycle, repair plan, and verification ofacceptability for replacements. RNP is committed toperforming corrective actionsfor augmented inspections usinigrepair and replacement procedures "equivalent to those requirements in ASME B&PV Code, Section XI.

Conclusion

This element complies with the corresponding aging management attribute in NUREG-1801,

Section X1.M16A [43] and Commitment 33 in the RNP 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 lO 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" [43].


5-14 WESTINGHOUSE NON7PROPRIETARY CLASS 3

RNP Confirmation Process.-

RNP has an established 0. CFR Part 50, Appendix B, Program [33, 34, 35] that addresses the elements ofcorrective actions, confirmation process, and administrative controls. The RNP Program includes non-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 -omplies with or exceeds the corresponding aging management attribute in NUREG-1 801,Section XI.M16A [43] and Commitment 33.in the RNP 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 " [43].

RNPAdministrative Controls

RNP has an established 10 CFR 50, Appendix.,B Program [33, 34,3•5] that addresses the elements.ofcorrective actions, confirmation process, and administrative controls. The RNP program includes non-safety-related structures, systems, and components.. Quality. assurance (QA) procedures, review. and...approval processes, and administrative contro.ls are implemented in accordance withthe requirements.of10 CFR 50, Appendix B.

Conclusion

This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.M16A [43] and Commitment 33 in the RNP SER.

5.10 GALL REVISION 2 ELEMENT 10: OPERATING EXPERIENCE


"Relatively few incidents ofPWR. internals aging degradation have been reported, in operatingU.S. commercial PWR plants. A summary of observations to date is provided in Appendix A ofMRP-227-A. The applicant is expected to review subsequent operating experience for impact onits program or to participate in industry initiatives that perform this function.


WESTINGHOUSE NqON-PROPRIETARY CLASS 3 5-15

The application of the MRP-22 7 guidance will establish a considerable'amount of operating'experience over the next few years. Section 7 ofMRP-227 describes the reporting requirementsf6r these applications, and the planfobr evaluating the accuwhulaied additional operatingexperience" [43].

RNP Operating Experience

Extensive industry and RNP operating experience has been reviewed during the development of theReactor Vessel Internals Aging Management Program. The experience reviewed includes NRCInformation Notices 84-18, "Stress Corrosion Cracking in PWR Systems" and 98-11, "Cracking ofReactor Vessel Internal Baffle Former Bolts in.Foreign Plants." Most of the industry-operatingexperience reviewed has involved cracking ofaustenitic stainless steel baffle-former bolts or SCC ofhigh-strength internals bolting. SCC of control rod guide tube split pins has also been reported.

Early plant operating experience related to hot functional testing and reactor internals is documented inplant historical records. Inspections performed as part of the 10-year IS Iprogram have been conducted asdesignated by existing commitments and would be expected to discover overall general internals structuredegradation. To date, very little degradation has been observed industry-wide.

Industry operating experience is routinely reviewed by Progress Energy engineers using INPO OperatingExperience (OE), the Nuclear Network, and other information sources as directed under the applicableprocedure [29], 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-specificexperiencewith re~ictorvessel internals reveals thattheU.S.: industry, including Progress Energy and RNP,- has responded proactively to industry issues relativeto reactor internals degradation. -Two examples that demohstrate this proactive response are thereplacement of control rod guide tube split pins in 1990 and 2010, as well as participation by ProgressEnergy in the augmented examinations performed by the PWROG on control rod guide tube guide cardsin spring 2010. These are briefly described in the following paragraphs.

* RNP Control Rod Guide Tubes Split Pins

The control rod guide tube split pins were replaced at RNP'diuring the 1990 refueling outage. The newpins were considered a replacement in kind and an improvement over the previous pins. Ongoingindustry experience has' shown continued ddgradation of this cbmponent and, in response, RNP performeda second split pin replacement activity in spring 2010 RO-26. The replacement included a materialupgrade from X-750 to 316 SS in support of managing aging in the component.

RNP Participation as a Representative Pilot Plant in the PWROG Control Rod Guide CardsInspection Prop-ram

The PWROG has conducted upper internals control rod guide tube guide card wear measurements on asample of guide tubes from selected representative pilot plants to approximate the remaining life of theguide tube guide cards. This was a proactive effort by the U.S. industry to establish criteria for inspection


5-16 WESTTNGHOUSE NON-PROPRIETARY CLASS 3

and gather data to support aging management of the component. RNP.p-oactively inspectedguide cardsin 1990 RO-13, and was a key participant as one of the representative pilot inspection plants. TheRNP inspection in support of the industry program. was completed in spring 2010 RO-26, with anevaluation of the guide card wear being completed in WCAP-.17277-P [53]. Inspection outcomes wereevaluated to ensure compliance with MRP-227 specifications. The.PWROG pilot program has beencompleted, and a draft WCAP [54], which gives the generic inspection criteria, guidelines andrecommendations for guide card wear, has been issued for review by the utilities.

A key element of the MRP-227-A Guideline is the reporting of age-related degradation of reactor vesselcomponents. Progress Energy, through its participation in PWROG and EPRI-MRP activities, willcontinue to benefit from the reporting of inspection information and will share its own operatingexperiencewith the industry through the reportingrequirements of Section 7 of MRP-227-A. RNPpresented the results of its initial MRP-227, Revision 0 inspection at the April 2012 PWROG MaterialsSubcommittee (MSC) meetings [55]. The collected information from MRP-227-A augmentedinspections 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.MI6A [43] and Commitment 33 in the RNP SER.

WCAP- I 7077-NP * August 2012WCAP-1i7077-NP August 2012

Revision I

WESTINGHOUSE -NON-PROPPJETARY CLASS 3 .6-1

6 DEMONSTRATION...

Robinson Nuclear Plant has demonstrated a lnfig-term commnitment to aging management of reactorinternals. This AMP is' based on an established history of programs to identify and nmonitor 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 RNP reactor vessel internals have beenperformed during each 10-year interval since plant operations "commenced.

Asdocumented in RNP operational procedures, reports are continuously reviewedhbyRNP personnel foi applicable issues that indicate operating procedures or prbgrams requireupdates based on new OE. " . -

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

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

Replacement control rod guide tube split pins for RNP in 1990 were fabricated from moreresistant Alloy X-750 materials, with modified geometry and heat treatment to increase resistanceto SCC (versus original pins). The new Type 316 SS split pins that were used in the replacementin spring 2010 will provide additional resistance to PWSCC.

Replacement of the Type 304 SS hold down spring is planned for spring 2013. The replacementspring will be fabricated from a modified Type 403 SS, which is more resistant to stressrelaxation than the original Type 304 material.

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

This AMP fulfills the approved license renewal methodology requirement to identify the most 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 wereperformed at RNP in spring 2012 RO-27. The augmented inspections discussed in compliance withMRP-227-A requirements have been integrated in the inspection procedures, which were used to perform



the ASME Section XI, 10-year ISI examinations.. Integration of therequired 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 RNPprograms, and use of Operating ExperienceReports (OERs), provide reasonable assurance that the reactor internals will continue to perform theirintended 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 cameout of the NRC review of MRP-227, as listed in [11], 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 Conditions Compliance to SE on MRP-227Topical Condition Applicable/Not Compliance in AMP

_ _ _'_ _ _ Applicable

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 applicablecomponents subject to multiple Applicabledegradation mechanisms

4. Imposition of minimum . •Applicable Notes 2 through 4 were added to Table C-I,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 inspection 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.

7. Updating of MRP-227, Revision 0, Applicable Section 5 is updated to reflect XI.M 16AAppendix A from GALL Revision 2 [43].


WESTINGHOUSE NON-'PROPMETARY CLASS 3 6-3-

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 respbnsiblefor assessingits plant's design and operating history and demonstrating that the approved version' ofMRP-22 7is applicable to the facility. Each applicant/licensee shall refer, in'paiticular, 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-227 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. The*applicant/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 I" [I11].

RNP Compliance

The process used to verify that RNP 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 is as follows:

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

2'. Identification of RNP PWR internal comporients.

3. Comparison of the typical Westinghouse PWR internal components to the RNP PWRinterihai components.

a. Confirm that no additional items were identified by this comparison (primarilysupports Applicant/Licensee Action Item 2).

b. Confirm that the materials identified for RNP are consistent with those materialsidentified in MRP-191, Table 4-4.

c. Confirm that the RNP internals are the same ,as, or equivalent to, the typicalWestinghouse PWR internals'regarding design and fabrication.

4. Confirmation that the RNPoperating history is consistent with the assumptions inMRP-227-A regarding core loading patterns.

5. Confirm that the RNP RVI materials operated at temperatures within the original designbasis parameters.

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



7. Confirmation that any changes to the RNP RVI components do not impact the applicationof the MRP-227-A generic aging management strategy.

RNP reactor internal components are represented by the design and operating history assumptionsregarding neutron fluence, temperature, materials, and stress values in the MRP-191 generic FMECA and

the MRP-232 functionality analyses based on the following:

I . RNP operating history is consistent with the assumptions in MRP-227-A with regard toneutron 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 patternsfollowed by 30 years of low-leakage core fuel management strategy. In fuelcycle C16 (November 12, 1993) [48] at 23 years of operation, RNP switched touse of a low-leakage core design. It is projected that RNP will continue to use alow-leakage core design for all subsequent fuel cycles, as discussed in subsection4.3.2.4. Therefore, RNP meets the fluence and fuel management assumptions inMRP- 191 and requirements for MRP-227-A application.

b. RNP has operated under base load conditions over the life of the plant as

discussed in subsection 4.3.2.4. Therefore, RNP satisfies the assumptions inMRP documents regarding operational parameters affecting fluence.

2. The RNP RVI operate between Thot and ,Told [48, 49], which are approximately equal to,but not less than 544.5'F for TCOId or higher than 596.5/610.3'F for Th.t. The designtemperature for the vessel is 650'F [50, 51]. RNP operating history is within originaldesign basis parameters, and therefore consistent with the assumptions use8l to ýdevelop eothe MRP-227-A aging management strategy with regard to temperature operationalparameters.

3. RNP internal components and materials are comparable to the typical Westinghouse PWRinternal components (MRP-191, Table 4-4) as summarized in [57].

a. No additional components were identified for RNP by this comparison [2].

b. Materials identified for RNP are consistent or reasonably eqUivalent with thosematerials identified in MRP- 19 1, Table 4-4 for Westinghouse-designed plants[32]. Where differences exist there is no impact on the RNP RVI program or thecomponent is already credited as being managed under an alternate RNP agingmanagement program.

c. RNP internals are consistent with the typical Westinghouse PWR'nternalsregarding design and fabrication [57].

4. Modifications to the RNP reactor internals made over the lifetime of the plant are thosespecifically directed by Westinghouse, the Original Equipment Manufacturer (OEM) asdiscussed in subsection 4.3.2.4. The design has been maintained over the lifetime of theplant as specified by the OEM, operational parameters with regard to fluence and


WESTINGHOUSE NONIPROPRIETARY CLASS 3 6-5

* temperature are compliant with MRP-227-A-requirements and the components andmaterials are the same as those considered in MRP-191. Therefore, the RNP stress valuesare represented by the assumptions in MRP-191, MRP-232, and MRP-227-A, confirmingthe applicability of the generic FMECA.

Conclusion

RNP complies with Licensee/Action Item 1 of the NRC SE on MRP-227, Revision 0, .and therefore meetsthe requirement for application of MRP-227-A as'a strategy for managing age-related materialdegradation in reactor internals components.

6.2.2 SE Applicant/LicenseeAction 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-1 91 and identify whether these tablescontain all of the R VI 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-22 7, 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" [1.].....

RNP Compliance

This action item requires comparison of the RVI cbmponents that are within the scope of license renewalfor RNP to those components contained in MRP-191, Table 4-4. A detailed tabulationof the RNP RVIcomponents [57] was completed and compared favorably to the typical Westinghouse PWR internalcomponents in MRP- 191. All components required ter be included in the RNP program [2] are consistentwith those contained in MRP- 191. Several components have different materials than specified but thesehave no effect on the recommended MRP aging strategy or are already managed by an alternate RNPprogram; therefore, no modifications to the program detailed in MRP-227-A need to be proposed. Thissupports the requirement that the AMP shall provide assurance that the effects of aging on the RNP RVIcomponents within the scope of license renewal, but not included in Table 4-4 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 RNP with no modifications.


6-6 WESTINGHOUSE NON-PROPRIUARY CLASS 3

Conclusion

RNP complies with Licensee/Action Item 2 of the NRC SE on MRP-227, Revision 0, and therefore meetsthe requirement for application of MRP-227-A as a strategy for managing age-related materialdegradation 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, gpplicants/licensees of CE and Westinghouse arerequired to perform plant-specific analysis either to justify the acceptability ofanapplicant '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 the

applicant 's/licensee's AMP application. The CE and Westinghouse components identified for thistype ofplant-specific evaluation include: CE thermal shield positioning pins and CE in-coreinstrumentation thimble tubes (Section 4.3.2 in MRP-22 7), and Westinghouse guide tube supportpins (split pins) (Section 4.3.3 in MRP-22 7). This is Applicant/Licensee Action Item 3" [11].

RNP Compliance

RNP is compliant with the requirements in Table 4-9 of MRP-227-A as applicable to RNP, is shown inAppendix C, Table C-3. This is detailed in the plant-specific RNP program documents for ASME SectionXI [ 13, 14, 1.7] and the plant-specific flux thimb!e program [ 18].

Conclusion

RNP complies with Licensee/Action Item 3 of the NRC SE on MRP-227, Revision 0, and therefore meetsthe requirement for application of MRP-227-A as a strategy for managing age-related materialdegradation 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 examination

methods 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" [ 11 ].


. WESTINGHfOUSE NON-PROPRIETARY CLASS 3 ... 6'-7

RNP Compliance

This applicant/licensee action item is not applicable'to RNP since itonly applies to B&W plants.

Conclusion

Licensee/Action Item 4 of the NRC SE on MRP-227, Revision 0 is not applicable to RNP.:

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 the

NRC-approved version ofMRP-227 for loss of compressibility for Westinghouse hold down

springs, and for distortion in the gap between the top and bottom core 'shroud segments in CE

units with core barrel shrouds assembled in two vertical sections. The applicant/licensee Shall

include its proposed acceptance criteria and an explanation of how the proposed acceptancecriteria are consistent with the plants' licensing basis and the need to maintain the functionality

of the component being inspected under all licensing basis conditions ofoperation 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" [11].

RNP Compliance

See Table 7-1. RNP is planning to perform a pr~empti'~ei64lacemenftof the hold down spring in fall2013 instead of the MRP-227-A inspections/physical measurements.

Conclusion

Licensee/Actiof :Item 5 Of the NRC SE on MRP-227, Revision 0 is required for a 304 SS hold downspring. RNP plans to replace their hold down spring in fall 2013; therefore, Licensee/Applicant ActionItem 5 is not applicable to RNP.

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

"As addiessed in Section 3.3.6 in this SE, MRP-22 7 does not propose to inspect the following

inaccessible components: the B& W core barrel cylinders (including vertical and circumferential

seam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices,

B& W core barrel-tb-former bolts and their locking devices, and B& W core barrel assemblyinter"nal baffle-to-baffle bolts. The MRP also identified that although the B& W core barrel

assembly internal baffle-to-baffle bolts are accessible; the bolts are non-inspectable usingcurrently available examination techniques:

Applicants/licensees shalljustify the acceptability of these components for continued operationthrough the period of extended operation by performing an evaluation, or by proposing a



scheduled replacement of the components. As part of their application to implement the approvedversion of MRP-227, applicants/licensees shall provide their justification 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" [11].

RNP Compliance

This applicant/licensee action item is not applicable to RNP since it only applies to B&W plants.

Conclusion

Licensee/Action Item 6 of the NRC SE on MRP-227, Revision 0 is not applicable to RNP.

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 their

facilities to demonstrate that B& W IMI guide tube assembly spiders and CRGT spacer castings,CE lower support columns, and Westinghouse lower support column bodies will maintain their

functionality during the period of extended operation or for additional R VI components that may

befabricatedfrom CASS, martensitic stainless steel or precipitation hardened stainless steelmaterials. These analyses shall also consider the possible loss offracture toughness in thesecomponents due to thermal and irradiation embrittlement, and may also need to considerlimitations on accessibility for inspection and the resolution/sensitivity of the inspection.techniques. The requirement may not apply to components that were previously evaluated as not.requiring aging management during development of MRP-22-7. That is, the requirement.would...apply to components fabricated from susceptible materials for which an individual, licensee has,

determined aging management is required, for example during their review performed inaccordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall beconsistentwith 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-22 7. This is Applicant/Licensee Action Item 7" [11].

RNP Compliance

Applicant/Licensee Action Item 7 from' the NRC's final SE on MRP-227, Revision 0 states that forassessment of CASS materials, the applicant/licensee for license renewal may apply the criteria in theNRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of CastAustenitic Stainless Steel Components," [52] as the basis for determining whether the CASS materials aresusceptible to the thermal aging mechanism. If the application of the applicable screening criteria [52] forthe component's material demonstrates that the components are not susceptible to either thermalembrittlement (TE) or irradiation embrittlement, or the synergistic effects of TE and irradiationembrittlement combined, then no other evaluation would be necessary. The RNP CASS components andthe assessment of their susceptibility to TE are summarized in Table 6-2.



Based on the criteria of [52], the RNP.CASS flow mixers,- lower support columns, and BMI. cruciforms,butt, and special columns are not susceptible to TE. Conclusive Confirmation of.material compositionunder TE susceptibility thresholds was not demonstrated for the CASS bases of the upper supportcolumns; thus, it is conservatively assumed that they are potentially susceptible to TE. Under MRP-191the bases were designated as CASS, screened-in for TE, SCC, and IE, and dispositioned for thesemechanisms under the FMECA. Thus, the conservative assumption that the upper support column basesare susceptible to TE has already been addressed in the development of the MRP-227-A inspectionrequirements.

Conclusion

RNP complies with Licensee/Action Item 7 of the NRC SE on MRP-227, Revision 0, and therefore meetsthe requirement for application of MRP-227-A as a strategy for managing age-related materialdegradation in reactor internals components.

The results of this CASS evaluation do not conflict with the MRP-227-A strategy for aging managementof RVI. It is concluded that continued application of the strategy of MRP-227-A will meet therequirement for managing age-related degradation of the RNP CASS RVI components.

Table 6-2 Summary of RNP CASS Components and their Susceptibility to TE

Susceptibility to TEMolybdenum (Based on the NRC

CASS Component Content Casting Ferrite Content • Criteria 152])Flow mixer devices, Low 0.5 max Static _<20% Not susceptible to TEwith and withoutthermocouple

Upper support column Low 0.5 max • Static Possible>20% Potentially susceptibleBases, with and. without,flow mixer ......

Lower support columns Low 0.5 max Static _<20% Not susceptible to TE

Bottom-mounted Low 0.5 max Static 5<20% Not susceptible to TE

instrumentation (BMI)cruciforms, butt, andspecial columns

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

and 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-22 7, as amended by this SE, as anAMP for the RVI components at their facility. This submittal shall include the information

identified in Section 3.5.1 of this SE. This is Applicant/Licensee Action Item 8" [11].



RNP Compliance

RNP has completed the initial submittal of their AMP, and according to [44], because RNP has submittedtheir AMP for approval by the NRC, RNP may withdraw their submittal and provide new and revisedcommitments to the NRC to resubmit their AMP in accordance with MRP-227-A [11] no later thanOctober 1, 2012.

Conclusion

RNP complies with Licensee/Action Item 8 of the NRC SE on MRP-227, Revision 0, and therefore meetsthe requirement:for application of MRP-227-A as a strategy for managing age-related materialdegradation in reactor intemals components.



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 inspection,pertaining to the reactor internals AMP. Should a change occur in plant operational practices or operating experience result in changes to the,projections, appropriate updates will be performed on affected plant documentation in accordance with approved procedures.

Table 7-1 Aging Management Program Enhancement and Inspection Implementation Summary

Refueling Project EstimatedOutage Month/Year EFPY AMP-Related ScopeO') Inspection Method and Criteria Comments

26 Spring 2010 28.9 Initial MRP-227, Revision 0 MRP-227, Revision 0 Guide card wear inspection wasaugmented inspection for inspections in accordance with completed at RNP during RO-26guide plates (cards) MRP-228 specifications

27 Spring 2012 30.4 ASME Code Section XI MRP-227, Revision 0 Third period of fourth inspection

Initial MRP-227, Revision 0 inspections in accordance with intervalaugmented inspections for MRP-228 specifications Inspections were completed at RNPcontrol rod guide tube lower during RO-27flange welds, upper core barrelflange weld, baffle-edge bolts,baffle-former assembly, andthermal shield flexures

28 Fall 2013 31.9 Initial MRP-227-A augmented MRP-227-A inspections in RNP plans to do a preemptiveinspections for upper and accordance with MRP-228 replacement of the hold down springlower core barrel cylinder girth specifications instead of MRP-227-A inspectionswelds, lower core barrel flangeweld, hold down spring, andbaffle-former bolts completedduring or before this outage

29 Spring 2015 33.4 Not applicable Not applicable Not applicable

30 Fall 2016 34.9 Not applicable Not applicable Not applicable.



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

31 Spring 2018 36.4 Not applicable Not applicable Not applicable

32 Fall 2019 37.9 Not applicable Not applicable Not applicable

33 Spring 2021 39.4 Subsequent MRP-227-A MRP-227-A inspections in Not applicableaugmented inspection for accordance with MRP-228guide plates (cards) specifications

34 Fall 2022 40.9 ASME Code Section XI MRP-227-A inspections in Third period of fifth inspection

Subsequent MRP-227-A accordance with MRP-228 interval

augmented inspections for specifications

control rod guide tube lowerflange welds, upper core barrelflange weld, baffle-edge bolts,baffle-former assembly, andthermal shield flexures

35 Spring 2024 42.4 Subsequent MRP-227-A MRP-227-A inspections in Not applicableaugmented inspections for accordance with MRP-228upper and lower core barrel specificationscylinder girth welds, lowercore barrel flange weld, andhold down spring completedduring or before this outage

36 Fall 2025 43.9 Not applicable Not applicable Not applicable

37 Spring 2027 45.4 Subsequent MRP-227-A MRP-227-A inspections in Not applicableaugmented inspections for the accordance with MRP-228baffle-former bolts completed, specificationsduring or before this outage

38 Fall 2028 46.9 Not applicable ... Not applicable. Not applicable .

39 Spring 2030 48.4 Not applicable ..... Not applicable Renewed Operating License expires.7/31/2030

•............... ......... -.......



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

Refueling Project Estimated . .

Outage Month/Year EFPY AMP-Related Scope-1 ) Inspection Method and Criteria Comments

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


WESTINGHOUSE NON-PROPRILETARY CLASS 3 8-1

8 IMPLEMENTING DOCUMENTS

As noted within this RNP AMP document, the PWR Vessel Internals Program is a part of the "ReactorCoolant System Material Integrity Management Program" as documented in ADM-NGGC-01 12 [7]. TheRNP AMP also references the Water Chemistry Program and the ASME Section XI Inservice Inspection,Subsections IWB, IWC, and IWD Program. MRP-227-A augmented examinations (Appendix C),recommended as a result of industry programs, will be included in the existing ASME Section XIprogram.

RNP documents associated with the existing RNP programs and considered to be implementingdocuments of the PWR Vessel Internals Program are:

0 CP-200, Chemistry Program Implementation [15]0 PLP-025, Inservice Inspection Programs [13]0 TMM-038, Inservice Examination Program [14]* EGR-NGGC-0210, ASME Section XI Inservice Inspection Examination Program/Plan

Administration [56]

The PWR Vessel Internals AMP relies on the Water Chemistry Program for maintaining high waterpurity to reduce susceptibility to cracking due to SCC. The Water Chemistry Program was evaluated[2, 22] and found to be consistent with GALL with some exceptions related to augmented inspectionsexpected to be defined through industry programs. Additional procedures may be updated or created asOE for augmented examinations is accumulated.

Based on this information, the updated AMP for RNP 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. 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear PowerPlants."

2. U.S. Nuclear Regulatory Commission, NUREG-1785, "Safety Evaluation Report Related tothe License Renewal of H. B. Robinson Steam Electric Plant, Unit 2," Docket No. 50-261.Carolina Power & Light Company, March 2004.

3. U.S. Nuclear Regulatory Commission, NUREG-1800, U.S. NRC Standard Review Plan forthe Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), April2001.

4. U.S. Nuclear Regulatory Commission, NUREG-1801, "Generic Aging Lessons Learned(GALL) Report," April 2001.

5. NGG Document, RNP-L/LR-0354A, Rev. 3, "Aging Management Review Reactor Vessel."

6. NGG Document, RNP-L/LR-0354B, Rev. 2, "Aging Management Review ReactorInternals."

7. NGG Standard Procedure, ADM-NGGC-01 12, Rev. 5, "Reactor Coolant System MaterialIntegrity Management Program."

8. NGG Document, RNP-L/LR-0614, Rev. 3, "Aging Management Program PWR VesselInternals Program."

9. EGR-NGGC-0504, Rev. 8, "Mechanical System Aging Management Review for LicenseRenewal."

10. EGR-NGGC-0506, Rev. 7, "Civil/Structural Screening and Aging Management Review forLicense Renewal."

1I. Materials Reliability Program: "Pressurized Water Reactor Internals Inspection andEvaluation Guidelines" (MRP-22 7-A). EPRI, Palo Alto, CA: 2011. 1022863.

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

13. RNP Plant Operating Manual, PLP-025, Rev. 22, "Inservice Inspection Programs."

14. RNP Plant Operating Manual, TMM-038, Rev. 19, "Inservice Examination Program."

15. RNP Plant Operating Manual, CP-200, Rev. 17, "Chemistry Program Implementation."

16. NGG Document, RNP-L/LR-0600, Rev. 12, "Aging Management Program Water ChemistryProgram."

17. NGG Document, RNP-L/LR-0606, Rev. 5, "Aging Management Program ASME Section XI,Subsections IWB, IWC, and IWD Inservice Inspection Program."

18. NGG Document, RNP-L/LR-0609, Rev. 2, "Aging Management Flux Thimble Eddy CurrentInspection Program."

19. "Pressurized Water Reactor Primary Water Chemistry Guidelines," Volumes I and 2,Revision 6, EPRI, Palo Alto, CA; 2007, 1014986.


9-2 WESTINGHOUSE:NON-PROPRIETARY CLASS 3

20. U.S. NRC Bulletin 88-09, "Thimble. Tube Thinning in WestinghouseReactors,"

July 26, 1988.

21. EGR-NGGC-0503, Rev. 9, "Mechanical Component Screening for License Renewal."

_22. H. B. Robinson UFSAR, Rev. 23; Section 3.9.5; Section 4.1; Section 18.1.

23. Westinghouse Report, WCAP-12202, "Bottom Mounted Instrumentation Double Wall Flux. Thimble Tube Combination Thimble Tube Wear Program," February 1989.

24. EC56266, Rev. 11, "Reactor Head & Service Structure Replacement."

25. Engineering Evaluation 90-079, Rev. 0, "Control Rode Guide Tube Support PinReplacement," September 20, 1990.

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

27. Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228). EPRI,Palo Alto, CA: 2009.1016609.

28. NGG Standard Procedure, CAP-NGGC-0200, Rev. 34, "Condition Identification andScreening Process."

29. NGG Standard Procedure, CAP-NGGC-0202, Rev. 20, "Operating Experience andConstruction Experience Program."

30. NGG Standard Procedure, CAP-NGGC-0205, Rev. 15, "Condition Evaluation and CorrectiveAction Process."

31. NGG Standard Procedure, CAP-NGGC-0206, Rev. 6, "Performance Assessment.andTrending."

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

33. NGG Program Manual, NGGM-PM-0007, Rev. 21, "Quality Assurance Program Manual."

34. NGG Standard Procedure, PRO-NGGC-0200, Rev. 14, "Procedure Useand Adherence."

35. NGG Standard Procedure, PRO-NGGC-0204, Rev. 22, "Procedure Review and Approval."

36. Westinghouse Report,.WCAP-15028, "Guide Tube Cold-Worked 316 Replacement SupportPin Development Program," March 1998.

37. Letter J.W. Moyer, Carolina:Power and Light Company, to the U.S. Nuclear RegulatoryCommission, Subject: Application for Renewal of Operating License, H.B. Robinson SteamElectric Plant, Unit 2, June 14, 2002. (Serial: RNP-RA/02-0089)

38. Westinghouse Report, WCAP- 1.5664:"Determination of Acceptable Baffle-Barrel-Bolting forThree-Loop Westinghouse 15x 15 Downflow and 17x 17 Standard Upflow Domestic Plants,"December 2001.

39. EC 50826, Rev. 1, "Cut and Cap Flux Thimbles N-5 and N-12 During RO-21."

40. 5379-04731, Rev. 7, "Thimble Assembly - Flux Bottom Mounted Instrumentation."


WESTINGHOUSENON-PROPRIETARY CLASS 3 9-3

41. ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition-with 2008 Addenda. (RNPInternal Update Valid After July 20, 2012).

42'. RNP Plant Operating Manual, TMM-015, Rev. 35, "Inservice Relpair and Replacement."

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

44. U.S. Nuclear Regulatory Commission ML 111990086, "NRC Regulatory Issue Summary2011-07 License Renewal Submittal Information for Pressurized Water Reactor Internals

Aging Management," July 21, 2011.

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

46. Engineering Change, PCHG-CMU, 72122R1, "Replace Upper Internals Splii Pins DuringRO-26. This EC will incorporate the NSSS OEM Replacement Component."

47. Westinghouse Letter, CPL-89-570, "Carolina Power & Light Company, H. B. Robinson Unit2, BMI Thimble Wear Program Reports," May 25, 1989.

48. Progress Energy Letter: David J. Martrano to Daniel C. Beddingfield, "Transmittal of InputDocuments for Reactor Vessel Leak Before Break Analysis and Acceptable Baffle BoltingAnalysis," June 19, 2012.

49. Carolina Power & Light Company Calculation, RNP-M/MECH- 165 , Rev. 12, "ContainmentAnalysis Inputs for H.B. Robinson - Unit NO. 2."

50. Westinghouse Equipment Specification, 676367, Rev. 0, "Reactor Vessel - Reactor Coolant' System," June 8, 1966.

51. Westinghouse Equipment Specification, 676487, Rev. 0, "Reactor Vessel - Reactor CoolantSystemi," February 20, 1967: .

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

53. Westinghouse Report, WCAP-17277-P, Rev. 0, "H.B. Robinson Unit 2 - 15xl 5UpperInternals Guide Tube - Guide Card Wear Evaluation," October 2010.

54. Westinghouse Report, WCAP- 17451 -P, Rev. 0-B, "Reactor Internals Guide Tube Wear -Westinghouse Domestic Fleet Operational Projections," August 2012.

55. PWROG Document, OG-12-176, "Meeting held from April 16-19, 2012 in Las Vegas,Nevada Summary of the Materials Subcommittee," May 9, 2012.

56. NGG Standard Procedure, EGR-NGGC-0210, Rev. 0, "ASME Section XI InserviceInspection Examination Progralm/Plan Administration."

57. Westinghouse Letter, PGN-12-71, Rev. 1, "MRP-227-A Aging Management Program PlanUpdate - Revision 1," August 22, 2012.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1

APPENDIX AILLUSTRATIONS

CONTROL RODDRIVE MECHANISM

UPPER SUPPORTPLATE

INTERNALS-SUPPORT

LEDGE

CORE BARREL

SUPPORT COLUMN -

UPPER COREPLATE

OUTLET NOZZLE

BAFFLE RADIALSUPPORT

BAFFLE

CORE SUPPORTCOLUMNS

INSTRUMENTATIONTHIMBLE GUIDES

RADIAL SUPPORT

CORE SUPPORT

ROD TRAVELHOUSING

INSTRUMENTATIONPORTS

THERMAL SLEEVE

LIFTING LUG

CLOSURE HEADASSEMBLY

HOLD-DOWN SPRING

CONTROL RODGUIDE TUBE

CONTROL RODDRIVE SHAFT

INLET NOZZLE

CONTROL RODCLUSTER (WITHDRAWI

ACCESS PORT

REACTOR VESSEL

LOWER CORE PLATE

Figure A-1 Illustration of a Typical Westinghouse Internals


A-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3A-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3

Wear Area

Figure A-2 Typical Westinghouse Control Rod Guide Card

WCAP-17077-NP August 2012Revision 1

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

UpperGuide Tube

LowerGuide tube

Upper Support

Plate

Sheaths andC-Tubes

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


A-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3A-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3

Flange Weld

Upper Core Barrel toLower Core BarrelCircumferential Weld

Lower BarrelCircumferential Weld

Core Barrel toSupport Plate Weld

, Axial Weld

r BarrelWeld

Lower BarrelAxial Weld

Figure A-4 Major Core Barrel Welds



0000c

0000c

0000c

00000

000co

09994

4)

0fon

Figure A-5 Bolting Systems used in Westinghouse Core Baffles


A-6 WESTINGHOUSE NONIPROPRIETARY CLASS 3A-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3

INTERNALSSUPPORT LEDGE -

THERMAL SHIELD

CORE SUPPORTCOLUMN

CORE SUPPORTFORGING

Figure A-6 Core Baffle/Barrel Structure


WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7

COMMEU EDGE BRhACETBAFFLE TO FORMER 301.T

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


A-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3

Figure A-8 Vertical Displacement between the BaUJ. Platesland Bracket at the Bottom of theBaffle-Former-Barrtl g~sesgIA4



-CORE BARREL

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

Weld

Figure A-10 Typical Thermal Shield Flexure


A-10 WESTINGHOUSE NON-,PROPRIETARY CLASS 3

Core Plate

Lower CoreSupport Structure

Core SupportPlate (Forging)

Figure A-11 Lower Core Support Structure


WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-llIWESTINGHOUSE NON-PROPRIETARY CLASS 3 A-il

LOWER CORE PLATE

DIFFUSER PLATE

CORE SUPPORTPLATE/FORGING

BOTTOM MOUNTEDINSTRUMENTATIONCOLUMN

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

I

Figure A-13 Typical Core Support Column


A-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3A- 12 WESTINGHOUSE NON-PROPRIETARY CLASS 3

. a

/

Figure A-14 Examples of BMI Column Designs


WESTINGHOUSE NONýPROPRIETARY CLASS 3 B-1

APPENDIX BROBINSON LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY TABLES

The content and numerical identifiers in Tables B-I and B-2 of Appendix B are extracted fromTable 3.1-1 (AMPs evaluated in the GALL report that are relied on for license renewal) and Table 3.1-2(evaluations that are different from or not addressed in the GALL report) of the license renewalapplication approved by the NRC.

Table B-1 LRA Aging Management Review Summary Table 3.1-1 Robinson LRA

Component/Commodity ; Aging ManagementGroup"• Aging Effect/Mechanism Program Comments

5. Baffle-Former Assembly Loss of fracture PWR Vessel Internals

Baffle-Former Bolts toughness due to Programneutron irradiationembrittlement and voidswelling

8. Reactor Internals Changes in dimension PWR Vessel Internalsdue to void swelling Program

Baffle-Former Assembly Crack initiation and PWR Vessel Internals

12. Baffle-Former Bolts growth due to SCC and Program and WaterIASCC Chemistry Program

Baffle-Former Assembly Loss of preload due to ASME Section XI,13. Baffle-Former Bolts stress relaxation Subsection IWB,

IWC, and IWDProgramn.nd PWRVessel InternalsProgram

25. Reactor Vessel Internals Loss of fracture PWR Vessel Internals See also Table B-2,CASS Components toughness due to Program Item 14

thermal aging, neutronirradiationembrittlement, and voidswelling

28. Reactor Internals, Loss of material due to ASME Section XI, NRC Bulletin 88-Reactor Vessel Closure wear Subsection IWB, 09 (Eddy current)Studs, and Core Support IWC, and IWDPads Program and Flux

Thimble EddyCurrent InspectionProgram

30. Upper and Lower Loss of preload due to ASME Section XI, See also Table B-2,Internals Assembly stress relaxation Subsection IWB, Item 15

(upper internals hold-down IWC, and IWDspring and lower internals Program and PWRassembly clevis insert bolts) Vessel Internals

Program


B-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3B-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3

Table B-I LRA Aging Management Review Summary Table 3.1-1 Robinson LRA

Component/Commodity Aging ManagementGroup') Aging Effect/Mechanism Program.• Comments

31. Reactor Vessel Internals Loss of fracture PWR Vessel Internalsin Fuel Zone Region (except toughness due to Program and Primarybaffle bolts) neutron irradiation Water Chemistry

embrittlement and voidswelling:

33. Reactor Vessel Internals Crack initiation and PWR Vessel Internals(except baffle former bolts) growth due to SCC and Program and Primary

IASCC Water Chemistry

35. Reactor Internals (upper Loss of preload due to ASME Section XI, See Table B-2,and lower internal stress relaxation Subsection IWB; Item 15assemblies) IWC, and IWD

Program and PWRVessel InternalsProgram.

Note:

1. The numbers contained in this column reflect the identical numbers in the RNP LRA table referenced.

Table B-2 LRA Aging Management Review Summary Table 3.1-2 Robinson LRA

Component/Commodity Aging ManagementGroup"•. .Aging Effect/Mechanism Program "Comments,

14. Reactor Vessel Internals- Reduction of fractu'e - PWR VesselCASS Components toughness from thermal Internals Program . .

embrittlement andneutron irradiationembrittlement

15. Reactor Internals: Loss of-preload due to ASME Section XI,Upper Support Column stress relaxation Subsection IWB,Bolts, Hold-down Spring, IWC, and IWDLower Support Plate Program and PWRColumn Bolts, and Clevis Vessel InternalsInsert Bolts Program "_ _ _

16. Flux Thimbles Cracking from SCC. Primary WaterChemistry

Note:

1. The numbers contained in this column reflect the identical numbers in the RNP LRA table referenced.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-1

APPENDIX CMRP-227-A AUGMENTED INSPECTIONS

Table C-i 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 Visual (VT-3) examination 20% examination of theTube Assembly (Wear) no later than 2 refueling number of CRGTGuide plates (cards) outages from the beginning assemblies, with all guide

of the license renewal cards within each selectedperiod, and no earlier than CRGT assembly examined..two refueling outagesprior See Figure A-2to the start of the license -

renewal period. Subsequent•examinations are required ona ten-year interval.

Control Rod Guide All plants Cracking (SCC,. Bottom-mounted Enhanced visual (EVT-1) 100% of outer (accessible)Tube Assembly. Fatigue) - :instrumentation examination to determine the CRGT lower flange weldLower flange welds Aging "(BMI) column presence of crack-like surfaces and adjacent base

Management (iE bo.dies,;Lower surface flaws in flange welds metal on the individualand TE)..- .support column no later than 2 refueling periphery CRGTb... odies (cast), outages from the beginning assemblies.

- Uipper core plate, of the license renewal period (Note 2)-Lower support and subsequent examination See Figure A-3forging/casting on a ten-year interval.

Core Barrel Assembly All plants Cracking (SCC) Lower support Periodic enhanced visual 100% of one side of theUpper core barrel flange column bodies (EVT-1) examination, no accessible surfaces of theweld (non-cast) later than 2 refueling outages selected weld and adjacent

Core barrel outlet from the beginning of the base-metal (Note 4).nozzle welds license renewal period and See Figure A-4.

subsequent examination on a.- . ten-year interval. -.


C-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3C-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3

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

Effect Expansion Link ExaminationItem Applicability (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 theUpper and lower core IASCC, Fatigue) core barrel (EVT-1) examination, no accessible surfaces of thebarrel cylinder girth cylinder axial later than 2 refueling outages selected weld and adjacentwelds welds from the beginning of the base metal (Note 4).

license renewal period and See Figure A-4subsequent examination on aten-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 outages selected weld and adjacentfrom the beginning of the base metal (Note 4).license renewal period andsubsequent examinations ona ten-year interval.

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 components

results in . and subsequent . accessible from~core side

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

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

Failed or A-7missing boltsiProtrusion ofbolt heads•~Aging,'Management.. E. ...... . ..

and ISRY" '

(Note 6) . _ _ _ _ _ __._....

..C"P" .I•7077-NP August. '2012

WCAP- I17077-NP August 2012Revision I


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

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

Baffle-Former All plants Crhcking Lower support Baseline volumetric (UT) 100% of accessible boltsAssembly (IASCC, column bolts, examination between 25 and (Note 3). Heads accessibleBaffle-former bolts Fatigue) Barrel-former 35 EFPY, with subsequent from the core side. UT

Aging bolts examination on a ten-year accessibility may beManagement (IE interval, affected by complexity of

and ISR) head and locking device

(Note 6) designs.. .. .See Figures A-5 and A-6

Baffle-Former All plants Distortion (Void None Visual (VT-3) examination Core side-surface, asAssembly . Swelling),.or: 'o check for evidence of indicated.Assembly Cracking distortion, withbaseline See Figure A-8(Includes: Baffle plates, (IASCC) that examination between 20.and•baffle edge bolts and results in: 40 EFPY and subsequentindirect effects of void Abnormal examinations on a ten-year:.swelling in former plates) interaction with interval.

fuel assemblies

Gaps along highfluence' bafflejointVerticaldisplacement ofbaffle platesnear highfluence joint "

Broken or• * . ' •damaged edge ". . .:.. bolt locking * "

systems alonghigh fluencebaffle joints -. .



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

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

Alignment and All plants Distortion (Loss None Direct measurement of Measurements should beInterfacing Components with 304 of Load) spring height within three taken at several pointsInternals hold down stainless Note: This cycles of the beginning of around the circumferencespring steel hold mechanism was the license renewal period. If of the spring, with a

down not strictly the first set of measurements statistically adequatesprings identified in the is not sufficient to determine number of measurements at

original list of life, spring height each point to minimize

age-related measurements must be taken uncertainty.

degradation during the next two outages, See Figure A-9mechanisms. in order to extrapolate the

expected spring height to 60years.

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-6 and A-10

(Wear) that renewal period. Subsequentresults in examinations on a ten-yearthermal shield interval.

" . flexures,excessive wear,fracture, orcompleteseparation

.. . .. . . . ..


WESTINGHOUSE NON-PROPRIETARY CLASS 3 . I . . C'-5WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-S

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

I A Effect Expansion Link ExaminationItem Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage

Notes:

I. 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.

i,

- 1< Os,


C-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3

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

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. " _ _ _I

Lower Internals All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessibleAssembly Aging flange weld examination, surfaces (Note 2).Lower support forging or Management Re-inspection every 10 years See Figure A-12.castings (TE in Casting) following initial inspection.

Core Barrel Assembly All plants Cracking Baffle-former Volumetric (UT) 100% of accessible bolts.Barrel-former bolts (iASCc, bolts examination. Accessibility may be

Fatigue) I . Re-inspection every 10 years limited by-presence of

Aging following initial inspection. thermal shields orneutronManagement 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 supportedby plant-Lower support column -Fatigue) Re-inspection every 10 years specific justification (Notebolts Aging following initial inspection. 2).

Management (1E See Figures A- 11, A-12andISR) and A-13

Core Barrel Assembly All plants. Cracking-(SCC, Upper core barrel Enhanced visual (EVT-l) 100% of one side of theCore barrel outlet nozzle Fatigue). flange weld examination. accessible surfaces of thewelds Aging. . Re-inspection every 10 years selected weld and adjacent

Management (LE .- following initial inspection, base metal (Note 2).

of lower .... - See Figure A-4sections)



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

Effect Primary Link Examination MethodItem Applicability (Mechanism) (Note 1) (Note 1) Examination Coverage

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 (IASCC) flange weld examination. surfaces (Note 2).Lower support column Aging Re-inspection every 10 years See Figures A-1l, A- 12,bodies Management, following initial inspection, and A-13(non cast) (IE)

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 l'dwer flanges Re-inspection every 10 years See Figures A-I I, A-12,bodies detection of . . following initial inspection. and A-13(cast) fractured

support columns .

AgingManagement(IE) "

WCAP- 1 7077-NP August 2012Revision I


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

Item Applicability Effect Primary Link Examination Method Examination Coverage(Mechanism) (Note 1) (Note 1) _

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

fractured Re-inspection every 10 years See Figures A- 12 and A-column bodies following initial inspection. 14

Aging Flux thimbleManagement insertion/withdrawal to be

monitored at each inspectioninterval.

Notes:

I. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table C-4.2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination

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



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 atCore barrel flange (Wear) Section Xl 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, f 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. I examination, as defined inFlux thimble tubes plant response to IEB 88-

09.

Alignment and All plants Loss of material ASME Code Visual (VT-3) examination. All accessible surfaces atInterfacing Components (Wear) Section XI specified frequency.Clevis insert bolts (Note 2)

(Not 2)


C-i o WESTINGHOUSE NON-PROPRIETARY CLASS 3

Table C-3 MRP-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 Components (Wear) Section XI specified frequency.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.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-11WESTINGHOUSE NON-PROPRIETARY CLASS 3 c-Il

Table C-4 MRP-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__riteria

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.

WCAP- 17077-NP .August 2012Revision 1


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 Acceptance Criteria

(Note 1)Control Rod Guide All plants Enhanced visual a. Bottom- a. Confirmation of a. For BMI column.Tube Assembly (EVT-1) mounted surface-breaking bodies, the specific

Lower flange welds examination instrumentation indications in two or more relevant condition for

The specific .(BMI) column CRGT lower flange welds, the VT-3 examination is

relevant condition bodies combined with flux completely fracturedis a detectable ýb. Lower support thi*hble 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 weldsshall require EVT-1examination of cast lowersupport column bodies,upper core plate and lowersupport forging/castingswithin threep fuel cyclesfollowing the initidl

. . observation.

WCAP-17077-NP .- August 2012Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 - C-13



(Note 1) Acceptance Criteria

Core Barrel Assembly All plants Periodic enhanced Core barrel outlet a. The confirmed detection a and b. The specific

Upper core barrel flange visual.(EYT- 1) nozzle welds and sizing of a surface- relevant condition forweld examination. Lower support breaking, indication with a the expansion core

column bodies length greater than two barrel outlet nozzle weld

The specific (non cast) inches in the upper core and lower supportrelevant condition barrel flange weld shall column body

require that the EVT-I examination is ais a detectable examination be expanded detectable crack-likecrack-like surface to include the core outlet surface indication.indication, nozzle welds by the

completion of the nextrefueling.outage.

b. If extensive cracking inthe remaining core barreloutlet nozzle welds isdetected, EVT- 1examination shall beexpanded to include the

.. upper six inches of theaccessible surfaces of thenon-cast lower supportcolumn bodies withinthree fuel cycles followthe initial observation.


C- 14 WESTINGHOUSE NON-PROPRIETARY CLASS 3


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

(Note 1) Acceptance CriteriaCore 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 coreT The specific length greater than two barrel cylinder axialrelevant condition inches in the upper core weld examination is a

is a detectable barrel cylinder girth welds detectable crack-likecrack-like surface shall require that the EVT- surface indication.indication. I examination be

expanded to include theupper core barrel cylinderaxial welds by~thecompletion of the nextrefueling outage. .-.

-7....... ....


WESTINGHOUSE NON-PROPRIETARY CLASS 3 . Cý-15


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


Core Barrel Assembly All plants Periodic enhanced Lower core barrel The confirmed detection The specific relevant

Lower core barrel visual (EVT-1) cylinder axial and sizing of a surface- condition for the

cylinder girth welds examination, welds breaking indication with a expansion lower coreThe specific . .length greater than two barrel cylinder axial

relevant condition inches in the lower core weld examination is a

is a detectable barrel cylinder girth welds detectable crack-likecrack-like surface shall require thatthe EVT- surface indication.indication. 1 examination be

expanded to include thelower core barrel cylinderaxial welds by thecompletion of the next -

refueling outage.

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

Baffle-edge bolt edge bolts The specific . . .

relevant conditionsare missing orbrdken lockingdevices, failed ormissing bolts, andprotrusion of boltheads.

WCAP- 1 7077-NP August 2012Revision I

C- 16 WESTINGHOUSE NON-PROPRIETARY CLASS 3




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. examined on the four lower support column

for the UT of the Barrel-former 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 .f... 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.


WESTINGHOUSE NON-PROPRIETARY CLASS 3 .C-17


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

(Note 1) Acceptance CriteriaBaffle-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 near.-high fluence joints',and broken ordamaged. edge boltlocking systemsalong-high fluencebaffle plate joints.






Alignment and All plants Direct physical None N/A N/AInterfacing Components with 304 measurement or

Internals hold down stainless spring height.spring steel hold The examination

down acceptancesprings criterion for this

NOTE: measurement is that

RNP hold the remainingdown spring compressibleis 304 SS, height of the springbut RNP shall provide hold-plans to do a down forces withinreplacement the plant-specificin fall 2013 design 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:

1. 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.
