ML12278A399

WCAP-17077-NP, Rev 1, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant

ML12278A399
Person / Time
Site:	Robinson
Issue date:	08/31/2012
From:	Boggess C, Mckinley J, Szweda K, Basel R Westinghouse
To:	Office of Nuclear Reactor Regulation
References
RNP/RA/12-0100 WCAP-17077-NP, Rev 1
Download: ML12278A399 (101)

Text

Enclosure to Serial: RNP-RA/12-0100 103 Pages (including cover page)

ENCLOSURE PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant

Westinghouse Non-Proprietary Class 3 WCAP-17077-NP August 2012 Revision 1 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Robinson Nuclear Plant Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17077-NP Revision 1 PWR Vessel Internals Program Plan for Aging Management of 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*, Manager Reactor Internals Aging Management

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

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

© 2012 Westinghouse Electric Company LLC All Rights Reserved

WESTINGHOUSE NON-PROPRIETARY CLASS 3 iii TABLE OF CONTENTS L IST O F TA B L E S ........................................................................................................................................ v L IST O F FIG U RE S .................................................................................................................................... vii L IST O F A C R O N YM S ................................................................................................................................ ix ACKNOWLEDGEMENTS ......................................................................................................................... xi P U RPO S E ..................................................................................................................................... 1-1 2 BA C K G R O UN D .......................................................................................................................... 2-1 3 PRO G RA M O W N E R ................................................................................................................... 3-1 3.1 Corporate Nuclear Engineering Services Chief Engineering Section .............................. 3-1 3.2 RN P Engineering ............................................................................................................. 3-1 3.3 RNP Environmental and Chemistry ................................................................................. 3-1 4 DESCRIPTION OF THE H. B. ROBINSON REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS ............................................. 4-1 4.1 Existing RNP Program s ................................................................................................... 4-3 4.1.1 Primary Water Chemistry Program ................................................................. 4-4 4.1.2 ASME Section XI Inservice Inspection Program ............................................ 4-4 4.1.3 Flux Thimble Eddy Current Program .............................................................. 4-4 4.2 Supporting RNP Programs and Aging Management Supportive Plant Enhancements ... 4-5 4.2.1 Reactor Internals Aging Management Review Process ................................... 4-5 4.2.2 Flux Thim ble Tubes ........................................................................................ 4-5 4.2.3 Reactor Head and Control Rod Drive Mechanism Replacement .................... 4-6 4.2.4 Control Rod Guide Tube Split Pin Replacement Project ................................ 4-6 4.3 Industry Program s ............................................................................................................ 4-6 4.3.1 WCAP-14577, Aging Management for Reactor Internals ............................... 4-6 4.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines ........... 4-7 4.3.3 Ongoing Industry Programs ........................................................................... 4-10 4.4 Sum m ary ........................................................................................................................ 4-11 5 H.B. ROBINSON REACTOR INTERNALS AGING MANAGEMENT PROGRAM AT T R IB U T ES .............................................................................................................................. 5-1 5.1 GALL Revision 2 Element 1: Scope of Program ........................................................... 5-1 5.2 GALL Revision 2 Element 2: Preventive Actions .......................................................... 5-3 5.3 GALL Revision 2 Element 3: Parameters Monitored or Inspected ................................ 5-4 5.4 GALL Revision 2 Element 4: Detection of Aging Effects ............................................. 5-5 5.5 GALL Revision 2 Element 5: Monitoring and Trending .............................................. 5-10 5.6 GALL Revision 2 Element 6: Acceptance Criteria ...................................................... 5-11 5.7 GALL Revision 2 Element 7: Corrective Actions ........................................................ 5-12 WCAP- 17077-NP August 2012 Revision I

iv WESTINGHOUSE NON-PROPRIETARY CLASS 5.8 GALL Revision 2 Element 8:, Confirmation Process .................................................... 5-13 5.9 GALL Revision 2 Element 9: Administrative Controls ................................................ 5-14 5.10 GALL Revision 2 Element 10; Operating Experience ................................................. 5-14 6 DEMONSTRATION ..... '...... r .......................... ........................ :6-1 6.1 Demonstration of Topical Report Conditions Compliance to SE on MRP-227, Revision 0 ...................... v ...................... 6-2 6.2 Demonstration of Applicant/Licensee Action Item Compliance to SE on MRP-227, R ev ision 0 ........................................ ........................................................................ ... 6-3 6.2.1 SE Applicant/Licensee Action Item 1: ADplicability of FMECA and Functionality Analysis Assumptions ............................................................... 6-3 6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal .......................................................... 6-5 6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Program s .............................................................................. 6-6 6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief ......................................................................................... 6-6 6.2.5 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&W C om ponents ..................................................................................................... 6-7 6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS M aterials .......................................................................................................... 6-8 6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff Review and A pproval ...................................................................................... 6-9 7 PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE ............................... 7-1 8 IMPLEM ENTING DOCUM ENTS .............................................................................................. 8-1 9 REFE REN C ES ............................................................................................................................. 9-1 APPENDIX A ILLUSTRA TIONS ......................................................................................... A-I APPENDIX B ROBINSON LICENSE RENEWAL AGING MANAGEMENT REVIEW SUM MA RY TA BLES ................................................................................... B-I APPENDIX C MRP-227-A AUGMENTED INSPECTIONS ............................................... C-I WCAP- I 7077-NP August 2012 Revision 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 Implementation S 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 for W 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 in Recommendations for Westinghouse-DeSigned Internals". .............................................. C-9 Table. C4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for W estinghouse-D esigned Internals ........................................................................... C-11

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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 WCAP- 17077-NP August 2012 Revision I

3.; --

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

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

FMECA failure mode, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC irradiation-assisted stress corrosion cracking IGSCC intergranular stress corrosion cracking INPO Institute of Nuclear Power Operations ISI inservice inspection ISR irradiation-enhanced stress relaxation LRA License Renewal Application MRP Materials Reliability Program NDE nondestructive examination NEI Nuclear Energy Institute NES Nuclear Engineering & Services NOS Nuclear Oversight Section NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system OBE operating basis earthquake OE Operating Experience OEM Original Equipment Manufacturer OER Operating Experience Report PH precipitation-hardenable (heat treatment)

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

PWSCC primary water stress corrosion cracking QA quality assurance RCS reactor coolant system RI-FG Reactor Internals Focus Group WCAP- 17077-NP August 2012 Revision I

X WESTINGHOUSE. NON-PROPRIETARY CLASS 3.

LIST OF ACRONYMS-(cont.)

RIS Regulatory Issue Summary RO refueling outage .

RNP H. B. Robinson Steam Electric Plant, Unit Number 2, or Robinson Nuclear Plant RV reactor vessel RVI reactor vessel internals SCC stress corrosion cracking SER Safety Evaluation Report SRP Standard Review Plan SS stainless steel TE thermal embrittlement UT ultrasonic testing (a volumetric NDE method)

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

WOG Westinghouse Owners Group XL Extra-long Westinghouse Fuel Trademark Statement:

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

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. WESTINGHOUSE NON-PROPRIETARY CLASS 3 xi ACKNOWLEDGEMENTS The authors would like to thank the members of the Progres's Enetgy Aging Management Program Team led by David Martrano and Anthony James and our associates at Westinghouse, including Dr. Randy Lott, for their efforts in supporting development of this WCAP.

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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 Program Plan (AMP). The purpose of the AMP is to manage the effects of aging on reactor vessel internals through the license renewal period. RNP entered the license renewal period on August 1, 2010. This document provides a description of the program as it relates to the management of aging effects identified in various regulatory and updated industry-generated documents in addition to the program documented in RNP calculation RNP-L/LR-0614 [8] in support of license renewal program evaluations. This AMP is supported by existing RNP documents and procedures and, as needed by industry experience or directive in the future, will be updated or supported by additional documents to provide clear and concise direction for the effective management of aging degradation in reactor internals components. These actions provide assurance that operations at RNP will continue to be conducted in accordance with the current licensing 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 Section XI Inservice Inspection (ISI) programs [17], and industry requirements [I I]. This AMP fully captures the intent of the additional industry guidance for reactor internals augmented inspections, based on the programs sponsored by U.S. utilities through the Electric Power Research Institute (EPRI) managed Materials Reliability Program (MRP) and the Pressurized Water Reactor Owners Group (PWROG).

The main objectives for the RNP RVI AMP are to:

• Demonstrate that the effects of aging on the RVI will be adequately managed for the period of extended operation in accordance with 10 CFR 54 [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. Participatein industryprograms to investigate aging effects and determine the appropriateAMP activities to addressbaffle andformer assembly issues, and to address changes in dimensions due to void swelling.

2. Evaluate the results of completed researchprojectsfrom the Westinghouse Owners Group (formerly WOG, now PWROG) and the EPRI MRP, andfactor them into the PWR Vessel InternalsProgramas appropriate.

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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, will become part of the RNP ASME B&PV Code,Section XI program [.17]. Corrective actions for augmented inspections willbe deyeloped using repair and replacement procedures equivalent to those requirements in ASME B&PV Code,Section XI, or as determined independently by Progress Energy, or in cooperation with 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 of aging for reactor internals components and establishes the basis for providing reasonable assurance that the internals components will continue to perform their intended function through the RNP license renewal period of extended operation. Revision 0 of this WCAP supported the RNP license renewal commitment to submit to the U.S. Nuclear Regulatory Commission (NRC) an inspection plan for the PWR Vessel .Internals Program, as 'itwould be implemented, from RNP's participation in industry initiatives, 24 months priorto the augmented inspection. Since MRP-227-A was released, the NRC publ.ished.aRegulatory Issue Summary (RIS) that implemented new guidelines as to when plants must submit an inspection plan. According to the NRC RIS [44], because; RNP has submitted their AMP for approval by the NRC, RNP may withdraw their submittal and provide new and revised commitments to the 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].

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

nuclear power industry has been actively engaged in recentyears in a significant effort to support the industry goal of responding to these requirements' Various programs have been underwiy witfiin the industry over the past decade to develop guidelines for managing the effects of aging within PWR reactor' internals. In 1997, the WOG issued WCAP-14577 [26], "License Renewal Evaluation: Aging Management for Reactor Internals," which was reissued as Revision I-A in 2001 after receiving NRC Staff review-and approval. Later, an effort was engaged by the EPRI MRP to address the PWR internals aging management issue for the three currently Operating U.S. reactor designs'- Westinghouse, Combustion Engineering (CE), and Babcock & Wilcox (B&W).

The MRP first established a framework and strategy for the aging management'of PWR internals components 'using proven'and familiar methods for inspection, monitoring, surveillance, and communication. Based upon that framework and strategy, and on the accumulated industry research data, the following elements of an Aging Management Program Were further developed [26, 32]:

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

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

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

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

- Components that are significantly susceptible to the aging effects.

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

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

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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 of reactor internals components requiring inspection, type of NDE .required for each component, timing for initial inspections, and criteria for evaluating inspection results. MRP-227-A provides a standardized approach to PWR internals aging management for each unique reactor design (Westinghouse, B&W, and CE).

MRP-228 [27], "Inspection Standard for Reactor Internals Components," provides guidance on the 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 Data Requirements" for the MRP-227 inspections, where feasible [45]. This document has been submitted to the 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 renewal inspection commitments. In some cases, individual plants will develop plant-specific acceptance criteria for some internals components where a generic approach is not practical.

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

As described in NUREG- 1785 [2], the RNP internals are designed to support, align, and guide the core components 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 obtainnaccess to 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-mating surface and is closely guided at the bottom by radial support/clevis assemblies. The upper internals assembly is clamped at this same ledge by the reactor vessel head. The bottom of the upper internals assembly is closely guided by the core barrel alignment pins of the lower internals assembly.

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 beam sections, support columns, control rod guide tube assemblies, and the upper core plate. Upon upper internals 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 movable detectors for measurement of the neutron flux distribution within the reactor core. Movable miniature neutron flux detectors are available to scan the active length of selected fuel assemblies to provide remote reading of the relative three-dimensional flux distribution. The thimbles are inserted into the reactor core through guide tubes, or conduits, extending from the bottom of the reactor vessel (RV) through the concrete shield area and then up to a thimble seal table. Since the movable detector thimbles are closed at WCAP- 17077-NP August 2012 Revision I

WESTINGHou'st NON-pROPRIETARY CLASS 3 2m3 the leading (reactor) end, they are dryinside. The thimbles thus serve as a pressure barrier between the reactor coolant pressure and the atmosphere. Mechanical seals between the retractable thimbles and the conduits are provided at the seal table.

RNP was granted a license for extended operation by the NRC through the issuance of a SER in NUREG-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 required by 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 in Tables B-I and B-2.

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

continued reliable function. As designated by the protocols of NEI 03-08 [12], "Guidelines for the, Management of Materials Issues," each plant will be required to use MRP-227-A and MRP-228 to develop and implement an AMP for reactor internals no later than three years after the initial industry issuance of MRP-227, Revision 0. MRP-227, Revision 0 was issued in December 2008, and plant-AMPs must therefore be completed by December 2011, or sooner, if required by plant-specific License Renewal Commitments. According to [44], because RNP has.submitted their AMP forapproval by the NRC, RNP may withdraw their submittal and provide new and revised commitments to the NRC to resubmit their AMP 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 functionswill be maintained consistent with the -cufrent'licensing bdis for the: pei'od:of extended operation.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 3-1 3 PROGRAM OWNER The PWR Vessel Internals Program is part of the "Reactor Coolant System Material Integrity Management Program" [7]. The successful implementation and comprehensive long-term management of the 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 are provided in the following paragraphs. Progress Energy will maintain cognizance of industry activities related to PWR internals inspection and aging management, and will address/implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices.

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

Additional responsibilities and the appropriate responsible personnel are discussed in the following subsections.

3.1 CORPORATE NUCLEAR ENGINEERING SERVICES CHIEF ENGINEERING SECTION The Nuclear Engineering Services (NES) Chief Engineering Section (CES) is responsible for providing governance and oversight of the implementation of the PWR Vessel Internals Program, including:

Ensuring that appropriate programs are established and maintained to support inspection and mitigation 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 aging management.

3.2 RNP ENGINEERING RNP Engineering is responsible for the overall development and implementation of the PWR Vessel Internals Aging Management Program, including:

Planning, control, and implementation of the RVI AMP mitigation, inspection, and repair activities, 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 times specified.

3.3 RNP ENVIRONMENTAL AND CHEMISTRY RNP Environmental and Chemistry is responsible for:

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3-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Maintaining primary water chemistry inlaccordance with: approved RNP procedures and specifications. 9-.

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

2'.

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WESTINdHOUSENON-PROPRIETARY CLASS 3 ._47i 4 DESCRIPTION OF THE H. B. ROBINSON REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS The U.S. nuclear industry, through the combined efforts of utilities, vendors, and independent consultants, has defined a generic guideline to assist utilities in developing reactor internals plant-specific aging management programs based on inspection and evaluation. As noted in Section 3 of this AMP, the PWR Vessel Internals Program is a part of the "Progress Energy Reactor Coolant System Material Integrity Management Program" [7]. The intent of this program is to ensure the long-term integrity and safe operation of the reactor internals components. RNP has developed this AMP in conformance with the 10 Generic 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], inspections prescribed by the ASME Section XI Inservice Inspection Program [17], thimble tube inspections [18], and mitigation projects such as split pin replacement [25] and [46], combined with augmented inspections or evaluations as recommended by MRP-227-A.

Aging degradation mechanisms that impact internals have been identified and documented in RNP Aging Management Reviews [5, 6] prepared using the corporate procedural guidance document [9] in support of the License Renewal effort. The overall outcome of the reviews and the additional work performed by the industry, as summarized in MRP-227-A, is to provide appropriate augmented inspections for reactor internals components to provide early detection of the degradation mechanisms of concern. Therefore, this AMP is consistent with the existing RNP AMR methodology and the additional industry work summarized in MRP-227-A. All sources are consistent and address concerns about component degradation resulting from the following eight material aging degradation mechanisms identified as affecting reactor internals:

• Stress Corrosion Cracking (SCC)

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

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

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

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4-.2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Fatigue -

Fatigue is defined as the structural deterioration that can.occur as the result of repeated stress/strain cycles causedby fluctuating loads and temperatures. After repeated cyclic loading of sufficient, magnitude, microstructural damage can accumulate, leading to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical orthermal cyclic loading can lead to growth of the, initiated crack. Corrosion fatigue is included in the degradation description.

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

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

Thermal Aging Embrittlement Thermal aging embrittlement is the exposure of delta ferrite within cast austenitic stainless steel (CASS) and. precipitation-hardenable (PH) stainless steel to high inservice temperatures, which can result in an increase in tensile, strength,, a decrease in ductility, and a loss of fracture toughness. Some degree of thermal aging embrittlement can also occur at no .rma!operating temperatures for CASS and PH stainless steel internals. CASS components have a duplex microstructure 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 is present and the local applied stress intensity exceeds the reduced fracture toughness.

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

• changed. Such changes in mechanical properties include increasing yield strength, increasing ultimate strength, decreasing ductility, and, a loss of fracture toughness. The irradiation embrittlement aging mechanism is a function of both temperature and neutron fluence. . While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity exceeds the reduced fracture toughness.

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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 formation of microscopic cavities in the material These cavities :result from the nucleation and growth of clusters of irradiation-produced vacancies. *-lelium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material. Void swelling may produce dimensional changes that exceed the tolerances on a component. Strain gradients produced by differential swelling in the system may produce significant stresses. Severe swelling

(>5 percent by volume) has been correlated with extremely low fracture toughness values. Also included in this mechanism is irradiation growth of anisotropic materials, which is known to cause significant dimensional changes within ifi-c6re instrumentation tubes that ar6 fabricated from'zirconium alloys. While the initial aging effect is'dimensional change and distortion, severe void swelling may result in cracking under stress.

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

Creep (or more precisely, secondary creep) is a slow, time- and temperature-dependent, plastic deformation of materials that can occur at stres's§e-*els below'the ,ield strength (elaýtic limit).

Creep occurs at elevated temperatures where: ontinuous defdrmaitiori takes place undeit constant strain. Secbndary creep iii austenitic stainless stelels is: associated with temlierature* higher than those 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 can also be affected by void swelling should it occur. The aging effect is a loss of mechanical closure integrity (or'preload) that can lead to unahticipatted loading that, in turn, may eventually cause subsequent 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 management program as described in NUREG-1801, GALL Report Section XI.M16A for PWR Vessel Internals. In the RNP RVI AMP, this is demonstrated through application of existing RNP AMR methodology that credits inspections prescribed by the ASME Section XIllnservice Inspection Program, existing RNP programs, and additional augmented inspections based on MRP-227-A recommendations.'A description of the' applicable existing RNP programs and compliance With the elements of the GALL is contained in the following subsections. .

4.1 EXISTING RNP PROGRAMS RNP's overall strategy for managing aging in reactor internals components is supported by the following existing programs:

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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 the RVI components. Although affiliated with and supporting the RVI AMP, they will be managed under the existing programs.

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

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

The evaluation. of this program against the 10 attributes in the GALL for Program XI.M2 [43] in support of 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 as cracking, loss of preload due to stress relaxation or irradiation creep, loss of material, and reduction of fracture toughness due to thermal embrittlement. For RNP, inspections conducted under the reactor internals AMPs will be controlled as a combination of ASME Section XI ISI exams on core support structures and augmented exams performed under that ISI Program for the remaining reactor internals components addressed within MiRP-227-A. The RNP Section XI, 10-year ISI examinations supporting the 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 support of 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 thimble tube 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 assemblies provide a path for the flux thimbles into the core from the bottom of the vessel and protect the flux thimbles during operation of the reactor. The flux thimble, provides a path for the neutron flux detector into the core and is subject to reactor coolant pressure on the outside and containment pressure on the inside.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3' 4-5 The RNP thimble tube inspection program is an existing plant-specific program that satisfies NRC Bulletin 88-09 requirements that a tube wear inspection procedure be established [18] and maintained for Westinghouse-supplied reactors that use bottom-mounted flux'thimble tube in'strumentAtion. The program follows Branch Technical Position RLSB-1, Aging Management Review - Generic, which is included 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, and corrective 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 support of 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 the requirements 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 procedure directs the use of calculations as the vehicle to implement the license renewal aging management review and to document the results. Calculation RNP-L/LR-0354A [5] and RNP-L/LR-0354B [6] document the results of the aging management review performed in support of RNP license renewal for reactor internals. 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 B Tables B- I and B-2 of this AMP. .  : * .. . .

The calculations supported the' LRA, as follows'!

1. Ideniifiedapplicableaging effects requiringmanagemeni

2. Associated aging'management programsto manage those aging effects

3. Identified enhancements or modifications to existing programs,new aging management programs,or any other actions requiredto support the conclusions reachedin the calculation Aging management reviews were performed for each RNP system that contained long-lived, passive components requiring aging management review, in accordance with the screening process of RNP EGR-NGGC-0503, "Mechanical Component Screening for LicenSe Renewal" [21 ]. This'review is not repeated 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 combination thimble tubes at RNP, and the results are documented in WCAP-12202 [23], which wasprovided in Westinghouse Letter, CPL-89-570, "Carolina Power & Light Company, H. B. Robinson Unit 2, BMI Thimble Wear Program Reports," May 25, 1989 [47]. As a result of the evaluation and subsequent WCAP-17077-NP August 2012 Revision I

4-6 WESTINGHOUSE. NON-PROPRIETARY CLASS 3 operation, a plant modification was performed. The modification consisted of a cut-and-cap operation to secure 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 of thimbles N,5 and N-I12 were removed through the core in a subsequent refueling outage with the final configuration 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 and documented in the industry. As a result of the identification of the existence of the concern at RNP, a program was undertaken to replace the undesirable material with one more resistant to the effects of PWSCC .for the reactor head. A side benefit noted to the selected component replacement option was the introduction of design features to facilitate refueling outage activities. Detailed descriptions of all facets of 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 tube assembly into the top of the core plate. In general, SCC prevention is aided by adherence to strict primary water chemistry limits that effectively prevent SCC and greatly reduce the probability of IASCC. The limits imposed by the Primary Water Chemistry Program at RNP are consistent with the latest EPRI 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 treatments were 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 significantly reduced the susceptibility to IGSCC while maintaining the fatigue and wear requirements necessary to support continued uninterrupted service.

In response to the industry concern, split pins were replaced at RNP during the 1990 RO-.I 3. A second replacement occtqrred in spring.2010 RO-26, which replaced the existing X-750 support pins with the Westinghouse-designed cold-worked Type 316 SS support pins to mitigate the concern for potential SCC inherent with the Alloy X-750 material [46]. Detailed descriptions of all facets of the replacement are retained 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 a technical evaluation of aging degradation mechanisms and aging effects for Westinghouse RVI components. The WOG sent the report to the NRC staff to demonstrate that WOG member plant owners that subscribed to the WCAP could adequately manage effects of aging on RVI during the period of WCAP-1 7077-NP August 2012 Revision I

ME!MNGHOUSENON-PROPRIETARY CLASS 3- 477 extended operation, using approved aging management rhetho6dologies of the WCAP to develop plant-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 includingutility representatives, NSSS vendors, independent consultants, and international committee representatives who reviewed available ddta;and industry experience on materials aging. The objective'of the group was to develop a consistent, systematic approach for identifying and prioritizing inspection and evaluatibn requirements 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 for specific RVI components. MRP-227-A credited existing component inspections, when they were deemed adequate, as a result of detailed expert panel assessments conducted in conjunction with the development of the industry document. Through the elements of the process, the reactor internals for all currently licensed and operating PWR designs in the United States were evaluated in the MRP program; and for reactOr internals were defined.

appropriate inspection, evaluation, and implementation requirements 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 aging mechanisms were placed in the Primary group. The aging management requirements that are needed to ensure functionality of Primary components are described-in the I&E guidelines4 . The Primary group also includes components that have shown a degree of tolerance toa specific aging degradation effect, but for which no highly susceptible component exists or for which no highly susceptible component is accessible. "

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4-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Expansion Those PWR internals that are highly or moderately! susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components depends on the findings from the examinations of the Primary components at individual plants.

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

No Additional Measures Programs Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the No Additional Measures group. Additional components were placed in the No Additional Measures group as a result of a failure mode, effects, and criticality analysis (FMECA) and the functionality assessment. No further action is required by these guidelines for managing the aging of the No Additional Measures components.

The categorization and analysis used in the development of MRP-227-A are not intended to supersede any ASME B&PV Code Section XI requirements. Anycomponents that are classified as core support structures, as defined in ASME B&PV Code Section XI IWB.2500, Category B-N-3, have requirements that remain in effect and may only be altered as allowed by 10 CFR 50.55a.

4.3.2.2 NEI 03-08 Guidance within MRP-227-A ,

The industry program requirements of MRP-227-A are classified in accordance with the requirements of the NEI 03-08 protocols. The MRP-227. 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 programfor management of aging of reactor internalscomponents within thirty-six months following.issuance ofMRP-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 this AMP for RNP to meet its license renewal commitment that pre-dates the implementation date contained in MRP-227, Revision 0.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 4-9 According to the NRC Regulatory Issue Summary (RIS) [44], because RNP has submitted their AMP for approval by the NRC, RNP may withdraw their submittal and provide new and revised commitments to the NRC to resubmit their AMP in accordance with MRP-227-A [11] no later than 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 and Tables 5-1 through 5-3for the applicable design within twenty-four monthsfollowing issuance of MRP-227-A.

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

2. Examinationsspecified in the MRP-227-A guidelines shall be conducted in accordancewith Inspection 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 as applicable where required by MRP-227-A directives. . * . . .

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

RNP Applicability: RNP will comply with this requirement.

4. Each commercial U.S. PWR unit shallprovide a summary reportof all inspections and monitoring, items requiringevaluation, and new repairsto the MRP ProgramManager within 120 days of the completion of an outage during which PWR internals within the scope of MRP-22 7-A are examined.

RNP Applicability: As discussed in Section 4, Progress Energy will participate in future industry efforts 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 the attributes that make up an acceptable AMP. These attributes are consistent with the RNP Aging Management Review process. Evaluation of the RNP RVI AMP against GALL attribute elements is provided in Section 5 of this program plan.

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• August 2012 Revision I

4-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 As part of License Renewal, RNP agreed to participate in industry activities associated with the development of the standard Industry Guideline for Inspection and Evaluation of Reactor Internals, The industry efforts have defined the required inspections and examination techniques for those components critical to aging management of RVI. The results of the industry recommended inspections, as published in MRP-227-A, serve as the basis for identifying any augmented inspections that are required to complete the 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-A assumptions:*

30 years of operationwith high-leakagecore loadingpatterns (freshfuel assemblies loaded in peripherallocations)followed by implementation of a low-leakagefuel management strategyfor the remaining30 years of operation.

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

Base load operation,i.e., typically operates atfixed power levels and does not usually vary power on a calendaror 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 the originalvendors.

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

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

4.3.3 Ongoing Industry Programs The U.S. industry, through both the EPRI/MRP and the PWROG, continues to sponsor activities related to RVI aging management, including planned development of a standard NRC submittal template, development of a plant-specific implementation program template for currently licensed U.S. PWR plants, and development of acceptance criteria and inspection disposition processes. Progress Energy will maintain cognizance of industry activities related to PWR internals inspection and aging management; WCAP- 17077-NP August 2012 Revision 1

WESTINGHQUSE-NON-'PROPRJETARY CLASS 3 4-11 and will address/implement industry guidance, stemming from those activities, as appropriate under NEI 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 a determination of which reactor internals passive components are most susceptible to the aging mechanisms of concern and then determination of the proper inspection or mitigating program that provides reasonable assurance that the component will continue to perform its intended function through the period of extended operation. The GALL-based approach was used at RNP for the initial basis of the LRA that resulted in the NRC SER in NUREG-1785 [2].

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

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

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I, WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-1 5 H.B. ROBINSON REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES The RNP RVI AMP is credited for aging management of RVI components for the following eight aging degradation mechanisms and their associated effects:

• Stress corrosion cracking 0 Irradiation-assisted stress corrosion cracking

• Wear

• Fatigue

• Thermal aging embrittlement 0 Irradiation embrittlement

• Void swelling and irradiation growth 0 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 10 attributes for successful component aging management. The framework for assessing the effectiveness of the projected program is established by the use of the 10 elements of the GALL.

RNP fully utilized the GALL process contained in NUREG- 1801 [4] in performing the aging management review of the reactor internals in the license renewal process. However, RNP made a commitment (see NUREG-1785 [2]) to incorporate the following: (1) RNP will continue to participate in industry programs to investigate aging effects and determine the appropriate AMP activities to address baffle and former assembly issues, and to address change in dimensions due to void swelling, (2) as WOG and EPRI MRP research projects are completed, RNP will evaluate the results and factor them into the PWR Vessel Internals Program as appropriate, and (3) RNP will implement an augmented inspection during the license renewal term. Augmented inspections, based on required program enhancements resulting 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 inspections identified 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 Nuclear Plant, which is built to a Westinghouse NSSS design. The scope of the program applies the methodology and guidance in the most recently NRC-endorsed version of MRP-22 7, which provides augmented inspection andflaw evaluation methodologyfor assuringthe functional integrity ofsafety-related internals in commercial operating U.S. P WR nuclearpower plants designed by B& W, CE, and Westinghouse. The scope of components consideredfor inspection under MRP-22 7 guidance includes core support structures (typically denoted as Examination CategoryB-N-3 by the ASME Code,Section XI), those RVI components that serve an intended WCAP-17077-NP August 2012 Revision I

5-2 WESTTNGHOUSF-.)NON-PROPRIETARY CLAS$ I license renewal safety function pursuantto criteriain 10,CFR 54.4(a) (1), and other R VI.

components whose failure couldprevent satisfactoryaccomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i) , (ii), or (iii). The scope of the program does not include consumable items, such asfuel assemblies, reactivity control assemblies, and nuclear instrumentation,because these components are not typically within the scope of the components that are requiredto .be subject,to an aging management review (AMR), as defined by the criteria set in 10CFR 54.21(a)(1), The scope of the program also does not include welded attachments.

to the internalsurface of the reactorvessel because these components are considered to be ASME Code Class I appurtenancesto the reactor vessel and are adequately managed in accordance with an applicant'sAMP that correspondsto GALL AMP XI.M1, "ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC, and IWD."

The scope of the program includes the response bases to applicable license renewal applicant action items (LRAAIs) on the MRP-22 7 methodology, and any additionalprograms,actions, or activities that are discussed in these LRAAI responses and creditedfor aging management of the applicant'sRVI components. The LRAAIs are identified in the staffs safety evaluation on MRP-227 and include applicable action items on meeting those assumptions thatformed the basis of the MRP's augmented inspection andflaw evaluation methodology (as discussed in Section 2.4 of MRP-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 ofMRP-22 7 specifies applicabilitylimitations to base-loadedplants and the fuel loading management assumptions upon which the functionality analyses were -based. These*.

limitations andassumptions require a determination ofapplicabilityby the applicantfor each reactorand 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 during each 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 in Sections 3.9.5 and 4.1 of the RNP UFSAR.

The RNP RyI subcomponents that required aging management review are indicated in the previously submitted 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 subcomponent passive function(s) and reference(s) to the corresponding AMR table(s) in Section 3 ,of the RNP License Renewal 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 also included 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 to address the component and aging effect during the period of extended operation. The NRC has reviewed WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NOMPROPRIETARY CLASS 3. 5,-3 and approved the aging management strategy presented in the Appendix B tables as documented in the SER 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 and characterizing of the aging effects (cracks, lossof material, loss of preload, etc.) of interest, prescribed frequency of inspection, and examination acceptance criteria. The information provided in MRP-227-A is rooted in the GALL methodology. The basic assumptions of MRP-227-A, Section 2.4 are metby Robinson Unit 2 and are addressed in subsection 4.312.4 of this AMP. The Topical Report Conditions and Applicant/Licensee Action Items provided by the NRC in the SE on MRP-227, Revision 0 [11] are met by RNP and demonstration of compliance is addressed in Section 6.1 for the Topical Report Conditions and in Section 6.2 for the Applicant/Licensee Action Items. The RNP RVI AMP scope is additionally based on previously established and approved GALL Report approaches through application of 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 GALL Report AMP Element Description "The guidance in MRP-227 relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive agingmechanisms (e.g., loss ofmaterial induced by general,pittingcorrosion,crevice corrosion,or stress corrosioncracking or any'ofitsformns

[SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintainedin accordancewith the Water Chemistry Program. The program description,evaluation, and technical 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 existihg program that complies with the requirements of this element. A description and applicability to the RNP reactor 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, chemistry programs are used to control water chemistry for impurities (e.g., dissolved oxygen, chloride, fluoride, and sulfate) that accelerate corrosion. This program relies on monitoring and control of Water chemistry to keep peak levels of various contaminants below the system-specific limits. The RNP PWR Primary' Water Chemistry Program [15, 16] is based on the current, approved revisions of EPRI PWR Primary Water Chemistry Guidelines.

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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 are consistent 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 OR INSPECTED GALL Report AMP Element Description "The program manages the following age-relateddegradationeffects and mechanisms that are applicablein general to the R VI compqnents at the facility: (a) cracking induced by SCC, PWSCC, L4SCC, orfatigue/cyclical loading, (b) loss of materialinduced by wear, (c) loss of fracture toughness induced by either thermal agingor neutron irradiationembrittlement; (d) changes in dimension due to void swelling and irradiationgrowth, distortion,or deflection; and (e)loss ofpreloadcaused by thermal and irradiation-enhanced stress relaxationor creep. For the management of cracking, the program monitors the evidence of surface breaking linear discontinuitiesif a visual inspection technique is used as the non-destructionexamination (NDE) method, orfor relevantflaw presentationsignals if a volumetric UT method is used as the NDE method For the management of loss of material,the program monitorsfor gross or abnormal surface conditions that may. be indicative.of !oss of material occurringin the components. For the management ofloss ofpreload,the program monitorsfor gross surface.conditions.that may.

be indicative of loosening in applicable bolted,fastened, keyed, or pinned connections. The programdoes not directly monitorfor loss offracture toughness that is induced by thermal aging or neutron irradiationembrittlement, or by void swelling and irradiationgrowth; instead,the impact of loss offracture toughness on component integrity is indirectly manqged by using visual or volumetric examination techniques to monitorfor cracking in the components and by applying applicablereducedfracture toughness propertiesin the flaw evaluations if cracking is detected in the components and is extensive enough to warranta supplementalflawgrowth orflaw tolerance evaluation underMRP-227 guidance or ASME Code,Section XI requirements. The program uses physical measurements to monitorfor any dimensional changes due to void swelling, irradiationgrowth, distortion,or deflection.

Specifically, the program implements the parameters monitored/inspectedcriteriafor Westinghouse designed PrimaryComponents in Table 4-3 ofMRP-227. Additionally, the program implements the parametersmonitored/inspectedcriteriafor Westinghouse designed Expansion Components in Table 4-6 of MRP-227. The parameters monitored/inspectedfor Existing Program Componentsfollow.the basesfor referenced Existingprograms,such as the requirementsfor ASME Code Class RVI components in ASME Code,Section XI, Table IWB-2500-1, Examination CategoriesB-N-3, as implemented through the applicant'sASME Code,Section XI program, or the recommendedprogramfor inspecting Westinghouse-designedflux WCAP-1 7077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 - -5 thimble tubes in GALL AMP XIM3 7, "Flux Thimble Tube Inspection." No inspeetions, except for those specified in ASME Code,Section XI, are requiredfor components that are identifiedas requie'ing "No Additional Measure;'",inaccordancewith the analyses reportedin 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 mechanisms on the intended function of the RNP PWR internals components through inspection and condition monitoring activities in accordance with the augmented requirements defined under industry directives as contained in MRP-227-: and ASME Seciion XI. ".

This AMP implements the requirements for the Primary Component inspections from Table 4-3 of MRP-227-A (included in Appendix C of this AMP as Table C-1),' the 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 Existing Component inspections from Table 4-9 of MRP-227-A (included in Appendix C of this AMP as Table C-3). These tables contain requirements to monitor and inspeci the RVI through the period of extended 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 visual examination of components for assessment, signs of degradation, and corrective actions. 'The requirements of MRP- 227-A only augment and do not replace or modify the requirements of ASME Section 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 and the 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 GALL Report AMP Element Description "The detection of agingeffects is covered in two places: (a) the guidance in Section 4 of MRP-227 provides an introductorydiscussion andjustification of the examination methods selectedfor detecting the aging effects ofinterest; and (b) standardsfor examination methods, procedures, andpersonnel are provided in'a companion document, MRP:228. In all cases, well-established methods were selected. These methods include volumetric UT examination methods for detecting flaws in bolting,physical measurementsfor detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) examinationsfor detecting effects rangingfrom general conditibns to detection and sizing of surface-breakingdiscontinuities. Surface examinations may also be used WCAP- 17077-NP August 2012 Revision 1

5-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 as an alternative to visual examinationsfordetection and sizing of surface-breaking discontinuities.

Crackingcaused by SCC, -IASCC,andfatigue is. monitored/inspectedby either VT-1 or EVT-1 examination (for internalsother than bolting) or by volumetric UT examination (bolting). The VT-3 visual methods may be-appliedfor the detection of cracking only when the flaw.tolerance.of the component or affected assembly, as evaluatedfor reducedfracture toughness properties, is known and has been shown to be tolerantof easily detected largeflaws, even :under reduced

.fracturetoughness conditions. In addition, VT-3 examinations are used to monitor/inspectfor

-loss of materialinduced by wear andfor generalaging conditions, such as gross distortion caused by void swelling and irradiationgrowth or by gross effects of loss ofpreloadcaused by thermal and irradiation-enhancedstress relaxationand creep.

In addition, the program adopts the recommended guidance in MRP-22 7for defining the Expansion criteriathat needed to be applied to inspections of PrimaryComponents and Existing Requirement Components andfor expanding the examinations to include additionalExpansion Components. As a result, inspections performed on the RVI components areperformed consistent with the inspectionfrequency and sampling basesfor PrimaryComponents, Existing Requirement Components, and Expansion Components in MRP-227, which have been demonstratedto be in conformance with the inspection criteria,sampling basis criteria,and sample Expansion criteriain Section A. 1.2.3.4 ofNRC Branch Position RLSB-1.

Specifically, the program implements the parameters monitored/inspectedcriteriaand basesfor inspecting the relevant parameterconditionsfor Westinghouse designed Primary Components in Table 4-3 of MRP-22 7 andfor Westinghousle.designed Expansion Components in.Table 4-6 of MRP-22 7.

The program is supplemented.by the followingplant-specifc Primary Component and Expansion Component inspectionsfor the program (as applicable).: for RNP, no additional.Primaryor.

Expansion components are relevant to the scope of aging managementfor the R VI.:

-In addition, in some cases (as defined in MRP-227), physical measurements are used as supplemental techniques to managefor the gross effects of wear, loss ofpreloaddue to stress relaxation,or for changes in dimension due to void swelling, deflection or distortion. The. ...

physical measurements methods applied in accordancewith.this program include no specific.

component at RNP. The hold down spring at RNP isfabricatedfrom Type 304 SS that does require inspection by physical measurement; however, RNP intends to replace the hold down spring in fall 2013. Therefore, RNP will not be requiredto perform physical measurements of the

hold down springper 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 and field-proven through the application of the Section XI ISI Program. Those augmented inspections that are taken from the MRP-227-A recommendations will be applied through use of the MRP-228 Inspection Standard. This AMP implements the augmented inspection requirements of Table 4-3, Table 4-6, and WCAP- 17077-NP August 2012 Revision 1

WESTMHOUSENONý,PROPRIETARY CLASS 3 5-7 Table 4-9 from MRP-227-A forthe Primary, Ex'pansion,'and Existing Components, respectively. These are included in Appendix C of this AMP for reference. These tables include the inspection frequency and sampling bases. For the Expansion Components of MRP-227-A, this AMP implements the expansion requirements of Table 5-3 of MRP-227-A (included in Appendix C of this AMP as Table C-4).

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

Inspection standards developed by the industry for the application of these techniques for augmrented reactor 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 support structures and interior attachments to reactor vessels) and B-N-3 (removable core support structures) are defined 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 the allowable linear flaw standards of IWB-3510

• -Structural degradation of attachment welds such that the original cross-sectional area: is reduced by 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 intent is to ensure that the structure will meet minimum flaw tolerance'fracture requirements. In the MRP-227-A recommendations, VT-I examinations have been identified for components requiring close visual examinations with some estimate of the scale of deformation or wear. In MRP-227-A note that VT-I has only been selected to detect distortion as evidenced by small gaps between the upper-to-lower mating surfaces of CE-welded core shrouds assembled in two vertical sections. Therefore, no additional VT-I inspections over and above those required by ASME Section XI ISI have been specified.

EVT-I Enhanced Visual Examination for the Detection of Surface Breaking Flaws In the augmented inspections detailed in the MRP-227-A for reactor internals, the EVT- I enhanced visual examination has been identified for inspection of components where surface-breaking flaws are a potential concern. Any visual inspection for cracking requires a reasonable expectation that the flaw length and crack mouth opening displacement meet the resolution requirements of the observation technique. The EVT-I specification augments the VT-I requirements to provide more rigorous inspection standards for stress corrosion cracking and has been demonstrated for similar inspections in boiling water reactor (BWR) internals. Enhanced visual examination (i.e., EVT-I) is also:conducted in accordance with the requirements described for visual examination (i.e., VT-I) with additional WCAP-17077-NP August 2012 Revision I

5-8 WESTINGHOUSE NON-,PROPRIETARY CLASS 3 requirements (such as camera scanning speed) currently being developed by the industry. Any recommendation for EVT-1 inspection will require additional analysis to establish flaw-tolerance criteria, which must take into account potential embrittlement due to thermal aging or neutron irradiation. The industry, through the *PWROG, has developed an approach for acceptance criteria methodologies to support plant-specific augmented examinations. This work is summarized in WCAP- 17096-NP, "Reactor Internals Acceptance Criteria Methodology and.Data Requirements" [45]. The acceptance criteria developed using these methodologies maybe created on either a generic or plant-specific basis because both loads and component dimensions may vary from plant-to-plant within a typical PWR design.

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

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

• 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 rod motion or could result in blockage of coolant flow through fuel; Corrosion or erosion that reduces the nominal section thickness by more than 5 percent; Wear of mating surfaces that may lead to loss of function;

• Structural degradation of interior attachments such that the original cross-sectional area is .

reduced more than 5 percent.

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

WCAP- I7077-NP August 2012 Revision I

WESTIN614OUSE NON PROPRIETARY CLASS I 5-9 Ultrasonic Testing .

Volumetric examinations in the form of ultrasonic testing (UT) techniques can be used to identify and determine the length and depth of a crack in a component. Although access to the surface of the component is required to apply the ultrasonic signals, the flaw may exist in the bulk of the material. In this proposed strategy,.UT inspections have been recommendedexclusively for detection of flaws, in.

bolts. For the bolt inspections, any bolt with a detected flaw should be assumed to have failed. The size of the flaw in the bolt is not.critical because crack growth rates are generally high, and itsis assumed that the observed flaw will result in failure prior to the next inspection opportunity. It has generally been observed through examination performance demonstrations that UT can reliably (90 percent or greater reliability) detect flaws that reduce the cross-sectional area of a bolt by 35 percent.

Failure of a single bolt does not compromise the function of the entire assembly. Bolting systems in the reactor internals are highly redundant. For any system of bolts, it is possible to demonstrate multiple minimum 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 continuation of 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 intended function of the component. This analysis is highly plant-specific. Therefore, any recommendation for UT inspection of bolts assumes that the plant owner will work with the designer to establish minimum' acceptable bolting patterns prior to the inspection to support continued operation. For Westinghouse designed plants, minimum acceptable bolting patterns for baffle-fornier: and barrel-former bolts are available 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 by various degradation mechanisms such as wear or loss of functionality as a result of loss of preload or material deformation. For RNP, direct physical measurements are required only for:the hold down spring; however, RNP is planning to replace the current hold down spring in fall 2013. Therefore, RNP will 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.

WCAP- I7077-NP August 2012 Revision I

5-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 5.5 GALL REVISION,2 ELEMENT 5: MONITORING AND TRENDING GALL Report AMP Element Description "The methods for monitoring, recording,evaluating,and trending the data that resultfrom the program's inspections are given in Section 6 of MRP-227 and its subsections. The evaluation methods include recommendationsforflaw depth sizing andfor crackgrowth determinations as wellfor performing applicable limit load, linear elastic and elastic-plasticfracture analyses of relevantflaw. indications. The examinations and re-examinationsrequired by the MRP-22 7 guidance, together with the requirementsspecified in MRP-228for inspectionmethodologies, inspectionprocedures, and inspectionpersonnel,provide timely detection, reporting,and corrective actions.with respect to the effects ofthe age-relateddegradationmechanisms within the scope ofthe program. The extent of the examinations, beginning with the sample of susceptible PWR internals component locations identifiedas PrimaryComponent locations, with the potentialfor inclusion of Expansion Component locations if the effects are greaterthan

• anticipated,plus the continuation of the Existing Programsactivities, such as the ASME Code,Section XI, Examination Category.B-N-3 examinationsfor core support structures,provides a high degree of confidence in the totalprogram" [43.].,

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

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

Appendix C, Tables C-1, C-2, and C-3 identify the augmented Primary andExpansion inspection and monitoring recommendations, and the Existing programs credited for inspection and aging management.

As discussed in MRP-227-A, inspection of the "Primary" components provides reasonable assurance for demonstrating component current capacity to perform the intended functions. Table C-4 in Appendix C identifies the MRP-227-A expansion criteria from the Primary components. If these expansion criteria are met for a component, the associated Expansion component is to be inspected to manage the aging degradation.

WCAP- 17077-NP August 2012 Revision I

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

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

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

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

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

. "0 'For visual examination (andsurface examihation'asan alternativeto visual examination),

the examination acceptance criterion is the absehbe& of any of the specif descriptive relevant conditions; inhaddition,there are reqidir-mneht&'to-recordand dispositionsurface' breaking indicationsthat are detected'and sizedfor length by VT-1/E VT-1 examinations, '

%For volumetric examination, the examination acceptance criterionis the capabilityfor reliable detection ofindications in bolting, as demonstratedin the examination Technical

Justification;in addition,there are requirementsfor system-level assessment of bolted or pinnedassemblies with unacceptable volumetric (UT) examination indications that exceed specified limits, and Forphysicalmeasurements, the excamination acceptancecriterionfor the acceptable '

tolerance in the measured differential heightfrom the top of the plenum rib pads to the vessel seatingsurface in B& Wplants are given in Table 5-1 ofMRP-22 7.' The acceptance criterion for physical measurementsperformed on the height limits of the Westinghouse-designedhold-down springs are requiredfor 304 SS hold down springs. RNP is planning to replace the

• hold down spring in Fall2013; therefore, RNP is not requiredto produce acceptance criteria for the physical measurements on the hold down spring" *[43].

• WCAP- 17077-NP August 2012 Revision I

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 program scope are evaluated in accordance with the, applicable requirements of the ASME Code through the existing Corrective Action Program [28]..

Inspection acceptance andexpansion criteria are provided in Appendix C, Table C-4. These criteria will be reviewed periodically as the industry continues to develop and refine the information and will be included in updates to RNP procedures to enable the examiner to identify examination acceptance criteria considering state-of-the-art information and techniques .

Augmented inspections, as defined by the MRP-227-A requirements included in this AMP as Appendix C, Table C-1,.Table C-2, and Table C-3, that result in recordable relevant conditions will be entered into the plant Corrective Action Program and addressed by appropriate actions that may include enhanced inspection, repair, replacement,. mitigation actions, or analytical evaluations. An example of an analytical evaluation, is using a minimum bolting WCAP approach such as those commonly used to support continued component or assembly functionality. Additional analysis to establish acceptable bolting pattern evaluation criteria for the baffle-former bolt assembly, as contained in various industry documents

[38], is also considered in determining the acceptance of inspection results to support continued component or assembly functionality.

The industry, through various cooperative efforts, is working to construct a consensus set of tools in line with accepted and proven methodologies to support this element. One of these tools is the PWROG document WCAP417096-NP, "Reactor Internals Acceptance Criteria Methodology and Data, -

Requirements,'" [45], which details acceptance criteria methodology for the MRP-227 Primary and Expansion'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 GALL Report AMP Element Description "Corrective actionsfollowing the detection of unacceptable conditions are fundamentally providedfor in each plant's corrective actionprogram. Any detected conditions that do. not satisfy the examination acceptance criteriaare requiredto be dispositionedthrough the plant corrective actionprogram, which may require.repair,replacement, or analyticalevaluationfor continued service until the next inspection. The. dispositionwill ensure that design basisfunctions of the reactorinternals components will continue to befulfilled for all licensing basis loads and events. Examples of methodologies that can be used to analyticallydisposition unacceptable conditions arefound in the ASME Code,Section XI or in Section 6 of MRP-22 7. Section 6 of MRP-227 describes the options that are availablefor dispositionof detected conditions that WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 5-13 exceed the examination acceptance criteriaof Section 5 of the report. These include engineering' evaluation methods, as well as supplementary examinations tofurther characterizethe detected

• condition, or the alternativeof component repairand replacementprocedures. The latterare

.subject to the requirements of the ASME Code,Section XI. The implementation of the guidance in MRP-22 7, plus the implementationof any ASME Code requirements,provides an acceptable level ofaging management ofsafety-relatedcomponents addressedin accordancewith the corrective actions of 10 CFR Part50, Appendix B or its equivalent, as applicable.

- *. l  : '.

Other alternative corrective action bases may be used to disposition relevant conditions if they.

have been previously approved or endorsedby the NRC. Examples ofpreviously NRC-endorsed alternativecorrective actions bases include those corrective actions basesfor Westinghouse-design RVI components that are defined in Tables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of..

Westinghouse Report No. WCAP-14577-Rev. ]-A, orfor B& W-designed RVI components in B& W Report No. BA W-2248. Westinghouse Report No. WCAP-14577-Rev. 1-A was endorsedfor use in an NRC SE to the Westinghouse Owners Group, dated February10; 2001. B& WReport No.

BA W-2248 was endorsedfor use in an SE to FramatomeTechnologies on behalfofthe B& W, Owners Group, dated December 9, 1999. Alternative corrective action bases not approved or endorsed by the NRC will be submittedfor NRC approvalprior to their implementation " [43].

RNP Corrective Action The existing RNP .procedure for corrective actions, the "Condition Evaluation and Corrective Action Process," [30] and the ASME Section XI ISI program [17], will be credited for this element. These procedures 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 to performing 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 GALL Report AMP Element Description "Site quality assuranceprocedures, review and approvalprocesses, and administrativecontrols are implemented in accordancewith the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable.It is expected that the implementation of the guidance in MRP-227 will provide an acceptable level of qualityfor inspection,flaw evaluation, and other elements of aging management of the PWR internals that are addressedin accordance'withthe 10 CFR Part 50, Appendix B, or their equivalent (as applicable), confirmationprocess, and administrative controls" [43].

WCAP- 17077-NP August 2012 Revision I

5-14 WESTINGHOUSE NON 7 PROPRIETARY CLASS 3 RNP Confirmation Process.-

RNP has an established 0.CFR Part 50, Appendix B, Program [33, 34, 35] that addresses the elements of corrective 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 controls are implemented in accordance with the requirements of 10 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 GALL Report AMP Element Description "The administrativecontrolsfor such programs, including their implementingproceduresand review and approvalprocesses,are under existing site 10 CFR 50 Appendix B QualityAssurance Programs,or their,equivalent, as applicable.Such a program is thus expected to be established with a sufficient level of documentation and administrativecontrols to ensure effective long-term implementation" [43].

RNPAdministrative Controls RNP has an established 10 CFR 50, Appendix.,B Program [33, 34,3*5] that addresses the elements.of corrective 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.of 10 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 GALL Report AMP Element Description "Relativelyfew incidents ofPWR.internals agingdegradationhave been reported,in operating U.S. commercial PWR plants. A summary of observationsto date is provided in Appendix A of MRP-227-A. The applicant is expected to review subsequent operatingexperiencefor impact on its program or to participatein industry initiativesthat perform thisfunction.

WCAP-17077-NP August 2012 Revision I

WESTINGHOUSE NqON-PROPRIETARY CLASS 3 5-15 The applicationof the MRP-22 7 guidance will establish a considerable'amountof operating' experience over the next few years. Section 7 ofMRP-227 describes the reportingrequirements f6r these applications,and the planfobr evaluating the accuwhulaiedadditionaloperating experience" [43].

RNP Operating Experience Extensive industry and RNP operating experience has been reviewed during the development of the Reactor Vessel Internals Aging Management Program. The experience reviewed includes NRC Information Notices 84-18, "Stress Corrosion Cracking in PWR Systems" and 98-11, "Cracking of Reactor Vessel Internal Baffle Former Bolts in.Foreign Plants." Most of the industry-operating experience reviewed has involved cracking ofaustenitic stainless steel baffle-former bolts or SCC of high-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 in plant historical records. Inspections performed as part of the 10-year IS Iprogram have been conducted as designated by existing commitments and would be expected to discover overall general internals structure degradation. To date, very little degradation has been observed industry-wide.

Industry operating experience is routinely reviewed by Progress Energy engineers using INPO Operating Experience (OE), the Nuclear Network, and other information sources as directed under the applicable procedure [29], for the determination of additional actions and lessons learned. These insights, as applicable, can be incorporated into the plant systems quarterly health reports and further evaluated for incorporation into plant programs.

A review of industry and plant-specificexperiencewith re~ictorvessel internals reveals thatthe U.S.: industry, including Progress Energy and RNP,- has responded proactively to industry issues relative to reactor internals degradation. -Two examples that demohstrate this proactive response are the replacement of control rod guide tube split pins in 1990 and 2010, as well as participation by Progress Energy in the augmented examinations performed by the PWROG on control rod guide tube guide cards in 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 new pins were considered a replacement in kind and an improvement over the previous pins. Ongoing industry experience has' shown continued ddgradation of this cbmponent and, in response, RNP performed a second split pin replacement activity in spring 2010 RO-26. The replacement included a material upgrade 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 Cards Inspection Prop-ram The PWROG has conducted upper internals control rod guide tube guide card wear measurements on a sample of guide tubes from selected representative pilot plants to approximate the remaining life of the guide tube guide cards. This was a proactive effort by the U.S. industry to establish criteria for inspection WCAP- I7077-NP August 2012 Revision I

5-16 WESTTNGHOUSE NON-PROPRIETARY CLASS 3 and gather data to support aging management of the component. RNP.p-oactively inspectedguide cards in 1990 RO-13, and was a key participant as one of the representative pilot inspection plants. The RNP inspection in support of the industry program. was completed in spring 2010 RO-26, with an evaluation of the guide card wear being completed in WCAP-.17277-P [53]. Inspection outcomes were evaluated to ensure compliance with MRP-227 specifications. The.PWROG pilot program has been completed, and a draft WCAP [54], which gives the generic inspection criteria, guidelines and recommendations 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 vessel components. Progress Energy, through its participation in PWROG and EPRI-MRP activities, will continue to benefit from the reporting of inspection information and will share its own operating experiencewith the industry through the reportingrequirements of Section 7 of MRP-227-A. RNP presented the results of its initial MRP-227, Revision 0 inspection at the April 2012 PWROG Materials Subcommittee (MSC) meetings [55]. The collected information from MRP-227-A augmented inspections will benefit the industry.in its continued response to RVI aging degradation.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.MI6A [43] and Commitment 33 in the RNP SER.

• August 2012 WCAP- I 7077-NP WCAP-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 reactor internals. This AMP is' based on an established history of programs to identify and nmonitor potential aging degradation in the reactor internals. Programs and activities undertaken in the course of fulfilling that commitment include:

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

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

Review of Nuclear Oversight Section (NOS) audit reports, NRC inspection reports, and INPO evaluations indicate no unacceptable issues related to 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 more resistant Alloy X-750 materials, with modified geometry and heat treatment to increase resistance to SCC (versus original pins). The new Type 316 SS split pins that were used in the replacement in spring 2010 will provide additional resistance to PWSCC.

Replacement of the Type 304 SS hold down spring is planned for spring 2013. The replacement spring will be fabricated from a modified Type 403 SS, which is more resistant to stress relaxation 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 PWR internals inspection and aging management; and will address/implement industry guidance, stemming from those activities, as appropriate under NEI 03-08 practices.

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

Typical ASME Section XI examinations identified in the AMP for the period of extended operation were performed at RNP in spring 2012 RO-27. The augmented inspections discussed in compliance with MRP-227-A requirements have been integrated in the inspection procedures, which were used to perform WCAP-1 7077-NP August 2012 Revision I

6-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 the ASME Section XI, 10-year ISI examinations.. Integration of therequired inspections will be tracked to completion. As discussed, the industry MRP-227-A. guidelines also provide for updates as experience is gained through inspection results. This feedback loop will enable updates based on actual inspection experience..

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

Table 6-1 lists the seven topical report conditions and Section 6.2 lists the.eight applicant action items that cameout of the NRC review of MRP-227, as listed in [11], as well as their compliance within this AMP.

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

_ _ _'_ _ _ Applicable

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

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

3. Inspection of high consequence Not Not applicable components subject to multiple Applicable degradation 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, the former 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 every components 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 16A Appendix A from GALL Revision 2 [43].

WCAP-1 7077-NP August 2012 Revision I

WESTINGHOUSE NON-'PROPMETARY CLASS 3 6 6.2 DEMONSTRATION OF APPLICANT/LICENSEE ACTION ITEM COMPLIANCE TO SE ON MRP-227,:REVISION 0 6.2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions "As addressedin Section 3.2.5.1 of this SE, each applicant/licenseeis respbnsibleforassessing its plant's design and operatinghistory anddemonstratingthat the approved version'ofMRP-22 7 is applicableto the facility. Each applicant/licenseeshallrefer, in'paiticular,to the assumptions regardingplant design and operatinghistory made in the FMECA andfunctionality analysesfor reactors of their design (i.e.; Westinghouse, CE, or B& W) which support MRP-227 and describe the process usedfor determiningplant-specific differences in the design oftheir RVI components or plant operatingconditions, which result in different component inspection categories. The*

applicant/licenseeshall submit this evaluationfor NRC review and approvalas partof its application to implement the approved version ofMRP-227. This is Applicant/Licensee Action Item I" [I11].

RNP Compliance The process used to verify that RNP is reasonably represented by the generic industry program assumptions with regard to neutron fluence, temperature, materials, and stress values used in the development of MRP-227-A 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 PWR interihai components.

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

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

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

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

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

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

WCAP- 17077-NP August 2012 Revision I

6-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3

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

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

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

a., The FMECA and functionality analyses for MRP-227-A were based on the assumption of 30 years of operation with high-leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy. In fuel cycle C16 (November 12, 1993) [48] at 23 years of operation, RNP switched to use of a low-leakage core design. It is projected that RNP will continue to use a low-leakage core design for all subsequent fuel cycles, as discussed in subsection 4.3.2.4. Therefore, RNP meets the fluence and fuel management assumptions in MRP- 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 in MRP 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 design temperature for the vessel is 650'F [50, 51]. RNP operating history is within original design basis parameters, and therefore consistent with the assumptions use8l to ýdevelop eo the MRP-227-A aging management strategy with regard to temperature operational parameters.

3. RNP internal components and materials are comparable to the typical Westinghouse PWR internal 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 those materials 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 the component is already credited as being managed under an alternate RNP aging management program.

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

4. Modifications to the RNP reactor internals made over the lifetime of the plant are those specifically directed by Westinghouse, the Original Equipment Manufacturer (OEM) as discussed in subsection 4.3.2.4. The design has been maintained over the lifetime of the plant as specified by the OEM, operational parameters with regard to fluence and WCAP- I7077-NP August 2012 Revision I

WESTINGHOUSE NONIPROPRIETARY CLASS 3 6-5

• temperature are compliant with MRP-227-A-requirements and the components and materials are the same as those considered in MRP-191. Therefore, the RNP stress values are represented by the assumptions in MRP-191, MRP-232, and MRP-227-A, confirming the applicability of the generic FMECA.

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

6.2.2 SE Applicant/LicenseeAction Item 2: PWR Vessel Internal Components within the Scope of License Renewal "As discussed in Section 3.2.5.2 of this SE, consistent with the requirementsaddressed in 10 CFR 54.4, each applicant/licenseeis responsiblefor identifying which RVI components are within the scope of LR for itsfacility. Applicants/licenseesshall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-1 91 and identify whether these tables contain all of the R VI components that are within the scope of LR for theirfacilities in accordancewith 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for itsfacility, the applicantor licensee shall identify the missing component(s) andpropose any necessary modifications to the program defined in MRP-22 7, as modified by this SE, when submitting its plant-specificAMP. The AMP shallprovide assurance that the effects of agingon the missing component(s),will be managedfor the periodof extended operation. This issue 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 renewal for RNP to those components contained in MRP-191, Table 4-4. A detailed tabulationof the RNP RVI components [57] was completed and compared favorably to the typical Westinghouse PWR internal components in MRP- 191. All components required ter be included in the RNP program [2] are consistent with those contained in MRP- 191. Several components have different materials than specified but these have no effect on the recommended MRP aging strategy or are already managed by an alternate RNP program; therefore, no modifications to the program detailed in MRP-227-A need to be proposed. This supports the requirement that the AMP shall provide assurance that the effects of aging on the RNP RVI components within the scope of license renewal, but not included in Table 4-4 will be managed for the period of extended operation.

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

WCAP-17077-NP August 2012 Revision I

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 meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.

6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs "As addressedin Section 3.2.5.3 in this SE, gpplicants/licenseesof CE and Westinghouse are requiredto perform plant-specific analysis either to justify the acceptabilityofan applicant 's/licensee's existingprograms, or to identify changes to the programsthat should be implemented to manage the aging of these componentsfor the period of extended operation. The results of thisplant-specific analyses and a descriptionof the plant-specificprograms being relied on to manage aging of these components shall be submitted as part of the applicant's/licensee'sAMP application. The CE and Westinghouse components identifiedfor this type ofplant-specific evaluation include: CE thermal shieldpositioningpins and CE in-core instrumentationthimble tubes (Section 4.3.2 in MRP-22 7), and Westinghouse guide tube support pins (splitpins) (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 in Appendix C, Table C-3. This is detailed in the plant-specific RNP program documents for ASME Section XI [ 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 meets the requirement for application of MRP-227-A as a strategy for managing age-related material degradation in reactor internals components.,

6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief "As discussed in Section 3.2.5.4 of this SE, the B& W applicants/licenseesshall confirm that the core support structure upperflange weld was stress relieved during the originalfabricationof the Reactor Pressure Vessel in order.to confirm the applicabilityof MRP-227, as approved by the NRC, to theirfacility. If the upperflange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component. If necessary, the examination methods andfrequencyfor non-stress relieved B& W core support structure upperflange welds shall be consistent with the recommendations in MRP-227, as approved by the NRC, for the Westinghouse and CE upper core support barrelwelds. The examination coveragefor this B& W flange weld shall conform to the staff's imposed criteriaas describedin Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee'sresolution of this plant-specificaction item shall be submitted to the NRC for review and approval.This is Applicant/Licensee Action Item 4" [11 ].

WCAP-17077-NP August 2012 Revision I

. 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 as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components "As addressedin Section 3.3.5 in this SE, applicants/licenseesshall identify plant-specific acceptance criteriato be applied when performing the physicalmeasurements requiredby the NRC-approved version ofMRP-227for loss ofcompressibilityfor Westinghouse hold down springs, andfor distortion in the gap between the top and bottom core 'shroudsegments in CE units with core barrelshrouds assembled in two vertical sections. The applicant/licenseeShall include its proposedacceptance criteriaand an explanation of how the proposedacceptance criteriaare consistent with the plants' licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions ofoperation during the period of extended operationas 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 fall 2013 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 down spring. RNP plans to replace their hold down spring in fall 2013; therefore, Licensee/Applicant Action Item 5 is not applicable to RNP.

6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components "As addiessed in Section 3.3.6 in this SE, MRP-22 7 does not propose to inspect thefollowing inaccessiblecomponents: the B& W core barrelcylinders (including vertical and circumferential seam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices, B& W core barrel-tb-formerbolts and their locking devices, and B& W core barrelassembly inter"nal baffle-to-baffle bolts. The MRP also identifiedthat although the B& W core barrel assembly internalbaffle-to-baffle bolts are accessible; the bolts are non-inspectable using currently available examination techniques:

Applicants/licenseesshalljustify the acceptabilityof these components for continued operation through the period ofextended operation by performing an evaluation, or by proposing a WCAP-17077-NP August 2012 Revision I

6-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 scheduled replacement of the components. As part of theirapplication to implement the approved version of MRP-227, applicants/licenseesshall provide theirjustificationfor the continued operability ofeach of the inaccessiblecomponents and, if necessary,provide theirplanfor the replacementof the componentsfor NRC review and approval.This is Applicant/Licensee Action Item 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/licenseesof B& W, CE, and Westinghouse reactors are requiredto develop plant-specific analyses to be appliedfor 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 periodof extended operationorfor additionalR VI components that may befabricatedfromCASS, martensiticstainless steel or precipitationhardened stainless steel materials. These analyses shall also consider the possible loss offracture toughness in these components due to thermal and irradiationembrittlement, and may also need to consider limitations on accessibilityfor inspection and the resolution/sensitivityof the inspection.

techniques. The requirement may not apply to components that were previously evaluated as not

.requiringaging management duringdevelopment ofMRP-22-7. That is, the requirement.would...

apply to components fabricatedfrom susceptible materialsfor which an individual,licensee has, determined aging management is required,for example during their review performed in accordancewith Applicant/LicenseeAction Item 2. The plant-specific analysis shall beconsistent with the plant's licensing basis and the need to maintain thefunctionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licenseeshall include the plant-specific analysis as part of theirsubmittal to apply the approvedversion of MRP-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 for assessment of CASS materials, the applicant/licensee for license renewal may apply the criteria in the NRC letter of May 19, 2000, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," [52] as the basis for determining whether the CASS materials are susceptible to the thermal aging mechanism. If the application of the applicable screening criteria [52] for the component's material demonstrates that the components are not susceptible to either thermal embrittlement (TE) or irradiation embrittlement, or the synergistic effects of TE and irradiation embrittlement combined, then no other evaluation would be necessary. The RNP CASS components and the assessment of their susceptibility to TE are summarized in Table 6-2.

WCAP- I7077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 6-9 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 composition under TE susceptibility thresholds was not demonstrated for the CASS bases of the upper support columns; thus, it is conservatively assumed that they are potentially susceptible to TE. Under MRP-191 the bases were designated as CASS, screened-in for TE, SCC, and IE, and dispositioned for these mechanisms under the FMECA. Thus, the conservative assumption that the upper support column bases are susceptible to TE has already been addressed in the development of the MRP-227-A inspection requirements.

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

The results of this CASS evaluation do not conflict with the MRP-227-A strategy for aging management of RVI. It is concluded that continued application of the strategy of MRP-227-A will meet the requirement 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 TE Molybdenum (Based on the NRC CASS Component Content Casting Ferrite Content

• Criteria 152])

Flow mixer devices, Low 0.5 max Static _<20% Not susceptible to TE with and without thermocouple Upper support column Low 0.5 max

• Static Possible>20% Potentially susceptible Bases, 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, and special 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/licenseesshall make. a submittalfor NRC review and approvalto credit their implementation of MRP-22 7, as amended by this SE, as an AMP for the RVI components at theirfacility. This submittal shall include the information identified in Section 3.5.1 of this SE. This is Applicant/Licensee Action Item 8" [11].

WCAP- I7077-NP August 2012 Revision I

6-10 WESTINGHOUSE NON-PROPRIETARY CLASS 3 RNP Compliance RNP has completed the initial submittal of their AMP, and according to [44], because RNP has submitted their AMP for approval by the NRC, RNP may withdraw their submittal and provide new and revised commitments to the NRC to resubmit their AMP in accordance with MRP-227-A [11] no later than October 1, 2012.

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

WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 7-1 7 PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE The requirements of MRP-227-A are based on an 18-month refueling cycle and consider both EFPY and cumulative operation. The information contained in Table 7-1 is based on this information and includes a description of the past inspections, as well as the latest scope of 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 Estimated Outage 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 was augmented inspection for inspections in accordance with completed at RNP during RO-26 guide 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 interval augmented inspections for MRP-228 specifications Inspections were completed at RNP control rod guide tube lower during RO-27 flange welds, upper core barrel flange weld, baffle-edge bolts, baffle-former assembly, and thermal shield flexures 28 Fall 2013 31.9 Initial MRP-227-A augmented MRP-227-A inspections in RNP plans to do a preemptive inspections for upper and accordance with MRP-228 replacement of the hold down spring lower core barrel cylinder girth specifications instead of MRP-227-A inspections welds, lower core barrel flange weld, hold down spring, and baffle-former bolts completed during 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.

WCAP- 17077-NP August 2012 Revision I

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

Refueling Project Estimated Outage Month/Year EFPY AMP-Related Scope(') Inspection Method and Criteria Comments 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 applicable augmented inspection for accordance with MRP-228 guide 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 lower flange welds, upper core barrel flange weld, baffle-edge bolts, baffle-former assembly, and thermal shield flexures 35 Spring 2024 42.4 Subsequent MRP-227-A MRP-227-A inspections in Not applicable augmented inspections for accordance with MRP-228 upper and lower core barrel specifications cylinder girth welds, lower core barrel flange weld, and hold down spring completed during 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 applicable augmented inspections for the accordance with MRP-228 baffle-former bolts completed, specifications during 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 WCAP- 17077-NP August 2012 Revision 1

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

Refueling Project Estimated . .

Outage Month/Year EFPY AMP-Related 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.

WCAP- I 7077-NP August 2012 Revision 1

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 "Reactor Coolant System Material Integrity Management Program" as documented in ADM-NGGC-01 12 [7]. The RNP 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 XI program.

RNP documents associated with the existing RNP programs and considered to be implementing documents 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 water purity 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 inspections expected to be defined through industry programs. Additional procedures may be updated or created as OE for augmented examinations is accumulated.

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

WCAP- I7077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 9-1 9 REFERENCES

1. 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."

2. U.S. Nuclear Regulatory Commission, NUREG-1785, "Safety Evaluation Report Related to the 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 for the Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), April 2001.

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 Reactor Internals."

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

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

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

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

1I. MaterialsReliability Program: "PressurizedWater Reactor Internals Inspection and Evaluation Guidelines" (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.

12. NEI 03-08, Rev. 2, "Guideline for the Management of Materials Issues," Nuclear Energy Institute, 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 Chemistry Program."

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 Current Inspection Program."

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

WCAP- I7077-NP August 2012 Revision I

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 Pin Replacement," September 20, 1990.

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

27. MaterialsReliability Program: Inspection Standardfor PWR Internals (MRP-228). EPRI, Palo Alto, CA: 2009.1016609.

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

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

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

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

32. Materials Reliability Program: Screening, Categorizationand Ranking of Reactor Internals Componentsfor Westinghouse and Combustion EngineeringPWR 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 Support Pin Development Program," March 1998.

37. Letter J.W. Moyer, Carolina:Power and Light Company, to the U.S. Nuclear Regulatory Commission,

Subject:

Application for Renewal of Operating License, H.B. Robinson Steam Electric Plant, Unit 2, June 14, 2002. (Serial: RNP-RA/02-0089)

38. Westinghouse Report, WCAP- 1.5664:"Determination of Acceptable Baffle-Barrel-Bolting for Three-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."

WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSENON-PROPRIETARY CLASS 3 9-3

41. ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition-with 2008 Addenda. (RNP Internal 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 ML111990086, "NRC Regulatory Issue Summary 2011-07 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 Criteria Methodology and Data Requirements," December 2009.

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

47. Westinghouse Letter, CPL-89-570, "Carolina Power & Light Company, H. B. Robinson Unit 2, BMI Thimble Wear Program Reports," May 25, 1989.

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

49. Carolina Power & Light Company Calculation, RNP-M/MECH- 165 , Rev. 12, "Containment Analysis 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 Coolant Systemi," February 20, 1967: .

52. U.S. Nuclear Regulatory Commission Letter, "License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of CastAustenitic Stainless Steel Components," May 19, 2000. (NRC ADAMS Accession No: ML003717179).

53. Westinghouse Report, WCAP-17277-P, Rev. 0, "H.B. Robinson Unit 2 - 15xl 5Upper Internals 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 Inservice Inspection Examination Progralm/Plan Administration."

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

WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-1 APPENDIX A ILLUSTRATIONS ROD TRAVEL HOUSING CONTROL ROD INSTRUMENTATION DRIVE MECHANISM PORTS THERMAL SLEEVE UPPER SUPPORT PLATE LIFTING LUG INTERNALS- CLOSURE HEAD SUPPORT ASSEMBLY LEDGE HOLD-DOWN SPRING CORE BARREL CONTROL ROD GUIDE TUBE SUPPORT COLUMN -

CONTROL ROD DRIVE SHAFT UPPER CORE PLATE OUTLET NOZZLE INLET NOZZLE BAFFLE RADIAL CONTROL ROD SUPPORT CLUSTER (WITHDRAWI BAFFLE CORE SUPPORT ACCESS PORT COLUMNS INSTRUMENTATION REACTOR VESSEL THIMBLE GUIDES RADIAL SUPPORT CORE SUPPORT LOWER CORE PLATE Figure A-1 Illustration of a Typical Westinghouse Internals WCAP-17077-NP August 2012 Revision I

A-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Wear Area Figure A-2 Typical Westinghouse Control Rod Guide Card WCAP-17077-NP August 2012 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-3 Upper Guide Tube Upper Support Plate Lower Guide tube Sheaths and C-Tubes Figure A-3 Lower Section of Control Rod Guide Tube Assembly WCAP-1 7077-NP August 2012 Revision I

A-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-4 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Flange Weld , Axial Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld r Barrel Weld Lower Barrel Circumferential Weld Lower Barrel Axial Weld Core Barrel to Support Plate Weld Figure A-4 Major Core Barrel Welds WCAP- 17077-NP August 2012 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-5 0000c 0000c 0000c 00000 000co 09994 4) 0f on Figure A-5 Bolting Systems used in Westinghouse Core Baffles WCAP- 17077-NP August 2012 Revision 1

A-6 WESTINGHOUSE NONIPROPRIETARY CLASS 3 A-6 WESTINGHOUSE NON-PROPRIETARY CLASS 3 INTERNALS SUPPORT LEDGE -

THERMAL SHIELD CORE SUPPORT COLUMN CORE SUPPORT FORGING Figure A-6 Core Baffle/Barrel Structure WCAP-17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7 WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-7 COMMEU EDGE BRhACET BAFFLE TO FORMER 301.T Figure A-7 Bolting in a Typical Westinghouse Baffle-Former Structure WCAP-17077-NP August 2012 Revision I

A-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Figure A-8 Vertical Displacement between the BaUJ. Platesland Bracket at the Bottom of the Baffle-Former-Barrtl g~sesgIA4 WCAP- 17077-NP August 2012 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-9

-CORE BARREL Figure A-9 Schematic Cross-Sections of the Westinghouse Hold Down Springs Weld Figure A-10 Typical Thermal Shield Flexure WCAP- 17077-NP August 2012 Revision 1

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

Figure A-11 Lower Core Support Structure WCAP- 17077-NP August 2012 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-llI WESTINGHOUSE NON-PROPRIETARY CLASS 3 A-il LOWER CORE PLATE DIFFUSER PLATE CORE SUPPORT PLATE/FORGING BOTTOM MOUNTED INSTRUMENTATION COLUMN Figure A-12 Lower Core Support Structure - Core Support Plate Cross-Section I

Figure A-13 Typical Core Support Column WCAP-17077-NP August 2012 Revision I

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

/

. a Figure A-14 Examples of BMI Column Designs WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NONýPROPRIETARY CLASS 3 B-1 APPENDIX B ROBINSON LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

TABLES The content and numerical identifiers in Tables B-I and B-2 of Appendix B are extracted from Table 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 renewal application approved by the NRC.

Table B-1 LRA Aging Management Review Summary Table 3.1-1 Robinson LRA Component/Commodity ; Aging Management Group"* Aging Effect/Mechanism Program Comments

5. Baffle-Former Assembly Loss of fracture PWR Vessel Internals Baffle-Former Bolts toughness due to Program neutron irradiation embrittlement and void swelling

8. Reactor Internals Changes in dimension PWR Vessel Internals due to void swelling Program Baffle-Former Assembly Crack initiation and PWR Vessel Internals

12. Baffle-Former Bolts growth due to SCC and Program and Water IASCC Chemistry Program Baffle-Former Assembly Loss of preload due to ASME Section XI,

13. Baffle-Former Bolts stress relaxation Subsection IWB, IWC, and IWD Programn.nd PWR Vessel Internals Program

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, neutron irradiation embrittlement, and void swelling

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 IWD Pads Program and Flux Thimble Eddy Current Inspection Program

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 IWD spring and lower internals Program and PWR assembly clevis insert bolts) Vessel Internals Program WCAP-1 7077-NP August 2012 Revision I

B-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 B-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table B-I LRA Aging Management Review Summary Table 3.1-1 Robinson LRA Component/Commodity Aging Management Group') Aging Effect/Mechanism Program.* Comments

31. Reactor Vessel Internals Loss of fracture PWR Vessel Internals in Fuel Zone Region (except toughness due to Program and Primary baffle bolts) neutron irradiation Water Chemistry embrittlement and void swelling:

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 15 assemblies) IWC, and IWD Program and PWR Vessel Internals Program.

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 Management Group"*. .Aging Effect/Mechanism Program "Comments,

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

embrittlement and neutron irradiation embrittlement

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

16. Flux Thimbles Cracking from SCC. Primary Water Chemistry Note:

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

WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-1 APPENDIX C MRP-227-A AUGMENTED INSPECTIONS Table C-i MRP-227-A Primary Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Expansion Link Examination Item

• Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Control Rod Guide All plants - -Loss of Material None Visual (VT-3) examination 20% examination of the Tube Assembly (Wear) no later than 2 refueling number of CRGT Guide plates (cards) outages from the beginning assemblies, with all guide of the license renewal cards within each selected period, and no earlier than CRGT assembly examined..

two refueling outagesprior See Figure A-2 to the start of the license -

renewal period. Subsequent*

examinations are required on a 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 weld Lower 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 individual and TE)..- .support b... column odies (cast), no later than 2 refueling periphery CRGT outages from the beginning assemblies.

- Uipper core plate, of the license renewal period (Note 2)

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

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

WCAP- 17077-NP August 2012 Revision I

C-2 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-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 Examination Item 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 the Upper and lower core IASCC, Fatigue) core barrel (EVT-1) examination, no accessible surfaces of the barrel cylinder girth cylinder axial later than 2 refueling outages selected weld and adjacent welds welds from the beginning of the base metal (Note 4).

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

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

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

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

Baffle-edge bolts edge bolts Fatigue) that between 20 and 40 EFPY 100% of components results in . and subsequent . accessible from~core side Lost or broken " examinations on-a ten-year -(Note 3).

locking devices interval. See Figures A-5, A-6, and Failed or A-7 missing boltsi Protrusion of bolt heads

• ~Aging,'

Management.. E. ...... . ..

and ISRY" '

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

C"P" .. .I*7077-NP '2012 August.

WCAP- I17077-NP August 2012 Revision I

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

Effect Expansion Link Examination Item iApplicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Baffle-Former All plants Crhcking Lower support Baseline volumetric (UT) 100% of accessible bolts Assembly (IASCC, column bolts, examination between 25 and (Note 3). Heads accessible Baffle-former bolts Fatigue) Barrel-former 35 EFPY, with subsequent from the core side. UT Aging bolts examination on a ten-year accessibility may be Management (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, as Assembly . 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 subsequent indirect effects of void Abnormal examinations on a ten-year:.

swelling in former plates) interaction with interval.

fuel assemblies Gaps along high fluence' baffle joint Vertical displacement of baffle plates near high fluence joint "

Broken or

• damaged ' edge "

. . .:.. bolt locking "

• systems along high fluence baffle joints -. .

WCAP- 17077-NP August 2012 Revision I

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

Item Applicability Effect Expansion Link Examination Item____Applicability____ (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Alignment and All plants Distortion (Loss None Direct measurement of Measurements should be Interfacing Components with 304 of Load) spring height within three taken at several points Internals hold down stainless Note: This cycles of the beginning of around the circumference spring steel hold mechanism was the license renewal period. If of the spring, with a down not strictly the first set of measurements statistically adequate springs 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-9 mechanisms. in order to extrapolate the expected spring height to 60 years.

Thermal Shield All plants Cracking None Visual (VT-3) no later than 2 100% of thermal shield Assembly with thermal (Fatigue) or refueling outages from the flexures.

Thermal shield flexures shields Loss of Material beginning of the license See Figures A-6 and A-10 (Wear) that renewal period. Subsequent results in examinations on a ten-year thermal shield interval.

" . flexures

,excessive wear, fracture, or complete separation WCAP- 17077-NP August 2012 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 . I . . C'-5 WESTINGHOUSE 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 Examination Item 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 for inspection credit.

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

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

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

i,

- 1< Os, WCAP- 17077-NP August 2012 Revision I

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 Examination Item Applicability (Mechanism) (Note 1) Method/Frequency (Note 1) Examination Coverage Upper Internals All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessible Assembly (Fatigue, Wear) flange weld examination, surfaces (Note 2).

Upper Core Plate Re-inspection every 10 years following initial inspection. " _ _ _I Lower Internals All plants Cracking CRGT lower Enhanced visual (EVT-1) 100% of accessible Assembly 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 orneutron Management pads (Note 2).

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

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

Management (1E See Figures A- 11, A-12 andISR) and A-13 Core Barrel Assembly All plants. Cracking-(SCC, Upper core barrel Enhanced visual (EVT-l) 100% of one side of the Core barrel outlet nozzle Fatigue). flange weld examination. accessible surfaces of the welds Aging. . Re-inspection every 10 years selected weld and adjacent Management (LE .- following initial inspection, base metal (Note 2).

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

WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-7 Table C-2 MRP-227-A Expansion Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Effect Primary Link Examination Method Item 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 accessible Assembly (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 accessible Assembly (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 .

Aging Management (IE) "

WCAP- 17077-NP August 2012 Revision I

C-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-8 WESTINGHOUSE NON-PROPRIETARY CLASS 3 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 column Instrumentation System (Fatigue) tube (CRGT) of BMI column bodies as bodies for which difficulty Bottom-mounted including the lower flanges indicatedby difficulty of is detected during flux instrumentation (BMI) detection of insertion/withdrawal of flux thimble column bodies completely thimbles. insertion/withdrawal.

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

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).

WCAP- 17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-9 Table C-3 MRP-227-A Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Core barrel flange (Wear) Section Xl to determine general specified frequency.

condition for excessive wear.

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

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

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

(Note 1) Management (IE)

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

Lower core plate XL lower core plate (Note 1) .. .,.

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

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

Clevis insert bolts (Note (Not 2) 2)

WCAP- I7077-NP August 2012 Revision 1

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-Designed Internals (cont.)

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

Upper core plate alignment pins Notes:

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

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

WCAP-17077-NP August 2012 Revision I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 C-11 WESTINGHOUSE NON-PROPRIETARY CLASS 3 c-Il Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1) Acceptance__riteria Control Rod Guide All plants Visual (VT-3) None N/A N/A Tube Assembly Examination Guide plates (cards) The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

WCAP- 17077-NP 2012 Revision 1

C-12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link (s) Expansion Criteria Acceptance Criteria (Note 1)

Control Rod Guide All plants Enhanced visual a. Bottom- a. Confirmation of a. For BMI column

.Tube Assembly (EVT-1) mounted surface-breaking bodies, the specific Lower flange welds examination instrumentation indications in two or more relevant condition for The specific .(BMI) column CRGT lower flange welds, the VT-3 examination is relevant condition bodies combined with flux completely fractured is a detectable ýb. Lower support thi*hble column bodies.

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

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

indications in two or more CRGT lower flange welds shall require EVT-1 examination of cast lower support column bodies, upper core plate and lower support forging/castings within threep fuel cycles following the initidl

. . observation.

WCAP-17077-NP .- August 2012 Revision 1

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

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (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 for weld 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 support relevant condition barrel flange weld shall column body require that the EVT-I examination is a is a detectable examination be expanded detectable crack-like crack-like surface to include the core outlet surface indication.

indication, nozzle welds by the completion of the next refueling.outage.

b. If extensive cracking in the remaining core barrel outlet nozzle welds is detected, EVT- 1 examination shall be expanded to include the

.. upper six inches of the accessible surfaces of the non-cast lower support column bodies within three fuel cycles follow the initial observation.

WCAP- I 7077-NP August 2012 Revision 1

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

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

weld (Note 2) examination.

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

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

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

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

WCAP- 17077-NP August 2012 Revision I

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

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

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

refueling outage.

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

Baffle-edge bolt edge bolts The specific . . .

relevant conditions are missing or brdken locking devices, failed or missing bolts, and protrusion of bolt heads.

WCAP- 17077-NP August 2012 Revision I

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

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1) Acceptance Criteria Baffle-Former All plants Volumetric (UT) a. Lower support a. Confirmation that more a and b. The Assembly examination, column bolts than 5% of the baffle- examination acceptance Baffle-former bolts The examination former bolts actually criteria for the UT of the acceptance criteria b. 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 lower support column bolts within the next three fuel cycles.

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

WCAP- I7077-NP August 2012 Revision 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 .C-17 Table C-4 MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination ExamiationAdditional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Additinal Eaitio (Note 1) Acceptance Criteria Baffle-Former All plants. Visual (VT-3) None N/A N/A Assembly examination.

Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, vertical displacement of shroud plates near.

-high fluence joints',

and broken or damaged. edge bolt locking systems along-high fluence baffle plate joints.

WCAP- 17077-NP August 2012 Revision 1

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

Examination Additional Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Acceptance Criteria (Note 1) Acceptance Criteria Alignment and All plants Direct physical None N/A N/A Interfacing Components with 304 measurement or Internals hold down stainless spring height.

spring steel hold The examination down acceptance springs criterion for this NOTE: measurement is that RNP hold the remaining down spring compressible is 304 SS, height of the spring but RNP shall provide hold-plans to do a down forces within replacement the plant-specific in fall 2013 design tolerance.

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

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

Notes:

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.

WCAP- 17077-NP August 2012 Revision 1