ML15226A229

PWR Vessel Internals Program Plan for Aging Management of Reactor Internals

ML15226A229
Person / Time
Site:	Farley
Issue date:	07/31/2015
From:	Carpenter B, Szweda K, Mary Walsh Westinghouse
To:	Office of Nuclear Reactor Regulation
References
NL-15-1507 WCAP-18011-NP, Rev. 0
Download: ML15226A229 (109)

Text

Enclosure ito NL-l 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 WCAP-18011--NP July 2015 Revision 0 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at J.M. Farley Nuclear Plant Unit 1 Westinghouse El-i

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 WCAP-18011-NP Revision 0 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at J.M. Farley Nuclear Plant Unit 1 Bradley T. Carpenter*

Reactor Internals Aging Management Mary Ann T. Walsh*

Reactor Internals Aging Management Karli N. Szweda*

Reactor Internals Aging Management July 2015 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, USA

© 2015 Westinghouse Electric Company LLC All Rights Reserved WCAP- 1801 1-NP El1-2

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 i TABLE OF CONTENTS LIST OF TABLES.................................................................................................. iv LIST OF FIGURES.................................................................................................. v LIST OF ACRONYMS............................................................................................. vi 1 PURPOSE................................................................................................. 1-1 2 BACKGROUND ......................................................................................... 2-1 3 PWR VESSEL INTERNALS PROGRAM OWNER................................................. 3-1 3.1 SNC -EXECUTIVE........................................................................... 3-1 3.2 SNC -CORPORATE.......................................................................... 3-1 3.3 SNC -FNP SITE ................................................................................ 3-3 3.4 PWR PRIMARY SYSTEM INTEGRITY PROGRAM TECHNICAL TEAM ........... 3-4 4 DESCRIPTION OF THE FARLEY NUCLEAR PLANT UNIT 1 REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS ........................ 4-1 4.1 EXISTING FARLEY NUCLEAR PLANT UNIT 1 PROGRAMS ........................ 4-4 4.1 .1 Water Chemistry Control Program................................................ 4-4 4.1.2 Inservice Inspection Program...................................................... 4-4 4.2 SUPPORTING FARLEY NUCLEAR PLANT UNIT 1 PROGRAMS AND AGING MANAGEMENT SUPPORTIVE PLANT ENHANCEMENTS .......................... 4-5 4.2.1 Reactor Internals Aging Management Review Process ......................... 4-5 4.2.2 Reactor Vessel Internals Program ................................................. 4-5 4.2.3 Flux Detector Thimble Inspection Program...................................... 4-5 4.2.4 Control Rod Guide Tube Support Pin Replacement Project .................... 4-6 4.2.5 Power Uprating Project ............................................................ 4-6 4.3 INDUSTRY PROGRAMS..................................................................... 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

SUMMARY

................................................................................... 4-11 5 FARLEY NUCLEAR PLANT REACTOR INTERNALS AGING MANAGEMENT PROGRAM ATTRIBUTES ........................................................................................... 5-1 5.1 GALL REVISION 2 ELEMENT 1: SCOPE OF PROGRAM ............................. 5-1 5.2 GALL REVISION 2 ELEMENT 2: PREVENTATIVE 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 ELEMENTS5: 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.......................... 5-13 5.8 GALL REVISION 2 ELEMENT 8: CONFIRMATION PROCESS ..................... 5-14 5.9 GALL REVISION 2 ELEMENT 9: ADMINISTRATIVE CONTROLS ................ 5-14 5.10 GALL REVISION 2 ELEMENT 10: OPERATING EXPERIENCE..................... 5-15 6 DEMON STRATION..................................................................................... 6-1 6.1 DEMONSTRATION OF TOPICAL REPORT CONDITIONS COMPLIANCE TO SE ON MRP-227, REVISION 0.................................................................. 6-3 WCAP-1801 1-NP July 2015 El1-3 Revision 0

Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 iii 6.2 DEMONSTRATION OF APPLICANT/LICENSEE ACTION ITEM COMPLIANCE TO SE ON MRP-227, REVISION 0 .............................................................. 6-4 6.2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions.............................................. 6-4 6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal..................................................... 6-6 6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs........................................................ 6-7 6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief ................................................................ 6-8 6.2.5 SE Applicant/Licensee Action Item 5: Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI Components ..

.. .......................................................................... 6 -9 6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components......................................................................... 6-9 6.2.7 SE Applicant/Licensee Action Item 7: Plant-Specific Evaluation of CASS Materials ........................................................................... 6-10 6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff Review and Approval............................................................. 6-14 7 PROGRAM ENHANCEMENT AND IMPLEMENTATION SCHEDULE ....................... 7-i 8 IMPLEMENTING DOCUMENTS ..... .............................................................. 8-i 9 REFERENCES .......................................................................................... 9-i APPENDIX A ILLUSTRATIONS............................................................................. A-i1 APPENDIX B FARLEY UNIT 1 LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

TABLE .......................................................................... B-I1 APPENDIX C MRP-227-A AUGMENTED INSPECTIONS............................................... C-i WCAP-I88011-NP ,July 2015 El1-4 Revision 0 Ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 6-1. Topical Report Condition Compliance to SE on MRP-227 ...................................... 6-3 Table 6-2. Summary of Joseph M. Farley Unit 1 CASS Components and Their Susceptibility to TE...6-12 Table 7-1. Aging Management Program Enhancement and Inspection Implementation Summary......7-1 Table B-i. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA ..... B-i Table C-i1. MRP-22 7-A Primary Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals ............................................................ C-i Table C-2. MVIRP-227-A Expansion Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals ............................................................ C-7 Table C-3. MRP-22 7-A Components Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals................................. C- 10 Table C-4. MIRP-22 7-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals........................................................... C- 12 WCAP-1801 1-NP July 2015 El-5 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 V LIST OF FIGURES Figure A-i. Illustration of Typical Westinghouse 4-Loop Plant Internals Assembly ...................... A-i Figure A-2. Typical Westinghouse Control Rod Guide Card................................................. A-2 Figure A-3. Typical Lower Section of Control Rod Guide Tube Assembly ................................ A-3 Figure A-4. Major Core Barrel Welds.......................................................................... 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-Former-Barrel Assembly ...................................................................... A-8 Figure A-9. Schematic Cross-Sections of the Westinghouse Hold Down Springs......................... A-8 Figure A- 10. Typical Thermal Shield Flexure ................................................................ A-9 Figure A-il. Lower Core Support Structure ................................................................. A- 10 Figure A- 12. Lower Core Support Structure - Core Support Plate Cross-Section.i ..................... A-il1 Figure A- 13. Typical Core Support Column ................................................................. A-il1 Figure A- 14. Examples of BMI Column.Designs ........................................................... A- 12 WCAP- 1801 1-NP July 2015 El1-6 Revision 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 vi 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 CFR Code of Federal Regulations._

CLB -- current licensing basis CRGT control rod guide tube ECP Engineering Change Package EFPY effective full-power years EPRI Electric Power Research Institute ET electromagnetic testing (eddy current)

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

FMECA failure modes, effects, and criticality analysis FNP Farley Nuclear Plant GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC irradiation-assisted stress corrosion cracking IE irradiation embrittlement.

INPO Institute of Nuclear Power Operations ISI inservic¢ inspection ISR irradiation-enhanced stress relaxation LRA License Renewal Application LRAAI license renewal applicant action items MRP Materials Reliability Program NDE nondestructive examination NEI Nuclear Energy Institute NOS Nuclear Oversight Section NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system OE Operating Experience OEM Original Equipment Manufacturer OER Operating Experience Report PH precipitation-hardenable (heat treatment)

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

PWSCC primary water stress corrosion cracking WCAP-18011-NP July 2015 El-7 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 vii LIST OF ACRONYMS (cont.)

QA quality assurance RCS reactor coolant system RIS Regulatory Issue Summary RO refueling outage RV reactor vessel RVI reactor vessel internals SCC stress corrosion cracking SE Safety Evaluation SER Safety Evaluation Report SNC Southern Nuclear Company SRP Standard Review Plan SS stainless steel TE thermal embrittlement UFSAR Updated Final Safety Analysis Report UT ultrasonic testing (a volumetric NDE method)

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

WCAP Westinghouse Commercial Atomic Power WOG Westinghouse Owners Group XL Extra-long Westinghouse Fuel Trademark Statement:

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

WCAP-1 801 1-NP July 20150 El1-8 Revision

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 1-1 1 PURPOSE The purpose of this report is to document the Joseph M. Farley Nuclear Plant Unit 1, hereafter referred to as Farley Nuclear Plant (FNP) Unit 1, Reactor Vessel (RV) 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. FNP Unit 1 enters the license renewal period on June 25, 2017. 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 Southern Nuclear Company (SNC) Procedure NMP-ES-029-GL02 [1] in support of license renewal program evaluations. This AMP is supported by existing FNP Unit 1 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 FNP Unit 1 will continue to be conducted in accordance with the current licensing basis (CLB) for the reactor vessel internals by fulfilling License Renewal commitments [2], U. S. Nuclear Regulatory Commission (NRC) expectations in the Regulatory Issue Summary (RIS) [3], American Society of Mechanical Engineers (ASMVE) Boiler and Pressure Vessel (B&PV) Code Section XI Inservice Inspection (ISI) programs [4] and industry requirements [5].

This AMP fully captures the intent of the additional industry guidance for reactor internals augmented inspections, based on the programs sponsored by U. S. utilities through the Electric Power Research Institute (EPRI)-managed Materials Reliability Program (MRP) and the Pressurized Water Reactor Owners Group (PWROG).

The main objectives for the FNP Unit 1 RVL 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 the Code of Federal Regulations, Title 10, Part 54 (10 CFR 54) [6].

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

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

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

FNP Unit 1 License Renewal Commitment 6 [2], "FNP Reactor Vessel Internals Program" commits FNP Unit i to:

1. Implement the FNP Reactor Vessel Internals Programpriorto entering the period of extended operation;

2. Participatein industry initiatives intended to clarify the nature and intent of aging mechanisms potentially affecting the FNP reactor internals;

3. Incorporatethe results of these initiatives into the RVI Program; and, WCAP- 1801 1-NP July 2015 El1-9 Revision 0

Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 1-2

4. Submit an inspection planfor the RVI Program for NRC review and approval at least 24 months prior to entering the periods of extended operationfor the FNP units.

Augmented inspections, based on the required program enhancements resulting from industry programs, will be implemented as part of the FNP Unit 1 ISI Engineering Program [41. Corrective actions for augmented inspections will be developed using repair and replacement procedures equivalent to those requirements in ASME B&PV Code,Section XI or as determined independently by Southern Nuclear Operating Company (SNC), or in cooperation with the industry, to be equivalent or more rigorous than currently defined procedures.

This AMP for the FNP Unit 1 reactor internals demonstrates that the program adequately manages the effects of aging for reactor internals components and establishes the basis for providing reasonable assurance that the internals components will continue to perform their intended function through the FNP Unit 1 license renewal period of extended operation. This Westinghouse Commercial Atomic Power (WCAP) topical report supports the FNP Unit 1 License Renewal Commitment 6, which includes a submission to the NRC of an inspection plan for the Reactor Vessel Internals Program, as it would be implemented from the participation of FNP Unit 1 in industry initiatives 24 months prior to entering the period of extended operation. The implementation schedule for this commitment requires submission to the NRC no later than June 25, 2015; however the NRC granted SNC an extension until August 2015.

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

WCAP- 1801 I-NP July 2015 El-l0 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program 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 (SRP) [7]. The U. S. nuclear power industry has been actively engaged in recent years in a significant effort *to support the industry goal of responding to these requirements. Various programs have been underway within the industry over the past decade to develop guidelines for managing the effects of aging within PWR reactor internals. In 1997, the Westinghouse Owners Group (WOG) issued WCAP-14577, "License Renewal Evaluation: Aging Management for Reactor Internals" which was reissued as Revision 1-A in 200 1[8]1 after receiving NRC Staff review and approval. Later, an effort was engaged by the EPRI MLRP to address the PWR internals aging management issue for the three currently operating U. S. reactor designs -

Westinghouse, Combustion Engineering (CE) and Babcock & Wilcox (B&W).

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

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

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

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

- Components that are moderately susceptible to the aging effects

- 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 (GE), existing evaluations and prior examination results.

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

• MIRP-227-A [5], "PWR Internals Inspection and Evaluation Guidelines" (hereafter referred to as the "I&E Guidelines" or simply "MRP-227-A") provides the industry background, listing of reactor internals components requiring inspection, type of Nondestructive Examination (NDE) required for each component, timing for initial inspections and criteria for evaluating inspection WCAP- 8011-NP July 2015 El-i1 Revision 0

Enclosure 1ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 2-2 results. MRP-227-A provides a standardized approach to PWR internals aging management for each unique reactor design (Westinghouse, B&W and CE).

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

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

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

As described in NUREG-1825 [2], subsection 2.3.1.2.1, the FNP Unit 1 RVI consists of the lower core support structure, the upper core support structure and the in-core instrumentation support structures. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and Control Rod Drive Mechanism (CRDM), direct coolant flow past the fuel elements, direct coolant flow to the pressure vessel head, provide gamma and neutron shielding and provide guides for the in-core instrumentation.

The lower core support structure consists of the core barrel, the core baffle assemblies, the lower core plate, the neutron shield panels, the lower core support forging, the secondary support assembly and associated support columns. The lower core support structure is supported at its upper flange from a ledge in the reactor vessel, and is restrained at its lower end by a radial support system attached to the vessel wall. The upper core support structure consists of the upper support assembly, the upper core plate, support columns and control rod guide tube assemblies. The in-core instrumentation support structures consist of an upper system to convey and support thermocouples penetrating the vessel through the upper closure head and a lower system to convey and support flux thimbles penetrating the vessel through the bottom head.

The reactor vessel internals functions include structural support, flow distribution and radiation shielding.

FNP Unit 1 was granted a license for extended operation by the NRC through the issuance of a Safety Evaluation Report (SER) in NURiEG-1825 [2]. In the SER, the NRC concluded that the FNP Unit 1 License Renewal Application (LRA) adequately identified the RVI components that are within the scope of license renewal, as required by 10 CFR 54.4(a), and those subject to an Aging Management Review (AMR), as required by 10 CFR 54.2 1(a)(1) [6], and is therefore acceptable. A listing of the FNP Unit 1 RVI components and subcomponents, already reviewed by the NRC in the SER and that are subject to AMP requirements, is included in Tables B-i and B-2.

In accordance with 10 CFR Part 54 [6], frequently referred to as the License Renewal Rule, FNP Unit 1 has developed a program to direct the performance of aging management reviews of mechanical WCAP-18OI1-NP July 2015 El1-12 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 2-3 structures and components [27]. 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 Nuclear Energy Institute (NEI) 03-08 [13],

"Guidelines for the Management of Materials Issues," each plant will be required to use MRP-227-A and MRP-228 to develop and implement an AMP for reactor internals no later than three years after the initial industry issuance of MRP-227, Revision 0. MIRP-227, Revision 0 was issued in December 2008, and plant AMVPs [1 ] must therefore be completed by December 2011, or sooner, if required by plant-specific License Renewal Commitments. According to [3], FNP Unit 1 is a Category B plant that is expected to submit their RVI AMP based on the guidance of MRP-227-A, which is consistent with their commitments. Per the SER [2], FNP Unit 1 has a commitment to submit their AMP for approval by the NRC no later than June 25, 2015; however the NRC granted SNC an extension until August 2015.

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

WCAP- 1801 1-NP July 2015 E1-13 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 3-1 3 PWR VESSEL INTERNALS PROGRAM OWNER The SNC "PWR Reactor Internals Program Strategic Plan" [1], which is a sub-tier document of the PWR Primary System Integrity Program [341, manages the effects of age-related degradation mechanisms of reactor vessel internals. The successful implementation and comprehensive long-term management of the FNP Unit 1 RVI AMP will require the integration of SNC, corporately and at Farley, and interaction with multiple industry organizations including, but not limited to, the ASME, MRP, NRC and PWROG. The responsibilities of the individual SNC corporate and Farley groups are provided in the following paragraphs. SNC 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 NEL 03-08 practices.

The overall responsibility for the scheduling and conducting of the PWR Primary System Integrity Program, including the RVI AMP, is the PWR Primary System Integrity Program owner in the Corporate Engineering Programs department.

Additional responsibilities and the appropriate responsible personnel, as described in [34], are discussed in the following subsections.

3.1 SNC - EXECUTIVE

• The overall responsibility for successful implementation of the PWR Primary System Integrity program (including reactor internals) resides with the Chief Nuclear Officer. As such, that individual establishes expectations for the implementation of the PRW Primary System Integrity Program.

• Approval of any deviations from mandatory or needed elements in industry documents that affect Farley.

3.2 SNC - CORPORATE

• The PWR Primary System Integrity Program owner resides in the Corporate Engineering Programs department and has overall responsibility for the development and maintenance of the PWR Primary System Integrity Program and for the following activities:

- Development of implementing instructions and guidelines, as needed.

- Development of a ten outage plan for reactor internals material management. This plan provides inspection and mitigation schedule for each unit over the next ten outages.

- Providing technical expertise and oversight to the SNC fleet and/or serve as the subject matter expert for reactor internals.

- Participate in industry programs for reactor internals aging management and addressing primary water stress corrosion cracking (PWSCC) issues.

WCAP-1l801 1-NP July 2015 El1-14 Revision 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 3-2

- Participate in industry assessments; ensuring program is in alignment with industry guidance and implementing best practices.

- Utilize the technical team to drive best practices and provide oversight.

- Ensuring that industry best practices, industry operating experience from Institute of Nuclear Power Operations (INPO), EPRI, Owners groups and others ( e.g., NSSS vendors and regulatory requirements) are communicated to the fleet and incorporated into the applicable program documentation.

- Review examination results, operating conditions, material properties and fabrication history for use in projecting future conditions and actions.

- Processing formal transmittals from the MVRP.

- Identifying areas for standardization between the sites/projects with respect to the PWR Primary System Integrity Program.

- Documenting and processing deviations from mandatory or needed elements in industry documents.

- Promptly communicating with the industry issue program Chairman or Project Manager emergent issues that could have safety significance, or represent a new degradation type that may have an effect on industry guidance or the existing knowledge base.

- Participating in program self-assessments and benchmarking activities.

- Providing input to MIRP industry inspection surveys.

- Drive susceptible components items towards long term resolution (asset management).

- Communicating program performance gaps to management.

- Periodically observe work activities and provide feedback to individuals and lessons learned to fleet.

- Updating Program Notebook.

- In addition to the above, provide oversight to the site programs, as needed.

The Engineering Integrity Programs group responsibilities include:

- Updating the reactors internals inspection plan.

- Provide the results of augmented examinations, which require reporting to the regulatory authority to Nuclear Licensing.

WCAP- 1801 I-NP July 2015 El-15 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 3-3

• The Fleet Chemistry group is responsible for sharing information obtained from industry participation with the appropriate SNC personnel on primary chemistry, as well as chemical mitigation experience.

3.3 SNC - FNP SITE Plant Management responsibilities include:

• Providing sufficient resources and oversight to~the PWR Primary System Integrity Program to ensure PWR Primary System materials degradation do not compromise the integrity of the primary system pressure boundary.

• Ensuring that the responsibility for implementing the site elements of the Program has been clearly defined for each department and assigned to the trained and qualified personnel.

Site Engineering Programs Department responsibilities include:

• Designating a Site Program Owner and backup. Site Program Owners responsibilities are described in NMP-ES-009 [14].

• Provide updates to the Reactor Intemnals ten outage plan and budget estimates to support the overall program.

• Site Implementation of the PWR Primary System Integrity Program and MRP Guidelines.

• Coordination of engineering evaluations and disposition of indications discovered during vessel examinations.

• Maintaining knowledge of significant operating evolutions that might impact the integrity of the Reactor Pressure Vessel (RPV) upper and lower heads.

• Reviewing and responding to industry OE.

• Coordinating vendor support for any specialized equipment needed to complete the required inspections.

• Outage planning for RPV inspections.

• Develop and implement corrective action plans for PWR Primary System Integrity Program issues, as requested by the FNP Engineering Programs Manager.

• Performing site assessments in accordance with NMP-GM-003 [15].

• Updating Program Notebook.

WCAP-1801 1-NP July 2015 E1-16 Revision 0

Enclosure i to NL-1 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 3-4

• Generate commitment notebooks in accordance with NMP-ES-063-GL02 [16], 1 year prior to the license renewal period.

3.4 PWR PRIMARY SYSTEM INTEGRITY PROGRAM TECHNICAL TEAM

• Support management on PWR Primary System Integrity issues, including recommending optimum technical and management practices for nuclear safety, plant availability and equipment reliability.

• Provide a technical forum for the integration of the various elements needed to implement an effective Program.

• Develop long range plans for assessment, inspection, mitigation and repairs, taking into account material condition, associated projections, industry insight and SNC strategic plans.

• Ownership of the strategic plan for inspection, mitigation, repair and chemistry initiatives.

• Ensure timely review of PWR Primary System Integrity issues by meeting at least once per year.

• Evaluate inspection, mitigation, repair and maintenance technologies with respect to the benefit of primary system integrity and cost.

• Establish strategic goals.

• Evaluate degradation mechanisms and operating conditions.

• Be knowledgeable of industry PWR Primary System Integrity issues and address potential impacts to FNP.

• Drive and adopt industry best practices.

• Provide oversight of implementation ofReactor Internals activities.

WCAP-1801 1-NP July 2015 E1-17 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-1l 4 DESCRIPTION OF THE FARLEY NUCLEAR PLANT UNIT 1 REACTOR INTERNALS AGING MANAGEMENT PROGRAMS AND INDUSTRY PROGRAMS The U. S. nuclear industry, through the combined efforts of utilities, vendors and independent consultants, has defined a generic guideline to assist utilities in developing reactor internals plant-specific aging management programs based on inspection and evaluation. The intent of this program is to ensure the long-term integrity and safe operation of the reactor internals components. SNC has developed this AMP in conformance with the 10 Generic Aging Lessons Learned (GALL) [17] attributes and MRP-227-A [5].

The LRA was based on Rev. 0 of the GALL [12], where this AMP is reconciled to Rev. 2 of the GALL

[17].

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

Where applicable, credit is taken for existing programs such as water chemistry [18] and inspections prescribed by the ASMIE Section XI Inservice Inspection Program [4], as well as mitigation projects such as support pin replacement [20] and baffle bolt replacement [42], combined with augmented inspections or evaluations as recommended by MRP-227-A.

Aging degradation mechanisms that impact internals have been identified and documented in FNP Unit 1 Aging Management Reviews [21 ]. 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 FNP Unit 1 AMR methodology and the additional industry work summarized in MRP-227-A. All sources are consistent and address concerns about component degradation resulting from the following eight material aging degradation mechanisms identified as affecting reactor internals:

• Stress Corrosion Cracking (SCC)

Stress corrosion cracking (SCC) refers to local, non-ductile 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.

WCAP-1801 1-NP July 2015 El1-18 Revision 0 Ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-2 Fatigue Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads and temperatures. After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack. Corrosion fatigue is included in the degradation description.

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

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

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

• Irradiation Embrittlement Irradiation embrittlement (IE) 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 ifra crack is present and the local applied stress intensity exceeds the reduced fracture toughness.

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Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-3 aVoid Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation of microscopic cavities in the material. These cavities result from the nucleation and growth of clusters of irradiation-produced vacancies. Helium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material. Void swelling may produce dimensional changes that exceed the tolerances on a component. Strain gradients produced by differential swelling in the system may produce significant stresses. Severe swelling

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

• Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation-Enhanced Creep The loss of preload aging effect can be caused by the aging mechanisms of stress relaxation or creep. Thermal stress relaxation (or primary creep) is defined as the unloading of preloaded components due to long-term exposure to elevated temperatures, as seen in PWR internals. Stress relaxation occurs under conditions of constant strain where part of the elastic strain is replaced with plastic strain. Available data show that thermal stress relaxation appears to reach saturation in a short time (< 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />) at PWR internals temperatures.

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

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

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

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Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-4 4.1 EXISTING FARLEY NUCLEAR PLANT UNIT 1 PROGRAMS The overall strategy of SNC for managing aging in reactor internals components is supported by the following existing programs [23]:

• Water Chemistry Control Program

• Inservice Inspection 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 Water Chemistry Control Program The FNP Water Chemistry Program [18] will manage loss of material and cracking within system components and structures, thereby ensuring continued structural integrity, reliability and availability. The program includes monitoring of detrimental species and addition of chemical additives. The program utilizes the EPRI water chemistry guidelines [25] in establishing chemistry control procedures for FNP.

These documents are updated as necessary to reflect improved guidance and industry experience. Prior to adopting a later revision, SNC evaluates the acceptability of implementing requirements.

With one exception, the FNP closed cycle cooling water monitoring and chemistry control methods are consistent with those described in NUREG-1801 [17]. The closed cycle cooling water program described in NUREG- 1801 [17] places emphasis on thermal-hydraulic performance testing for pumps and heat exchangers. The FNP program deals with performance monitoring as outlined in Section 5 of EPRI TR-107396 [32] regarding chemistry monitoring.

4.1.2 Inservice Inspection Program The FNP Unit 1 ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD, Program [4] is in accordance with ASME Section XI 2001 Edition with the 2003 Addenda [22]. The FNP ISI Program is implemented in accordance with 10 CFR 50.55a and is subject to its limitations and modifications. The program manages loss of material, cracking, changes in material properties, loss of preload/stress relaxation, loss of fracture toughness and change in strength in concrete. The program inspections include periodic visual, surface and/or volumetric examinations and leakage tests of Class 1, 2 and 3 pressure-retaining components and their integral attachments, including welds, pump casings, valve bodies and pressure-retaining bolting. The program is updated as required by 10 CFR 50.55a.

The FNP Unit 1 ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD, Program is consistent with the collection of acceptable ASME Section XI subprograms described in NUREG- 1801

[17].

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Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-5 4.2 SUPPORTING FARLEY NUCLEAR PLANT UNIT 1 PROGRAMS AND AGING MANAGEMENT SUPPORTIVE PLANT ENHANCEMENTS 4.2.1 Reactor Internals Aging Management Review Process A comprehensive review of aging management of reactor internals was performed according to the requirements of the License Renewal Rule [6] as directed by the Plant Farley commodity review procedure [27]. The Plant Farley License Renewal Commodity Group Review Document [21 ] documents the results of the aging management review performed in support of FNP Unit 1 license renewal for reactor internals. The FNP Unit 1 LRA was approved by the NRC in NUREG- 1825 [2]. RVI components specifically noted as requiring aging management, as identified in the NUJREG, are summarized in Table B-I of this AMP.

The AMiR supported the LRA as follows:

• Identified applicable aging effects requiring management

• Associated aging management programs to manage those aging effects

• Identified enhancements or modifications to existing programs, new aging management programs, or any other actions required to support the conclusions reached in the review Aging management reviews were performed for each FNP Unit 1 system that contained long-lived, passive components requiring aging management review, in accordance with the Plant Farley commodity review procedure [27]. This review is not repeated here, but the results are fully incorporated into the FNP Unit 1 RVI AMP.

4.2.2 Reactor Vessel Internals Program The FNP Reactor Vessel Internals Program [1] will be implemented prior to entering the period of extended operation to provide an integrated inspection program that addresses the reactor internals. The program will be used during the period of extended operation to manage the effects of crack initiation and growth due to IASCC; loss of fracture toughness due to irradiation embrittlement, thermal embrittlement (TE) or void swelling; or changes in material properties due to void swelling.

4.2.3 Flux Detector Thimble Inspection Program The FNP Flux Detector Thimble Inspection Program [19] will be implemented prior to entering the period of extended operation to formalize examinations already being performed. It will be used to identify loss of material resulting from fretting/wear in the detector thimbles during the period of extended operations.

The program is in response to NRC Bulletin 88-09 [24] with the intention to ensure that pressure boundary integrity of the in-core system flux thimble tubes is maintained.

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Enclosure 1ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-6 4.2.4 Control Rod Guide Tube Support Pin Replacement Project The control rod guide tube support pins are used to align the bottom of the control rod guide tube assembly into the top of the core plate. In general, SCC prevention is aided by adherence to strict primary water chemistry limits that effectively prevent SCC and greatly reduce the probability of IASCC. The limits imposed by the Water Chemistry Control Program at FNP Unit 1 are consistent with the latest EPRI guidelines as described in Section 4.1.

Since 1990, ultrasonic testing has indicated that SCC has occurred in certain second generation alloy X-750 (Grade 688) support pins in various plants with greater than 55,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation. Prior to replacement, numerous support pins at other plants using alloy X-750 material failed during removal or during operation between 110,900 and 149,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of operation.

In response to the industry concern for SCC of the alloy X-750 material, SNC replaced all of the upper internals guide tube support pins at FNP Unit 1 (October 2004) with Westinghouse-supplied strain hardened austenitic type 316 stainless steel support pins to mitigate the possibility of continued SCC of these components. Detailed descriptions of the replacement are contained within the Field Change Notice

[20].

4.2.5 Power Uprating Project FNP Unit 1 was originally licensed to operate at 2652 MWt core power (2660 MWt thermal); however, most safety analyses calculations had been performed assuming a higher core power. The FNP Unit 1 power uprate project increased the core operating power to 2775 MWt (2785 MWt thermal). Safety analysis assumed 2831 MWt core power for analyses supporting the power uprate project demonstrating margin to the uprated licensed core power output. Information on the power uprate and supporting analyses can be found in the licensing report [43] and NRC safety evaluation (SE) of the associated FNP license amendment [44].

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

The AMR for the FNP Unit 1 internals documented in [21] utilized WCAP-14577 [8] as an input source regarding applicable aging affects and aging management programs. FNP reactor internal components, plant operating and loading conditions, temperature, pressure and water chemistry are consistent with or bounded by those reflected in [8]. Therefore, the NRC approved topical report [8] is applicable to the FNP AMP.

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Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-7 4.3.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines MRP-227-A, as discussed in Section 2, was developed by a team of industry experts including utility representatives, NSSS vendors, independent consultants and international committee representatives who reviewed available data and industry experience on materials aging. The objective of the group was to develop a consistent, systematic approach for identifyring and prioritizing inspection and evaluation requirements for reactor internals. The following subsections briefly describe the industry process.

4.3.2.1 MRLP-227-A, RVI Component Category 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; appropriate inspection, evaluation and implementation requirements for reactor internals were defined.

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

• Primary Those PWR internals that are highly susceptible to the effects of at least one of the eight aging mechanisms were placed in the Primary group. The aging management requirements that are needed to ensure functionality of Primary components are-described in the Inspection &

Evaluation (I&E) guidelines. The Primary group also includes components that have shown a degree of tolerance to a specific aging degradation effect, but for which no highly susceptible component exists or for which no highly susceptible component is accessible.

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

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

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Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-8 aNo Additional Measures Programs Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the No Additional Measures group. Additional components were placed in the No Additional Measures group as a result of a failure mode, effects, and criticality analysis (FMECA) and the functionality assessment. No further action is required by these guidelines for managing the aging of the No Additional Measures components.

The categorization and analysis used in the development of MRP-227-A are not intended to supersede any ASME B&PV Code Section XI [22] requirements. Any components that are classified as core support structures, as defined in ASME B&PV Code Section XI IWB-25 00, Category B-N-3, have requirements that remain in effect and may only be altered as allowed by 10 CFR 50.5 5a.

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

Mandatory There is one Mandatory element:

1. Each commercial U. S. PWR unit shall develop and document aprogram for management of aging of reactor internals components within thirty-six months following issuance of MRP-22 7-Rev. 0 (that is, no later than December 31, 2011).

FNP Unit I Applicability: MRP-227, Revision 0 was officially issued by the industry in December 2008. An AMP must be developed within thirty-six months following issuance of MRP-227, Revision 0. To fulfill this requirement and the license renewal commitments provided in Section 1, SNC developed NMP-ES-029-GL02, "PWR Reactor Internals Program Strategic Plan" [ 1]. This program was implemented to meet this requirement as documented in [1 ].

According to the NRC Regulatory Issue Summary (RIS) [3], FNP Unit 1 qualifies as a Category 13 plant because they have a renewed license with a commitment to submit an AMP/inspection plan based on MRP-227, but have not yet been required to do so by their commitment. This AMP fulfills the license renewal commitment to submit an implementation schedule for FNP Unit I in accordance with MRP-22'7-A [5] to the NRC no later than permitted. The original date for submission was June 25, 2015; however the NRC granted SNC an extension until August 2015.

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Enclosure i to NL-15-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-9 Needed There are five Needed elements:

1. Each commercial U.S. PWR unit shall implement MRP-22 7-A, Tables 4-1 through 4-9 and Tables 5-1 through 5 -3 for the applicable design within twenty-four months following issuance of MRP-22 7-A.

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

2. Examinations specified in the MRP-22 7-A guidelines shall be conducted in accordancewith Inspection Standard, MRP-228 [10].

FNP Unit 1 Applicability: SNC has developed fleet NDE procedure NMP-ES-024-1 12 [38] to detail the process for the implementation of MRP-228 [ 10] for PWR Internals NDE requirements at Southern Nuclear facilities. The procedure defines a process to ensure that the combinations of equipment, procedures and personnel used to perform examinations of reactor internals at SNC sites meet the implementation requirements of MRP-228.

3. Examination results that do not meet the examination acceptance criteriadefined in Section 5 of the MRP-22 7-A guidelines shall be recordedand entered in the plant corrective action program and dispositioned.

FNP Unit 1 Applicability: FNP Unit 1 will comply with this requirement.

4. Each commercial U.S. PWR unit shallprovide a summary report of 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.

FNP Unit 1 Applicability: As discussed in subsection 4.3.3, SNC will participate in future industry efforts and will adhere to industry directives for reporting, response and follow-up.

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

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

WCAP-I18011-NP July 2015 E1-26 Revision 0 ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-10 4.3.2.3 GALL AMIP Development Guidance It should be noted that Section XI.M16A ofNUREG-1801, Revision 2 [17] includes a description of the attributes that make up an acceptable AMP. These attributes are consistent with the FNP Unit 1 Aging Management Review process. Evaluation of the FNP Unit 1 RVI AMP against GALL attribute elements is provided in Section 5 of this AMP.

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

4.3.2.4 MRP-227-A Applicability to FNP Unit 1 The applicability of MRP-227-A to FNP Unit 1 requires compliance with the following MRP-227-A assumptions:

30 years of operation with high-leakage core loadingpatterns (freshfuel assemblies loaded in peripherallocations)followed by implementation of a low-leakage fuel management strategyfor the remaining 30 years of operation.

FNP Unit 1 Applicability: According to the SNC RVI Program [1], the SNC fuel management program changed from a high to a low leakage core loading pattern prior to 30 years of operation of FNP Unit 1.

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

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

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

FNP Unit I 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. SNC has not made any modifications to the Unit 1 reactor internals components since May 2007 [1].

Based on the plant-specific applicability, as stated, the MRP-22-7-A work is representative for FNP Unit 1.

4.3.3 Ongoing Industry Programs The U. S. industry, through both the EPRII/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. SNC will maintain WCAP-1801 1-NP July 2015 E1-27 Revision 0

Enclosure ito NL-l 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 4-11 RVI 1Aging Enclosure FNP-1 Management Program to NL-15-1507 cognizance of industry activities related to PWR internals inspection and aging management. SNC will also address/implement industry guidance, stemming from those activities, as appropriate under NEI 03-08 practices.

4.4

SUMMARY

It should be noted that the SNC FNP Unit 1, the MRP and the PWROG approaches to aging management are based on the GALL approach to aging management strategies. This approach includes a determination of which reactor internals passive components are most susceptible to the aging mechanisms of concern and then a 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 Farley for the initial basis of the LRA that resulted in the NRC SER in NUREG- 1825 [2].

The approach used to develop the FNP Unit 1 AMP is fully compliant with regulatory directives and approved documents. The additional evaluations and analysis completed by the MRP industry group have provided clarification to the level of inspection quality needed to determine the proper examination method and frequencies. The tables provided in MRP-227-A and included as Appendix C of this AMP provide the level of examination required for each of the components evaluated.

It is the Farley 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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Enciosure 1 to NL-15-1 507 FNP-I RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-1 5 FARLEY NUCLEAR PLANT REACTOR INTERNALS\AGING MANAGEMENT PROGRAM ATTRIBUTES The FNP Unit 1 RVI AMP is credited for aging management of RVI Components for the following eight aging degradation mechanisms and their associated effects:

• Stress corrosion cracking

• Irradiation-assisted stress corrosion cracking

• Wear \

• Fatigue \

• Thermal aging embrittlement

• Irradiation embrittl'ement

• Void swelling and irradiation growth

• Thermal and irradiation-enhanced stres's relaxation or irradiation-enhanced creep The attributes of the FNP Unit 1 RVI AMP and compliance with NUREG-1801 (GALL Repo3rt),

Section XI.M 16A, "PWR Vessel Internals" [ 17] are described in this section. The GALL idenatifies 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.

SNC fully utilized the GALL process contained in NUREG-1801 [ 17] in performing the agin *,

management review of the reactor internals in the license renewal process. However, SNC ma*de a commitment (see NUJREG-1825 [2]) to incorporate the following: (1) implement the new FNP Reactor Vessel Internals Program prior to entering the period of extended operation, (2) continue to par;ticipate in industry initiatives intended to clarify the nature and extent of aging mechanisms affecting the1 FNP reactor internals, (3) incorporate the results 'of these initiatives into the RVI program and (4) si~mit a inspection plan for the RVI Program for NRC review and approval at least 24 months prior to entering*the periods of extended operation for the FNP units.

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

Specific details of the FNP Unit 1 reactor internals AMP are summarized in .the following subse btions.

5.1 GALL REVISION 2 ELEMENT 1: SCOPE OF PROGRAM GALL Report AMIP Element Description "The scope of the program includes all RVI components at the FarleyNuclear Plant Uni 1 Nuclear Plant, which is built to a Westinghouse NSSS design. The scope of the program ap~plies the methodology andguidance in the most recently NRC-endorsed version of MRP-22 7, ivhich provides augmented inspection andflaw evaluation methodology for assuringthe fun ctioncal integrity of safety-relatedinternals in commercial operating U.S. PWR nuclearpower plantrs 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 Category B-N-3 by the ASME Code, Section Xl), those RVI components that serve an intended license renewal safetyfunction pursuantto criteriain 10 CFR 54.4(a)(1), and other RV1I; WCAP-1801 1-NP Jut'~ 2015 E1-29 Rev'ision 0

Enclosure i to NL-l 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-2 components whose failure couldprevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a) (1)(i), (ii), or (iii). The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are requiredto be subject to an aging management review (AMR), as defined by the criteria set in 10 CFR 54.21(a)(1). The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenancesto the reactor vessel and are adequately managed in accordance with an applicant'sAMiPthat corresponds to 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-227 methodology, and any additionalprograms, actions, or activities that are discussed in these LRAAI responses and creditedfor aging management of the applicant'sR VI components. The LRAAIs are identified in the staff 'ssafety 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 of MRP-227 specifies applicability limitations to base-loadedplants and the fuel loading management assumptions upon which thefunctionality analyses were based These limitations and assumptions require a determination of applicabilityby the applicantfor each reactor and are covered in Section 2.4 of MRP-22 7" [17].

FNP Unit 1 Program Scope The FNP Unit 1 reactor internals consist of the lower core support structure, the upper core support structure and the in-core instrumentation support structures. The reactor internals support the core, maintain fuel alignment, limit fuel assembly movement, maintain alignment between fuel assemblies and CRDMs, direct coolant flow past the fuel elements, direct coolant flow to the pressure vessel head, provide gamma and neutron shielding, and provide guides for the in-core instrumentation. The lower core support structure consists of the core barrel, the core baffle assemblies, the lower core plate, the neutron shield panels, the lower core support forging, the secondary support assembly and associated support columns. The lower core support structure is supported at its upper flange from a ledge in the reactor vessel and, at its lower end, is restrained by a radial support system attached to the vessel wall. The upper core support structure consists of the upper support assembly, the upper core plate, support columns and control rod guide tube assemblies. The in-core instrumentation support structures consist of an upper system to convey and support thermocouples penetrating the vessel through the upper closure head, and a lower system to convey and support flux thimbles penetrating the vessel through the bottom head.

Additional RVI details are discussed in FNP Unit 1 updated final safety analysis report (UFSAR) subsection 4.2.2, Reactor Vessel Internals.

The FNP Unit 1 RVI subcomponents that required aging management review are indicated in the previously submitted Table 2.3.1-2 of the FNP Unit 1 LRA [23]. The components listed in Table 2.3.1.2 are consistent with those in Appendix B of this report.

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Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-3 The FNP Unit 1 Reactor Internals AMR was conducted and documented in [21]. The table summarizing the results of that review was also documented in Table 3.1.2-2 of the FNP Unit 1 LRA [23].This table is included in Appendix B of this AMP. The table identifies the aging effects that require management for the components requiring AMR. A column in the tables lists the program/activity that is credited to address the component and aging effect during the period of extended operation. The NRC has reviewed and approved the aging management strategy presented in the Appendix B tables as documented in the SER on license renewal [2].

The results of the industry research provided by MRP-22 7-A, summarized in the tables of Appendix C, provide the basis for the required augmented inspections, inspection techniques to permit detection and characterizing of the aging effects (cracks, loss of material, loss of preload, etc.) of interest, prescribed frequency of inspection and examination acceptance criteria. The information provided in MRP-227-A is rooted in the GALL methodology. The basic assumptions of MRP-227-A, Section 2.4 are met by IFNP Unit 1 and are addressed in subsection 4.3.2.4 of this AMP. The Topical Report Conditions and Applicant/Licensee Action Items provided by the NRC in the SE on MIRP-227, Revision 0 [5] are met by FNP, 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 FNP Unit 1 RVI AMP scope is additionally based on previously established and approved GALL Report approaches through application of the MRP-227-A [5] methodologies to determine those components that require aging management.

Conclusion This element complies with the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [17] and Commitment 6 in the FNP SER.

5.2 GALL REVISION 2 ELEMENT 2: PREVENTATIVE ACTIONS GALL Report AMP Element Description "The guidance in MRP-227 relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of itsforms

[SCC, PWSCC, or 1ASCCJ). Reactor coolant water chemistry is monitored and maintainedin accordance with the Water Chemistry Program. The program description, evaluation, and technical basis of water chemistry arepresented in GALL AMP XI.M2, 'Water Chemistry"'" [17].

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

FNP Unit 1 Primary Water Chemistry Program The FNP Water Chemistry Program [18] will manage loss of material and cracking within system components and structures, thereby ensuring continued structural integrity, reliability and availability. The WCAP-1801 1-NP July 2015 E1-31 Revision 0 ito NL-1 5-1 507 FNP-1 RVi Aging Management Program Westinghouse Non-Proprietary Class 3 5-4 program includes monitoring of detrimental species and addition of chemical additives. The FNP program utilizes the EPRI P'WR Primary Water Chemistry Guidelines [25] in establishing chemistry control procedures for FNP. Prior to adopting later revisions of the EPRI guidelines, SNC evaluates the acceptability of any changes in implementing requirements. The FNP Water Chemistry Program incorporates the best practices of industry organizations, vendors and utilities.

Conclusion This element complies with the corresponding aging management attribute in NUREG-1801,Section XI.M16A [17] and Commitment 6 in the FNP Unit 1 SER.

5.3 GALL REVISION 2 ELEMENT 3: PARAMETERS MONITORED OR INSPECTED GALL Report AMP Element Description "The program manages the following age-relateddegradationeffects and mechanisms that are applicable in general to the R171 components at the facility: (a) cracking induced by SCC, PWSCC, LASCC, orfatigue/cyclical loading, (b) loss of material induced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiationembrittlement,"

(d) changes in dimension due to void swelling and irradiationgrowth, distortion, or deflection; and (e) loss ofpreload caused by thermal and irradiation-enhancedstress relaxation or creep.

For the management of cracking, the program monitors the evidence of surface breaking linear discontinuities if a visual inspection technique is used as the non-destruction examination (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 loss of material occurring in the components. For the management of loss ofpreload, the program monitorsfor gross surface conditions that may be indicative of loosening in applicable bolted,fastened, keyed, or pinned connections. The program does 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 managed by using visual or volumetric examination techniques to monitorfor cracking in the components and by applying applicable reducedfracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warranta supplementalflaw growth orfiaw tolerance evaluation under MRP-22 7 guidance or ASME Code, Section Xl requirements. The program uses physical measurements to monitorfor any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.

Specifically, the program implements the parametersmonitored/inspectedcriteriafor Westinghouse designed Primary Components in Table 4-3 of MRP-22 7. Additionally, the program implements the parameters monitored/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 X1, Table 1WB-2500-1, Examination CategoriesB-N-3, as implemented through the applicant'~sASME Code,Section XI program, or WCAP- 1801 1-NP July 2015 E1-32 Revision 0

Enclosure 1Ito NL-15-1 507 FNP-1 RVl Aging Management Program Westinghouse Non-Proprietary Class 3 5-5 the recommendedprogram for inspecting Westinghouse-designedflux thimble tubes in GALL AMP XI.M3 7, "Flux Thimble Tube Inspection. "No inspections, except for those specified in ASME Code,Section XI, are requiredfor components that are identified as requiring "No Additional Measure," in accordance with the analyses reported in MRP-22 7" [17].

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

This AMP implements the requirements for the Primary Component inspections from Table 4-3 of MRP-227-A (included in Appendix C of this AMP as Table C-1), the Expansion Component inspections from Table 4-6 of MRP-227-A (included in Appendix C of this AMP as Table C-2) and the Existing Component inspections from Table 4-9 of MRP-227-A (included in Appendix C of this AMP as Table C-3). These tables contain requirements to monitor and inspect the RVI through the period of extended operation to address the effects of the eight aging degradation mechanisms.

For license renewal, the ASME Section XI Program [4] includes periodic visual, surface and/or volumetric examinations and leakage tests of Class 1, 2 and 3 pressure-retaining components and their integral attachments, including welds, pump casings, valve bodies and pressure-retaining bolting. The requirements of MRP-227-A only augment and do not replace or modify the requirements of ASME Section XI. This program is consistent with the corresponding program described in the GALL Report

[17].

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

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.MI6A [17] and Commitment 6 in the FNP Unit 1 SER.

5.4 GALL REVISION 2 ELEMENT 4: DETECTION OF AGING EFFECTS GALL Report AMP Element Description "The detection of aging effects is covered in two places."(a) the guidance in Section 4 of MiRP-227provides an introductory discussion andjustificationof the examination methods selectedfor detecting the aging effects of interest; and (b) standardsfor examination methods, procedures, andpersonnel are provided in a companion document, MRP-228. In all cases, well-establishedmethods were selected. These methods include volumetric UT examination methods for detectingflaws in bolting, physical measurementsfor detecting changes in dimension, and various visual (VT-3, VT-i, and EVT-1) examinationsfor detecting effects rangingfrom general conditions to detection and sizing ofsurface-breaking discontinuities. Surface examinations may WCAP-1801 1-NP July 2015 E1-33 Revision 0

Enclosure 1ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-6 also be used as an alternative to visual examinationsfor detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, andfatigue is monitored/inspectedby either VT-i or EVT-]

examination (for internals other 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 tolerant of easily detected largeflaws, even under reduced fracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspectfor loss of material induced by wear andfor general aging conditions, such as gross distortion caused by void swelling and irradiationgrowth or by gross effects of loss of preloadcaused by thermal and irradiation-enhancedstress relaxation and 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, inspectionsperformed on the R VI components are performed consistent with the inspectionfrequency and sampling basesfor Primary Components, Existing Requirement Components, and Expansion Components in MRP-22 7, which have been demonstrated to be in conformance with the inspection criteria,sampling basis criteria,and sample Expansion criteria in Section A. 1.2.3.4 of NRC Branch PositionRLSB-J.

Specifically, the program implements the parametersmonitored/inspectedcriteriaand basesfor inspecting the relevantparameterconditionsfor Westinghouse designed Primary Components in Table 4-3 of MRP-227 andfor Westinghouse designed Expansion Components in Table 4-6 of MRP-22 7.

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

In addition, in some cases (as defined in MRP-22 7), physical measurements are used as supplemental techniques to managefor the gross effects of wear; loss ofpreload due to stress relaxation, orfor changes in dimension due to void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program include thatfor the hold down spring. The hold down spring at FNP Unit 1 isfabricatedfrom Type 304 SS that requires inspection by physical measurement" [17].

FNP Unit 1 Detection of Aging Effects Detection of indications required by the ASME Section XI 1S1 Program [4] is well established and field-proven through the application of the Section XI ISI Program. Those augmented inspections that are taken from the MRP-227-A recommendations will be applied through use of the MRP-228 inspection standard. This AMP implements the augmented inspection requirements of Table 4-3, Table 4-6 and Table 4-9 from MRP-227-A for the Primary, Expansion and Existing Components, respectively. These are included in Appendix C of this AMP for reference. These tables include the inspection frequency and WCAP-l8O1 1-NP July 2015 E1-34 Revision 0

Enclosure i to NL-15-1 507 FNP-l RVI Aging Management Program\

Westinghouse Non-Proprietary Class 3 5-7 sampling basis. For the Expansion Components of MRP-227-A, this AMP implements the expansion requirements of Table 5-3 of MRP-227-A (included in Appendix C of this AMP as Table C-4).

Inspection can be used to detect physical effects of degradation in cluding cracking, fracture, wear and distortion. The choice of an inspection technique depends on the n~ature 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) visural, (2) ultrasonic and (3) physical measurement. The three different visual techniques include VT-3, V\T['- and EVT-l. 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 ofithese techniques for augmented reactor internals inspections are documented in MRP-228 [10]. SNC has developed a fleet NDE procedure [38] which details the SNC process for implementing the techniques per the requirements prescribed in MRP-228.

VT-I1 Visual Examinations The acceptance criteria for visual examinations conducted under categories .B-N-2 (welded core support struictures and interior attachments to reactor vessels) and B-N-3 (remoyable core support structures) are defined in IWB-3 520 [22]. VT-i visual examination is intended to iden~ify crack-like sulrface 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-35 10 [22]

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

Although the IWB-3 520 criteria do not directly apply to austenitic stainles~s 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 fo~r components requiring close visual examinations with some estimate of the scale of deformation or weair. Note that in MRP-227-A, VT- 1 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 "etn.Teeoe oadtoa VT-i inspections over and above those required by ASME Section XI ISI have been specified.

EVT- 1 Enhanced Visual Examination for the Detection of Surface Breaking 1Flaws In the augmented inspections detailed in the MRP-227-A for reactor intemnals,*.lthe EVT- 1 enhanced visual examination has been identified for inspection of components where surface-brieaking flaws are a potential concern. Any visual inspection for cracking requires a reasonable exp.ectation that the flaw length and crack mouth opening displacement meet the resolution requirements of the observation technique. The EVT-1 specification augments the VT-i requirements to provide~more rigorous inspection standards for stress corrosion cracking, and has been demonstrated for similar in~;pections in boiling water reactor (BWR) intemnals. Enhanced visual examination (i.e., EVT-1) is also condl.icted in accordance with the requirements described for visual examination (i.e., VT-i) with additional req~ulrements (such as WCAP- 1801 1-NP iJuly 2015 E1-35 Revision 0

Enclosure i to NL-15-1507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-8 camera scanning speed). Any recommendation for EVT-1I 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" [11]. The acceptance criteria developed using these methodologies may be created on either a generic or plant-specific basis because both loads and component dimensions may vary from plant-to-plant within a typical PWR design.

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

The acceptance criteria for visual examinations conducted under categories B-N-2 (welded core support structures and interior attachments to reactor vessels) and B-N-3 (removable core support structures) are defined in IWB-3 520. These criteria are designed to provide general guidelines. The unacceptable conditions for a VT-3 examination are listed below:

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

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

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

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

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

• Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5 percent The VT-3 examination is intended for use in situations where the degradation is readily observable. It is meant to provide an indication of condition, and quantitative acceptance criteria are not generally required. In any particular recommendation for VT-3 visual examination, it should be possible to identify the specific conditions of concern. For instance, the unacceptable conditions for wear indicate wear that might lead to loss of function. Guidelines for wear in a critical-alignment component may be very different from the guidelines for wear in a large structural component.

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Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-9 Surface Examination In order to further characterize discontinuities on the surface of components, surface examination can supplement either visual (VT-3) or (VT-l/E VT-i) examinations specified in these guidelines. This supplemental examination may thus be used to reject or accept relevant indications. A surface examination is an examination that indicates the presence of surface discontinuities, and the ASME B&PV Code [22] lists magnetic particle, liquid penetrant, eddy current and ultrasonic examination methods as surface examination alternatives. Here, only the electromagnetic testing (ET), also called eddy current surface examination method, is covered.

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

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

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

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

Establishment of the acceptable bolting pattern for any system of bolts requires analysis to demonstrate that the system will maintain reliability and integrity in continuing to perform the intended function of the component. This analysis is highly plant-specific. Therefore, any recommendation for UT inspection of bolts assumes that the plant owner will work with the designer to establish acceptable bolting patterns prior to the inspection to support continued operation. For Westinghouse-designed plants, acceptable bolting patterns for baffle-former and barrel-former bolts are available through the PWROG (e.g., [41]).

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Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-10 SNC 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 offunctionality, as a result of loss of preload or material deformation. For FNP Unit 1, direct physical measurements are required only for the hold down spring.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.M16A [17] and Commitment 6 in the FNP Unit 1 SER.

5.5 GALL REVISION 2 ELEMENT 5: MONITORING AND TRENDING GALL Report AMP Element Description "The methods for monitoring,recording, evaluating, and trending the data that resultfrom the program 's'inspections are given in Section 6 of MRP-22 7 and its subsections. The evaluation methods include recommendationsfor flaw depth sizing andfor crack growth determinations as well for performing applicable limit load, linear elastic and elastic-plasticfracture analyses of relevantflaw indications. The examinations and re-examinations required by the MRP-227 guidance, together with the requirements specified in MRP-228for inspection methodologies, inspectionprocedures, and inspection personnel,provide timely detection, reporting, and corrective actions with respect to the effects of the age-relateddegradationmechanisms within the scope of the program. The extent of the examinations, beginning with the sample of susceptible P WR internals component locations identified as Primary Component locations, with the potentialfor inclusion of Expansion Component locations if the effects are greater than anticipated,plus the continuation of the ExistingPrograms activities, such as the ASME Code,Section XI, Examination Category B-N-3 examinationsfor core support sfructures, provides a high degree of confidence in the totalprogram" [17].

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

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Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-11 Inspections credited in Appendix B are based on utilizing the FNP Unit 1 10-year ISI program and the augmented inspections derived from MRP-227-A as documented in Appendix C. The MRP-227-A inspections only augment and do not replace the existing ASME Section XI 1SI requirements. These inspections, where practical, are scheduled to be conducted in conjunction with typical 10-year ISI examinations.

Tables C-i, C-2 and C-3 of this document identify the augmented Primary and Expansion inspection and monitoring recommendations, and the Existing programs credited for inspection and aging management.

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

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

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.MI6A [17] and Commitment 6 in the FNP Unit 1 SER.

5.6 GALL REVISION 2 ELEMENT 6: ACCEPTANCE CRITERIA GALL Report AMP Element Description "Section 5 of MRP-22 7 provides specific examination acceptance criteriafor the Primary and Expansion Component examinations. For components addressedby examinations referencedto ASME Code,Section XI,, the IWB-3500 acceptance criteriaapply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Programreference document.

The guidance in MRP-227 contains three types of examination acceptance criteria:

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

the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sizedfor length by VT-1/E VT-i examinations,

• For volumetric examination, the examination acceptance criterion is the capabilityfor reliable detection of indications in bolting, as demonstrated in the examination Technical Justification,"in addition, there are requirementsfor system-level assessment of bolted or WCAP- 1801 1-NP July 2015 E1-39 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-12 J

pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits;~and For physical measurements, the examination acceptance criterionfor the acceptable tolerance in the measured differential heightfrom the top of the plenum rib pads to the vessel seating surface in B& Wplants are given in Table 5-1 of MRP-22 7. The acceptance criterion for physical measurementsperformed on the height limits of the Westinghouse-designedhold down springs are requiredfor 304 SS hold down springs. FNP Unit 1 has a 304 SS hold down spring," therefore, FNP Unit 1 is requiredto produce acceptance criteriafor the physical measurements on the hold down spring" [17].

FNIP Unit 1 Acceptance Criteria Those recordable indications that are the result of inspections required by the existing FNP Unit 1 ISI program scope are evaluated in accordance with the applicable requirements of the ASME Code through the existing Corrective Action Program [26].

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

Augmented inspections, as defined by the MRP-227-A requirements included in this AMP as Table C-i1, Table C-2 and Table C-3, that result in recordable relevant conditions will be entered into the plant Corrective Action Program and addressed by appropriate actions that may include enhanced inspection, repair, replacement, mitigation actions or analytical evaluations. An example of an analytical evaluation is using an acceptable bolting WCAP approach, such as those commonly used to support continued component or assembly functionality. Additional analysis to establish acceptable bolting pattemn evaluation criteria for the baffle-former bolt assembly, as contained in various industry documents [41 ], 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 ofthese tools is the PWROG document WCAP-17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements" [11], which details acceptance criteria methodology for the MRP-227 Primary and Expansion components. Status is monitored through direct SNC 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-1 801,Section XI.M16A [17] and Commitment 6 in the FNP Unit 1 SER.

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Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westin* ihouse Non-Proprietary Class 3 5-13 5.7 GALL REVISION 2 ELEMEI '*T 7: CORRECTIVE ACTIONS GALL Report AMP Element Description "Corrective actionsfollowing the de 'tection of unacceptable conditions arefundamentally providedfor in each plant's correctii* ' action program.Any detected conditions that do not satis~fy the examination acceptance criteria* tre requiredto be dispositionedthrough the plant corrective actionprogram, which may require re piair, replacement, or analytical evaluationfor continued service until the next inspection. The L 'isposition will ensure that design basis functions of the reactor internals components will corn *inue to be fulfilledfor all licensing basis loads and events.

Examples of methodologies that can be used to analyticallydisposition unacceptable conditions arefound in the ASME Code, Section A "[orin Section 6 of MRP-22 7. Section 6 of MRP-22 7 describes the options that are available, for disposition of detected conditions that exceed the examination acceptance criteriaof Sect, ion 5 of the report. These include engineering evaluation methods, as well as supplementary exa* *inations to further characterize the detected condition, or the alternative of component repairand i *'eplacement procedures. The latter are subject to the requirements of the ASME Code,Section XI. The implementation of the guidance in MRP-227, plus the implementation of any ASME Co, de requirements,provides an acceptable level of aging management of safety-relatedcomponent* addressedin accordance with the corrective actions of 10 CFR Part50, Appendix B or its equiva, lent, as applicable.

Other alternative corrective action bases n. )ay be used to disposition relevant conditions if they have been previously approved or endorsec* ! by the NRC. Examples ofpreviously NRC-endorsed alternative corrective actions bases include those corrective actions basesfor Westinghouse-design RVI components that are defined in "itables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghouse Report No. WCAP-J 45 77-Rev . 1-A, orforB& W-designed RVl components in B&W Report No. BAW-2248. Westinghouse Reporr No. WCAP-14577-Rev. 1-A was endorsedfor use in an NRC SE to the Westinghouse Owners. Gro "ip, dated February 10, 2001. B&W Report No.

BA W-2248 was endorsedfor use in an SE to~* ,ramatome Technologies on behalfof the B& W Owners Group, dated December 9, 1999. Alte ':,native corrective action bases not approved or endorsed by the NRC will be submittedfor NA *'C approvalpriorto their implementation" [17].

FNP Unit 1 Corrective Action The existing FNP procedure for corrective actions, the" ' Corrective Action Program" [26] and the ASME Section XI ISI program [4], will be credited for this elen bent. These procedures establish the FNP Unit 1 repair and replacement requirements of ASME Code Sec tfion XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" [22]. These requirem, ents include the identification of a repair cycle and a repair plan, and verification of acceptability for rep lacements. FNP Unit 1 is committed to performing corrective actions for augmented inspections ,sing i repair and replacement procedures equivalent to those requirements in ASME B&PV Code, S iection XI [22] and MRP-227-A, Section 6 [5].

Conclusion This element complies with the corresponding aging manag ement attribute in NUREG-1801,Section XI.M16A [17] and Commitment 6 in the FNP Unit i!SER.

WCAP- 1801 1-NP July 2015 El1-41 Revision 0

Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-14 5.8 GALL REVISION 2 ELEMENT 8: CONFIRMATION PROCESS GALL Report AMP Element Description "Site quality assuranceprocedures, review and approvalprocesses, and administrative controls 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-22 7 will provide an acceptable level of qualityfor inspection, flaw evaluation, and other elements of aging management of the PWJ? internals that are addressedin accordance with the 10 CFR Part 50, Appendix B, or their 'equivalent (as applicable), confirmationprocess, and administrative controls" [17].

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

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [17] and Commitment 6 in the FNP Unit 1 SER.

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

FNP Unit 1 Administrative' Controls FNP Unit 1 has an established 10 CFR Part 50, Appendix B Program [28] that addresses the elements of corrective actions, confirmation process and administrative controls. QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B.

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [17] and Commitment 6 in the FNP Unit 1 SER.

WCAP-1801 1-NP July 2015 El1-42 Revision 0

Enclosure 1ito NL-15-1507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-15 5.10 GALL REVISION 2 ELEMENT 10: OPERATING EXPERIENCE GALL Report AMP Element Description "Relatively few incidents of PWR internals aging degradationhave been reported in operating U.S. commercialP WR plants. A summary of observations to date is provided in Appendix A of MRP-22 7-A. The applicantis expected to review subsequent operating experiencefor impact on its program or to participatein industry initiatives thatperform this fun ction.

The applicationof the MRP-22 7 guidance will establish a considerable amount of operating experience over the next few years. Section 7 of MiRP-227 describes the reportingrequirements for these applications, and the planfor evaluating the accumulatedadditional operating experience" [171.

FNP Unit 1 Operating Experience Extensive industry and FNP Unit 10GE has been reviewed during the development of the RVI AMP. The experience reviewed includes NRC Information Notices 84-18, "Stress Corrosion Cracking in PWR Systems" [29] and 98-11, "Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants"

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

Early plant GE related to hot functional testing and reactor internals is documented in plant historical records. Inspections performed as part of the 10-year 1S1 program have been conducted as designated by existing commitments, and would be expected to discover overall general internals structure degradation.

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

Industry GE is routinely reviewed by SNC engineers using Institute of Nuclear Power Operations (TNPO)

OE, the Nuclear Network, and other information sources as directed under the applicable procedure [3 1]

for the determination of additional actions and lessons learned.

A review of industry and plant-specific experience with RVI reveals that the U. S. industry, including SNC and FNP Unit 1, has responded proactively to industry issues relative to reactor internals degradation. Three examples that demonstrate this proactive response is the replacement of the Unit 1 control rod guide tube split pins in 2004, the replacement of baffle bolts in 1998, and the upflow conversion of reactor internals in 1982, which are briefly described in the following paragraphs.

FNP Unit 1 Control Rod Guide Tubes Support Pins In response to the industry concern for SCC of the alloy X-750 material, SNC replaced all of the upper internals guide tube support pins at FNP Unit 1 (October 2004) with Westinghouse-supplied, cold worked Type 316 S5S support pins to mitigate the possibility of continued SCC of these components. Detailed descriptions of the replacement are contained within the Field Change Notice [20], and documents referenced within, as well as the plant records [47].

WCAP-1801 I-NP July 2015 E1-43 Revision 0 ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 5-16 FNP Unit 1 Baffle Bolts During the Fall 1998 Outage, a proactive decision was made to replace a portion of the 1088 baffle former bolts in response to indications of cracking in 316 Type SS baffle-former bolts observed in a number of plants outside of the U.S. Detailed descriptions of the replacement are contained in the Field Change Notice [42], and documents referenced within, as well as the plant records [46].

During the same outage as the replacement campaign, FNP Unit 1 also conducted a UT examination of the original baffle-former bolt population. A total of 1086 of the 1088 baffle-former bolt population was examined with two bolts not inspected due to bolt head configuration. No indications were detected in the bolts examined [39]. In addition, the 211 baffle bolts removed were mechanically tested and six of these received further testing in the Westinghouse hot cell testing facility to advance industry irradiated material knowledge [40].

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

FNP Unit 1 Upflow Conversion In response to the fuel rod failures resulting from flow-induced vibration initiated by reactor coolant crossflow jetting through joints between baffle plates, several plants with Westinghouse-designed reactor internals were field modified to reverse the secondary coolant flow pattern in the baffle/barrel region, in order to reduce the jet-driving differential pressure. The original baffle/barrel region coolant flow pattern is known as "downflow" while the modified flow pattern is described as "upflow." Farley Unit 1 reactor internals have been modified for upflow conversion [43].

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG- 1801,Section XI.M16A [171 and Commitment 6 in the FNP Unit 1 SER.

WCAP- 1801 1-NP July 2015 E1-44 Revision 0 ito NL-1 5-1 507 FNP-1 RVJ Aging Management Program Westinghouse Non-Proprietary Class 3 6-1 6 DEMONSTRATION FNP Unit 1 has demonstrated a long-term commitment to aging management of reactor internals. This AMP is based on an established history of programs to identify and monitor potential aging degradation in the reactor internals. Programs and activities undertaken in the course of fulfilling that commitment include:

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

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

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

• The Water Chemistry Control Program at FNP has been effective in maintaining oxygen, halogens and sulfate at levels sufficiently low to prevent SCC, therefore maintaining structural integrity of the reactor vessel internals.

• Replacement control rod guide tube support pins for FNP Unit 1 in 2004 were fabricated from strain hardened austenitic type 316 stainless steel materials to increase resistance to SCC (versus original pins) [20].

• Replaced a portion of the 1,088 baffle former bolts during 1998 outage in response to indications of cracking in Type 316 S S baffle-former bolts observed in a number of plants outside of the U.S.

[42].

• Completed core power uprate for FNP Unit 1 in 1998 from 2652 MWt to 2775 MWt.

• Completed conversion of reactor internals coolant flow from "downflow" to "upflow" for Unit 1 in 1982.

• SNC has actively participated in past and ongoing EPRI and PWROG RVI activities. SNC 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, which is consistent with the ASMIE approach to inspections. This approach provides WCAP- 1801 1-NP July 2015 E1-45 Revision 0

Enclosure i to NL-15-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-2 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 are to be performed in the outage prior to entering the period of extended operation (Fall 2016, refueling outage 27 (RO-27). The previous ISI for FNP Unit 1 was performed in Fall 2007 (RO-2 1). The augmented inspections discussed in compliance with MRP-227-A requirements have been integrated in the implementation schedule, which is shown in Section 7. Integration of the required inspections will be tracked to completion. As discussed, the industry MRP-227-A guidelines also provide for updates as experience is gained through inspection results. This feedback loop will enable updates based on actual inspection experience.

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

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

WCAP- 1801 1-NP July 2015 E1-46 Revision 0

Enclosure i to NL-15-1E;07 FNP-1 RVI Aging Managiement Program.

Westinghouse Non-Proprietary Class 3 6-3 6.1 DEMONS1I'RATION OF TOPICAL REPORT CONDITIONS COMPLIANCE TO SE ON MRAP-227, REVISION 0 Table* 6-1. Topical Report Condition Compliance to SE on MIRP-227

~Applicable/

! Not Topical Con,'dition Applicable Compliance in AMP

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

_____________________________________guide tube (CRGT) lower flange weld.

2. Inspection of compo;oents subject Applicable The upper and lower core barrel cylinder girth to irradiation-assistedl stress welds and the lower core barrel flange weld corrosion cracking '.are moved from Table C-2 "Expansion Components" to Table C-i "Primary Components."

3. Inspection of high con.sequence Not Not applicable for FNP Unit 1.

components subject to :multiple Applicable degradation mechanisrn~s

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

components

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

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

7. Updating of MRP-227, Applicable Section 5 is updated to reflect XI.M 16A from Revision 0, Appendix A GALL Revision 2 [17].

WCAP- 1801 1-NP July 2015 E1-47 Revision 0 Ito NL-1 5-1 507 FNP-1 RVJ Aging Management Program Westinghouse Non-Proprietary Class 3 6-4 6.2 DEMONSTRATION OF APPLICANT/LICENSEE ACTION ITEM COMPLIANCE TO SE ON MRP-227, REVISION 0 6.2.1 SE Applicant/Licensee Action Item 1: Applicability of FMECA and Functionality Analysis Assumptions

"¶As addressedin Section 3.2.5.1 of this SE, each applicant/licenseeis responsiblefor assessing its plant's design and operatinghistory and demonstrating that the approved version of MRP-22 7 is applicable to thefacility. Each applicant/licenseeshall refer, in particular,to the assumptions regardingplant design and operatinghistory made in the FMECA andfunctionality analysesfor reactors of their design (i. e., Westinghouse, CE, or B& W) which support MRP-227 and describe the process usedfor determiningplant-specific differences in the design of their RVI components or plant operatingconditions, which result in different component inspection categories. The applicant/licenseeshall submit this evaluationfor NRC review and approvalas part of fits application to implement the approved version of MRP-227. This is Applicant/LicenseeAction Item 1" [5].

FNP Unit 1 Compliance The process used to verify that the RVI components at FNP Unit 1 are reasonably represented by the generic industry program assumptions (with regard to neutron fluence, temperature, stress values and materials used in the development of MRP-227-A [5]) is:

1. Identification of typical Westinghouse-designed pressurized water reactor (PWR) RVI components (MRP-191, Table 4-4 [9]).

2. Identification of FNP Unit I RVI components.

3. Comparison of the typical Westinghouse-designed PWR RVL components to the FNP Unit 1 RVI components identified in [23]:

a. Confirmation that no additional items were identified by this comparison (primarily supports A/LAI 2).

b. Confirmation that the materials for FNP Unit 1 are consistent with those materials identified in MRP-191, Table 4-4 [9].

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

4. Confirmation that the FNP Unit 1 operating history is consistent with the assumptions in MRP-227-A [5] regarding core loading patterns.

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

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

7. Confirmation that any changes to the FNP Unit 1 RVI components do not impact the application of the MIRP-227-A [5] generic aging management strategy.

WCAP- 1801 1-NP July 2015 E1-48 Revision 0 Ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-5 The FNP Unit 1 RVI components are reasonably represented by the design and operating history assumptions regarding neutron fluence, temperature, materials and stress values in the MRP- 191 [9]

generic FMECA and in the MRP-232 [33] functionality analysis based on the following:

1. FNP Unit 1 operating history is consistent with the assumptions in M\RP-227-A [5] with regard to neutron fluence and fuel management.

a. The FMECA and functionality analysis for MVRP-227-A [5] were based on the assumption of 30 years of operation with high-leakage core loading pattemns followed by 30 years of low-leakage core fuel management strategy. As stated in [1], FNP Unit I fuel management program changed from a high to a low-leakage core loading pattern prior to 30 years of operation. By operating with a low-leakage core design prior to 30 years, FNP Unit 1 meets the fluence and fuel management assumptions in MRP-191 [9] and requirements for MRP-227-A [5] application.

b. As stated in [1], FNP Unit 1 has always operated as a base load unit. Therefore, FNP Unit 1 satisfies the assumptions in material reliability program (MRP) documents regarding operational parameters affecting fluence.

2. The FNP Unit 1 reactor coolant system operates between TcoId and Thor [35, Table 5.1-1]. Teold is no lower than 530.6°F and Thor is no higher than 613.3°F [35, Table 5.1-1]. The design temperature for the vessel is 650°F [35, Table 5.4-1]. Therefore, FNP Unit 1 operating history is within original design basis parameters and is consistent with the assumptions used to develop the MRP-227-A [5] aging management strategy with regard to temperature operational parameters.

3. The FNP Unit I RVI components and materials are comparable to the typical Westinghouse-designed PWR RVI components (MRP-191, Table 4-4 [9]).

a. The components required to be in the FNP Unit 1 program [23] are consistent with those contained in MRP-191 [9]. No additional components are identified for FNP Unit 1.

b. FNP Unit I RVI component materials are consistent with, or equivalent to, those materials identified in MRP- 191, Table 4-4 [9] for Westinghouse- designed plants. The exceptions are:

the CRGT guide plates/cards; and the upper instrumentation conduit and supports - brackets, clamps, terminal blocks, and conduit straps, which are identified as having a material different than specified in MRP-191 and involve CF8. Several additional components have slightly different materials than those specified in MRP- 191; however, they have been determined to have no effect on the recommended MRP aging management inspection sampling strategy. These are dispositioned in the response to A/LAI 2.

c. FNP Unit 1 internals are the same, or equivalent to, the typical Westinghouse PWR internals regarding design and fabrication.

4. FNP Station is a two unit site with Westinghouse three-loop pressurized water reactors. A power uprate, in which the rated thermal power was increased from 2652 to the present 2775 megawatts thermal, has been implemented since initial commercial operation [43].

The guide tube assembly split pins were replaced in Unit 1 [20].

The vessel internals designs were converted from downflow to upflow [43].

A portion of the Unit 1 baffle former bolts were replaced during the fall 1998 outage [42].

WCAP- 18011-NP July 2015 E1-49 Revision 0 ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-6 SNC has not made any other modifications to the Unit 1 reactor internals components since May 2007 [1]. Therefore, modifications to the FNP Unit 1 RVI made over the lifetime of the plant are those specifically directed by the original equipment manufacturer (OEM). The OEM has developed or evaluated design changes and satisfied assumptions for A/LAI 1.

The design has been maintained over the lifetime of the plant as specified by the OEM, operational parameters are compliant with MIRP-227-A [5] requirements with regard to fluence and temperature, and the components are consistent with those considered in MRP-! 91 [9]. The materials for the components are consistent with those considered in MRP- 191 [9]. Therefore, the FNP Unit 1 RVI stress values are represented by the assumptions in MRP-191 [9], MRP-227-A

[5] and MRP-232 [33], confirming the applicability of the generic FMECA.

Conclusion The FNP Unit I evaluation for the AILAI I of the NRC SE regarding MRP-227, Revision 0 confirms that MRP-227-A is applicable to FNP Unit 1, with the exception of the material difference for the CRGT plates/cards. In order to address this material difference, an evaluation was completed, and documented in

[48], and concluded that the MRP-227-A aging management strategy for the CRGT guide plates is unchanged as a result of the potential for the component to be fabricated from CASS material.

6.2.2 SE Applicant/Licensee Action Item 2: PWR Vessel Internal Components within the Scope of License Renewal "As discussed in Section 3.2.5.2 of this SE, consistent with the requirements addressed in 10 CFR 54.4, each applicant/licensee is responsiblefor identifying which RVIlcomponents are within the scope of LR for its facility. Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identif whether these tables contain all of the RVI components that are within the scope of LR for their facilities in accordance with 10 CFR 54.4. If the tables do not identif all the RVIlcomponents that are within the scope of LR for itsfacility, the applicant or 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-specific AMP. The AMP shall provide assurance that the effects of aging on the missing component(s) will be managedfor the period of extended operation. This issue is Applicant/LicenseeAction Item 2" [5].

FNP Unit 1 Compliance This AiLAI requires comparison of the FNP Unit 1 RVI components that are within the scope of license renewal for FNP Unit I to those components contained in MIRP-191, Table 4-4 [9]. A detailed tabulation of the FNP Unit 1 RVI components [23] was completed, and it was compared to the typical Westinghouse PWR components in MRP-191, Table 4-4 [9]. All components required to be included in the FNP Unit 1 program are consistent with those contained in MRP- 191 [9].

Several components have different materials than that specified in MiRP- 191 [9] assessment.

WCAP- 1801 1-NP July 2015 El-5O Revision 0 ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-7 The potential for alternate materials, specifically CF8, to be used for the guide plates/cards was identified.

An evaluation of the material differences was completed and documented [48] which concluded that the MRP-227-A aging management strategy for CRGT guide plates is unchanged as a result as a result of the potential for the component to be fabricated from CASS material.

The upper instrumentation conduit and supports - bolting; upper instrumentation conduit and supports -

brackets, clamps, terminal blocks, and conduit straps are CF8. Using the FMECA process, the use of cast austenitic stainless steel (CASS) materials for the component: upper instrumentation conduit and supports

- brackets, clamps, terminal blocks, and conduit straps was evaluated. The FMECA concluded that the components could be classified as "No Additional Measures" based on a consideration of the likelihood of failure and the likelihood of damage. There is no change to the FNP Unit 1 MRP-227-A inspection requirements as a result of the inclusion of CF8 for these components (brackets, clamps, terminal blocks, and conduit straps).

Several additional components have slightly different materials (i.e., different grades of austenitic stainless steel) than those specified in MRP- 191; however, they have been determined to have no effect on the recommended MIRP aging management inspection sampling strategy.

This does not support the requirement that the AMP shall provide assurance that the effects of aging on the FNP Unit 1 RVI components within the scope of license renewal, but not included in the generic Westinghouse-designed PWR RVI components from MRP-191, Table 4-4 [9], will be managed for the period of extended operation. The inconsistent material for the guide plates/cards will be evaluated and it will be determined if it falls under the requirements for application of the MRP-227-A strategy for managing age-related material degradation in reactor internal components.

The generic scoping and screening of the RVI, as summarized in MRP-191 [9] and MiRP-232 [33], to support the inspection sampling approach for aging management of the RVI specified in MRP-227-A [5]

are not applicable to FNP Unit 1 with no modifications for the FNP Unit 1 components.

Conclusion The FNP Unit I evaluation for the A/LAI 2 of the NRC SE regarding MR!P-227, Revision 0 confirms that MRP-227-A is applicable to FNP Unit 1, with the exception of the material difference for the CRGT plates/cards. In order to address this material difference, an evaluation was completed, and documented in [48], and concluded that the MRP-227-A aging management strategy for the CRGT guide plates is unchanged as a result of the potential for the component to be fabricated from CASS material.

6.2.3 SE Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs "As addressed in Section 3.2.5.3 in this SE, applicants/licenseesof CE and Westinghouse are requiredto perform plant-specfifc analysis either to justify the acceptabilityoffan applicant's/licensee'sexisting programs, or to identify changes to the programs that should be implemented to manage the aging of these componentsfor the period of extended operation. The results of this plant-specific analyses and a description of the plant-specific programs being relied on to manage aging of these components shall be submitted as part of the WCAP- 18011-NP July 2015 El-5i Revision 0 ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-8 applicant's/licensee 'sAMP application. The CE and Westinghouse components identifedfor this type of plant-specfijc evaluation include: CE thermal shield positioningpins and CE in-core instrumentation thimble tubes (Section 4.3.2 in MRP-22 7), and Westinghouse guide tube support pins (split pins) (Section 4.3.3 in MRP-22 7). This is Applicant/Licensee Action Item 3" [5].

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

In response to the industry concern, the control rod guide tube support pins fabricated from INCONEL Alloy X-750 were replaced at FNP Unit 1 during the Fall 2004 outage; the replacement support pins utilized improved materials (strain hardened austenitic stainless steel) that support the proactive management of aging in reactor internals components. Detailed descriptions of the replacement are retained in the plant records [47].

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

6.2.4 SE Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief "As discussed in Section 3.2.5.4 of this SE, the B& W applicants/licenseesshall confirm that the core support structure upperflange weld was stress relieved during the originalfabrication of the Reactor Pressure Vessel in order to confirm the applicability of MRP-22 7, 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 andfrequency for non-stress relieved B& W core support structure upperflange welds shall be consistent with the recommendations in MRP-22 7, 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 criteria as described in Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee's resolution of this plant-specific action item shall be submitted to the NRC for review and approval. This is Applicant/LicenseeAction Item 4" [5].

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

Conclusion Applicant/Licensee Action Item 4 of the NRC SE on MRP-227, Revision 0 is not applicable to FNP Unit 1.

WCAP- 801 1-NP July 2015 E1-52 Revision 0 ito NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-9 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 addressed in Section 3.3.5 in this SE, applicants/licenseesshall identify plant-specifc acceptance criteriato be applied when performing the physical measurements requiredby the NRC-approved version of MRP-22 7for loss of compressibilityfor Westinghouse hold down springs, andfor distortion in the gap between the top and bottom core shroud segments in CE units with core barrelshrouds assembled in two vertical sections. The applicant/licenseeshall include its proposed acceptance criteriaand an explanation of how the proposed acceptance criteriaare consistent with the plants 'licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 5" [5].

FNP Unit 1 Compliance See Table 7-1. FNP Unit 1 utilizes a Type 304 SS hold down spring; therefore, SNC is planning to perform inspections/physical measurements on the FNP Unit 1 hold down spring according to MRP-227-A. SNC has an internal corrective action program tracking item to obtain the acceptance criteria for the hold down spring in advance of the outage in which measurements will be taken.

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

6.2.6 SE Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components "As addressedin Section 3.3.6 in this SE, MRP-22 7 does not propose to inspect the following inaccessible components: the B& W core barrel cylinders (including vertical and circumferential

'seamwelds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices, B& W core barrel-to-formerbolts and their locking devices, and B& W core barrelassembly internal baffle-to-baffle bolts. The MRP also identified that although the B&W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using curr'ently available examination techniques.

Applicants/licensees shalljustify the acceptability of these components for continued operation throu~gh the period of extended operation by performing an evaluation, or by proposing a scheduled replacement of the components. As part of their applicationto implement the approved version of MRP-227, applicants/licenseesshall provide theirjusti~ficationfor the continued operability of each of the inaccessible components and, if necessary, provide theirplan for the replacement of the components for NRC review and approval. This is Applicant/Licensee Action Item 6' [5].

WCAP- 18011-NP July 2015 E1-53 Revision 0

Enclosure 1Ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-10 FNP Unit 1 Compliance This Applicant/Licensee Action Item is not applicable to FNP Unit 1 since it only applies to B&W plants.

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

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 theirfacilities to demonstrate thatB& WIMI guide tube assembly spiders and CRGTspacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain theirfunctionality during the period of extended operation orfor additionalRVI components that may be fabricatedfrom CASS, martensitic stainless steel or precipitationhardenedstainless steel materials. These analyses shall also consider the possible loss of fracture 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 requirementmay not apply to components that were previously evaluated as not requiring aging management during development of MRP-22 7. That is, the requirement would apply to components fabricatedfrom susceptible materialsfor which an individual licensee has determined aging management is required,for example during their review performed in accordancewith Applicant/Licensee Action Item

2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licenseeshall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-22 7. This is Applicant/Licensee Action Item 7" [5].

FNP Unit 1 Compliance The NRC final safety evaluation (SE) on MRP-227, subsection 3.3.7 [5] states that, for assessment of cast austenitic stainless steel (CASS) materials, the applicant/licensee for license renewal may apply the criteria in the NRC letter of May 19, 2000, "License Renewal Issue No.98-003 0, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components" [36] as the basis for determining whether the CASS materials are susceptible to the thermal aging mechanism. If the application of the applicable screening criteria for the components material demonstrates that the components are not susceptible to either TE or IE, or to the synergistic effects of TB and IE combined, then no other evaluation would be necessary.

The FNP Unit 1 upper internals assembly - control rod guide tube (CRGT) - intermediate flanges, lower flanges and guide plates/cards are either wrought material or the alternate CASS material. The FNP Unit 1 upper instrumentation conduit and supports (stops, gussets, clamps, and support blocks) are also are either wrought material or alternate CASS material. For the evaluation to support LAI 7, it is conservative to assume they are CASS.

WCAP-1801 1-NP July 2015 E1-54 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-I11 The FNP Unit 1 mixing devices, upper support column assemblies - bases (mixer and orifice base) and bottom-mounted instrumentation (BMI) column assemblies - cruciform (standard and special) are CASS.

For each of the CASS components, the elemental percentages from the chemical data retrieved from certified material test reports (CMTRs) for the CASS component are input into Hull's formula (per guidance of NUREG/CR-45 13 [37]) to calculate the delta ferrite content of the CASS material. The CMTRs do not list the element percentage for nitrogen; thus, per the guidance of NUREG/CR-45 13, nitrogen is assumed to be 0.04 percent [37]. The CMTRs do not list an elemental percentage for molybdenum. A-35 1, Grade CF8 did not have a requirement for percent molybdenum in 1974. The 2013 Edition of the American Society of Mechanical Engineers (ASME) Code has SA-35 1, Grade CF8 chemistry requirements that specify' a maximum of 0.5 percent molybdenum; thus, this maximum value is input into Hull's formula. Where CMTRs were not located, a conservative combination of ASME A35 1, Grade CF8 chemical requirements was input into Hull's formula. The results of the TE evaluation for the FNP Unit 1 CASS components are summarized in Table 6-2.

Based on the criteria of the NRC letter dated May 19, 2000 [36]:

• The CRGT - intermediate flanges, lower flanges and guide plates/cards are considered as potentially susceptible to TE.

• The upper instrumentation conduit and supports (stops, gussets, clamps and support blocks) are considered as potentially susceptible to TE.

• 19 of the 23 mixing devices are not susceptible to TE and 4 of the 23 are potentially susceptible to TE.

• All of the upper support column - bases (mixing style) are not susceptible to TE.

• All of the upper support column - bases (orifice style) are not susceptible to TE.

• The BMI column cruciforms (standard) are not susceptible to TE.

• The BMI column cruciforms (special) are considered as potentially susceptible to TE.

All the above components were considered in MRP-191 and were screened for susceptibility to material degradation, including consideration of TE and JE. With the exception of the control rod guide tube -

guide plates/cards and the upper instrumentation conduit and supports (stops, gussets, clamps and support blocks), the above components were screened as CASS and considered for TE in MRP- 191. The assessments of the CASS CRGT guide plates/cards and the upper instrumentation conduit and supports (stops, gussets, clamps, and support blocks), taking into consideration their potential susceptibility to TE and their impact on the FNP aging management strategy, are discussed in the response to A/LAI 2.

No martensitic stainless steel (SS) or martensitic precipitation hardening (PH)-SS components were identified for the FNP Unit 1 reactor vessel internals.

WCAP- 1801 1-NP July 201!5 E1-55 Revision 0 Ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-12 Conclusion It is concluded that continued application of the MRP-227-A [5] strategy will meet the requirement for managing age-related degradation of the FNP Unit 1 CASS reactor vessel internals components.

Table 6-2. Summary of Joseph M. Farley Unit 1 CASS Components and Their Susceptibility to TE Susceptibility Molybdenum Ferrite to TE (Based CASS Component Content Casting Content on NRC Letter MRP-191 [9] Name Material (Percent) Method (Percent) [361)

Control Rod Guide Tube ASTM 0.5 Maximum Static Possible Potentially Assemblies and Flow A240 or > 20%(2 Susceptible( 2 )

Downcomers, alternate Flanges - intermediate ASTM A351, CF8 Control Rod Guide Tube ASTM 0.5 Maximum Static Possible Potentially Assemblies and Flow A240 or > 20%(2 Susceptible( 2 )

Downcomers, alternate Flanges - lower ASTM A351, CF8 Control Rod Guide Tube ASTM 0.5 Maximum Static Possible Potentially Assemblies and Flow A240 or > 20%(2 Susceptible( 2 )

Downcomers, alternate Guide plates/cards ASTM A3 51, Grade CF8 Mixing Devices ASTM 0.5 Maximum Static 19 of 23 19 of 23 Not A351, < 20%(') Susceptible(')

Grade CF8 4 of 23 4 of23 Possible Potentially

> 20%(2) Susceptible( 2 )

WCAP-1801 1-NP July 2015 E1-56 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-13 Table 6-2. Summary of Joseph M. Farley Unit 1 CASS Components and Their Susceptibility to TE (cont.)

Susceptibility Molybdenum Ferrite to TE (Based CASS Component Content Casting Content on NRC Letter MIRP-191 [91 Name Material (Percent) Method (Percent) [361)

Upper Instrumentation ASTM 0.5 Maximum Static Possible Potentially Conduit and Supports A240, > 20%(2 Susceptible( 2 )

(Thermocouple stops, A479, or*

supports, gussets and ASTM clamps) A351, Grade CF8 Upper Support Column ASTM 0.5 Maximum Static < 20%('* Not Assemblies, A351, Susceptible( 1

• Column bases Grade CF8 Upper Support Column ASTM 0.5 Maximum Static < 20%(') Not Assemblies, A351, susceptible('*

Mixer bases Grade CF8 Bottom-Mounted ASTM 0.5 Maximum Static < 20%(I Not Instrumentation (BMI) A35 1, susceptible(')

Column Assemblies, Grade CF8 column cruciform (standard cruciform)

Bottom-Mounted ASTM 0.5 Maximum Static Possible Potentially Instrumentation (BMI) A3 51, > 20%(2 Susceptible( 2 )

Column Assemblies, Grade CF8 column cruciform (special cruciform)

Notes:

1. Conclusion is based on CMTR chemistry data.

2. Where CMTR not located, conservative combination of ASME A351, Grade CF8 chemical requirements input into Hull's formula shows ferrite content can exceed 20 percent.

WCAP- 1801 1-NP July 2015 E1-57 Revision 0 Ito NL-I 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 6-14 6.2.8 SE Applicant/Licensee Action Item 8: Submittal of Information for Staff Review and Approval "As addressedin Section 3.5.1l in this SE, applicants/licenseesshall make a subm ittalfor NRC review and approval to credit their implementation ofJMRP-227, 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" [5].

FNP Unit 1 Compliance FNP Unit 1, per the RIS [3], is considered a Category B plant that is expected to submit their RVI AMP based on the guidance of MRP-227-A, which is consistent with their commitments. Per the LR SER [2],

FNP Unit 1 has a commitment to submit their AMIP for approval by the NRC no later than June 25, 2015.

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

WCAP- 1801 1-NP July 2015 E1-58 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program 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 inspection information and requirements from MRP-227-A and includes a description of 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 Outage Project Estimated Inspection Method and No. Outage/Year EFPY( 2 ) AMP-Related Scope") Criteria Comments 26 Spring 2015 31.19 Not Applicable Not Applicable Not Applicable 27 Fall 2016 32.58 Initial MRP-227-A MRP-227-A visual (EVT- 1) The initial inspection window for augmented inspections of inspection in accordance these components is no later than the upper and lower core with MiRP-228 two refueling outages from the barrel flange welds, and the specifications. beginning of extended operation.

upper and lower core barrel While the inspections are planned cylinder girth welds. for RO-27, FNP has the option to perform these inspections until RO-29.

These inspections are to be conduct~d in conjunction with a planned core barrel removal to perform cold leg DM Welds volumetric examinations and ASME Code Section XI Category B-N-3 visual examinations.

Extended period of operation begins at midnight on June 25, 2017.

WCAP- 1801 1-NP July 2015 E1-59 Revision 0

Enclosure i to NL-1 5-1 507 SFNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 7-2 Table 7-1. Aging Management Program Enhancement and Inspection Implementation Summary (cont.)

Refueling Project Estimated Inspection Method and Outage Month/Year EFPY<2 ) AMP-Related Scope0) ~ Criteria Comments 27 (cont.) Fall 2016 32.58 ASME Section XI 10 Year ASME Code Section XI 28 Spring 2018 33.98 Initial MRP-227-A Inspect and measure in The initial inspection window for augmented inspections of accordance with the guide plates (cards) is no later guide plates (cards). WCAP- 17451 requirements. than two refueling outages from the beginning of extended operation.

FNP has the option to perform these inspections until RO-29.

FNP Unit 1 can meet the requirements of both MRP-22 7-A and WCAP- 17451.

Initial MRP-227-A MRP-227-A inspections in Thintaispconwdwfr augmented inspections of accordance with MRP-228 the control rod guide tube lower control rod guide tube lower specifications, flange welds is no later than two flange welds. refueling outages from the beginning of extended operation.

FNP has the option to perform these inspections until RO-29.

Initial M1RP-227-A Direct measurement of hold The initial inspection window for augmented inspections of down spring, the hold down spring is within hold down spring. three cycles of the beginning of the license renewal period. While the inspection is planned for RO-28, FNP has the option to perform this inspection until RO-30.

WCAP- 1801 1-NP July 2015 El-S0 Revision 0 ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 7-3 Table 7-1. Aging Management Program Enhancement and Inspection Implementation Summary (cont.)

Refueling Project Estimated Inspection Method and Outage Month/Year EFPY*2 ) AMP-Related Scope") Criteria Comments 28 (cont.) Spring 2018 33.98 Initial MRP-227-A MRP-227-A inspections in The initial inspection window for augmented inspections for accordance with MRP-228 baffle-edge bolts and the baffle-edge bolts and the specifications. baffle-former assembly is between baffle-former assembly 20 and 40 EFPY. While the completed before or during inspections are planned for RO-28, this outage. FNP has the option to perform these inspections until RO-31I.

29 Fall 2019 35.37 Not Applicable Not Applicable Not Applicable 30 Spring 2021 36.77 Not Applicable Not Applicable Not Applicable 31 Fall 2022 38.16 Not Applicable Not Applicable Not Applicable 32 Spring 2024 39.56 Not Applicable Not Applicable Not Applicable WCAP- 1801 1-NP July 2015 El1-61 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 7-4 Table 7-1. Aging Management Program Enhancement and Inspection Implementation Summary (cont.)

Refueling Project Estimated Inspection Method and Outage Month/Year EFPY<2 ) AMP-Related Scope") Criteria Comments 33 Fall 2025 40.95 Initial MRP-227-A I\RP-227-A inspections in The initial inspection window for augmented inspections for accordance with MRP-228 the baffle-former bolts is between baffle-former boltsO) ~ specifications. 25 and 35 EFPY. The replacement completed before or during baffle bolts will be at this outage, approximately 25 EFPY at the time of inspection. A technical justification will document the acceptability of performing the inspection of the original bolts aged beyond 35 EFPY.

Subsequent MRP-227-A The subsequent inspection window augmented inspections of the MRP-227-A visual (EVT- 1) for these components is ten years upper and lower core barrel inspection in accordance after the initial inspection.

with MRP-228 flange welds, and the upper and lower core barrel specifications.

cylinder .girth welds.

ASME Section XI 10 Year ASME Code Section XI WCAP-18011-NP July 2015 E1-62 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 7-5 Table 7-1. Aging Management Program Enhancement and Inspection Implementation Summary (cont.)

Refueling Project Estimated Inspection Method and Outage Month/Year EFPY( 2 ) AMP-Related Scope(') Criteria Comments 34 Spring 2027 42.35 Subsequent MRP-227-A Inspect and measure in The subsequent inspection window augmented inspections of accordance with for these components is ten years guide plates (cards). WCAP- 17451 requirements. after the initial inspection.

Subsequent MRP-227-A MRP-227-A inspections in The subsequent inspection window augmented inspections of accordance with MRP-228 for these components is ten years control rod guide tube lower specifications, after the initial inspection.

flange welds.

Subsequent MRP-227-A MRP-227-A inspections in The subsequent inspection window augmented inspections for accordance with IVRP-228 for these components is ten years baffle-edge bolts and the spcfctosatethiniliseto.

baffle-former assemblyspcfctosafethiniliseto.

completed before or during this outage.

35 Fall 2028 43.74 Not Applicable Not Applicable Not Applicable 36 Spring 2030 45.14 Not Applicable Not Applicable Not Applicable 37 Fall 2031 46.53 Not Applicable Not Applicable Not Applicable 38 Spring 2033 47.93 Not Applicable Not Applicable Not Applicable 39 Fall 2034 49.32 ASME Section XI 10 Year ASME Code Section XI ISI(4) 40 Spring 2036 50.72 Not Applicable Not Applicable Not Applicable N/A N/A N/A Not Applicable Not Applicable Renewed Operating License

_________ _________________________________expires____Jepire Jun ,25,2037 WCAP- 1801 1-NP July 2015 E1-63 Revision 0 ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 7-6 Table 7-1. Aging Management Program Enhancement and Inspection Implementation Summary (cont.)

IRefueling Outage Project Month/Year EFPYt 2 )

Estimated AMP-Related Scope"*) Inspection Method and Criteria Comments Notes:

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

2. From [jI], FNP Unit I is at 27 EFPY during the Fall of 2010 and is anticipated to be at approximately 41 EFPY during Fall 2025. Therefore, each calendar year is the equivalent of 0.93 EFPY.

3. A portion of the baffle-former bolts were replaced during the Fall 1998 Outage. Therefore, at the time of the Fall 2025 outage the original baffle-former bolts will be at approximately 41 EFPY while the replacement baffle-former bolts will be at approximately 25 EFPY. A technical justification will document the acceptability of performing the MRP-227-A inspection during this outage with the original bolts aged beyond 35 EFPY.

4. ASME Section XI rules are followed for the In-Service Inspections, which allows for adjustment from the 10-year subsequent inspection requirement in order to align with a scheduled plant outage. The subsequent ASME Section XI inspection dates provided in this table could be adjusted as a result, but will comply with the Code.

WCAP- 1801 1-NP July 2015 E1-64 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVJ Aging Management Program Westinghouse Non-Proprietary Class 3 8-1 8 IMPLEMENTING DOCUMENTS As noted within this AMP document, the FNP Unit 1 PWR Vessel Internals Program is documented in [1].

The FNP Unit 1 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. SNC has also developed a fleet NDE procedure NMP-ES-024- 112 [38] "Materials Reliability Program (MRP) MRP-228 Implementation PWR RPV Internals Inspections" to establish a process for implementing the requirements of MIRP-228.

SNC documents associated with the existing FNP programs and considered to be implementing documents of the PWR Vessel Internals Program are:

• NMP-CH- 100-GL01, "Farley Primary Water Chemistry Strategic Plan" [ 18]

• FNP-0-SYP-22.0, "Flux Thimble Tube Examination Program" [19]

• NMP-ES-0 18, "ASME Section XI ISI Program" [4]

• NMP-ES-029, "PWR Primary System Integrity" [34]

• NMP-ES-024-1 12, "Materials Reliability Program (MRP) MRP-228 Implementation PWR RPV Internals Inspections" [38]

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

Based on this information, the AMP for FNP Unit 1 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- 18011-NP July 2015 E1-65 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES

1. Southern Nuclear Company Procedure, NMP-ES-029-GL02, "PWR Reactor Vessel Internals Program Strategic Plan," Version 3.0.

2. U.S. Nuclear Regulatory Commission, NUREG-1825, "Safety Evaluation Report Related to the License Renewal of Joseph M. Farley Nuclear Plant, Units 1 and 2," Docket Nos. 50-348 and 50-3 64, Southern Nuclear Operating Company, Inc., May 2005.

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

4. Southern Nuclear Company Procedure, NMP-ES-0 18, "SNC Inservice Inspection Engineering Program."

5. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-22 7-A). EPRI, Palo Alto, CA: 2011. 1022863.

6. U.S. Nuclear Regulatory Commission, Code of Federal Regulations, 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants," Washington D.C.,

Federal Register, Volume 77, No. 39907, dated May 8, 1995 and last updated on July 6, 2012.

7. U.S. Nuclear Regulatory Commission Document, NUREG-1800, Rev. 2, "Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plants (SRP-LR),"

December 2010.

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

9. Materials Reliability Program:"Screening, Categorization and Ranking of Reactor Internals Componentsfor Westinghouse and Combustion EngineeringPWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.

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

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

12. U.S. Nuclear Regulatory Commission Document, NUREG-1801, Rev. 0, "Generic Aging Lessons Learned (GALL) Report," July 2001.

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

14. Southern Nuclear Company Procedure, NMP-ES-009, "Engineering Programs."

15. Southern Nuclear Company Procedure, NMiP-GM-003, "Self-Assessments and Benchmark Procedure."

16. Southern Nuclear Company Procedure, NMP-ES-063-GL02, "Farley License Renewal Program Manual," Version 1.2, June 2015.

WCAP- 1801 1-NP July 2015 E1-66 Revision 0 ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 9-2

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

18. Southern Nuclear Company Procedure, NMP-CH-100-GL01, "Farley Primary Water Chemistry Strategic Plan."

19. Farley Nuclear Plant Procedure, FNP-0-SYP-22.0, "Flux Detector Thimble Inspection Program,"

Version 1.0, April 10, 2006. (Implemented via Preventative Maintenance Activity N1C56007)

20. Farley Unit 1 Field Change Notice, ALAO-405 77, "Control Rod Guide Tube /Probe Holder Shroud Support Pin Replacement," June 2005.

21. Farley Nuclear Plant Procedure, CGR-RVI-101, "Plant Farley License Renewal Commodity Group Review: Reactor Vessel Internals," Rev. 0, November 15, 2005.

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

23. SNC Report, "Joseph M. Farley License Renewal Application," September 2003 (ADAMS Accession Nos. ML032721356, ML032721360).

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

25. Pressurized Water Reactor Primary Water Chemistry Guidelines, Revision 7, EPRI, Palo Alto, CA:

2014. 3002000505.

26. Southern Nuclear Company Procedure, "Corrective Action Program," NMP-GM-002.

27. Farley Nuclear Plant Procedure, LR-2-12, "Plant Farley Commodity Review Procedure,"

Version 1.0.

28. Southern Nuclear Operating Company, Inc., "Quality Assurance Topical Report," Version 14.0, June 2, 2015.

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

30. U.S. Nuclear Regulatory Commission Information Notice, 98-1 1, "Cracking of Reactor Vessel Internal Baffle Former Bolts in Foreign Plants," March 25, 1998.

3 1. Southern Nuclear Company Procedure, "Operating Experience Program," NMP-GM-008.

32. Closed Cooling Water Chemistry Guideline, Revision 0, EPRI, Palo Alto, CA: 2013. 3002000590.

33. Materials Reliability Program."Aging Management Strategiesfor Westinghouse and Combustion EngineeringPWR Internal Components (MRP-232, Rev. 1). EPRI, Palo Alto, CA: 2012. 1021029.

34. Southern Nuclear Company Procedure, NMP-ES-029, "PWR Primary System Integrity Program,"

Version 8.1, March 6, 2014.

35. Updated Final Safety Analysis Report (UFSAR) FNP-FSAR-4, Rev. 26, December 2014.

36. U.S. Nuclear Regulatory Commission Letter, "License Renewal Issue No. 98-0030, "Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components," May 19, 2000.

(NRC ADAMS Accession No. ML003717179)

WCAP- 1801 1-NP July 2015 E1-67 Revision 0

Enclosure i to NL-15-1507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 9-3

37. U.S. Nuclear Regulatory Commission, NUREG/CR-45 13, Rev. 1, "Estimation of Fracture.

Toughness of Cast Stainless Steel During Thermal Aging in LWR Systems," August 1994.

(NRC ADAMS Accession No. ML052360554)

38. Southern Nuclear Company Procedure, NMP-ES-024-1 12, "Materials Reliability Program (MRP)

MRP-228 Implementation PWR RPV Internals Inspections."

39. Farley Inspection Report, EDRE-EMT-12 10, Rev. 0, "Analysis of the Mechanical Properties of CW 316 SS Baffle Bolts Removed from Farley Unit 1," 1999. (Westinghouse Proprietary Class 2)

40. Material Reliability Program: Hot Cell Testing of Baffle/Former Bolts Removed from Two Lead PWR Plants (MRP-51), November 2001. 1003069.

41. Westinghouse Report, WCAP- 15664, Rev. 0, "Determination of Acceptable Baffle-Barrel-Bolting for Three-Loop Westinghouse 1Sx1 5 Downflow and 17x1 7 Standard Upflow Domestic Plants,"

December 2001. (Westinghouse Proprietary Class 2)

42. Farley Unit 1 Field Change Notice, ALAO-40576, "Replacement Baffle Bolts," March 2000.

43. Westinghouse Report, WCAP- 14723, "Farley Nuclear Plant Units 1 and 2 Power Uprate Project

• NSSS Licensing Report," January 1997.

44. NRC SER on Parley Units 1 and 2 License Amendment No. 129, April 29, 1998. (ADAMS Accession No. ML01!2140259)

45. Westinghouse Report, WNEP-83 10, Rev. 0, "Specific Design Report of Reactor Internals Upflow Conversion," September 1983. (Westinghouse Proprietary Class 2)

46. Parley Nuclear Plant Design Change Package (DCP) No. 98-1-9344, "Baffle Former Bolt Replacement (Unit 1)."

47. Parley Nuclear Plant Design Change Package (DCP) No. 1039994301. (Unit 1 Split Pin Replacement)

48. Westinghouse Letter, LTR-RIAM-15-62, Rev. 0, "Transmittal of the Farley Unit 1 CASS Guide Plate Disposition Evaluation," July 29, 2015. (Westinghouse Proprietary Class 2)

WCAP-1801 1-NP July 2015 E1-68 Revision 0

Enclosure ito NL-1 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A-I FNP-1 RVI 1Aging Enclosure Management Program to NL-15-1507 APPENDIX A ILLUSTRATIONS HOUSING

~ROD TRAVEL INSTRUMENTATION CONTROL ROD - /*PORTS DRIVE MECHANISM THERMAL SLEEVE UPPER SUPPORT -

PLATE CLOSURE HEAD SUPPORT ASSEMBLY LEDGE

'//////*HOLD-DOWN SPRING INTERNALS \

CORE BARREL-

• CONTROL ROD GUIDE TUBE CONTROL ROD DRIVE SHAFT UPPER CORE--*

PLATE OUTLET NOZZLE-- INLET NOZZLE

• CONTROL ROD BAFFLE RADIAL *.

SUPPORT CLUSTER (WITHDRAWI

• ACCESS PORT INSTRUMENTATION REACTOR VESSEL THIMBLE GU[DES CORE SUPPORT -----

=PLATE Figure A-i. Illustration of Typical Westinghouse 4-Loop Plant Internals Assembly WCAP- 1801 1-NP July 2015 E1-69 Revision 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A-2 Wear Area Figure A-2. Typical Westinghouse Control Rod Guide Card WCAP- 18011I-NP July 2015 E1-70 Revision 0

Enclosure ito NL-15-l 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A-3 Upper Guide Tube Upper Support Plate Lower Guide tube I

Guide Cards Sheaths and C-Tubes Guide Tube Lower Flank Figure A-.3. Typical Lower Section of Control Rod Guide Tube Assembly WCAP- 1801 1-NP July 2015 El1-71 Revision 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A-4 Flarnge Weld Axial Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld Lower Barrel Axial Weld Lower Barrel Circumferential Wl SLower Barrel N.Axial Weld Core Barrel to Support Plate Weld Figure A-4. Major Core Barrel Welds WCAP- 18011I-NP July 2015 E1-72 Revision 0

Enclosure ito NL-l 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A-5 Cu Figure A-5. Bolting Systems used in Westinghouse Core Baffles WCAP-1801 1-NP July 2015 E1-73 Revision 0

--

• w Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program d

R Westinghouse Non-Proprietary w Class 3 A-6 Westinghouse Non-Proprietary Class 3 A-6 INTERNALS SUPPORT LEDGE THERMAL SHIE"LD BAFFLE FORMER.

LOWER CORE PLATE CORE SUPPORT COLUMN DIFFUSER PLATE CORE SUPPORT FORG ING Figure A-6. Core Baffle/Barrel Structure WCAP- 18011I-NP July 2015 E1-74 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A-7 DAFFLETOFORM OL*T Figure A-7. Bolting in a Typical Westinghouse Baffle-Former Structure WCAP- 1801 1-NP July 2015 E1-75 Revision 0

Enclosure ito NL-15-l 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A-8 Figure A-8. Vertical Displacement between the Baffle Plates and Bracket at the Bottom of the Baffle-Former-Barrel Assembly Figure A-9. Schematic Cross-Sections of the Westinghouse Hold Down Springs WCAP-1801 1-NP July 2015 E1-76 Revision 0

Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Pronrietarv Class 3 A-9 A-9 Figure A-tO. Typical Thermal Shield Flexure Note: Figure A- 10 is not applicable as the FNP Unit 1RVI design does not utilize a thermal shield.

WCAP- 1801 1-NP July 2015 E1-77 Revision 0

Enclosure 1ito NL-1 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proorietarv Class 3 A-Ill A~1fl Lower Core Plate Lower Core Support Structure Core Support Plate (Forging)

Figure A-lII. Lower Core Support Structure WCAP-fU180I1-NP July 2015 E1-78 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVJ Aging Management Program Westinghouse Non-Proprietary Class 3 A-ll

-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 Figure A-13. Typical Core Support Column WCAP-1801 1-NP July 2015 E1-79 Revision 0

Enclosure 1Ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 A- 12 a 0 Figure A-14. Examples of BMI Column Designs

.WCAP- 1801 1-NP July 2015 E1-80 Revision 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary class 3 B-i APPENDIX B FARLEY UNIT 1 LICENSE RENEWAL AGING MANAGEMENT REVIEW

SUMMARY

TABLE The content in Table B-i of Appendix B is extracted from Table 3.1.2-2 of the license renewal application approved by the NRC.

Table B-1. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA Aging Effect Requiring Component Type Management Aging Management Program01 )

1. Baffle and Former Plates Change in Material Reactor Vessel Internals Program (B.5. 1)

Properties

2. Baffle and Former Plates Cracking Reactor Vessel Internals Program (B.5. 1)

3. Baffle and Former Plates Cracking Water Chemistry Control Program (B.3.2)

4. Baffle and Former Plates Loss of Fracture Reactor Vessel Internals Program (B.5. 1)

Toughness

5. Baffle and Former Plates Loss of Material Water Chemistry Control Program (B.3.2)

6. Baffle Bolts Change in Material Reactor Vessel Internals Program (B.5. 1)

Properties

7. Baffle Bolts Cracking Reactor Vessel Internals Program (B.5.1)

8. Baffle Bolts Cracking Water Chemistry Control Program (B.3.2)

9. Baffle Bolts Loss of Fracture Reactor Vessel Internals Program (B.5. 1)

Toughness

10. Baffle Bolts Loss of Preload/ Inservice Inspection Program (B.3.1)

Stress Relaxation

11. Baffle Bolts Loss of Preload! Reactor Vessel Internals Program (B.5.1)

Stress Relaxation

12. Baffle Bolts Loss of Material Water Chemistry Control Program

_________________(B.3.2)

13. BMJ Column Cruciforms Cracking Water Chemistry Control Program

____ ___(B.3 .2)

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Enclosure i to NL-1 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 B-2 Table B-i. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA (cont.)

Aging Effect Requiring Component Type Management Aging Management Program"x)

14. BMI Column Cruciforms Cracking Reactor Vessel Internals Program (B.5.1)

15. BMI Column Cruciforms Loss of Fracture Reactor Vessel Internals Program (B.5.1)

Toughness

16. BMI Column Cruciforms Loss of Material Water Chemistry Control Program

____ ___(B.3 .2)

17. BMI Columns (with Cracking Water Chemistry Control Program associated fasteners) (B.3.2)

18. BMI Columns (with Cracking Reactor Vessel Internals Program (B.5.1) associated fasteners)

19. BMI Columns (with Loss of Material Water Chemistry Control Program associated fasteners) (B.3.2)

20. BMI Columns (with Loss of Preload / Inservice Inspection Program (B.3.1) associated fasteners) Stress Relaxation

21. Clevis Inserts and Fasteners Cracking Water Chemistry Control Program (B.3 .2)

22. Clevis Inserts and Fasteners Cracking Reactor Vessel Internals Program (B.5.1)

23. Clevis Inserts and Fasteners Loss of Material Inservice Inspection Program (B.3.1)

24. Clevis Inserts and Fasteners Loss of Material Water Chemistry Control Program (B.3.2)

25. Clevis Inserts and Fasteners Loss of Preload / Inservice Inspection Program (B.3.1)

Stress Relaxation

26. Control Rod Guide Tube Cracking Water Chemistry Control Program Assemblies (with associated (B.3.2) fasteners)

27. Control Rod Guide Tube Cracking Reactor Vessel Internals Program (B.5.1)

Assemblies (with associated fasteners)

28. Control Rod Guide Tube Loss of Material Water Chemistry Control Program Assemblies (with associated (B.3.2) fasteners)

29. Control Rod Guide Tube Loss of Preload/ Inservice Inspection Program (B.3.1)

Assemblies (with associated Stress Relaxation fasteners)

WCAP- 1801 1-NP July 2015 E1-82 Revision 0

Enclosure i to NL-15-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 B-3 Table B-1. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA (cont.)

Component Type Management Aging Management Program0)~

30. Core Barrel and Core Barrel Cracking Water Chemistry Control Program Flange(B.3 .2)

31. Core Barrel and Core Barrel Cracking Reactor Vessel Internals Program (B.5. 1)

Flange __________________

32. Core Barrel and Core Barrel Loss of Fracture Reactor Vessel Internals Program (B.5.1)

Flange Toughness _________________

33. Core Barrel and Core Barrel Loss of Material Water Chemistry Control Program Flange __________(B.3.2)

34. Core Barrel Outlet Nozzles Cracking Water Chemistry Control Program

__________ ________(B.3 .2)

35. Core Barrel Outlet Nozzles Cracking Reactor Vessel Internals Program (B.5.1)

36. Core Barrel Outlet Nozzles Loss of Material Water Chemistry Control Program

____ ____ ____ (B.3.2)

37. CRGT Support Pins Cracking Water Chemistry Control Program (B.3 .2)

38. CRGT Support Pins Cracking Reactor Vessel Internals Program (B.5. 1)

39. CRGT Support Pins Loss of Material Water Chemistry Control Program (B.3 .2)

40. CRGT Support Pins Loss of Preload/ Inservice Inspection Program (B.3.1)

Stress Relaxation

41. Flux Thimble Tubes Cracking Water Chemistry Control Program (B.3.2)

42. Flux Thimble Tubes Cracking Reactor Vessel Internals Program (B.5.1)

43. Flux Thimble Tubes Loss of Material Flux Detector Thimble Inspection Program (B.5.2)

44. Flux Thimble Tubes Loss of Material Water Chemistry Control Program (B.3.2)

45. HeadfRPV Alignment Pins Cracking Water Chemistry Control Program (with associated fasteners) (B.3 .2)

46. Head/RIPV Alignment Pins Cracking Reactor Vessel Internals Program (B.5. 1)

(with associated fasteners)

47. Head/RPV Alignment Pins Loss of Material Water Chemistry Control Program (with associated fasteners) (B.3 .2)

WCAP- 1801 1-NP July 2015 E1-83 Revision 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 B-4 Table B-i. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA (con t.)

Aging Effect Requiring Component Type Management Aging Management Program"l)

48. Head/RPV Alignment Pins Loss of Preload/ Inservice Inspection Program (B.3.1)

(with associated fasteners) Stress Relaxation

49. Head Cooling Spray Cracking Water Chemistry Control Program Nozzles (B.3 .2)

50. Head Cooling Spray Cracking Reactor Vessel Internals Program (B.5.1)

Nozzles

51. Head Cooling Spray Loss of Material Water Chemistry Control Program Nozzles (B.3 .2)

52. HJTC Probe Holder Cracking Water Chemistry Control Program Extension, and Probe (B.3 .2)

Holder Shroud Assemblies (with associated fasteners)

53. HJTC Probe Holder Cracking Reactor Vessel Internals Program (B.5.1)

Extension, and Probe Holder Shroud Assemblies (with associated fasteners)

54. HJTC Probe Holder Loss of Material Water Chemistry Control Program Extension, and Probe (B.3 .2)

Holder Shroud Assemblies (with associated fasteners)

55. HJTC Probe Holder Loss of Preload/ Inservice Inspection Program (B.3.1)

Extension, and Probe Stress Relaxation Holder Shroud Assemblies (with associated fasteners)

56. Internals Holddown Spring Cracking Water Chemistry Control Program

____ ___(B.3.2)

57. Internals Hoiddown Spring Cracking Reactor Vessel Internals Program (B.5. 1)

58. Internals Holddown Spring Loss of Material Water Chemistry Control Program (B.3 .2)

59. Internals Holddown Spring Loss of Material Inservice Inspection Program (B.3.1)

60. Internals Holddown Spring Loss of Preload/ Inservice Inspection Program (B.3.1)

Stress Relaxation

61. Lower Core Plate and Fuel Cracking Water Chemistry Control Program Alignment Pins (with (B.3 .2) associated fasteners)

WCAP- 1801 1-NP July 2015 E1-84 Revision 0 Ito NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 B-5 Table B-i. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA (co nt.)

Aging Effect Requiring Component Type Management Aging Management Program")

62. Lower Core Plate and Fuel Cracking Reactor Vessel Internals Program (B.5. 1)

Alignment Pins (with associated fasteners)

63. Lower Core Plate and Fuel Loss of Fracture Reactor Vessel Internals Program (B.5.1)

Alignment Pins (with Toughness associated fasteners)

64. Lower Core Plate and Fuel Loss of Material Water Chemistry Control Program Alignment Pins (with (B.3 .2) associated fasteners)

65. Lower Core Plate and Fuel Loss of Preload/ Inservice Inspection Program (B.3.1)

Alignment Pins (with Stress Relaxation associated fasteners)

66. Lower Support Columns Cracking Water Chemistry Control Program (with associated fasteners) (B.3.2)

67. Lower Support Columns Cracking Reactor Vessel Internals Program (B.5.1)

(with associated fasteners)

68. Lower Support Columns Loss of Material Water Chemistry Control Program (with associated fasteners) (B.3.2)

69. Lower Support Columns Loss of Preload / Inservice Inspection Program (B.3.1)

(with associated fasteners) Stress Relaxation

70. Lower Support Forging Cracking Water Chemistry Control Program (B.3.2)

71. Lower Support Forging Cracking Reactor Vessel Internals Program (B .5.1)

72. Lower Support Forging Loss of Fracture Reactor Vessel Internals Program (B.5.1)

Toughness

73. Lower Support Forging Loss of Material Water Chemistry Control Program (B.3 .2)

74. Neutron Panels (with Cracking Water Chemistry Control Program associated fasteners) (B.3 .2)

75. Neutron Panels (with Cracking Reactor Vessel Internals Program (B.5. 1) associated fasteners)

76. Neutron Panels (with Loss of Material Water Chemistry Control Program associated fasteners) ________(B.3 .2)

WCAP- 1801 1-NP July 2015 E1-85 Revision 0

Enclosure i to NL-15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 B-6 Table B-i. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA (cont.)

Aging Effect Requiring Component Type Management Aging Management Program"l)

77. Neutron Panels (with Loss of Preload/ Inservice Inspection Program (B.3.1) associated fasteners) Stress Relaxation

78. Radial Support Keys and Cracking Water Chemistry Control Program Fasteners (B.3 .2)

79. Radial Support Keys and Cracking Reactor Vessel Internals Program (B.5.1)

Fasteners

80. Radial Support Keys and Loss of Material Inservice Inspection Program (B.3. 1)

Fasteners

81. Radial Support Keys and Loss of Material Water Chemistry Control Program Fasteners (B.3 .2)

82. Radial Support Keys and Loss of Preload! Inservice Inspection Program (B.3.1)

Fasteners Stress Relaxation

83. Secondary Core Support Cracking Water Chemistry Control Program Assembly (with associated (B.3.2) fasteners)

84. Secondary Core Support Cracking Reactor Vessel Internals Program (B.5.1)

Assembly (with associated fasteners)

85. Secondary Core Support Loss of Material Water Chemistry Control Program Assembly (with associated (B .3.2) fasteners)

86. Secondary Core Support Loss of Preload/ Inservice Inspection Program (B.3.1)

Assembly (with associated Stress Relaxation fasteners)

87. Upper Core Plate Alignment Cracking Water Chemistry Control Program Pins (with associated (B.3 .2) fasteners)

88. Upper Core Plate Alignment Cracking Reactor Vessel Internals Program (B.5. 1)

Pins (with associated fasteners)

89. Upper Core Plate Alignment Loss of Material Inservice Inspection Program (B.3.1)

Pins (with associated fasteners)

WCAP- 1801 1-NP July 2015 E1-86 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 B-7 Table B-1. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA (cont.)

Aging Effect Requiring Component Type Management Aging Management Program")

90. Upper Core Plate Alignment Loss of Material Water Chemistry Control Program Pins (with associated (B.3 .2) fasteners)

91. Upper Core Plate Alignment Loss of Preload/ Inservice Inspection Program (B.3.1)

Pins (with associated Stress Relaxation fasteners)

92. Upper Core Plate and Fuel Cracking Reactor Vessel Internals Program (B.5.1)

Alignment Pins (with associated fasteners)

93. Upper Core Plate and Fuel Cracking Water Chemistry Control Program Alignment Pins (with (B.3.2) associated fasteners)

94. Upper Core Plate and Fuel Loss of Material Water Chemistry Control Program Alignment Pins (with (B.3 .2) associated fasteners)

95. Upper Core Plate and Fuel Loss of Preload/ Inservice Inspection Program (B.3.1)

Alignment Pins (with Stress Relaxation associated fasteners)

96. Upper Instrumentation Cracking Water Chemistry Control Program Conduit and Supports (with (B.3.2) associated fasteners)

97. Upper Instrumentation Cracking Reactor Vessel Internals Program (B.5. 1)

Conduit and Supports (with associated fasteners)

98. Upper Instrumentation Loss of Material Water Chemistry Control Program Conduit and Supports (with (B.3.2) associated fasteners)

99. Upper Instrumentation Loss of Preload/ Inservice Inspection Program (B.3.1)

Conduit and Supports (with Stress Relaxation associated fasteners) 100. Upper Support Assembly Cracking Water Chemistry Control Program (with associated fasteners) (B.3.2) 101. Upper Support Assembly Cracking Reactor Vessel Internals Program (B.5.1)

(with associated fasteners) 102. Upper Support Assembly Loss of Material Water Chemistry Control Program (with associated fasteners) (B.3.2)

WCAP- 18011-NP July 2015 E1-87 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 B-8 Table B-i. LRA Aging Management Review Summary Table 3.1.2-2 Farley Nuclear Plant LRA (cont.)

Aging Effect Requiring Component Type Management Aging Management Program(1) 103. Upper Support Assembly Loss of Preload/ Inservice Inspection Program (B.3.1)

(with associated fasteners) Stress Relaxation 104. Upper Support Column Cracking Water Chemistry Control Program Bases (B.3 .2) 105. Upper Support Column Cracking Reactor Vessel Internals Program (B.5.1)

Bases 106. Upper Support Column Loss of Fracture Reactor Vessel Internals Program (B.5.1)

Bases Toughness 107. Upper Support Column Loss of Material Water Chemistry Control Program Bases ______ _(B.3 .2) 108. Upper Support Columns Cracking Water Chemistry Control Program (with associated fasteners) (B.3.2) 109. Upper Support Columns Cracking Reactor Vessel Internals Program (B.5.1)

(with associated fasteners) 110. Upper Support Columns Loss of Material Water Chemistry Control Program (with associated fasteners) (B .3.2) 111. Upper Support Columns Loss of Preload/ Inservice Inspection Program (B.3.1)

(with associated fasteners) Stress Relaxation Notes

1. Information in parentheses are the Appendix B section numbers in the Farley LRA.

WCAP- 1801 1-NP July 2015 E1-88 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVl Aging Management Program Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C MRP-2 27-A AUGMENTED INSPECTIONS Table C-1. MIRP-227-A Primary Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Examination Item Applicability (Mechanism) Expansion Link(1) Method/Frequency(1) Examination Coverage Control Rod Guide All plants Loss of None Visual (VT-3) examination 20% examination of the Tube Assembly Material no later than 2 refueling number of CRGT Guide plates (cards) (Wear) outages from the beginning assemblies, with all guide of the license renewal period, cards within each and no earlier than two selected CRGT assembly refueling outages prior to the examined.

start of the license renewal - See Figure A-2 period. Subsequent examinations are required on a ten-year interval.

Control Rod Guide All plants Cracking Bottom-mounted Enhanced visual (EVT-l) 100% of outer Tube Assembly (SCC, Fatigue) instrumentation examination to determine the (accessible) CRGT lower Lower flange welds Aging (BMI) column presence of crack-like surface flange weld surfaces and Management bodies, Lower flaws in flange welds no later adjacent base metal on (IE and TE) support column than 2 refueling outages from the individual periphery bodies (cast), Upper the beginning of the license CRGT assemblies.(2 core plate, Lower renewal period and See Figure A-3 support subsequent examination on a

___________________________forging/casting ten-year interval.

WCAP- 1801 1-NP July 2015 E1-89 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-2 Table C-1. MRP-227-A Primary Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

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

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

Core Barrel Assembly All plants Cracking Upper and lower core Periodic enhanced visual 100% of one side of the Upper and lower core (SCC, IASCC, barrel cylinder axial (EVT-1) examination, no accessible surfaces of the barrel cylinder girth Fatigue) welds later than 2 refueling selected weld and welds outages from the beginning adjacent base metal( 4 )

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

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

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

WCAP- 1801 1-NP July 2015 E1-90 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RV! Aging Management Program Westinghouse Non-Proprietary Class 3 C-3 Table C-i. MRP-227-A Primary Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Effect Examination Item Applicability (Mechanism) Expansion Link(') Method/Frequency0)~ Examination Coverage Baffle-Former All plants Cracking None Visual (VT-3) examination, Bolts and locking devices Assembly with baffle- (IASCC, with baseline examination on high-fluence seams.

Baffle-edge bolts edge bolts Fatigue) that between 20 and 40 EFPY 100% of components Note: results in and subsequent accessible from core FNP Unit 1

• Lost or examinations on a ten-year side( 3 ).

has baffle broken interval. See Figures A-5, A-6, and edge bolts locking A-7 devices

• Failed or missing bolts

• Protrusion of bolt heads Aging Management (IE and ISR)( 6 )

Baffle-Former All plants Cracking Lower support Baseline volumetric (UT) 100% of accessible bolts Assembly (IASCC, column bolts, Barrel- examination between 25 and (Note 3). Heads Baffle-former bolts Fatigue) former bolts 35 EFPY, with subsequent accessible from the core Aging examination on a ten-year side. UT accessibility Management interval. may be affected by (IE and ISR)( 6 ) Note: Farley Unit 1 will complexity of head and perform a baseline locking device designs.

examination of the See Figures A-5 and A-6 replacement baffle-former bolts within this EFPY range.

WCAP- 18011-NP July 2015 E1-91 Revision 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-4 Table C-1. MRP-227-A Primary Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

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

Assembly Swelling), or distortion, with baseline See Figure A-8 (Includes: Baffle pats Cracking examination between 20 and bfeedeotsad(IASCC) that 40 EFPY and subsequent indirect effects of void results in: examinations on a ten-year swelling in former

• Abnormal interval.

plates) interaction with fuel assemblies

• Gaps along high fluence baffle joint

• Vertical displacemen t of baffle plates near high fluence joint

• Broken or damaged edge bolt locking systems along high fluence baffle joints __________

WCAP-18011-NP July 2015 E1-92 Revision 0

Enclosure i to NL-1 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Glass 3 C-5 Table C-i. MRP-227-A Primary Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

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

holrdongs derdto during the next two outages, See Figure A-9 304 SS 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 beginning of the license See Figures A-6 and A-10 Note: Material renewal period. Subsequent FNP Unit 1 (Wear) that examinations on a ten-year results in interval.

does not have temlsil a thermal feue shield, excessive wear, fracture, or complete separation WCAP- 18011-NP July 2015 E1-93 Revision 0 ito NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-6 Table C-1. MRP-227-A Primary Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Item I Applicability Effct (Mechanism) Expansion Link~'} Examination Method/Frequency0)~ Examination Coverage Notes:

l. 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.

WCAP- 18011-NP July 2015 E1-94 Revisioyi 0

Enclosure i to NL-l15-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-7 Table C-2. MiRP-227-A Expansion Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals Effect Examination Item Applicability (Mechanism) Primary Link(1) Method/Frequency") Examination Coverage Upper Internals All plants Cracking CRGT lower flange Enhanced visual (EVT-1) 100% of accessible Assembly (Fatigue, Wear) weld examination. surfaces( 2 ).

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

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

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

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

support forging_________________________

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

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

(IE, Void See Figure A-7 Swelling and

_____________________ISR)__ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

WCAP- 1801 I-NP July 2015 E1-95 Revision 0 to NL-1 5-1507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-8 Table C-2. MIRP-227-A Expansion Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Effect Examination Item Applicability (Mechanism) Primary LinkC') Method/Frequency0)~ Examination Coverage Lower Support All plants Cracking Baffle-former bolts Volumetric (UT) 100% of accessible bolts Assembly (IASCC, examination. or as supported by Lower support column Fatigue) Re-inspection every 10 years plant-specific bolts Aging following initial inspection. Justification( 2 ).

Management See Figures A-li1, A- 12 (IE and ISR) and A-13 Core Barrel Assembly All plants Cracking (SCC, Upper core barrel Enhanced visual (EVT-1) 100% of one side of the Core barrel outlet Fatigue) flange weld examination, accessible surfaces of the nozzle welds Aging Re-inspection every 10 years selected weld and Management following initial inspection, adjacent base metalC2 ).

(IE of lower See Figure A-4 sections)

Core Barrel Assembly All plants

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

Lower support column Aging Re-inspection every 10 years See Figures A-l1, A-12, bodies Management following initial inspection. and A-i13 (non cast) (IE)

WCAP- 1801 1-NP July 2015 E1-96 Revision 0 to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-9 Table C-2. MRP-227-A Expansion Component Inspection and Monitoring Recommendations for Westinghouse-Designed Internals (cont.)

Effect Examination Item Applicability (Mechanism) Primary Link( 1 ) Method/Frequency"l) Examination Coverage Lower Support All plants Cracking Control rod guide Visual (EVT-1) 100% of accessible Assembly Note: (IASCC) tube (CRGT) lower examination, support columns(2 ).

Lower support column FNP Unit 1 including the flanges Re-inspection every 10 years See Figures A1il, A-12, detection of floigiiiliseto, adA1 bodies lower support fractured floigiiiliseto. adA1 (cast) column bodies are non-cast. support columns Aging Management (LE)

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

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

Notes:

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

2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both the accessible and inaccessible portions).

WCAP- 18011-NP July 2015 E1-97 Revision 0

Enclosure i to NL-! 5-1!507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-IO0 Table C-3. MRP-227-A Components Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage Core Barrel Assembly All plants Loss of material ASME Code Visual (VT-3) examination All accessible surfaces at Core barrel flange (Wear) Section XI to determine general specified frequency.

condition for excessive wear.

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

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

Lower core plate Fatigue) detect evidence of distortion XL lower core plate0) ~ Aging and/or loss of bolt integrity.

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(')

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

WCAP- 1801 1-NP July 2015 E1-98 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-il Table C-3. MIRP-227-A Components Existing Inspection and Aging Management Programs Credited in Recommendations for Westinghouse-Designed Internals (cont.)

Effect Item Applicability (Mechanism) Reference Examination Method Examination Coverage Alignment and All plants Loss of material ASME Code Visual (VT-3) examination. All accessible surfaces at Interfacing (Wear)( 2 ) Section XI specified frequency.

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

Components Upper core plate alignment pins Notes:

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

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

WCAP- 1801 1-NP July 2015 E1-99 Revision 0

Enclosure i to NL-1 5-1 507 FNP-l RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-i12 Table C-4. MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals Item I Additional Examination Acceptance Criteria Control Rod Guide All plants Visual (VT-3) None N/A N/A Tube Assembly Examination Guide plates (cards) The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

WCAP- 1801 1-NP July 2015 E1-1 00 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Wetngos NnPrpitayCls C-i13 Table C-4. MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Acceptance Additional Examination Item Applicability Criteria(1 0 Expansion Link(s) Expansion Criteria Acceptance Criteria Control Rod Guide All plants Enhanced visual a. Bottom-mounted a. Confirmation of surface- a. For BMI column Tube Assembly (EVT- 1) instrumentation breaking indications in bodies, the specific Lower flange welds examination (BMI) column two or more CRGT relevant condition for The specific bodies lower flange welds, the VT-3 examination relevant b. Lower support combined with flux is completely condition is a column bodies thimble fractured column detectable (cast), upper insertion/withdrawal bodies.

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

b. Confirmation of surface- specific relevant breaking indications in condition is a two or more CRGT detectable crack-like lower flange welds shall surface indication.

require EVT- 1 examination of cast lower support column bodies, upper core plate and lower support forging/castings within three fuel cycles following the initial observation.

WCAP- 18011I-NP July 2015 El1-101 Revision 0

Endosure 1Ito NL-i 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-i14 Table C-4. MRP'-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Additional Examination Item Acceptance Criteria Expansion Criteria Core Barrel Assembly All plants Periodic a. Core barrel a. The confirmed detection a and b.

Upper core barrel enhanced visual outlet nozzle and sizing of a surface-(EVT-1) welds breaking indication with The specific relevant flange weld condition for the examination. b. Lower support a length greater than two expansion core barrel The specific column bodies inches in the upper core outlet nozzle weld relevant (non cast) barrel flange weld shall and lower support condition is a require that the EVT- 1 column body detectable examination be examination is a crack-like expanded to include the detectable crack-like surface core outlet nozzle welds surface indication.

indication. 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-1801 1-NP July 2015 E1-1 02 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-i15 Table C-4. MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

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

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

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

detectable examination be expanded to crack-like include the upper core barrel surface cylinder axial welds by the indication. completion of the next refueling outage.

WCAP- 1801 1-NP July 2015 El-i103 Revision 0

Enclosure i to NL-15-1507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-i16 Table C-4. MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Acceptance Additional Examination Item Applicability Criteria(0~ Expansion Link(s) Expansion Criteria Acceptance Criteria Core Barrel Assembly All plants Periodic Lower core barrel The confirmed detection and The specific relevant Lower core barrel enhanced visual cylinder axial welds sizing of a surface-breaking condition for the cylinder girth welds (EVT-1) indication with a length expansion lower core examination, greater than two inches in barrel cylinder axial weld The specific the lower core barrel examination is a relevant cylinder girth welds shall detectable crack-like condition is a require that the EVT-1 surface indication.

detectable examination be expanded to crack-like include the lower core barrel surface cylinder axial welds by the indication, 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 Note: relevant FNP Unit 1 conditions are has baffle missing or edge bolts broken locking devices, failed or missing bolts, and protrusion of bolt heads.

WCAP- 18011-NP July 2015 El1-104 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-I17 Table C-4. MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Acceptance Additional Examination Item Applicability Criteria(') Expansion Link(s) Expansion Criteria Acceptance Criteria Baffle-Former All plants Volumetric (UT) a. Lower support a. Confirmation that more a and b.

Assembly examination, column bolts than 5% of the baffle- The examination Baffle-former bolts The b. Barrel-former former bolts actually acceptance criteria for examination bolts examined on the four the UT of the lower acceptance baffle plates at the support column bolts criteria for the largest distance from the and the barrel-former UT of the core (presumed to be the bolts shall be baffle-former lowest dose locations) established as part of bolts shall be contain unacceptable the examination established as indications shall require technical justification.

part of the UT examination of the examination lower support column technical bolts within the next justification, 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- 1801 1-NP July 2015 El-1 05 Revision 0

Enclosure i to NL-1 5-i1507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-i18 Table C-4. MRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Acceptance Additional Examination Item Applicability Criteria~') Expansion Link(s) Expansion Criteria 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-1801 1-NP July 2015 El-i106 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-i19 Table C-4. MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

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

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

WCAP- 18011-NP July 2015 El-i107 Revision 0

Enclosure i to NL-1 5-1 507 FNP-1 RVI Aging Management Program Westinghouse Non-Proprietary Class 3 C-20 Table C-4. MIRP-227-A Acceptance Criteria and Expansion Criteria Recommendations for Westinghouse-Designed Internals (cont.)

Examination Acceptance Additional Examination Item Applicability Criteria~') Expansion Link(s) Expansion Criteria Acceptance Criteria Thermal Shield All plants Visual (VT-3) None N/A N/A Assembly with thermal examination.

Thermal shield flexures shields The specific Note: relevant FNP conditions for Unit Ildoes not thermal shield have a thermal flexures are shield. 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- 1801 1-NP July 2015 El-i108 Revision 0

Joseph M. Farley Nuclear Plant - Units 1 and 2 License Renewal Commitment Item 6 Enclosure 2 PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at J.M. Farley Nuclear Plant Unit 2