ML12278A100

Pressurized Water Reactor (PWR) Internals Aging Management Program Plan

ML12278A100
Person / Time
Site:	Palo Verde
Issue date:	09/07/2012
From:	Marthandam V Structural Integrity Associates
To:	Office of Nuclear Reactor Regulation, Arizona Public Service Co
References
500566453, Project 1200347 1200347.401
Download: ML12278A100 (139)

Text

Enclosure PWR Internals Aging Management Program Plan for Palo Verde Nuclear Generating Station Units 1, 2, and 3 Report No. 1200347.401 Revision 1 Project No. 1200347 September 2012 PWR Internals Aging Management Program Plan for Palo Verde Nuclear Generating Station Units 1, 2, and 3 Prepared for" Arizona Public Service Company Contract Order No: 500566453 Prepared by: Structural Integrity Associates, Inc.San Jose, California Prepared by: Date: 9/7/2012 Vikram Marthandam, Ph.D.Reviewed by: Approved by: Christopher Cruz Date: 9/7/2012 Date: 9/7/2012 Timothy J. Griesbach!jkSftud a ftww Ify Misacias, kIm?

REVISION CONTROL SHEET Document Number: 1200347.401 Title: PWR Internals Aging Management Program Plan for Palo Verde Nuclear Generating Station Units 1, 2, and 3 Client: Arizona Public Service SI Project Number: 1200347 Quality Program: F Nuclear Z Commercial Section Pages Revision Date Comments 1.0 1-1 16 0 8/27/12 Initial Issue 2.0 2-1-2-24 3.0 3-1 13 4.0 4-1 5 5.0 5-1 38 6.0 6-1 5 App. A A-1 -A-5 App. B B-1 -B-7 App. C C-1 -C-5 App. D D-1 -D-5 App. E E-1 -E-5 1.0 1-1 16 1 9/7/12 Comment Resolution 2.0 2-1 24 3.0 3-1 13 4.0 4-1 5 5.0 5-1 38 6.0 6-1 5 App. A A-1 -A-5 App. B B-1 -B-7 App. C C-1 -C-5 App. D D-1 -D-5 App. E E-1 -E-5 V~fG~Iu fwgdiy Auocatos, kIm?

Table of Contents Section Page

1.0 INTRODUCTION

.......................................................................................................

1-1 1.1 O bjectiv e .......................................................................................................................

1-1 1.2 PVNGS Reactor Vessel Internals Inspection Program Commitment

...........................

1-2 1.3 Palo Verde Reactor Vessel Internals Aging Management Program Background

........ 1-3 1.4 Palo Verde Reactor Vessel Internals Aging Management Program Elements .............

1-5 1.5 R esp on sib ilities .............................................................................................................

1-7 1.5.1 Palo Verde Nuclear Generating Station Engineering Programs Department

........ 1-7 1.5.2 Palo Verde Nuclear Generating Station Program Engineering

..............................

1-7 1.5.3 Palo Verde Nuclear Generating Station Chemistry

.................................................

1-8 1.6 Program Im plem entation ..............................................................................................

1-8 1.6.1 ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program .................

1-9 1.6.2 W ater Chem istry Program .......................................................................................

1-9 1.6.3 Reactor Coolant System Transient and Operating Cycles Procedure

.....................

1-9 1.7 Aging Management Review and Program Enhancements

..........................................

1-10 1.7.1 Reactor Internals Aging Management Review Process .........................................

1-10 1.8 Industry Program s .......................................................................................................

1-11 1.8.1 CE NPSD-1216, Aging Management of Reactor Internals

...................................

1-11 1.8.2 MRP-22 7-A, Reactor Internals Inspection and Evaluation Guidelines

.................

1-11 1.8.3 NEI 03-08 Guidance Within MRP-227-A

..............................................................

1-11 1.8.4 MRP-22 7-A AMP Development Guidance ............................................................

1-13 1.8.5 Ongoing Industry Programs ..................................................................................

1-15 1.9 Sum m ary .....................................................................................................................

1-15 2.0 AGING MANAGEMENT APPROACH ..................................................................

2-1 2.1 Mechanisms of Age-Related Degradation in PWR Internals

.......................................

2-1 2.1.1 Stress Corrosion Cracking .......................................................................................

2-1 2.1.2 Irradiation-Assisted Stress Corrosion Cracking .....................................................

2-2 Report No. 1200347.401.R1 iii v onkti fe itg MAsomtu, 2 .1.3 W ea r ........................................................................................................................

2 -2 2 .1.4 F a tig u e .....................................................................................................................

2 -2 2.1.5 Therm al Aging Em brittlem ent .................................................................................

2-3 2.1.6 Irradiation E m brittlem ent ........................................................................................

2-3 2.1.7 Void Swelling and Irradiation Growth ....................................................................

2-3 2.1.8 Thermal and Irradiation-Enhanced Stress Relaxation or Creep .............................

2-4 2.2 A ging M anagem ent Strategy ........................................................................................

2-4 2.3 Palo Verde Reactor Vessel Internals Aging Management Program Attributes

............

2-7 2.3.1 NUREG-1801/AMP Program Element 1: Scope of Program ..................................

2-8 2.3.2 NUREG-1801/AMP Program Element 2: Preventive Actions ...............................

2-10 2.3.3 NUREG-1801/AMP Program Element 3: Parameters Monitored/Inspected:

..... 2-11 2.3.4 NUREG-1801/AMP Program Element 4: Detection ofAging Effects: .................

2-13 2.3.5 NUREG-1801/AMP Program Element 5: Monitoring and Trending:

..................

2-15 2.3.6 NUREG-1801/AMP Program Element 6: Acceptance Criteria ............................

2-17 2.3.7 NUREG-1801/AMP Program Element 7: Corrective Actions: .............................

2-18 2.3.8 NUREG-1801/AMP Program Element 8: Confirmation Process .........................

2-20 2.3.9 NUREG-1801/AMP Program Element 9: Administrative Controls:

.....................

2-21 2.3.10 NUREG-1801/AMP Program Element 10: Operating Experience

.......................

2-21 3.0 PALO VERDE REACTOR VESSEL INTERNALS DESIGN AND OPERATING EXPERIEN CE ............................................................................................................

3-1 3.1 U pper G uide Structure A ssem bly .................................................................................

3-1 3.2 Core Support B arrel A ssem bly .....................................................................................

3-2 3.3 Lower Support Structure and Internals Nozzle Assembly ............................................

3-2 3.4 C ore Shroud A ssem bly ................................................................................................

3-3 3.5 Control Element Assembly Shroud Assemblies

...........................................................

3-4 3.6 PVNGS Units 1, 2, and 3 Design Distinctions

...........................................................

3-13 3.7 PVN G S U nit Operating Experience

...........................................................................

3-13 4.0 EXAMINATION AND ACCEPTANCE AND EXPANSION CRITERIA ...........

4-1 4.1 Exam ination A cceptance Criteria .................................................................................

4-1 Report No. 1200347.401.R1 iv v sk W h bteg ' Associates, 1W 4.1.1 Visual (VT-3) Examination

......................................................................................

4-1 4.1.2 Visual (VT-1) Examination

......................................................................................

4-2 4.1.3 Enhanced Visual (EVT-1) Examination

...................................................................

4-2 4.1.4 Surface Exam ination ................................................................................................

4-2 4.1.5 Volum etric Exam ination ..........................................................................................

4-3 4.1.6 Physical Measurements Examination

......................................................................

4-3 4.2 E xpansion C riteria ........................................................................................................

4-4 4.3 Evaluation, Repair, and Replacement Strategy.............................................................

4-4 4.3.1 R ep orting .............................................................................................................

.... 4-4 4.4 Im plem entation Schedule ..............................................................................................

4-5 4.5 C om m itm ent Tracking ..................................................................................................

4-5 5.0 RESPONSES TO NRC SAFETY EVALUATION APPLICANT/LICENSEE ACTION ITEMS .........................................................................................................

5-1 5.1 SE Section 4.2.1, Applicant/Licensee Action Item 1 (Applicability of FMECA and Functionality Analysis Assumptions):

..........................................................................

5-1 5.2 SE Section 4.2.2, Applicant/Licensee Action Item 2 (PWR Vessel Internal Components Within the Scope of License Renewal):

........................................................................

5-2 5.3 SE Section 4.2.3, Applicant/Licensee Action Item 3 (Evaluation of the Adequacy of Plant-Specific Existing Program s): .......................................................................................

5-3 5.4 SE Section 4.2.4, Applicant/Licensee Action Item 4 (B&W Core Support Structure Upper F lange Stress R elief): ....................................................................................................

5-3 5.5 SE Section 4.2.5, Applicant/Licensee Action Item 5 (Application of Physical Measurements as part of I&E Guidelines for B&W, CE, and Westinghouse RVI C om p on ents): ................................................................................................................

5-4 5.6 SE Section 4.2.6, Applicant/Licensee Action Item 6 (Evaluation of Inaccessible B&W C om pon ents): ...............................................................................................................

5-4 5.7 SE Section 4.2.7, Applicant/Licensee Action Item 7 (Plant-Specific Evaluation of CASS M aterials):

.....................................................................................................................

5 -5 5.8 SE Section 4.2.8, Applicant/Licensee Action Item 8 (Submittal of Information for Staff Review and Approval):

...........................................................................................

5-6 Report No. 1200347.40 1.R1 v v ambw IItt Assodctus, hW

6.0 REFERENCES

............................................................................................................

6-1 APPENDIX A SECTION XI 10 YEAR ISI EXAMINATIONS OF B-N-2 AND B-N-3 INTERNALS COMPONENTS FOR PVNGS [42] ...........................................

A-1 APPENDIX B AGING MANAGEMENT REVIEW PVNGS UNITS 1, 2, AND 3 [51 ........ B-1 APPENDIX C PVNGS UNIT 1 RVI SYSTEM, STRUCTURE, AND COMPONENT TABLES[121 ...............................................................................................................................

C -1 APPENDIX D PVNGS UNIT 2 RVI SYSTEM, STRUCTURE, AND COMPONENT TABLES 1121 ...............................................................................................................................

D -1 APPENDIX E PVNGS UNIT 3 RVI SYSTEM, STRUCTURE, AND COMPONENT TABLES[121 ................................................................................................................................

E -1 Report No. 1200347.401.R1 vi biitrw gnft AuSociatu, Inc.

List of Tables Table Page Table 1-1. Key Elements of the Reactor Vessel Internals Aging Management Program ...........

1-6 Table 5-1. C-E Plants Primary Category Components from Table 4-2 of MRP-227-A

[3] ..... 5-11 Table 5-2. C-E Plants Expansion Category Components from Table 4-5 of MRP-227-A

[3]. 5-17 Table 5-3. C-E Plants Existing Program Components Credited in Table 4-8 of M R P -227-A [3] ..................................................................................................................

5-20 Table 5-4. C-E Plants Examination Acceptance and Expansion Criteria from Table 5-2,3-fM RP-227-A [3] Applicable to PVNGS ........................................................

5-21 Table 5-5. PVNGS Response to the NRC Final Safety Evaluation of MRP-227-A

[4] ...........

5-26 Table 5-6. PVNGS Unit 1 Program Enhancement and Implementation Schedule ...................

5-29 Table 5-7. PVNGS Unit 2 Program Enhancement and Implementation Schedule ...................

5-32 Table 5-8. PVNGS Unit 3 Program Enhancement and Implementation Schedule ...................

5-35 Table 5-9. Summary of Actions Related to Aging Management of RVI for PV N G S U nits 1, 2, and 3 ...................................................................................................

5-38 Report No. 1200347.401.R1 vii V AssacIkItUs 1'

List of Figures Figure Figure 3-1., Figure 3-2.Figure 3-3.Figure 3-4.Figure 3-5.Figure 3-6.Figure 3-7.Figure 3-8.Page Illustration of the PVNGS Vessel and Internals

[38] ...............................................

3-5 PVNGS Reactor Vessel and Internals Assembly [38] .............................................

3-6 PVNGS Modified Upper Guide Structure Assembly [38] .......................................

3-7 PVNGS Core Support Barrel Assembly [38] .........................

3-8 PVNGS Core Shroud Assembly with Full Height Panels (with bottom-mounted IC I) [3 8 ] ..................................................................

.................................................

3 -9 (a) Illustration of a Typical C-E Fuel Alignment Plate, and (b) Radial view Schem atic Illustration of Guide Tubes ...................................................................

3-10 Isometric View of Lower Support Structure in the C-E Core Shroud with Full-H eight Shroud P lates ..............................................................................................

3-11 ICI Support Assembly (Palo Verde Units) [38] .....................................................

3-12 Report No. 1200347.401.Rl viii varnbw*~a hkftgl* Asokt ftW~

LIST OF ACRONYMS AMD Aging Management Document AMP Aging Management Program AMR Aging Management Review ARDM Age-related degradation mechanism ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel CASS Cast austenitic stainless steel C-E Combustion Engineering CEA Control Element Assembly CEOG Combustion Engineering Owners Group CFR Code of Federal Regulations CLB Current licensing basis EFPY Effective full power years EPRI Electric Power Research Institute EVT Enhanced visual testing (visual NDE method indicated as EVT-1)FMECA Failure modes, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC Irradiation Assisted Stress Corrosion Cracking ICI In-Core Instrumentation ISI Inservice Inspection ISR Irradiation-Enhanced Stress Relaxation LRA License Renewal Application MRP Materials Reliability Program MSC Materials Subcommittee NDE Nondestructive Examination NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OE Operating Experience PVNGS Palo Verde Nuclear Generating Station PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group RCS Reactor Coolant System RFO Refueling Outage RV Reactor Vessel RVI Reactor Vessel Internals Report No. 1200347.401.Rl ix VibW&9t l ly Asscltuw, /=P' SCC SE SER Ss TLAA TS UFSAR UT UGS VT Stress Corrosion Cracking Safety Evaluation Safety Evaluation Report Stainless Steel Time-limited Aging Analysis Technical Specifications Updated Final Safety Analysis Report Ultrasonic Testing Upper Guide Structure Visual Testing Report No. 1200347.401.Rl vItsfrcwIhilglyAs*~

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1.0 INTRODUCTION

1.1 Objective The purpose of this document is to describe the potential aging concerns in the reactor vessel internals (RVI) at Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3. This document also describes the mandatory and recommended guidance for managing potential aging concerns at PVNGS Units 1, 2, and 3 through the period of extended operation, which begins on June 1 2025 for Unit 1, April 24, 2026 for Unit 2, and November 25, 2027 for Unit 3.This Aging Management Program (AMP) document satisfies the license renewal and power uprate commitments as contained in the PVNGS license renewal application (LRA) [5]. This program coordinates with the ASME Section XI inservice inspection (ISI) program and supplements that program with augmented examinations for managing the potential aging effects of the RVI. This program plan establishes appropriate monitoring and inspections to maintain the reactor vessel internals functionality; the strategy is to assure nuclear safety and plant reliability.

This document provides assurance that operations at PVNGS Units 1, 2, and 3 will continue to be conducted in accordance with the current licensing bases (CLB) for the RVI, and it will provide the technical basis for managing the time-limited aging concerns for the duration of the plant by fulfilling the license renewal and power uprate commitments.

This document identifies the internals components that must be considered for aging management review and identifies the augmented inspection plan for PVNGS Units 1, 2, and 3 reactor vessel internals.

The program plan supports the NEI 03-08 Materials Initiative Process [ 1], the NEI 03-08 Guideline for the Management of Materials Issues [2], the EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A

[3]), and the Applicant/Licensee Action Items in the SE [4].The main objectives of the PVNGS RVI AMP are:* To demonstrate that the effects of aging on the RVI will be adequately managed for the period of extended operation in accordance with 10 CFR Part 54.Report No. 1200347.401 .R1 1-1 twbwh negrekty Assackles, W

• To summarize the role of existing PVNGS AMPs related to aging management of RV internals.

• To define and implement the industry-defined (EPRI/MRP and PWROG) requirements and guidance for managing aging of RV internals.

• To provide inspection plans for the PVNGS RV internals.

1.2 PVNGS Reactor Vessel Internals Inspection Program Commitment In order to meet the license renewal [5] and power uprate commitments

[6, 9], PVNGS will submit this aging management program plan. The license renewal and power uprate commitments listed below define the content and timeline for the program that PVNGS has committed to implement for the RVI Components:

By Letter No. 102-06423, dated October 11,2011 (ADAMS Accession No. ML11297A118)

[7], Arizona Public Service Company (APS) submitted to the Nuclear Regulatory Commission (NRC) the following updated commitment to submit a reactor vessel internals inspection program: APS will submit the PVNGS Units 1, 2, and 3 reactor vessel internals aging management program and inspection plans in accordance with MRP-22 7-A no later than October 1, 2012, for NRC review and approval.The Aging Management Program Units 1, 2, and, 3 has been established so that the aging effects of the RVI components are adequately managed and to provide reasonable assurance that the internals components will continue to perform their intended function through the license renewal period of extended operation.

Furthermore, this AMP will demonstrate the consistency of the program with the elements documented in NUREG- 1801, Revision 2 [10], Chapter XI.M. 16A, "PWR Vessel Internals." The operating experience provided by NUREG-1801, Revision 2 [10] will also be reviewed and incorporated into plant-specific programs.Report No. 1200347.401.R1 1-2 Wl ft I 1W' 1.3 Palo Verde Reactor Vessel Internals Aging Management Program Background The managing of aging degradation effects in RVI is required for nuclear plants considering or entering license renewal, as specified in the NRC Standard Review Plan (SRP) for License Renewal Applications

[ 11]. The U.S. nuclear industry has been actively engaged in supporting the industry goal of responding to these requirements.

Various programs have been established within the industry over the past decade to develop guidelines for managing the aging effects of PWR RV internals.

In 2001, Combustion Engineering Owners Group (CEOG) issued CE NPSD- 1216 "Generic Aging Management Review Report for the Reactor Vessel Internals

[ 12]." Later, in 2008, MRP-227, Revision 0 was published by EPRI MRP to address the PWR vessel internals aging management issue for the three currently operating U.S. PWR designs, namely, Combustion Engineering (C-E), Westinghouse, 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[13 -15]: 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.

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, to components that are moderately susceptible to the aging effects, to components that are significantly susceptible to the aging effects.Functionality assessments were performed to determine the effects of the degradation mechanisms on component functionality.

These assessments were based on representative plant designs of PWR internals components and assemblies of components using irradiated and aged material properties.

Report No. 1200347.401.Rl 1-3 VjItIiCIIrWE , 11W Aging management strategies for implementing the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections were developed.

Development of these strategies was based on combining the results of functionality assessment with several contributing factors including component accessibility, operating experience, existing evaluations, and prior examination results.The industry efforts, as coordinated by the EPRI MRP, has finalized the inspection and evaluation (I&E) guidelines for the RVI, and the NRC has endorsed this document by issuing a safety evaluation (SE). A supporting document addressing inspection requirements has also been completed.

The industry guidance is contained in the following documents:

• Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) provides the industry background, listing of reactor vessel internal components requiring inspection, the type or types of nondestructive examination (NDE) required for each component, timing for initial inspections, and criteria for evaluating inspection results. The NRC has endorsed MRP-227-A by issuing a safety evaluation (SE) [4].MRP-228 [14], "Inspection Standard for PWR Internals," provides guidance on the qualification and demonstration of the NDE techniques and other criteria pertaining to the actual performance of the inspection.

The PWROG has developed and submitted for NRC review and approval WCAP-17096,"Reactor Internals Acceptance Criteria Methodology and Data Requirements" [ 17] for MRP-227-A inspections, where feasible.

Plant specific acceptance criteria can also be developed for some internals components if a generic approach is not practical.

PVNGS Units 1, 2, and 3 RVI are a part of the primary reactor coolant system (RCS), which is a two-loop C-E designed nuclear steam supply system (NSSS).Report No. 1200347.401 .Rl 1-4 M M-l* VIris' A review of Section 2.3.1.1 of the PVNGS LRA specifies that the RVI are comprised of the following component groups:* Core Support Structure (CSS): o Core support barrel (CSB) assembly o Lower support structure assembly o Core shroud assembly" Upper Guide Structure (UGS): o UGS support barrel assembly o UGS Control Element Assembly (CEA) shroud assembly o UGS holddown ring* Flow Skirt (perforated with flow holes) and reinforced with two stiffening rings.* In-core instrumentation support structures 1.4 Palo Verde Reactor Vessel Internals Aging Management Program Elements The key elements of the PVNGS Reactor Vessel Internals Aging Management Program are outlined in Table 1-1. The program attributes are described in detail in Section 2.3 of this document.

Additionally, PVNGS participates in PWR Owners Group Materials Subcommittee (PWROG MSC) and the MRP to focus on preventing material degradation, improve plant performance, sharing lessons learned from operating experience, and provide an effective interface with the NRC. As RVI examination experiences are shared amongst other utilities, MRP, and PWROG MSC, the RVI AMP key elements will be updated to include any relevant OE or lessons learned.Report No. 1200347.401 .R 1-5 C AD taw Table 1-1. Key Elements of the Reactor Vessel Internals Aging Management Program Plan Attribute Attribute Description I Scope of Program The scope of this AMP is MRP-227-A

[3] and the SE for MRP-227, Rev. 0 [4].Supplemental inspections of RV internals are described in MRP-227-A

[3].Additional actions and long range plans for aging management of internals are defined within this document.

The scope of the program is described in more detail in Section 2.3.1 of this document.2 Preventive Actions Preventive measures are described in Section 2.3.2 of this document.3 Parameters PVNGS monitors, inspects, and/or tests for the effects of the eight aging Monitored/Inspected degradation mechanisms on the intended function of the reactor vessel internals components as described in Section 2.3.3 of this document.4 Detection of Aging The PVNGS ASME Section XI [21] ISI program for B-N-2 and B-N-3 Effects internals components (Appendix A), and the additional locations identified in MRP-227-A

[3], form the inspection plan for detection and monitoring of aging effects in the RV internals as described in Section 2.3.4 of this document.5 Inspection Program for This program, in combination with the ASME Section XI [21] ISI program, Monitoring and Trending provides direction for inspections required to support continued RV internals component reliability as described in Section 2.3.5 of this document.6 Acceptance Criteria Acceptance criteria used in the RV Internals Aging Management Program are based on the most appropriate ASME Section XI [21 ] and WCAP-17096

[17]criteria as described in Section 2.3.6 of this document.7 Corrective Actions Components with identified relevant conditions shall be dispositioned as described in Section 2.3.7 of this document.

The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition until the next planned inspection, or repair/replacement to remediate the relevant condition.

Additional inspections of expansion category components may also be required as specified in MRP-227-A.8 Confirmation Process The confirmation process for the RV Internals Program is described in Section 2.3.8 of this document.9 Administrative Controls Administrative controls that apply to the RVI AMP, procedures, reviews and approval processes is described in Section 2.3.9 of this document 10 Operating Experience Operating experience related to the PVNGS RV internals is described in Section 2.3.10 of this document.Report No. 1200347.401.R1 1-6 V U~SW~fJU iWgrfty Assacites,1IW.

1.5 Responsibilities The RVI Aging Management Program at PVNGS is guided by the PVNGS Outage Planning and Implementation Procedure

[22]. Implementation and comprehensive long term management of the RVI AMP will require the integration of Arizona Public Service Company's (APS) corporate organization, and interaction with multiple industry organizations including ASME, MRP, NRC, and PWROG. The responsibilities of PVNGS groups are provided in the following paragraphs.

PVNGS will maintain cognizance of industry activities related to PWR internals inspection and aging management and will address and implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices.

The overall responsibility for administration of RVI AMP rests with PVNGS Senior Management.

1.5.1 Palo Verde Nuclear Generating Station Engineering Programs Department The Engineering Department at PVNGS is responsible for providing governance and oversight of the implementation of the RVI AMP, including:

• Ensuring that appropriate programs are established and maintained to support inspection and mitigation activities for the RVI.* Providing oversight of plant implementation of inspection and mitigation activities.

• Maintaining cognizance of industry activities related to PWR internals inspection and aging management.

1.5.2 Palo Verde Nuclear Generating Station Program Engineering PVNGS Engineering is responsible for the overall development and implementation of the PWR Internals Aging Management Program, including:

Report No. 1200347.401.Rl 1-7 IOW ur UUV ASSOCtU, I

• Planning and implementation of the RVI AMP mitigation, inspection, and repair activities, as approved by site senior management.

• Reviewing and approving of vendor programs involved in various RVI activities.

• Ensuring that required inspections and supporting activities are implemented in the time period specified.

1.5.3 Palo Verde Nuclear Generating Station Chemistry The PVNGS System Chemistry Program is responsible for:* Maintaining primary water chemistry in accordance with approved PVNGS procedures

[23]and specifications.

'* Ensuring that the PVNGS documents supports and incorporates the guidance of industry programs including but not limited to the EPRI Water Chemistry Guidelines

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

1.6 Program Implementation PVNGS's overall strategy for managing aging in reactor vessel internals components is supported by the following existing programs:* ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program* Water Chemistry Program* Reactor Coolant System Transient and Operating Cycles Program These are established programs that support the aging management of RCS components in addition to the RVI components.

Report No. 1200347.401.Rl 1-8 c ILluI/WU' , Anadates, 1W 1.6.1 ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program The ASME Section XI IWB, IWC, IWD, IWF Inservice Inspection Program [21] is an existing program that facilitates inspections to identify degradation in Class 1, 2 and 3 piping, components, supports, and integral attachments.

The program includes periodic visual, surface and/or volumetric examinations and leakage tests of all Class 1, 2 and 3 pressure-retaining components, their supports and integral attachments.

Inspections of removable core support structures (Category B-N-3) are included in this existing program. These are identified in ASME Section XI [21 ], "Rules for Inservice Inspection of Nuclear Power Plant Components." 1.6.2 Water Chemistry Program The Water Chemistry Program [23] is an existing program that is credited for managing aging effects by controlling the environment to which internal surfaces of systems and components are exposed. Such effects include: " Loss of material due to general corrosion, pitting, and crevice corrosion* Cracking due to SCC* Other degradation such as intergranular attack, steam generator tube degradation and outer diameter stress corrosion cracking The aging effects are minimized by controlling the chemistry and by managing the causes of the underlying environmental degradation mechanisms.

This program minimizes the occurrences of these aging effects and contributes to maintaining each component's ability to perform the intended functions.

This is in accordance with the EPRI PWR Primary Water Chemistry Guidelines

[24].1.6.3 Reactor Coolant System Transient and Operating Cycles Procedure The PVNGS Reactor Coolant System Transient and Operating Cycles Procedure

[29] maintains the components in the UFSAR Section 3.9.1.1 within the cyclic or transient limits in the Report No. 1200347.401 .R1 1-9 V 'utwt*rAss

/t,1 UFSAR [30]. The procedure incorporates a cycle counting based approach for monitoring fatigue usage in plant components.

It is applicable for both the pressure boundary and component support locations for which monitoring or projecting fatigue usage is important for managing fatigue or component life. Several components in the reactor vessel internals for PVNGS have calculated fatigue usage factors in accordance with the ASME Section III, Subsection NG design.1.7 Aging Management Review and Program Enhancements

1. 7.1 Reactor Internals Aging Management Review Process A comprehensive review of aging management of RVI was performed as part of the PVNGS license renewal application

[5]. The AMR performed for the LRA submittal documents the results of the aging management review for the PVNGS RVI. The NRC indicated its approval of the PVNGS LRA in NUREG- 1961 [31]. The RVI components specifically noted as requiring aging management, as identified in the LRA, are summarized in Appendix B, Table B- 1 of this document.The assessments supporting the LRA performed the following:

1. Identified applicable aging effects requiring management

2. Associated aging management programs to manage those aging effects 3. Identified enhancements or modifications to existing programs, new aging management programs, or any other actions required to support the conclusions reached in the assessment AMRs were performed for each PVNGS system that contained long-lived, passive components requiring an aging management review, in accordance with the PVNGS screening process. The results of these reviews have been incorporated into the PVNGS RVI AMP.Report No. 1200347.401.Rl 1-10 V Oitugrit, Associates, 1W 1.8 Industry Programs 1.8.1 CE NPSD-1216, Aging Management of Reactor Internals The Combustion Engineering Owner's Group (CEOG) topical report CE NPSD-1216

[12]contains a technical evaluation of aging degradation mechanisms and aging effects for C-E RVI components.

The CEOG report provided guidance for CEOG member plant owners to manage effects of aging on RVI during the period of extended operation, using approved aging management methodologies to develop plant-specific AMPs.The AMR for the Palo Verde internals, documented in the PVNGS license renewal application

[5], was completed in a manner consistent with the approach of CE NPSD-1216

[12]. Both the Palo Verde specific AMR document and the generic C-E document were completed to facilitate plant license renewal in accordance with 10 CFR Part 54 [5].1.8.2 MRP-22 7-A, Reactor Internals Inspection and Evaluation Guidelines MRP-227-A was developed by a team of industry experts including utility representatives, NSSS vendors, independent consultants, and international representatives who reviewed available data and industry experience on materials aging. The objective of this project was to develop a consistent, systematic approach for identifying and prioritizing inspection requirements for reactor vessel internals.

1.8.3 NEI 03-08 Guidance Within MRP-227-A The industry program requirements of MRP-227 are classified in accordance with the requirements of the NEI 03-08 [2] protocols.

The MRP-227-A

[3] guideline includes"mandatory," "needed," and "good practice" requirements defined as the following:

Mandatory Each commercial U.S. PWR unit shall develop and document a PWR reactor internals aging management program within 36 months following issuance of MRP-227, Rev. 0.Report No. 1200347.40 L.R 1-11 ja hweduitr$

Assomters, 1W PVNGS Applicability:

MRP-227 was officially issued by the industry in December 2008 [16]. An aging management program was to be developed by December 2011. In order to meet this "Mandatory" requirement, an aging management program plan for PVNGS was completed in November 2011.Originally, there was a commitment for PVNGS to complete and submit the AMP by five years after the license amendment incorporating power uprate. The commitment was subsequently revised so that the NEI-03-08 commitment supersedes the power uprate commitment

[6, 9]. Subsequently, PVNGS updated this commitment to submit a RVI AMP based on the NRC-approved PWR reactor vessel internals aging management program of the MRP-227 no later than October 1, 2012 [7]. Therefore, this revised AMP was developed to meet MRP-227-A and replace the prior version.Needed 1. Each commercial U.S. PWR unit shall implement Tables 4-1 through 4-9 and Tables 5-1 through 5-3 of M.RP-22 7-A for the applicable design within 24 months following issuance of MIRP-227-A.

Palo Verde Applicability:

MRP-227 augmented inspections will be incorporated in the Palo Verde ISI for the license renewal period. The applicable C-E tables contained in MRP-227-A components are Table 4.2 (Primary), Table 4.5 (Expansion), and-Table 4.8 (Existing) and are attached herein as Table 5-1, Table 5-2, and Table 5-3, respectively.

This AMP has been developed in accordance with MRP-227-A

[3].2. Examinations specified in the MRP-227-A guidelines shall be conducted in accordance with Inspection Standard MRP-228.Report No. 1200347.401 .R1 1-12 kkp 1W Palo Verde Applicability:

Inspection standards will be in accordance with the requirements of MRP-228 [14]. These inspection standards will be used for augmented inspection at Palo Verde as applicable where required by MRP-227-A.

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

Palo Verde Applicability:

PVNGS will comply with this requirement

[8].4. Each commercial U.S. PWR unit shall provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals within the scope of MRP-227-A are examined.Palo Verde Applicability:

APS will participate in future industry efforts and will adhere to industry directives for reporting, response, and follow-up.

5. If an engineering evaluation is used to disposition an examination result that does not meet the examination acceptance criteria in Section 5 of MRP-22 7-A, this engineering evaluation shall be conducted in accordance with a NRC-approved evaluation methodology.

Palo Verde Applicability:

PVNGS will comply with this requirement by using NRC-approved evaluation methodology (e.g. WCAP-17096

[17]).1.8.4 MRP-227-A AMP Development Guidance In addition to the implementation of the requirements of MRP-227-A in accordance with NEI 03-08, this RVI AMP addresses the 10 program elements as defined in the GALL Report Chapter XI.Ml6A (provided in Section 2.3 of this Report)Report No. 1200347.40 l.R1 1-13 C ON&'W'/ "d,1 1.8.4.1 MRP-22 7-A Applicability to Palo Verde The applicability of MRP-227-A to PVNGS requires compliance with the following MRP-227 assumptions:

• Operation of 30 years or less with high-leakage core loading patterns (fresh fuel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation

[25 -27].Applicability:

PVNGS Units 1, 2, and 3 historic core management practices meet the requirements of MRP-227-A.

• Base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule [29].Applicability:

PVNGS Units 1, 2, and 3 operate as base load units.* No design changes beyond those identified in general industry guidance or recommended by the original vendors.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.

PVNGS Units 1, 2, and 3 have not made any modifications of the RVI components beyond those identified in general industry guidance or recommended by the vendor (C-E) since the May 2007 effective date of this statement, and therefore meets this requirement of MRP-227-A.

Hence, it is evident that operations at PVNGS conform to the assumptions in Section 2.4 of MRP 227-A.Report No. 1200347.401 .R 1-14 Associates k W, Inc' 1.8.5 Ongoing Industry Programs APS actively participates in the EPRI MRP, PWR Owners Group, and other 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.

1.9 Summary The GALL Report identifies which reactor internals passive components are most susceptible to the aging mechanisms of concern. Additionally, this report identifies the appropriate inspections or mitigation programs needed to manage the aging mechanisms of the reactor vessel internals to assure these components will maintain their functionality through the period of extended operation.

The GALL Report was used at Palo Verde for the initial basis of their LRA. The NRC has reviewed Palo Verde's LRA and their approval is documented in NUREG-1961

[31].The Palo Verde RVI AMP has been created to address the reactor vessel internals aging concerns consistent with the information identified in the GALL Report and the guidance in MRP-227-A.

PVNGS will manage their RVI inspections through their augmented ISI program and will complete any repairs and/or replacements in accordance with ASME Code requirements and any NRC approved methodologies.

The PVNGS AMP will be updated accordingly as operating experiences and new inspection requirements and technologies evolve associated with managing reactor vessel aging concerns.Report No. 1200347.401.Rl 1-15 .a eLWIe hr Assowkits, 1W 2.0 AGING MANAGEMENT APPROACH The reactor vessel internals is a part of the reactor coolant system (RCS). The reactor vessel internals are passive structural components designed to support the functions of the RCS core cooling, control element assembly (CEA) insertion, and the integrity of the fuel and pressure vessel boundary.

The core support structures provide support and restraint of the core. Static (i.e. deadweight and mechanical) loads from the assembled components, fuel assemblies, dynamic loads (i.e. hydraulic flow, flow-induced vibration, thermal expansion, and seismic) and LOCA loads are all carried by the core support structures.

In addition to core support, the various internals assemblies provide a flow boundary to direct coolant flow from the cold leg inlet nozzles, down the annulus to the lower plenum, upward past the core, into the upper plenum region above the core, and out the outlet nozzles to the hot leg piping.2.1 Mechanisms of Age-Related Degradation in PWR Internals The potential aging mechanisms that could affect the long term operation of PWR reactor vessel internals are discussed in this section. Initial screening performed as part of MRP-227-A was on the basis of susceptibility of PWR RVI to eight different age-related degradation mechanisms

-stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (IASCC), wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling, and the combination of thermal and irradiation-enhanced stress relaxation.

2.1.1 Stress Corrosion Cracking 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.

If components are susceptible to SCC and require aging management, EVT-I exams will be performed per MRP-227-A.

Report No. 1200347.401.R1 2-1 .aftft Wlt fly Assate,1 2.1.2 Irradiation-Assisted Stress Corrosion Cracking Irradiation-assisted stress corrosion cracking (IASCC) is a unique form if SCC that occurs only in highly-irradiated components.

The aging effect is cracking.2.1.3 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.2.1.4 Fatigue Fatigue is defined as the structural deterioration that can occur as a result of repeated stress/strain cycles caused by fluctuating loads and temperatures.

After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading eventually to macroscopic crack initiation at the most highly affected locations.

Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack.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. From a design perspective, the aging effects of low-cycle fatigue and high-cycle fatigue are additive.Fatigue crack initiation and growth resistance is 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.Report No. 1200347.401.R1 2-2 J OSIuWFgI/WU 2.1.5 Thermal Aging Embrittlement Thermal aging embrittlement is the exposure of delta ferrite within cast austenitic stainless steel (CASS) and precipitation-hardenable (PH) stainless steel to high inservice temperatures, which can result in an increase in tensile strength, a decrease in ductility, and a loss of fracture toughness.

Some degree of thermal aging embrittlement can also occur at normal operating temperatures for CASS 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.

There are no CASS or PH components in the PVNGS lower core support structures

[12].2.1.6 Irradiation Embrittlement Irradiation embrittlement is also referred to as neutron embrittlement.

When exposed to high energy neutrons, the mechanical properties of stainless steel and nickel-base alloys can be changed. Such changes in mechanical properties include increasing yield strength, increasing ultimate strength, decreasing ductility, and a loss of fracture toughness.

The irradiation embrittlement aging mechanism is a function of both temperature and neutron fluence. While the initial aging effect is loss of ductility and toughness, unstable crack extension is the eventual aging effect if a crack is present and the local applied stress intensity factor exceeds the reduced fracture toughness.

2.1.7 Void Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation of microscopic cavities in the material.

These cavities result from the nucleation and growth of clusters of irradiation produced vacancies.

Helium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material.

Void swelling may produce dimensional changes that exceed the tolerances on a component.

Strain gradients produced by differential swelling in the system may produce significant stresses.

Severe Report No. 1200347.401.Rl 2-3 jviLaLn u hr It Assowatus, I swelling (>5% 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 in in-core instrumentation tubes fabricated from zirconium alloys. (Note: The PVNGS reactor vessel internals do not include any zirconium alloys.) While the initial aging effect is dimensional change and distortion, severe void swelling may eventually result in cracking under stress. Aging management of void swelling is by visual inspection targeted at locations where swelling is most likely to occur.2.1.8 Thermal and Irradiation-Enhanced Stress Relaxation or 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, such as seen in PWR internals.

Stress relaxation occurs under conditions of constant strain where part of the elastic strain is replaced with plastic strain. Available data show that thermal stress relaxation appears to reach saturation in a short time (<100 hours) at PWR internals temperatures.

Creep (or more precisely, secondary creep) is a slow, time and temperature dependent, plastic deformation of materials that can occur when subjected to stress levels below the yield strength (elastic limit). Creep occurs at elevated temperatures where continuous deformation takes place under constant strain. 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 a thermal process that depends on the neutron fluence and stress; and, it can also be affected by void swelling, should it occur. The aging effect is a loss of mechanical closure integrity (or, preload) that can lead to unanticipated loading which, in turn, may eventually cause subsequent degradation by fatigue or wear and result in cracking.2.2 Aging Management Strategy The guidelines provided in MRP-227-A

[3] define a supplemental inspection program for managing aging effects and provide generic guidance to help develop this aging management Report No. 1200347.401 .Rl 2-4 .IAwb Wag AsocItUs, InW program for PVNGS. The EPRI MRP Reactor Internals Focus Group developed these guidelines to support the continued functionality of RVI. The focus group also developed MRP-228, which addresses the inspection standard for the RVI. The aging management strategy used to develop the guidelines combined with the results of functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results. The aging management strategy that was developed was used in the development of an appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections.

Additionally, it was also used to identify the components and locations for supplemental examinations by categorization.

MRP-227-A used a screening and ranking process to aid the identification of required inspections for specific RVI components.

The screening and categorization process also credited existing component inspections, when they were deemed adequate.

Through the screening and categorization process, the RVI for all currently licensed and operating PWR designs in the U.S.were evaluated, and appropriate inspection, evaluation and implementation requirements for RVI were defined.The RVI components are categorized in MRP-227-A as "Primary" components, "Expansion" components, "Existing Programs" components, or "No Additional Measures" components, described as follows:* Primary: Those PWR internals that are highly susceptible to the effects of at least one of the eight aging mechanisms were placed in the Primary group. The aging management requirements that are needed to ensure functionality of Primary components are described in these I&E guidelines.

The Primary group also includes components which have shown a degree of tolerance to a specific aging degradation effect, but for which no highly susceptible component exists 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 Report No. 1200347.401.R1 2-5 5Vj 1 iD f r ft* AssOGad , Inc implementation of aging management requirements for Expansion components will depend on the findings from the examination of the Primary components at individual plants.* Existing Programs:

Those PWR internals that are susceptible to the effects of at least one of the eight aging mechanisms and for which generic and plant-specific existing AMP elements are capable of managing those effects, were placed in the Existing Programs group.* No Additional Measures:

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 FMECA and the functionality assessment.

No further action is required by these guidelines for managing the aging of the No Additional Measures components.

A description of the categorization process used to develop the Guidelines is given below. The approach in these guidelines has been used to develop the PVNGS AMP.In accordance with the MRP-227-A I&E Guidelines

[3], this inspection strategy consists of the following:

• Selection of items for aging management

• Selection of the type of examination appropriate for each degradation mechanism* Specification of the required level of examination qualification

• Schedule of first inspection and frequency of any subsequent inspections

• Requirements for sampling and coverage* Requirements for expansion of scope if unanticipated indications are found" Inspection acceptance criteria" Methods for evaluating examination results not meeting the acceptance criteria" Updating the program based on industry-wide results* Contingency measures to repair, replace, or mitigate Report No. 1200347.401.Rl 2-6 V- Slynkif kkp* Asscu" 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 on the framework and strategy and on the accumulated industry research data, the following elements of an AMP were further developed

[ 13, 24]: " 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.

• PWR Internals components were categorized, based on the screening criteria as follows:-Components for which the effects of 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 the functionality assessment with several contributing factors to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections.

Factors considered included component accessibility, operating experience (OE), existing evaluations, and prior examination results.2.3 Palo Verde Reactor Vessel Internals Aging Management Program Attributes The attributes of the PVNGS RVI AMP and their compliance with the ten elements of NUREG-1801 (GALL Report), Revision 2, Chapter XI.M 16A, "PWR Vessel Internals" [ 10] are essential for successful management of component aging are described in this section.Report No. 1200347.40 L.R1 2-7 IMIWgrktt AssaCates, In This AMP is consistent with the GALL process and includes consideration of the augmented inspections identified in MRP-227-A

[3]. Specific details of the PVNGS RVI AMP are summarized in the following subsections.

2.3.1 NUREG-1801/AMP Program Element 1: Scope of Program"The scope of the program includes all R VI components at the Palo Verde Nuclear Generating Station, which is built to a C-E NSSS design. The scope of the program applies to the methodology and guidance in the most recently NRC-endorsed version of MRP-22 7, which provides augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S.PWR nuclear power plants designed by B& W, C-E, and Westinghouse.

The scope of components considered for inspection under MiRP-22 7 guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code,Section XI), those R VI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a) (1), and other R VI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a) (1) (1), (ii), or (iii). The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review (AMR), as defined by the acceptance 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 I appurtenances to the reactor vessel and are adequately managed in accordance with an applicant's AMP that corresponds to GALL AMP XI.M1, "ASME Code,Section XI Inservice Inspections, Subsections IWB, IWC, and IWD. ""The scope of the program includes the response bases to applicable license renewal applicant action items (LRAAIs) on the MRP-22 7 methodology, and any additional programs, actions, or activities that are discussed in these LRAAI responses and credited Report No. 1200347.401.Rl 2-8 V I Iu Itl.y ASSOates.

1cW.

for aging management of the applicant's R VI components.

The LRAAIs are identified in the staff's safety evaluation on MRP-227 and include applicable action items on meeting those assumptions that formed the basis of the MRP 's augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 of MRP-22 7), and NSSS vendor-specific orplant-specific LRAAIs as well. The responses to the LRAAIs on MIRP-227 are provided in Appendix C of the LRA.""The guidance in MRP-22 7 specifies applicability limitations to base-loaded plants and the fuel loading management assumptions upon which the functionality analyses were based. These limitations and assumptions require a determination of applicability by the applicant for each reactor and are covered in Section 2.4 of MRP-22 7." 2.3.1.1 PVNGS Program Scope A description of the PVNGS RVI design is provided in Section 3.0 of this program plan.Additional details regarding the RVI are provided in the PVNGS UFSAR [30]. The PVNGS RVI subcomponents that require aging management review are indicated in the PVNGS LRA[5]. Table 3.1.2-1 from the PVNGS LRA is also located in Appendix B of this program plan.Table B-I includes a summary of the results of the AMR. This table identifies the aging effects that require management.

A column in the table lists the programs and activities at PVNGS that are credited to address the aging effects for each management strategy presented in Appendix B, Table B-1 as documented in Table 3.1.2-1 of the PVNGS LRA and Section 3.1.2.3.2 of the NRC's license renewal SER [31].MRP-227-A provides the inspection and evaluation guidelines to develop plant specific programs to manage the effects of aging in PWR internals.

MRP-227-A is also used as a guidance to develop an aging management program to satisfy license renewal commitments for the PWR fleet. A summary of the inspections required to be performed, the appropriate inspection techniques used to detect aging (i.e. cracking, loss of material, loss of preload, etc.), frequency of inspections, and the acceptance criteria for the inspections are provided in MRP-Report No. 1200347.401.Rl 2-9 ItC1Lnh Wft *I. Assawats, In 227-A (summarized in Tables 2 through 5 of this AMP). Guidance provided in MRP-227-A in conjunction with the guidance provided in the NRC SE [4] for MRP-227 and the GALL Report were reviewed to establish the basis for the PVNGS RVI AMP. In addition, plant specific existing programs such as the Section XI ISI program for PVNGS will complement the augmented inspection requirements provided in MRP-227-A in successfully managing the effects of aging for the PVNGS Units during the period of extended operation.

2.3.1.2 Conclusion This element is consistent with .he corresponding aging management attribute in Revision 2 of NUREG-1801

[10], Chapter XI.M16A and the list of commitments in the PVNGS License Renewal SER, as updated in Table 19.5-1 in the PVNGS UFSAR [30].2.3.2 NUREG-1801/AMP Program Element 2: Preventive Actions"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 its forms [SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program. The program description, evaluation and technical bases of water chemistry are presented in GALL AMP X1.M2, "Water Chemistry." 2.3.2.1 PVNGS Preventive Action The PVNGS RVI AMP includes the following existing programs that comply with the requirement of this element. This program is a support program of 81 DP-9RC03 [1]. A description and applicability to the PVNGS RVI AMP is provided in the following subsection.

Report No. 1200347.401.R1 2-10 -Associatus, Inc..

2.3.2.2 Primary Water Chemistry Program The primary goal of this program is to mitigate loss of material due to general, pitting, and crevice corrosion, cracking due to Stress Corrosion Cracking (SCC) by controlling the internal environment of systems and components.

This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below the system-specification limits. The PVNGS system chemistry program [23] is based on current, approved revisions of EPRI PWR Primary Water Chemistry Guidelines.

This program is consistent with the corresponding program described in Revision 2 for GALL Report [10]. The program description, evaluation, and 'technical basis of water chemistry are presented in GALL AMP XI.M2, "Water Chemistry." The limits of known detrimental contaminants imposed by the water chemistry program are consistent with the EPRI PWR Primary Water Chemistry Guidelines

[24].2.3.2.3 Conclusion This element is consistent with the corresponding aging management program attribute in Revision 2 of NUREG- 1801 [10], Chapter XI.M 16A.2.3.3 NUREG-1801/AMP Program Element 3: Parameters Monitored/Inspected: "The program monitors and manages the following age-related degradation effects and mechanisms that are applicable in general to RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss offracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling and irradiation growth, distortion, or deflection; and (e) loss ofpreload caused by thermal and irradiation-enhanced stress relaxation or creep. For the management of cracking, the program monitors for evidence of surface breaking linear discontinuities if a visual inspection technique is used as the non-destructive examination (NDE) method, or for Report No. 1200347.401.Rl 2-11 .ski huI &Iityp ASSOwa'tes, I' relevant flaw presentation signals if a volumetric UT method is used as the NDE method.For the management of loss of material, the program monitors for 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 monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections.

The program does not directly monitor for loss offracture toughness that is induced by thermal aging or neutron irradiation embrittlement, or by void swelling and irradiation growth; instead, the impact of loss offracture toughness on component integrity is directly managed by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable reduced fracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation under the MiRP-22 7 guidance or ASME Code,Section XI requirements.

The program uses physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection.""Specifically, the program implements the parameters monitored/inspected criteria for CE designed Primary Components in Table 4-2 of MRP-227. Additionally, the program implements the parameters monitored/inspected criteria for CE designed Expansion Components in Table 4-5 of MRP-227. The parameters monitored/inspected for Existing Program Components follow the bases for referenced Existing Programs, such as the requirements of the ASME Code Class R VI components in ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-N-3, as implemented through the applicant's ASME Code,Section XI program, or the recommended program for inspecting Westinghouse designed flux thimble tubes in GALL AMP XI.M37, "Flux Thimble Tube Inspection. "No inspections, except for those specified in ASME Code, Section Xl, are required for components that are identified as requiring "No Additional Measures, " in accordance with the analyses reported in MRP-22 7. " Report No. 1200347.401.R1 2-12 V OW&W M*' 1 2.3.3.1 PVNGS Parameters Monitored/Inspected PVNGS monitors, inspects, and/or tests for the effects of the eight aging degradation mechanisms on the intended function of the reactor vessel internals components through inspection and condition monitoring activities in accordance with the augmented requirements defined under industry directives as contained in MRP-227-A

[3] and ASME Section XI [21].This program is a support program of 81 DP-9RC03 [1].2.3.3.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG- 1801 [10], Chapter XI.M 16A.2.3.4 NUREG-1801/AMP Program Element 4: Detection of Aging Effects: "The detection of aging effects is covered in two places: (a) the guidance in Section 4 of MRP-22 7 provides and introductory discussion and justification of the examination methods selected for detecting the aging effects of interest; and (b) standards for examination methods, procedures, and personnel are provided in a companion document, MRP-228. In all cases, well established methods were selected.

These methods include volumetric UT examination methods for detecting flaws in bolting, physical measurements for detecting changes in dimensions, and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities.

Surface examinations may also be used as an alternative to visual examinations for detecting and sizing of surface breaking discontinuities.""Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-i or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting).

The VT-3 visual methods may be applied for the detection of cracking only Report No. 1200347.401 .R1 2-13 OWIIW UFuV/ ASsowatu, Inc when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions.

In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss ofpreload caused by thermal and irradiation-enhanced stress relaxation and creep."In addition, the program adopts the recommended guidance in MRP-227for defining the Expansion criteria that need to be applied to inspections of Primary Components and Existing Requirement Components and for expanding the examinations to include additional Expansion Components.

As a result, inspections performed on the R VI components are performed consistent with the inspection frequency and sampling bases for Primary Components, Existing Requirement Components, and Expansion Components in MRP-227, 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 Position RSLB-1.""Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for CE designed Primary Components in Table 4-2 ofMRP-227 and for CE designed expansion components in Table 4-5 of MRP-227." Physical measurements:

Per Revision 2 of NUREG-1801, this is not applicable for C-E designed plants.2.3.4.1 PVNGS Detection ofAging Effects Detection of indications required by the ASME Section XI ISI Program is well-established and field-proven through application of the Section XI ISI Program. Those augmented inspections Report No. 1200347.40 L.Rl 2-14 V oIIIDUTlII /W that are taken from the MRP-227-A recommendations will be applied through use of the MRP-228 [14] Inspection Standard.Inspection can be used to detect physical effects of degradation including cracking, fracture, wear and distortion.

The choice of an inspection technique depends on the nature and extent of the expected damage. The recommendations supporting aging management of the RVI, as contained in this program, are built around three basic inspection techniques:

visual, ultrasonic, and physical measurement.

The visual techniques include VT-3, VT-1, and EVT-1. Inspection standards developed by the industry for application of these techniques in augmented RVI inspections are documented in MRP-228 [14]. Continued functionality can be confirmed by physical measurements to detect degradation mechanisms such as wear, or loss of functionality as a result of loss of preload or material deformation.

If components have been shown to be flaw tolerant, the scope of the inspections for detection of aging effects may be modified.2.3.4.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG- 1801 [ 10], Chapter XI.M 16A.2.3.5 NUREG-1801/AMP Program Element 5: Monitoring and Trending: "The methods for monitoring, recording, evaluating, and trending the data that result from the program's inspections are given in Section 6 of MRP-22 7 and its subsections.

The evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as wellfor performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications.

The examinations and re-examinations required by the MRP-22 7 guidance, together with the requirements specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects of the age-related degradation mechanisms within the scope of the Report No. 1200347.40 L.R 2-15 luw ftwg d Assoiates, 1W program. The extent of the examinations, beginning with the sample of susceptible PWR internals component locations identified as Primary Component locations, with the potential for inclusion of Expansion Component locations if the effects are greater than anticipated, plus the continuation of the Existing Programs activities, such as the ASME Code,Section XI, Examination Category B-N-3 examinations for core support structures, provides a high degree of confidence in the total program." 2.3.5.1 PVNGS Monitoring and Trending Reporting operating experience with PWR internals has been generally proactive.

The majority of materials aging degradation models and analyses used to develop the MRP-227-A guidelines are based on test data from RVI 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 operating experience through the auspices of the MRP and PWROG. PVNGS has in the past and will continue to maintain cognizance of industry activities and will continue to share operating experience information related to PWR internals inspection and aging management.

Inspections credited as part of the existing programs, where practical, are scheduled to be conducted in conjunction with typical 10-year ISI examinations.

Inspections performed at PVNGS as part of the ISI program are provided in Appendix A. Tables 2 and 3 identify the inspection requirements for Primary and Expansion category components credited for aging management of RVI. As discussed in MRP-227-A

[3], the sampling inspections of the "Primary" components, with the potential for expanding the sampling program if unexpected effects are found, provides reasonable assurance for demonstrating the ability of the reactor vessel internal components to perform the intended functions.

Reporting requirements are included as part of MRP-227-A guidelines.

Consistent reporting of inspection results across all PWR designs will enable the industry to monitor RVI degradation on an ongoing basis as plants enter the period of extended operation.

Reporting of examination Report No. 1200347.401 .Rl 2-16 -OW' grkklg Assaaktas, '

results will allow the industry to monitor and trend results and take appropriate preemptive action through update of the MRP guidelines.

2.3.5.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801

[10], Chapter XI.M16A.2.3.6 NUREG-1801/AMP Program Element 6: Acceptance Criteria"Section 5 of MRP-227 provides specific examination acceptance criteria for the Primary and Expansion Component examinations.

For components addressed by examinations referenced to ASME Code,Section XI, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.The guidance provided 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 sized for length by VT-i/EVT-1 examinations;

• For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment for bolted or pinned assemblies with unacceptable volumetric (UT)examination indications that exceed specified limits; and" Physical measurements.

Per Revision 2 of NUREG-1801, this is not applicable for C-E designed plants." Report No. 1200347.401-R1 2-17 bitugrily

ASO0t, 2.3.6.1 PVNGS Acceptance Criteria Recordable indications that are the result of inspections required by PVNGS existing ISI program [ 18 -20] are evaluated in accordance with the requirements of the ASME Code and documented in the PVNGS Corrective Action Program [8].Inspection acceptance and expansion criteria are provided in Table 5-4 of this document.

These criteria will be reviewed whenever new revisions of the NRC approved versions of MRP-227 and WCAP- 17096 are published and as the industry continues to develop and refine the information.

Changes applicable to the PVNGS RVI will be included as part of updates to this AMP.Recordable indications found during the MRP-227-A augmented inspections will be entered into the PVNGS correction action program. These indications will be addressed by additional inspections, repair, replacement, mitigation, or analytical evaluations to further disposition these indications.

Industry groups are working to develop a consistent set of tools compliant with approved methodologies to support this element. Additional analysis to establish evaluation acceptance criteria for "Expansion" category components has been developed by the WCAP-17096-NP.

The status of these ongoing processes is monitored via PVNGS participation in various industry programs related to aging management of PWR internals.

2.3.6.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801

[10], Chapter XI.M16A.2.3.7 NUREG-1801/AMP Program Element 7: Corrective Actions: "Corrective actions following the detection of unacceptable conditions are fundamentally provided for in each plant's corrective action program. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through Report No. 1200347.401.R1 2-18 Assowatkttus.

I ASnc.',

the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection.

The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilledfor all licensing basis loads and events. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code, Section XJ or in 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 criteria of Section 5 of the report. These include engineering evaluation methods, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures.

The latter are subject to the requirements of the ASME Code,Section XI. The implementation of the guidance in MRP-227, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable." 2.3.7.1 PVNGS Corrective Actions The PVNGS Correction Action Procedure

[8] addresses this element of the GALL attributes.

Indications that require repair and replacement will be addressed through the PVNGS Correction Action Program. Repair and replacement activities will be performed in accordance with methodologies provided in Section 6 of MRP-227-A

[3] and ASME Code Section XI [21].. The corrective actions for existing Section XI (B-N-3) examinations will include the identification of a repair plan and verification of acceptability of replacements.

Any indications found during the Section XI examinations for the RVI will be documented in the corrective action program [8].These indications will be addressed by additional inspections, repair, replacement, or analytical evaluations to further disposition the indications.

Any associated relevant indications will be developed using evaluation methods or repair and replacement procedures in accordance with or equivalent to the requirements of ASME Code Section XI. Actions to evaluate and monitor flaws or indications will be a part of the corrective action process. This evaluation guidance is included in MRP-227-A and WCAP-17096-NP

[ 17]. For example, the guidance provided in Report No. 1200347.401.Ri 2-19 c airn&*Il l Assawatus, 1W" WCAP- 17096-NP may be used to evaluate component degradation that exceeds acceptance criteria in Section 5 of MRP-227-A when it is observed during required inspections.

Other methods may also be used if approved by NRC.2.3.7.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801

[10], Chapter XI.M16A.2.3.8 NUREG-1801/AMP Program Element 8: Confirmation Process"Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, or equivalent, as applicable.

It is expected that the implementation of the guidance in MRP-227 will provide an acceptable level of quality for inspection, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with 10 CFR Part 50, Appendix B or equivalent, as applicable, confirmation process, and administrative controls." 2.3.8.1 PVNGS Confirmation Process The PVNGS RVI Aging Management Program meets the "Mandatory" and "Needed" requirements under NEI 03-08. This program conforms to the PVNGS NEI 03-08 Materials Initiative Process [1], and it ensures that deviations, self-assessments and benchmarks are conducted as necessary to support the NEI Materials Initiative.

The PVNGS Section XI Inspection Program and Corrective Action Process meet the requirements for QA programs.

In particular, all QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B [34].Report No. 1200347.401 .R 2-20 1111W W k 2.3.8.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG- 1801 [10], Chapter XI.M 16A.2.3.9 NUREG-1801/AMP Program Element 9: Administrative Controls: "The administrative controls for such programs, including their implementing procedures and review and approval processes, are under the existing site 10 CFR 50 Appendix B Quality Assurance Programs, or their equivalent, as applicable.

Such a program is thus expected to be established with a sufficient level of documentation and administrative controls to ensure effective long-term implementation." 2.3.9.1 PVNGS Administrative Controls PVNGS QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B which are acceptable in addressing administrative controls.2.3.9.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG- 1801 [ 10], Chapter XI.M 16A.2.3.10 NUREG-1801/AMP Program Element 10: Operating Experience"Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-22 7-A. The applicant is expected to review subsequent operating experience for impact on its program or to participate in industry initiatives that perform this function.Report No. 1200347.401.R1 2-21 ISiiOq4 InW, The application of MRP-22 7 guidance will establish a considerable amount of operating experience over the next few years. Section 7 of MRP-22 7 describes the reporting requirements for these applications, and the plan for evaluating the accumulated additional operating experience." 2.3.10.1 PVNGS Operating Experience Extensive industry and PVNGS operating experience (OE) has been reviewed during the developmert of the PVNGS RVI AMP.Industry and PVNGS specific information relevant to aging has been compiled into a PVNGS OE database [35]. Industry operating experience sources in the database include the following:

applicable NRC Generic Publications (including Information Notices, Circulars, Bulletins and Generic Letters), NRC Generic Aging Lessons Learned (GALL) Report, etc. Plant specific operating experience sources in the database include applicable maintenance work history, licensee event reports (LERs), and, Corrective Action Process documents (CAPs, CRs, DR, ERs), etc.Early plant operating experience related to hot functional testing and RVI is documented in plant historical records. Inspections performed as part of the 10-year ISI program have been conducted as designated by commitments and would be expected to discover general internals structure degradation.

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

Industry operating experience is routinely reviewed by PVNGS engineers using the Institute of Nuclear Power Operations (INPO) OE database, the Nuclear Network, and other information sources, as directed under the applicable procedure

[35], for the determination of additional actions and lessons learned. RVI operating experiences will be incorporated into the plant system health reports and included in the applicable plant programs.Report No. 1200347.401 .R 2-22 W klq 1W A review of industry and plant-specific experience with RVI reveals that the U.S. nuclear industry, including PVNGS, has responded proactively to issues relative to RVI degradation.

Examples of PVNGS proactivity are briefly described in the following paragraphs:

• Participation in PWROG OE Activities:

PVNGS participated in a PWROG project, which documented RVI aging degradation OE from the domestic and international PWR plants. The PWROG members were asked about prior inspections and results for the MRP-227-A RVI components.

PVNGS submitted survey responses for all three units, detailing previous inspections, specific findings, and inspection timing. The results of this survey are documented in WCAP-17435-NP

[36]. The OE in this report provides a benchmark from which to evaluate further RVI aging management events.Palo Verde reported the following information for the MRP-227-A RVI components in the survey: o Number of inspections per component o Details of each inspection per component o Record of indications for a given inspection o Subsequent corrective actions (response) or indications found* Cognizance of Industry OE: PVNGS is committed to monitoring specific industry OE that could potentially affect the RVI during the period of extended operating at PVNGS and at other domestic PWR facilities.

For example, PVNGS has monitored the emerging OE from the fuel leakage at a domestic C-E designed power plant in fuel assemblies adjacent to the core shroud. Higher radiation levels may increase the susceptibility of stainless steel in the RVI to various material degradation mechanisms.

These first burned peripheral fuel assemblies were detected at the specific plant and dispositioned.

PVNGS will incorporate related OE to ensure safe and reliable operation.

Report No. 1200347.401 .Rl 2-23 &ltr d MsowI as," In A key element of the MRP-227-A guideline is the reporting of age-related degradation of RVI components.

PVNGS, through its participation in EPRI MRP activities, will continue to benefit from the reporting of inspection information and will share its own OE with the industry through those groups or INPO, as appropriate.

2.3.10.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG- 1801 [10], Chapter XI.M 16A.Report No. 1200347.401.Rl 2-24 V OWNWO kLb Assudates, Wn 3.0 PALO VERDE REACTOR VESSEL INTERNALS DESIGN AND OPERATING EXPERIENCE The Palo Verde RPV and internals were designed by Combustion Engineering (System 80 model). The Palo Verde Units 1, 2, and 3 vessel internals were designed to ASME Section III, Subsection NG, (1974 Edition) [39]. The PVNGS design consists of a core shroud assembly with full height core shroud plates. The components of the reactor vessel internals are divided into smaller sub-assemblies consisting of the upper guide structure, CEA shroud assemblies, core support barrel assembly, core shroud assembly, lower internals assembly, and in-core instrumentation system. The general arrangement of the PVNGS reactor vessel internals and RVI internals assembly are shown in Figure 3-1 and Figure 3-2 respectively.

Schematic representations of other reactor vessel internals components are provided in Figure 3-3 through Figure 3-8.3.1 Upper Guide Structure Assembly The upper guide structure (UGS) assembly consists of a support barrel assembly and a CEA shroud assembly.

The upper guide structure is supported by and external flange at the top of the support barrel assembly.

This flange is supported above the core support barrel flange, on the holddown ring. The PVNGS hold-down ring is located between the upper side of the upper internals assembly flange and the top of the core support barrel flange [12]. The UGS flange has four equally spaced, machined alignment keyways which engage the alignment keys. The CEA shroud assembly is attached to the support barrel assembly by 8 tie rod assemblies.

Alignment between the lower end of the CEA shrouds and the upper end of the fuel assemblies is maintained by tubes which extend down from the support barrel assembly fuel alignment plate.The fuel alignment plate has four equally spaced slots machined in its periphery.

These slots engage the guide lugs on the core shroud and provide lateral stability for the UGS assembly.

A schematic representation of the PVNGS UGS assembly is provided in Figure 3-3. The upper guide structure (UGS) aligns and supports the upper end of the fuel assemblies to prevent movement during pressure transients or a severe accident condition.

The UGS also protects the Report No. 1200347.401.R1 3-1 ASSOOIttUS Ass Iae, CEAs from the effects of coolant cross flow into the outlet nozzle region. The UGS assembly can be is removed during refueling as a single component in order to provide access to the fuel assemblies

[38].3.2 Core Support Barrel Assembly The core support barrel assembly consists of the core support barrel (CSB), the lower support structure and instrument nozzle assembly, the core shroud assembly, the core support barrel snubber lugs, and alignment keys. The core support barrel assembly is supported at its upper flange from a ledge in the reactor vessel flange. The lower end is res,,rained in its lateral movement by six sets of core support barrel snubber lugs. Within the core support barrel is the core shroud, which is attached to the lower support structure.

The core shroud forms the perimeter of the core and acts as the transition structure between the rectilinear polygon core cross section and the cylindrical core support barrel. Four guide lugs are located at the top of the core shroud to maintain alignment between the upper guide structure and the lower support structure to prevent excessive motion between the core shroud and the fuel alignment plate. The lower support structure and instrument nozzle assembly are positioned within the barrel at the lower end and is composed of the grid beam structure, columns, cylinder, bottom plate and instrument nozzles [38]. The support beams and insert pins provide support and orientation for the fuel assemblies.

A schematic representation of the PVNGS core support barrel assembly is provided in Figure 3-4.3.3 Lower Support Structure and Internals Nozzle Assembly The lower support structure and instrument nozzle assembly consists of a welded grid structure fabricated from 13/4 inch thick main and cross beams with fuel assembly locating pins inserted into holes in the top of the grid and 2/2 thick perforated plates connecting the main beams at the bottom. The grid and plate structure is welded to a 149 inch ID cylinder, 161/4 inch long and 3 21/32 inches thick. The instrument nozzle support plate is located below the grid structure and is connected to the lower support structure by support columns and instrument nozzles. The lower support structure and instrument nozzle assembly is supported by the core support barrel lower Report No. 1200347.40,1.Rl 3-2 .""kmw ,L%'t Ar I flange and is welded into the CSB after being oriented during internals installation

[38]. For plants with bottom-entry ICI such as the PVNGS units, the ICI assemblies consist of an ICI nozzle, which is essentially the same as the ICI guide tube for the top-entry plants. The ICI nozzle contains the in-core instrumentation.

There are no ICI thimbles in the Palo Verde Units[12]. The PVNGS ICI support assembly is welded to the Lower Internals Assembly (Lower Support Structure).

The guide tubes are connected to and supported by the lower internals assembly, from which the in-core instrumentation enters the core. The ICI nozzles, as they are called in the Palo Verde Units, contain the in-core instrumentation and penetrate up through the precision holes in the ICI Nozzle Support Plate and into the Lower Internals Assembly, as shown in Figure 3-6. The ICI nozzles terminate at the grid located at the bottom of the fuel assemblies with a minimal gap to prevent cross vibration.

The lower support structure and instrument nozzle assembly aligns the fuel assemblies, provides a guide path for the in-core instrumentation and directs coolant flow into the fuel assemblies (Figure 3-7 and Figure 3-8). A support grid, bottom plates and a cylindrical skirt provide support for the instrumentation assembly and transmits the core load to the bottom flange of the core support barrel. The in-core instrumentation then enters a guide tube located within the center of the fuel assemblies.

The ICI support assembly in the PVNGS units also has the following functions:

• To provide guidance and support to the In-core Instrumentation within the Lower Internals Assembly.* To protect the In-core Detector Assemblies from flow effects within the Reactor Vessel.There are no core support plates and core support columns in the PVNGS Units [37]. PVNGS Units 1, 2, and 3 have a welded shroud and bottom-mounted ICI have no core support plate, in which case the fuel alignment pins are attached directly to the core support deep beams.3.4 Core Shroud Assembly The core shroud assembly is located within the core support barrel and directly below the upper internals assembly (Figure 3-5). The core shroud assembly is a welded assembly fabricated from Report No. 1200347.40 L.R 3-3 AISOtWWtUt

• A ssJdates, 1W Type 304 stainless steel. The core shroud consists of thirty six vertical panels arranged to form a polygon enclosing the core perimeter, a top and bottom plate and five intermediate horizontal rings with vertical reinforcing ribs and channels connecting the horizontal rings. Four guide lugs located 900 apart on the top plate engage and provide lateral restraint for the UGS fuel alignment plate. The core shroud is oriented and welded to the lower support structure as shown in Figure 3-5. The core shroud assembly functions are to provide a boundary between reactor coolant flow on the outside of the core support barrel and the reactor coolant flow through the fuel assemblies, to limit the amount of coolant bypass flow, and to reduce the lateral motion of the fuel assemblies.

Holes are provided in the core shroud structure to allow coolant to flow upward between the core shroud and the core support barrel, thereby minimizing thermal stresses in the shroud and eliminating stagnant flow conditions

[38].3.5 Control Element Assembly Shroud Assemblies The control element shroud assemblies consist of control element assembly shrouds, the control element assembly shroud bolts, and the control element assembly shroud extension shaft guides.The CEA shroud assembly is a welded assembly consisting of sixty one (61) 9-inch ID cylinders approximately 160 inches long located on 16.36 inch centers in a square array. The cylinders are connected to each other by full length plates at 90' to each other, forming square channels with a cylinder at each comer. The plates and cylinders are perforated to permit cross flow between channels for pressure equalization.

The CEA shrouds shield the CEAs from the effects of the coolant cross flow and provide lateral support for the CEA extension shaft when the closure head is removed [38].Report No. 1200347.401.Rl 3-4 V aj C kItIU Assocates, I' INSTRUMEN~TATION ASSEMBLY INSTR~UMENT TUBE iNOZZLE)Figure 3-1. Illustration of the PVNGS Vessel and Internals

[38]Report No. 1200347.401 .R1 3-5 Vsftu g kw!blgri Associates, 1Wc UPPER GUIDE STRUCTURE SUPPORT BARREL UPRGUIDE-+-

ASSEMBLY ______TUBE Ii-r-un, ~r~r Figure 3-2. PVNGS Reactor Vessel and Internals Assembly [38]Report No. 1200347.401.Rl 3-6 V OW&W

• Assoctas, 1W Figure 3-3. PVNGS Modified Upper Guide Structure Assembly [38]Report No. 1200347.401.R1 3-7 jstrsiwu twi*r AIsaOts, I' Figure 3-4. PVNGS Core Support Barrel Assembly [38]Report No. 1200347.401.R1 3-8 V o~Snm&*w hitgdl Assoadaes IncW Figure 3-5. PVNGS Core Shroud Assembly with Full Height Panels (with bottom-mounted ICI) [38]Report No. 1200347.401.Rl V OKOW kuIbdWgit Assomtfes, IncW 3-9 (a)UGS SUPPORT PLATE-, A F j CEA GUIDE TUBES J, L CEA GUIDE TUBE EXTENSIONS (b,)FUEL ALIGNMENT PLATE Figure 3-6. (a) Illustration of a Typical C-E Fuel Alignment Plate, and (b) Radial view Schematic Illustration of Guide Tubes Report No. 1200347.401.Rl 3-10 C aftuchngu hitwur Assadates, 1Wc~

.lulkrate the deep beam guid (nmnber 3). a: well as the fel alignment pim (numbers i and 2)Figure 3-7. Isometric View of Lower Support Structure in the C-E Core Shroud with Full-Height Shroud Plates Report No. 1200347.401 .Rl 3-11 r l t9I1Y Aur t A caatus, W SNUBBER--LUG SUPPORT " BEAMS ICI NOZZLES-LOWER SUPPORT STRUCTURE ICi NOZZLE SUPPORT PLATE Figure 3-8. ICI Support Assembly (Palo Verde Units) [38]Report No. 1200347.401.Rl 3-12 .V Oi&Wuktr Assacates.

IW 3.6 PVNGS Units 1, 2, and 3 Design Distinctions The PVNGS Units are a C-E System 80 design reactor. The three PVNGS units are unique in the C-E fleet and are the only System 80 units in the U.S. PWR fleet. This design has characteristics that are not found in prior C-E units. Some of the unique System 80 design features include:* Bottom-mounted instrumentation (unique for the C-E design).* CEA guide tubes were rolled and welded to the fuel alignment plate to provide protection for the individual CEAs from cross-flow (unlike prior CEA shrouds for cross-flow protection).

• Full-height welded core shroud plate (unlike half-height welded or full-height bolted).* No zirconium alloy based top mounted thimble tubes.o In the C-E designed plants (except PVNGS), zirconium alloy thimble tubes exhibited growth due to irradiation.

This was a degradation mechanism of concern for C-E designed plants and were subsequently replaced.

However, there are no ICI thimbles in the Palo Verde units. Therefore, the thimble growth issue is not applicable to PVNGS." No cast austenitic stainless steel (CASS) materials are present in the PVNGS RVI." No baffle or core shroud bolts are present in the PVNGS core shroud.* There are no core support columns in the PVNGS units." There are no core support plates in the PVNGS units [37].3.7 PVNGS Unit Operating Experience PVNGS Units 1, 2, and 3 have not had any operating experience with degradations or non-conforming conditions

[40].Report No. 1200347.401.R1 3-13 U irnhw kkg* ,sso In c.,

4.0 EXAMINATION AND ACCEPTANCE AND EXPANSION CRITERIA 4.1 Examination Acceptance Criteria 4.1.1 Visual (VT-3) Examination Visual (VT-3) examination has been determined to be an appropriate NDE method for the detection of general degradation conditions in many of the susceptible components.

The ASME Code Section XI, Examination Category B-N-3 [21 ], provides a set of relevant conditions for the visual (VT-3) examination of removable core support structures in IWB-3520.2.

These are: 1. Structural distortion or displacement of parts to the extent that component function may be impaired 2. Loose, missing, cracked, or fractured parts, bolting, or fasteners 3. Corrosion or erosion that reduces the nominal section thickness by more than 5%4. Wear of mating surface that may lead to loss of functionality

5. Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5%For components in the Existing Programs group, these general relevant conditions are sufficient.

However, for components where visual (VT-3) is specified in the Primary or the Expansion group, more specific descriptions of the relevant conditions are provided in Tables 5-1 through 5-4 of this document.

Typical examples are "fractured material" and "completely separated material." One or more of these specific relevant condition descriptions may be applicable to the Primary and Expansion components listed in Tables 5-1 and 5-2 of this document.

The examination acceptance criteria for components requiring visual (VT-3) examinations is thus the absence of the relevant condition(s) specified in Table 5-4 of this document.

The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition, or repair/replacement to remediate the relevant condition.

Report No. 1200347.401.R1 4-1 Aktes,1 4.1.2 Visual (VT-i) Examination Visual (VT-1) examination is defined in the ASME Code Section XI as an examination"conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion." For these guidelines VT-I has only been selected to detect distortions as evidenced by small gaps between the upper-to-lower mating surfaces of C-E welded core shroud assembled in two vertical sections.

The examination acceptance criterion is thus the absence of the relevant condition of gaps that would be indicative of distortion from void swelling.

Visual (VT-1) examinations do not apply to PVNGS reactor vessel internals.

4.1.3 Enhanced Visual (EVT-1) Examination Enhanced visual (EVT-1) examination has the same requirements as the ASME Code Section XI visual (VT-1) examination, with additional requirements given in MRP-228 [14]. These enhancements are intended to improve the detection and characterization of discontinuities taking into account the remote visual aspect of reactor internals examinations.

As a result, EVT-1 examinations are capable of detecting small surface-breaking cracks and sizing surface crack length when used in conjunction with sizing aides (e.g. landmarks, ruler, and tape measure).EVT- 1 examination is the appropriate NDE method for detection of cracking in plates or their welded joints. Thus the relevant condition applied for EVT- 1 examination is the same as for cracking in Section XI which is crack-like surface-breaking indications.

The examination acceptance criterion for EVT- 1 examination is the absence of any detectable surface-breaking indication.

4.1.4 Surface Examination Surface ET (eddy current) examinations are specified as an alternative or as a supplement to visual examinations.

No specific acceptance criteria for surface (ET) examination of PWR internals locations are provided in the ASME Code Section XI. Since surface ET is employed as a signal-based examination, a technical justification is documented in MRP-228 [14]. MRP-228 Report No. 1200347.401 .Ri 4-2 I-Mu hWe kftt Asso1atus, W provides the basis for detection and length sizing of surface-breaking or near-surface cracks.The signal-based relevant indication for surface (ET) is thus the same as the relevant condition for enhanced visual (EVT- 1) examination.

The acceptance criteria for enhanced visual (EVT- 1)examinations are therefore applied when this method is used as an alternative or supplement to visual examination.

4.1.5 Volumetric Examination The intent of volumetric examinations specified for bolts and pins is to detect planar defects. No flaw sizing measurements are recorded or assumed in the acceptance or rejection of individual bolts or pins. Individual bolts or pins are accepted based on the lack of detection of any relevant indications established as part of the examination technical justification.

When a relevant indication is detected in the cross-sectional area of the bolt or pin, that bolt or pin is assumed to be non-functional and the indication is recorded.

A bolt or pin that passes the criterion of the examination is assumed to be functional.

Because there are no baffle-former bolts in the PVNGS design, no volumetric examinations of the internals are needed to meet MRP-227-A requirements.

4.1.6 Physical Measurements Examination Physical measurements can be applied to confirm loss of material due to wear, loss of pre-load, or distortion/deflection caused by void swelling.

The visual inspections are targeted at the locations where displacement or separation of plates is most likely to be noted. The extent and character of the distortion at these locations is discussed in MRP-230. These inspections are included to provide physical validation of the swelling calculations.

If distortion at these locations is not observed, it is reasonable to assume that the MRP-230 analysis continues to bound the behavior of the structure.

Report No. 1200347.401.Rl 4-3 ft" kItwgll Assoutus, 1W 4.2 Expansion Criteria The criteria for expanding the scope of examination from the Primary components to their linked Expansion components are contained in Table 5-4.4.3 Evaluation, Repair, and Replacement Strategy Any condition detected during examinations that do not satisfy the examination acceptance criteria of Section 4.1 shall be entered and dispositioned in the corrective action program.The options listed below will be considered for disposition of such conditions.

Selection of the most appropriate option(s) will be dependent on the nature and location of the indication detected.1. Supplemental examinations will be used in order to further characterize and disposition of a detected condition 2. Engineering evaluations that demonstrate the acceptability of detected conditions

3. Repair to restore a component with a detected condition to acceptable status 4. Replacement of a component The methodology used to perform engineering evaluations to determine the acceptability of a detected condition (item 2 above) shall be conducted in accordance with an NRC approved evaluation methodology.

WCAP- 17096-NP [ 17] and other NRC approved methodologies will be used to provide acceptance criteria for Primary and Expansion category items.4.3.1 Reporting Reporting and documentation of relevant conditions and disposition of indications that do not meet the examination acceptance criteria will be performed consistent with MRP-227-A and the PVNGS Corrective Action Program. APS shall provide a summary report to the EPRI MRP Program Manager of all inspections and monitoring, items requiring evaluation, and new repairs.Report No. 1200347.401 .Rl 4-4 .IWWtWsI W *t AsSocautes, I This report shall be provided within 120 days of the completion of the outage during which the activities occur. This is part of the "Needed" requirement 7.6 under MRP-227-A.

Inspection results having potential industry significance shall be expeditiously reported to the RCS Materials Degradation Program Manager for consideration of reporting under the NEI 03-08, Materials Initiative Protocol [2].4.4 Implementation Schedule The Program Enhancement and Implementation Schedule for PVNGS Units 1, 2, and 3 is provided in Table 5-6, Table 5-7, and Table 5-8, respectively

[44].4.5 Commitment Tracking A summary of actions related to the Aging Management of Reactor Vessel Internals for PVNGS Units 1, 2, and 3 is provided in Table 5-9.Report No. 1200347.401.Rl 4-5 VjjauIr Ifut ,W 5.0 RESPONSES TO NRC SAFETY EVALUATION APPLICANT/LICENSEE ACTION ITEMS As part of the NRC Revision 1 of the Final Safety Evaluation of MRP-227 [4], a number of action items and conditions were specified by the staff. Table 5-5 documents PVNGS's conformance to the Topical Report Conditions and the Applicant/Licensee Action Items in the NRC Safety Evaluation of MRP-227 [4]. Wherever possible, these items have been addressed in the appropriate sections of this document.

All NRC action items and conditions not addressed elsewhere in this document are discussed in this section.5.1 SE Section 4.2.1, Applicant/Licensee Action Item 1 (Applicability of FMECA and Functionality Analysis Assumptions):

As addressed in Section 3.2.5.1 of this SE, each applicant/licensee is responsible for assessing its plant's design and operating history and demonstrating that the approved version of MRP-22 7 is applicable to the facility.

Each applicant/licensee shall refer, in particular, to the assumptions regarding plant design and operating history made in the FMECA and functionality analyses for reactors of their design (i.e., Westinghouse, CE or B& W) which support MRP-22 7 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories.

The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version of MRP-22 7.The assumptions regarding plant design and operating history made in MRP-191 [13] are appropriate for PVNGS. The FMECA and functionality analyses were based on the assumption of 30 years of operation with high leakage core loading patterns; therefore, PVNGS Units 1, 2, and 3 are bounded by the assumption in MRP-191 [13].As discussed in Section 1.8.4.1 of this document, operations at PVNGS conform to the assumptions in Section 2.4 of MRP-227-A

[3].Report No. 1200347.401.R1 5-1 cJS ftow ft*lwgly Assoatus, W'

• PVNGS Units 1, 2, and 3 historic core management practices meet the requirements of MRP-227-A* PVNGS Units 1, 2, and 3 operate as base load units* No design changes were implemented beyond those identified in general industry guidance or recommended by the vendor (C-E or Westinghouse) 5.2 SE Section 4.2.2, 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 responsible for identifying which R VI components 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 identify whether these tables contain all of the R VI components that are within the scope of the LR for their facilities in accordance with 10 CFR 54.4. If the tables do not identify all the R VI components that are within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) and propose 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 managed for the period of extended operation.

The information contained in Table 4-5 of MRP-191 [13] was reviewed and that review determined that this table contained all of the RVI components that are within the scope of license renewal for PVNGS. The aging management review performed as part of the PVNGS LRA is described in Section 1.7.1 and summarized in Table B-1 of this document.Report No. 1200347.401.R1 5-2 .Asscitus tnc 5.3 SE Section 4.2.3, 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/licensees of CE and Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an applicant

's/licensee

's existing programs, or to identify changes to programs that should be implemented to manage the aging of these components for 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 applicant

's/licensee's AMP application.

The CE and Westinghouse components identified for this type ofplant-specific evaluation include: CE thermal shield positioning pins 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-227).The SE for MRP-227 [4] requires C-E plants to evaluate whether existing plant-specific programs are adequate to manage the aging effects of thermal shield positioning pins and in-core instrument thimble tubes. These actions are not applicable because PVNGS does not have ICI thimble tubes or a thermal shield.5.4 SE Section 4.2.4, 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/licensees shall confirm that the core support structure upper flange weld was stress relieved during the original fabrication of Reactor Pressure Vessel in order to confirm the applicability of MRP-227, as approved by the NRC, to their facility.

If the upper flange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component.

If necessary, the examination methods andfrequency for non-stress relieved B& W core support structure upper flange welds shall be consistent with the recommendations in MRP-22 7, as approved by the NRC, for the Westinghouse and CE upper core support barrel welds. The examination coverage for this B& W flange weld shall conform to the staff's imposed criteria as described in Section 3.3.1 and 4.3.1 Report No. 1200347.401 .R 5-3 Ujj MM&ru W b i Anocates, W 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 action does not apply to C-E designed units.5.5 SE Section 4.2.5, 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/licensees shall identify plant-specific acceptance criteria to be applied when performing the physical measurements required by the NRC-approved version of MRP-227for loss of compressibility for Westinghouse hold down springs, and for distortion in the gap between the top and bottom core shroud segments in CE units with core barrel shrouds assembled in two vertical sections.

The applicant/licensee shall include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are 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 of MRP-22 7.There are no gaps in the core shroud segments for the Palo Verde Units 1, 2, and 3 designs because the shrouds are made from full height panels. Therefore, this action to measure distortion of the gaps does not apply for PVNGS.5.6 SE Section 4.2.6, Applicant/Licensee Action Item 6 (Evaluation of Inaccessible B&W Components):

As addressed in Section 3.3.6 in this SE, MiRP-22 7 does not propose to inspect the following inaccessible components:

the B& W core barrel cylinders (including vertical and circumferential seam welds), B& Wformer plates, B& W external baffle-to-baffle bolts and their locking devices, B& W core barrel-to-former bolts and their locking devices, and B& W core barrel assembly internal baffle-to-baffle bolts. The MRP also identified that although the B& W core barrel Report No. 1200347.401.R1 5-4 h wMeigf Assadates.

nW assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques.

Applicants/licensees shall justify the acceptability of these components for continued operation through the period of extended operation by performing an evaluation, or by proposing a scheduled replacement of the components.

As part of their application to implement the approved version of MRP-227, applicants/licensees shall provide theirjustification for the continued operability of each of the inaccessible components and, if necessary, provide their plan for the replacement of the components for NRC review and approval.This action does not apply to the C-E designed units.5.7 SE Section 4.2.7, Applicant/Licensee Action Item 7 (Plant-Specific Evaluation of CASS Materials):

As discussed in Section 3.3.7 of this SE, the applicants/licensees of B& W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B& WIMI guide tube assembly spiders and CRGTspacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operation or for additional R VI components that may befabricatedfrom CASS, martensitic stainless steel or precipitation hardened stainless steel materials.

These analyses shall also consider the possible lo.s offracture toughness in these components due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques.

The requirement may not apply to components that were previously evaluated as not requiring aging management during development ofMRP-227.

That is, the requirement would apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation.

The Report No. 1200347.401.R1 5-5 U sir* ASSOCkuts, 1W applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-22 7.The SE for MRP-227 [4] requires the applicants/licensees of C-E reactors to develop plant-specific analyses to be applied for their facilities to demonstrate that the C-E lower support columns will maintain their functionality during the period of extended operation.

It also requires that the licensee provide technical justification that RVI components that may be fabricated from CASS, martensitic stainless steel, or precipitation hardened stainless steel materials will maintain their functionality during the period of extended operation.

The PVNGS units do not have lower support plates or lower support columns as part of the reactor vessel internals.

The welds in the deep beam supports carry loads into the core support barrel lower flange. Inspection requirements provided in the Primary components tables of MRP-227-A (identified in Table 5-1 of this document) addresses this requirement.

The beam-to-beam welds, in the axial elevation from the beam top surface to four inches below would be inspected using EVT-1 examination, no later than two refueling outages from the beginning of the license renewal period with subsequent examinations to be performed on a ten-year interval, if adequacy of remaining fatigue live cannot be demonstrated.

The PVNGS RVI components do not contain CASS, martensitic stainless steel, or precipitation hardened stainless steel materials

[12]. This was confirmed by reviewing the list of materials in the Palo Verde Units 1, 2, and 3 RV internals components given in Tables C-l, D- 1, and E- 1.5.8 SE Section 4.2.8, Applicant/Licensee Action Item 8 (Submittal of Information for Staff Review and Approval):

As addressed in Section 3.5.1 in this SE, applicants/licensee shall make a submittal for NRC review and approval to credit their implementation of MRP-22 7, as amended by this SE, as an AMP for the RVI Components at their facility.

This submittal shall include the information identified in Section 3.5.1 of this SE.Section 3.5.1 of SE (Submittal of Information for Staff Review and Approval):

Report No. 1200347.40 L.R1 5-6 -S&weu M r*r Associtus, 1W In addition to the implementation of MRP-227 in accordance with NEI 03-08, applicants/licensees whose licensing basis contains a commitment to submit a PWR R VI AMP and/or inspection program shall also make a submittal for NRC review and approval to credit their implementation of MRP-22 7, as amended by this SE. An applicant

's/licensee's application to implement MRP-22 7, as amended by this SE shall include the following items (1) and (2).Applicants who submit applications for LR after issuance of this SE shall, in accordance with the NUREG-1801, Revision 2, submit the information provided in the following items (1) through (5)for staff review and approval.1. An AMP for the facility that addresses the 10 program elements as defined in NUREG-1801, Revision 2, AMP XI.M16A.The attributes of the PVNGS RVI AMP and their compliance with the ten elements of NUREG- 1801 (GALL Report), Revision 2, Chapter XI.M 16A, "PWR Vessel Internals" [ 10]that are essential for successful management of component aging are described in Section 2.3 of this document.2. To ensure the MRP-22 7 program and plant-specific action items will be carried out by applicants/licensees, applicants/licensees are to submit an inspection plan which addresses the identifiedplant-specific action items for staff review and approval consistent with the licensing basis for the plant. If an applicant/licensee plans to implement an AMP which deviates from the guidance provided in MRP-227, as approved by the NRC, the applicant/licensee shall identify where their program deviates from the recommendations of MRP-227, as approved by the NRC, and shall provide ajustification for any deviation which includes a consideration of how the deviation affects both"Primary" and "Expansion" inspection category components.

Report No. 1200347.401.R1 5-7 V OWWWklutwkUt, AssktaS, Ic The aging management program plan for the PVNGS Units 1, 2, and 3 will not deviate from the recommendations of MRP-227-A.

Inspection of Primary, Expansion, and components credited as part of plant specific existing programs provided in Table 5-1 through Table 5-4 of this document will be performed in accordance with the requirements of MRP-227-A.

3. The regulation at 10 CFR 54.21(d) requires that an FSAR supplement for the facility contain a summary description of the programs and activities for managing the effects of aging and the evaluation of TLAAs for the period of extended operation.

Those applicants for LR refirencing MRP-22 7, as approved by the NRC, for their R VI component AMP shall ensure that the programs and activities specified as necessary in MRP-227, as approved by the NRC, are summarily described in the FSAR supplement.

APS will incorporate a summary description of the PVNGS Units 1, 2, and 3 reactor vessel internals aging management program into the Updated Final Safety Analysis Report (UFSAR) no later than the next scheduled update required by 10 CFR 50.71 (e) following NRC approval of the program. This summary description shall reference the "PWR Internals Aging Management Program Plan for Palo Verde Nuclear Generating Station Units 1, 2, and 3.,'4. The regulation at 10 CFR 54.22 requires each applicant for LR to submit any TS changes (and the justification for the changes) that are necessary to manage the effects of aging during the period of extended operation as part of its LR application (LRA). For the plant CLBs that include mandated inspection or analysis requirements for R VI either in the operating license for the facility or in the facility TS, the applicant/licensee shall compare the mandated requirements with the recommendations in the NRC-approved version of MRP-22 7. If the mandated requirements differ from the recommended criteria in MRP-22 7, as approved by the NRC, the conditions in the applicable license conditions Report No. 1200347.401.Rl 5-8 Ov owtCILaI I g Assiatus, 1Wc.'

or TS requirements take precedence over the MRP recommendations and shall be complied with.No changes to the Technical Specifications (TS) are required.5. Pursuant to 10 CFR 54.21(c)(1), the applicant is required to identify all analyses in the CLBfor their RVI components that conform to the definition of a TLAA in 10 CFR 54.3 and shall identify these analyses as TLAAs for the application in accordance with the TLAA identification requirement in 10 CFR 54.21(c) (1). MPRP-22 7 does not specifically address the resolution of TLAAs that may apply to applicant/licensee R VI components.

Hence, applicants/licensees who implement MARP-227, as approved by the NRC, shall still evaluate the CLBfor their facilities to determine if they have plant-specific TLAAs that shall be addressed.

If so, the applicant

's/licensee

's TLAA shall be submitted for NRC review along with the applicant

's/licensee

's application to implement the NRC-approved version of MRP-22 7.For those cumulative usage factor (CUF) analyses that are TLAAs, the applicant may use the PWR Vessel Internals Program as the basis for accepting these CUF analyses in accordance with 10 CFR 54.21(c) (1) (iii) only if the R VI components in the CUF analyses are periodically inspected for fatigue-induced cracking in the components during the period of extended operation.

The periodicity of the inspections of these components shall bejustified to be adequate to resolve the TLAA. Otherwise, acceptance of these TLAAs shall be done in accordance with either 10 CFR 54.21 (c) (1) (i) or (ii), or in accordance with 10 CFR 54.21(c) (1) (iii) using the applicant's program that corresponds to NUREG-1801, Revision 2, AMP XM1, "Metal Fatigue of Reactor Coolant Pressure Boundary Program. " To satisfy the evaluation requirements of ASME Code,Section III, Subsection NG-2160 and NG-3121, the existing fatigue CUF analyses shall include the effects of the reactor coolant system water environment.

Report No. 1200347.401 .R1 5-9 v aIIWIUh"If ASSowatas, I.

This action item addresses the cumulative usage factor (CUF) analyses that are TLAAs for the reactor vessel internals.

PVNGS will use the PWR Vessel Internals Program and the Reactor Coolant System Transient and Operating Cycles Program as the basis for verifying these CUF analyses in accordance with 10 CFR 54.2 1 (c)(1)(iii).

The logic for this approach contains four steps: (1) identifying the core support structures with existing CUF fatigue analyses; (2) using some form of cycle monitoring to determine a conservative schedule for potential periodic examinations of those components; (3) carrying out the periodic examinations; and (4) providing an engineering evaluation of relevant conditions detected during those examinations to justify continued operation, or carrying out repair or replacement of internals components as required.

In the SER for license renewal, the 10 CFR 54.21 (c)(1)(iii) approach has been accepted (low cycle monitoring, high cycle vibration is not a TLAA), and no requirement for reactor water environmental effects was imposed for reactor vessel internals.

Therefore, no further action is required to satisfy this A/LAI other than the commitment to use the PWR Vessel Internals Program and the Reactor Coolant System Transient and Operating Cycles Program as the basis for verifying any CUF analyses in accordance with 10 CFR 54.21 (c)(1)(iii).

Report No. 1200347.401.Rl 5-10 V on&* ' 'u ,1 Table 5-1. C-E Plants Primary Category Components from Table 4-2 of MRP-227-A

[3]Effect Examination Item Applicability (Mechanism)

Expansion Link(') Method/FrequencyU 1 Examination Coverage Comments Core Shroud Bolted plant designs Cracking Core support column Baseline volumetric (UT) 100% of accessible bolts(3), or as Assembly (IASCC, bolts, barrel-shroud examination between 25 and 35 supported by plant-specific (Bolted) Fatigue) bolts EFPY, with subsequent justification.

Heads are Core shroud examination after 10 to 15 accessible from the core side.bolts Aging additional EFPY to confirm stability UT accessibility may be affected Management of bolting pattern. Re-examination by complexity of head and (Not applicable (IE and ISR)(2) for high-leakage core designs locking device designs.for PVNGS) requires continuing inspections on a ten-year interval.See Figure 4-24 of MRP-227-A N/A Core Shroud Plant designs with core Cracking Remaining axial Enhanced visual (EVT-1) Axial and horizontal weld seams Assembly shrouds assembled in (IASCC) welds examination no later than 2 at the core shroud re-entrant (Welded) two vertical sections refueling outages from the comers as visible from the core Core shroud Aging beginning of the license renewal side of the shroud, within six plate-former Management period and subsequent inches of central flange and plate weld (IE) (2? examination on a ten-year interval, horizontal stiffeners.

N/A (Not applicable See Figures 4-12 and 4-14 of for PVNGS) MRP-227-A Report No. 1200347.401.R1 5-11!VSftnia tuiw ~~y Assocates, kIm?

Table 5-1. C-E Plants Primary Category Components from Table 4-2 of MRP-227-A

[3] (continued)

Effect Examination Item Applicability (Mechanism)

Expansion Link(l) Method/Frequency(1)

Examination Coverage Comments Core Shroud Plant designs with Cracking (IASCC) Remaining axial Enhanced visual (EVT-1) Axial weld seams at the core PVNGS Unit 1: Assembly core shrouds welds, ribs and rings examination no later than 2 shroud re-entrant corners, at the Enhanced visual (Welded) assembled with full- Aging Management refueling outages from the core mid-plane

(+/- three feet in inspections to be Shroud plates height shroud (IE)( beginning of the license renewal height) as visible from the core performed in plates. period and subsequent side of the shroud. 2026.examination on a ten-year interval.

2026.See Figure 4-13 of MRP-227-A PVNGS Unit 2: Enhanced visual inspections to be performed in 2027.PVNGS Unit 3: Enhanced visual inspections to be performed in 2028.Core Shroud Bolted plant designs Distortion None Visual (VT-3) examination no later Core side surfaces as indicated.

Assembly (Void Swelling), than 2 refueling outages from the (Bolted) including:

beginning of the license renewal See Figures 4-25 and 4-26 of Assembly

• Abnormal period. Subsequent examinations MRP-227-A interaction with fuel on a ten-year interval.(Not Applicable assemblies for PVNGS)

• Gaps along high fluence shroud plate joints 9 Vertical N/A displacement of shroud plates near high fluence joint Aging Management (IE)Report No. 1200347.401.Rl 5-12 VjJ1b7AatrIfwhudly Associaes, Inc?

Table 5-i: C-E Plants Primary Category Components from Table 4-2 of MRP-227-A

[3] (continued)

Item Applicability Effect (Mechanism)

Expansion Link(l) Examination Examination Coverage Comments Method/Frequency(I)ExmntoCveaeomns Core Shroud Plant designs with Distortion None Visual (VT-1) examination no later If a gap exists, make three to five Assembly core shrouds than 2 refueling outages from the measurements of gap opening (Welded) assembled in two (Void Swelling), as beginning of the license renewal from the core side at the core Assembly vertical sections evidenced by period. Subsequent examinations shroud re-entrant comers. Then, separation between on a ten-year interval, evaluate the swelling on a plant-(Not the upper and lower specific basis to determine Applicable for core shroud frequency and method for PVNGS) segments additional examinations.

Aging Management See Figures 4-12 and 4-14 of (IE) MRP-227-A Core Support All plants Cracking (SCC) Lower core support Enhanced visual (EVT-1) 100% of the accessible surfaces PVNGS Unit 1: Barrel beams examination no later than 2 of the upper flange weld. (4) Enhanced visual Assembly Core support barrel refueling outages from the inspections to be Upper(core assembly upper beginning of the license renewal performed in support barrel) cylinder period. Subsequent examinations See Figure 4-15 of MRP-227-A 2026.flange weld Upper core barrel on a ten-year interval.flange PVNGS Unit 2: Enhanced visual inspections to be performed in 2027.PVNGS Unit 3: Enhanced visual inspections to be performed in 2028.Report No. 1200347.401.Rl 5-13 VjjSfruc htuIntd~y AssacWIats, IWO~

Table 5-1. C-E Plants Primary Category Components from Table 4-2 of MRP-227-A

[3] (continued)

Examination Item Applicability Effect (Mechanism)

Expansion Link(l) Method/Frequency(i)

Examination Coverage Comments Core Support All plants Cracking (SCC, IASCC) Lower cylinder axial Enhanced visual (EVT-1) 100% of the accessible surfaces PVNGS Unit 1: Barrel welds examination no later than 2 of the lower cylinder welds (4) Enhanced visual Assembly Aging Management (IE) refueling outages from the inspections to be Lower cylinder beginning of the license renewal performed in girth welds period. Subsequent examinations See Figure 4-15 of MRP-227-A 2026.on a ten-year interval.

2026.PVNGS Unit 2: Enhanced visual inspections to be performed in 2027.PVNGS Unit 3: Enhanced visual inspections to be performed in 2028.Lower All plants Cracking (SCC, IASCC, None Visual (VT-3) examination no later 100% of the accessible surfaces Support Fatigue including than 2 refueling outages from the of the core support column Structure damaged or fractured beginning of the license renewal welds(5)Core support material) period.column welds Aging Management (IE, Subsequent examinations on a (Not Applicable TE) ten-year interval.for PVNGS)Report No. 1200347.401.R1 5-14 Vjj rf~t9WIfifwudffAssociateS, lao Table 5-1. C-E Plants Primary Category Components from Table 4-2 of MRP-227-A

[3] (continued)

EffectExamination Item Applicability Effect Exaso iklExmntoExmainCvegeomns (Mechanism)

Expansion Link(1) Method/Frequency(I)

Examination Coverage Comments Core Support All plants Cracking None If fatigue life cannot be Examination coverage to be PVNGS Unit 1: Barrel (Fatigue) demonstrated by time limited aging defined by evaluation to Enhanced visual Assembly analysis (TLAA), enhanced visual determine the potential location inspections to be Lower flange (EVT-1) examination, no later than and extent of fatigue cracking.

performed in weld 2 refueling outages from the beginning of the license renewal See Figures 4-15 and 4-16 of 2026.period. Subsequent examination MRP-227-A PVNGS Unit 2: on a ten-year interval.

Enhanced visual inspections to be performed in 2027.PVNGS Unit 3: Enhanced visual inspections to be performed in 2028.Lower All plants with a core Cracking None If fatigue life cannot be Examination coverage to be Support support plate (Fatigue) demonstrated by time limited aging defined by evaluation to Structure analysis (TLAA), enhanced visual determine the potential location Core support Aging (EVT-1) examination, no later than and extent of fatigue cracking.

N/A plate Management 2 refueling outages from the (IE) beginning of the license renewal See Figure 4-16 of MRP-227-A (Not applicable period. Subsequent examination for PVNGS) on a ten-year interval.Upper All plants with core Cracking None If fatigue life cannot be Examination coverage to be PVNGS Unit 1: Internals shrouds assembled with (Fatigue) demonstrated by time limited aging defined by evaluation to Enhanced visual Assembly full-height shroud plates analysis (TLAA), enhanced visual determine the potential location inspections to be Fuel alignment (EVT-1) examination, no later than and extent of fatigue cracking.

performed in plate 2 refueling outages from the beginning of the license renewal See Figure 4-17 of MRP-227-A 2026.period. Subsequent examination PVNGS Unit 2: on a ten-year interval.

Enhanced visual inspections to be performed in 2027.PVNGS Unit 3: Enhanced visual insl~ections to be performed in 2028.Report No. 1200347.401.Rl 5-15 7sbt, Iftg AOMtS, so, Table 5-1. C-E Plants Primary Category Components from Table 4-2 of MRP-227-A

[3] (continued)

Item Applicability Effect (Mechanism)

Expansion Link(l) Examination Examination Coverage Comments Method/Frequency(1)

Control All plants with Cracking (SCC, Remaining instrument Visual (VT-3) examination, no later 100% of tubes in peripheral CEA N/A Element instrument guide Fatigue) that results in guide tubes within the than 2 refueling outages from the shroud assemblies (i.e., those Assembly tubes in the CEA missing supports or CEA shroud beginning of the license renewal adjacent to the perimeter of the Instrument shroud assembly separation at the assemblies, period. Subsequent examination fuel alignment plate).guide tubes welded joint between on a ten-year interval.the tubes and (Not applicable supports Plant-specific component integrity for PVNGS) assessments may be required if degradation is detected and remedial action is needed.Lower All plants with Cracking (Fatigue)

None Enhanced visual (EVT-1) Examine beam-to-beam welds, in PVNGS Unit 1: Support core shrouds that results in a examination, no later than 2 the axial elevationi from the beam Enhanced visual Structure assembled with detectable surface- refueling outages from the top surface to four inches below, inspections to be Deep beams full-height shroud breaking indication in beginning of the license renewal performed in plates. the welds or beams period. Subsequent examination See Figure 4-19 of MRP-227-A on a ten-year interval, if adequacy 2026.Aging Management of remaining fatigue life cannot be (IE) demonstrated.

PVNGS Unit 2: Enhanced visual inspections to be performed in 2027.PVNGS Unit 3: Enhanced visual inspections to be performed in 2028.Notes: 1. Examination acceptance criteria and expansion criteria for the C-E components are in Table 5-4 (MRP-227-A Table 5-2)2. Void swelling effects on this component is managed through management of void swelling on the entire core shroud assembly.3. A minimum of 75% of the total population (examined

+ unexamined), including coverage consistent with the Expansion criteria in Table 5-4, 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 5-4, must be examined from either the inner or outer diameter for inspection credit.5. A minimum of the total population of core support welds.Report No. 1200347.401.R1 5-16 cjjsfntildrd IIteg* Assockaes, Ina?

Table 5-2. C-E Plants Expansion Category Components from Table 4-5 of MRP-227-A

[3]Item Applicability Effect (Mechanism)

Primary Link Examination Examination Method/Frequency (1) Coverage/Frequency (I) Comments Core Shroud Bolted plant Cracking (IASCC, Core shroud Volumetric (UT) 100 % (or as supported by Assembly designs Fatigue) bolts examination, plant-specific justification)

(2)(Bolted) of barrel-shroud and guide Barrel-shroud bolts Aging Management Re-inspection every 10 lug insert bolts with neutron (IE and ISR) years following initial fluence exposures

> 3 (Not applicable for inspection.

displacements per atom PVNGS) (dpa). N/A See Figure 4-23 of MRP-227-A Core Support Barrel All plants Cracking (SCC, Upper (core Enhanced visual (EVT-1) 100 % of accessible welds Contingency if Assembly Fatigue) support barrel) examination, and adjacent base metal (2) indications are found Lower core barrel flange flange weld. in EVT-1 exam of Re-inspection every 10 See Figure 4-15 of Upper (core support years following initial MRP-227-A barrel) flange welds inspection, in 2026, 2027 and 2028 for PVNGS 1, 2, and 3, respectively.

Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1) 100 % of accessible surface Contingency if Assembly support barrel) examination.

of the weld and adjacent indications are found Upper Cylinder Aging Management flange weld base.(2) in EVT-1 exam of (including welds) (IE) Re-inspection every 10 Upper (core support years following initial See Figure 4-15 of barrel) flange welds inspection MRP-227-A in 2026, 2027 and 2028 for PVNGS 1,2, and 3, respectively.

Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1) 100% of accessible bottom Contingency if Assembly support barrel) examination.

surface of the flange (2) indications are found Upper Core Barrel flange weld Re-inspection every 10 in EVT-1 exam of Flange years following initial See Figure 4-15 of Upper (core support yearscfollwing-initAl barrel) flange welds inspection MRP-227-A in 2026, 2027 and 2028 for PVNGS 1, 2,_and 3, respectively.

Report No. 1200347.401.R1 5-17 Vj1Sbitww

~1fgdy Mssaciues, kW~

Table 5-2. C-E Plants Expansion Category Components from Table 4-5 of MRP-227-A

[3] (continued)

Item Applicability Effect (Mechanism)

Primary Link Examination Examination Comments Method/Frequency (1) Coverage/Frequency (1)Core Support Barrel All plants Cracking (SCC) Core barrel Enhanced visual (EVT-1) 100% of one side of the Contingency if Assembly assembly girth examination, with initial and accessible weld and adjacent indications are found Core barrel assembly welds subsequent examinations base metal surfaces for the in EVT-1 exam of axial welds dependent on the results of weld with the highest Core barrel assembly core barrel assembly girth calculated operating stress. girth welds in 2026, weld examinations.

2027 and 2028 for See Figures 4-15 of PVNGS 1, 2, and 3, MRP-227-A.

respectively.

Lower Support All plants except Cracking (SCC, Upper (core Visual (EVT-1) examination.

100% of accessible Structure those with core Fatigue) including support barrel) surface.(2)

Lower core support shrouds assembled damaged or fractured flange weld Re-inspection every 10 N/A beams with full-height material years following initial See Figures 4-16 and 4-31 of shrod pltesMRP-227-A.(Not applicable for shroud plates inspection.

PVNGS)Core Shroud Bolted plant Cracking (IASCC, Core shroud Ultrasonic (UT) 100 % (or as supported by N/A Assembly designs Fatigue) bolts examination, plant-specific analysis) of (Bolted) core support column bolts Core support column Aging Management Re-inspection every 10 with neutron fluence bolts (IE) years following initial exposures

> 3 dpa. (2)inspection.(Not applicable for See Figures 4-16 and 4-33 of PVNGS) MRP-227-A Core Shroud Plant designs with Cracking (IASCC) Shroud plates of Enhanced visual (EVT-1) Axial weld seams other than Contingency if Assembly core shrouds welded core examination, the core shroud re-entrant indications are found (Welded) assembled with Aging Management shroud comer welds at the core mid- in EVT-1 exam of Remaining axial welds, full-height shroud (IE) assemblies Re-inspection every 10 plane, plus ribs and rings. Shroud plates of Ribs and rings plates. years following initial See Figure 4-13 of welded core shroud inspection.

MRP-227-A assemblies in 2026, 2027 and 2028 for PVNGS 1, 2, and 3, respectively.

Report No. 1200347.401.R8 5-18 Cj~fruturw hnifgr* Assocala'ls, Inc?

Table 5-2. C-E Plants Expansion Category Components from Table 4-5 of MRP-227-A

[3] (continued)

Item Applicability Effect (Mechanism)

Primary Link Examination Examination Comments Method/Frequency (1) Coverage/Frequency (1)Control Element All plants with Cracking (SCC, Peripheral Visual (VT-3) examination.

100% of tubes in CEA shroud N/A Assembly instrument guide Fatigue) that results in instrument guide assemblies.

(2)Remaining instrument tubes in the CEA missing supports or tubes within the Re-inspection every 10 guide tubes shroud assembly.

separation at the CEA shroud years following initial See Figure 12 welded joint between assemblies, inspection.(Not applicable for the tubes and supports.PVNGS)1.2.Examination acceptance criteria for the C-E components are in Table 5-4 (MRP-227-A Table 5-2).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 accessible and inaccessible portions).

Report No. 1200347.401.R1 5-19!VjSbutwwI hitdwqy Associates, Inc?

Table 5-3. C-E Plants Existing Program Components Credited in Table 4-8 of MRP-227-A

[3]Item Applicability Effect Primary Examination Method Examination Coverage Comments (Mechanism)

Link Core Shroud All plants Loss of material ASME Visual (VT-3) examination, general condition First 10-year ISI after 40 years To be inspected Assembly (Wear) Code examination for detection of excessive or of operation, and at each Guide lugs Section Xl asymmetrical wear. subsequent inspection interval.Guide lug Aging inserts and Management Accessible surfaces at bolts (ISR) specified frequency Lower All plants with core Cracking (SCC, ASME Visual (VT-3) examination to detect severed Accessible surfaces at Support shrouds assembled with IASCC, Fatigue) Code fuel alignment pins, missing locking tabs, or specified frequency Structure full-height shroud plates Section Xl excessive wear on the fuel alignment pin Fuel Aging nose or flange.alignment pins Management To be inspected (IE and ISR)Lower All plants with core Loss of Material ASME Visual (VT-3) examination Accessible surfaces at Support shrouds assembled in two (Wear) Code specified frequency Structure vertical sections Section Xl Fuel Aging alignment pins Management N/A (IE and ISR)(Not applicable for PVNGS)Core Barrel All plants Loss of Material ASME Visual (VT-3) examination Area of the upper flange To be inspected Assembly (Wear) Code potentially susceptible to wear Upper flange Section Xl Report No. 1200347.401.Rl 5-20 Cj fS~tXWW ~ 1ygrI~ Associaes, hwO Table 5-4. C-E Plants Examination Acceptance and Expansion Criteria from Table 5-2 of MRP-227-A

[3] Applicable to PVNGS Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Additional Examination (Note 1) Acceptance Criteria Core Shroud Assembly Bolted plant designs Volumetric (UT) a. Core support a. Confirmation that >5% of the a and b. The examination (Bolted) examination, column bolts core shroud bolts in the four acceptance criteria for the UT of plates at the largest distance the core support column bolts Core The examination

b. Barrel-shroud from the core contain and barrel-shroud bolts shall be acceptance criteria for the bolts unacceptable indications shall established as part of the (Not applicable for PVNGS) UT of the core shroud bolts require UT examination of the examination technical shall be established as part lower support column bolts justification.

of the examination technical barrel within the next 3 refueling justification, cycles.b. Confirmation that > 5% of the core support column bolts contain unacceptable indications shall require UT examination of the barrel-shroud bolts within the next 3 refueling cycles.Core Shroud Assembly Plant designs with core Visual (EVT-1) examination.

Remaining axial Confirmation that a surface- The specific relevant condition is (Welded) shrouds assembled in two welds breaking indication

> 2 inches in a detectable crack-like surface vertical sections length has been detected and indication.

The specific relevant sized in the core shroud plate-Core shroud plate-former condition is a detectable former plate weld at the core crack-like surface indication soudere-entrat core plate weld shroud re-entrant comers (as visible from the core side of the shroud), within 6 inches of the (Not applicable for PVNGS) central flange and horizontal stiffeners, shall require EVT-1 examination of all remaining axial welds by the completion of the next refueling outage.Report No. 1200347.401.R1 5-21 V a&noud"i ~1fgy Assocates, Inc?

Table 5-4. C-E Plants Examination Acceptance Criteria from Table 5-2 of MRP-227-A

[3] Applicable to PVNGS (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Core Shroud Assembly Plant designs with core Visual (EVT-1) examination, a. Remaining axial a. Confirmation that a surface The specific relevant condition is (Welded) shrouds assembled with full- welds breaking indication

> 2 inches in a detectable crack-like surface height shroud plates length has been detected and indication.

Shroud plates The specific relevant b. Ribs and rings sized in the axial weld seams at condition is a detectable the core shroud re-entrant crack-like surface indication corners at the core mid-plane shall require EVT-1 or UT examination of all remaining axial welds by the completion of the next refueling outage.b. If extensive cracking is detected in the remaining axial welds, an EVT-1 examination shall be required of all accessible rib and ring welds by the completion of the next refueling outage.Core Shroud Assembly Bolted plant designs Visual (VT-3) examination.

None N/A N/A (Bolted) The specific relevant conditions are evidence of Assembly abnormal interaction with (Not applicable for PVNGS) fuel assemblies, gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near high fluence joints Report No. 1200347.401.R1 5-22 V f~SuturI hifsgdiy Assaciaes, kIm?

Table 5-4. C-E Plants Examination Acceptance Criteria from Table 5-2 of MRP-227-A

[3] Applicable to PVNGS (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Core Shroud Assembly Plant designs with core Visual (VT-1) examination.

None N/A N/A (Welded) shrouds assembled in two The specific relevant vertical sections condition is evidence of Assembly physical separation between (Not applicable for PVNGS) the upper and lower core shroud sections.Core Support Barrel All plants Visual (EVT-1) examination.

Lower core support Confirmation that a surface The specific relevant condition is Assembly The specific relevant beams breaking indication

>2 inches in a detectable crack-like surface Upper (core support barrel) condition is a detectable length has been detected and indication.

flange weld crack-like surface indication.

Upper core barrel sized in the upper flange weld cylinder (including shall require that an EVT-1 welds) examination of the lower core support beams, upper core Upper core barrel barrel cylinder and upper core flange barrel flange be performed by the completion of the next refueling outage Core Support Barrel All plants Visual (EVT-1) examination.

Lower cylinder axial Confirmation that a surface The specific relevant condition Assembly The specific relevant welds breaking indication

>2 inches in for the expansion lower cylinder Lower cylinder girth welds condition is a detectable the length has been detected axial welds is a detectable crack-like surface indication, and sized in the lower cylinder crack-like surface indication girth weld shall require an EVT-1 examination of all accessible lower cylinder axial welds by the completion of the next refueling outage.Report No. 1200347.401.RI 5-23 C an"wowu ~wrly Assaowlaes, Inc?

Table 5-4. C-E Plants Examination Acceptance Criteria from Table 5-2 of MRP-227-A

[3] Applicable to PVNGS (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Lower Support Structure All plants Visual (VT-3) examination.

None None N/A Core support column welds The specific relevant (Not Applicable for PVNGS) condition is missing or separated welds.Core Support Barrel All plants Visual (EVT-1) examination.

None N/A N/A Assembly The specific relevant Lower flange weld condition is a detectable crack-like indication.

Lower Support Structure All plants with a core support Visual (EVT-1) examination.

None N/A N/A Core support plate plate The specific relevant (Not applicable for PVNGS) condition is a detectable crack-like surface indication.

Upper Internals All plants with core shrouds Visual (EVT-1) examination.

None N/A N/A AssembIyJ2) assembled with full-height The specific relevant Fuel alignment plate shroud plates condition is a detectable crack-like surface indication.

Report No. 1200347.401.R1 5-24 VjSfritd"u Iutigdy Assocaaes, MOc Table 5-4. C-E Plants Examination Acceptance Criteria from Table 5-2 of MRP-227-A

[3] Applicable to PVNGS (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Control Element All plants with instrument tubes Visual (VT-3) examination.

Remaining Confirmed evidence of missing The specific relevant conditions Assembly(3) in the CEA shroud assembly The specific relevant instrument tubes supports or separation at the are missing supports and Instrument guide tubes conditions are missing within CEA shroud welded joint between the tubes separation at the welded joint supports and separation at assemblies and supports shall require the between the tubes and the (Not applicable for PVNGS) the welded joint between the visual (VT-3) examination to be supports.tubes and the supports.

expanded to the remaining instrument tubes within the CEA shroud assemblies by completion of the next refueling outage.Lower Support Structure All plants with core shrouds Visual (EVT-1) examination.

None N/A N/A Deep Beams assembled with full-height shroud plates The specific relevant condition is a detectable crack-like indication Note: 1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

Report No. 1200347.401.Rl 5-25 Report o. 120047.401 R1 5-2 hifwgdy Associaes, h70' Table 5-5. PVNGS Response to the NRC Final Safety Evaluation of MRP-227-A

[4]MRP-227 SE Item PVNGS Response SE Section 4.1.1, Topical Report In accordance with SE Section 4.1.1, the Lower Core Support Beams, Core Condition 1: Moving components Support Barrel Assembly Upper Cylinder and Upper Core Barrel Flange have from "No Additional Measures" to been added to the PVNGS "Expansion" inspection category and are contained"Expansion" category.

in Table 5-2. The components are linked to the "Primary" components Lower Support Structure Deep Beams and Core Support Barrel Upper (core support barrel) flange weld.SE Section 4.1.2, Topical Report In accordance with SE Section 4.1.2, the Core Support Barrel Assembly Condition 2: Inspection of Lower Cylinder Girth Welds have been added to the PVNGS "Primary" components subject to irradiation-inspection category and are contained in Table 5-1. The examination method assisted stress corrosion cracking.

is consistent with the MRP recommendations for these components, the examination coverage conforms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination frequency is on a 10-year interval consistent with other "Primary" inspection category components.

SE Section 4.1.3, Topical Report Core support columns and core support column welds are not applicable to Condition 3: Inspection of high the PVNGS Units. However, in accordance with SE Section 4.1.3, deep consequence components subject to beams present in the lower support structure have been added to the PVNGS multiple degradation mechanisms "Primary" inspection category and are contained in Table 5-1. The examination method is consistent with MRP recommendations for these components.

The coverage confirms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination frequency is on a 10-year interval consistent with the other "Primary" inspection category components.

SE Section 4.1.4, Topical Report In accordance with SE Section 4.1.4, PVNGS will meet the minimum Condition 4: Minimum inspection coverage specified in the SE. The appropriate wording has been examination coverage criteria for added to Table 5-2 examination coverage."expansion" inspection category components SE 4.1.5, Topical Report Condition Not applicable for PVNGS.5: Examination frequencies for baffle former bolts and core shroud bolts V:j-sw hyt frli Assoclata, k=Report No. 1200347.401.R1 5-26 Table 5-5. PVNGS Response to NRC Final Safety Evaluation of MRP-227-A

[4] (continued)

MRP-227 SE Item PVNGS Response SE 4.1.6, Topical Report Condition In accordance with SE Section 4.1.6, Table 5-2 requires a 10-year re-6: Periodicity of the re-examination examination interval for all "Expansion" inspection category components of "expansion" inspection category once degradation is identified in the associated "Primary" inspection category components component and examination of the expansion category component commences.

SE Section 4.1.7, Topical Report This condition applies to update of the industry guidelines.

No plant-specific Condition 7: Updating of industry actions are required.guideline SE Section 4.2.1, The evaluation of design and operating history demonstrating that MRP-227-Applicant/Licensee Action Item 1: A is applicable to PVNGS is contained in Section 1.8.4.1 and Section 5.1 of Applicability of FMECA and this document.Functionality Analysis Assumptions SE Section 4.2.2, The PVNGS review of components within the scope of license renewal was Applicant/Licensee Action Item 2: compared against the information contained in MRP-191. Table 4-5 is PWR Vessel Internals Components provided in Table B-1. The Aging Management Review performed as part of Within the Scope of License the PVNGS LRA is described in Section 1.7.1 of this document and Renewal summarized as part of Applicant/Licensee Action Item 2 in Section 5.2.SE Section 4.2.3, No action required.

Neither ICI thimble tubes nor thermal shields are present Applicant/Licensee Action Item 3: in the PVNGS reactor vessel internals.

Evaluation of the Adequacy of Plant-Specific Existing Programs SE Section 4.2.4, No action required.

This action does not apply to C-E designed units.Applicant/Licensee Action Item 4: B&W Core Support Structure Upper Flange Stress Relief SE Section 4.2.5, PVNGS has a core shroud assembled with full-height shroud plates and hence Applicant/Licensee Action Item 5: no action is required.Application of Physical Measurements as part of I&E Guidelines for B&W, CE and Westinghouse RVI Components R o Nu04w.R-Int grlty Associates, Inc.'Report No. 1200347.401.Rl 5-27 Table 5-5. PVNGS Response to NRC Final Safety Evaluation of MRP-227-A

[4] (continued)

MRP-227 SE Item PVNGS Response SE Section 4.2.6, No action required.

This action does not apply to C-E designed units.Applicant/Licensee Action Item 6: Evaluation of Inaccessible B&W Components SE Section 4.2.7, PVNGS does not have CASS, martensitic stainless steel or precipitation Applicant/Licensee Action Item 7: hardened stainless steel materials in the reactor vessel internals.

The PVNGS Plant Specific Evaluation of CASS Units do not have lower support plates or lower support columns as part of the Materials RVI. However, deep beam supports are considered Primary components and are included for the MRP-227-A augmented inspections (Table 5-1) and summarized as part of Applicant/Licensee Action Item 7 in Section 5.7 of this document.SE Section 4.2.8, The responses to meet A/LAI No. 8 are contained in Section 5.8 of this Applicant/Licensee Action Item 8 document.-VitniotuW Nateg,*t Associates, hmG~Report No. 1200347.401.R5 5-28 Table 5-6. PVNGS Unit 1 Program Enhancement and Implementation Schedule Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 1R17 Spring 2013 N/A Not applicable Not applicable ot applicable 1R18 Fall 2014 N/A Not applicable Not applicable ot applicable 1R19 Spring 2016 /A ASME Section XI ISI Not applicable Not applicable 1R20 Fall 2017 N/A Not applicable Not applicable Not applicable 1 R21 Spring 2019 N/A Not applicable Not applicable Not applicable 1 R22 Fall 2020 N/A Not applicable Not applicable Not applicable 1 R23 Spring 2022 N/A Not applicable Not applicable Not applicable 1 R24 Fall 2023 N/A Not applicable Not applicable Not applicable 1 R25 Spring 2025 N/A Not applicable Not applicable Renewed Operating License begins June 1, 2025 1 R26 Fall 2026 N/A MRP-227-A augmented inspections for core

• MRP-227-A inspections in accordance ot applicable shroud assembly (shroud plates), core support with MRP-228.barrel assembly (upper core support barrel flange weld, lower cylinder girth welds, lower flange weld), lower support structure (core support column weld, deep beams), upper internals assembly (fuel alignment plate),* ASME Section XI 10 Year ISI inspections of

• Inspections in accordance with PVNGS core shroud assembly (guide lugs, guide lug ISI Program inserts and bolts) and core barrel assembly (upper flange)1R27 Spring 2028 N/A Not applicable Not applicable Not applicable lRtf Integrity Associates, 1ncW.Report No. 1200347.401.Rl 5-29 Table 5-6. PVNGS Unit 1 Program Enhancement and Implementation Schedule (continued)

Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 1 R28 Fall 2029 N/A Not applicable Not applicable Not applicable 1 R29 Spring 2031 N/A Not applicable Not applicable Not applicable 1R30 Fall 2032 /A Not applicable Not applicable Not applicable 1 R31 Spring 2034 N/A Not applicable Not applicable Not applicable 1 R321" Fall 2035 N/A

• MRP-227-A augmented inspections for core

• MRP-227-A inspections in accordance Not applicable shroud assembly (shroud plates), core support with MRP-228.barrel assembly (upper core support barrel flange weld, lower cylinder girth welds, lower flange weld), lower support structure (core support column weld, deep beams), upper internals assembly (fuel alignment plate),* ASME Section XI 10 Year lS1 inspections of

• Inspections in accordance with PVNGS core shroud assembly (guide lugs, guide lug ISI Program inserts and bolts) and core barrel assembly (upper flange)1R33 Spring 2037 N/A Not applicable Not applicable Not applicable 1R34 Fall 2038 N/A Not applicable Not applicable Not applicable 1R35 Spring 2040 N/A Not applicable Not applicable Not applicable 1R36 Fall 2041 N/A Not applicable Not applicable Not applicable 1R37 Spring 2043 N/A Not applicable Not applicable Not applicable jjsnCturei

~1tgy Associaes, lnW Report No. 1200347.401.Rl 5-30 Table 5-6. PVNGS Unit 1 Program Enhancement and Implementation Schedule (continued)

Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 1 R38 Fall 2044 N/A 0 MRP-227-A augmented inspections for core 0 MRP-227-A inspections in accordance Renewed Operating License shroud assembly (shroud plates), core support with MRP-228. expires June 1, 2045.barrel assembly (upper core support barrel flange weld, lower cylinder girth welds, lower flange weld), lower support structure (core support column weld, deep beams), upper intemals assembly (fuel alignment plate),* ASME Section X1 10 Year ISI inspections of

• Inspections in accordance with PVNGS core shroud assembly (guide lugs, guide lug IS1 Program inserts and bolts) and core barrel assembly (upper flange)Note: (1) Volumetric inspections would be performed during Refueling Outage 1R32.vjsm"tou lilt*~k Assocates, InW Report No. 1200347.401.Rl 5-31 Table 5-7. PVNGS Unit 2 Program Enhancement and Implementation Schedule Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 2R17 Fall 2012 N/A Not applicable Not applicable Not applicable 2Rl8 Spring 2014 N/A Not applicable Not applicable Not applicable 2RI 9 Fall 2015 N/A Not applicable Not applicable Not applicable 2R20 Spring 2017 N/A Not applicable Not applicable Not applicable 2R21 Fall 2018 N/A Not applicable Not applicable Not applicable 2R22 Spring 2020 N/A Not applicable Not applicable Not applicable 2R23 Fall 2021 N/A Not applicable Not applicable Not applicable 2R24 Spring 2023 N/A Not applicable Not applicable Not applicable 2R25 Fall 2024 N/A Not applicable Not applicable Not applicable 2R26(') Spring 2026 N/A

• MRP-227-A augmented inspections for core S MRP-227-A inspections in accordance Renewed Operating License shroud assembly (shroud plates), core support with MRP-228. begins April 24, 2026 barrel assembly (upper core support barrel flange weld, lower cylinder girth welds, lower flange Contingency exams of relevant weld), lower support structure (core support expansion components if column weld, deep beams), upper internals indications are found in assembly (fuel alignment plate), examination of upper flange weld* ASME Section XI 10 Year ISI inspections of

• Inspections in accordance with PVNGS core shroud assembly (guide lugs, guide lug ISI Program inserts and bolts) and core barrel assembly (upper flange)2R27 Fall 2027 N/A Not applicable Not applicable Not applicable 2R28 Spring 2029 N/A Not applicable Not applicable Not applicable 2m. ~ Vrily Associates, Inc.Report No. 1200347.401.Rl 5-32 Table 5-7. PVNGS Unit 2 Program Enhancement and Implementation Schedule (continued)

Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 2R29 Fall 2030 N/A Not applicable Not applicable Not applicable 2R30 Spring 2032 N/A Not applicable Not applicable Not applicable 2R31 Fall 2033 N/A Not applicable ot applicable Not applicable 2R32(') Spring 2035 N/A MRP-227-A augmented inspections for core 0 MRP-227-A inspections in accordance Contingency exams of relevant shroud assembly (shroud plates), core support with MRP-228. expansion components if barrel assembly (upper core support barrel flange indications are found in weld, lower cylinder girth welds, lower flange examination of upper flange weld weld), lower support structure (core support column weld, deep beams), upper internals assembly (fuel alignment plate), ASME Section XI 10 Year ISI inspections of

• Inspections in accordance with PVNGS core shroud assembly (guide lugs, guide lug IS1 Program inserts and bolts) and core barrel assembly (upper flange)2R33 Fall 2036 N/A Not applicable Not applicable Not applicable 2R34 Spring 2038 N/A Not applicable Not applicable Not applicable 2R35 Fall 2039 N/A Not applicable Not applicable Not applicable 2R36 Spring 2041 N/A Not applicable Not applicable Not applicable 2R37 Fall 2042 N/A Not applicable Not applicable Not applicable C~t1OWUrEiff 1dy Assaclats, IncO Report No. 1200347.401.Rl 5-33 Table 5-7. PVNGS Unit 2 Program Enhancement and Implementation Schedule (continued)

Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 2R38t 2) Spring 2044 IA 0 MRP-227-A augmented inspections for core 0 MRP-227-A inspections in accordance Contingency exams of relevant shroud assembly (shroud plates), core support with MRP-228. expansion components if barrel assembly (upper core support barrel flange indications are found in weld, lower cylinder girth welds, lower flange examination of upper flange weld weld), lower support structure (core support column weld, deep beams), upper internals assembly (fuel alignment plate), 0 ASME Section Xl 10 Year ISI inspections of

• Inspections in accordance with PVNGS core shroud assembly (guide lugs, guide lug ISI Program inserts and bolts) and core barrel assembly (upper flange)2R39 Fall 2045 N/A Not applicable Not applicable Renewed Operating License expires April 24, 2046 Note: (1) Relief Request #40 allows the Volumetric and Visual inspections to be moved to Refueling Outage 2R26 (fall), that would be the preferred outage. The downstream outages would then change from Refueling Outage 2R32 to 2R33.(2) Volumetric inspections would be performed during Refueling Outage 2R32.jjSbcalmu" ~1tuy Assac~ats, 1Wc Report No. 1200347.401.Rl 5-34 Table 5-8. PVNGS Unit 3 Program Enhancement and Implementation Schedule Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 3R16 Spring 2012 N/A Not applicable Not applicable ot applicable 3R17 Fall 2013 N/A Not applicable Not applicable ot applicable 3R18 Spring 2015 N/A Not applicable Not applicable Not applicable 3R19 Fall 2016 N/A Not applicable Not applicable Not applicable 3R20 Spring 2018 N/A Not applicable Not applicable Not applicable 3R21 Fall 2019 N/A Not applicable Not applicable Not applicable 3R22 Spring 2021 N/A Not applicable Not applicable Not applicable 3R23 Fall 2022 N/A Not applicable Not applicable Not applicable 3R24 Spring 2024 N/A Not applicable Not applicable Not applicable 3R25 Fall 2025 N/A Not applicable Not applicable Not applicable 3R26 Spring 2027 N/A Not applicable Not applicable Renewed Operating License begins November 25, 2027!j~snCftxmwh11sr*

Associats, Wnc Report No. 1200347.401.Rl 5-35 Table 5-8. PVNGS Unit 3 Program Enhancement and Implementation Schedule (continued)

Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 3R27 Fall 2028 N/A

• MRP-227-A augmented inspections for Core 0 MRP-227-A inspections in accordance Contingency exams of relevant shroud assembly (shroud plates), Core support with MRP-228. expansion components if barrel assembly (upper core support barrel flange indications are found in weld, lower cylinder girth welds, lower flange examination of upper flange weld weld), Lower support structure (core support column weld, deep beams), Upper internals assembly (fuel alignment plate),* ASME Section X1 10 Year ISI inspections of 0 Inspections in accordance with PVNGS Core shroud assembly (Guide lugs, guide lug ISI Program inserts and bolts) and Core barrel assembly (upper flange)3R28 Spring 2030 N/A Not applicable Not applicable Not applicable 3R29 Fall 2031 N/A Not applicable Not applicable Not applicable 3R30 Spring 2033 N/A Not applicable Not applicable Not applicable 3R31 Fall 2034 N/A Not applicable Not applicable Not applicable 3R32 Spring 2036 N/A Not applicable Not applicable Not applicable 3R33 Fall 2037 N/A 0 MRP-227-A augmented inspections for Core

• MRP-227-A inspections in accordance Contingency exams of relevant shroud assembly (shroud plates), Core support with MRP-228. expansion components if barrel assembly (upper core support barrel flange indications are found in weld, lower cylinder girth welds, lower flange examination of upper flange weld weld), Lower support structure (core support column weld, deep beams), Upper internals assembly (fuel alignment plate),* ASME Section XI 10 Year IS] inspections of 0 Inspections in accordance with PVNGS Core shroud assembly (Guide lugs, guide lug ISI Program inserts and bolts) and Core barrel assembly (upper flange)V &8tWture ~1tgy AssoA6tS, /Wc Report No. 1200347.401.R1 5-36 Table 5-8. PVNGS Unit 3 Program Enhancement and Implementation Schedule (continued)

Refueling Cycle End Estimated AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year EFPY 3R34 Spring 2039 N/A Not applicable Not applicable ot applicable 3R35 Fall 2040 N/A- Not applicable Not applicable ot applicable 3R36 Spring 2042 N/A Not applicable Not applicable Not applicable 3R37 Fall 2043 N/A Not applicable Not applicable Not applicable 3R38 Spring 2045 N/A Not applicable Not applicable.

Not applicable 3R39(') Fall 2046 N/A a MRP-227-A augmented inspections for Core 0 MRP-227-A inspections in accordance Renewed Operating License shroud assembly (shroud plates), Core support with MRP-228. expires November 25, 2047 barrel assembly (upper core support barrel flange weld, lower cylinder girth welds, lower flange Contingency exams of relevant weld), Lower support structure (core support expansion components if column weld, deep beams), Upper internals indications are found in assembly (fuel alignment plate), indication ar e f land in examination of upper flange weld ASME Section XI 10 Year ISI inspections of

• Inspections in accordance with PVNGS Core shroud assembly (Guide lugs, guide lug ISI Program inserts and bolts) and Core barrel assembly (upper flange)Note: (1) Volumetric inspections would be performed during Refueling Outage 3R33 V a~sn"tr ~1yg~j Assock~es, InW Report No. 1200347.401.Rl 5-37 Table 5-9. Summary of Actions Related to Aging Management of RVI for PVNGS Units 1, 2, and 3 Item PVNGS Action Program/Action Description No.1 PVNGS Units 1, 2, and 3 Update to the Reactor Vessel Internals Inspection Program Commitments Submit the PVNGS Units 1, 2, and 3 reactor vessel internals aging management program and inspection plans in accordance with MRP-227-A no later than October 1, 2012, for NRC review and approval.2 APS will incorporate a summary description of the PVNGS Units 1, 2, and 3 reactor vessel internals Reactor Vessel Internals Aging Management Program aging management program into the Updated Final Safety Analysis Report (UFSAR) no later than the next scheduled update required by 10 CFR 50.71(e) following NRC approval of the program. This summary description shall reference the "PWR Internals Aging Management Program Plan for Palo Verde Nuclear Generating Station Units 1, 2, and 3." 3 PVNGS will review plant specific and fleet operating experience based on updates to Appendix A of Reactor Vessel Internals Aging Management Program MRP-227-A and make updates to the RVI AMP as necessary.

4 Participation in Industry Groups (e.g. PWR Owners Group Materials Subcommittee, EPRI MRP Reactor Vessel Internals Aging Management Program 5 APS Personnel Responsibilities:

Ensure department specific actions are performed as it* PVNGS Engineering Programs Department:

relates to the aging management of Reactor Vessel* Program Engineering Department Internals* Chemistry Department 6 Plant Specific Programs:

The RVI AMP takes credit for plant specific* ASME Section XI Inservice Inspection Program programs.

RVI AMP items credited from plant* PVNGS Water Chemistry Program specific programs (e.g. ASME Section XI ISI program* RCS Transient and Operating Cycles Procedure components credited as part of MRP-227-A inspections)

R NR5sn t I~ 1urity Assochies, Inc Report No. 1200347.401.Rl 5-38 Table 5-9. Summary of Actions Related to Aging Management of RVI for PVNGS Units 1, 2, and 3 (continued)

Item PVNGS Action Program/Action Description No.7 Examinations specified in the MRP-227-A guidelines shall be conducted in accordance with Reactor Vessel Internals Aging Management Inspection Standard MRP-228. Prograrn/ASME Section XI ISI Program 8 Examination results that do not meet the examination acceptance criteria defined in Section 5 of Nuclear Administrative Technical Manual, 90DP-MRP-227-A guidelines shall be recorded and entered in the plant corrective action program and O0P 10, Revision 48, "Condition Reporting." dispositioned.

9 Each commercial U.S. PWR unit shall provide a summary report of all inspections and monitoring, Reactor Vessel Internals Aging Management Program items requiring evaluation, and new repairs to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals within the scope of MRP-227-A are examined 10 If an engineering evaluation is used to disposition an examination result that does not meet the PVNGS will comply with this requirement by using examination acceptance criteria in Section 5 of MRP-227-A, this engineering evaluation shall be NRC-approved evaluation methodology (e.g. WCAP-conducted in accordance with an NRC-approved evaluation methodology.

17096)11 Inspection acceptance and expansion criteria will be reviewed whenever new versions of the NRC Reactor Vessel Internals Aging Management Program approved versions of MRP-227 and WCAP- 17096 are published, as the industry continues to develop and refine the information.

Relevant changes based on the review of these NRC approved documents will be included as updates to the RVI AMP.V auhclawe ~1try Assaclates, inW Report No. 1200347.401.R1 5-39

6.0 REFERENCES

1. Nuclear Administrative and Technical, "PVNGS Integrated Materials Management Program," Document No. 81DP-9RC03, Revision 3 (SI File No. 1200347.234).

2. Nuclear Energy Institute, "Revision 2 to NEI 03-08, Guideline for the Management of Materials Issues," dated January, 2010 (SI File No. 1200347.212).

3. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), EPRI, Palo Alto, CA: 2011. 1022863 (SI File No.1200347.214P).

4. Letter from Robert A. Nelson (NRC) to Neil Wilmshurst (EPRI) dated December 16, 2011, "Revision 1 to the Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR)Internals Inspection and Evaluation Guidelines" (TAC No. ME0680)," NRC ADAMS Accession No. ML1 1308A770 (SI File No. 1200347.205).

5. "License Renewal Application, Palo Verde Nuclear Generating Station Unit 1, Unit 2 and Unit 3," Facility Operating License Nos. NPF-41, NPF-51, and NPF-74, Supplement 1, April 10, 2009 (SI File No. 1200347.208).

6. Letter from APS to U.S. NRC Dated May 13, 2010, "Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3: Docket Nos. SN 50-528, 50-529, and 50-530, Revision to Commitment Completion Date Associated with Power Uprate, PVNGS Units 1 and 3," ADAMS Accession No. ML101410260 (SI File No. 1200347.235).

7. Letter from APS to US NRC Dated October 11, 2011, "Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, Docket Nos. STN 50-528, 50-529 and 50-530: Update to the Reactor Vessel Internals Inspection Program Commitments," Letter No.102-06-423-DCM/GAM, NRC ADAMS Accession No. MLl 1297A1 18, (SI File No.1200347.204).

8. Nuclear Administrative and Technical Manual, "Condition Reporting," Document No.90DP-OIP 10, Revision 48 (SI File No. 1200347.202).

9. Letter from APS to US NRC Dated July 9, 2004, "Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3, Docket Nos. STN 50-528, 50.529 and 50-530: Request for a enort No. 1200347.401.R1 6-1 MIh /n R F- -] -...........

License Amendment to Support Replacement of Steam Generators and Uprated Power Operations in Units 1 and 3, and Associated Administrative Changes for Unit 2," Letter No. 102-05116-CDM/TNW/RAB, NRC ADAMS Accession No. ML042010289 (SI File No. 1200347.206).

10. NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," Rev. 2, U. S. Nuclear Regulatory Commission, December 2010 (SI File No. 1200347.222).

11. NUREG- 1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, September 2005.12. Combustion Engineering Owners Group Report, CE NPSD-1216, Revision 0, "Generic Aging Management Review Report for the Reactor Vessel Internals," CEOG Task 1185, March 2001 (SI File No. 1200347.207).

13. Materials Reliability Program: "Screening Categorization and Ranking of Reactor Internals Components of Westinghouse and Combustion Engineering PWR Design (MRP-191)," Electric Power Research Institute, Palo Alto, CA: 2007. 1013234 (SI File No. 1200347.218).

14. EPRI Report MRP-228, "Materials Reliability Program Inspection Standard for Reactor Internals," Latest Revision (SI File No. 1200347.219P).

EPRI Proprietary Information

15. Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values (MRP-175).

EPRI, Palo Alto, CA: 2005. 1012081 (SI File No. 1200347.228).

16. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev.

0), EPRI, Palo Alto, CA: 2008. 1016596 (SI File No. 1200347.213).

17. WCAP-17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009 (SI File No. 1200347.215).

18. 3 rP Inspection Interval:

Inservice Inspection Program Summary Manual, PVNGS Unit 1, Program No: 3TNT-ISI-1, Rev. 1 (SI File No. 1200347.201).

19. 3 rd Inspection Interval:

Inservice Inspection Program Summary Manual, PVNGS Unit 2, Program No: 3INT-ISI-2, Rev. 2 (SI File No. 1200347.201).

Report No. 1200347.401.R1 6-2 hf1* Assocbs, krc

20. 3d Inspection Interval:

Inservice Inspection Program Summary Manual, PVNGS Unit 2, Program No: 3INT-ISI-3, Rev. 1 (SI File No. 1200347.201).

21. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 2001 Edition through 2003 Addenda.22. PVNGS Nuclear Administrative and Technical Manual, "Outage Planning and Implementation," Document No 51DP-90M09, Rev. 14 (SI File No. 1200347.203).

23. Nuclear Administrative and Technical Manual, "System Chemistry Specifications," Document No. 74DP-9CY04, Revision 72 (SI File No. 1200347.238).

24. "Pressurized Water Reactor Primary Water Chemistry Guidelines," Volumes 1 and 2, Revision 6, Electric Power Research Institute, Palo Alto, CA: 2007, 1014986.25. Letter from US NRC to Arizona Nuclear Power Project"Dated October 21, 1987,"Issuance of Amendment No. 24 to Facility Operating License No. NPF-41 for the Palo Verde Nuclear Generating Station Unit No. 1 (TAC Nos. 65460, 65461, 65462 and 65691 through 65706)," ADAMS Accession No. ML021690079 (SI File No.1200347.231).

26. Letter from APS to US NRC Dated June 3, 2005, "Palo Verde Nuclear Generating Station (PVNGS) Units 1 and 3, Docket Nos. STN 50-528 and STN 50-530 Response to Request for Additional Information Regarding Steam Generator Replacement and Power Uprate License Amendment Request," ADAMS Accession No. ML051660184 (SI File No. 1200347.231).

27. Letter from US NRC to Arizona Nuclear Power Project Dated June 9, 1989, "Issuance of Amendment No. 18 to Facility Operating License No. NPF-74 for the Palo Verde Nuclear Generating Station, Unit 3 (TAC No. 71574)," ADAMS Accession No.ML0223 80393 (SI File No. 1200347.231)

28. Letter from US NRC to APS Dated June 2, 2010, "Request for Additional Information for the Review of the Palo Verde Nuclear Generating Station, Units 1, 2, and 3 License Renewal Application (TAC Nos. ME0254, MC0255, and ME0256)," ADAMS Accession No. ML10 1340100 (SI File No. 1200347.232)

Report No. 1200347.401.Rl 6-3 v aniwk

29. Nuclear Administrative and Technical Manual, "PVNGS RCS Transient and Operating Cycles Procedures," Document No. 73ST-9RC02, Latest Revision (SI File No.1200347.235).

30. PVNGS Updated Final Safety Analysis Report, Revision 16, June 2011 (SI File No.1200347.210).

31. "Safety Evaluation Report Related to the License Renewal of Palo Verde Nuclear Generating Station, Units 1, 2, and 3," NUREG-1961 (SI File No. 1200347.209).

32. U.S. Code of Federal Regulations, Title 10, "Energy," Part 50, "Domestic Licensing of Production and Utilization Facilities," 50.55a, "Codes and Standards." 33. NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," Rev. 1, U. S. Nuclear Regulatory Commission, September 2005.34. U.S. Code of Federal Regulations, Title 10, "Energy," Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants." 35. Nuclear Administrative and Technical Manual, "Industry Operating Experience Review," Document No. 65DP-OQQ01, Revision 29 (SI File No. 1200347.239).

36. Westinghouse Report, WCAP-17435-NP, Rev. 0, "Results of the Reactor Internals Operating Experience Survey Conducted under PWROG Project Authorization PA-MSC-0568," September 2011.37. Letter from APS to USNRC Dated March 1, 2010, "Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529 and 50-530 Response to January 28, 2010, Request for Additional Information (RAI) Regarding Reactor Internals, Half-Nozzle Repairs, Metal Fatigue, and Copper Alloy Piping, and Supplement to Balance of Plant RAI response, for the Review of the PVNGS License Renewal Application," ADAMS Accession No. ML100680517 (SI File No.1200347.240).

38. "Arizona Nuclear Power Project Palo Verde Unit 2 Reactor Vessel Internals Instruction Manual," Document No. IG-14373-RCE-400, Rev. 4 (SI File No. 120047.233).

39. ASME Boiler and Pressure Vessel Code,Section III, 1974 Edition.Report No. 1200347.401.R1 6-4 hft i

40. SECY-01-0065 Dated April 20, 2001, "Weekly Information Report -Week Ending April 13, 2001." (SI File No. 1200347.227).

41. Combustion Engineering Stress Report No. 14273-MD-001, "Evaluation of Reactor Core Support Structures," Date: 3/31/81 (SI File No. 1200347.217P).

C-E Proprietary Information.

42. Nuclear Administrative and Technical Manual, "Visual Examination of Reactor Vessel Internals," Document No. 73TI-9ZZ20, Revision 9 (SI File No. 1200347.229)

43. Letter from US NRC to APS dated February 22, 2010, "Palo Verde Nuclear Generating Station, Units 1, 2, and 3 -Relief Request No. RR-40, Reactor Vessel Weld Examination Interval Extension (TAC Nos. ME1634, ME1635, and ME1636)," ADAMS Accession No. ML 10029415 (SI File No. 1200347.236).

44. Email from Ed Fernandez (APS) to Tim Griesbach (SI) dated August 16, 2012, "

Subject:

FW: MRP-227 feedback summary," (SI File No. 1200347.237).

Report No. 1200347.401 .R1 6-5 V &ults n/

APPENDIX A SECTION XI 10 YEAR ISI EXAMINATIONS OF B-N-2 AND B-N-3 INTERNALS COMPONENTS FOR PVNGS [421 Report No. 1200347.401.R1 A-1 kjj*Ihl u ASSOatUS, In.'

Table A-1.Section XI 10 Year ISI Examinations of B-N-2 and B-N-3 Internals Components for PVNGS Code Exam Fabrication ID Examination Area Description Extent of Exam Category Method Lower Internals

-Exterior (Core Barrel)Closure Head Alignment Keys @ 0' B-N-3 VT-3 All accessible surfaces Closure Head Alignment Keys @ 90' B-N-3 VT-3 All accessible surfaces Closure Head Alignment Keys @ 1800 B-N-3 VT-3 All accessible surfaces Closure Head Alignment Keys @ 270' B-N-3 VT-3 All accessible surfaces Core Shroud and Lower Support Structure Top surface of shroud Alignment keys @ 0* B-N-3 VT-3 All accessible surfaces Top surface of shroud Alignment keys @ 90' B-N-3 VT-3 All accessible surfaces Top surface of shroud Alignment keys @ 1800 B-N-3 VT-3 All accessible surfaces Top surface of shroud Alignment keys @ 2700 B-N-3 VT-3 All accessible surfaces Periphery section of shroud, fuel alignment, and insert pins, support beams, welds and instrument B-N-3 VT-3 All accessible surfaces nozzles @ 0'Periphery section of shroud, fuel alignment, and insert pins, support beams, welds and instrument B-N-3 VT-3 All accessible surfaces nozzles @ 900 Periphery section of shroud, fuel alignment, and insert pins, support beams, welds and instrument B-N-3 VT-3 All accessible surfaces nozzles @ 1800 Periphery section of shroud, fuel alignment, and insert pins, support beams, welds and instrument B-N-3 VT-3 All accessible surfaces nozzles @ 270'Guide lugs, inserts, cap screw, dowel pins and B-N-3 VT-3 All accessible surfaces welds (@ 00, 900, 1800, and 2700)vItstWNc&W kf hd AssaafatUS, IWc Report No. 1200347.401.R1 A-2 Table A-1.Section XI 10 Year ISI Examinations of B-N-2 and B-N-3 Internals Components for PVNGS (continued)

Code Exam Fabrication ID Examination Area Description C M Extent of Exam Category JMethod Core Barrel Core barrel flange B-N-3 VT-3 All accessible surfaces Snubber lugs (@ 00, 600, 1200, 180', 240', and B-N-3 VT-3 All accessible bolts 270-)Flexture weld B-N-3 VT-3 All accessible surfaces Nozzle areas B-N-3 VT-3 All accessible surfaces Support flange keyways (@ 00, 900, 180', and B-N-3 VT-3 All accessible surfaces 2700)Upper Guide Structure

-Exterior Support flange keyways (@ 00, 400, 1800, and B-N-3 VT-3 All accessible surfaces 2700)Hold down ring clips (@ 80, 980, 1880, and 2780) B-N-3 VT-3 All accessible bolts Fuel alignment plate keyways (@ 0', 900, 180', and B-N-3 VT-3 All accessible surfaces 2700)Tie rod assemblies (8) (including nuts, locking B-N-3 VT-3 All accessible surfaces straps and welds)Top Hat to Support flange bolts B-N-3 VT-3 All accessible surfaces Shroud assembly B-N-3 VT-3 All accessible surfaces Reactor Vessel -Interior Core stabilizing lugs (include shims, lock pins, B-N-2 VT-3 All accessible surfaces bolts, welds) @ 00 Core stabilizing lugs (include shims, lock pins, B-N-2 VT-3 All accessible surfaces bolts, welds) @ 600 Report No. 1200347.40 1.R1 A-3 vanr &bdy* AssOGIates, 1Wc Table A-1.Section XI 10 Year ISI Examinations of B-N-2 and B-N-3 Internals Components for PVNGS (continued)

Code Exam Fabrication ID Examination Area Description Extent of Exam Category Method Core stabilizing lugs (include shims, lock pins, B-N-2 VT-3 All accessible surfaces bolts, welds) @ 1200 Core stabilizing lugs (include shims, lock pins, B-N-2 VT-3 All accessible surfaces bolts, welds) @ 1800 Core stabilizing lugs (include shims, lock pins, B-N-2 VT-3 All accessible surfaces bolts, welds)@ 2400 Core stabilizing lugs (include shims, lock pins, B-N-2 VT-3 All accessible surfaces bolts, welds)@ 3000 Core stop lugs (include top surface, welds, and B-N-2 VT-3 All accessible surfaces surrounding reactor cladding)

@ 16'Core stop lugs (include top surface, welds, and B-N-2 VT-3 All accessible surfaces surrounding reactor cladding)

@ 76'Core stop lugs (include top surface, welds, and B-N-2 VT-3 All accessible surfaces surrounding reactor cladding)

@ 1360 Core stop lugs (include top surface, welds, and B-N-2 VT-3 All accessible surfaces surrounding reactor cladding)

@ 1960 Core stop lugs (include top surface, welds, and B-N-2 VT-3 All accessible surfaces surrounding reactor cladding)

@ 2560 Core stop lugs (include top surface, welds, and B-N-2 VT-3 All accessible surfaces surrounding reactor cladding)

@ 3160 Flow baffle weld attachments

(@ 20', 60', 1000, B-N-2 VT-3 All accessible surfaces 1400, 1800, 2200, 2600, 3000, and 3400)Baffle segment welds (3) B-N-2 VT-3 All accessible surfaces Flange assembly to baffle welds (2) B-N-2 VT-3 All accessible surfaces Flange segment welds (3/flange)

B-N-2 VT-3 All accessible surfaces Incore Instrument Nozzles (including welds and B-N-2 VT-3 All accessible surfaces surrounding reactor cladding)Surveillance Capsule Holder Assemblies

@ 380 B-N-2 VT-3 All accessible surfaces Report No. 1200347.401.R1 A-4~§SUcn!~wgre kkq Assadfiteu, 1Wc Table A-1.Section XI 10 Year ISI Examinations of B-N-2 and B-N-3 Internals Components for PVNGS (continued)

Code Exam Fabrication ID Examination Area Description Extent of Exam Category Method Surveillance Capsule Holder Assemblies

@ 430 B-N-2 VT-3 All accessible surfaces Surveillance Capsule Holder Assemblies

@ 137' B-N-2 VT-3 All accessible surfaces Surveillance Capsule Holder Assemblies

@ 142' B-N-2 VT-3 All accessible surfaces Surveillance Capsule Holder Assemblies

@ 230' B-N-2 VT-3 All accessible surfaces Surveillance Capsule Holder Assemblies

@ 3100 B-N-2 VT-3 All accessible surfaces Report No. 1200347.401.R1 A-5 Usfrch~we bitugi AssaWIte/ncIW APPENDIX B AGING MANAGEMENT REVIEW PVNGS UNITS 1, 2, AND 3 [51 Report No. 1200347.401.R1 B-1 Iibhw flt

• Assocites, Inc" Table B-1. Summary of Aging Management Evaluations for PVNGS Reactor Vessel Internals

[5]Component Type Intended Material Environment Aging Effect Aging Management Program NUREG- Table 3.1.1 Notes Function Requiring 1801 Vol. 2 (of LRA)Management

[11] Item Item RV ICI Guide Tube Pressure Stainless Steel Borated Water None None IV.E-3 3.1.1.86 A Boundary Leakage (Int)RV ICI Guide Tube Pressure Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.A2-1 3.1.1.23 E Boundary (Int) 19.1.2) and ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD (UFSAR 19.1.1)RV ICI Guide Tube Pressure Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.A2-14 3.1.1.83 A Boundary (Int) 19.1.2)RV Core Stop Lug and Structural Nickel Alloys Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-23 3.1.1.37 C Surv Capsule Holder Support (Ext) 19.1.2) and Reactor Coolant System Supplement (UFSAR 19.1.21)RV Core Stop Lug and Structural Nickel Alloys Reactor Coolant Loss of Material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Surv Capsule Holder Support (Ext) 19.1.2)RVI Core Support Structural Stainless Steel Reactor Coolant Cumulative Time-Limited Aging Analysis IV.B3.24 3.1.1.05 A Structure Support (Ext) fatigue damage evaluated for the period of extended operation RVI CSS Core Shroud Direct Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-11 3.1.1.30 A Assembly Flow, (Ext) 19.1.2) and Reactor Coolant Structural System Supplement (UFSAR Support 19.1.21)RVI CSS Core Shroud Direct Stainless Steel Reactor Coolant Loss of fracture Reactor Coolant System IV.B3-12 3.1.1.22 A Assembly Flow, (Ext) toughness Supplement (UFSAR 19.1.21)Structural Support Report No. 1200347.401.R1 B-2~jsnCOWui

~1gdy Associaes, hInO Table B-1. Summary of Aging Management Evaluations for PVNGS Reactor Vessel Internals 15] (continued)

Component Type Intended Material Environment.

Aging Effect Aging Management Program NUREG- Table 3.1.1 Notes Function Requiring 1801 Vol. 2 (of LRA)Management

[111 Item Item RVI CSS Core Shroud Direct Flow, Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-13 3.1.1.33 A Assembly Structural (Ext) dimensions Supplement (UFSAR 19.1.21)Support RVI CSS Core Shroud Direct Flow, Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Assembly Structural (Ext) 19.1.2)Support RVI CSS Core Support Direct Flow, Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-14 3.1.1.33 A Barrel Assembly Structural (Ext) dimensions Supplement (UFSAR 19.1.21)Support RVI CSS Core Support Direct Flow, Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-15 3.1.1.30 A Barrel Assembly Structural (Ext) 19.1.2) and Reactor Coolant Support System Supplement (UFSAR 19.1.21)RVI CSS Core Support Direct Flow, Stainless Steel Reactor Coolant Loss of fracture Reactor Coolant System IV.B3-16 3.1.1.22 A Barrel Assembly Structural (Ext) toughness Supplement (UFSAR 19.1.21)Support RVI CSS Core Support Direct Flow, Stainless Steel Reactor Coolant Loss of material ASME Section XI Inservice IV.B3-17 3.1.1.63 A Barrel Assembly Structural (Ext) Inspection, Subsections IWB, Support IWC, and IWD (UFSAR 19.1.1)RVI CSS Core Support Direct Flow, Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Barrel Assembly Structural (Ext) 19.1.2)Support RVI CSS Core Support Structural Nickel Alloys Reactor Coolant Changes in Reactor Coolant System IV.B3-19 3.1.1.33 A Barrel Snubber Support (Ext) dimensions Supplement (UFSAR 19.1.21)Assembly RVI CSS Core Support Structural Nickel Alloys Reactor Coolant Loss of fracture Reactor Coolant System IV.B3-20 3.1.1.22 A Barrel Snubber Support (Ext) toughness Supplement (UFSAR 19.1.21)Assembly Report No. 1200347.401.R1 B-3!VjShwtwu

~1fgdy Assaciaes, Inc?

Table B-1. Summary of Aging Management Evaluations for PVNGS Reactor Vessel Internals

[5] (continued)

Component Type Intended Material Environment Aging Effect Aging Management Program NUREG- Table 3.1.1 Notes Function Requiring 1801 Vol. 2 (of LRA)Management

[11] Item Item RVI CSS Core Support Structural Nickel Alloys Reactor Coolant Loss of material ASME Section XI Inservice IV.B3-22 3.1.1.63 A Barrel Snubber Support (Ext) Inspection, Subsections IWB, Assembly IWC, and IWD (UFSAR 19.1.1)RVI CSS Core Support Structural Nickel Alloys Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-23 3.1.1.37 C Barrel Snubber Support (Ext) 19.1.2) and Reactor Coolant Assembly System Supplement (UFSAR 19.1.21)RVI CSS Core Support Structural Nickel Alloys Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Barrel Snubber Support (Ext) 19.1.2)Assembly RVI CSS Core Support Structural Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-19 3.1.1.33 A Barrel Snubber Support (Ext) dimensions Supplement (UFSAR 19.1.21)Assembly RVI CSS Core Support Structural Stainless Steel Reactor Coolant Loss of fracture Reactor Coolant System IV.B3-20 3.1.1.22 A Barrel Snubber Support (Ext) toughness Supplement (UFSAR 19.1.21)Assembly RVI CSS Core Support Structural Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-21 3.1.1.30 A Barrel Snubber Support (Ext) 19.1.2) and Reactor Coolant Assembly System Supplement (UFSAR 19.1.21)RVI CSS Core Support Structural Stainless Steel Reactor Coolant Loss of material ASME Section XI Inservice IV.B3-22 3.1.1.63 A Barrel Snubber Support (Ext) Inspection, Subsections IWB, Assembly IWC, and IWD (UFSAR 19.1.1)RVI CSS Core Support Structural Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Barrel Snubber Support (Ext) 19.1.2)Assembly RVI CSS Lower Support Direct Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-19 3.1.1.33 A Structure Assembly Flow, (Ext) dimension Supplement (UFSAR 19.1.21)Structural Support RVI CSS Lower Support Direct Stainless Steel Reactor Coolant Loss of fracture Reactor Coolant System IV.B3-20 3.1.1.22 A Structure Assembly Flow, (Ext) toughness Supplement (UFSAR 19.1.21)Structural Support Report No. 1200347.40 1.R1 B-4 C OWIJ " rIIffu1~y A&sOciaes, hic 6 Table B-1. Summary of Aging Management Evaluations for PVNGS Reactor Vessel Internals

[51 (continued)

Component Type Intended Material Environment Aging Effect Aging Management Program NUREG- Table 3.1.1 Notes Function Requiring 1801 Vol. 2 (of LRA)Management

[11] Item Item RVI CSS Lower Support Direct Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-21 3.1.1.30 A Structure Assembly Flow, (Ext) 19.1.2) and Reactor Coolant Structural System Supplement (UFSAR Support 19.1.21)RVI CSS Lower Support Direct Stainless Steel Reactor Coolant Loss of material ASME Section XI InservicelV.B3-22 3.1.1.63 A Structure Assembly Flow, (Ext) Inspection, Subsections IWB, Structural IWC, and IWD (UFSAR 19.1.1)Support I I RVI CSS Lower Support Direct Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Structure Assembly Flow, (Ext) 19.1.2)Structural Support RVI Flow Skirt Direct Flow Nickel Alloys Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-23 3.1.1.37 C (Ext) 19.1.2) and Reactor Coolant System Supplement (UFSAR 19.1.21)RVI Flow Skirt Direct Flow Nickel Alloys Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A (Ext) 19.1.2)RVI ICI Support Structural Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-19 3.1.1.33 A Structures Support (Ext) dimension Supplement (UFSAR 19.1.21)RVI ICI Support Structural Stainless Steel Reactor Coolant Loss of fracture Reactor Coolant System IV.B3-20 3.1.1.22 A Structures Support (Ext) toughness Supplement (UFSAR 19.1.21)RVI ICI Support Structural Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-21 3.1.1.30 A Structures Support (Ext) 19.1.2) and Reactor Coolant System Supplement (UFSAR 19.1.21)RVI ICI Support Structural Stainless Steel Reactor Coolant Loss of material ASME Section XI Inservice IV.B3-22 3.1.1.63 A Structures Support (Ext) Inspection, Subsections IWB, IWC, and IWD (UFSAR 19.1.1)RVI ICI Support Structural Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Structures Support (Ext) 19.1.2)Report No. 1200347.401.R1 B-5 C~SijturuIW In~ 1dy AssociAtes, kIm?

Table B-1. Summary of Aging Management Evaluations for PVNGS Reactor Vessel Internals

[51 (continued)

Component Type Intended Material" Environment Aging Effect Aging Management Program NUREG- Table 3.1.1 Notes Function Requiring 1801 Vol. 2 (of LRA)Management

[11] Item Item RVI UGS CEA Shroud Structural Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-2 3.1.1.30 A Assembly Support (Ext) 19.1.2) and Reactor Coolant System Supplement (UFSAR 19.1.21)RVI UGS CEA Shroud Structural Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-4 3.1.1.33 A Assembly Support (Ext) dimension Supplement (UFSAR 19.1.21)RVI UGS CEA Shroud Structural Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-5 3.1.1.37 A.Assembly Support (Ext) 19.1.2) and Reactor Coolant System Supplement (UFSAR 19.1.21)RVI UGS CEA Shroud Structural Stainless Steel Reactor Coolant Loss of preload Reactor Coolant System IV.B3-6 3.1.1.27 A Assembly Support (Ext) Supplement (UFSAR 19.1.21)RVI UGS CEA Shroud Structural Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Assembly Support (Ext) 19.1.2)RVI UGS Holddown Structural Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Ring Support (Ext) 19.1.2)RVI UGS Holddown Structural Stainless Steel Reactor Coolant Loss of material ASME Section XI Inservice IV.B3-26 3.1.1.63 A Ring Support (Ext) Inspection, Subsections IWB, IWC, and IWD (UFSAR 19.1.1)RVI UGS Holddown Structural Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-27 3.1.1.33 A Ring Support (Ext) dimensions Supplement (UFSAR 19.1.21)RVI UGS Holddown Structural Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-28 3.1.1.30 A Ring Support (Ext) 19.1.2) and Reactor Coolant System Supplement (UFSAR 19.1.21)RVI UGS Support Structural Stainless Steel Reactor Coolant Loss of material Water Chemistry (UFSAR IV.B3-25 3.1.1.83 A Barrel Assembly Support (Ext) 19.1.2)RVI UGS Support Structural Stainless Steel Reactor Coolant Loss of material ASME Section XI Inservice IV.B3-26 3.1.1.63 A Barrel Assembly Support (Ext) Inspection, Subsections IWB, I_ 1IWC, and IWD (UFSAR 19.1.1)Report No. 1200347.401.Rl B-6 R4Sfrcwu Iftr y Assoate, Inc?

Table B-1. Summary of Aging Management Evaluations for PVNGS Reactor Vessel Internals

[51 (continued)

Component Type Intended Material Environment Aging Effect Aging Management Program NUREG- Table 3.1.1 Notes Function Requiring 1801 Vol. 2 (of LRA)Management

[11] Item Item RVI UGS Support Structural Stainless Steel Reactor Coolant Changes in Reactor Coolant System IV.B3-27 3.1.1.33 A Barrel Assembly Support (Ext) dimensions Supplement (UFSAR 19.1.21)RVI UGS Support Structural Stainless Steel Reactor Coolant Cracking Water Chemistry (UFSAR IV.B3-28 3.1.1.30 A Barrel Assembly Support (Ext) 19.1.2) and Reactor Coolant System Supplement (UFSAR 19.1.21)RVI Upper Guide Structural Stainless Steel Reactor Coolant Cumulative Time-Limited Aging AnalysislV.B3-24 3.1,1.05 A Structure Assembly Support (Ext) fatigue damage evaluated for the period of extended operation Notes: A. Consistent withNUREG-1 801 item for component, material, environment, and aging effect.C. Component is different, but consistent with NUREG- 1801 item for material, environment, and aging effect.E. Consistent with NUREG-1801 for material, environment, and aging effect, but a different aging management program is credited or NUREG-1801 identifies a plant-specific aging management program.Report No. 1200347.401.R1 B-7 VjjfSfruture lit gdiy MsocWes, hio?

APPENDIX C PVNGS UNIT 1 RVI SYSTEM, STRUCTURE, AND COMPONENT TABLES [121 Report No. 1200347.401.R1 C-i1 Iovy scasIc Table C-1. List of Materials in Palo Verde Unit 1 [12]RVI Component 1 Fabrication RVI Component Group D iption Material I Material Specifications/

ASME Code Steel Type GSSS Sthcrical Washer Nitronic 60 SA479, S21800 Stainless Steel Upper Internals Assembly Top Plug 304SS SA479, 304SS Stainless Steel GSSS Support Cylinder 304SS SA240, 304SS Stainless Steel Upper Flange 304SS SA182, 304 SS Stainless Steel Tie Rod Nose Adapter 304SS ASTM A479, 304SS Stainless Steel Tube 304SS SA312, 304SS Stainless Steel GSSS Base Flange Forging 304SS SA240 or SAl 82, 304SS Stainless Steel Cylinder 304SS SA240, 304SS Stainless Steel Flange Block 304SS A240, 304SS Stainless Steel Lock Pin SS ASTM A479, 300 Series SS Stainless Steel GSSS Guide Structure Plate 304SS SA240, 304SS Stainless Steel Lower Flange 304SS SA 182, 304 SS Stainless Steel Pin A286 SS SA638, Grade 660 Super Alloy Steel Hex Nut 304SS SA194, B8, 304SS Stainless Steel Plate 304SS SA240, 304SS Stainless Steel Plug 304SS A479, 304SS Stainless Steel Plug 304SS SA,,',79, TP304SS Stainless Steel Bottom Plug 304SS SA479,304SS Stainless Steel Collar 304SS SA240, 304SS Stainless Steel GSSS Shim 304SS SA240, 304SS Stainless Steel Tube 304SS SA213, 304SS Stainless Steel Shim XM-29 A240, XM-29 Stainless Steel Report No. 1200347.401.R1 C-2 v an"uoww Inqftdy Assoackfs, hIm?

Table C-1. List of Materials in Palo Verde Unit 1 [121 (continued)

RVI Component Group RVI Component Description Fabrication Material Material Specifications/

ASME Code Steel Type+ I + 4 Locking Strap 304SS SA240, 304 SS Stainless Steel Upper Internals Assembly Fuel Alignment Plate 304SS SA240. 304SS Stainless Steel Guide 304SS SA 182, 304 SS Stainless Steel Socket Head Cap Screw 41OSS A 193, Grade B6 Stainless Steel Insert 304SS SA213, 304SS Stainless Steel Stop 304SS SA479, 304SS Stainless Steel Lifting Hook 304SS SA240, 304SS Stainless Steel FAP Sleeve Insert 304SS SA213, Grade TP, 304SS Stainless Steel Socket Head Cap Screw B8M A 193, Grade B8M Stainless Steel Sleeve Plug 304SS SA240 or SA479, Grade TP, 304SS Stainless Steel Pin Austenitic SS SA479, 300 Series Stainless Steel Locking Cup 304SS SA479 or SA213, 304SS Stainless Steel Round Nut 8M SA194, Grade 8M Stainless Steel Stud B8M SA193, Grade B8M Stainless Steel Guide Lug Insert 348SS A240, 348SS Stainless Steel Alignment Key A286 SA638, Grade 660 Super Alloy Steel Hold-Down Ring SS SA 182, Code Case 1747 Stainless Steel Socket Head Cap Screw A286 SA453, Grade 660 Super Alloy Steel Dowel Pin A286 SA638, Grade 660 Super Alloy Steel Dowel Pin 304SS SA479, 304SS Stainless Steel Snubber Block 304SS A240, 304SS Stainless Steel CEA Shroud Assemblies Plate 304SS A240, 304SS Stainless Steel Shroud Tube 304SS SA240, 304SS Stainless Steel Report No. 1200347.401.R1 C-3 CjjSfrucbirf Itwdiy Assaokftos, kic?

Table C-1. List of Materials in Palo Verde Unit 1 [12] (continued)

RVI Component Group RVI Component Description Fabrication Material Specifications!

ASME Code Steel Type RVICmponetGrupRVCompnentDescrptio Material MateralSecifiatios/ASM_

CodSteeTyp Shroud Flange 304SS SA240, 304SS Stainless Steel Shroud Web 304SS SA240, 304SS Stainless Steel Quarter Round 304SS SA240, 304SS Stainless Steel Half Round 304SS SA240, 304SS Stainless Steel CEA Shroud Assemblies Web 304SS SA240,304SS Stainless Steel Gusset 304SS SA240, 304SS Stainless Steel Shroud Comer Web 304SS SA240, 304SS Stainless Steel Top Plate 304SS SA240, 304SS Stainless Steel Base Plate 304SS SA240, 304SS Stainless Steel Tie Rod Tube 304SS SA240,304SS Stainless Steel Upper Flange 304SS SA 182, 304 SS Stainless Steel Upper Cylinder 304SS SA240, 304SS Stainless Steel Nozzle 304SS SA182, 304 SS Stainless Steel Nozzle Cylinder 304SS SA240, 304SS Stainless Steel Core Support Barrel Assembly Center Cylinder 304SS SA240, 304SS Stainless Steel Lower Cylinder 304SS SA240, 304SS Stainless Steel Snubber Lug 304SS SA182, 304 SS Stainless Steel Lower Flange 304SS SA182, 304 SS Stainless Steel Lift Bolt Insert Nitronic 60 Spec 00000-MCD-006, Rev. 01 Stainless Steel Intermediate Ring 304SS SA240, 304SS Stainless Steel Core Shroud Assembly Top Plate 304SS SA240, 304SS Stainless Steel Bottom Plate 304SS SA240, 304SS Stainless Steel Shroud Plate 304SS SA240, 304SS Stainless Steel Report No. 1200347.401.R1 C-4 V~Inwt~wui Initdfy Asac~als, Inc?

Table C-1. List of Materials in Palo Verde Unit 1 [121 (continued)

Fabrication RVI Component Group RVI Component Description Material Material Specifications/

ASME Code Steel Type Rib 304SS SA240, 304SS Stainless Steel Guide Lug 348SS SA182, 348SS Stainless Steel Core Shroud Channel Assy 304SS SA240, 304SS Stainless Steel Core Shroud Assembly Ring 304SS SA240, 304SS Stainless Steel End Plate 304SS SA240, 304SS Stainless Steel Channel 304SS SA240, 304SS Stainless Steel Core Shroud Channel 304SS SA240, 304SS Stainless Steel Lifting Block 304SS SA240, 304SS Stainless Steel Insert Pin A286 SA638, Grade 660 Super Alloy Steel Cylinder 304SS SA240, 304SS Stainless Steel Support Beams 304SS SA240, 304SS Stainless Steel Column 304SS SA479, 304SS Stainless Steel Lower Internals Assembly Bottom Plate 304SS SA240, 304SS Stainless Steel Lower Plate 304SS SA240, 304SS Stainless Steel Gusset 304SS SA240, 304SS Stainless Steel Support Pad 304SS SA240, 304SS Stainless Steel Lockbar 304SS SA479, 304SS Stainless Steel Sleeve 304SS SA213, Grade TP 304SS Stainless Steel Instrumentation Nozzle 304SS SA479, 304SS Stainless Steel ICI Instrument Nozzle Nut 304SS SA479, 304SS Stainless Steel Instrument Nozzle Support Plate 304SS SA240, 304SS Stainless Steel Instrument Support Beam 304SS SA182, 304 SS Stainless Steel Report No. 1200347.401.Rl C-5 R4txwh I tItidfy Assacu s, thim APPENDIX D PVNGS UNIT 2 RVI SYSTEM, STRUCTURE, AND COMPONENT TABLES [121 Report No. 1200347.401.R1 D-1 uJXSUwLhI'FIhag*

YMSW , IW Table D-1. List of Materials in Palo Verde Unit 2 [12]RVI Component RVI Component Description Fabrication Material Material Specilications/

ASME Code Steel Type Group I _ _ _ _ _ _ _ _ _ I _ _ _ __ _ _ I1 ___ __I Tie Rod Nose Adapter 304SS SA479, 304SS Stainless Steel Upper Internals Assembly Lock Pin Austenitic SS SA479, 300 Series SS Stainless Steel GSSS Sperical Washer Nitronic 60 SA479, S21800 Stainless Steel Top Plug 304SS SA479, 304SS Stainless Steel Tube 304SS SA312, 304SS Stainless Steel Bottom Plug 304SS SA479, 304SS Stainless Steel Hex Nut 304SS SA 194, B8, 304SS Stainless Steel Collar 304SS SA240, 304SS Stainless Steel Locking Strap 304SS SA240, 304 SS Stainless Steel Plug 304SS SA479, 304SS Stainless Steel Pin A286 SA638, Grade 660 Super Alloy Steel Socket Head Cap Screw 41OSS SA 193, Grade B6 Stainless Steel Socket Head Cap Screw B8M SA193-71, Grade B8M Stainless Steel Pin Austenitic SS SA479, 300 Series Stainless Steel Locking Cup 304SS SA479 or SA213, 304SS Stainless Steel Round Nut 8M SA 194, Grade 8M Stainless Steel Stud B8M SA193, Grade B8M Stainless Steel Flange Block 304SS SA240, 304SS Stainless Steel Shim Nitronic 33 SA240, XM-29 Stainless Steel GSSS Support Cylinder 304SS SA240, 304SS Stainless Steel GSSS Base Flange Forging 304SS SA240 or SA 182, 304SS Stainless Steel GSSS Guide Structure Plate 304SS SA240, 304SS Stainless Steel Report No. 1200347.401.Rl D-2 Rs74cbouIn IL*e Auociates, Inm.

Table D-1. List of Materials in Palo Verde Unit 2 [121 (continued)

RVI Component Group RVI Component Description Fabrication Material Material Specilications/

ASME Code Steel Type GSSS Shim 304SS SA240, 304SS Stainless Steel Upper Flange 304SS SA 182, 304 SS Stainless Steel Cylinder 304SS SA240,304SS Stainless Steel Lower Flange 304SS SAI82, 304 SS Stainless Steel Plate 304SS SA240, 304SS Stainless Steel Tube 304SS SA213, 304SS Stainless Steel Fuel Alignment Plate 304SS SA240, 304SS Stainless Steel Guide 304SS SA182, 304 SS Stainless Steel Insert 304SS SA213, 304SS Stainless Steel Upper Internals Assembly Stop 304SS SA479, 304SS Stainless Steel Lifting Hook 304SS SA240, 304SS Stainless Steel FAP Sleeve Insert 304SS SA213, Grade TP, 304SS Stainless Steel Sleeve Plug 304SS SA240 or SA479, 304SS Stainless Steel Socket Head Cap Screw A286 SA453, Grade 660 Super Alloy Steel Guide Lug Insert 348SS SA240, 348SS Stainless Steel Dowel Pin A286 SA638, Grade 660 Super Alloy Steel Dowel Pin 304SS SA479, 304SS Stainless Steel Alignment Key A286 SA638, Grade 660 Super Alloy Steel Hold-Down Ring SS SA 182, Code Case 1747 Stainless Steel Shroud Flange 304SS SA240, 304SS Stainless Steel Quarter Round 304SS SA240, 304SS Stainless Steel CEA Shroud Assemblies Half Round 304SS SA240, 304SS Stainless Steel Web 304SS SA240, 304SS Stainless Steel Gusset 304SS SA240, 304SS Stainless Steel Report No. 1200347.401.R1 D-3 Cjsftwo~n Inftorily Assowates, Wnc Table D-1. List of Materials in Palo Verde Unit 2 [121 (continued)

RVI Component Group RVI Component Description Fabrication Material Material Specilications/

ASME Code Steel Type Snubber Block 304SS SA240, 304SS Stainless Steel Plate 304SS SA240, 304SS Stainless Steel Shroud Tube 304SS SA240, 304SS Stainless Steel CEA Shroud Assemblies Shroud Web 304SS SA240, 304SS Stainless Steel Shroud Comer Web 304SS SA240, 304SS Stainless Steel Top Plate 304SS SA240, 304SS Stainless Steel Base Plate 304SS SA240, 304SS Stainless Steel Tie Rod Tube 304SS SA240, 304SS Stainless Steel Upper Flange 304SS SA182, 304 SS Stainless Steel Upper Cylinder 304SS SA240, 304SS Stainless Steel Nozzle 304SS SA182, 304 SS Stainless Steel Nozzle Cylinder 304SS SA240, 304SS Stainless Steel Core Support Barrel Assembly Center Cylinder 304SS SA240, 304SS Stainless Steel Lower Cylinder 304SS SA240, 304SS Stainless Steel Snubber Lug 304SS SA182, 304 SS Stainless Steel Lower Flange 304SS SA182, 304 SS Stainless Steel Lift Bolt Insert Nitronic 60 Spec 00000-MCD-006, Rev. 01 Stainless Steel Intermediate Ring 304SS SA240, 304SS Stainless Steel Top Plate 304SS SA240. 304SS Stainless Steel Bottom Plate 304SS SA240, 304SS Stainless Steel Core Shroud Assembly Shroud Plate 304SS _ SA240, 304SS Stainless Steel Rib 304SS SA240, 304SS Stainless Steel Guide Lug 348SS SA182, 348SS Stainless Steel Report No. 1200347.401.R1 D-4 V OW&W bvI ue di Assocktas, Inc Table D-1. List of Materials in Palo Verde Unit 2 [12] (continued)

RVI Component Group RVI Component Description Fabrication Material Material Specilications/

ASME Code Steel Type Ring 304SS SA240, 304SS Stainless Steel End Plate 304SS SA240, 304SS Stainless Steel Core Shroud Assembly Channel 304SS SA240, 304SS Stainless Steel Core Shroud Channel 304SS SA240, 304SS Stainless Steel Lifting Block 304SS SA240, 304SS Stainless Steel Column 304SS SA479, 304SS Stainless Steel Gusset 304SS SA240, 304SS Stainless Steel Support Pad 304SS SA240, 304SS Stainless Steel Comer Support Pad 304SS SA240, 304SS Stainless Steel Cylinder 304SS SA240,304SS Stainless Steel Lower Internals Assembly Support Beams 304SS SA240, 304SS Stainless Steel Bottom Plate 304SS SA240,304SS Stainless Steel Lower Plate 304SS SA240, 304SS Stainless Steel Lockbar 304SS SA479, 304SS Stainless Steel Sleeve 304SS SA213, Grade TP 304SS Stainless Steel Insert Pin A286 SA638, Grade 660 Super Alloy Steel Instrumentation Nozzle 304SS SA479, 304SS Stainless Steel ICI Instrument Nozzle Nut 304SS SA479, 304SS Stainless Steel Instrument Nozzle Support Plate 304SS SA240, 304SS Stainless Steel Instrument Support Beam 304SS SA182, 304 SS Stainless Steel Report No. 1200347.401.R1 D-5 CjSbucftn Iu~fwgly Assodatss

/nc APPENDIX E PVNGS UNIT 3 RVI SYSTEM, STRUCTURE, AND COMPONENT TABLES [12]Report No. 1200347.401.R1 E-1 Asrsoc Mo, ,Aisatdes, In Table E-1. List of Materials in Palo Verde Unit 3 [12]RVI Component Group RVI Component Fabrication Material Specillcations/

ASME Code Steel Type I Description I Material I Spherical Washer Martensitic SS SA479, Type 410, Condition 2 Stainless Steel Upper Internals Assembly Top Plug 304SS SA479, 304SS Stainless Steel GSSS Support Cylinder 304SS SA240, 304SS Stainless Steel Upper Flange 304SS SA 182, 304 SS Stainless Steel Tube 304SS SA312, 304SS Stainless Steel Flange Block 304SS SA240, 304SS Stainless Steel GSSS Base Flange Forging 304SS SA240 or SA182, 304SS Stainless Steel Cylinder 304SS SA240, 304SS Stainless Steel Bottom Plug 304SS SA479, 304SS Stainless Steel Pin A286 SA638, Grade 660 Super Alloy Steel GSSS Guide Structure Plate 304SS SA240, 304SS Stainless Steel Lower Flange 304SS SA 182, 304 SS Stainless Steel Hex Nut 304SS SA479, 304SS Stainless Steel Plug 304SS SA479, 304SS Stainless Steel Plate 304SS SA240, 304SS Stainless Steel Collar 304SS SA240, 304SS Stainless Steel Locking Strap 304SS SA240, 304 SS Stainless Steel Shim Nitronic 33 SA240, XM-29 Stainless Steel Tube 304SS SA213, 304SS Stainless Steel Fuel Aligtnment Plate 304SS SA240,304SS Stainless Steel Guide 304SS SA 182, 304 SS Stainless Steel LockinE Cuo 304SS SA479 or SA213. 304SS Stainless Steel& I L t Report No. 1200347.401.Rl E-2 ReportNo.

120347.41.R1 E2 ~j~IShc&wa Inftogi Assowates WOc Table E-1. List of Materials in Palo Verde Unit 3 [121 (continued)

RVI Component Group RVI Component Description Fabrication Material Specilications!

ASME Code Steel Type Material MaterialSpecillcations/ASMECode SteelType Shear Pin A286 SA638, Grade 660 Super Alloy Steel Modified Socket Head Cap Screw B8M SA193, Grade B8M Stainless Steel Insert 304SS SA213, 304SS Stainless Steel Stop 304SS SA479, 304SS Stainless Steel Lifting Hook 304SS SA240, 304SS Stainless Steel FAP Sleeve Insert 304SS SA213, Grade TP, 304SS Stainless Steel Sleeve Plug 304SS SA240 or SA479, 304SS Stainless Steel Upper Internals Assembly Socket Head Cap Screw 410SS SA193. Grade B6 Stainless Steel Washer 304SS SA479, 304SS Stainless Steel Socket Head Cap Screw B8M SA193-71, Grade B8M Stainless Steel Pin Austenitic SS SA479, 300 Series Stainless Steel Guide Lug Insert 348SS SA240, 348SS Stainless Steel Alignment Key A286 SA638, Grade 660 Super Alloy Steel Hold-Down Ring SS SA 182, Code Case 1747 Stainless Steel Socket Head Cap Screw A286 SA453, Grade 660 Super Alloy Steel Dowel Pin A286 SA638, Grade 660 Super.Alloy Steel Dowel Pin 304SS SA479, 304SS Stainless Steel Snubber Block 304SS SA240, 304SS Stainless Steel Plate 304SS SA240, 304SS Stainless Steel CEA Shroud Assemblies Shroud Tube 304SS SA240, 304SS Stainless Steel Shroud Flange 304SS SA240, 304SS Stainless Steel Shroud Web 304SS SA240,304SS Stainless Steel Report No. 1200347.401.Rl E-3 Report No. 1200347.40 1 .R1 E-3 ~j InsWc& I tgly ASSOOM~US moP Table E-1. List of Materials in Palo Verde Unit 3 [12] (continued)

RVI Component Group RVI Component Description Fabrication Material Material Specillcations/

ASME Code Steel Type Quarter Round 304SS SA240, 304SS Stainless Steel Half Round 304SS SA240, 304SS Stainless Steel Web 304SS SA240, 304SS Stainless Steel CEA Shroud Assemblies Gusset 304SS SA240,304SS Stainless Steel Shroud Comer Web 304SS SA240, 304SS Stainless Steel Top Plate 304SS SA240, 304SS Stainless Steel Base Plate 304SS SA240, 304SS Stainless Steel Tie Rod Tube 304SS SA240, 304SS Stainless Steel Lift Bolt Insert Nitronic 60 Spec 00000-MCD-006, Rev. 01 Stainless Steel Upper Flange 304SS SA182, 304 SS Stainless Steel Upper Cylinder 304SS SA240, 304SS Stainless Steel Nozzle 304SS SA182, 304 SS Stainless Steel Core Support Barrel Assembly Nozzle Cylinder 304SS SA240, 304SS Stainless Steel Center Cylinder 304SS SA240, 304SS Stainless Steel Lower Cylinder 304SS SA240, 304SS Stainless Steel Snubber Lug 304SS SA 182, 304 SS Stainless Steel Lower Flange 304SS SA182, 304 SS Stainless Steel Intermediate Ring 304SS SA240, 304SS Stainless Steel Top Plate 304SS SA240, 304SS Stainless Steel Core Shroud Assembly Bottom Plate 304SS SA240, 304SS Stainless Steel Shroud Plate 304SS SA240, 304SS Stainless Steel Rib 304SS SA240, 304SS Stainless Steel Guide Lug 348SS SA 182, 348SS Stainless Steel Report No. 1200347.401.R1 E-4 CSWnj&Wu Intogr* Assoadet~

Inc Table E-1. List of Materials in Palo Verde Unit 3 [12] (continued)

RVI Component Group RVI Component Description Fabrication Material Material Specillcations/

ASME Code Steel Type Core Shroud Channel Assy 304SS SA240, 304SS Stainless Steel Core Shroud Assembly Ring 304SS SA240, 304SS Stainless Steel Lifting Block 304SS SA240, 304SS Stainless Steel Insert Pin A286 SA638, Grade 660 Super Alloy Steel Cylinder 304SS SA240, 304SS Stainless Steel Support Beams 304SS SA240, 304SS Stainless Steel Column 304SS SA479, 304SS Stainless Steel Lower Internals Assembly Bottom Plate 304SS SA240, 304SS Stainless Steel Lower Plate 304SS SA240, 304SS Stainless Steel Gusset 304SS SA240, 304SS Stainless Steel Support Pad 304SS SA240, 304SS Stainless Steel Comer Support Pad 304SS SA240, 304SS Stainless Steel Lockbar 304SS SA479, 304SS Stainless Steel Instrumentation Nozzle 304SS SA479, 304SS Stainless Steel ICI Instrument Nozzle Nut 304SS SA479, 304SS Stainless Steel Instrument Nozzle Support Plate 304SS SA240, 304SS Stainless Steel__Instrument Support Beam 304SS SA182, 304 SS Stainless Steel Report No. 1200347.401.R1 E-5 cj(Sbuuwi 1119 W* Assocktat, WDc.