ML14141A555

Arkansas Nuclear One, Unit 1, Attachment 1 to 1CAN051403 PWR Internals Aging Management Program Plan

ML14141A555
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/01/2014
From:	Lohse C S Structural Integrity Associates
To:	Entergy Nuclear Operations
References
1CAN051403 1200459.401, Rev 1
Download: ML14141A555 (161)

Text

Attachment 1 to 1 CAN051403 PWR Internals Aging Management Program Plan For Arkansas Nuclear One, Unit 1 Report No. 1200459.401 Revision 1 Project No. 1301511 May 2014 PWR Internals Aging Management Program Plan for Arkansas Nuclear One, Unit 1 Prepared for: Entergy Nuclear Operations, Inc.Contract Order Nos: 10354474 and 10406996 Prepared by: Structural Integrity Associates, Inc.San Jose, California Prepared by: Reviewed by: Approved by: Christopher Lohse, P.E.

Griesbach T .Griesbach Date: 5/1/2014 Date: 5/1/2014 Date: 5/1/2014 Structural Integrity Associates, Inc.

REVISION CONTROL SHEET Document Number: 1200459.401 Title: PWR Internals Aging Management Program Plan for Arkansas Nuclear One, Unit 1 Client: Entergy Nuclear Operations SI Project Number: 1301511 Quality Program: N Nuclear E Commercial Section Pages Revision Date Comments 1.0 1-1 16 0 1/3/13 Initial Issue 2.0 2-1-2-25 3.0 3-1 36 4.0 4-1 5 5.0 5-1-5-30 6.0 6-1 3 App. A A-1 -A-3 App. B B-1 -B-12 App. C C-1 -C-7 1.0 1-3, 1-8, 1 5/1/2014 Revised to add confirmation of upper 1-12, 1-13, flange heat treatment, orphan components, 1-15 operational experience from vent valves, 2.0 2-1, 2-4, 2-7, Appendix D, and other comments 2-9, 2-10, 2-13, 2-19, 2-20, 2-24, 2-25 3.0 3-1, 3-3, 3-9, 3-22 4.0 4-2 5 5.0 5-1 5-3 9, 5-12 33 6.0 6-1-6-4 App. A A-2 App. B B-1 -B-12 App. D D-1 -D-3 r Structural Integrity Associates, Inc!

Table of Contents Section Page

1.0 INTRODUCTION

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

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

1-1 1.2 ANO- 1 License Renewal Background

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

1-2 1.3 ANO- 1 Vessel Internals Aging Management Program Background

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

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

1-6 1.5 R esp on sib ilities .............................................................................................................

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

1-9 1.6.1 ASME Section XI Inservice Inspection Program Subsections IWB, IWC, IWD a n d I W F ......................................................................................................................

1-1 0 1.6.2 W ater Chem istty Program .....................................................................................

1-10 1.7 Aging Management Review and Program Enhancements

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

1-11 1.7.1 Reactor Internals Aging Management Review Process .........................................

1-11 1.8 Industry Program s .......................................................................................................

1-11 1.8.1 BA W-2248A, Demonstration of the Management ofAging Effects for the R eactor Vessel Internals

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

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

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

1-12 1.8.3 NEI 03-08 Guidance Within MRP-227-A

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

1-12 1.8.4 MRP-22 7-A AMP Development Guidance ............................................................

1-14 1.8.5 O ngoing Industry Program s ..................................................................................

1-15 1.9 S um 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 C orrosion C racking .......................................................................................

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

2-1 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 ...................................................................................

2-2 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 ANO-1 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 ...............................

2-11 2.3.3 NUREG-1801/AMP Program Element 3: Parameters Monitored/Inspected

........ 2-12 2.3.4 NUREG-1801/AMP Program Element 4: Detection ofAging Effects ...................

2-14 2.3.5 NUREG-1801/AMP Program Element 5: Monitoring and Trending ....................

2-17 2.3.6 NUREG-1801/AMP Program Element 6: Acceptance Criteria ............................

2-18 2.3.7 NUREG-1801/AMP Program Element 7: Corrective Actions ...............................

2-20 Report No. 1200459.401 .Rl iii r Structural Integrity Associates, Inc 2.3.8 NUREG-1801/AMP Program Element 8." Confirmation Process .........................

2-22 2.3.9 NUREG-1801/AMP Program Element 9." Administrative Controls ......................

2-23 2.3.10 NUREG-1801/AMP Program Element 10: Operating Experience

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

2-23 3.0 ANO-1 REACTOR VESSEL INTERNALS DESIGN AND OPERATING EXPERIENCE

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

3-1 3A1 Description of ANO- 1 Reactor Vessel Internals

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

3-1 3.1.1 P lenum A ssem blv ................................................................................................

3-1 3.1.2 P lenum C over A ssem bly ..........................................................................................

3-1 3.1.3 P lenum Cylinder Assem bly ......................................................................................

3-3 3.1.4 Upp er G rid A ssem bly ...............................................................................................

3-5 3.1.5 Control Rod Guide Tube (CRGT) Assemblies

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

3-6 3.1.6 Core Support Shield Assembly (CSS) ......................................................................

3-8 3.1.7 Core Barrel Assembly (CBA) .................................................................................

3-12 3.1.8 Low er Internals Assem bly ......................................................................................

3-15 3.1.9 F low D istributor A ssem bly ....................................................................................

3-19 3.1.10 Incore Guide Tube Assemblies

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

3-21 3.2 ANO-1 Design Modifications and Distinctions

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

3-35 4.0 EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA ...................

4-1 4.1 Examination Acceptance Criteria .................................................................................

4-1 4.1.1 Visual (VT-3) Exam ination ......................................................................................

4-1 4.1.2 Visual (VT-1) Exam ination ......................................................................................

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

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

4-2 4.1.4 Sutface E xam ination ................................................................................................

4-3 4.1.5 Volum etric Exam ination ..........................................................................................

4-3 4.1.6 Physical Measurements Examination

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

4-4 4 .2 E xpansion C riteria ........................................................................................................

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

4-4 4.3.1 R ep o rting ..................................................................................................................

4 -5 4.4 Im plem entation Schedule ..............................................................................................

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 Programs):

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

5-4 5.4 SE Section 4.2.4, Applicant/Licensee Action Item 4 (B&W Core Support Structure U pper Flange Stress R elief): ....................................................................................

5-4 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 ponents):

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

.....5-5 Report No. 1200459.401 .Rl iv $Structural Integrity Associates, Inc!04egiy1-oiteI1 5.6 SE Section 4.2.6, Applicant/Licensee Action Item 6 (Evaluation of Inaccessible B & W C om ponents):

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

5-6 5.7 SE Section 4.2.7, Applicant/Licensee Action Item 7 (Plant-Specific Evaluation of C A S S M aterials):

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

5-7 5.8 SE Section 4.2.8, Applicant/Licensee Action Item 8 (Submittal of Information for Staff R eview and A pproval):

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

5-10

6.0 REFERENCES

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

6-1 APPENDIX A SECTION XI 10 YEAR ISI EXAMINATIONS OF B-N-3 INTERNALS COMPONENTS FOR ANO-1 115, 18] ...................

A-1 APPENDIX B MRP-189, REV. 1 COMPONENT REVIEW FOR ANO-1 13, 10, 181 ........ B-1 APPENDIX C REACTOR VESSEL INTERNALS MATERIALS OF CONSTRUCTION FOR ANO-1 1181 ..................................................................

C-1 APPENDIX D MANAGING LOSS OF DUCTILITY

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

D-1 Report No. 1200459.401 .R1 V Structural Integrity Associates, Inc?V List of Tables Table Page Table 1-1. Key Elements of the Reactor Vessel Internals Aging Management Program ...........

1-7 Table 5-1. B&W Plants Primary Category Components from Table 4-1 of M R P -227-A [3, 32] ..................................................................................................

5-12 Table 5-2. B&W Plants Expansion Category Components from Table 4-4 of M R P -227-A [3] ........................................................................................................

5-17 Table 5-3. B&W Plants Existing Program Components from AREVA Guidance [32] ...........

5-20 Table 5-4. B&W Plants Examination Acceptance and Expansion Criteria from Table 5-1 of M RP-227-A [3] Applicable to A N O -1 ....................................................................

5-21 Table 5-5. ANO-1 Response to the NRC Final Safety Evaluation of MRP-227-A

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

5-27 Table 5-6. ANO-1 Program Enhancement and Implementation Schedule ...............................

5-30 Table A-1.Section XI 10 Year ISI Examinations of B-N-3 Internals

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

A-2 Table B-1. Summary of Aging Management Evaluations for ANO-1 Reactor Vessel Intern als [3, 10 , 18] ....................................................................................................

B -2 Table C- 1. RVI M aterials of Construction for ANO- 1 [ 18] .......................................................

C-2 Report No. 1200459.401 .Rl vi Ijj-StructuraI Integrity Associates, Inc!Inegiy-soiteIc List of Figures Figur~e Page Figure 3-1. General Arrangement of Typical B&W Reactor Vessel Internals

[3] .........................

3-23 Figure 3-2. Plenum Assembly [9] ...................................................................................................

3-24 Figure 3-3. Plenum Cover Assembly [9] ........................................................................................

3-25 Figure 3-4. Plenum Cylinder Assembly [9] ....................................................................................

3-26 Figure 3-5. Upper Grid Assembly [9] .............................................................................................

3-27 Figure 3-6. Control Rod Assembly [9] ...........................................................................................

3-28 Figure 3-7. CRGT Assembly Spacer Castings and Rod Guide Brazement Configuration

[9] ....... 3-29 Figure 3-8. Control Rod Guide Tube Assembly Pipes [9] ..............................................................

3-30 Figure 3-9. Core Support Assembly [9] ..........................................................................................

3-31 Figure 3-10. Vent Valve Assembly [9] ...........................................................................................

3-32 Figure 3-11. Core Barrel Interior with Baffle Plates [9] .................................................................

3-33 Figure 3-12. Lower Internals Assembly [9] ....................................................................................

3-34 V Structural Integrity Associates, Inc?Report No. 1200459.401.R1 vii LIST OF ACRONYMS AMP Aging Management Program AMR Aging Management Review ANO-1 Arkansas Nuclear One -Unit 1 ARDM Age-related Degradation Mechanism ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox BWOG Babcock & Wilcox Owners Group CASS Cast austenitic stainless steel CBA Core Barrel Assembly CFR Code of Federal Regulations CLB Current licensing basis CRGT Control Rod Guide Tube CSA Core support assembly CSS Core support shield CUF Cumulative Usage Factor EFPY Effective full power years ENO Entergy Nuclear Operations EPRI Electric Power Research Institute ET Eddy Current Testing EVT Enhanced visual testing (visual NDE method indicated as EVT- 1)FD Flow Distributor 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 IE Irradiation Embrittlement Report No. 1200459.401 .Rl viii LJjStructural Integrity Associates, Inc?

IMI Incore Monitoring Instrumentation INPO Institute of Nuclear Power Operations IP Issue Programs ISI Inservice Inspection ISR Irradiation-Enhanced Stress Relaxation LCB Lower Core Barrel LRA License Renewal Application LTS Lower Thermal Shield MRP Materials Reliability Program MSC Materials Subcommittee NDE Nondestructive Examination NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System NUREG Nuclear Regulatory Guide OE Operating Experience PH Precipitation Hardened PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group QA Quality Assurance RCS Reactor Coolant System RFO Refueling Outage RPV Reactor Pressure Vessel RVI Reactor Vessel Internals RVLMS Reactor Vessel Level Monitoring System SCC Stress Corrosion Cracking SE Safety Evaluation SER Safety Evaluation Report SRP Standard Review Plan SS Stainless Steel Report No. 1200459.401 .Rl ix r Structural Integrity Associates, Inc!

SSHT TAC TLAA TS UCB UFSAR UGS UT UTS VT Surveillance Specimen Holder Tube Technical Advisory Committee Time-limited Aging Analysis Technical Specifications Upper Core Barrel Updated Final Safety Analysis Report Upper Guide Structure Ultrasonic Testing Upper Thermal Shield Visual Testing Report No. 1200459.401.R1 rStructural Integrity Associates, Inc.X

1.0 INTRODUCTION

1.1 Objective The purpose of this document is to provide an aging management program for managing aging effects in the reactor vessel internals (RVI) at Arkansas Nuclear One, Unit 1 (ANO- 1) through the period of extended operation, which begins on May 21, 2014. The Aging Management Program (AMP) document coordinates with the existing ASME Section XI inservice inspection.(ISI) program and supplements that program with augmented examinations.

This document identifies the internals components that must be included for aging management review and identifies the augmented inspection plan for the ANO-1 RVI. The program plan implements 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 NRC Safety Evaluation (SE) [4] for ANO- 1. This program document also meets the license renewal commitment for ANO- I for aging management of reactor vessel internals

[5] as specified in the Updated Final Safety Analysis Report (UFSAR), Section 16.1.5[6].The main objectives of the ANO-1 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.* To summarize the role of existing ANO-1 programs related to the 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.

Report No. 1200459.401 .Rl [-1 jStructural Integrity Associates, In0 1.2 ANO-1 License Renewal Background In order to meet a license renewal commitment

[5, 6], ANO-1 will submit this aging management program plan. The license renewal commitment listed below defines the content and timeline for the program that ANO-1 has committed to implement for the RVI Components:

Entergy Operations will participate in the B WOG Reactor Internals Aging Management Program and other industry programs, as appropriate, to continue investigation of aging effects requiring management for the reactor vessel internals.

These activities will assist in establishing the appropriate monitoring and inspection programs for the reactor vessel internals.

Entergy Operations will provide periodic updates after the completion of significant milestones in the preparation of the Reactor Vessel Internals Inspection, commencing within one year of the issuance of the renewed license. Entergy Operations will submit a report to the NRC at, or about, the end of the initial 40-year operating license term. This report will summarize the current understanding of aging effects applicable to the reactor vessel internals and will contain the Entergy Operations' inspection plan, including methods for each inspection.

Entergy Operations will perform the Reactor Vessel Internals Inspection as provided in the following summaty description.

Should data or evaluations indicate that this inspection can be modified or eliminated, Entergy Operations will provide plant-specific justification to demonstrate the basis for the modification or elimination.

The aging management program for ANO- 1 will be 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 period of extended operation.

Furthermore, this AMP will demonstrate the consistency of the program with the elements documented in NUREG- 1801, Revision 2 [7], Chapter XI.M. 16A, "PWR Vessel Report No. 1200459.401 .Rl 1-2 C structural Integrity Associates, Inc Internals." The operating experience provided by NUREG- 1801, Revision 2 will also be reviewed and incorporated into plant-specific programs.ANO-1 also has a specific commitment to demonstrate that the RVI will have sufficient ductility through the period of extended operation.

This commitment is to be included in the RVI AMP.Appendix D of this document contains the discussion of the analyses performed to show sufficient ductility of the RVI.1.3 ANO-1 Vessel Internals Aging Management Program Background Management of aging effects in RVI is required for nuclear plants applying for renewed operating licenses, as specified in the NRC Standard Review Plan (SRP) for License Renewal Applications

[8]. 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 2000, the B&W Owners Group (BWOG) issued BAW-2248A, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals

[9]." 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[10- 12].Screening criteria were developed, considering chemical composition, neutron fluence exposure, temperature history, and representative stress levels, for determining the Report No. 1200459.401 .Rl 1-3 I Structural Integrity Associates, ln0"Ueriysscats1n.

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.

Aging management strategies for implementing the appropriate aging management methodology, baseline examination timing, and the need and timing for 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 EPRI MRP, has finalized the inspection and evaluation (IE) 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].Report No. 1200459.401.Rl 1-4 V Structural Integrity Associates, Inc!

MRP-228 [11], "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-NP,"Reactor Internals Acceptance Criteria Methodology and Data Requirements" [13] 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.

ANO-1 reactor vessel internals are a part of the primary reactor coolant system (RCS), which is a two-loop B&W designed nuclear steam supply system (NSSS).A review of Section 3.2.5 of the ANO-1 LRA specifies that the following reactor vessel internals within the scope of license renewal are subject to aging management review:* Plenum assembly* Core support shield assembly" Core barrel assembly" Lower internals assembly" Reactor vessel level monitoring system (RVLMS) probe supports* Remaining portions of the surveillance specimen holder tubes* Thermal shield and thermal shield upper restraint As described in Section 2.3.1.6 of the license renewal application (LRA), ANO-1 is bounded by BAW-2248A

[9] with regard to reactor vessel internals within the first four groups listed above.As a result of NRC review of BAW-2248A, twelve applicant action items related to license renewal were identified.

ANO-I specific responses to the license renewal applicant action items relevant to the reactor vessel internals are provided in Table 2.3-5 of the ANO-I LRA [5]. The Report No. 1200459.401 .Rl 1-5 V Structural Integrity Associates, Inc RVLMS probe supports, surveillance specimen holder tubes, and thermal shield and thermal shield upper restraint are not within the scope of BAW-2248A.

However, based on the license renewal action item discussed in Table 2.3-5 of the license renewal application (RVI Applicant Action Item No. 3) the last three items listed are within the scope of license renewal and are subject to aging management review for ANO-1.1.4 ANO-1 Reactor Vessel Internals Aging Management Program Elements The key elements of the ANO-1 Reactor Vessel Internals Aging Management Program are outlined in Table 1. The program attributes are described in detail in Section 2.3 of this document.

Additionally, ANO-1 participates in PWR Owners Group Materials Subcommittee (PWROG MSC) and EPRI 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, EPRI MRP, and PWROG MSC, the RVI AMP key elements will be updated to include any relevant operational experience (OE) or lessons learned.Report No. 1200459.401 .Rl 1-6 V Structural Integrity Associates, In0 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 ANO-1 monitors, inspects, and/or tests for the effects of the eight aging degradation Monitored/Inspected 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 ANO-I ASME Section XI [14] ISI program for B-N-3 internals components Effects (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 Monitoring and Augmented inspections in accordance with MRP-227-A provided in this document, Trending in combination with the ASME Section XI [14] ISI program, provides reasonable assurance for demonstrating the ability of RVI components to perform the intended functions.

Reporting requirements consistent with the guidance provided in MRP-227-A allows the industry to monitor and trend results and take appropriate preemptive actions through the EPRI MRP guidelines 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 appropriate ASME Section XI [14] and WCAP-17096

[13] 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

[3].8 Confirmation The confirmation process for the RV Internals Program is described in Section Process 2.3.8 of this document.9 Administrative Administrative controls that apply to the RVI AMP, procedures, reviews and Controls approval processes is described in Section 2.3.9 of this document 10 Operating Operating experience related to the ANO-l RV internals is described in Section Experience 2.3.10 of this document.Report No. 1200459.401.R1 1-7 V Structural Integrity Associates, Inc.

1.5 Responsibilities

The RVI Program Manager has overall responsibility for the development and implementation of the RVI aging management plan. The responsibilities for implementing the NEI 03-08, Materials Initiative Process, are described in Reference

1. The RVI program is implemented in accordance with EN-DC-133

[33]. ENO actively participates in industry programs related to materials initiatives such as PWROG, EPRI MRP, and other programs related to aging management of reactor vessel internals.

The Reactor Vessel Internals Program Manager is responsible for: " Overall development of the RVI aging management plan" Administering and overseeing the implementation of the RVI aging management plan* Ensuring that regulatory requirements related to inspection activities, if any, are met and incorporated into the plan" Communicating with senior management on periodic updates to the plan" Maintaining the RVI aging management plan to incorporate changes and updates based on new knowledge and experience gained* Reviewing and approving industry and vendor programs related to RVI aging management activities

• Processing of any deviations taken from issue programs (IP) guidelines in accordance with NEI 03-08 [2] requirements" Ensure prompt notification of RCS Materials Degradation Management Program Manager whenever an issue or indication of potential generic industry significance is identified

• Participate in the planning and implementation of inspections of the internals.

Report No. 1200459.401.R1 1-8 j Structural Integrity Associates, Inc!

The ISI Engineer is responsible for:* Planning and implementing inspections required by Section XI for B-N-3 components

[ 15], the supplemental inspections identified in this program plan, and any other plant-specific commitments for inspection for managing aging of RVI.* Participating in industry groups such as PDI, EPRI MRP TAC Inspection Subcommittee, etc.The ISI Engineer and Level III NDE Coordinator are responsible for:* Providing NDE services* Reviewing and approving vendor NDE procedures and personnel qualifications

• Providing direction and oversight of contracted NDE activities 1.6 Program Implementation ANO-1 's overall strategy for managing aging in reactor vessel internals components is supported by the following existing programs: " ASME Section XI Inservice Inspection Program, Subsections IWB, IWC, IWD, IWF" Water Chemistry Program These are established programs that support the aging management of RCS components in addition to the RVI components.

Report No. 1200459.401 .Rl 1-9 Structural Integrity Associates, Inc!

1.6.1 ASME Section XI Inservice Inspection Program Subsections IWB, IWC, IWD and IWF The ASME Section XI Inservice Inspection Program, Subsections IWB, IWC, IWD, IWF [15] 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, "Rules for Inservice Inspection of Nuclear Power Plant Components" [14].1.6.2 Water Chemistry Program The water chemistry program [16] is an existing program that is credited for managing aging effects by controlling the environment to which internals surfaces of systems and components are exposed. Aging 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 These 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 EPRI PWR Primary Water Chemistry Guidelines

[17].Report No. 1200459.401 .R1 1-10 C Structural Integrity Associates, Inc!

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 ANO-1 license renewal application

[5]. BWOG report BAW-2248A

[9] performed an aging management review (AMR) of the reactor vessel internals used in B&W plants. As part of the ANO-1 license renewal application, ENO performed a comparison of the information in BAW-2248A to the ANO-1 site specific documentation to verify the applicability to ANO-I[18]. The summary of the aging management review (AMR) is documented in Section 3.2.5 of the ANO-1 LRA. The NRC has indicated its approval of the ANO-1 LRA in Safety Evaluation Report SR-1743 [19].The aging management review process supporting the LRA performed the following:

1. Identified applicable aging effects requiring management.

2. Identified appropriate 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 manage the aging effects identified in the review.AMRs were performed for each ANO- 1 system that contained long-lived, passive components requiring an aging management review, in accordance with the ANO-1 screening process. The results of these reviews have been incorporated into the ANO-1 RVI AMP.1.8 Industry Programs 1.8.1 BA W-2248A, Demonstration of the Management ofAging Effects for the Reactor Vessel Internals The Babcock & Wilcox Owners Group (BWOG) topical report BAW-2248A

[9] provides a technical evaluation of the effects of aging of the reactor vessel internals and demonstrates that aging effects within the scope of the report are adequately managed for the period of extended Report No. 1200459.401.R1 1-11 C Structural Integrity Associates, Inc!

operation associated with license renewal. The BWOG report provided guidance for BWOG member plant owners to manage effects of aging of RVI during the period of extended operation, using approved aging management methodologies to develop plant-specific AMPs.The AMR for the ANO-1 internals, documented in the ANO-l license renewal application

[5], was completed in a manner consistent with the approach of BAW-2248A

[9]. Both the ANO-1 specific AMR document and the generic BAW-2248A documents were completed to facilitate plant license renewal in accordance with 10 CFR Part 54 [20].1.8.2 MRP-227-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 NEI 03-08 [2] protocols.

The MRP-227-A

[3] guidelines include "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.ANO-1 Applicability:

MRP-227 was officially issued by the industry in December 2008 [21].An aging management program was to be developed by December 2011. An aging management program plan for the ANO- 1 reactor vessel internals was originally developed to meet this"Mandatory" requirement (WCAP- 17495 [22]). This document supersedes the original aging management program plan for the ANO-l reactor vessel internals.

Report No. 1200459.401.Rl 1-12 rStructural Integrity Associates, InO Inegit1soiaesI1~

Needed 1. Each commercial U.S. PWR unit shall implement Tables 4-1 through 4-9 and Tables 5-1 through 5-3 of MRP-227-A for the applicable design within 24 months following issuance of MRP-227-A.

ANO-1 Applicability:

MRP-227 augmented inspections will be incorporated in the ANO-1 ISI program for the license renewal period. The applicable B&W tables contained in MRP-227-A components are Table 4-1 (Primary) and Table 4-4 (Expansion) and are attached herein as Table 5-1 and Table 5-2, 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.ANO-1 Applicability:

Inspection standards will be in accordance with the requirements of MRP-228 [11]. These inspection standards will be used in augmented inspections at ANO-1 as applicable where required by MRP-227-A.

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

ANO-1 Applicability:

ANO-1 will comply with this requirement.

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 completion of an outage during which PWR internals within the scope of MRP-22 7-A are examined.ANO-1 Applicability:

ANO-I will comply with this requirement.

Report No. 1200459.401 .Rl 1-13 jstructural Integrity Associates, Inc!

5. If an engineering evaluation is used to disposition an examination result that does not meet the examination acceptance criteria in Section 5 of MUP-227-A, this engineering evaluation shall be conducted in accordance with an NRC-approved evaluation methodology.

ANO-1 Applicability:

ANO-I will comply with this requirement by using NRC-approved evaluation methodology (e.g. WCAP- 17096 [13]).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.M16A (provided in Section 2.3 of this document)1.8.4.1 MRP-227-A Applicability to ANO-1 The applicability of MRP-227-A to ANO- 1 requires compliance with the following MRP-227 assumptions: " Operation of 30 years or less with high-leakage core loading patterns (firesh ftel assemblies loaded in peripheral locations) followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation.

Applicability:

ANO-1 historic core management practices meet the requirements of MRP-227-A

[23, 24].* Base load operation, i.e., typically operates atfixedpower levels and does not usually vary power on a calendar or load demand schedule.Applicability:

ANO- I operates as a base load unit [25].Report No. 1200459.401 .R1 1-14 I Structural Integrity Associates, Inc!

• 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.

ANO-1 has not made any modifications of the RVI components beyond those identified in general industry guidance or recommended by the vendor (B&W) since the May 2007 effective date of this statement, and therefore meets this requirement of MRP-227-A.

Further discussion on modifications that were made based on general industry guidance or recommended by the vendor (B&W) is provided in Section 3.2 (ANO-1 Design Modifications and Distinctions).

1. 8.5 Ongoing Industry Programs ENO actively participates in 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 susceptible to aging mechanisms of concern. Additionally, this report identifies appropriate inspections or mitigation programs needed to manage the aging mechanisms of the reactor vessel internals to provide assurance that these components will maintain their functionality through the period of extended operation.

The NRC has reviewed the ANO-1 LRA [5] and their approval is documented in Safety Evaluation Report [19].The ANO- 1 RVI AMP has been created to address the reactor vessel internals aging concerns consistent with the information identified in Revision 2 of the GALL Report [7], the guidance provided in MRP-227-A

[3], and the SE to MRP-227 issued by the NRC. ANO-1 will manage their RVI inspections through their augmented ISI program and will complete any repairs and/or Report No. 1200459.401 .Rl 1.-15 rStructural Integrity Associates, M0 replacements in accordance with ASME Code requirements and any NRC approved methodologies.

The ANO-1 RVI AMP will be updated accordingly as operating experiences and new inspection requirements and technologies evolve associated with managing reactor vessel internals aging concerns.Report No. 1200459.401 .R1 1-16 rStructural Integrity Associates, Inc 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, maintaining integrity of the fuel and pressure vessel boundary.

The core support structures provide support, guidance, and protection for the reactor core, provide a passageway for the distribution of the reactor coolant flow to the reactor core, provide a passageway for support, guidance and protection for control elements and in-vessel/

core instrumentation, and provide gamma and neutron shielding for the reactor vessel.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.2.1.2 Irradiation-Assisted Stress Corrosion Cracking Irradiation-assisted stress corrosion cracking (IASCC) is a unique form of SCC that occurs only in highly-irradiated components.

The aging effect is cracking.Repor No.120049.40 .RlStructural Integrity Associates, Inc?Report No. 1200459.401 .R1 2-1 trtua 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.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 Report No. 1200459.401 .Rl 2-2 C Structural Integrity Associates, InO 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.

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 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. While the initial aging effect is dimensional change and distortion, severe void Report No. 1200459.401 .R1 2-3 fStructural Integrity Associates, Inc!

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 program for ANO-1. 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 the results of the functionality assessment with component accessibility, Report No. 1200459.401 .Rl 2-4 VStructural Integrity Associates, Inc!

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 for 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 or which no highly susceptible component is accessible.

• Expansion:

Those PWR internals that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components will depend on the findings from the examination of the Primary components at individual plants.Report No. 1200459.401 .Rl 2-5 C Structural Integrity Associates, Inc

• Existin2 Proprams:

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 ANO-1 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. 1200459.401 .R1 2-6 V Structural Integrity Associates, Inc?

EPRI 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

[10, 12]: " 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 ANO-1 Reactor Vessel Internals Aging Management Program Attributes The attributes of the ANO-1 RVI AMP and compliance comparison with the ten elements of NUREG- 1801 (GALL Report), Revision 2, Chapter XI.M 16A, "PWR Vessel Internals" [7] are included in this section to ensure successful management of component aging.Report No. 1200459.401 .R1 2-7 ?Structural Integrity Associates, Inc InegnysoiteI, 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 ANO-1 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 Arkansas Nuclear One, Unit 1, which is built to a B&WNSSS design. The scope of the program applies the methodology and guidance in the most recently NRC-endorsed version of MRP-22 7, which provides augmented inspection and flaw evaluation methodology for assuring the finctional 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 MRP-22 7 guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code, Section XW), those R VI components that serv'e an intended license renewal safetyfiunction pursuant to criteria in 10 CFR 54.4(a)(1), and other R VI components whose failure could prevent satisfactoty accomplishment of any of the finctions identified in 10 CFR 54.4(a) (1) (i), (ii), or (iii). The scope of the program does not include consumable items, such as fiel 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 1 appurtenances to the reactor vessel and are adequately managed in accordance with an applicant's AMP that corresponds to GALL AMP XI.M1, "ASAME 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-227 methodology, and any additional programs, actions, or activities that are discussed in these LRAAI responses and credited Report No. 1200459.401 .R1 2-8 dj:Structural Integrity Associates, Inc for aging management of the applicant's R VI components.

The LRAAIs are identified in the staff's safet 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 MiRP-22 7), and NSSS vendor-specific or plant-specific LRAAIs as well. The responses to the LRAAIs on MRP-227 are provided in Appendix C of the LRA. ""The guidance in MRP-227 specifies applicability limitations to base-loaded plants and the fuel loading management assumptions upon which the finctionality 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 ANO-1 Program Scope A description of the RVI design is provided in Section 3.0 of this program plan. Additional details regarding the RVI are provided in the ANO-1 UFSAR [6]. The ANO-I RVI subcomponents that require aging management review are indicated in the ANO-1 LRA [5].Table 3.2-1 and Section 3.5 in Appendix B of the ANO-1 LRA provide a summary of the aging effects requiring management for the reactor vessel internals.

A review of the ANO-1 RVI components was performed as part of this document by comparing the components that screened-in (as "Category B" or "Category C") for the aging mechanisms that affect RVI from Table 5-1 of MRP-189, Revision 1 [10] with Table 3-1, Table 4-1 and Table 4-4 of MRP-227-A[3] and the aging management review (AMR) [18] that was performed as part of the ANO-1 LRA [5]. The purpose of this AMR is to identify components that are listed as "Category B" and"Category C" in MRP- 189, Revision 1 [ 10] and ensure that the effects of aging for these RVI components were adequately managed. This review only looked at the corresponding components.

The weld materials listed in MRP- 189, Revision 1 were not considered in this review. The RVI components that screened in as Categories "B" and "C" were compared to the Primary and Expansion category components in MRP-227-A

[3] and the aging management review tables that were developed as part of the ANO-1 license renewal application

[5]. The effects of aging for the RVI components identified in Reference

[5] are managed by the ASME Report No. 1200459.401 .R1 2-9 V Structural Integrity Associates, Inc Section XI ISI Program and MRP-227-A.

The results of this review are shown in Appendix B and Table B-I includes a summary of the results. This table identifies the aging effects that require management.

A column in the table lists the programs and activities at ANO-1 that are credited to address the aging effects for each management strategy presented in Appendix B.The aging management of cast austenitic stainless steel (CASS) RVI components is presented in this program plan as part of the MRP-227-A

[3] with additional screening, functional evaluations, and possibly augmented inspections to manage component aging resulting from loss of fracture toughness due to thermal aging and/or neutron irradiation embrittlement.

The CASS components are the internal vent valve bodies, the CRGT assembly spacer castings, and the incore monitoring instrument guide tube assembly spider castings.

The vent valve retaining rings are fabricated from precipitation-hardened stainless steel, which is also susceptible to thermal aging and neutron embrittlement.

Visual inspections under this program are described in MRP-228 [11].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 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-227-A (summarized in Tables 5-1 through 5-3 of this AMP). Guidance provided in MRP-227-A

[3] 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 ANO-1 RVI AMP. In addition, plant specific existing programs such as the Section XI ISI program for ANO-1 will complement the augmented inspection requirements provided in MRP-227-A in successfully managing the effects of aging for ANO- 1 during the period of extended operation.

Report No. 1200459.401 .R1 2-10 V Structural Integrity Associates, Inc?Inegit-socatsnc 2.3.1.2 Conclusion The ANO-1 program implements the corresponding aging management attribute in Revision 2 of NUREG-1801

[7], Chapter XI.M16A, commitments made in the ANO-1 LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9]. This information has also been updated in the ANO- I Safety Analysis Report [6].2.3.2 NUREG-1801/AMP Program Element 2: Preventive Actions"The guidance in MRP-22 7 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 ANO-1 Preventive Action The ANO-1 RVI AMP credits the following existing programs that comply with the requirement of this element. A description of the applicability to the ANO-1 RVI AMP is provided in the following subsection.

2.3.2.2 Primary Water Chemistry Program The primary goal of this program is to mitigate loss of material due to general, pitting, crevice corrosion, and 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 ANO-1 water chemistry program [16] is based on current, approved revisions of EPRI PWR Water Chemistry Guidelines

[ 17].Report No. 1200459.401 .Rl 2-11 rStructural Integrity Associates, Inc!

The limits of known detrimental contaminants imposed by the primary water chemistry program are consistent with the EPRI PWR Primary Water Chemistry Guidelines

[17].2.3.2.3 Conclusion The ANO-1 program implements the corresponding aging management attribute in Revision 2 of NUREG- 1801 [7], Chapter XI.M 16A, commitments made in the ANO- l LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9].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 the R VI components at the facility: (a)cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture 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 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 Report No. 1200459.401 .R1 2-12 VStructural Integrity Associates, Inc?

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 MP-22 7 guidance or ASME Code, Section M 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 B& W designed Primary Components in Table 4-1 of MRP-227. Additionally, the program implements the parameters monitored/inspected criteria for B& W designed Expansion Components in Table 4-4 of MRP-22 7. The parameters monitored/inspected fior Existing Program Components follow the bases for referenced Existing Programs, such as the requirements of the ASME Code Class RVI components in ASME Code, Section Al, Table IWB-2500-1, Examination Categories B-N-3, as implemented through the applicant's ASMiE 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 X, are required for components that are identified as requiring "No Additional Measures," in accordance with the analyses reported in MRP-227." 2.3.3.1 ANO-1 Parameters Monitored/Inspected ANO- 1 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 inspection requirements under industry directives as contained in MRP-227-A

[3] and ASME Section XI [14].2.3.3.2 Conclusion The ANO- 1 program implements the corresponding aging management attribute in Revision 2 of NUREG-1801

[7], Chapter XI.M 16A, commitments made in the ANO-l LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9].Report No. 1200459.401 .R1 2-13 r Structural Integrity Associates, Inc!

2.3.4 NUREG-1801/AMP Program Element 4: Detection ofAging Effects"The detection of aging effects is covered in two places: (a) the guidance in Section 4 of MRP-22 7 provides an introductory discussion and justification of the examination methods selected for detecting the aging effects of interest; and (b) standards for examination methods, procedures, and personnel are provided in a companion document, MRP-228. In all cases, well 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 suface-breaking discontinuities.

Sutface examinations may also be used as an alternative to visual examinations jbr detecting and sizing of surface breaking discontinuities.""Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 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 when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is klnown 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-22 7for defining the Expansion criteria that need to be applied to inspections of Primary Components and Existing Requirement Components and jbr expanding the examinations to include additional Expansion Components.

As a result, inspections performed on the R VI Report No. 1200459.401.R1 2-14 $Structural Integrity Associates, Inc?"a,"iysocaesIc components are peijbrmed 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 B& W designed Primary Components in Table 4-1 of AMP-227 and for B& W designed expansion components in Table 4-4 of MRP-227. ""The program is supplemented by thefollowing plant-specific Primary Component and Expansion Component inspections for the program (as applicable):

for the ANO-1 program, there are no supplemental plant-specific Primary or Expansion components.

However, ANO-1 specific components (RVLMS probe supports, surveillance specimen holder tubes, and thermal shield and thermal shield upper restraint) that were outside the scope of BA W-2248A were identified as part of the ANO-1 LRA as components that are within the scope of the aging management review. Augmented inspections of these components are addressed as part of the ANO-1 plant specific programs.In addition, in some cases (as defined in MRP-22 7), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss ofpreload due to stress relaxation, or for changes in dimension due to void swelling, deflection or distortion.

The physical measurements method applied in accordance with this program include.-(i) One-time measurement of the differential height of the top of the plenum rib pads to reactor vessel seating sumface, with plenum in the reactor vessel with the fiel assemblies removed.(ii) One-time physical measurement of the core clamping region to evaluate if wear is a degradation mechanism that is operative.

The purpose of the clamping is to No. 1200459.401.RI 2-15 VStructural Integrity Associates, Inc!Report stabilize and significantly restrict rigid body pendulum motion of the core support assembly.

Wear at these locations will progress from motions generated by fluid flow once the loss of core clamping is initiated.

The one-time physical examination is to beiollowed by subsequent visual (VT-3) examination.

2.3.4.1 ANO- 1 Detection ofAging Effects Detection of indications required by the ASME Section XI ISI Program is well-established and field-proven through application of the ASME Section XI ISI Program. Augmented inspections taken from the MRP-227-A recommendations will be applied through use of MRP-228 [11]Inspection Standard.Inspections 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 RVI, as contained in this program, are built around the following 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 [11]. 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 The ANO-1 program implements the corresponding aging management attribute in Revision 2 of NUREG-1801

[7], Chapter XI.M16A, commitments made in the ANO-1 LRA and the action items identified in the ANO-l License Renewal SER to BAW-2248A

[9].Report No. 1200459.401 .Rl 2-16 N Structural Integrity Associates, lnc?InegiysoiteIc 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 AMPf-227 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 MP-22 7 guidance, together with the requirements specified in MRfP-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 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 ANO-1 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 the test data from RVI components removed from service. This data is used to identify trends in materials degradation and forecast potential component degradation.

The industry continues to share both material test data and operating experience through the auspices of the EPRI MRP and PWROG. ANO-1 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 of reactor vessel internal components, where practical, are scheduled to be conducted in conjunction with typical 10-year ISI examinations.

Report No. 1200459.401 .Rl 2-17 r Structural Integrity Associates, Inc?

Inspections performed at ANO- 1 as part of the ISI program are provided in Appendix A. Tables 5-1 and 5-2 of this document 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 results will allow the industry to monitor and trend results and take appropriate preemptive action through update of the EPRI MRP guidelines.

2.3.5.2 Conclusion The ANO-1 program implements the corresponding aging management attribute in Revision 2 of NUREG-1801

[7], Chapter XI.M 16A, commitments made in the ANO-1 LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9].2.3.6 NUREG-1801/AMP Program Element 6: Acceptance Criteria"Section 5 of MRP-22 7 provides specific examination acceptance criteria for the Primaty and Expansion Component examinations.

For components addressed by examinations referenced to ASME Code, Section M, 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-22 7 contains three types of examination acceptance criteria: Report No. 1200459.40 L.R1 2-18 V Structural Integrity Associates, lnc Inegiy1soiteIc

• For visual examination (and sutface 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-J/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 of bolted or pinned assemblies with unacceptable volumetric (UT)examination indications that exceed specified limits; and* Forphysical measurements, the examination acceptance criterion for the acceptable tolerance is measured in the differential height fiom the top of the plenum ribs to the vessel seating surface in B& Wplants are given in Table 5-1 of MRP-22 7.2.3.6.1 ANO-1 Acceptance Criteria Recordable indications that are the result of inspections required by the ANO-1 existing ISI program [ 15] are evaluated in accordance with the requirements of the ASME Code and documented in the ANO-1 Corrective Action Process [27].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 ANO-1 RVI will be included as part of updates to the AMP.Recordable indications found during the MRP-227-A augmented inspections will be entered into the ANO- 1 corrective 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 the Report No. 1200459.401 .R1 2-19 jstructural Integrity Associates, 1nc Inegiy4soiteIc approved methodologies to support this element. Additional analysis to establish evaluation acceptance criteria for "Expansion" category components has been developed by the PWROG and published in WCAP-17096-NP

[13]. The status of these ongoing processes is monitored via ENO's participation in various industry programs related to aging management of PWR internals.

2.3.6.2 Conclusion The ANO-1 RVI program implements the corresponding aging management attribute in Revision 2 of NUREG- 1801 [7], Chapter XI.M 16A, commitments made in the ANO- 1 LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9].2.3.7 NUREG-1801/AMP Program Element 7: Corrective Actions"Corrective actions following the detection of unacceptable conditions are findamentally 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 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 finctions of the reactor internals components will continue to be fiufilled for all licensing basis loads and events. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code, Section XJ or in Section 6 of MRP-227. Section 6 ofMRP-227 describes the options that are available for disposition of detected conditions that exceed the examination acceptance criteria of Section 5 of the report. These include engineering evaluation methods, as well as supplementaty 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 Report No. 1200459.401.R1 2-20 V Structural Integrity Associates, 1=8 addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

""Other alternative corrective action bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Examples of previously NRC-endorsed alternative corrective action bases include those corrective actions bases for Westinghouse-design R VI components that are identified in Tables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghouse Report No. WCAP-145 77-Rev. 1-A, or for B&W-designed R VI components in B& W Report No. BA W-2248A. Westinghouse Report No.WCAP-14577-Rev.

1-A was endorsed for use in an NRC SE to the Westinghouse Owners Group, dated February 10, 2001. B& W Report No. BA W-2248A was endorsed for use in an SE to Framatome Technologies on behalf of the B& W Owners Group, dated December 9, 1999. Alternative corrective action bases not approved or endorsed by the NRC will be submitted for NRC approval prior to their implementation." 2.3.7.1 ANO-1 Corrective Actions The ANO-1 Corrective Action Process [27] addresses this element of the GALL attributes.

Indications that require repair and replacement will be addressed through the ANO- 1 corrective 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 [14]. The corrective actions for existing Section XI (B-N-3) examinations will include the identification of a repair and verification of acceptability of replacements.

Any indications found during the Section XI examinations for the RVI will be documented in the correction action program. This evaluation guidance is included in MRP-227-A and WCAP-17096-NP.

For example, the guidance provided in 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.Report No. 1200459.401 .R1 2-21 r Structural Integrity Associates, Inc?

2.3.7.2 Conclusion The ANO-1 program implements the corresponding aging management attribute in Revision 2 of NUREG-1801

[7], Chapter XI.M16A, commitments made in the ANO-1 LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9].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-22 7 will provide an acceptable level of quality for inspection, flaw evaltation, and other elements of aging management of the PWR internals that are addressed in accordance with 10 CFR Part 50, Appendix B or their equivalent (as applicable), confirmation process, and administrative controls." 2.3.8.1 ANO-1 Confirmation Process The ANO-1 RVI Aging Management Program meets the "Mandatory" and "Needed" requirements under NEI 03-08 [2]. This program conforms to the ENO 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 ANO-1 Section XI Inservice 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 [28].2.3.8.2 Conclusion The ANO- 1 program implements the corresponding aging management attribute in Revision 2 of NUREG-1801

[7], Chapter XI.M16A, commitments made in the ANO-1 LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9].Report No. 1200459.401 .Rl 2-22 L Structural Integrity Associates, Inc!InertysoitenV 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 ANO-1 Administrative Controls ANO- 1 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 The ANO-1 program implements the corresponding aging management attribute in Revision 2 of NUREG- 1801 [7], Chapter XI.M 1 6A, commitments made in the ANO- 1 LRA and the action items identified in the ANO-l License Renewal SER to BAW-2248A

[9].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 summaty 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 thisfiunction.

The application of MRP-22 7 guidance will establish a considerable amount of operating experience over the next few years. Section 7 of MRP-227 describes the reporting Report No. 1200459.401.R1 2-23 V Structural Integrity Associates, Inc!

requirements for these applications, and the plan for evaluating the accumulated additional operating experience." 2.3.10.1 ANO- 1 Operating Experience Industry and ANO-1 operating experience (OE) has been reviewed during the development of the ANO-1 RVI AMP.Industry and ANO- 1 specific information relevant to aging has been compiled into the ANO- 1 OE program [29]. Industry operating experience sources in this program include 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), corrective action process documents (CAPs, CRs, DRs, ERs), etc.A review of the industry operating experience (OE) revealed several instances of cracked bolts or lost bolts in reactor vessel internals.

An evaluation of ANO-1 OE is highlighted in this section of the AMP and the details of plant specific design modifications based on OE are provided in Section 3.2 of this document.

One of the ANO-1 control rod drive mechanisms (CRDMs) was removed and the control rod guide assembly in the plenum was modified to accept the reactor vessel level monitoring probe [5]. Portions of the surveillance specimen holder tubes (SSHT) at ANO-1 are attached to the internals.

Although all the specimens have been removed, portions of the shroud tube and the supports that are bolted to the core support shield remain. These items only have the function of preventing loose parts in the RCS. Removal of the SSHT was due to a design flaw and was considered not to be caused by aging related degradation

[5]. Based on the discussions in BAW-2248A

[9] and the review of ANO-1 operating data using station information management system, condition reporting system, and licensee event database, cracking of the thermal shield bolting and core barrel bolting fabricated from Alloy A-286 was identified as an issue. These failures were attributed to intergranular stress corrosion cracking (IGSCC), and were not detected by visual examinations.

In addition, during recent inspection of RVI at Oconee Nuclear Station, Unit 1, a failed vent valve jack screw was identified

[32]. Other Report No. 1200459.401 .R1 2-24 W Structural Integrity Associates, Inc Ineriysscatsin.

than these few failures, the review of operating experiences did not identify any new aging issues related to the ANO-1 RVI [5, 19].Inspections performed as part of the 10-year ISI program have been conducted as designed by existing commitments and would be expected to discover general internals structure degradation.

Industry OE is routinely reviewed by the Institute of Nuclear Power Operations (INPO) and other informational sources, as directed under the applicable procedure for the determination of additional actions and lessons learned. These insights, as applicable, can be incorporated into the plant system health reports and further evaluated for incorporation into the applicable plant programs.A review of industry and plant-specific experience with RVI reveals that the U.S. Nuclear fleet (including ANO-1) has responded proactively to issues related to degradation of RVI by participation in the PWROG and EPRI MRP.2.3.10.2 Conclusion The ANO-1 program implements the corresponding aging management attribute in Revision 2 of NUREG- 1801 [7], Chapter XI.M I6A, commitments made in the ANO- I LRA and the action items identified in the ANO-1 License Renewal SER to BAW-2248A

[9].Report No. 1200459.40 L.R 1 2-25 Structural Integrity Associates, Inc?

3.0 ANO-1 REACTOR VESSEL INTERNALS DESIGN AND OPERATING EXPERIENCE The ANO- 1 RPV and internals were designed by Babcock & Wilcox. The ANO- 1 reactor vessel internals were designed to ASME Section III, 1965 Edition with Addenda through and including the 1967 Summer Addenda [ 18]. The ANO- I design consists of two major structural assemblies that are located within, but are not welded to the reactor vessel. These two major assemblies are the plenum assembly and core support assembly (CSA). The latter includes three principal sub-assemblies

-the core support shield (CSS) assembly, the core barrel assembly, and the lower internals assembly [3, 9]. The general arrangement of the ANO-1 reactor vessel internals is shown in Figure 3-1. Schematic representations of specific reactor vessel internals components and assemblies are provided in Figure 3-2 through Figure 3-12.3.1 Description of ANO-1 Reactor Vessel Internals 3.1.1 Plenum Assembly The plenum assembly is a cylindrical structure with perforated grid plates on top and bottom, and is comprised of: (1) the plenum cover assembly; (2) the plenum cylinder assembly; (3) the upper grid assembly; and (4) the control rod guide tube assemblies, as shown in Figure 3-2. The plenum assembly fits inside the core support shield, positions the top of the fuel assemblies, supports the control rod guide tube assemblies, and provides the core hold-down required for hydraulic lift forces. The plenum assembly also provides continuous guidance and protection for the control rods, and directs flow out of the core to reactor vessel outlet nozzles. The plenum assembly is removed at the beginning of every refueling outage, in order to permit access to the fuel assemblies

[3, 9].3.1.2 Plenum Cover Assembly The plenum cover assembly is bolted to the top of the plenum cylinder, and consists of a weldment, a bottom flange, a support ring and flange, and lifting lugs. The weldment is a series Report No. 1200459.401.Rl 3-1 r Structural Integrity Associates, Inc!

of parallel flat plates (ribs) assembled to form a square lattice. The rings and flanges provide the vertical and horizontal seating surface for the plenum cover and assembly.

The cover assembly is bolted to the plenum cylinder top flange. The perforated top plate has matching holes to position the upper end of the 69 CRGT assemblies.

The plenum cover assembly provides support for the top of the control rod guide tube assemblies.

The lifting lugs are used to lift the plenum assembly out of the reactor vessel. A schematic representation of the plenum cover assembly is shown in Figure 3-3 [9].3.1.2.1 Plenum Cover Weldment The plenum cover weldment is a lattice construction assembled from two sets of ten parallel flat plates intersecting perpendicularly with 10-inch spacing between ribs. The individual ribs are 2-inch thick flat stainless steel plates of varying lengths and heights. Small rib or compression pads are welded to the top outer edge of each rib where it mates with the plenum ring, forming a mating surface on which the reactor vessel head sits [9].3.1.2.2 Plenum Cover Bottom Flange The plenum cover bottom flange is a flat ring welded to the bottom of the weldment to provide a surface to attach the plenum cover to the plenum cylinder.

It is smaller and located inside of the plenum cover support flange. The bottom flange has 64 tapped holes to which the upper flange of the plenum cylinder is bolted [9].3.1.2.3 Plenum Cover Support Flange The plenum cover support flange is also welded to the bottom of plenum cover weldment assembly.

It provides the seating surface that rests on the top of the core barrel shield assembly and against the inner RV wall. At each of the four axes locations, the support flange has keyways which mate with reactor vessel flange keys to align the plenum assembly with the reactor vessel, the reactor closure head control rod drive penetrations, and the CSA [9].Report No. 1200459.401 .Rl 3-2 r $StrItctural Integrity Associates, Inc Inegiy:soiteIc 3.1.2.4 Plenum Cover Support Ring The plenum cover support ring is an 81/2-inch high, 2-inch thick ring welded onto the top of the support flange and outer vertical edges of the plenum cover weldment.

The support ring provides a surface which mates with the reactor vessel head. At each of the four axes locations, the support ring has keyways which mate with reactor vessel flange keys to align the plenum assembly [9].3.1.2.5 Lifting Lugs Three lifting lugs are located about the plenum cover assembly to be used when removing the plenum assembly during refueling outages. ANO- 1 has T-shaped lifting lugs that are fastened with two 1%A-inch diameter bolts to the base blocks. These bolts are secured with locking cups.The base blocks are welded between two of the weldment ribs [9].3.1.3 Plenum Cylinder Assembly The plenum cylinder assembly is bolted to the bottom of the plenum cover assembly and consists of a cylinder, top and bottom flanges, reinforcing plates, and round bars (Figure 3-4). Its primary function is to direct the flow of reactor coolant from the core region to the reactor vessel outlet nozzles [3, 9].3.1.3.1 Plenum Cylinder The plenum cylinder is a large 7-foot high cylinder made from 1 /2-inch thick stainless steel plate. Flanges for connecting the cylinder to the plenum cover and the upper grid are welded to the plenum cylinder ends. Two reinforcing plates are welded to the lower inner surface opposite to the core support shield outlet nozzles. Both the reinforcing plates and the plenum cylinder have twenty-four small holes to permit some of the reactor coolant coming up into the plenum to flow directly to the outlet nozzles. Ten large holes (six 34 inches in diameter and four 22 inches in diameter) at the top of the cylinder let the majority of the reactor coolant pass out into the Report No. 1200459.401 .RI 3-3 Structural Integrity Associates, lnc.

annulus between the plenum cylinder and the core support shield and ultimately down and out through the outlet nozzles [9].3.1.3.2 Top Flange The plenum top flange is welded to the top of the plenum cylinder.

The plenum cover assembly is bolted to the upper flange with sixty-four 1 1/8-inch large hex head bolts held in place with locking cups [9].3.1.3.3 Bottom Flange The plenum bottom flange is welded to the bottom of the plenum cylinder.

The plenum upper grid assembly is bolted to the bottom flange with thirty-six 1-inch diameter large hex head bolts held in place with locking cups [9].3.1.3.4 Reinforcing Plates The two 3-inch thick reinforcing plates are welded to the inner surface of the plenum cylinder.Twenty-four holes permit some RCS flow directly out from the plenum area to the outlet nozzles[9].3.1.3.5 Round Bars In order to insure that the space between the plenum cylinder and the core support shield does not collapse and prevent flow through the outlet nozzles, a set of 13 small stainless steel round bars or lugs are welded to the outer surface of the plenum cylinder at each of the outlet nozzle areas. These lugs fit against similar lugs welded to the inner surface of the core support shield.The round bars are 4 inches long and 21/2/2 inches in diameter [9].Report No. 1200459.401.Rl 3-4 :j§ Itructural Integrity Associates, Inc!

3.1.4 Upper Grid Assembly The upper grid assembly sits inside the lower flange of the core support shield and is bolted to the plenum cylinder bottom flange. It is comprised of an upper grid ring forging, an upper grid rib section, and fuel assembly support pads, as shown in Figure 3-5. Its function is to support and provide a seating surface for the top of the fuel assemblies located within the core barrel below, and to restrain and align the bottoms of the control rod guide tubes [9].3.1.4.1 Upper Grid Ring Forging The upper grid ring forging is a ring with an inward flange on the upper end. The top of the upper grid ring forging is machined to accept the thirty-six 1-inch bolts fastening the upper grid assembly to the plenum cylinder bottom flange. The upper grid rib section is bolted to the bottom of the ring assembly with thirty-six 3/4A-inch diameter screws held in place with welded lockpins [9].3.1.4.2 Upper Grid Rib Section The upper grid rib section is a 3-inch thick, 136-inch diameter disk, with 177 squares machined out, leaving a grid with 1-inch wide ribs. The square holes align with the fuel assembly locations in the core below. Pads to support and align the fuel assemblies are doweled and bolted into the ribs on the bottom side. The topside of the rib section is drilled and tapped to accept the dowels and screws, which hold the bottom flange of the 69 control rod guide tube (CRGT) assemblies to the upper grid [9].3.1.4.3 Fuel Assembly Support Pads There are 384 small 2-inch high pads attached to the bottom of the grid rib section to provide a seating surface and support for the tops of the fuel assemblies.

The pads are each held in place by two Alloy X-750 dowels and a 1/4-inch diameter cap screw. The dowels and cap screws are all welded in place [9].Report No. 1200459.401 .Rl 3-5 Structuralegrt Associates, Inc 3.1.5 Control Rod Guide Tube (CRGT) Assemblies The control rod guide tube assemblies each consist of a pipe (the guide housing), a flange, spacer castings, guide tubes, and rod guide sectors. The assemblies are welded to the plenum cover plate and bolted to the upper grid assembly.

Their function is to provide control rod assembly guidance, protect the control rod assembly from the effects of potential coolant cross-flow, and structurally connect the upper grid assembly to the plenum cover [9].The end of each of the 69 control rod assemblies consists of a spider plate through which 16 individual control rods are suspended.

The 139-inch long control rods are arranged in two concentric rings, four rods in the middle ring and twelve in the outer ring, as shown in Figure 3-6. The rods have no other support other than the spider head at the top. The CRGT assemblies provide support both for the CRGT assemblies as a whole and for each of the 16 individual control rods within each control rod assembly (CRA), as well as accommodating the control rod spider that travels the entire length of the CRGT assembly [9].The outer portion of the CRGT assemblies consists of tall pipes (or guide housings) welded to the CRGT assembly flanges at their bottoms. The insides of each CRGT assembly consists of an internal sub-assembly with ten parallel horizontal spacer castings to which are brazed 12 perforated vertical rod guide tubes and 4 pairs of vertical rod tube guide sectors, also called C-tubes. These internal sub-assemblies of spacers, rod guide tubes and rod guide sectors are referred to as the rod guide brazements.

Figure 3-7 shows the CRGT assembly spacer castings and the rod guide brazement configuration

[9].3.1.5.1 CRGT Assembly Pipes The CRGT assembly pipes (or guide housings) are approximately 12-foot tall, 8-inch diameter, Schedule 40 stainless steel pipes. At ten elevations, they are drilled at 4 equally spaced locations to accommodate the 3/8-inch diameter cap screws that hold the CRGT assembly spacer castings in place [9].Report No. 1200459.401 .Rl 3-6 Structurale Four equally spaced 3-inch diameter holes are located 2-inches from the bottom of the CRTG assembly pipes. Above them are two rows of four 3-inch wide, 8%A-inch high oval shaped holes.These holes allow some of the reactor coolant traveling up the CRGT assembly pipes to exit out into the plenum and to ensure that the pressures are equalized on both sides of the CRGT assembly pipes and prevent hydraulic effects from impeding control rod travel.The pipes are welded to the CRGT assembly flanges at the bottom, and to the top of the plenum cover plate. The top of the CRGT assembly pipes extend 21 1/8-inches above the plenum cover plate into the upper head area. The CRGT assembly pipes are shown in Figures 3-2 and 3-8 [9].3.1.5.2 CRGT Assembly Flanges The CRGT assembly flanges are 11/4- inch thick, 41/4-inch sided square plates with a hole in the center to match the inner diameter of the CRGT assembly pipes. Four additional small semicircular flow paths are equally spaced about the center to permit RCS flow upward through the flanges on the outside of the CRGT assembly pipe. Each flange is drilled to accept two 1/2-inch diameter dowels and four '/2-inch diameter hex cap screws which are torque into the upper grid rib section [9].3.1.5.3 CRGT Assembly Spacer Castings The CRGT assembly spacer castings are 3/4-inch thick disks, with internal spaces to conform to the general shape of the control rod spider, with margins to permit RCS flow and to accommodate the rod guide tubes and rod guide sectors [9].3.1.5.4 CRGTAssembly Rod Guide Tubes Within each CRGT assembly are 12 guide tubes. These are long 0.750-inch ID, 0.095-inch thick tubes with a 5/16-inch wide vertical slot. The tubes have a vertical row of '/4-inch holes to permit RCS flow into the area. The CRGT assembly guide tubes are brazed into holes in the CRGT Report No. 1200459.401.R1 3-7 Structural Integrity Associates, Inc?

assembly spacer castings, with the slots aligned to match where the control rod assembly spider arms pass [9].3.1.5.5 CRGTAssembly Rod Guide Sectors The CRGT assembly rod guide sectors are similar to the CRGT assembly rod guide tubes, however, they are for the 4 inner individual control rods in each assembly which are suspended from the middle of a spider arm and thus require slots on both sides of the tubes. They are long, 0.109-inch thick plates with a curved cross section. They are brazed in pairs in holes in the CRGT assembly spacer castings, facing each other with a gap between them to permit travel of the spider arm between them. The rod guide sectors do not have cooling holes like the rod guide tubes, since they are open on two sides [9].3.1.6 Core Support Shield Assembly (CSS)The CSS is a large flanged cylinder, which sits on top of the core barrel and fuel assemblies.

The CSS provides a boundary between the incoming cold reactor coolant, which is directed downward on the outside of the CSS, and the heated reactor coolant flowing upward from the core area into the plenum area and out through the reactor outlet nozzles, bounded by the inside core support shield wall and nozzles. The top flange of the CSS rests on and is supported by a circumferential ledge in the reactor vessel closure flange. This provides support for the entire reactor vessel internals assembly [3, 9].The plenum assembly is supported by and fits inside the CSS. The bottom flange of the CSS is bolted to the core barrel. The inside surface of the core support shield bottom flange provides the lower seating surface and aligns the plenum assembly.The core support shield cylinder wall has two openings with nozzles for RCS outlet flow. These openings are formed by two forged rings which seal the reactor vessel outlet nozzles by the differential thermal expansion between the CSS and the low alloy steel reactor vessel. The nozzle seal surfaces are finished and fitted to a predetermined cold gap providing clearance for Report No. 1200459.40 l.R1 3-8 Vstructural Integrity Associates, Inc!

CSA installation and removal. At operating temperature, the mating metal surfaces are in.contact to make a seal without exceeding allowable stresses in either the reactor vessel or internals

[9].Eight vent valve mounting rings are welded in the cylinder wall. Internals vent valves are installed in the CSS cylinder wall to allow steam flow from the core should a cold leg (reactor coolant inlet) pipe rupture occur [9].The core support assembly is fabricated by bolting together the core support shield assembly, the core barrel assembly, and the lower internals assembly to form a tall cylinder.

The core support assembly remains in place in the reactor vessel during refueling, and is removed only to perform scheduled inspections of the reactor vessel interior surfaces or of the core support assembly itself[3, 9].The top portion of the core support assembly is the core support shield assembly, a cylinder with an upper flange that rests on a circumferential support ledge in the reactor vessel closure flange, thereby supporting the entire core support assembly.

It sits directly on top of the core barrel, and consists of a cylinder, top and bottom flanges, outlet nozzles, vent valve nozzles, vent valves, round bars, flow deflectors, and lifting lugs. Its function is to provide a boundary between the incoming cold reactor coolant on the outside of the cylinder and the heated reactor coolant flowing on the inside of the cylinder [3, 9].3.1.6.1 Core Support Shield Cylinder The core support shield cylinder is a 145-inch ID, approximately 7-foot high cylinder.

There are ten major penetrations through the core support shield cylinder:

two 67-inch OD core outlet nozzles and eight vent nozzles. Figure 3-9 represents the core support shield assembly.The core support shield top flange is welded to the top of the core support cylinder.

The core support shield top flange extends out from the inner diameter.

The bottom of the top flange rests on a circumferential ledge in the reactor vessel closure flange. The top of the flange provides the Report No. 1200459.401 .Rl 3-9 r structural Integrity Associates, Inc!

seating surface to support the bottom of the plenum cover support flange, and thus supports the entire plenum assembly.

The bottom of the top flange is penetrated in eight places by the vent nozzles [9].The core support shield bottom flange is welded to the bottom of the core support shield cylinder.

The bottom of the plenum assembly is guided by the inside surface of the bottom flange of the core support shield. The core support shield bottom flange is bolted to the top flange of the core barrel with one hundred and twenty 1%-inch diameter core barrel bolts, secured with locking clips or locking cups [9].3.1.6.2 Outlet Nozzles The two forged outlet nozzles are 67-inch OD, 83%-inch thick curved ring-shaped inserts. The outlet nozzles are welded into the support shield cylinder with full penetration welds (i.e., the inner surfaces are welded flush with the inner cylinder wall and extend out horizontally 8 inches towards the inner reactor vessel wall). The wall thickness of the nozzle tapers, with the inner hole having an oval shape [9].3,1.6.3 Vent Valve Nozzles The eight vent nozzles are flat ring-shaped inserts welded into the core barrel support cylinder to provide support for the internal vent valve sub-assemblies.

The nozzles are welded into the core barrel cylinder using full penetration welds with their inner edges matching the inner core cylinder wall (when viewed from above, the inner flat surfaces of the nozzles crosses the inside of the core support cylinder).

The nozzles are approximately 38-inch OD and are 61/4-inches thick in cross section. To accommodate the internal vent valves, the inner surfaces of the rings have lips and flanges [9].Two small guide blocks are welded to the top outside surface of each vent nozzle. The guide blocks are machined to provide a small triangular seating surface for the vent valve assemblies.

Report No. 1200459.401.Rl 3-10 jStructural Integrity Associates, Inc 3.1.6.4 Internal Vent Valves Eight internal vent valve assemblies are installed in the core support shield as shown in Figure 3-10. For all normal operating conditions, the vent valve is closed. In the event of a pipe rupture in the reactor vessel inlet pipe, the valve will open to permit steam generated in the core to flow directly into the break, and will permit the core to be flooded and adequately cooled after emergency core coolant has been supplied to the reactor vessel [9].Each valve assembly consists of a hinged disc, valve body with sealing surfaces, a split-retaining ring and fasteners, which retain and seal the perimeter of the valve assembly, and an alignment device to maintain the correct orientation of the valve assembly for hinged-disc operating.

Each valve assembly can be remotely handled as a unit for removal or installation.

Valve component parts, including the disc are designed to minimize the possibility of loss of parts to the coolant system, and all operating fasteners including a positive locking device. The hinged-disc includes a device for remote testing and verification of proper disc function.

The external side of the disc is contoured to absorb the impact load of the disc on the reactor vessel inside wall without transmitting excessive impact loads to the hinge parts as a result of a loss-of-coolant accident [9].The hinge assembly consists of a shaft, two valve body journal receptacles, two valve disc journal receptacles, and four fanged shaft journals (brushings).

Loose clearances are used between the shaft and journal inside diameters, and between the journal outside diameters and their receptacles.

The valve disc hinge journal contains integral exercise lugs for remote operation of the disc with the valve installed in the core support shield. The hinge assembly provides eight loose rotational clearances to minimize any possibility of impairment of free movement of the disc in service. In addition, the valve disc hinge loose clearances permit disc self-alignment so that the external differential pressure adjusts the disc seal face to the valve body seal face. This feature minimizes the possibility of increased leakage and pressure induced deflection loadings on the hinge parts in service. The hinge assembly and disc are active components and are exercised at each refueling outage to evaluate performance

[9].Report No. 1200459.401.Rl 3-11 :Structurai Integrity Associates, Inc?

3.1.6.5 Round Bars At each outlet nozzle area, thirteen round bars are located on the inner side of the core support shield to mate with the similar lugs welded on the outer side of the plenum cylinder.

The round bars insure that the two cylinders are kept apart so RCS flow is not disrupted under any condition

[9].3.1.6.6 Flow Detectors A baffle consisting of three 1-inch thick plates shaped to form an inverted "U" is welded to the outer side of the core support cylinder around the area opposite each of the four inlet (cold leg)nozzles. These baffles help divert the incoming flow downward to the bottom of the core, and minimize the upward flow that might damage the internal vent valve assemblies

[9].3.1.6.7 Lifting Lugs Three lifting lugs are welded on the inside of the core support shield top flange. These lugs permit lifting of the CSA out of the core when required, such as for vessel inspections

[9].3.1.7 Core Barrel Assembly (CRA)The core barrel assembly is a second flanged cylinder, with its top flange bolted to the bottom flange of the core support shield assembly and its bottom flange bolted to the top flange of the lower internals assembly.

The core barrel assembly consists of a cylinder, top and bottom flanges, baffle and former plates, and a thermal shield cylinder.

Its functions are to direct the flow of coolant and to support the lower internals assembly.

In addition, the thermal shield reduces the amount of radiation that reaches the reactor vessel. The incoming reactor coolant is directed downward along the outside of the core barrel cylinder and upward through the fuel assemblies contained inside the core barrel. A small amount of coolant flows upward through the space between the core barrel cylinder and the baffle plates. A small portion of the coolant also runs down the annulus between the thermal shield and the core barrel cylinder, through holes drilled in the core barrel cylinder bottom flange, and then upward through the core [3, 9].Report No. 1200459.401 .Rl 3-12 $Structural Integrity Associates, lnc IneriysocaesIc The core barrel supports the fuel assemblies.

The core barrel consists of flanged cylinder, rings of internal horizontal former plates bolted to the cylinder at eight elevations, and vertical baffle plates bolted to the former plates to produce an inner wall enclosing the fuel assemblies, as shown in Figure 3-9. The bottom flange of the CSS is bolted to the top flange of the core barrel cylinder and the lower grid assembly bolts to the core barrel bottom flange [9].At the fourth elevation from the bottom, near the hottest section of the core, the ring of former plates are narrower than those at the other elevations, and the baffle plates are bolted to these narrower formers with special shoulder bolts that maintain a 1/4-inch gap between the baffle plates and former plates. This arrangement provides additional cooling flow to the hottest portion of the baffle plates and some flexibility to the assembly under unusual loads. The thermal shield is attached to the outside of the core barrel assembly by an upper thermal shield restraint assembly [9].3.1.7.1 Core Barrel The core barrel is a cylinder approximately 121/4 feet high and 2 inches thick. It is formed from two cylinders welded together circumferentially.

The core barrel top and bottom flanges are welded to the ends of the cylinder.

The core support shield assembly bolts to the top flange with one hundred and twenty 13/4-inch diameter core barrel bolts secured with locking clips or locking cups. The lower grid shell forging is bolted to the core barrel bottom flange with one hundred and eight 13/4%-inch diameter core barrel bolts secured with locking clips or locking cups [9].At twenty equal spaced locations, the thermal shield upper restraint assemblies are bolted on to the outer vertical wall of the core barrel top flange with three 1 '/2-inch diameter bolts secured with locking clips [9].Report No. 1200459.40 l.R1 3-13 C Structural Integrity Associates, Inc?

3.1.7.2 Baffle Plates The vertical baffle plates from an outer perimeter of the core area to confine and direct the flow of reactor coolant. The baffle plates do not ordinarily provide any structural support to or affect the alignment of the fuel assemblies since there is a clearance between the outer fuel assemblies and baffle plates. The baffle plates are approximately 13'/4 feet high, 3/4/4-inch thick, with width varying from about 8 to 45 inches. At various elevations the baffle plates have rows of 1 3/8-inch diameter flow holes to help equalize the coolant behind the baffle plates with the main coolant flow. A schematic representation of the core barrel with the baffle plates is provided in Figure 3-11 [9].The baffle plates are bolted to the formers with seven hundred and fifty-six 5/8-inch diameter hex head bolts at seven elevations secured in place by stainless steel locking pins.At the fourth highest elevation, the narrow baffle plates are bolted to the formers with one hundred and eight 5/8-inch diameter core barrel shoulder screws which have a shoulder which holds the baffle plates '/4-inch out from the former. The shoulder screws are secured in place by 1/8-inch diameter dowels which are welded in place [9].At the tall vertical joints where two baffle plates meet the form comers, a total of six hundred and twelve 7/16-inch diameter bolts secured with locking rings hold the plates together [9].3.1.7.3 Formers The 1 1/4-inch thick former plates provide horizontal framing to support the vertical baffle plates at eight elevations.

The outside edges of the formers curve to match the inside surface of the core barrel to which they are bolted. Inside surfaces of the formers are either flat or step shaped to support the various baffle plates. The formers have small holes to permit some reactor coolant to flow up through and cool the void behind the baffles [9].Report No. 1200459.401.RI 3-14 r StructuraI Integrity Associates, Inc To hold the formers in place, a total of seven hundred and four 5/8-inch diameter socket head cap screws are bolted through the outer side of the core barrel into the formers. The screws are held in place with locking pins [9].At 16 locations on the top and bottom rows of formers there are 0.625-inch diameter Alloy X-750 dowels used to locate the formers on the core barrel [9].3.1.8 Lower Internals Assembly The lower internals assembly consists of a lower grid assembly, a flow distributor assembly, and in-core monitoring instrumentation guide tube assemblies, as shown in Figure 3-12. The lower internals assembly is bolted to the bottom flange of the core barrel cylinder, and its function is to direct coolant flow upwards through the fuel assemblies

[3, 9].3.1.8.1 Lower Grid Assembly The lower grid assembly consists of three grid structures or flow plates: (1) the lower grid rib section, (2) the flow distributor plate, and (3) the lower grid forging. The lower grid assembly provides alignment and support for the fuel assemblies, supports the thermal shield and flow distributor, and aligns the incore instrument guide tubes with the fuel assembly instrument tubes.Each of these flow plates has holes or flow ports to direct coolant flow upward toward the fuel assemblies.

The lower grid assembly is surrounded by the lower grid shell forging. The lower grid shell forging is a forged flanged ring cylinder, which supports the various horizontal grid structures and flow plates [9].The top flange of the lower grid shell forging is bolted to the lower flange of the core barrel by 108 (13/4-inch diameter) core barrel bolts [9].Alignment between fuel assemblies and incore instruments is provided by pads bolted to the top of the lower grid rib section [9].Report No. 1200459.401 .Rl 3-15 IV Structural Integrity Associates, InO Twelve pairs of guide blocks bolted to the outer surface of the lower grid shell forging mate with the guide lugs welded on the inside RV wall. The guide blocks are not credited with preventing internals motion or as an internals support during normal and upset operating conditions and also for seismic and design break events [9].3.1.8.2 Lower Grid Rib Section The lower grid rib section is a 5-inch thick, 141-inch diameter disk through which 177 squares are machined out, leaving a grid with 1-inch wide ribs. The square holes align with the fuel assembly locations in the core above. There are additional holes about the periphery of the disk to permit a small bypass flow of reactor coolant up behind plates in the core barrel [9].There are 384 small fuel assembly support pads attached to the top of the rib section to provide seating surface and support for the bottoms of the fuel assemblies.

A '/2-inch diameter, 2 1/4-inch long cap screw is used to hold each pad in place. Two 3/8-inch diameter Alloy X-750 dowels position each pad. Below the rib plate at 48 grid intersections, there are support post assemblies that provide support from the lower grid forging. The support post assemblies are bolted in place with 1-inch diameter socket head cap screws secured with welded locking pins [9].Incore guide tube spider castings are welded in 52 of the holes to provide support for the tops of the incore instrument guide tubes. The spider castings are cylinders with four legs that are welded to the walls of the holes in the lower grid rib section [9].3.1.8.3 Lower Grid Flow Distributor Plate The lower grid flow distributor plate, located midway between the lower grid rib section and the lower grid forging, aids the distributing coolant flow. It is a flat 1-inch thick, 135 7/8-inch diameter perforated plate with a 1/8-inch lip around the bottom. The flow distributor plate rests on and is welded to a 1/2-inch lip on the lower grid shell forging [9].Report No. 1200459.401 .Rl 3-16 *jjStructural IntegrIty Associates, lnc InegiysoiteIc The flow distributor plate has six hundred and seventy-seven 3 3/8-inch diameter flow holes (177 of which are aligned with the center of the fuel assemblies).

Twelve of the normal flow holes near the center of the flow distributor plate are fitted with orifice plugs which reduce the diameter of the flow port down to 1 7/8-inches.

There are also 24 smaller flow holes and 48 holes to accommodate the support posts. The support posts are welded to the lower grid flow distributor plate [9].3.1.8.4 Lower Grid Forging The lower grid forging is a single 135-inch diameter forged disk that serves as the main weight-bearing structure in the lower grid. The majority of the lower grid forging, i.e., the center 96 inches of the disc, is a 131/22 inches thick. The disc tapers up to 6 inches thick at its edges. There are 177 flow holes machined out of the lower grid forging, aligned with the fuel assemblies.

The lower grid forging is welded to the lower grid shell forging. The lower ends of the 48 support post assemblies are welded to the top of the lower grid forging [9].3.1.8.5 Lower Grid Shell Forging The lower grid shell forging is a 2-foot high, 136-inch ID cylinder with numerous internal and external flanges and lips that support the various items of the lower grid assembly.

The lower grid shell forging is 4-inches thick at its thinnest cross-section

[9].The lower grid shell forging is bolted to the core barrel lower flange with 108 core barrel bolts, described previously.

The lower end of the thermal shield is shrunk fit on the lower grid flange and fastened by ninety-six 1-inch diameter bolts or studs and nuts secured with locking clips or locking cups. The lower grid rib section is bolted to the shell forging with thirty-six 3/4-inch diameter socket head cap screws secured with welded locking pins. The flow distributor plate rests on and is welded to a 1/2-inch lip on the lower grid shell forging. The lower grid forging rests on and is welded to the top surface of the lower grid shell forging lower flange. The flow distributor assembly bolts into the bottom of the shell forging with ninety-six 1-inch diameter high strength bolts secured with locking clips. The lower surface of the bottom flange of the Report No. 1200459.40 l.R1 3-17 V Structural Integrity Associates, Inc.'

lower grid shell forging holds the clamping ring in place, which holds the incore guide support plate in place against the flow distributor flange [9].Guide blocks are bolted at twelve equidistant locations around the outside vertical wall of the lower grid shell forging. These blocks engage the guide lugs welded to the wall of the reactor vessel and serve to maintain alignment and prevent rotation of core internals.

Shock pads are bolted to the underside of the upper flange of the lower grid shell forging, directly above the guide blocks [9].3.1.8.6 Guide Blocks The 24 guide blocks are each 61/2-inches wide, 5-inches high with beveled guiding/mating surfaces extending out 3 inches from the shell forging wall. Each is held in place with a 1-inch diameter hex head bolt and washer and a 11/2/2-inch diameter Alloy X-750 dowel [9].3.1.8.7 Shock Pads Twelve shock pads are bolted to the lower surface of the upper flange of the lower grid shell forging, located directly above the reactor vessel guide lugs. In the unlikely event of a core barrel joint failure, the reactor vessel core guide lugs and lower grid shock pads will limit the core to drop approximately one-half inch [9].3.1.8.8 Support Post Assemblies The support post assemblies are 48 cylinders placed between the lower grid forging and the lower grid rib section to provide support. The support post assemblies consist of the support pipes and the associated bolting plugs. The support pipes are made from 101/2-inch high sections of 4-inch schedule 160 pipe. There are four equally spaced notches at the bottom of the cylinders, where they are welded to the top of the lower grid forging that allow coolant flow up from below. The bolting plugs are 1%A-inch high disks welded to the top of the support pipes.The bolting plugs have four scallops shaped holes machined out of the edges so that the tops Report No. 1200459.401 .Rl 3-18 r Structural Integrity Associates, Inc?

have a cruciform shape through which coolant can flow. The top of each bolting plug is drilled and tapped to accept the cap screw used to hold it to the lower grid rib section [9].The support post assemblies rest on top of the lower grid forging, straddling over grid intersections.

They are fitted through matching holes in the flow distributor plate and rest against the bottom of the lower grid rib section. The support posts are welded to the top of the lower grid forging and on both sides of the penetration of the lower grid flow distribution plate.They are bolted to the bottom of the lower grid rib section by 1-inch diameter socket head screws secured with locking pins [9].3.1.9 Flow Distributor Assembly The flow distributor assembly supports the IMI guide tubes and directs the inlet coolant entering the bottom of the core. It consists of a perforated head (plate), a flange, an IMI guide support plate, and a clamping ring [10].3.1.9.1 Flow Distributor Head and Flange The flow distributor is a perforated dished head with an external flange that is bolted to the bottom flange of the lower grid. The flow distributor supports the incore instrument guide tubes and directs the inlet coolant entering the bottom of the core [9].The flow distributor head is a 2-inch thick, 136-inch ID bowl-shaped plate that bows downward about 20 inches. The head is welded to the flow distributor flange, which is 5 inches high, with an approximately 3-inch thick flange extending out to a 142 inch OD. The incore guide support plate fits across the flange, resting in a lip in the flange. The clamping ring fits against the inside diameter of the flange on top of and holding the incore guide support plate in place. This whole assembly is bolted to the bottom of the lower grid shell forging with ninety-six 1-inch diameter high strength bolts secured with locking clips [9].Report No. 1200459.40 l.Rl 3-19 VStructural Integrity Associates, Inc?

There are 52 approximately 4'/2-inch diameter holes through which the incore instrument guide tube pass. Fifteen of these incore holes have shallow counter-bores on the bottom edge to permit welding the instrument guide tubes directly to the flow distributor head plate. The remaining 37 guide tubes are secured by a set of four gussets which are 3/4-inch thick triangular shaped pieces, 6-inches high and 1%-inches wide. The long sides of the gussets are welded to the guide tubes and the bases are welded to the distributor head [9].There are one hundred and fifty-six 6-inch diameter holes and five 31/2/2-inch diameter holes in the flow distributor head to permit reactor coolant flow upward through the lower grid assembly [9].3.1.9.2 Incore Guide Support Plate The incore guide support plate is a 134-inch diameter, 2-inch thick disk, with 52 shaped holes to accommodate the incore instrument guide tubes. The guide tubes are held in place by washers and guide tube nuts secured by welded locking clips [9].At 46 of the guide tube holes there are also four oval-shaped flow ports machined through the guide support plate to permit reactor coolant flow parallel to the incore guide tubes. There are also numerous holes between 61/2 and 71/2 /2inches in diameter for reactor coolant flow [9].The incore guide support plate rests on a lip in the top of the flow distributor assembly.

The clamping ring sits on top of the plate and holds it in place [9].3.1.9.3 Clamping Ring The clamping ring is an approximately 4-inch high, 132-inch diameter, 1-inch thick ring that fits against the inner surface of the flow distributor ring forging, atop the incore support plate. When the flow distributor assembly is bolted to the lower grid shell forging, this clamping ring holds the incore support plate in place [9].Report No. 1200459.401 .Rl 3-20 CStructural Integrity Associates, Inc?

3.1.10 Incore Guide Tube Assemblies The incore instrument guide tube assemblies guide the 52 incore instrument assemblies from the instrumentation nozzles in the reactor vessel bottom head to the instrument tubes in the fuel assemblies.

Horizontal clearances are provided between the reactor vessel instrumentation nozzles and the instrument guide tubes in the flow distributor to accommodate misalignment.

The incore instrument guide tubes are designed so they will not be affected by core drop [9].The guide tubes are long tapered tubes through which the incore nuclear detectors and thermocouples are fed into the fuel assemblies.

The diameters vary along the length of the guide tubes. At the top, where they are held in place by the spiders welded into the lower grid rib section, the guide tubes have a 1-inch OD with a 0.60 to 0.67-inch center bore. At the bottom, the guide tubes have a 41/2-inch OD with a 3'/2-inch ID. The top 32-inches of all 52 guide tubes, from where they penetrate the flow distributor up to the spiders in the lower grid rib section, are essentially identical.

There are ten different guide tube models, however, which differ in their overall length, varying from 77% to 51 1/4inches.

The length required depends upon the location within the core, as the distances vary between the incore guide support plate and the flow distributor head and between the flow distributor head and the bottom of the reactor vessel [9].The guide tube assemblies are attached to the bottom of the flow distributor head either by a weld bead around the full circumference of the guide tube, or by four gussets which are welded to the flow head and the guide tubes. The guide tubes then have an interference fit through holes in the incore guide support plate. The guide tubes are held to the top of the incore support plate with washers and the guide tube nuts. The outside of the guide tubes have a 1 3/4-inch section of threading at this location to engage with the guide tube nuts. The guide tubes have an approximate 2-inch diameter where they pass up through 61/2-inch diameter holes in the lower grid forging and the 3 3/8-inch diameter holes in the flow distributor plate [9].Report No. 1200459.401.R1 3-21 C structural Integrity Associates, Inc 3.1.10.1 Gussets Thirty-seven guide tubes whose projected length below the flow distributor require additional stiffness are secured by sets of four gussets which are 3/4-inch thick triangular shaped pieces, 6-inches high and 13/4-inches wide. The long sides of the gussets are welded to the guide tubes and the bases are welded to the distributor head [9].3.1.10.2 Guide Tube Nuts The guide nuts are 21/2-inch tall, 1/4-inch thick nuts that fit over the guide tubes and secure them to the top of the incore support plate [9].3.1.10.3 Guide Tube Spiders Spider castings are welded in 52 of the holes to provide support for the incore instrument guide tubes. The spider castings are 13/4-inch high, 1-inch ID cylinders with four '/4-inch thick L-shaped legs that extend out and are welded to the walls of the holes in the lower grid rib section.The inner diameters of the spider tube cylinders are chrome plated 0.0002 to 0.0004 inches thick.The chrome-plated bore of the spider hub forms a guide bushing for the top of the incore instrument guide tube assembly to accommodate longitudinal thermal expansion

[9].Report No. 1200459.401.R1 3-22 J $Structural Integrity Associates, Inc!

Plenum Assembly C L OE CiV 0 U 2i c. E Figure 3-1. General Arrangement of Typical B&W Reactor Vessel Internals

[3]Report No. 1200459.401 .RI 3-23 V Structural Integrity Associates, IncO Plenum Cover Assembly N'y Plenum Cylinder Upper Grid Assembly-

>1): I, ( N 7' {/' Cover Plate N-, >\ .',-'-,",-Side View CRA Tube Lifting Lug-z R "--" -Weldment Upper Flange Reinforcement Plate'" Lower Flange T Upper Grid...... Rib Section F pprRing Forging FA Support Pads Figure 3-2. Plenum Assembly [9]3-24 IjjfStructural Integrity Associates, Ic.3-241 Report No. 1200459.401.R1 Top View Keyway --Bottom Flange Support Ring Segment Support Flange Cover Plate Support Pad Side View Lifting Lug Support Ring Segment Support Flange Weldment Ribs Figure 3-3. Plenum Cover Assembly [9]Report No. 1200459.401 .Rl 3-25 C Structural Integrity Associates, Inc!

Top View Top Flan~ge Re in o(ameint Plate Round Bars Swie view..4 -- Top FlanrgeFlow Holes.7',.J-plenum Clyinder______ RetnforngPlate-Bottom Flange Figure 3-4. Plenum Cylinder Assembly [9]Report No. 1200459.401.Rl 3-26 C Structural Integrity Associates, Inc!

Top View K> ~ If/:i F-Rtib SectionRod Side View Bottorm Vk.w Support Pad Figure 3-5. Upper Grid Assembly [9]3-27 qjj-Structural Integrity Associates, IncO Report No. 1200459.401.R1 ax:~C>Thp ~ew maemn Abaorbing MDWW Cordmi Rod Figure 3-6. Control Rod Assembly [9](Not in Scope)Report No. 1200459.401.R1 3-28 V Structural Integrity Associates, IncO Rod Gude Smmi Rod GOwde Setorm Rod GuideTub" Rvd Gtu. Spacer Castings Rod Guide Seacru -Rcd GusTov <-- Spam Cmtinfg x/72~ 7/...FC A ... .......Figure 3-7. CRGT Assembly Spacer Castings and Rod Guide Brazement Configuration

[9]Report No. 1200459.401.Rl 3-29 VStructural Integrity Associates, Inc.

a Guide Oarzement Gi-idaTube Guide Haushg Figure 3-8. Control Rod Guide Tube Assembly Pipes [9]Report No. 1200459.401.R1 3-30 C Structural Integrity Associates, Inc?

Lug Core suppoint Shiald Outlet Nozzle Rouind Bars Core $*"*I Asemarbly Lowr Iviamats Arnmb CSS Bottom Range CB Top Fla~nge 4,, -Thernal Shield (Not In Scope)CB Cylirwfr.... .Formers Baffle PlalftCa Bottomn Figure 3-9. Core Support Assembly [9]3-31 C structural Integrity Associates, IncO Report No. 1200459.401.Rl ExrciseLug (r fieg Rector Vessel WaI Rirg (M in pe)Mount" Ring O Noznzl Care Supr"", Sbd Sertion Z-Z Figure 3-10. Vent Valve Assembly [9]3-32 j Structural Integrity Associates, IncO Report No. 1200459.401.Rl Figure 3-11. Core Barrel Interior with Baffle Plates [9]Report No. 1200459.401 .Rl 3-33 C structural Integrify Associates, InO FA Suppodt Fa Figure 3-12. Lower Intemals Assembly [9]Report No. 1200459.401.R1 3-34 C Structural Integrity Associates, Inc 3.2 ANO-1 Design Modifications and Distinctions Some of the design distinctions in the ANO- 1 reactor vessel internals include [18]: " The ANO-1 reactor vessel internals has one additional component function of supporting the reactor vessel level monitoring probes. One of the control rod drive mechanisms (CRDM-30) was removed and the control rod guide assembly in the plenum was modified in order to install the reactor vessel level monitoring system (RVLMS). Therefore, support of the RVLMS probes is a site-specific function for ANO-1 that is in addition to the functions listed in BAW-2248A

[9]." The ANO- 1 Safety Analysis Report (SAR) lists "provide gamma and neutron shielding" as a separate function.

This function was not within the scope of BAW-2248A

[9] and is an intended function of the internals.

Items that support this intended function include the thermal shield, thermal shield upper restraints and associated bolting. The thermal shield and upper restraint are fabricated from austenitic Type 304 Stainless Steel (ASTM A240, Type 304 [18]).* The ANO-1 SAR section 3.2.4.1 lists the "support for the surveillance specimen assemblies in the annulus between the thermal shield and the reactor vessel wall" as a function of the reactor internals.

Although all the specimens have been removed, portions of the shroud tube and the supports that are bolted to the core support shield remain. These components only have the remaining function of being secured and hence prevent loose parts in the RCS [5].* The ANO-1 plenum cover assembly is entirely stainless steel.* The ANO- 1 plenum cylinder is entirely stainless steel except for the Inconel dowels.* The control rod guide tube assembly is entirely stainless steel." The core support shield assembly is entirely stainless steel except for some of the miscellaneous locking device parts that are fabricated from Inconel.* The core barrel assembly is entirely stainless steel except for the Inconel dowels.* The lower internals assembly is entirely stainless steel except for the Inconel dowels." The incore guide tube assembly is entirely stainless steel.Report No. 1200459.401 .Rl 3-35 rStructural Integrity Associates, 1n0 4.0 EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA 4.1 Examination Acceptance Criteria 4.1.1 Visual (VT-3) Examination Visual (VT-3) examination has been detennined 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 [14], 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. Foreign materials or accumulation of corrosion products that could interfere with control rod motion or could result in blockage of coolant flow through fuel 5. Wear of mating surface that may lead to loss of functionality

6. Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5%For components where visual (VT-3) is specified in the Primary and Expansion group, specific descriptions of the relevant conditions are provided in Tables 5-1 and 5-2 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 Primary and Expansion Report No. 1200459.401 .Rl 4-1 Structural Integrity Associates, Inc!

components listed in Tables 5-1 and 5-2 of this document.

A specific relevant condition refers to a more comprehensive description of an indication that goes beyond the description presented in ASME Section XI for B-N-3 components.

The examination acceptance criteria for components requiring (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.

Relevant conditions are defined in ASME Section IX, IWA-9000 [14]; they do not include fabrication marks, material roughness, and other conditions acceptable by material design, and manufacturing specifications of the component.

4.1.2 Visual (VT-1) 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 the components, including such conditions as cracks, wear, corrosion, or erosion." The acceptance criterion for any visual (VT-1) examination is the absence of any relevant conditions defined by the ASME Code, as supplemented by more specific plant inservice inspection requirements. (Note: MRP-227-A does not specify visual (VT-1) examination for B&W designed RVI components.)

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 in MRP-228 [11]. 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 length when used in conjunction with sizing aids (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 Report No. 1200459.401.R1 4-2 ij tstructural Integrity Associates, Inc!

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 criteria for EVT-1 examination are the absence of any detectable surface-breaking indication. (Note: MRP-227-A does not specify enhanced visual (EVT- 1) examination for B&W designed RVI components.)

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 [11 ]. MRP-228 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)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, it 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 of the pass/fail acceptance of individual bolts or pins, the examination acceptance criterion for volumetric (UT) examination of bolts and pins is based on a reliable detection of indications as established by the individual technical justification for the proposed examination.

This is in keeping with current industry practice.

For example, planar flaws on the order of 30%Report No. 1200459.401.RI 4-3 I Structural Integrity Associates, lnO of the cross-sectional area have been demonstrated to be reliably detectable in previous bolt NDE technical justifications for baffle-former bolting [3].Bolted and pinned assemblies are evaluated for acceptance based on meeting a specified number and distribution of functional bolts and pins. The criteria for this evaluation can be: 1) found in previous Owners Group reports, 2) developed for use by the PWROG or 3) developed on a plant-specific basis by the applicable NSSS vendor.4.1.6 Physical Measurements Examination Continued functionality can be confirmed by physical measurements where, for example, loss of material caused by wear, loss of pre-load of clamping force caused by various degradation mechanisms, or distortion/deflection caused by void swelling may occur.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 in order to further characterize and disposition a detected condition 2. Engineering evaluations that demonstrate the acceptability of detected conditions Report No. 1200459.401 .RI 4-4 ýStructural IntegrIty Associates, M0

3. Repair and 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

[13] and/or 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

[3] and the ANO-1 Corrective Action Process [27]. Entergy Nuclear Operations shall provide a summary report to the EPRI MRP Program Manager of all inspections and monitoring, items requiring evaluation, and new repairs.This report shall be provided within 120 days of the completion of the outage during which the activities occur. This is a 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 ANO-1 is provided in Table 5-6.Report No. 1200459.401 .R1 4-5 jstructurai Integrity Associates, Inc!

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 ANO-I'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-227 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 finctionality 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 R VI 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- 189 [10] are appropriate for ANO- 1. The FMECA and functionality analyses were based on the assumption of 30 years of operation with high leakage core loading patterns; therefore, ANO-1 is bounded by the assumption in MRP-227-A

[3].As discussed in Section 1.8.4.1 of this document, operations at ANO-1 conform to the following assumptions in Section 2.4 of MRP-227-A

[3]: Report No. 1200459.401 .Rl 5-1 jStructurai Integrity Associates, Inc?

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

Applicability:

ANO-1 historic core management practices meet the requirements of MRP-227-A

[23, 24]." Base load operation, i.e., typically operates at fixed power levels and does not usually vary power on a calendar or load demand schedule.Applicability:

ANO-1 operates as a base load unit [25].* No design changes beyond those identified in general industty 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.

ANO- 1 has not made any modifications of the RVI components beyond those identified in general industry guidance or recommended by the vendor (B&W) since the May 2007 effective date of this statement, and therefore meets this requirement of MRP-227-A.

Further discussion on modifications that were made based on general industry guidance or recommended by the vendor (B&W) is provided in Section 3.2 of this document.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 ficilities in accordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are Report No. 1200459.401 .RI 5-2 C Structural Integrity Associates, Inc within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) and propose necessar, 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.

As part of the license renewal application, ENO reviewed the current design and operation of the ANO-l reactor vessel internals using processes to identify RCS system components subject to aging management review and to incorporate approved BWOG Topical Report BAW-2248A

[9]and determined that ANO-1 RVI have three additional components that were not listed in BAW-2248A, but were included in the ANO-I Section XI program [18, 34]:* Reactor vessel level monitoring system (RVLMS) probe supports* Remaining portions of the surveillance specimen holder tubes (SSHT)* Thermal shield and thermal shield upper restraint These components will undergo a future screening and categorization, and based on the screening results, they will be removed from future inspections if they screen out, or added to the primary or expansion categories if they screen in. Until that time, in order to ensure that these components are adequately managed during the period of extended operation, they will be inspected during the 10-year intervals based on their aging effects. For the RVLMS probe supports, cracking and stress relaxation were identified as aging effects for the RVLMS brazement guide j-bolt and nut [34, pgs 13, 14]]. For the remaining portions of the SSHTs, they need to remain secured to prevent loose parts in the RCS. Cracking and stress relaxation were identified as aging effects of the SSHT bolting [34, pg. 14]. For the thermal shield and thermal shield upper restraint, cracking and reduction of fracture toughness were identified as aging effects for the thermal shield and thermal shield upper restraint assemblies

[34, pg. 14]. Based on a review of the aging effects, the orphan components will be visually inspected (VT-3) during the 10-year ISI inspections.

Report No. 1200459.401.Rl 5-3 1jIFStructural Integrity Associates, Mc Inegittsocatsnc In addition to these components, AREVA has also advised the B&W utilities that the vent valve locking devices for the original and modified designs be included as an existing program under MRP-227-A based on recent OE from ONS-1 [32]. Based on this recommendation, Table 5-1 was modified and Table 5-3 was added to list the existing programs under the current Section XI program.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 petformn plant-specific analysis either tojustify 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-227), and Westinghouse guide tube support pins (split pins) (Section 4.3.3 in MRP-227).This action does not apply to B&W designed units.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 cotnfirm 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 faicility.

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

If necessary, the examination methods andfrequency for non-stress relieved B& W core support structure upper flange welds Report No. 1200459.401 .Rl 5-4 $Structural Integrity Associates, Inc!

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 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.A review of the fabrication records was performed by AREVA [30]. The review of the records indicates that the upper flange was stress relieved during the fabrication process [30, 31]. This review satisfies the requirements of Applicant/Licensee Action Item 4 and the upper flange does not need to be inspected as a "Primary" inspection category component.

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 finctionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227.In accordance with MRP-227-A, physical measurements are to be performed for the following B&W designed RVI components:

• Plenum cover weldment rib pads Report No. 1200459.401 .RI 5-5 Structural Integrity Associates, Inc

• Plenum cover support flange" CSS top flange Physical measurements are required to determine the differential height of top of the plenum rib pads to the reactor vessel seating surface with all three items inside the reactor vessel, but with the fuel assemblies removed. The acceptance criteria for this one time physical measurement shall be an average measured differential height from the top of the plenum rib pads to the vessel seating surface of 0.004 inches relative to the as-built condition in accordance with Table 5-4 of this document [3].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, MRP-227 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 bqffle-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 identijied that although the B& W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques.

Applicants/licensees shalljustify 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-22 7, applicants/licensees shall provide theirjustification for the continued operability of each of the inaccessible components and, if necessaiy provide their plan for the replacement of the components for NRC review and approval.ANO- 1 will justify the acceptability of inaccessible and non-inspectable components (core barrel cylinder including vertical and circumferential seam welds, former plates, external baffle-to-Report No. 1200459.401 .R1 5-6 rStructural Integrity Associates, Inc baffle bolts and their locking devices, core barrel-to-former bolts and their locking devices, and internal baffle-to-baffle bolts) for continued operation through the period of extended operation by performing an evaluation, proposing a schedule for replacement, or justification for some other alternative process for these components.

The evaluation, schedule for replacement, or justification for some other alternative process will be submitted to the NRC by the end of one year from the initial inspection of the linked Primary component items if the inspection results indicate aging, which is the implementation date for this condition.

Any "other alternative process" shall include justification of operation in the degraded condition on a generic or plant-specific basis. Any "aging" detected during the inspection is defined to mean when the expansion criteria for the linked Primary component are met.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& W IMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their finctionality during the period of extended operation or for additional R VI components that may be fabricated friom CASS, martensitic stainless steel or precipitation hardened stainless steel materials.

These analyses shall also consider the possible loss 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 of MRP-227. That is, the requirement would applv to components fabricated firom 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 finctionality of the components being evaluated under all licensing basis conditions of operation.

The Report No. 1200459.401 .Rl 5-7 Z Structural Integrity Associates, Inc!

applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-22 7.The ANO- 1 RVI components that are fabricated from CASS and precipitation hardenable stainless steel materials include: CASS Materials:

-Control Rod Guide Tube (CRGT) Assembly Spacer Castings (A351, Grade CF-3M)(Primary component per MRP-227-A

[3])-Core Support Shield Assembly -Vent Valve Bodies (A35 1, Grade CF-8) (ferrite content determined to be below the screening criteria)-Instrument Guide Tube Assembly -Spider Castings (A351, Grade CF-8) (Primary component per MRP-227-A

[3])* Precipitation Hardenable Stainless Steel:-Core Support Shield Assembly -Vent Valve Retaining Rings (AMS 5658, Type 15-5PH)(Primary component per MRP-227-A

[3])The applicable aging effects for the items fabricated from CASS, martensitic and precipitation hardenable stainless steels are reduction in fracture toughness by either thermal or irradiation embrittlement or a combination of the two. An analytical approach to assess the effect of reduction of fracture toughness on the applicable reactor vessel internals will be performed.

The analysis will include an assessment of the potential for synergistic thermal aging and neutron irradiation embrittlement for each affected CASS component by comparing its nominal 60-year neutron exposure to significant thresholds established in regulatory guidance or, conversely, to industry-established thresholds justified by experimental data and corroborated by industry operating experience.

For CASS components with the 60-year neutron exposure below the justified threshold, the reduction in fracture toughness used in the analytical assessments will be based on thermal aging embrittlement data. Otherwise, an additional reduction in fracture Report No. 1200459.401 .Rl 5-8 Structural Integrity Associates, Inc?

toughness to account for the synergistic effects of irradiation and thermal aging will be documented.

Industry efforts are underway to reach agreement on generic screening thresholds for combinations of thermal aging and neutron irradiation embrittlement, with active participation in those efforts by Entergy personnel.

In addition to the establishment of generic screening thresholds, the industry efforts include semi-generic procedures for the functional evaluations of those component locations that screen in, based upon the established screening thresholds, with the intent to minimize the plant-specific requirements for those evaluations.

Depending upon the outcome of discussions with the regulator, the ANO-1 commitment for the evaluation of RV internals CASS component locations may include some degree of functionality assessment or flaw tolerance analysis of susceptible CASS component locations, and possibly the development of an inspection plan for some of those locations, depending upon the results of both industry generic evaluations and plant-specific evaluations.

Such an inspection plan will consider scope of the inspections, particular inspection methods to be employed, accessibility and sample size issues, and inspection acceptance criteria.Should data or the analytical evaluations indicate that the inspections can be modified or eliminated, ENO will provide plant-specific justification to demonstrate the basis for the modification or elimination, as applicable.

The generic or plant-specific assessments for the CASS components will be completed 12 months prior to the second refueling outage after entering the period of extended operation.

For components fabricated from CASS, martensitic and precipitation hardenable stainless steel materials, the acceptance criteria (critical crack size) will be determined by analysis using an NRC approved methodology.

Report No. 1200459.401.R1 5-9 C StructuraI Integrity Associates, Inc?IneriysocaesIc 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 submittalfor 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):

In addition to the implementation of MRP-22 7 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)Jor 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 ANO- 1 RVI AMP and the implementation of the ten elements of NUREG- 1801 (GALL Report), Revision 2, Chapter XI.M 16A, "PWR Vessel Internals" [7]that are essential for successful management of component aging are described in Section 2.3 of this document.2. To ensure the MRP-227 program and plant-specific action items will be carried out by applicants/licensees, applicants/licensees are to submit an inspection plan which Report No. 1200459.401.R1 5-10 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-22 7, as approved by the NRC, the applicant/licensee shall identify where their program deviates from the recommendations of MRP-22 7, as approved by the NRC, and shall provide ajustification fior any deviation which includes a consideration of how the deviation affects both"Prinany" and "Expansion" inspection categoty components.

The aging management program plan for ANO-I will not deviate from the recommendations of MRP-227-A.

Inspection of Primary and Expansion category components are provided in Tables 5-1 and 5-2 of this document will be performed in accordance with the requirements of MRP-227-A.

Report No. 1200459.401 .R1 5-11 Structural Integrity Associates, Inc!

Table 5-1. B&W Plants Primary Category Components from Table 4-1 of MRP-227-A

[3, 32]Item Applicability Effect (Mechanism)

Expansion Link(') Examination Examination Coverage Inspection Method/Frequency(l)

Schedule Plenum Cover All plants Loss of material and None One-time physical measurement Determination of differential Physical Assembly & associated loss of core no later than two refueling outages height of top of plenum rib pads measurements to Shield Core clamping pre-load from the beginning of the license to reactor vessel seating surface, be performed in Support Shield (Wear) renewal period, with plenum in reactor vessel. 1R26.Assembly Plenum cover Perform subsequent visual (VT-3) See Figure 4-1 of MRP-227-A weldment rib pads examination on the 10-year ISI Plenum cover interval.support flange CSS top flange Control Rod All plants Cracking (TE), None Visual (VT-3) examination during Accessible surfaces at each of VT-3 examination Guide Tube including the detection the next 10-year IS1. the 4 screw locations (at every To be performed Assembly of fractured spacers or Subsequent examinations on the 900) of 100% of the CRGT spacer in 1R26.CRGT spacer missing screws 10-year ISI interval, castings castings (limited accessibility).

See Figure 4-5 of MRP-227-A I Report No. 1200459.401.Rl 5-12 Situctural Integrity Associates, Inc.

Table 5-1. B&W Plants Primary Category Components from Table 4-1 of MIRP-227-A

[3, 32] (continued)

Examination Inspection Item Applicability Effect (Mechanism)

Expansion LinkV 1) MethodIFrequencyati Examination Coverage Schedule Core Support All plants Cracking (TE), including None Visual (VT-3) examination during 100% of accessible surfaces VT-3 examination Shield the detection of surface the next 10-year ISI. (see BAW-2248A, page 4.3 and To be performed Assembly irregularities, such as Table 4-1) in 1 R26.CSS vent valve damaged, fractured Subsequent examinations on the top retaining ring material, or missing 10-year ISI interval.

See Figure 4.11 of MRP-227-A CSS vent valve items bottom retaining ring Core Support All plants Bolts: Cracking (SCC) UTS bolts and LTS Volumetric examination (UT) of the 100% of accessible bolts and UT examination Shield Locking Devices: Loss of studs/nuts or bolts bolts within two refueling outages their locking devices(2).

To be performed Assembly material, damaged, and their locking from 1/1/2006 or next 10-year ISI in 1R26.Upper core distorted or missing devices interval, whichever is first. See Figure 4-7 of MRP-227-A.

barrel (UCB) locking devices (Wear or bolts and their Fatigue damage by failed SSHT studs/nuts or Subsequent examination on the locking devices bolts). bolts and their locking 10-year ISI interval unless an devices (not evaluation of the baseline results applicable to ANO-1) submitted for NRC staff approval justifies a longer interval between Lower grid shock pad examinations.

bolts and their locking devices (not Visual (VT-3) examination of bolt applicable to ANO-1) locking devices on the 10-year ISI interval.Report No. 1200459.401.Rl 5-13 l Structural Integrity Associates, Inc.'

Table 5-1. B&W Plants Primary Category Components from Table 4-1 of MRP-227-A

[3, 32] (continued)

Item Applicability Effect (Mechanism)

Expansion Link(1)Examination Method/Frequency(1)Examination Coverage Inspection Schedule Core Barrel All plants Bolt: Cracking (SCC) UTS bolts and LTS Volumetric (UT) of the bolts during 100% of accessible bolts and UT examination Assembly Locking Devices: Loss of studs/nuts or bolts the next 10-year ISI interval from their locking devices(2). To be performed Lower core material, damaged, and their locking 1/1/2006.

in 1R26.barrel (LCB) distorted or missing devices See Figure 4-8 of MRP-227-A.

bolts and their locking devices (Wear or Subsequent examination on the locking devices Fatigue damage by failed SSHT studs/nuts or 10-year ISI interval unless an bolts). bolts and their locking evaluation of the baseline results devices (not submitted for NRC staff approval applicable to ANO-1) justifies a longer interval between examinations.

Lower grid shock pad bolts and their locking Visual (VT-3) examination of bolt devices (not locking devices on the 10-year ISI applicable to ANO-1) interval.Core Barrel All plants Cracking (IASCC, IE, Baffle-to-former bolts, Baseline volumetric (UT) no later 100% of accessible bolts"'. UT examination Assembly Overload)

3) Core barrel-to-former than two refueling outages from To be performed Baffle-to-former bolts the beginning of the license See Figure 4-2 of MRP-227-A in 1 R26.bolts renewal period with subsequent examination after 10 additional years.Core Barrel All plants Cracking (IE), including Core barrel cylinder Visual (VT-3) examination during 100 % of the accessible surface VT-3 examination Assembly the detection of readily (including vertical and the next 10-year SI. within 1 inch around each flow To be performed detectable cracking in the circumferential seam and bolt hole. in 1 R26.Baffle plates baffle plates welds), Subsequent examinations on the Former plates 10-year ISI interval.

See Figure 4-2 of MRP-227-A Report No. 1200459.401.R1 CStructural Integrity Associates, Inc?5-14 Table 5-1. B&W Plants Primary Category Components from Table 4-1 of MRP-227-A

[3, 32] (continued)

Item Applicability Effect (Mechanism)

Expansion Link) MEx m ination VI) EnspnCoverage sct MethodIFrequency

~ ~ EaiainCvrg Schedule Core Barrel All plants Cracking (IASCC, Locking devices, Visual (VT-3) examination during 100% of accessible baffle-to-VT-3 examination Assembly IE, Overload), including locking welds, the next 10-year ISI. former and internal baffle-to-baffle To be performed Locking devices, including the for the external baffle-to-bolt locking devices(2). in 1R26.including locking detection of missing, baffle bolts and Core Subsequent examinations on the welds, of baffle- non-functional, or barrel-to-former bolts 10-year ISI interval.

See Figure 4-2 of MRP-227-A.

to-former bolts removed locking and internal devices or welds baffle-to-baffle bolts Flow All plants Bolt: Cracking (SCC) UTS bolts and LTS Volumetric examination (UT) of the 100% of accessible bolts and UT examination Distributor Locking Devices: studs/nuts or bolts and bolts during the next 10-year ISI their locking devices(2). To be performed Assembly Loss of material, their locking devices, interval from 1/1/2006.

in 1R26.Flow distributor damaged or See Figure 4-8 of MRP-227-A.(FD) bolts and distorted or missing SSHT studs/nuts or bolts Subsequent examination on the their locking locking devices and their locking devices 10-year ISI interval unless an devices (Wear or Fatigue (not applicable to ANO-1) evaluation of the baseline results, damage by failed submitted for NRC staff approval, bolts). Lower grid shock pad justifies a longer interval between bolts and their locking examinations.

devices (not applicable to ANO-1) Visual (VT-3) examination of bolt locking devices on the 10-year ISI interval.Lower Grid All plants Cracking (SCC), Alloy X-750 dowel locking Initial visual (VT-3) examination no Accessible surface of 100% of the VT-3 examination Assembly including the welds to the upper and later than two refueling outages 24 dowel-to-guide block welds. To be performed Alloy X-750 detection of lower grid fuel assembly from the beginning of the license in 1 R26.dowel-to-guide separated or missing support pads. renewal period. See Figure 4-4 of MRP-227-A.

block welds locking welds, or missing dowels Subsequent examinations on the 10-year ISI interval Report No. 1200459.401.R1 5-15 CStructural Integrity Associates, Inc?

Table 5-1. B&W Plants Primary Category Components from Table 4-1 of MRP-227-A

[3, 32] (continued)

Item Applicability Effect (Mechanism)

Expansion LinkO)Examination Method/Frequency(1)Examination Coverage Inspection Schedule Incore Monitoring All plants Cracking (TE/IE), Lower grid fuel Initial visual (VT-3) examination no 100% of top surfaces of 52 spider VT-3 examination Instrumentation including the detection assembly support later than two refueling outages castings and welds to the To be performed (IMI) Guide Tube of fractured or missing pad items: pad, pad- from the beginning of the license adjacent lower grid rib section. in 1R26.Assembly spider arms or, to-rib section welds, renewal period.IMI guide tube Cracking (IE), Alloy X-750 dowel, See Figures 4-3 and 4-6 of spiders including separation cap screw, and their Subsequent examinations on the MRP-227-A.

IMI guide tube of spider arms from locking welds 10-year ISI interval.spider-to-lower grid the lower grid rib (Note: the pads, rib section welds section at the weld dowels, and cap screws are included because of IE of the welds)Notes: I. Examination acceptance criteria and expansion criteria for the B&W components are in Table 5-4 (Table 5-1 of MRP-227-A).

2. A minimum of 75% of the total population (examined

+ unexamined), including coverage consistent with the Expansion criteria in Table 5-4 (Table 5-1 of MRP-227-A), must be examined for inspection credit.3. The primary aging degradation mechanisms for loss ofjoint tightness for this item are IC and ISR. Fatigue and Wear, which can also lead to cracking, are secondary aging degradation mechanisms after significant stress relaxation and loss of preload has occurred due to IC/TSR. Bolt stress relaxation cannot readily be inspected by NDE. Only bolt cracking is inspected by UT inspection.

The effect of loss of joint tightness on the functionality will be addressed by analysis of the core barrel assembly, which will be performed to address Applicant/Licensee action item 6 in the SE [4].Report No. 1200459.401.R1 5-16 3Structural Integrity Associates, lnc Table 5-2. B&W Plants Expansion Category Components from Table 4-4 of MIRP-227-A

[3]Item Applicability Effect (Mechanism)

Primary Link"' Examination Examination Inspection Method/Frequency'" CoveragelFrequency" Schedule Upper Grid Assembly All plants Cracking (SCC), Alloy X-750 Visual (VT-3) examination.

Accessible surfaces of 100% Contingency if Alloy X-750 dowel-to- (except Davis including the detection dowel-to-guide Subsequent examinations of the dowel locking welds. indications are found upper grid fuel Besse) of separated or missing block welds on the 10-year ISI interval during the initial (VT-assembly support pad locking welds, or unless an applicant/licensee See Figure 4-6 of 3) examination in welds missing dowels provides an evaluation for MRP-227-A (i.e., these are 1R26.NRC staff approval that similar to the lower grid fuel justifies a longer interval assembly support pads).between inspections.

Core Barrel Assembly All plants Bolt or Stud/Nut:

UCB, LCB or FD Bolt or Stud/Nut:

Volumetric 100% of accessible bolts or Contingency if Upper thermal shield Cracking (SCC) bolts and their examination (UT). studs/nuts and their locking indications are found (UTS) bolts and their Locking Devices: Loss locking devices devices(2). in UT exam of bolts in locking devices of material, damaged, Locking Devices: Visual 1R26.distorted or missing (VT-3) examination See Figure 4-7 of Core Barrel Assembly Not applicable to locking devices (Wear MRP-227-A Surveillance specimen ANO-1 or Fatigue damage by Subsequent examinations holder tube (SSHT) failed bolts). on the 10-year ISI interval studs/nuts (or bolts and unless an applicant/licensee their locking devices provides an evaluation for NRC staff approval that justifies a longer interval between inspections Core Barrel Assembly All plants Cracking (IE), including Baffle plates No examination Inaccessible.

Contingency if Core barrel cylinder readily detectable requirements.

Justify by indications are found (including vertical and cracking evaluation or by See Figure 4-2 of in VT-3 exam of circumferential seam replacement.

MRP-227-A Baffle plates in 1R26.welds)Former plates Report No. 1200459.401.R1 CStructural Integrity Associates, IncO 5-17 Table 5-2. B&W Plants Expansion Category Components from Table 4-4 of MRP-227-A

[3] (continued)

Item Applicability Effect (Mechanism)

Primary Link Examination Examination Inspection Method/Frequency (1) Coverage/Frequency (1) Schedule Core Barrel Assembly All plants Cracking (IASCC, IE, Baffle-to-former Internal baffle-to-baffle An acceptable examination Contingency if Baffle-to-baffle bolts Overload)(3) bolts bolts: technique currently not indications are found Core barrel-to-former No examination available, in UT exam of baffle-bolts requirements, to-former bolts in Justify by evaluation or by See Figure 4-2 of 1R26.replacement.

MRP-227-A.

External baffle-to-baffle Inaccessible.

bolts, core barrel-to-former bolts: No examination See Figure 4-2 of requirements.

Justify by MRP-227-A.

evaluation or by replacement.

Core Barrel Assembly All plants Cracking (IASCC, IE) Locking devices, No examination Inaccessible.

Contingency if Locking devices, including locking requirements.

indications are found including locking welds, welds, of baffle- Justify by evaluation or See Figure 4-2 of in VT-3 exam of for the external baffle- to-former bolts replacement.

MRP-227-A.

locking devices, to-baffle bolts and core including locking barrel-to-former bolts or internal baffle- welds of baffle-to-to baffle bolts former bolts or internal baffle-to baffle bolts in 1R26 Lower Grid Assembly All plants Cracking (IE), including IMI guide tube Visual (VT-3) examination.

Accessible surfaces of the Contingency if Lower grid fuel the detection of spiders and Subsequent examinations pads, dowels, and cap indications are found assembly support pad separated or missing spider-to-lower on the 10-year ISI interval screws, and associated in VT-3 examination items: pad, pad-to-rib welds, missing support grid rib section unless an applicant/licensee welds in 100% of the lower of IMI guide tube section welds, Alloy X- pads, dowel, cap welds provides an evaluation for grid fuel assembly support spiders and spider-to-750 dowel, cap screw, screws and locking NRC staff approval that pads. lower grid rib section and their locking welds welds, or misalignment justifies longer interval welds in 1R26.of the support pads between inspections.

See Figure 4-6 of (Note: the pads, dowels MRP-227-A.

and cap screws are included because of IE of the welds)Report No. 1200459.401.R1 5-18* Structural Integrity Associates, Inc?

Table 5-2. B&W Plants Expansion Category Components from Table 4-4 of MRP-227-A

[3] (continued)

Item Applicability Effect (Mechanism)

Primary Link Examination Examination Inspection Method/Frequency (1) Coverage/Frequency (1) Schedule Lower Grid Assembly All plants Cracking (SCC), including Alloy X-750 Visual (VT-3) examination.

Accessible surfaces of Contingency if Alloy X-750 dowel-to-the detection of separated dowel-to-guide Subsequent examinations on 100% of the support pad indications are found lower grid fuel assembly or missing locking welds, block welds. the 10-year ISI interval unless dowel locking welds, in VT-3 examination support pad welds or missing dowels. an applicant/licensee provides of Alloy X-750 dowel-an evaluation for the NRC staff See Figure 4-6 of to-guide block welds approval that justifies a longer MRP-227-A in 1R26 interval between inspections.

Lower Grid Assembly Not applicable to Bolts: Cracking (SCC) UCB, LCB or Bolt: Volumetric examination 100% of the accessible Contingency if Lower grid shock pad ANO-1 Locking Devices: Loss of FD bolts and (UT). bolts and their locking indications are found bolts and their locking material, damaged, their locking Locking Devices: Visual (VT-3) devices(2).

in UT exam of bolts in devices distorted or missing devices examination.

1R26.locking devices (Wear or Subsequent examinations on See Figure 4-4 of Fatigue damage of failed the 10-year ISI interval unless MRP-227-A.

bolts). an applicant/licensee provides an evaluation for NRC staff approval that justifies a longer interval between inspections.

Lower Grid Assembly All plants Bolts: Cracking (SCC) UCB, LCB or Bolt: Volumetric examination 100% of the accessible Contingency if Lower thermal shield FD bolts and (UT). bolts and their locking indications are found (LTS) bolts and their Locking Devices: Loss of their locking Locking Devices: Visual (VT-3) devices(2). in UT exam of bolts in locking devices material, damaged, devices examination.

1R26.distorted or missing Subsequent examinations on See Figure 4-8 of Studs/nuts: (Not locking devices (Wear or the 10-year ISI interval unless MRP-227-A.

applicable to ANO-1) Fatigue damage of failed an applicant/licensee provides bolts). an evaluation for NRC staff approval that justifies a longer interval between inspections.

Notes: I1. Examination acceptance criteria and expansion criteria for B&W components are in Table 5-4 (MRP-227-A Table 5-1).2. A minimum of 75% of the total population (examined

+ unexamined) must be examined for inspection credit.3. The primary aging degradation mechanisms for loss of joint tightness of these items are IC and ISR. Fatigue and Wear, which can also lead to cracking, are secondary aging degradation mechanisms after significant stress relaxation and loss of preload has occurred due to IC/ISR. Bolt stress relaxation cannot readily be inspected by NDE. Only bolt cracking could be inspected by UT inspection if it were possible for these bolts. Therefore, the effects of loss of joint tightness and/or cracking on the functionality of these bolts relative to the entire core barrel assembly will be addressed by analysis of the core barrel assembly, which will be performed to address Applicant/Licensee action item 6 of the SE [4].Report No. 1200459.401.R1 5-19 C Integrity Associates, IncO Table 5-3. B&W Plants Existing Program Components from AREVA Guidance [32]Item Applicability Effect (Mechanism)

Reference Examination Method Examination Coverage Core Support Shield All plants Loss of material from ASME Code Visual (VT-3) examination 100% of accessible surfaces Assembly locking device Section XI during each 10-year ISI CSS vent valve (associated with jack interval (Note 1)miscellaneous locking screw spring and devices (original design) pressure plate)Core Support Shield All plants Cracking (SCC), ASME Code Visual (VT-3) examination 100% of accessible surfaces Assembly including detection of Section XI during each 10-year ISI CSS vent valve fractured or missing interval (Note 1)miscellaneous locking locking cups and devices (modified welds, or the bolted design) block Note: 1. In addition to the Existing Program vent valve VT-3 examinations identified in this table, the following testing and inspection requirements are currently performed and will continue to be performed by the B&W units: A verification of the operation of each vent valve shall also be performed through manual actuation of the valve. Verify that the valves are not stuck in the open position and that no abnormal degradation has occurred.

Examine the valves for evidence of leakage between the valve disc and the valve body (i.e., flow lines across the sealing surface), cracking of lock welds and locking cups, jack screws for proper position, and wear. The frequency is defined in each unit's technical specifications or in their pump and valve inservice test programs (see BAW-2248A, page 4.3 and Table 4-1 [9]).Report No. 1200459.401.Rl 5-20 c Structural Integrity Associates, Inc?

Table 5-4. B&W Plants Examination Acceptance and Expansion Criteria from Table 5-1 of MRP-227-A

[3] Applicable to ANO-1 Examination Item Applicability Acceptance Criteria Expansion Link(s) Expansion Criteria Additional Examination (Note 1) Acceptance Criteria Plenum Cover Assembly All plants One-time physical None N/A N/A& Core Support Shield measurement.

In addition, a Assembly visual (VT-3) examination is Plenum cover weldment rib conducted for these items.pads Plenum cover support The measured differential flange height from the top of the CSS top flange plenum rib pads to the vessel seating surface shall average less than 0.004 inches compared to the as-built condition The specific relevant condition for these items is wear that may lead to loss of function.Core Support Shield All plants Visual (VT-3) examination.

None N/A N/A Assembly The specific relevant CSS vent valve top retaining condition is evidence of ring damaged or fractured CSS vent valve bottom retaining ring retaining ring material, and missing items.Control Rod Guide Tube All plants The specific relevant None N/A N/A Assembly condition for the VT-3 of the CRGT spacer castings CRGT spacer castings is evidence of fractured spacer or missing screws.V Structural Integrity Associates, Inc.?Report No. 1200459.401.Rl 1 5-21 Table 5-4. B&W Plants Examination Acceptance Criteria from Table 5-1 of MRP-227-A

[3] Applicable to ANO-1 (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Core Support Shield All plants 1) Volumetric (UT) UTS bolts and LTS 1) Confirmed unacceptable indications

1) The examination Assembly examination of the UCB bolts or studs/nuts exceeding 10% of the UCB bolts shall acceptance criteria for Upper core barrel (UCB) bolts. and their locking require that the UT examination be the UT of the expansion bolts and their locking devices expanded by the completion of the next bolting shall be devices. The examination refueling outage to include: established as part of acceptance criteria for the SSHT studs/nuts or the examination UT of the UCB bolts shall be bolts and their 100% of the accessible UTS bolts and technical justification.

established as part of the locking devices (not 100% of the accessible LTS bolts or examination technical applicable to ANO-1) studs/nuts, 2) The specific relevant justification.

2) Confirmed evidence of relevant condition for the VT-3 of conditions exceeding 10% of the UCB the expansion locking 2) Visual (VT-3) examination bolt locking devices shall require that devices is evidence of of the UCB bolt locking Lower grid shock pad the VT-3 examination be expanded by broken or missing bolt devices, bolts and their the completion of the next refueling locking devices.locking devices (not outage to include: The specific relevant applicable to ANO-1)condition for the VT-3 of the 100% of the accessible UTS bolt and UCB bolt locking devices is 100% of the accessible LTS bolt or evidence of broken or stud/nut locking devices, missing bolt locking devices.Report No. 1200459.401.Rl 5-22 VStructural Integrity Associates, Inc.

Table 5-4. B&W Plants Examination Acceptance Criteria from Table 5-1 of MRP-227-A

[3] Applicable to ANO-1 (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Core Barrel Assembly All plants 1) Volumetric (UT) UTS bolts and LTS 1) Confirmed unacceptable indications

1) The examination Lower core barrel (LCB) examination of the LCB bolts or studs/nuts exceeding 10% of the LCB bolts shall acceptance criteria for bolts and their locking bolts. and their locking require that the UT examination be the UT of the expansion devices The examination devices expanded by the completion of the next bolting shall be acceptance criteria for the SSHT studs/nuts or refueling outage to include: established as part of UT of the LCB bolts shall be bolts and their the examination established as part of the locking devices (not 100% of the accessible UTS bolts and technical justification.

examination technical applicable to ANO-1) 100% of the accessible LTS bolts or justification.

studs/nuts, 2) The specific relevant Lower grid shock pad condition for the VT-3 of 2) Visual (VT-3) examination bolts and their the expansion locking of the LOB bolt locking locking devices (not devices, applicable to ANO-1o) 2) Confirmed evidence of relevant devices is evidence of conditions exceeding 10% of the LCB broken or missing bolt The specific relevant bolt locking devices shall require that locking devices.condition for the VT-3 of the the VT-3 examination be expanded by LCB bolt locking devices is the completion of the next refueling evidence of broken or outage to include: missing bolt locking devices.100% of the accessible UTS bolt and 100% of the accessible LTS bolt or stud/nut locking devices, CStructural Integrity Associates, Inc!Report No. 1200459.401.Rl 5-23 Table 5-4. B&W Plants Examination Acceptance Criteria from Table 5-1 of MRP-227-A

[3] Applicable to ANO-1 (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Core Barrel Assembly All plants Baseline volumetric (UT) Baffle-to baffle bolts, Confirmed unacceptable indications N/A Baffle-to-former bolts examination of the baffle-to-greater than or equal to 5% (or 43) of the former bolts. Core barrel-to-former baffle-to-former bolts, provided that none bolts of the unacceptable bolts are on formed The examination elevations 3, 4, and 5, or greater than acceptance criteria for the 25% of the bolts on a single baffle plate, UT of the baffle-to-former shall require an elevation of the internal bolts shall be established as baffle-to-baffle bolts for the purpose of part of the examination determining whether to examine or technical justification.

replace the internal baffle-to-baffle bolts.The evaluation may include external baffle-to-baffle bolts and core barrel-to-former bolts for the purpose of determining whether to replace them Core Barrel Assembly All plants Visual (VT-3) examination

a. Former plates a and b. Confirmed cracking in multiple a and b. N/A Baffle plates (2 or more) locations in the baffle plates The specific relevant b. Core barrel cylinder shall require expansion, with continued condition is readily (including vertical and operation of former plates and the core detectable cracking in the circumferential seam barrel cylinder justified by elevation or by baffle plates. welds) replacement by the completion of the next refueling outage.Core Barrel Assembly All plants Visual (VT-3) examination.

Locking devices, Confirmed relevant conditions in greater N/A Locking devices, including including locking than or equal to 1% (or 11) of the baffle-locking welds, of baffle-to-This specific relevant welds for the external to-former or internal baffle-to-baffle bolt former bolts and internal condition is missing, non- baffle-to-baffle bolts locking devices, including locking welds, baffle-to-baffle bolts functional, or removed and core barrel-to-shall require an evaluation of the external locking devices, including former bolts baffle-to-baffle and core barrel-to-former locking welds. bolt locking devices for the purpose of determining continued operation or replacement.

Lower Grid Assembly All plants Initial visual (VT-3) Alloy X-750 dowel Confirmed evidence of relevant The specific relevant Alloy X-750 dowel-to-guide examination, locking welds to the conditions at two or more locations shall condition for VT-3 of the block welds The specific relevant upper and lower grid require that the VT-3 examination be expansion dowel locking condition is separated or fuel assembly support expanded to include Alloy X-750 dowel weld is separated or missing locking weld, or pads locking welds to the upper and lower grid missing locking weld, or missing dowel. fuel assembly support pads by the missing dowel.completion of the next refueling outage.Report No. 1200459.401.R5 5-24 CStructural Integrity Associates, Inc.8 Table 5-4. B&W Plants Examination Acceptance Criteria from Table 5-1 of MRP-227-A

[3] Applicable to ANO-1 (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Flow Distributor Assembly All plants 1) Volumetric (UT) UTS bolts and LTS bolts or 1) Confirmed unacceptable indications

1) The examination Flow distributor (FD) bolts examination of FD bolts. studs/nuts and their locking exceeding 10% of the FD bolts shall acceptance criteria for the and their locking devices devices require that the UT examination be UT of the expansion The examination acceptance expanded by the completion of the bolting shall be criteria for the UT of the FD SSHT studs/nuts or bolts expand bythe to ing shal be bolts shall be established as and their locking devices next refueling outage to include: established as part of the part of the examination (not applicable to ANO-1) examination technical technical justification.

100% of the accessible UTS bolts and justification.

Lower grid shock pad bolts 100% of the accessible LTS bolts or 2) Visual (VT-3) examination and their locking devices studs/nuts, 2) The specific relevant of the FD bolt locking (not applicable to ANO-1) condition for the VT-3 of devices. the expansion locking The specific relevant 2) Confirmed evidence of relevant devices is evidence of condition for the VT-3 of the conditions exceeding 10% of the FD broken or missing bolt FD bolt locking devices is bolt locking devices shall require that locking devices.evidence of broken or the VT-3 examination be expanded by missing bolt locking devices, the completion of the next refueling outage to include: 100% of the accessible UTS bolt and 100% of the accessible LTS bolt or stud/nut locking devices, Report No. 1200459.401.R2 5-25 4Structural Integrity Associates, Inc!

Table 5-4. B&W Plants Examination Acceptance Criteria from Table 5-1 of MRP-227-A

[3] Applicable to ANO-1I (continued)

Item Applicability Examination Expansion Link(s) Expansion Criteria Additional Examination Acceptance Criteria Acceptance Criteria (Note 1)Incore Monitoring All plants Initial visual (VT-3) Lower fuel grid Confirmed evidence of relevant The specific relevant Instrumentation (IMI) examination, assembly support pad conditions at two or more IMI guide tube conditions for the VT-3 of Guide Tube Assembly items: pad, pad-to-rib spider locations or IMI guide tube spider- the lower grid fuel IMI guide tube spiders The specific relevant section welds, Alloy to-lower grid rib section welds shall assembly support pad IMI guide tube spider-to-conditions for the IMI guide X-750 dowel, cap require that the VT-3 examination be items (pads, pad-to-rib lower grid rib section welds tube spiders are fractured or screw, and their expanded to include lower fuel assembly section welds, Alloy X-750 missing spider arms. locking welds support pad items by the completion of dowels, cap screws, and the next refueling outage, their locking welds) are The specific relevant separated or missing conditions for the IMI spider- welds, missing support to-lower grid rib section pads, dowels, cap screws welds are separated or and locking welds, or missing welds misalignment of the support pads.Note: 1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

Report No. 1200459.401.RI 5-26 CStructural Integrity Associates, Inc?

Table 5-5. ANO-1 Response to the NRC Final Safety Evaluation of MRP-227-A

[4]MRP-227 SE Item ANO-1 Response SE Section 4.1.1, Topical Report Not applicable to ANO- 1.Condition 1: Moving components from "No Additional Measures" to"Expansion" category.SE Section 4.1.2, Topical Report Not applicable to ANO-1.Condition 2: Inspection of components subject to irradiation-assisted stress corrosion cracking.SE Section 4.1.3, Topical Report In accordance with SE Section 4.1.3, flow distributor-to-shell forging bolts Condition 3: Inspection of high have been added to the ANO-1 "Primary" inspection category and are consequence components subject to contained in Table 5-1 of this document (consistent with Table 4-1 of multiple degradation mechanisms MRP-227-A).

Additionally, in accordance with Section 4.1.3 of the SE, reference to Note 3 has been removed from the "Expansion Link" column and the "Expansion Category" reference for the LCB bolts and FD bolts from the "Expansion Link" column of the UCB bolt line has also been deleted. Consistent with SE Sections 3.3.7 and 4.2.7, the CRGT spacer castings (Primary component) will be subject to Applicant/Licensee Action Item No. 7.SE Section 4.1.4, Topical Report In accordance with SE Section 4.1.4, ANO-1 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 of this document (consistent with Table 4-4 of"expansion" inspection category MRP-227-A) for examination coverage.components SE Section 4.1.5, Topical Report In accordance with SE Section 4.1.5, ANO-1 will meet the 10-year Condition 5: Examination inspection frequency of the specified components following the initial or frequencies for baffle former bolts baseline inspection unless an evaluation is provided for NRC staff approval and core shroud bolts that justifies a longer interval between inspections.

Table 5-1 of this document (consistent with Table 4-1 of MRP-227-A) reflects the changes made in the NRC approved version of MRP-227 [3].Structural Integrity Associates, Inc.Report No. 1200459.401.Rl 5-27 Table 5-5. ANO-1 Response to NRC Final Safety Evaluation of MRP-227-A

[4] (continued)

MRP-227 SE Item ANO-1 Response SE Section 4.1.6, Topical Report In accordance with SE Section 4.1.6, ANO-1 will meet the 10-year re-Condition 6: Periodicity of the re- examination interval to all "Expansion" inspection category components examination of "expansion" once degradation is identified in the associated "Primary" inspection inspection category components category component and examination of the "Expansion" category component commences unless an evaluation is provided for NRC staff approval that justifies a longer interval between inspections.

Table 5-2 of this document (consistent with Table 4-4 of MRP-227-A) reflects the changes made in the NRC approved version of MRP-227 [3].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 plant design and operating history demonstrating that Applicant/Licensee Action Item 1: MRP-227-A is applicable to ANO-1 is contained in Section 1.8.4.1 and Applicability of FMECA and Section 5.1 of this document.Functionality Analysis Assumptions SE Section 4.2.2, The ANO-1 review of components within the scope of license renewal was Applicant/Licensee Action Item 2: compared against the information contained in MRP-189, Tables 4-1 and 4-PWR Vessel Internals Components

2. In addition to this review, AREVA identified an existing program for the Within the Scope of License vent valve locking devices [32]. The existing program is contained in Table Renewal 5-3. The Aging Management Review performed as part of the ANO-1 LRA is described in Section 1.7.1 of this document and summarized as part of Applicant/Licensee Action Item 2 in Section 5.2. Three components were added as inspection items based on their aging mechanisms.

These components will be examined until the screening and categorization is completed.

SE Section 4.2.3, No action required.

This action does not apply to B&W designed units.Applicant/Licensee Action Item 3: Evaluation of the Adequacy of Plant-Specific Existing Programs SE Section 4.2.4, A records search was performed.

The records search confirmed that the Applicant/Licensee Action Item 4: upper flange was stress relieved [30, 31] as discussed in Section 5.4 of this B&W Core Support Structure document.Upper Flange Stress Relief rStructural Integrity Associates, Inc!Report No. 1200459.401.Rl 5-28 Table 5-5. ANO-1 Response to NRC Final Safety Evaluation of MRP-227-A

[4] (continued)

MRP-227 SE Item ANO-1 Response SE Section 4.2.5, The ANO-1 plant specific acceptance criteria for the plenum cover Applicant/Licensee Action Item 5: weldment rib pads, plenum cover support flange and CSS top flange for Application of Physical this one time physical measurement shall be an average measured Measurements as part of I&E differential height from the top of the plenum rib pads to the vessel seating Guidelines for B&W, CE and surface of 0.004 inches relative to the as-built condition.

The full Westinghouse RVI Components discussion of the physical measurement is discussed in Section 5.5 of this document.SE Section 4.2.6, ANO-l will justify the acceptability of inaccessible and non-inspectable Applicant/Licensee Action Item 6: components through the period of extended operation by performing an Evaluation of Inaccessible B&W evaluation, proposing a schedule for replacement, or justification for some Components other alternative process for these components as discussed in Section 5.6.SE Section 4.2.7, ANO-I has RVI components that are fabricated from CASS, martensitic Applicant/Licensee Action Item 7: stainless steel and precipitation hardened materials.

This A/LAI will be Plant Specific Evaluation of CASS addressed based on further screening and an analytical approach that Materials assesses the effect of reduction in fracture toughness.

The need for an inspection method will depend on the result of the analyses performed as described 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.VStructural Integrity Associates, Inc!Report No. 1200459.401.Rl 5-29 Table 5-6. ANO-I Program Enhancement and Implementation Schedule Refueling AMP-Related Scope Inspection Method and Criteria Comments Outage Period of Extended Operation begins I R24 Not Applicable Not Applicable May 20, 2014 1R25 Not Applicable Not Applicable Not Applicable" ASME Section XI ISI Program B-N-3 components:

e Inspections in accordance with Plenum assembly:

ANO-I ISI program o plenum cover assembly e MRP-227-A inspections in o plenum cylinder assembly accordance with MRP-228.o upper grid assembly o control rod guide tube assemblies Core support assembly: o core support shield assembly o core barrel assembly o lower internals assembly:-lower grid assembly-flow distributor assembly-IMI guide tube assembly" MRP-227-A Primary Components:

IR26 Plenum Cover Assembly:

Not Applicable o Plenum cover weldment rib pads o Plenum cover support flange Control Rod Guide Tube Assembly: o CRGT spacer castings Core Support Shield Assembly: o CSS top flange o CSS vent valve top retaining ring o CSS vent valve bottom retaining ring o UCB bolts and their locking devices Core Barrel Assembly: o LCB bolts and their locking devices o Baffle-to-former bolts o Baffle plates o Locking devices, including locking welds, of baffle-to-former bolts and internal baffle-to-baffle bolts CStructural Integrity Associates, Report No. 1200459.401.R1 5-30 Table 5-6. ANO-1 Program Enhancement and Implementation Schedule (continued)

Refueling AMP-Related Scope Inspection Method and Criteria Comments Outage Flow Distributor Assembly:

• Inspections in accordance with o FD bolts and their locking devices ANO-I ISI program Lower Grid Assembly:

e MRP-227-A inspections in o Alloy X-750 dowel-to-guide block welds accordance with MRP-228.Incore Monitoring Instrumentation Guide Tube Assembly: o IMI guide tube spiders o IMI guide tube spider-to-lower grid rib section welds I R26 Existing Program Components: (Continued)

Core Support Shield Assembly:

Not Applicable o CSS vent valve miscellaneous locking devices (original design)o CSS vent valve miscellaneous locking devices (modified design)ANO-I Orphan Components:

o RVLMS probe supports o Surveillance specimen holder tubes o Thermal shield and thermal shield upper restraint I R27 Not Applicable Not Applicable Not Applicable 1R28 Not Applicable Not Applicable Not Applicable 1R29 Not Applicable Not Applicable Not Applicable I R30 Not Applicable Not Applicable Not Applicable 1R31 Not Applicable Not Applicable Not Applicable 1 R32 Not Applicable Not Applicable Not Applicable.

• ASME Section XI IS Program B-N-3 components:

a Inspections in accordance with 1R33 Plenum assembly:

ANO-I ISI program Not Applicable o plenum cover assembly 9 MRP-227-A inspections in o plenum cylinder assembly accordance with MRP-228.CStructural Integrity Associates, Inc!Report No. 1200459.401.R1 5-31 Table 5-6. ANO-1 Program Enhancement and Implementation Schedule (continued)

Refueling AMP-Related Scope Inspection Method and Criteria Comments Outage o upper grid assembly

• Inspections in accordance with o control rod guide tube assemblies ANO-I ISI program Core support assembly:

• MRP-227-A inspections in o core support shield assembly accordance with MRP-228.o core barrel assembly o lower internals assembly:-lower grid assembly-flow distributor assembly-IMI guide tube assembly MRP-227-A Primary Components:

Plenum Cover Assembly: o Plenum cover weldment rib pads o Plenum cover support flange Control Rod Guide Tube Assembly: o CRGT spacer castings I R33 Core Support Shield Assembly:

Not Applicable (Continued) o CSS top flange o CSS vent valve top retaining ring o CSS vent valve bottom retaining ring o UCB bolts and their locking devices Core Barrel Assembly: o LCB bolts and their locking devices o Baffle-to-former bolts o Baffle plates o Locking devices, including locking welds, of baffle-to-former bolts and internal baffle-to-baffle bolts Flow Distributor Assembly: o FD bolts and their locking devices Lower Grid Assembly: o Alloy X-750 dowel-to-guide block welds Incore Monitoring Instrumentation Guide Tube Assembly: o IMI guide tube spiders o IMI guide tube spider-to-lower grid rib section welds IStructural Integrity Associates, Inc!Report No. 1200459.401.Rl 5-32 Table 5-6. ANO-1 Program Enhancement and Implementation Schedule (continued)

Refueling AMP-Related Scope Inspection Method and Criteria Comments Outage Existing Program Components:

e Inspections in accordance with Core Support Shield Assembly:

ANO-I ISI program o CSS vent valve miscellaneous locking devices (original 9 MRP-227-A inspections in design) accordance with MRP-228.IR33 o CSS vent valve miscellaneous locking devices (modified (Continued) design) Not Applicable ANO-1 Orphan Components:

o RVLMS probe supports o Surveillance specimen holder tubes o Thermal shield and thermal shield upper restraint 1R34 Not Applicable Not Applicable Not Applicable 1R35 Not Applicable Not Applicable Not Applicable I R36 Not Applicable Not Applicable Not Applicable IR37 Not Applicable Not Applicable Not Applicable 1R38 Not Applicable ot Applicable Renewed Operating License Expires p _May 20, 2034 CStructural Integrity Associates, Inc?Report No. 1200459.401.R3 5-33

6.0 REFERENCES

1. EN-DC-202, Rev. 6, "NEI 03-08 Materials Initiative Process," Entergy Nuclear Management Manual, 11/21/13 (SI File No. 1200459.201).

2. Nuclear Energy Institute, "Revision 2 to NEI 03-08, Guideline for the Management of Materials Issues," dated January, 2010 (SI File No. 1200459.202).

3. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), EPRI, Palo Alto, CA: 2008. 1022863 (SI File No.1200459.203P).

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.MLl 1308A770 (SI File No. 1200459.204).

5. License Renewal Application

-Arkansas Nuclear One -Unit 1 (SI File No. 1200459.205).

6. Arkansas Nuclear One -Unit 1 Safety Analysis Report Amendment 25, Facility Operating License Number DPR-5 1, Docket Number 50-313 (SI File No. 1200459.206).

7. NUREG- 1801, "Generic Aging Lessons Learned (GALL) Report," Rev. 2, U. S. Nuclear Regulatory Commission, December 2010 (SI File No. 1200459.207).

8. NUREG-1800, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, September 2005 (SI File No.1200459.208).

9. The B&W Owners Group Licensing Renewal Task Force, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," BAW-2248A, March 2000, ADAMS Accession No. ML003708443 (SI File No. 1200459.209).

10. Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals Component Items (MRP-189-Rev.

1), EPRI, Palo Alto, CA: 2006, 1018292 (SI File No. 1200459.2 1OP).I'I~StructuraI Integrity Associates, lnc?Report No. 1200459.401.Rl 6-1 Sruta

11. Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228).

EPRI, Palo Alto, CA: 2009. 1016609 (SI File No. 1200459.211).

EPRIPROPRIETARY.

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

EPRI, Palo Alto, CA: 2005.1012081 (SI File No. 1200459.212).

13. WCAP- 17096-NP, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009 (SI File No. 1200459.213).

14. ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition through 2003 Addenda.15. Entergy Nuclear Engineering Programs, SEP-ISI-ANOl

-101, Revision No. 0, "Program Section for ASME Section XI, Division 1 ANO 1 Inservice Inspection Program," (SI File No. 1200459.214).

16. Entergy Operations Incorporated Arkansas Nuclear One Document No. 1000.106, Change No. 010, "Primary Chemistry Monitoring Program," (SI File No. 1200459.215).

17. "Pressurized Water Reactor Primary Water Chemistry Guidelines," Volumes 1 and 2, Revision 6, Electric Power Research Institute, Palo Alto, CA: 2007, 1014986.18. Arkansas Nuclear One Engineering Report 93-R-1013-08, "Demonstration of the Management of Aging Effects for the ANO- 1 Reactor Vessel Internals," (SI File No.1200459.216).

19. Safety Evaluation Report Related to the License Renewal of Arkansas Nuclear One, Unit 1, Docket No. 50-313, SR1743, (SI File No. 1200459.217).

20. U.S. Code of Federal Regulations, "Title 10, Energy, Part 50.54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants." 21. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev.

0), EPRI, Palo Alto, CA: 2008. 1016596 (SI File No.1200459.218).

22. WCAP-17495-NP, Revision 0, "Reactor Vessel Internals Program Plan for Aging Management of Reactor internals at Arkansas Nuclear One, Unit 1, December 2011. (SI File No. 1200459.227)

Report No. 1200459.401 .R1 6-2 r Structural

23. Letter from US NRC to Arkansas Power and Light Company dated November 24, 1986,"Amendment to Facility Operating License: Amendment No. 105," ADAMS Accession No.ML021230036 (SI File No. 1200459.219).

24. Letter from US NRC to Arkansas Power and Light Company dated November 8, 1988,"Issuance of Amendment No. 113 to Facility Operating License No. DPR-51 -Arkansas Nuclear One, Unit No. 1 (TAC No. 69056)," ADAMS Accession No. ML021230267 (SI File No. 1200459.220).

25. "Applicant's Environmental Report -Operating License Renewal Stage -Arkansas Nuclear One -Unit I," (SI File No. 1200459.221).

26. NUREG-1801, "Generic Aging Lessons Learned (GALL) Report," Rev. 0, U. S. Nuclear Regulatory Commission, July 2001 (SI File No. 1200459.222).

27. EN-LI-102, Rev. 23, "Corrective Action Process," Entergy Nuclear Management Manual, 11/21/13 (SI File No. 1200459.223).

28. 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.29. EN-OE-100, Rev. 20, "Operating Experience Program," Entergy Nuclear Management Manual, 11/21/13 (SI File No. 1200459.224).

30. AREVA Engineering Information Record No. 51-9196971-000, Rev. 0, "Reactor Vessel Internals Welds Stress Relief Records Search for ANO- 1," AREVA PROPRIETARY (SI File No. 1200459.228P).

31. Entergy Corrective Action LR-LAR-2010-00176, SI File No. 1200459.228.

32. AREVA Letter No. AREVA- 13-01501, "Recommended Examination of Vent Valve Locking Devices for B&W Nuclear Units," (SI File No. 1200459.228).

33. EN-DC-133, Rev. 0, "PWR Vessel Internals Program," Entergy Nuclear Management Manual, 3/31/14 (SI File No. 1200459.234).

Report No. 1200459.401.R1 6-3 Structural-F

34. Arkansas Nuclear One Engineering Report No. CALC-ANO1-ME-i 1-00026, Rev. 0,"Review of the Reactor Vessel Internals Aging Management Program for License Renewal Implementation," (SI File No. 1200459.235).

35. "Information in Support of the EPRI Materials Reliability Program (MRP): Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev.

0) Review," October 29, 2010, NRC Accession Number ML103090248.

36. ANP-3281P, Revision 1, "Time-Limited Aging Analysis Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One Unit 1 at 60 Years," Licensing Report, March 2014. AREVA PROPRIETARY. (SI File No. 1200459.233P).

37. Letter from James F. Mallay to NRC Document Control Desk, "Submittal of BAW-2241 P, Revision 2, Fluence and Uncertainty Methodologies," June 2, 2003, NRC Accession Number ML031550365.

38. BAW-2241NP-A, Revision 2, "Fluence and Uncertainty Methodologies," April 2006, NRC Accession Number ML073310660.

39. Letter from Ronnie L. Gardner to NRC Document Control Desk, "Publication of Revision 1 of Appendix G to BAW-2241(P), Revision 2, "Fluence and Uncertainty Methodologies,'"" November 20, 2007, NRC Accession Number ML073310655.

Renort No. 1200459.401.Rl 6-4 4 Sructura i--r ......................

APPENDIX A SECTION XI 10 YEAR ISI EXAMINATIONS OF B-N-3 INTERNALS COMPONENTS FOR ANO-1 [15, 18]Report No. 1200459.401.R1 A-1 r Structural Integrity Associates, ln0 Table A-1.Section XI 10 Year ISI Examinations of B-N-3 Internals Components for ANO-1 115, 18]Item Fabrication ID Examination Area Code Exam , D i Examination Percentage Required No. Description Category Method REMOVABLE CORE SUPPORT STRUCTURES Plenum Assembly 100% of accessible surfaces -every B 13.70 Plenum Cover Assembly B-N-3 VT-3 interval 100% of accessible surfaces -every B 13.70 Plenum Cylinder Assembly B-N-3 VT-3 interval 100% of accessible surfaces -every B 13.70 Upper Grid Assembly B-N-3 VT-3 interval 100% of accessible surfaces -every B 13.70 Control Rod Guide Tube Assemblies B-N-3 VT-3 interval Core Support Assembly B 13 100% of accessible surfaces -every B13.70I Core Support Shield Assembly I B-N-3 VT-3 interval B 1l t r100% of accessible surfaces -every B 13.70 Core Barrel Assembly I B-N-3 VT-3 interval Lower Internals Assembly 100% of accessible surfaces -every B 13.70 Lower Grid Assembly B-N-3 VT-3 interval 100% of accessible surfaces -every B 13.70 Flow Distributor Assembly B-N-3 VT-3 interval 100% of accessible surfaces -every B13.70 IMI Guide Tube Assembly B-N-3 VT-3 interval Report No. 1200459.401.R1 A-2 VStructural Integrity Associates, Inc.

APPENDIX B MRP-189, REV. 1 COMPONENT REVIEW FOR ANO-1 [3, 10, 181 Report No. 1200459.401 .R1 B-1 W Structural Integrity Associates, Inc?Inegiy1soiteIc Table B-1. Summary of Aging Management Evaluations for ANO-1 Reactor Vessel Internals

[3, 10, 181 Component/

AigEfc gn urn Commodityi Aging Effect Aging Classification Classification Simil Final Augmented Inspection Current MRP-189, Rev A I Description Requiring Management in MRP-189 in MRP-227-A Classification Requirements eXams Table 5-1 Management Programs Exams Examination One-time physical Category B-N-Similar to measurement no later 3, Removable Plenum Cover Simi tonthan two refueling outages Core Support Aseby Plenum Rib Lsof ASME Section Xl AemblyRib Pads in Los f rPrimary per from the beginning of the Structures In MRP-227-A and Wedmen R abb e Mti oL/We ISI Program, Primary N/A MRP-227-A LR period. VT-3 visual Section Xl Programs PAbl24-A oPerform subsequent visual examination of BAW-2248A (VT-3) examination on the accessible 10-year IS[ interval, surface -Every interval Similar to the Examination FuelExmnto Assembly One-time physical Category B-N-Assembl tmeasurement no later 3, Removable Similar to Fuel Support Pads than two refueling outages Core Support Plenum Cover Assembly ASME Section Xl in the Section Primary per from the beginning of the Structures In MRP-227-A and Assembly Support Pads L/Wear ISI Program, Primary XlISI bMateal/Wear P C Pia MRP-227-A LR period. VT-3 visual Section Xl Programs Support Flange in Table 4-1 of RVI AMP Program and Perform subsequent visual examination of Assembly in (VT-3) examination on the accessible Table 4-1 of 10-year IS] interval, surface -Every BAW-2248A interval Control Rod Cracking GuiExaminationma Accessible surface Guide Tube (Thermal Visual (VT-3) examination Examination at each of the 4 (CRGT) Embrittlement), ASME Section X during the next 10-year Category AssembygNone includng theyeNar screw locations (at NoneP y ISI Program, Primary None ISI. Subsequent Remoabl every 90') of 100%Spacer Castings detection of MRP-227-A Removableexmntin nthToeSuprr (avg ferrite fractured RVI AMP examinations on the Core Support ofsthe (limited content 6.2% -spacers or 10-year ISI interval Structures assibility) 27.7%) missing screws accessibility)

Control Rod Examination Guide Tube Dispositioned Category (CRGT) None Loss of ASME Section Xl B as No No Additional N/A B-N-3, Assembly Material/Wear ISI Program Additional Measures Removable Rod Guide Tubes Measures Core Support Structures Report No. 1200459.401.R1 B-2 CStructural Integrity Associates, IncO Table B-1. Summary of Aging Management Evaluations for ANO-1 Reactor Vessel Internals 13, 10, 181 (Continued)

Componentl Aging Effect Aging Current Commodity Additional Requiring Management Classification Classification Similarity Final Augmented Inspection Section Xl Comments MRP-189, Rev. 1 Description Management Programs MRP-189 in MRP-227-A Classification Requirements Table 5-1 MExams Control Rod Guide Tube Examination (CRGT) Dispositioned Category Sectors None Loss of ASME Section Xl B as No No Additional N/A B-N-3, Material/Wear ISI Program Additional Measures Removable Measures Core Support Structures Examination One-time physical Category Core Support measurement no later B-N-3, than two refueling outages Core Sm o r Shield CSS Top Loss of ASME Section XI Primary per from the beginning of the upport In MRP-227-A and Assembly Flange Material/Wear ISI Program, Primary N/A MRP-227-A LR period. Structures Cylinder Top RVI AMP VT-3 visual ection Xl Programs Flange Perform subsequent visual (VT-3) examination on the examination of 10-year ISI interval.

acessiEve surface -Every interval Volumetric examination (UT) of the bolts within two refueling outages from 1/1/2006 or next 10-year Examination ISI interval, whichever is Category Core Support first. B-N-3, Shield Subsequent examination Removable Assembly CSS to Core ASME Section XI Primary per on the 10-year ISI interval Core Support In MRP-227-A and Upper CSS Barre SCC ISI Program, C Primary N/A unless an evaluation of Structures Upper Core Barrel Bolts RV-27AIecinAMPrgrm Barrel Bolts RVI AMP the baseline results VT-3 visual Section Xl Programs (Original) submitted for NRC staff examination of approval justifies a longer accessible interval between surface -Every examinations.

Visual interval (VT-3) examinations of the bolt locking devices on the 10-year ISI interval.Report No. 1200459.401.R1 B-3 V Structural Integrity Associates, Inc!

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

[3, 10, 18] (Continued)

Component/

la Final Augmented Inspection Current Commodity Additional Aging Effect Aging Classification Classification Final Agednptein Comments MRP-189, Rev. 1 Description in MRP-189 in MRP-227-A imi sification Requirements ection XC Table 5-1 Management Programs Exams Volumetric examination (UT) of the bolts within two refueling outages from 1/1/2006 or next 10-year Examination IS[ interval, whichever is Category Core Support first. B-N-3, Shield Subsequent examination Removable Assembly CSS to Core ASME Section X1 Primary per on the 10-year IS( interval Core Support In MRP-227-A and Upper Core Barrel Bolts SCC ISI Program, C Primary N/A MRP-227-A unless an evaluation of StructuresPrograms Barrel Bolts the baseline results VT-3 visual submitted for NRC staff examination of (Replacement) approval justifies a longer accessible interval between surface -Every examinations.

Visual interval (VT-3) examinations of the bolt locking devices on the 10-year ISI interval.Cracking Examination Core Support (Thermal Category Shield Embrittlement), B-N-3, Assembly Vent Valve including the Visual (VT-3) examination Removable Vent Valve Top Retaining detection of ASME Section XI during the next 10-year Core Support anRings in surface Primary per ISI. Subsequent Structures In MRP-227-A and and Vent Valve Table 4-1 of irregularitie Program, Primary N/A MRP-227-A examinations on the 10- VT-3 visual Section Xl Programs Bottom Retaining BAW-2248A such as year ISI interval.

examination of Rings (tempering damaged, aminio temperature fractured accessible 1100°F) material, or surface -Every missing items interval Core Support Examination Shield Not Category Assembly Thermal ASME Section XI Not specifically B-N-3, Vent Valve Disc None Embrittlement ISt Program B specifically N/A mentioned N/A Removable Note 1 (casting, ferrite mentioned Core Support unknown) Structures Note 1: The staff also determined that, consistent with the aging management review requirements in 10 CFR 54.21(a)(1), the scope of MRP-227 does not include any movable parts or a change in configuration (i.e., active RVI components, such as B&W-design vent valve discs, shafts or hinge pins, or RVI nuclear instrumentation) or components that would be subject to replacement based on qualified life or specified time period (i.e., consumable items, such as fuel assemblies or reactor control assemblies)

Report No. 1200459.401.RI B-4 C S tructural Integrity Associates, Inc!

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

[3, 10, 18] (Continued)

Component/

Augmented Commodity Additional Aging Aging Classification Classification in Final Current Section XI Comments MRP-189, Rev. 1 Description Requiring Management in MRP-189 MRP-227-A Classification Inspection Exams Table 51 Management Programs Requirements Core Support Shield Assembly Vent Valve Disc Examination Category Shaft or Hinge Pin None Thermal ASME Section Xl B Not specifically N/A Not specifically N/A B-N-3, Removable Note 1 (A 276 Type 431 Embrittlement ISI Program mentioned mentioned Core Support Cond. T) Structures Addressed in CALC Vent Valve Not specifically Examination Category 3-R-1013-08 of the Core Support Locking mentioned.

Existing B-N-3, Removable Management of Aging AREVA has program per Core Support Shield Assembly Devices Loss of ASME Section Xl Not specifically recomended N/A AREVA N/A Structures VT-3 visual Effects for the ANs-1 Vent Valve rcmedd NAAEANASrcue T3vsa Lcng Deve (Original) in Material ISI Program mentioned classification as recommendation examination of Reactor Vessel LcigDvcs Table 4-1 of Intras (Original)

BAW-2248A an existing [321. accessible surface -program (321. Every interval See AREVA letter[321.Addressed in CALC C9 3-R-1 013-08 Not specifically Examination Category (Demonstration of the Core Support Vent Valve mentioned.

Existing B-N-3, Removable Mnement of Agn SilAseby Locking ARV a rga e oeSpotManagement of Aging SilAseby Devices Crcig ASME Section XI Not specifically Areomede N/A program N/A Srctures Supot-vsa Effects for the ANO-1 Vent Valve Mdd n Cracking roamnted recommended N/A AREVA N/A Structures VT-3 visual RatrVse VetVle (Modified) in ISI Program mentioned Reactor Vessel Locking Devices Table g o classification as recommendation examination of (Mdfe) Table 4-1 of asenas BAW-2248A an existing [32]. accessible surface -program [32). Every interval See AREVA letter[32].Core Barrel No examination Core barrel cylinder Core Barrel Cylinder ASME Section Xi Expansion per requirements. (including vertical and Assembly (Including Irradiation Exanio Prga,6EpninNAM P-27A Jsiybercufrnilsa

/Core Barrel Vertical and Embrititement ISI Program, BxasonNA MRP-227-A Justify by circumferential seam N/A Corer Cicalernd RVI AMP evaluation or welds) inaccessible for Cylinder Circumferential replacement.

inspection.

Seam Welds)C Structurai Integrity Associates, Inc.Report No. 1200459.401.R1 I B-5 Table B-I. Summary of Aging Management Evaluations for ANO-1 Reactor Vessel Internals

[3, 10, 181 (Continued)

Component/

Current Commodity Additional Aging Effect Aging Classification Classification ilarity Final Augmented Inspection Section Xt Comments MRP-189, Rev. Description Requiring Management in MRP-189 in MRP-227-A Sim Classification Requirements I Table 5-1 Management Programs Exams Volumetric examination (UT) of the bolts during the next Examination Core Barrel Bolts: Cracking 10-year ISI interval from Category Assembly (SCC). 1/1/2006.

B-N-3, Lower Core Similar to Lower Reduction of ASME Subsequent examination on Removable Barrel (LCB) Internals Assembly Fracture Section Xl ISI Primary per the 10-year ISI interval unless Core Support In MRP-227-A to Core Barrel Bolts Toughness and L dMRP-227-A an evaluation of the baseline Structures and Section Xl Lower Grid in Table 4-1 of Loss of Closure Program, results submitted for NRC staff VT-3 visual Programs Assembly-to-BAW-2248A Integrity per RVI AMP approval justifies a longer examination Core Barrel BA-28Inertpr CoregBal Table 4-1 of interval between examinations, of accessible Bolts (Original)

BAW-2248A)

Visual (VT-3) examinations of surface -the bolt locking devices on the Every interval 10-year ISI interval.Examination Bolt or Stud/Nut:

Volumetric C ateor examination (UT). Category Core Barrel Similar to Core Locking Devices: Visual (VT-3) B-N-3R CoreBarrl Smila to oreRemovable Assembly Cracking (SCC), ASME examination UT 26 barrel-to-thermal rackingportSCC),P-A2ME UTS A286 shield olts in Loss of Closure Section Xl ISI SExpansion per Subsequent examinations on C res and Section Xl Thermal Shield- Table 4-1 of BAW- Integrity (LRA Program. B Expansion N/A ExPansion pe tSbsqe n 1-exam Sintrationlsso Cr Strcupprts In ecin MR- I7-shel ols n Inegiy LR roraMRP-227-A the 10-year 1SI interval unless V-T-3 visual Programs to-Core barrel 2248AAMR)

RVI AMP an applicant/licensee provides examination Bolts an evaluation for NRC staff of accessible approval that justifies a longer surface -interval between inspections.

Every interval Core Barrel Dispositioned Dispositioned as Examination as No ipoiind Category AsebyASME aNoNo Additional Ctgr Assembly ISR and Creep, Section Xl ISI Additional No addtoa B-N-3, N/A Thermal Shield None Wear, Fatigue Seto l11BMeasures per N/A Measures per N/A RemovableNA Upper Restraint Program Table pe Table 3-1 of Core Cap Screws Table 3-1 of MP27ACore Support sMRP-227-A Structures Cracking (IE), Examination including the ASME Visual (VT-3) examination Category Core Barrel detection of SI CA suent examination on temovb Assembly None readily Section X1 ISI Primary N/A Primary per during the next 10-year ISI. B-N-3, In MRP-227-A Baffle Plates detectable Program, MRP-227-A Subsequent examinations on Removable cracking in the RVI AMP the 10-year ISI interval.

Core Support baffle plates Structures Report No. 1200459.401.R1 B-6 i Structural Integrity Associates, Inc?

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

[3, 10, 18] (Continued)

Component/

Aging Effect Aging Augmented Commodity Additional Requiring Management Classification Classification Final mspartty Current Section ComExam MRP-189, Rev. 1 Description in MRP-189 in MRP-227-A Similarity Classification XI Exams Table 5-1 Management Programs Requirements Cracking (IE), Core Barrel including ASME Section Xl Expansion per Nexaintso former plates are Assembly None readily ISI Program, C Expansion N/A MRP227A requirements.

Justify inaccessible for In MRP-227-A Former Plates detectable RVI AMP by evaluation or by inspections.

cracking replacement.

Core Barrel IASCC, Listed as Assembly Thermal Stress Core Core Barrel-to-Relaxation,barrel-to-No examination Former Plate Cap Former Plate Cap None ISR and Creep, ISI Program, C Expansion former requ irements Justify Screws In MRP-227-A Screws Wear, Fatigue, bolts in by evaluation or by inaccessible for Irradiation RVI AMP Table 4-4 replacement.

inspection Embrittlement, of MRP-Void Swelling 227-A Listed as locking Core Barrel devices for Core Barrel-to-Assel the Core Former Plate Cap Assembly None IASCC, ASME Section Xl barrel-to-Not specifically Expansion Screws -Locking In MRP-227-A Core Barrel-to-C Irradiation SI Program, Expansion former mentioned Pins are Former Plate Cap Embrittlement RVI AMP bolts in inaccessible for Screws -Locking Pins Table 4-4 inspections of MRP-________________

__________

___________

____________

___________

227A _______22___-A W Structural Integrity Associates, Inc.Report No. 1200459.40 L.R1 B-7 Table B-1. Summary of Aging Management Evaluations for ANO-1 Reactor Vessel Internals

[3, 10, 18] (Continued)

Component/

Aging Effect Aging Augmented Current Commodity Additional ingClassification Classification Inspection Section X Comments MRP-189, Rev. 1 Description Management Programs in MRP-189 in MRP-227-A milarity Classification Requion Exame Table 5 Description Requirements Exams Core Barrel Core Barrel-to-Assembly ASME Section Dispositioned as No Additional Former Plate Core Barrel-to-None Cracking (IE) XI ISI Program No Additional N/A Measures N/A Dowels are N/A Former Plate Measures inaccessible for Dowels inspection Core Barrel Examination Assembly Baseline volumetric Category B-N-3, Baffle-to-Former (UT) no later than two Removable Hex Head Bolts Baffle-to-former Cracking ASME Section refueling outages from Core Support b ale- (IASCC, IE, XI Primary N/A Primary per the beginning of the Structures VT-3 (Note 2)bolts in Table 4- vraCC Program, MRP-227-A license renewal period visual 1 of BAW-2248A (Note 2) RVI AMP with subsequent examination of examination after 10 accessible additional years. surface -Every interval Cracking Locking devices, CoSCrcing including locking Core Barrel Lockin d (IASCC, IE, welds, of internal Assembly including locking Overload), Visual (VT-3) Examination baffle-to-baffle bolts Baffle-to-Former welds, of baifn including the ASME Section examination during the Category B-N-3, are primary.Hex Head Bolts -dedo af e tectioimfryl perCtegoryB1N0-yarerpimary Locking Pins and to-former bolts detection of XP ISI B Primary N/A Primary per next 10-year 151. Removable Locking devices, missing, non- Program, MRP-227-A Subsequent Core Support including locking Baffle-to-Baffle 12 functional, or RVI AMP examinations on the Structures welds, for the Pt Bolts Locking bolts removed 10-year ISI interval, external baffle-to-.

Rings locking devices baffle bolts are or welds expansion.

Note 2: The primary aging degradation mechanism for loss ofjoint tightness for this item are IC and ISR. Fatigue and wear, which can also lead to cracking, are secondary aging degradation mechanisms after significant stress relaxation of loss ofpreload has occurred due to IC/ISR. Bolt stress relaxation cannot readily be inspected by NDE. Only bolt cracking is inspected by UT inspection.

The effect of loss ofjoint tightness on the functionality will be addressed by analysis of the core barrel assembly, which will be performed to address Applicant/Licensee action item 6 in SE [4]Report No. 1200459.401 .Rl B-8 C Structural Integrity Associates, Inc?

Table B-1. Summary of Aging Management Evaluations for ANO-1 Reactor Vessel Internals 13, 10, 18] (Continued)

Component/

Aging Effect Agn Commodity Additional Aging Classification Classification Final Augmented Inspection Current Section MRP-18m, Rev. I Description Requiring Management in MRP-189 in MRP-227-A Similarity Classification Requirements Xl Exams Comments Table 5-1 Management Programs IASCC, Thermal Baseline volumetric Stress examination (UT) no Examination Core Barrel Relaxation, ISR ASME Section Listed as later than two refueling Category B-N-3, Assembly None and Creep, Xl 1IS Primary FB bolts in Primar outages from the Removable Core In MRP-227-A Baffle-to-Former Wear, Fatigue, Program, MRP-227-A beginning of the license Suoft Shoulder Screws Irradiation RVI AMP Table 3-1 renewal period with Structures Embrittlement, subsequent examination Void Swelling after 10 additional years.Listed as Visual (VT-3)Core Barrel ASME Section FB bolt examination during the Examination Assembly A SM loSking next 10-year ISI. Category B-N-3, Baffle-to-Former None IASCC, IE Xl B Primary c Primary Removable Core In MRP-227-A Shoulder Screws -Subsequent Support Locking Dowels RVI AMP MRP-227-A examinations on the 10- Structures year interval Report No. 1200459.401.R1 B-9 CStructural Integrity Associates, inc!

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

[3, 10, 18] (Continued)

Component/

urn C o m m o nenty A g in g E ff e ct A g in g Fi a u m n e n p c i nC u rre n t Commodity Additional Requiring Management Classification Classification Similarity Final Augmented Inspection Section XI Comments MRP-189, Rev. I Description in MRP-189 in MRP-227-A Classification Requirements Exams Table 5-1 Management Programs Bolts: Cracking (IASCC, Fatigue, IE, Overload), ISR, Creep, Wear, Void Locking devices, Swelling, Detection Internal baffle-to-baffle Examination including locking of missing, non- examination Category B-N-3, welds, or internal functional, or requirements, Justify by Removable baffle-to-baffle Core Barrel removed locking bolts are Primary Assembly devices or welds ASME Section evaluation or by Core Support components.

Baffle-to-Baffle 12 None for the internal X ISI C Expansion N/A Expansion per replacement.

Structures VT-3 Locking devices, pt Bolts and locking baffle-to-baffle Program, MRP-227-A External baffle-to-former visual in VIAPbolts:

No examination examination of icluding locking rings bolts and Cracking RVI AMP welds, for the (IASCC, IE) for the requirements.

Justify accessible external baffle-to-external baffle-to by evaluation or by surface -Every baffle bolts are baffle-bolts) replacement interval Expansion components.

Locking Device: Cracking (IASCC, IE)Lower Grid Cracking (IE), Assembly including the Visual (VT-3) Examination Lower Grid Fuel detection of examination.

Assembly Support Similar to separated or Subsequent examination Category B-N-3, Pads, dowels and Removable Pads, Dowel and lower grid missing welds, on the 10-year ISI Core Support cap screws are Cap Screw assembly, and missing support ASME Section interval unless an Core VT-3 included because fuel assembly pads, dowels, cap Proram B Expansion N/A Expansion applicantlicensee of Irradiation support pads screws and locking Program, provides an evaluation visual Embrittiement of in Table 4-1 of welds, or for NRC staff approval accessible the pad-to-rib BAW-2248A misalignment of the that justifies a longer surface -Every section welds support pads per interval between interval Table 4-4 of inspections.

MRP-227-A Report No. 1200459.401.R1 B-10 t Structural Integrity Associates, Inc?

Table B-1. Summnary of Aging Management Evaluations for ANO-1 Reactor Vessel Internals 13, 10, 181 (Continued)

Component/

Aging Effect Aging Classificatio Commodity Additional Ruing MAng Classification SimFinal Augmented Inspection Current Section Xl Comments MRP-1189, Rev. 1 Description Requiring Management n in MRP- in MRP-227-A ilarity Classification Requirements Exams Table 5-1 Management Programs 189 Lower Grid Examination Assembly Lower Grid Rib Category B-N-3, Rib-to-Shell Dispositioned Removable Core Forging Cap to Shell Forging ISR and Creep, ASME Section as No N/A No Additional N/A Support Structures.

Screws Screws in Table Wear, Fatigue X SI Program Additional Measures Visual T-3N/A 4-1 of Measures examination of BAW-2248A accessible surface -Every Interval Examination Lower Grid Category B-N-3, Assembly Dispositioned Removable Core Support Post None ISR and Creep, ASME Section B as No N/A No Additional N/A Support Structures.

N/A Pipes eap Wear, Fatigue XI iSI Program Additional Measures Visual VT-3 Screws Measures examination of accessible surface -Every Interval Bolt: Volumetric Bolts: Cracking examination (UT).Lower Grid (SCC). Locking Devices: Visual Examination (VT-3) examination.

C ate on Assembly Lower Internals Locking Devices: Subsequent Category B-N-3, LTS: Assembly to Loss of material, ASME Section Removable Core Replacement dsebyt xpninpr examinations on the 10- Rempovabl Srcoures Replacemer Thermal Shield damaged Xl ISl Expansion N/A year ISI interval unless VT-3 Sual N/A A286Bolts in Table 4- distorted or Program, MRP-227-A an applicant/lienseeS Assembly-to-Bissing locking RVl AMP examination of Thermal Shield 1 of BAW-2248A missing o provides an evaluation accessible surface -devices (Wear or for NRC staff approval Evessiblerfal Bolts) Fatigue damage that justifies a longer Every interval by failed bolts) interval between inspections Report No. 1200459.401.R1 B-1 1 4 Structural Integrity Associates, Inc!

Table B-1. Summary of Aging Management Evaluations for ANO-1 Reactor Vessel Internals 13, 10, 181 (Continued)

Component/

Commodity Additional Aging Effect Aging Classification Classification Final Augmented Inspection Current Section MRP°189, Rev. 1 Description Requiring Management in MRP-189 in MRP-227-A Similarity Classification Requirements XI Exams Comments Table 5-1 Management Programs Volumetric (UT) of the Shell Forging to bolts during the next 10-Flow Distributor year ISI interval from Bolts in Table 4-1 1/1/2006.

Subsequent Category B-N-3, Flow Distributor of BAW-2248A.

examination on the 10- Removb Core Assembly Flow Distributor ASME Section year IS[ interval unless an Remort ED: A286 Flow S F Cracking (SCC) XS ISI Program, Primary N/A Primary per evaluation of the baseline Structures VT-3 N/A Distributor-to-Lower Bolts (addressed RVI AMP MRP-227-A results, submitted for NRC visual examination Grid Shell Forging in Topical Report staff approval, justifies a of accessible Bolts Condition 3 in longer interval between surface -Every Section 4.1.3 of examinations.

Visual (VT- interval NRC SE on 3) examination of bolt MRP-227) locking devices on the 10-year ISI interval Cracking (TE/IE), including the Initial visual (VT-3)embe detection of examination no later than Examination Assembly fractured or missing ASME Section two refueling outages from Category B-N-3, IMl Guide Tube spider arms or, Primary per thbeingofheRmvleCrNA Spider Castings None Cracking (IE), Xl ISI Program, B Primary N/A MRP-227-A the beginning of the Removable Core N/A (Ferrite content including separation RVI AMP license renewal period. Support unknown) of spider arms from Subsequent examinations Structures the lower grid rib on the 10-year ISI interval section at the weld Report No. 1200459.401.R1 B-12 VStructural Integrity Associates, Inc?

APPENDIX C REACTOR VESSEL INTERNALS MATERIALS OF CONSTRUCTION FOR ANO-1 [181 Report No. 1200459.401.R1 C- 1 Structural Integrity Associates, Inc!

Table C-1. RVI Materials of Construction for ANO-1 [181 Item Material Specification Number Type or Grade/Class PLENUM ASSEMBLY Plenum Cover Assembly Plenum Cover Weldment Ribs Stainless Steel A240-63 Tp. 304 Rib Pads Stainless Steel A275-65 Tp. 304 Plenum Cover Bottom Flange Stainless Steel A240-63 Tp. 304 Plenum Cover Support Flange Stainless Steel A240-63 Tp. 304 Plenum Cover Plate Stainless Steel A240-63 Tp. 304 Plenum Cover Support Ring Stainless Steel A240-63 Tp. 304 Lifting Lugs Stainless Steel A240-63 Tp. 304 Base Blocks Stainless Steel A240-63 Tp. 304 Lifting Lug-to-Base Block Bolts Stainless Steel A193-65 Gr. B-8 Locking Cups Stainless Steel ASTM A 167 Tp. 304 Plenum Cylinder Cylinder Stainless Steel A240-63 Tp. 304 Top Flange, Bottom Flange Stainless Steel A473-63 Tp. 304 Reinforcing Plates Stainless Steel A240-63 Tp. 304 Round Bars Stainless Steel A276-65 Tp. 304 Top Flange-to-Cover Bolts Stainless Steel A193-65 Gr. B-8 Locking Cups Stainless Steel ASTMA167 Tp. 304 Bottom Flange-to-Upper Grid Screws Stainless Steel AI193-65 Gr. B-8 Locking Cups Stainless Steel ASTM A167 Tp. 304 Upper Grid Assembly Upper Grid Rib Section Stainless Steel A240-63 Tp. 304 Upper Grid Ring Forging Stainless Steel A473-63 Tp. 304 Rib-to-Ring Screw Stainless Steel A 193-65 Gr. B-8 Lockpins Stainless Steel AISI Tp. 304 Flow Assembly Support Pads Stainless Steel A276-65 Tp. 304 Dowels Ni-Cr Alloy (Inconel X-750) AMS 5667F Cap Screw Stainless Steel A193-65 Gr. B-8 Report No. 1200459.401.R1 C-2 V Structural Integrity Associates, Inc!

Table C-1. RVI Materials of Construction for ANO-1 [18] (Continued)

Item Material Specification Number Type or Grade/Class Control Rod Guide Tube Assembly CRGT Pipe Stainless Steel A312-64 Tp. 304 CRGT Flange Stainless Steel A240-63 Tp. 304 Flange-to-Upper Grid Screws Stainless Steel A 193-65 Gr. B-8 Dowels Stainless Steel A276-65 Tp. 304 CRGT Spacer Castings Stainless Steel A351-65 Gr. CF-3M Spacer Casting Screws Stainless Steel A 193-65 Gr. B-8 Spacer Casting Washers Stainless Steel AN Tp. 304 CRGT Rod Guide Tubes Stainless Steel SA 240-63 Tp. 304L CRGT Rod Guide Sectors Stainless Steel SA 240-63 Tp. 304L CORE SUPPORT SHIELD ASSEMBLY (CSS)Core Support Shield Cylinder Stainless Steel A240-63 Tp. 304 Top Flange Stainless Steel A473-63 Tp. 304 Bottom Flange Stainless Steel A473-63 Tp. 304 Core Support Shield-to-Core Barrel Stainless Steel (114- A453 (A286) Gr. 660, Cond A Replacement Bolts replacement)

Locking Cups & Tie Plates (replacement Stainless Steel A240-74 or A479-75 Tp. 304L bolts)Outlet Nozzles Stainless Steel A473-63 Tp. 304 Vent Valve Nozzles Stainless Steel A473-63 Tp. 304 Vent Valve Guide Blocks Stainless Steel A276-65 Tp. 304 Vent Valve Body Stainless Steel A351 Gr. CF-8 Retaining Rings Stainless Steel AMS 5658 Jack Screws Stainless Steel AMS 5737C A240 Tp. 304 Misc. locking device parts Stainless Steel SA479 Tp. 304 SA- 182 Tp. 304 SA-193 Gr. B8 or B8M Report No. 1200459.401.R1 C-3* Structural Integrity Associates, Inc!

Table C-1. RVI Materials of Construction for ANO-1 118] (Continued)

Item Material Specification Number Type or Grade/Class MIL-N-23228A or Cond. A SB-168 Misc. locking device parts (Continued)

Ni-Cr-Fe Alloy (Inconel 600) MIL-N-23229A Cond. A AM.2TI or SB- 166 Misc. 1k devic Ctd) Ni-Cr Alloy AMS-56622 ocking device pas oninue (Inconel 718) or SA-637 Gr. 718 Round Bars Stainless Steel A276-65 Tp. 304 Flow Deflectors Stainless Steel A240-63 Tp. 304 Lifting Lugs Stainless Steel A276-65 Tp. 304 RVLMS Brazement guide assembly-guide Stainless Steel A240 Tp. 304L assembly top block Brazement guide assembly manometer Stainless Steel A213 Tp. 304L housing Brazement guide assembly manometer Stainless Steel A479 Tp. 304L housing end piece Brazement guide assembly brazement Stainless Steel A240 Tp. 304L hub Brazement guide assembly tie rod Stainless Steel A479 Tp. 304L Brazement guide assembly j-bolt and nut Stainless Steel A479 Tp. 304L Brazement guide assembly arm Stainless Steel A240 Tp. 304L Brazement guide assembly tube ID Stainless Steel A479 Tp. 304L support Closure extension instrument guide tube Stainless Steel A213 Tp. 304L or Tp. 304 Closure extension instrument guide end Stainless Steel SA479 Tp. 304L or Tp. 304 plug Closure extension instrument guide Stainless Steel A240 Tp. 304L or Tp. 304 spacer Closure extension instrument guide tube Stainless Steel A479 Tp. 304L ID support Closure extension instrument guide Stainless Steel A213 Tp. 304L or Tp. 304 housing Report No. 1200459.401.Rl C-4 t Structural Integrity Associates, Inc!

Table C-1. RVI Materials of Construction for ANO-1 [181 (Continued)

Item Material I Specification Number Type or Grade/Class CORE BARREL ASSEMBLY Core Barrel Cylinders Stainless Steel A240-63 Tp. 304 Top Flange Stainless Steel A473-63 Tp. 304 Bottom Flange Stainless Steel A473-63 Tp. 304 Lower Internals Assembly-to-Core Barrel Stainless Steel (all original)

A453 (A286) Gr. 660, Cond. A Bolts Locking Clips Stainless Steel A240-63 Tp. 304L Core Barrel-to-Thermal Shield Bolts Stainless Steel (all original)

A453-63 (A286) Gr. 660, Cond. A Locking Clips Stainless Steel A240-63 Tp. 304L Baffle Plates Stainless Steel A240-63 Tp. 304 Formers Stainless Steel A240-63 Tp. 304 Barrel-to-Former Bolts Stainless Steel A 193-65 Gr. B8 Locking Pins Stainless Steel AISI Tp. 304 Dowels Ni-Cr Alloy (Inconel X-750) AMS 5667F Baffle-to-Former Bolts Stainless Steel A 193-65 Gr. B8 Locking Pins Stainless Steel AISI Tp. 304 Baffle-to-Former Shoulder Screws Stainless Steel A 193-65 Gr. B8 Locking Dowel Stainless Steel A276-65 Tp. 304 Baffle-to-Baffle Bolts Stainless Steel A 193-65 Gr. B8 Locking Rings Stainless Steel AISI Tp. 304 Lower Internals Assembly -Lower Grid Rib Section Lower Grid Rib Section Stainless Steel A240-63 I Tp. 304 Fuel Assembly Support Pads Stainless Steel A276-65I Tp. 304 Report No. 1200459.401.R1 C-5 CStructural Integrity Associates, Inc!° Table C-1. RVI Materials of Construction for ANO-1 1181 (Continued)

Item Material I Specification Number Type or Grade/Class Lower Internals Assembly -Lower Grid Rib Section (Continued)

Dowels Ni-Cr Alloy (Inconel X-750) AMS 5667F Cap Screw Stainless Steel A193-65 Gr. B-8 Rib-to-Shell Forging Screw Stainless Steel A193-65 Gr. B-8 Locking Pin Stainless Steel AISI Tp. 304 Lower Grid Flow Distributor Plate Stainless Steel A240-63 Tp. 304 Orifice Plugs Stainless Steel A276-65 Tp. 304 Lower Grid Forging Stainless Steel A473-63 Tp. 304 Lower Grid Shell Forging Stainless Steel A473-63 Tp. 304 Lower Internal Assembly-to-thermal Stainless Steel A453 (A286) Cr. 660, Cond A shield bolts (Replacement)

Locking Cups & Tie Plates (replacement Stainless Steel A240-74 or A479-75 bolts) StainlessSteelA240-4_orA479-75__p__304 Guide Blocks Stainless Steel A276-65 Tp. 304 Guide Block Bolts Stainless Steel A193-65 Gr. B-8 Guide Block Washers Stainless Steel A193-65 Gr. B-8 Dowel Ni-Cr Alloy (Inconel X-750) AMS 5667F Shock Pads Stainless Steel A276-65 Tp. 304 Shock Pad Bolts Stainless Steel A193-65 Gr. B-8 Support Post Pipes Stainless Steel A312-64 Tp. 304 Bolting Plugs Stainless Steel A276-65 Tp. 304 Support Post Cap Screws Stainless Steel A193-65 Gr. B-8 Locking Pin Stainless Steel AISI Tp. 304 Report No. 1200459.401.R1 C-6 k Structural Integrity Associates, Inc.P Table C-1. RVI Materials of Construction for ANO-1 1181 (Continued)

Item Material Specification Number Type or Grade/Class Flow Distributor Assembly Flow Distributor Head Stainless Steel A240-63 Tp. 304 Flow Distributor Flange Stainless Steel A473-63 Tp. 304 Shell Forging-to-Flow Distributor Bolts Stainless Steel A453 (A286) Gr. 660, Cond A Locking Clip Stainless Steel A240-63 Tp. 304 tncore Guide Support Plate Stainless Steel A240-63 Tp. 304 Clamping Ring Stainless Steel A240-63 Tp. 304 Dowel Stainless Steel A276-65 Tp. 304 Incore Guide Tube Assembly Incore Guide Tubes Stainless Steel A276-65 Tp. 304 Gussets Stainless Steel A276-65 Tp. 304 Guide Tube Nuts Stainless Steel A276-65 Tp. 304 Guide Tube Washers Stainless Steel A276-65 Tp. 304 Locking Clips Stainless Steel A240-63 Tp. 304 Spiders Stainless Steel A351-65 Gr. CF-8 Remaining Portions of the SSHT, Thermal Shield and Thermal Shield Upper Restraint Upper Tube (part 224) Stainless Steel A269-65 Tp. 304 Bracket (part 227) Stainless Steel A276-65 Tp. 304 Hex Head Cap Screw (part 114) Stainless Steel A193-66 Gr. B-8 C Brackets (part 385) Stainless Steel A276-65 Tp. 304 Brackets (B221 and B224) Stainless Steel A240-63 Tp. 304 Thermal Shield Upper Restraint Stainless Steel A240-63 Tp. 304 Upper Restraint Shim Stainless Steel A240-63 Tp. 304 Upper Restraint Dowel Stainless Steel A453-65 Grade 660 Report No. 1200459.401.R1 C-7 V Structural Integrity Associates, Inc!

Table C-1. RVI Materials of Construction for ANO-1 118] (Continued)

Item Material Specification Number Type or Grade/Class Remaining Portions of the SSHT, Thermal Shield and Thermal Shield Upper Restraint (Continued Upper Restraint Dowel Stainless Steel A276-65 Tp. 304 Upper Restraint Locking Clip Stainless Steel A240-63 Tp. 304 Thermal Shield Cylinder Stainless Steel A240-63 Tp. 304 Report No. 1200459.401.R1 C-8 C Structural Integrity Associates, lnO APPENDIX D MANAGING LOSS OF DUCTILITY Report No. 1200459.401.R1 D-1 C Structural Integrity Associates, In0 D.1 Background In response to the NRC review of the ANO-1 LRA, and the NRC staff's final SER of the topical report BAW-2248 [9], there were 12 action items regarding demonstration that the effects of aging will be adequately managed so that the intended function would be maintained consistent with the CLB for the period of extended operation for the reactor vessel internals

[ 19]. Of the 12 action items, only Item 12 was unresolved and required further action: "The applicant will develop the necessary data, and perform the necessary analyses, to demonstrate that the reactor vessel internals will have sufficient ductility to absorb local strain in the regions of the high stress intensity, and will meet the deformation limits at the end of the period of extended operation.

" This action item has been examined in more detail. In 2010, Appendix E of the 1970 RV internals topical report was updated by AREVA for 60 years on a generic basis for the B&W units and submitted, for information, to the NRC [35]. That updated report identified the locations of the maximum stress intensity where a loss of ductility due to neutron irradiation would be detrimental to be the core barrel flanges. However, in a subsequent licensing report for ANO- 1, a more detailed examination showed that the location where the highest stress intensity occurs is at the bottom core support shield flange [36]. Therefore, the bottom core support shield flange was reevaluated as the region of maximum stress intensity for ANO- 1 to meet license renewal applicant action item # 12 from BAW-2248A.

D.2 Analysis and Conclusions for ANO-1 Loss of Ductility in RV Internals In order to support an analysis, fluence values were specifically generated for ANO- 1 at 54 EFPY. The methodology used to determine the neutron fluence was based on AREVA's NRC approved fluence analysis methodology, described in topical report BAW-2241P-A (References 37, 38, and 39). The fluence methodology is consistent with the guidance of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Report No. 1200459.401 .R1 D-2 C Structural Integrity Associates, Inc Fluence." The projected 54 EFPY fluence was recalculated for the core support shield lower flange, which has been shown to be the region of maximum stress intensity for the RV internals.

Details of the actual fluence value and how it compares to other locations in the internals is contained in Reference

36. The second required input is the material used for the manufacturing of the core support shield top and bottom flanges. As detailed in Reference 9, the core support shield flanges are fabricated from American Society of Testing and Materials (ASTM) A 473-63 Type 304 austenitic stainless steel. Recent irradiated Type 304 test data was compared to previous data to validate that the 20 percent uniform elongation of irradiated material credited for 40 years in Appendix E of the 1970 RV internals topical report is conservative.

These inputs were used to show that the acceptance ductility and deformation limits for a 40-year lifetime remain valid for a 60-year lifetime.

Results and discussion of this analysis are contained in Reference 36.Based on the materials data included in a 1970 RV internals topical report and newer materials data and the projected fluence value for 54 EFPY for ANO-1, the conclusions from Appendix E to the 1970 RV internals topical report concerning the acceptable ductility and deformation limits for a 40-year lifetime remain valid for a 60-year lifetime for the ANO-1 RV internals.

The results of these evaluations have been submitted to the NRC under letter 1 CAN051401.

The submittal of this letter fulfills the Item 12 action for the "TLAA Regarding Reactor Vessel Internals Loss of Ductility for Arkansas Nuclear One, Unit 1 at 60 years." Report No. 1200459.401.Rl D-3 r Structural Integrity Associates, Inc Attachment 2 to 1 CAN051403 List of Commitments Attachment 2 1 CAN051403 Page 1 of 6 List of Commitments This table identifies actions discussed in this letter for which Entergy Operations, Inc. (Entergy)commits to perform for Arkansas Nuclear One, Unit 1 (ANO-1). Any other actions discussed in this submittal are described for the NRC's information and are not commitments.

TYPE SCHEDULED COMMITMENT (Check one) COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required)Inspection standards will be in accordance with the requirements of MRP-228. These inspection standards will be used in X augmented inspections at ANO-1 as applicable where required by MRP-227-A.

Examination results that do not meet the examination acceptance criteria defined in Section 5 of MRP-227-A guidelines shall be X recorded and entered in the plant corrective action program and dispositioned.

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 X within 120 days of completion of an outage during which PWR internals within the scope of MRP-227-A are examined.If an engineering evaluation is used to disposition an examination result that does not meet the examination acceptance criteria in Section 5 of MRP-227-A, this engineering X evaluation shall be conducted in accordance with an NRC-approved evaluation methodology.

The ANO-1 Reactor Vessel Internals (RVI)Aging Management Program (AMP) will be updated accordingly as operating experiences X and new inspection requirements and technologies evolve associated with managing reactor vessel internal aging concerns.

Attachment 2 1 CAN051403 Page 2 of 6 TYPE SCHEDULED COMMITMENT (Check one) COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required)Augmented inspections taken from the MRP-227-A recommendations will be applied through use of the MRP-228 Inspection Standard.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-1 7096 are published X and as the industry continues to develop and refine the information.

Changes applicable to the ANO-1 RVI will be included as part of updates to the AMP.Indications that require repair and replacement will be addressed through the ANO-1 corrective action program. Repair and replacement activities will be performed in accordance with methodologies provided in X Section 6 of MRP-227-A and ASME Code Section XI. The corrective actions for existing Section XI (B-N-3) examinations will include the identification of a repair and verification of acceptability of replacements.

Inspection results having potential industry significance shall be expeditiously reported to the RCS Materials Degradation Program X Manager for consideration of reporting under the NEI 03-08, Material Initiative Protocol.

Attachment 2 1 CAN051403 Page 3 of 6 TYPE SCHEDULED COMMITMENT (Check one) COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required)With regard to the three ANO-1 RVI components that were not listed in BAW-2248A, but were included in the ANO-1 Section Xl program, these components will undergo a future screening and categorization, and based on the screening results, will be removed from future inspections if the components screen out, or added to the primary or expansion categories if the components screen in. Until that time, in order to ensure that these components are adequately managed during the period of extended operation, the components will be inspected during the 10-year intervals based on the respective aging effects. For the reactor vessel level monitor system (RVLMS) X probe supports, cracking and stress relaxation were identified as aging effects for the RVLMS brazement guide j-bolt and nut. For the remaining portions of the surveillance specimen holder tubes (SSHTs), the SSHTs need to remain secured to prevent loose parts in the RCS. Cracking and stress relaxation were identified as aging effects of the SSHT bolting. For the thermal shield and thermal shield upper restraint, cracking and reduction of fracture toughness were identified as aging effects for the thermal shield and thermal shield upper restraint assemblies.

Based on a review of the aging effects, the orphan components will be visually inspected (VT-3)during the 10-year ISI inspections.

Attachment 2 1 CAN051403 Page 4 of 6 TYPE SCHEDULED COMMITMENT (Check one) COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required)Physical measurements are required to determine the differential height of top of the plenum rib pads to the reactor vessel seating surface with all three items inside the reactor vessel, but with the fuel assemblies removed.The acceptance criteria for this one time X physical measurement shall be an average measured differential height from the top of the plenum rib pads to the vessel seating surface of 0.004 inches relative to the as-built condition in accordance with Table 5-1 of this document.ANO-1 will justify the acceptability of inaccessible and non-inspectable components (core barrel cylinder including vertical and circumferential seam welds, former plates, external baffle-to-baffle bolts and their locking devices, core barrel-to-former bolts and their locking devices, and internal baffle-to-baffle bolts) for continued operation through the period of extended operation by performing an evaluation, proposing a schedule for replacement, or justification for some other alternative process for these components.

The evaluation, schedule for replacement, or X justification for some other alternative process will be submitted to the NRC by the end of one year from the initial inspection of the linked Primary component items if the inspection results indicate aging, which is the implementation date for this condition.

Any"other alternative process" shall include justification of operation in the degraded condition on a generic or plant specific basis.Any "aging" detected during the inspection is defined to mean when the expansion criteria for the linked Primary component are met.

Attachment 2 1 CAN051403 Page 5 of 6 TYPE SCHEDULED COMMITMENT (Check one) COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required)An analytical approach to assess the effect of reduction of fracture toughness on the applicable reactor vessel internals will be performed.

The analysis will include an assessment of the potential for synergistic thermal aging and neutron irradiation embrittlement for each affected cast austenitic stainless steel (CASS) component by comparing its nominal 60-year neutron exposure to significant thresholds established in regulatory guidance or, conversely, to industry-established thresholds justified by X experimental data and corroborated by industry operating experience.

For CASS components with the 60-year neutron exposure below the justified threshold, the reduction in fracture toughness used in the analytical assessments will be based on thermal aging embrittlement data. Otherwise, an additional reduction in fracture toughness to account for the synergistic effects of irradiation and thermal aging will be documented.

Should data or the analytical evaluations indicate that the inspections can be modified or eliminated, Entergy will provide plant-specific justification to demonstrate the basis for the modification or elimination, as X applicable.

The generic or plant-specific assessments for the CASS components will be completed 12 months prior to the second refueling outage after entering the period of extended operation.

For components fabricated from CASS, martensitic, and precipitation hardenable stainless steel materials, the acceptance X criteria (critical crack size) will be determined by analysis using an NRC approved methodology.

Attachment 2 1 CAN051403 Page 6 of 6 TYPE SCHEDULED COMMITMENT (Check one) COMPLETION ONE-TIME CONTINUING DATE ACTION COMPLIANCE (If Required)The aging management program plan for ANO-1 will not deviate from the recommendations of MRP-227-A.

Inspection of Primary and Expansion category X components are provided in Tables 5-1 and 5-2 of this document will be performed in accordance with the requirements of MRP-227-A.