ML16202A167

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Report 1500227.401, PWR Internals Aging Management Program Plan.
ML16202A167
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ATTACHMENT TO 2CAN071603 PWR INTERNALS AGING MANAGEMENT PROGRAM PLAN FOR ARKANSAS NUCLEAR ONE, UNIT 2

Report No. 1500227.401 Revision 0 Project No. 1500227 July 2016 PWR Internals Aging Management Program Plan for Arkansas Nuclear One, Unit 2 Prepared for:

Entergy Nuclear Operations Contract Order No: 10463207 Prepared by:

Structural Integrity Associates, Inc.

San Jose, California Prepared by: Date: 7/5/2016 Stephen M. Parker, P .E.

Reviewed by: Date: 7/5/2016 Timothy J. Griesbach Approved by: Date: 7/5/2016 Christopher S. Lohse, P.E.

e Structural Integrity Associates, Inc.

REVISION CONTROL SHEET Document Number: 1500227.401

~~~~~~~~~~~~~~~~~~~~~~~~~-

Title:

PWR Internals Aging Management Program Plan for Arkansas Nuclear One, Unit 2 Client: Entergy Nuclear Operations SI Project Number: 1500227 Quality Program: ~ Nuclear 0 Commercial Section Pages Revision Date Comments 1.0 1-1-1-16 0 7/5/16 Original Issue 2.0 2-1 24 3.0 3-1-3-17 4.0 4 4-5 5.0 5 5-29 6.0 6 6-4 e Structural Integrity Associates, Inc.

Table of Contents Section Page

1.0 INTRODUCTION

....................................................................................................... 1-1 1.1 Objective ....................................................................................................................... 1-1 1.2 AN0-2 Reactor Vessel Internals Inspection Program Commitment.. .......................... 1-2 1.3 AN0-2 Reactor Vessel Internals Aging Management Program Background .............. 1-3 1.4 AN0-2 Reactor Vessel Internals Aging Management Program Elements ................... 1-7 1.5 Responsibilities ............................................................................................................. 1-9 1.6 Program Implementation ............... :............................................................................ 1-10

1. 6.1 ASME Section XI Inservice Inspection Program ................................................... 1-10
1. 6.2 Primary Chemistry Monitoring Program .............................................................. 1-11 1.7 Aging Management Review and Program Enhancements .......................................... 1-11
1. 7.1 Reactor Internals Aging Management Review Process ......................................... 1-11 1.8 Industry Programs ....................................................................................................... 1-12
1. 8.1 CE NPSD-1216, Aging Management ofReactor Internals ................................... 1-12 1.8.2 MRP-227-A, Reactor Internals Inspection and Evaluation Guidelines ................. 1-12 1.8.3 NE! 03-08 Guidance Within MRP-227-A .............................................................. 1-12 1.8.4 MRP-227-A AMP Development Guidance ............................................................ 1-14 1.8.5 Ongoing Industry Programs .................................................................................. 1-16 1.9 Summary ..................................................................................................................... 1-16 2.0 AGING MANAGEMENT APPROACH .................................................................. 2-1 2.1 Mechanisms of Age-Related Degradation in PWR Internals ....................................... 2-1 2.1.l Stress Corrosion Cracking ....................................................................................... 2-1 2.1.2 Irradiation-Assisted Stress Corrosion Cracking ..................................................... 2-1 2.1.3 Wear ......................................................................................................................... 2-1 2.1.4 Fatigue ..................................................................................................................... 2-2 2.1.5 Thermal Aging Embrittlement.................................................................................. 2-2 2.1.6 Irradiation Embrittlement ........................................................................................ 2-3 Report No. 1500227.401.RO 111 e Structural Integrity Associates, Inc.

2.1. 7 Void Swelling and Irradiation Growth .................................................................... 2-3 2.1.8 Thermal and Irradiation-Enhanced Stress Relaxation or Creep ............................. 2-3 2.2 Aging Management Strategy ........................................................................................ 2-4 2.3 AN0-2 Reactor Vessel Internals Aging Management Program Attributes .................. 2-7 2.3.l NUREG-1801/AMP Program Element 1: Scope ofProgram .................................. 2-7 2.3.2 NUREG-1801/AMP Program Element 2: Preventive Actions............................... 2-10 2.3.3 NUREG-1801/AMP Program Element 3: Parameters Monitored/Inspected: ...... 2-11 2.3.4 NUREG-1801/AMP Program Element 4: Detection ofAging Effects: ................. 2-13 2.3.5 NUREG-1801/AMP Program Element 5: Monitoring and Trending: .................. 2-16 2.3.6 NUREG-1801/AMP Program Element 6: Acceptance Criteria ............................ 2-17 2.3.7 NUREG-1801/AMP Program Element 7: Corrective Actions: ............................. 2-19 2.3.8 NUREG-1801/AMP Program Element 8: Confirmation Process ......................... 2-21 2.3.9 NUREG-1801/AMP Program Element 9: Administrative Controls: ..................... 2-21 2.3.10 NUREG-1801/AMP Program Element JO: Operating Experience ....................... 2-22 3.0 AN0-2 REACTOR VESSEL INTERNALS DESIGN AND OPERATING EXPERIENCE ............................................................................................................ 3-1 3.1 Core Support Structure ................................................................................................. 3-1 3.2 Core Shroud Assembly ................................................................................................. 3-3 3.3 Flow Skirt...................................................................................................................... 3-4 3.4 Upper Guide Structure Assembly ................................................................................. 3-4 3.5 In-Core Instrumentation Support System ..................................................................... 3-5 3.6 AN0-2 Design.Distinctions ........................................................................................ 3-17 3.7 AN0-2 Unit Operating Experience ............................................................................ 3-17 4.0 EXAMINATION AND ACCEPTANCE AND EXPANSION CRITERIA ........... 4-1 4.1 Examination Acceptance Criteria ................................................................................. 4-1 4.1.1 Visual (VT-3) Examination ...................................................................................... 4-1 4.1.2 Visual (VT-1) Examination ...................................................................................... 4-2 4.1.3 Enhanced Visual (EVT-1) Examination ................................................................... 4-2 4.1.4 Suiface Examination ................................................................................................ 4-3 Report No. 1500227.401.RO

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IV

4.1.5 Volumetric Examination ....................*...................................................................... 4-3 I

4.1.6 Physical Measurements' Examination ......................................................................

4-3 4.2 Expansion Criteria ........................................................................................................ 4-4 4.3 Evaluation, Repair, and Replacement Strategy............................................................. 4-4 4.3.1 Reporting............ .'..................................................................................................... 4-5 4.4 Implementation Schedule .............................................................................................. 4-5 4.5 . Commitment Tracking .................................................................................................. 4-5 5.0 RESPONSES TO NRC SAFETY EVALUATION APPLICANT/LICENSEE ACTION ITEMS ......................................................................................................... 5-1 5.1 SE Section 4.2.1, Applicant/Licensee Action Item 1 (Applicability ofFMECA 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): .................................................. 5-3 5.3 SE Section 4.2.3, Applicant/Licensee Action Item 3 (Evaluation of the Adequacy of Plant-Specific Existing Programs): .......................................................................... 5-3 5.4 SE Section 4.2.4, Applicant/Licensee Action Item 4 (B&W Core Support Structure Upper Flange Stress Relief): ......................................................................................... 5-4 5.5 SE Section 4.2.5, Applicant/Licensee Action Item 5 (Application of Physical Measurements as part ofl&E Guidelines for B&W, CE, and Westinghouse RVI Components): ................................................................................................................ 5-5 5.6 SE Section 4.2.6, Applicant/Licensee Action Item 6 (Evaluation oflnaccessible B&W Components): ..................................................................................................... 5-5 5.7 SE Section 4.2.7, Applicant/Licensee Action Item 7 (Plant-Specific Evaluation of CASS Materials): .......................................................................................................... 5-6 5.8 SE Section 4.2.8, Applicant/Licensee Action Item 8 (Submittal of Information for Staff Review and Approval): ........................................................................................ 5-7

6.0 REFERENCES

............................................................................................................ 6-1 Report No. 1500227.401.RO v lJ Structural Integrity Associates, Inc.

List of Tables Table 1-1. Key Elements of the Reactor Vessel Internals Aging Management Program ........... 1-8 Table 5-1. CE Plants Primary Category Components from Table 4-2 ofMRP-227-A [4] ...... 5-10 Table 5-2. CE Plants Expansion Category Components from Table 4-5 ofMRP-227-A [4] .. 5-14 Table 5-3. CE Plants Existing Program Components Credited in Table 4-8 of MRP-227-A [4] .................................................................................................................. 5-17 Table 5-4. CE Plants Examination Acceptance and Expansion Criteria from Table 5-2 of MRP-227-A (4) .................................................................................................................. 5-18 Table 5-5. AN0-2 Response to the NRC Final Safety Evaluation ofMRP-227-A [5] ............ 5-23 Table 5-6. AN0-2 Program Enhancement and Implementation Schedule ............................... 5-26 Table 5-7. Suminary of Actions Related to Aging Management ofRVI for AN0-2 ............... 5-28 Report No. 1500227.401.RO Vl e Structural Integrity Associates, Inc.

List of Figures Figure Figure 3-1. Illustration of the AN0-2 Vessel and Internals [18, Figure 4.1-1] ........................... 3-7 Figure 3-2. AN0-2 Core Support Barrel Assembly [12, Figure 4.1-3] ....................................... 3-8 Figure 3-3. AN0-2 Lower Support Structure [12, Figure 4.1-5). ................................................ 3-9 Figure 3-4. AN0-2 Snubber Assembly [18, 4.2-10] ................................................................. 3-10 Figure 3-5. AN0-2 Core Shroud Assembly [18, Figure 4.2-11] ............................................... 3-11 Figure 3-6. Potential Crack Locations for CE Welded Core Shroud Assembled in Stacked Sections [4, Figure 4-12) .................................................................................................... 3-12 Figure 3-7. Locations of Potential Separation Between Core Shroud Sections Caused by Swelling Induced Warping of Thick Flange Plates in CE Welded Core Shroud Assembled in Stacked Sections [4, Figure 4-14] ............................................................... 3-13 Figure 3-8. AN0-2 Upper Guide Structure Assembly [12, Figure 4.1-2] ................................. 3-14 Figure 3-9. Control Element Assembly [31, Figure 42] ............................................................ 3-15 Figure 3-10. In-Core Support Assembly [12, Figure 4.1-8a] .................................................... 3-16 Report No. 1500227.401.RO Vll i)Structural Integrity Associates, Inc.

LIST OF ACRONYMS AMP Aging Management Program AMR Aging Management Review AN0-2 Arkansas Nuclear One, Unit 2 ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox CASS Cast austenitic stainless steel CE Combustion Engineering CEA Control Element Assembly CEOG Combustion Engineering Owners Group CFR Code of Federal Regulations CLB Current licensing basis CSB Core Support Barrel EFPY Effective full power years EPRI Electric Power Research Institute ET Eddy Current Testing EVT Enhanced visual testing (visual NDE method indicated as EVT-1)

FMECA Failure modes, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC Irradiation Assisted Stress Corrosion Cracking ICI In-Core Instrumentation IE Irradiation Embrittlement IGSCC Intergranular Stress Corrosion Cracking INPO Institute of Nuclear Power Operations IP Issue Programs ISi Inservice Inspection ISR Irradiation-Enhanced Stress Relaxation LRA License Renewal Application LRAAI License Renewal Application Action Item MRP Materials Reliability Program MSC Materials Subcommittee NDE Nondestructive Examination NGF Next Generation Fuel NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System Report No. 1500227.401.RO viii iJStructural Integrity Associates, Inc.

OE Operating Experience PDI Performance Demonstration Initiative PH Precipitation-Hardenable PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group PWSCC Primary Water Stress Corrosion Cracking RCS Reactor Coolant System RIS Regulatory Issue Summary RV Reactor Vessel RVI Reactor Vessel Internals sec Stress Corrosion Cracking SE Safety Evaluation SER *Safety Evaluation Report SRP Standard Review Plan SS Stainless Steel TLAA Time-limited Aging Analysis TE Thermal Embrittlement TS Technical Specifications UFSAR Updated Final Safety Analysis Report UT Ultrasonic Testing UGS Upper Guide Structure VT Visual Testing Report No. 1500227.401.RO IX

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1.0 INTRODUCTION

1.1 Objective The purpose of this document is to describe the potential aging concerns in the reactor vessel internals (RVI) at Arkansas Nuclear One, Unit 2 (AN0-2). This document also describes the mandatory and recommended guidance for managing potential aging concerns at AN0-2 through the period of extended operation, which begins at midnight on July 17, 2018. This Aging Management Program (AMP) document satisfies the license renewal commitment as contained in the AN0-2 license renewal application (LRA) [1]. This program coordinates with the AN0-2 inservice inspection (ISi) program [2] and supplements that program with augmented examinations for managing the potential aging effects of the RVI. This program plan establishes appropriate monitoring and inspections to maintain the reactor vessel internals functionality; the strategy is to assure nuclear safety and plant reliability. This document provides assurance that operations at AN0-2 will continue to be conducted in accordance with the current licensing bases (CLB) for the RVI, and it will provide the technical basis for managing the time-limited aging concerns for the duration of the piant by fulfilling the license renewal commitments. This document identifies the internals components that must be considered for aging management review and identifies the augmented inspection plan for the AN0-2 reactor vessel internals. The program plan supports the NEI 03-08 Guideline for the Management of Materials Issues [3], the EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A [4]), and the Applicant/Licensee Action Items in the NRC Safety Evaluation (SE) [5].

The main objectives of the AN0-2 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 [6].

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  • To define and implement the industry-defined (EPRI/MRP and PWROG) requirements and guidance for managing aging of RV internals.
  • To provide inspection plans for the AN0-2 RV internals.

1.2 AN0-2 Reactor Vessel Internals Inspection Program Commitment In order to meet license renewal Commitment 19 [1, 18], AN0-2 will submit this aging management program plan. This plan also addresses license renewal Commitment 15.

MRP-227-A [4] and the response to A/LAI 7 of the NRC SE address the CASS internals components. Therefore, the CASS internals components are managed under the MRP-227-A AMP. The license renewal commitments listed below define the content for the program that AN0-2 has committed to implement for the RVI Components:

Commitment 15: The Reactor Vessel Internals CASS Program will manage aging effects of cast austenitic stainless steel reactor vessel internals components. This program will supplement the reactor vessel internals inspections required by the ASME Section XI Inservice Inspection Program. The program will manage cracking, reduction offracture toughness, and dimensional changes using inspections of applicable components which will be determined based on the neutron jluence and thermal embrittlement susceptibility of the component. A description of the Reactor Vessel Internals CASS Program, which includes the inspection plan, will be submitted to the NRC for review and approval.

Commitment 19: The Reactor Vessel Internals Stainless Steel Plates, Forgings, Welds, and Bolting Program will manage aging effects of reactor vessel internals plates, forgings, welds, and bolting. This program will supplement the reactor vessel internals inspections required by the ASME Section Xl inservice inspection program. This program will manage the effects of crack initiation and growth due to stress corrosion cracking or irradiation assisted stress corrosion cracking, loss offracture toughness due to neµtron irradiation embrittlement, and distortion due to void swelling. This program will provide visual inspections and non-destructive examinations of reactor vessel internals.

In the development of this program, Entergy will support reactor vessel internals aging effects research through EPRI, the Materials Reliability Program, and other applicable Report No. 1500227 .40l.RO 1_2 !i)structura/ Integrity Associates, Inc.

industry efforts to better characterize the internals aging effects and to provide material property data to generate acceptance standards for inspections. Appropriate examination techniques will be selected based on the results of these industry efforts.

A description of this program, which includes the inspection plan, will be submitted to the NRC for review and approval.

Augmented inspections, based on required program enhancements resulting from the industry programs referred to in these license renewal commitments, will become part of the AN0-2 ASME B&PV Code,Section XI ISi program [2]. Corrective actions for augmented inspections will either be developed using repair and replacement procedures equivalent to those required in ASME B&PV Code,Section XI [9], or more rigorous procedures will be determined by AN0-2 independently or in cooperation with the industry. AN0-2 is currently committed to the 2001 Edition through 2003 Addenda of the ASME B&PV Code [9], and the development of this AMP is based on that Edition of the Code and MRP-227-A [4]. However, later Editions and Addenda of the Code or additional Code Cases or Safety Evaluation Reports (SERs) may be incorporated or invoked as necessary or required by 10 CFR 50.55a. The use ofMRP-227-A, as approved by the NRC, is consistent with current industry practice.

The Aging Management Program has been established so that the aging effects of the RVI components are adequately managed and to provide reasonable assurance that the internals components will continue to perform their intended function through the period of extended operation. Furthermore, this AMP will demonstrate the consistency of the program with the elements documented in NUREG-1801, Revision 2 [10], Chapter XI.M.16A, "PWR Vessel Internals." The operating experience provided by NUREG-1801, Revision 2 [l O] will also be reviewed and incorporated into plant-specific programs.

1.3 AN0-2 Reactor Vessel Internals Aging Management Program Background The managing of aging degradation effects in RVI is required for nuclear plants considering or entering license renewal, as specified in the NRC Standard Review Plan (SRP) for License Renewal Applications [11]. The U.S. nuclear industry has been actively engaged in supporting the industry goal ofresponding to these requirements. Various programs have been established Report No. 1500227.401.RO 1-3 iJ Structural Integrity Associates, Inc.

within the industry over the past decade to develop guidelines for managing the aging effects of Pressurized Water Reactor (PWR) RV internals. In 2001, Combustion Engineering Owners Group (CEOG) issued CE NPSD-1216 "Generic Aging Management Review Report for the Reactor Vessel Internals" [12]. Later, in 2008, MRP-227, Revision 0 [8] 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 (CE), Westinghouse, and Babcock & Wilcox (B&W).

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

[13 - 15]:

  • Screening criteria were developed, considering chemical composition, neutron fluence exposure, temperature history, and representative stress levels, for determining the relative susceptibility of PWR internals components to each of eight postulated aging mechanisms.
  • PWR internals components were categorized, based on the screening criteria, into categories that ranged from:

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

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

- 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 of subsequent inspections were developed.

Development of these strategies was based on combining the results of functionality assessment Report No. 1500227.401.RO 1-4

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with several contributing factors including component accessibility, operating experience, existing evaluations, and prior examination results.

The industry efforts, as coordinated_ by the EPRI MRP, has finalized the inspection and evaluation (I&E) guidelines for the RVI, and the NRC has endorsed this document by issuing a safety evaluation (SE). A supporting document addressing inspection requirements has also been completed. The industry guidance is contained in the following documents:

  • Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) [4] provides the industry background, listing ofreactor 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. MRP-227-A provides a standardized approach to PWR internals aging management for each unique reactor design (Westinghouse, B&W, and CE).

The NRC has endorsed MRP-227-A by issuing a safety evaluation (SE) [5].

  • MRP-228 [14], "Inspection Standard for PWR Internals," provides guidance on the qualification and demonstration of the NDE techniques and other criteria pertaining to the actual performance of the inspection.

The PWR Owners Group (PWROG) has developed and submitted for NRC review and approval WCAP-17096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements"

[16] for MRP-227-A inspections, where feasible. This document has been approved by the NRC through the issuance of a final safety evaluation [17]. Final reports are to be developed and be available for industry use in support of planned license renewal inspection commitments. Plant specific acceptance criteria can also be developed for some internals components if a generic approach is not practical.

The AN0-2 RVI are a part of the primary reactor coolant system (RCS), which is a two-loop CE designed nuclear steam supply system (NSSS).

A review of Section 2.3.1.2 of the AN0-2 LRA specifies that the RVI are comprised of the following component groups:

  • Control Element Assembly (CEA) Shroud Assembly Report No. 1500227.401.RO 1-5

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  • Core Shroud Assembly
  • Core Support Barrel (CSB) Assembly
  • Incore Instrumentation (ICI)
  • Lower Internals Assembly (Lower Support Structure Assembly)
  • Upper Internals Assembly (Upper Guide Structure Assembly)

The reactor coolant enters through the inlet nozzles of the reactor vessel, flows downward between the reactor vessel wall and the core barrel, passes through the flow skirt where the flow distribution is equalized, and then into the lower plenum. The coolant then flows upward through the core removing heat from the fuel rods. The heated coolant enters the outlet plenum where it flows around the outside of the CEA shrouds to the reactor vessel outlet nozzles. The CEA shrouds protect the CEAs from the effects of coolant crossflow in the outlet plenum.

The AN0-2 LRA lists the following functions for the reactor vessel internals [l]:

  • Provide support and orientation for the reactor core
  • Provide support, orientation, guidance, and protection of the CEAs
  • Provide a passageway for support, guidance, and protection for the incore instrumentation The function of the RVI is described in Section 4.1 of the AN0-2 Updated Final Safety Analysis Report (UFSAR) [18]. The reactor internals support and orient the fuel assemblies, CEAs, and incore instrumentation, and guide the reactor coolant through the reactor vessel. They also absorb the static and dynamic loads and transmit the loads to the reactor vessel flange. They will safely perform their functions during normal operating, upset and emergency conditions. The internals are designed to safely withstand the forces due to deadweight, handling, pressure differentials, flow impingement, temperature differentials, vibration and seismic acceleration.

AN0-2 was granted a license for extended operation by the NRC through the issuance of an SER in NUREG-1828 [19]. In the SER, the NRC concluded that, "On the basis of its evaluation of the Report No. 1500227.401.RO 1-6 e Structural Integrity Associates, Inc.

license renewal application, the NRC staff has determined that the requirements of 10 CFR 54.29(a) have been met." It was concluded that the AN0-2 LRA [1] adequately identified the RVI components that are subject to an Aging Management Review (AMR), and that the requirements of 10 CFR 54.21(a) [6] had been met. A listing of the AN0-2 RVI components and subcomponents subject to AMP requirements is summarized in Table 3.1.2-2 of the AN0-2 LRA [1].

1.4 AN0-2 Reactor Vessel Internals Aging Management Program Elements The key elements of the AN0-2 Reactor Vessel Internals Aging Management Program, based on the GALL process contained in NUREG-1801 [10], are outlined in Table 1-1. The program attributes are described in detail in Section 2.3 of this document. Additionally, Entergy participates in PWR Owners Group Materials Subcommittee 0PWROG MSC) and the MRP to focus on preventing material degradation, improve plant performance, sharing lessons learned from operating experience, and provide an effective interface with the NRC. As RVI examination experiences are shared amongst other utilities, MRP, and PWROG MSC, the RVI AMP key elements will be updated to include any relevant OE or lessons learned.

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Table 1-1. Key Elements of the Reactor Vessel Internals Aging Management Program Plan Attribute Attribute Description 1 Scope of Program The scope of this AMP is MRP-227-A [4] and the SE for MRP-227, Rev. 0

[5]. Supplemental inspections of RV internals are described in MRP-227-A

[4]. 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. l of this document.

2 Preventive Actions Preventive measures are described in Section 2.3.2 of this document.

3 Parameters AN0-2 monitors, inspects, and/or tests for the effects of the eight aging Monitored/Inspected degradation mechanisms on the intended function of the reactor vessel internals components as described in Section 2.3.3 of this document.

4 Detection of Aging The AN0-2 ASME Section XI ISI program [2] for B-N-3 internals Effects components and the additional locations identified in MRP-227-A [4], form the inspection plan for detection and monitoring of aging effects in the RV internals as described in Section 2.3.4 of this document.

5 Inspection Program for This program, in combination with the ASME Section XI ISI program [2],

Monitoring and provides direction for inspections required to support continued RV internals Trending component reliability as described in Section 2.3.5 of this document.

6 Acceptance Criteria Acceptance criteria used in the RV Internals Aging Management Program are based on the most appropriate ASME Section XI [9] and WCAP-17096

[16] criteria as described in Section 2.3.6 of this document.

7 Corrective Actions Components with identified relevant conditions shall be dispositioned as described in Section 2.3.7 of this document. The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition until the next planned inspection, or repair/replacement to remediate the relevant condition. Additional inspections of expansion category components may also be required as specified in MRP-227-A.

8 Confirmation Process The confirmation process for the RV Internals Program is described in Section 2.3.8 of this document.

9 Administrative Controls Administrative controls that apply to the RVI AMP, procedures, reviews and approval processes is described in Section 2.3.9 of this document 10 Operating Experience Operating experience related to the AN0-2 RV internals is described in Section 2.3.10 of this document.

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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 21. The RVI program is implemented in accordance with EN-DC-133 [20]. Entergy 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:

  • 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 [3] 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.

The ISI Engineer is responsible for:

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  • Planning and implementing inspections required by Section XI for B-N-3 components

[2], the supplemental inspections identified in this program plan, and any other plant-specific commitments for inspection for managing aging ofRVI.

  • Participating in industry groups such as Performance Demonstration Initiative (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 AN0-2' s overall strategy for managing aging in reactor vessel internals components is supported by the following existing programs:
  • Primary Chemistry Monitoring Program [22]

These are established programs that support the aging management of RCS components in addition to the RVI components.

1.6.1 ASME Section XI Inservice Inspection Program The AN0-2 Inservice Inspection (ISi) Program [2] is a plant-specific program encompassing ASME Section XI Subsections IWB, IWC, IWD and IWF requirements. The ISi Program manages cracking, loss of mechanical closure integrity, and loss of material due to wear of reactor coolant system piping and components. The program implements the applicable requirements of ASME Section XI, Subsections IWB, IWC, IWD, and IWF, and other requirements specified in 10CFR50.55a with approved NRC alternatives and reliefrequests. The ISi program supports aging management of the ASME Category B-N-3 RVI components. The ISi Program is updated as required to the latest ASME Section XI code edition and addendum approved by the Nuclear Regulatory Commission in 10CFR50.55a.

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1. 6.2 Primary Chemistry Monitoring Program The Primary Chemistry Monitoring Program at AN0-2 [22] is comparable to the program described in NUREG-1801,Section XI.M2, Water Chemistry [10].

The main objective of this program is to manage aging effects caused by corrosion and cracking mechanisms. The program relies on monitoring and control of water chemistry based on the EPRI guidelines for primary water chemistry [23].

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 AN0-2 license renewal application [1]. The AMR performed for the LRA submittal documents the results of the aging management review for the AN0-2 RVI. The NRC indicated its approval of the AN0-2 LRA in NUREG-1828 [19]. The RVI components specifically noted as requiring aging management areidentified in Table 3.1.2-2 of the LRA [1].

The assessments supporting the LRA performed the following:

1. Identified applicable aging effects requiring management
2. Associated aging management programs to manage those aging effects
3. Identified enhancements or modifications to existing programs, new aging management programs, or any other actions required to support the conclusions reached in the assessment AMRs were performed for each AN0-2 system that contained long-lived, passive components requiring an aging management review, in accordance wit.b the AN0-2 screening process. The results of these reviews have been incorporated into the AN0-2 RVI AMP.

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1.8 Industry Programs 1.8.1 CE NPSD-1216, Aging Management ofReactor Internals The Combustion Engineering Owner's Group (CEOG, now PWROG) topical report CE NPSD-1216 [12] contains a technical evaluation of aging degradation mechanisms and aging effects for CE RVI components. The CEOG report provided guidance for CEOG member plant owners to manage effects of aging on RVI during the period of extended operation, using approved aging management methodologies to develop plant-specific AMPs.

The AMR for the AN0-2 internals, documented in the AN0-2 license renewal application [1],

was completed in a manner consistent with the approach of CE NPSD-1216 [12]. Both the AN0-2 specific AMR document [24] and the generic CE document were completed to facilitate plant license renewal in accordance with 10 CFR Part 54 [6].

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. For details regarding this approach, refer to Section 2.2 of this document.

1.8.3 NEI 03-08 Guidance Within MRP-227-A The industry program requirements ofMRP-227-A are classified in accordance with the requirements of the NEI 03-08 [3] protocols. The MRP-227-A [4] guideline includes "mandatory," "needed," and "good practice" requirements defined as the following:

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  • Mandatory Each commercial US. PWR unit shall develop and document a PWR reactor internals aging management program within 36 months following issuance of MRP-227, Rev. 0 (that is, no later than December 31, 2011).

AN0-2 Applicability: MRP-227, Revision 0 was officially issued by the industry in December 2008 [8]. An aging management program was to be developed by December 2011. In order to meet this "Mandatory" requirement, an aging management program plan for AN0-2 was completed in November 2011 [7]. This AMP replaces the previous AMP [7] to conform to the updated requirements ofMRP-227-A.

AN0-2 qualifies as a Category B plant according to the NRC Regulatory Issue Summary (RIS) [25] being a plant with a renewed license that has a commitment to submit an AMP/inspection plan based on MRP-227-A, but has not yet been required to do so by their commitment. This AMP fulfills the license renewal commitment to submit a description of this program, including the inspection plan, to the NRC for review and approval.

  • Needed
1. Each commercial US. PWR unit shall implement Tables 4-1through4-9 and Tables 5-1through5-3 [of MRP-227-A] for the applicable design within twenty-four months following issuance of MRP-227-A.

AN0-2 Applicability: MRP-227 augmented inspections will be incorporated in the AN0-2 ISi program for the license renewal period. The applicable CE tables contained in MRP-227-A for RVI components are Table 4-2 (Primary), Table 4-5 (Expansion), Table 4-8 (Existing Programs), and Table 5-2 (Acceptance and Expansion Criteria) and are attached herein as Table 5-1, Table 5-2, Table 5-3, and Table 5-4, respectively.

This AMP has been developed in accordance with MRP-227-A [4].

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2. Examinations specified in the AfRP-227-A guidelines shall be conducted in accordance with Inspection Standard [AfRP-228].

AN0-2 Applicability: Inspections will be in accordance with the requirements of MRP-228 [14]. These inspection standards will be used for augmented inspections at AN0-2, as applicable, where required by MRP-227-A.

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

AN0-2 Applicability: AN0-2 will comply with this requirement [26].

4. Each commercial US. PWR unit shall provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the A1RP Program Manager within 120 days of the completion of an outage during which PWR internals within the scope of AfRP-227 are examined.

AN0-2 Applicability: AN0-2 will comply with this requirement.

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

AN0-2 Applicability: AN0-2 will comply with this requirement by using NRC-approved evaluation methodology (e.g. WCAP-17096 [16]).

1.8.4 MRP-227-A AMP Development Guidance In addition to the implementation of the requirements ofMRP-227-A in accordance with NEI Report No. 1500227.401.RO 1-14 e Structural Integrity Associates, Inc.

03-08, this RVI AMP addresses the 10 program elements as defined in the GALL Report Chapter Xl.M16A (provided in Section 2.3 of this Report) 1.8.4.1 MRP-227-A Applicability to AN0-2 The applicability ofMRP-227-A to AN0-2 requires compliance with the following MRP-227 assumptions:

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

Applicability: AN0-2 historic core management practices meet the requirements of MRP-227-A [27].

  • 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: AN0-2 operates as a base load unit [18, Section 10.2.1].

  • 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. AN0-2 has not made any modifications of the RVI components beyond those identified in general industry guidance or recommended by the vendor (CE) since the May 2007 effective date of this statement, and therefore meets this requirement ofMRP-227-A.

Hence, it is evident that operations at AN0-2 conform to the assumptions in Section 2.4 of MRP 227-A.

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1.8.5 Ongoing Industry Programs Entergy actively participates in the EPRI MRP, PWROG, and other activities related to PWR internals inspection and aging management and will address/implement industry guidance stemming from those activities, as appropriate under NEI 03-08 practices.

1.9 Summary The GALL Report identifies which reactor internals passive components are most susceptible to the aging mechanisms of concern. Additionally, this report identifies the appropriate inspections or mitigation programs needed to manage the aging mechanisms of the reactor vessel internals to assure these components will maintain their functionality through the period of extended operation. The GALL Report was used at AN0-2 for the initial basis of their LRA. The NRC has reviewed AN0-2's LRA and their approval is documented in NUREG-1828 [19].

The AN0-2 RVI AMP has been created to address the reactor vessel internals aging concerns consistent with the information identified in the GALL Report, the guidance in MRP-227-A, and the SE of MRP-227, Revision 0 issued by the NRC. AN0-2 will manage their RVI inspections through their augmented ISI program and will complete any repairs and/or replacements in accordance with ASME Code requirements and any NRC approved methodologies. The AN0-2 AMP will be updated accordingly as operating experiences and new inspection requirements and technologies evolve associated with managing reactor vessel aging concerns.

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2.0 AGING MANAGEMENT APPROACH 2.1 Mechanisms of Age-Related Degradation in PWR Internals Aging degradation mechanisms that impact RVI have been identified and documented in AN0-2-specific AMRs in support oflicense renewal [1, Table 3.1.2-2]. 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 ofMRP-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.

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.

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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 oft~e initiated crack. Corrosion fatigue is included in the degradation description.

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

Fatigue crack initiation and growth resistance 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 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.

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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 frqm 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 swelling may eventually result in cracking under stress.

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 Report No. 1500227.401.RO 2-3 IJ Structural Integrity Associates, Inc.

replaced with plastic,strain. Available data show that thermal stress relaxation appears to reach saturation in a short time (<1000 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 oc,curs 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 tum, 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 [4] define a supplemental inspection program for.

man~ging aging effects and provide generic guidance to help develop this aging management program for AN0-2. The EPRI MRP Reactor Internals Focus Group developed these guidelines to support the continued functionality ofRVI. 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 functionality assessment with component accessibility, operating experience, existing evaluations, and prior examination results. The aging management strategy that was developed was used in the development of an appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections. Additionally, it was also used to identify the components and locations for supplemental examinations by categorization.

MRP-227-A used a screening and ranking process to aid the identification ofrequired inspections for specific RVI components. The screening and categorization process also credited existing component inspections, when they were deemed adequate. Through the screening and Report No. 1500227 .40l.RO 2 _4 SJstructurallntegrity Associates, Inc.

/

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

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component exists or for which no highly susceptible component is accessible.

  • Expansion: Those PWR internals that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation o( aging management requirements for Expansion components will depend on the findings from the examination of the Primary components at individual plants.
  • Existing Programs: Those PWR internals that are susceptible to the effects of at least one of the eight aging mechanisms and for which generic and plant-specific existing AMP elements are capable of managing those effects, were placed in the Existing Programs group.
  • No Additional Measures: Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the No Additional Measures group. Additional components were placed in the No Additional Measures group as a result of FMECA and the functionality assessment. No further action is required by these guidelines for managing the aging of the No Additional Measures components.

A description of the categorization process used to develop the Guidelines is given below. The approach in these guidelines has been used to develop the AN0-2 AMP.

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In accordance with the MRP-227-A I&E Guidelines [4], this inspection strategy consists of the following:

  • Selection of the type of examination or other methodologies appropriate for each degradation mechanism
  • Specification of the required level of examination qualification
  • Schedule of first and frequency of any subsequent examinations
  • Sampling and coverage
  • Expansion of scope if sufficient evidence of degradation is observed
  • Examination acceptance criteria
  • Methods for evaluating examination results not meeting the examination acceptance criteria
  • Updating the program based on industry-wide results
  • Contingency measures to repair, replace, or mitigate The MRP first established a framework and strategy for the aging management of PWR internals components using proven and familiar methods for inspection, monitoring, surveillance, and communication. Based on the framework and strategy and on the accumulated industry research data, the following elements of an AMP were further developed [13, 15]:
  • Screening criteria were developed, considering chemical composition, neutron fluence exposure, temperature history, and representative stress levels, for determining the relative susceptibility of PWR internals components to each of eight postulated aging mechanisms.
  • PWR internals components were categorized, based on the screening criteria as follows:

- Components for which the effects of the postulated aging mechanisms are insignificant Report No. 1500227.401.RO 2-6 l)structural Integrity Associates, Inc.

- 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 AN0-2 Reactor Vessel Internals Aging Management Program Attributes The attributes of the AN0-2 RVI AMP and their compliance with the ten elements ofNUREG-1801 (GALL Report), Revision 2, Chapter XI.M16A, "PWR Vessel Internals" [10] are included in this section to ensure successful management of component aging.

This AMP is consistent with the GALL process and includes consideration of the augmented inspections identified in MRP-227-A [4]. Specific details of the AN0-2 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 RV! components at the Arkansas Nuclear One, Unit 2, which is built to a CE NSSS design. The scope of the program applies to the methodology and guidance in the most recently NRC endorsed version of MRP-227, which provides augmented inspection and.flaw evaluation methodology for assuring the functional integrity ofsafety-related internals in commercial operating U.S. PWR nuclear power plants designed by B&W, CE, and Westinghouse. The scope of components Report No. 1500227.401.RO 2-7 e Structural Integrity Associates, Inc.

consideredfor inspection under MRP-227 guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code,Section XI), those RV! components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(l), and other RV! components whose failure could prevent satisfactory accomplishment of any of the functions identified in JO CFR 54.4(a)(l)(i),

(ii), or (iii). The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review (AMR.), as defined by the acceptance criteria set in JO CFR 54.21(a)(l). 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.Ml, "ASME Code,Section XI Inservice Inspections, Subsections IWB, !WC, and I

IWD."

"The scope of the program includes the response bases to applicable license renewal applicant action items (LRAA!s) on the MRP-227 methodology, and any additional programs, actions, or activities that are discussed in these LRAAI responses and credited for aging management of the applicant's RV! components. The LRAA!s are identified in the staff's safety evaluation on MRP-227 and include applicable action items on meeting those assumptions that formed the basis of the MRP 's augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 of MRP-227), and NSSS vendor-specific or plant-specific LRAA!s as well. The responses to the LRAA!s 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 functionality analyses were based. These limitations and assumptions require a determination of applicability by the applicant for each reactor and are covered in Section 2.4 of MRP-227."

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2.3.1.1 AN0-2 Program Scope A description of the AN0-2 RVI design is provided in Section 3.0 of this program plan.

Additional details regarding the RVI are provided in the AN0-2 UFSAR [18]. The AN0-2 RVI subcomponents that require aging management review are indicated in the AN0-2 LRA [1].

Table 3.1.2-2 ofthe.AN0-2 LRA includes a summary of the results of the AMR. This table identifies the aging effects that require management. A column in the table lists the programs and activities at AN0-2 that are credited to address the aging effects for each management strategy presented in Table 3.1.2-2 of the AN0-2 LRA and Section 3.1.2.3.2 of the NRC's license renewal SER [19].

MRP-227-A provides the inspection and evaluation guidelines to develop plant specific programs to manage the effects of aging in PWR internals. MRP-227-A is also used as a guidance to develop an aging management program to satisfy license renewal commitments for the PWR fleet. A summary of the inspections required to be performed, the appropriate inspection techniques used to detect aging (i.e. cracking, loss of material, loss ofpreload, etc.),

frequency of inspections, and the acceptance criteria for the inspections are provided in MRP-227-A (summarized in Table 5-1 through Table 5-4 of this AMP). Guidance provided in MRP-227-A in conjunction with the guidance provided in the NRC SE [5] for MRP-227, Revision 0 and the GALL Report were reviewed to establish the basis for the AN0-2 RVI AMP.

The basic assumptions ofMRP-227-A, Section 2.4 are met by AN0-2 and are addressed in Section 1.8.4.1 of this AMP. The Topical Report Conditions and Applicant/Licensee Action Items provided by the NRC in the SE on MRP-227, Revision 0 [5] are met by AN0-2 and demonstration of compliance is addressed in Section 5.0. In addition, plant specific existing programs such as the Section XI ISi program for AN0-2 will complement the augmented inspection requirements provided in MRP-227-A in successfully managing the effects of aging for AN0-2 dµring the period of extended operation.

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2.3.1.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter XI.Ml6A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19], and the AN0-2 UFSAR [18]..

2.3.2 NUREG-1801/AMP Program Element 2: Preventive Actions "The guidance in MRP-227 relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material

  • induced by general, pitting corrosion, crevice corrosiOn, or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program. The program description, evaluation and technical bases of water chemistry are presented in GALL AMP Xl.M2, "Water Chemistry."

2.3.2.1 AN0-2 Preventive Action The AN0-2 RVI AMP includes the Primary Chemistry Monitoring Program [22] as an existing program that complies with the requirement of this element. A description and applicability to the AN0-2 RVI AMP is provided in the following subsection.

2.3.2.2 Primary Chemistry Monitoring Program The primary goal of this program is to mitigate loss of material due to general, pitting, and 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 AN0-2 water chemistry program [22] is based on current, approved revisions of EPRI PWR Primary Water Chemistry Guidelines [23].

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This program is consistent with the corresponding program described in Revision 2 for GALL Report [10]. The program description, evaluation, and technical basis of water chemistry are presented in GALL AMP XI.M2, "Water Chemistry."

The limits of known detrimental contaminants imposed by the water chemistry program are consistent with the EPRI PWR Primary Water Chemistry Guidelines [23].

2.3.2.3 Conclusion This element is consistent with the corresponding aging management program attribute in Revision 2 ofNUREG-1801 [10], Chapter XI.Ml6A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

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 RV! components at the facility: (a) cracking induced by SCC, PWSCC, JASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss offracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimension due to void swelling and irradiation growth, distortion, or deflection; and (e) loss ofpre load 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 ofmaterial, 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 Report No. 1500227.401.RO 2-11 s; Structural Integrity Associates, Inc.

toughness that is induced by thermal aging or neutron irradiation embrittlement, or by void swelling and irradiation growth; instead, the impact of loss of.fracture toughness on component integrity is directly managed by using visual or volumetric examination techniques to monitor for cracking in the components and by applying applicable reduced.fracture toughness properties in the flaw evaluations if cracking is detected in the components and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation under the MRP-227 guidance or ASME Code,Section XI requirements. The program uses physical measurements to monitor for any dimensional changes due to void swelling, irradiation growth, distortion, or deflection. "

"Specifically, the program implements the parameters monitored/inspected criteria for CE designed Primary Components in Table 4-2 of MRP-227. Additionally, the program implements the parameters monitored/inspected criteria for CE designed Expansion Components in Table 4-5 of MRP-227. The parameters monitored/inspectedfor Existing Program Components follow the bases for referenced Existing Programs, such as the requirements of the ASME Code Class RV! components in ASME Code, Section XL Table IWB-2500-1, Examination Categories B-N-3, as implemented through the applicant's ASME Code,Section XI program, or the recommended program for inspecting Westinghouse designed flux thimble tubes in GALL AMP Xl.M37, "Flux Thimble Tube Inspection. "No inspections, except for those specified in ASME Code,Section XI, 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 AN0-2 Parameters Monitored/Inspected AN0-2 monitors, inspects, and/or tests for the effects of the eight aging degradation mechanisms on the intended function of the reactor vessel internals components through inspection and condition monitoring activities in accordance with the augmented requirements defined under industry directives as contained in MRP-227-A [4] and ASME Section XI [9].

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This AMP implements the requirements for the Primary Component inspections from Table 4-2 ofMRP-227-A (Table 5-1 of this AMP), the Expansion Component inspections from Table 4-5 ofMRP-227-A (Table 5-2 of this AMP), and the Existing Component inspections from Table 4-8 ofMRP-227-A (Table 5-3 of this AMP). These tables contain requirements to monitor and inspect the RVI through the period of extended operation to address the effects of the eight aging degradation mechanisms.

For license renewal, the ASME Section XI Program [2] consists of periodic volumetric, surface, and/or visual examination of components for assessment, signs of degradation, and corrective actions. This program is consistent with the corresponding program described in the GALL Report [1 O].

2.3.3.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [IO], Chapter XI.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

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 Ji.,JRP-227 provides an introductory discussion andjustification 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, Ji.,JRP-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/ram general conditions to detection and sizing ofsurface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detecting and sizing of surface breaking e

discontinuities. "

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"Cracking caused by SCC, IASCC, andfatigue 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/or reduced fracture toughness properties, is known and has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss ofpre load caused by thermal and irradiation-enhanced stress relaxation and creep. "

"Jn addition, the program adopts the recommended guidance in MRP-227 for defining the Expansion criteria that need to be applied to inspections ofPrimary Components and Existing Requirement Components and for expanding the examinations to include additional Expansion Components. As a result, inspections performed on the R VJ components are performed consistent with the inspection frequency and sampling bases for Primary Components, Existing Requirement Components, and Expansion Components in MRP-227, which have been demonstrated to be in conformance with the inspection criteria, sampling basis criteria, and sample Expansion criteria in Section A.1.2.3.4 ofNRC Branch Position RSLB-1."

"Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for CE designed Primary Components in Table 4-2 of MRP-227 andfor CE designed expansion components in Table 4-5 of MRP-227."

"The program is supplemented by the following plant-specific Primary Component and Expansion Component inspections for the program (as applicable): for AN0-2, no Report No. 1500227.401.RO 2-14 e Structural Integrity Associates, Inc.

additional Primary or Expansion components are relevant to the scope of aging management for the RV!. "

Physical measurements: Per Revis,ion 2 ofNUREG-1801, this is not applicable for CE designed plants.

2.3.4.l AN0-2 Detection ofAging Effects Detection of indications required by the ASME Section XI ISi Program is well-established and field-proven through application of the Section XI ISi Program [2]. Those augmented inspections that are taken from the MRP-227-A recommendations will be applied through use of the MRP-228 [14] 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 the RVI, as contained in this program, are built around three basic inspection techniques: visual, ultrasonic, and physical measurement. The visual techniques include VT-3, VT-1, and EVT-1. hispection standards developed by the industry for application of these techniques in augmented RVI inspections are documented in MRP-228 [14]. Continued functionality can be confirmed by physical measurements to detect degradation mechanisms such as wear, or loss of functionality as a result of loss of preload or material deformation. If components have been shown to be flaw tolerant, the scope of the inspections for detection of aging effects may be modified. Acceptance criteria for each inspection technique are provided in Section 4.1 of this AMP.

2.3.4.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter XI.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

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

"The methods for monitoring, recording, evaluating, and trending the data that result from the program's inspections are given in Section 6 of MRP-227 and its subsections.

The evaluation methods include recommendations for flaw depth sizing and for crack growth determinations as well for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications. The examinations and re-examinations required by the MRP-227 guidance, together with the requirements specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects of the age-related degradation mechanisms within the scope of the 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 AN0-2 Monitoring and Trending Reporting operating experience with PWR internals has been generally proactive. Flux thimble wear and control rod, guide tube split pin cracking issues were identified by the industry and continue to be actively managed. The extremely low frequency of failure in reactor internals makes monitoring and trending based on operating experience somewhat impractical. The majority of materials aging degradation models and analyses used to develop the MRP-227-A guidelines are based on test data from RVI components removed from service. The data are used to identify trends in materials degradation and forecast potential component degradation. The industry continues to share both material test data and operating experience through the auspices of the MRP and PWROG. AN0-2 has in the past and will continue to maintain cognizance of Report No. 1500227.401.RO 2-16 e Str~ctura/ Integrity Associates, Inc.

industry activities and will continue to share operating experience information related to PWR internals inspection and aging management.

Inspections credited as part of the existing programs, where practical, are scheduled to be conducted in conjunction with typical IO-year ISi examinations, as documented in Section 5.0 of the AN0-2 ISi program [2].

Table 5-1 and Table 5-2 identify the inspection requirements for Primary and Expansion category components credited for aging management ofRVI. As discussed in MRP-227-A [4],

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 ofMRP-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 MRP guidelines.

2.3.5.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter XI.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

2.3.6 NUREG-1801/AMP Program Element 6: Acceptance Criteria "Section 5 of MRP-227 provides specific examination acceptance criteria for the Primary and Expansion Component examinations. For components addressed by Report No. 1500227.401.RO 2-17

!I)Structural Integrity Associates, Inc.

examinations referenced to ASME Code,Section XI, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.

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

  • For visual examination (and surface examination as an alternative to visual examination), the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sizedfor length by VT- I IEVT-1 examinations;
  • For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment for bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and
  • Physical measurements: Per Revision 2 of NUREG-1801, this is not applicable for CE designed plants. "

2.3.6.1 AN0-2 Acceptance Criteria Recordable indications that are the result of inspections required by AN0-2 existing ISi program scope [2] are evaluated in accordance with the requirements of the ASME Code and documented in the AN0-2 Corrective Action Process [26].

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 ofMRP-227 and WCAP-17096 are published and as the industry continues to develop and refine the information. Changes applicable to the AN0-2 RVI will be included as part of updates to this AMP.

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Recordable indications found during the MRP-227-A augmented inspections will be entered into the AN0-2 correction action program. These indications will be addressed by additional inspections, repair, replacement, mitigation, or analytical evaluations to further disposition these indications. Industry groups are working to develop a consistent set of tools compliant with approved methodologies to support this element. Additional analysis to establish evaluation acceptance criteria for "Expansion" category components has been developed by the PWROG in WCAP-17096-NP [16]. The status of these ongoing processes is monitored via Entergy participation in various industry programs related to aging management of PWR internals.

2.3.6.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter XI.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

2.3. 7 NUREG-1801/AMP Program Element 7: Corrective Actions:

"Corrective actions following the detection of unacceptable conditions are fundamentally provided for in each plant's corrective action program. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled 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 XI 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 ofSection 5 of the report. These include engineering evaluation methods, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures. The latter are subject to the requirements of the ASME Code,Section XI. The implementation of the Report No. 1500227 .40l.RO 2_19 iJstructural/ntegrity Associates, Inc.

guidance in MRP-227, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 5 0, Appendix B or its equivalent, as applicable. "

2.3. 7.1 AN0-2 Corrective Actions The existing AN0-2 procedure for the plant-specific Corrective Actions Program [26] is credited for this element. Repair and replacement activities will be performed in accordance with methodologies provided in Section 6 ofMRP-227-A [4] and ASME Code Section XI [9]. The corrective actions for existing Section XI (B-N-3) examinations will include the identification of a repair plan and verification of acceptability of replacements. Any indications found during the Section XI examinations for the RVI will be documented in the corrective action program [26].

These indications will be addressed by additional inspections, repair, replacement, or analytical evaluations in accordance with or equivalent to the requirements of the ASME Code,Section XI.

Actions to evaluate and monitor flaws or indications will be a part of the corrective action process. This evaluation guidance is included in MRP-227-A [4] and WCAP-17096-NP [16].

For example, the guidance provided in WCAP-17096-NP [16] may be used to evaluate component degradation that exceeds acceptance criteria in Section 5 ofMRP-227-A [4] when it is observed during required inspections. Other methods may also be used if approved by the NRC.

2.3.7.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter Xl.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

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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 JO CFR Part 50,,

Appendix B, or equivalent, as applicable. It is expected that the implementation of the guidance in MRP-227 will provide an acceptable level of quality for inspection, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with 10 CFR Part 5 0, Appendix B or their equivalent (as applicable), confirmation process, and administrative controls. "

2.3.8.J AN0-2 Confirmation Process AN0-2 has an established 10 CFR Part 50, Appendix B Program [28] that addresses the elements of corrective actions, confirmation process, and administrative controls. The AN0-2 Section XI Inspection Program [2] and Corrective Action Process [26] 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 [29].

2.3.8.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter XI.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

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 5 0 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. "

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2.3.9.1 AN0-2 Administrative Controls AN0-2 QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B which are acceptable in addressing administrative controls.

2.3.9.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter XI.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

2.3.10 NUREG-1801/AMP Program Element 10: Operating Experience "Relatively few incidents ofPWR internals aging degradation have been reported in operating US. commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-227-A. The applicant is expected to review subsequent operating experience for impact on its program or to participate in industry initiatives that perform this function.

The application of MRP-227 guidance will establish a considerable amount of operating experience over the nextfew years. Section 7 of MRP-227 describes the reporting requirements for these applications, and the plan for evaluating the accumulated additional operating experience. "

2.3.10.1 AN0-2 Operating Experience Extensive industry and AN0-2 operating experience (OE) has been reviewed during the development of the AN0-2 RVI AMP.

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Early plant operating experience related to hot functional testing and RVI is documented in plant historical records. Inspections performed as part of the 10-year ISi program have been conducted as designated by commitments and would be expected to discover general internals structure degradation. To date, little degradation has been observed industry-wide.

Industry and AN0-2 specific information relevant to aging has been compiled into the AN0-2 OE program [30]. 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 industry and plant-specific experience with RVI reveals that the U.S. nuclear industry, including AN0-2, has responded proactively to issues relative to RVI degradation.

Examples of AN0-2 proactivity are briefly described in the following paragraphs:

  • AN0-2 Flux Thimble Tube Replacement In the CE-designed plants, zirconium-base alloy thimbles exhibited growth due to irradiation.

This thimble growth was a major aging management issue, and the thimbles were subsequently replaced. AN0-2 has monitored the growth of the Zircaloy section of the thimble tube due to the high level of neutron radiation exposure and replaced ICI thimble tubes [32].

  • Participation in PWROG OE Activities:

Entergy participated in a PWROG project, which documented RVI aging degradation OE from the domestic and international PWR plants. The PWROG members were asked about prior inspections and results for the MRP-227-A RVI components. Entergy submitted survey responses detailing previous inspections, specific findings, and inspection timing. The results of this survey are documented in WCAP-17435-NP [33]. The OE in this report

'provides a benchmark from which to evaluate further RVI aging management events.

.\

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Entergy reported the following information for the MRP-227-A RVI components in the survey:

o Number of inspections per component o Details of each inspection per component o Record of indications for a given inspection o Subsequent corrective actions (response) or indications found

  • Cognizance oflndustry OE:

AN0-2 is committed to monitoring specific industry OE that could potentially affect the RVI during the period of extended operating at AN0-2 and at other domestic PWR facilities. For example, AN0-2 has monitored the emerging OE from the fuel leakage at a domestic CE designed power plant in fuel assemblies adjacent to the core shroud. Higher radiation levels may increase the susceptibility of stainless steel in the RVI to various material degradation mechanisms. These first burned peripheral fuel assemblies were detected at the specific plant and dispositioned. AN0-2 will incorporate related OE to ensure safe and reliable operation.

Industry OE published by the Institute of Nuclear Power Operations (INPO) and other informational sources is routinely reviewed, 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 key element of the MRP-227-A guideline is the reporting of age-related degradation of RVI components. Entergy, through its participation in EPRI MRP activities, will continue to benefit from the reporting of inspection information and will share its own OE with the industry through those groups or INPO, as appropriate.

2.3.10.2 Conclusion This element is consistent with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Chapter XI.M16A and Appendix A (Commitment 19) of the AN0-2 license renewal SER [19].

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3.0 AN0-2 REACTOR VESSEL INTERNALS DESIGN AND OPERATING EXPERIENCE The AN0-2 RVI are a part of the primary RCS, which is a two-loop CE designed NSSS. The AN0-2 reactor internals are designed to support and position the reactor core fuel assemblies and CEAs, provide hold-down for the fuel assemblies, absorb the dynamic loads and transmit these and other loads to the reactor vessel flange, provide flow paths for the reactor coolant and guide in-core instrumentation. The components of the reactor vessel internals are divided into sub-assemblies consisting of the core support structure, core shroud, flow skirt, upper guide structure assembly, and in-core instrumentation support system. The general arrangement of the AN0-2 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 3-10.

Descriptions of the reactor vessel internals assemblies are obtained from the AN0-2 UFSAR [18, Section 4.2.2].

3.1 Core Support Structure The major structural member of the reactor internals is the core support structure shown in Figure 3-2 and Figure 3-3. The core support structure consists of the core support barrel and the lower support structure. The material for the assembly is Type 304 stainless steel.

The core support structure is supported at its upper end by the upper flange of the core support barrel, which rests on a ledge in the reactor vessel. Alignment is accomplished by means of four equally spaced keys in the flange which fit into the keyways in the vessel ledge and closure head.

The lower flange of the core support barrel supports, secures and positions the lower support structure and is attached to the lower support structure by means of a welded flexural type connection. The lower support structure provides support for the core by means of a core support plate supported by columns mounted on support beams which transmit the load to the core support barrel lower flange. The core support plate provides support and orientation for the lower Report No. 1500227.401.RO 3-1 i.J Structural Integrity Associates, Inc.

ends of the fuel assemblies. The core shroud, which provides a flow path for the coolant and lateral support for the fuel assemblies, is also supported and positioned by the core support plate.

The lower end of the core support barrel is restricted from excessive radial and torsional movement by six snubbers which interface with the pressure vessel wall.

A. Core Support Barrel The core support barrel is a right circular cylinder including a heavy ring flange at the top

. ___ end and an internal ring flange at the lower end. The core barrel is supported from a ledge on the pressure vessel. The core support barrel, in turn, supports the lower support structure upon which the fuel assemblies rest. Press-fitted into the flange of the core support barrel are four alig~ent keys located 90 degrees apart. The reactor vessel, closure head and upper guide structure assembly flange are slotted in locations corresponding to the alignment key locations to provide proper alignment between these components in the vessel flange region.

The upper section of the barrel contains two outlet nozzles which interface with internal projections on the vessel nozzles to minimize leakage of coolant from inlet to outlet.

Since the weight of the core support barrel is supported at its upper end, it is possible that coolant flow could induce vibrations in the structure. Therefore, amplitude limiting devices, or snubbers, are installed on the outside of the core support barrel near the bottom end. The snubbers consist of six equally spaced lugs around the circumference of the barrel and act as a tongue-and-groove assembly with the mating lugs on the pressure vessel. Minimizing the clearance between the two mating pieces limits the amplitude of vibration. During assembly, as the internals are lowered into the pressure vessel, the pressure vessel lugs engage the core support barrel lugs in an axial direction. Radial and axial expansion of the core support barrel are accommodated, but lateral movement of the core support barrel is restricted. The pressure vessel lugs have bolted, captured, Inconel X shims and the core support barrel lug mating surfaces are hardfaced with Stellite to 3-2 estructural Integrity Associates, Inc.

Report No. 1500227.401.RO

minimize wear. The shims are machined during initial installation to provide minimum clearance. The snubber assembly is shown in Figure 3-4.

B. Core Support Plate and Lower Support Structure The core support plate is a Type 304 stainless steel plate into which the necessary flow distributor holes for the fuel assemblies have been machined. Fuel assembly locating pins are inserted into this plate.

The fuel assemblies and core shroud are positioned on the core support plate. This plate is welded to the top .of a cylindrical structure at the base of which is welded a bottom plate.

This structure seats on the lower flange of the core support barrel and transmits the lower support structure loads to the core support barrel. The core support plate is supported by an arrangement of columns welded at the base to support beams as shown in Figure 3-3.

The bottoms of the beams are welded to a bottom plate which contains flow holes for primary coolant flow. The ends of the beams are welded to the lower cylinder. The cylinder guides the main coolant flow and provides core shroud bypass flow by means of holes in the cylinder.

3.2 Core Shroud Assembly The core shroud provides an envelope for the core and limits the amounts of coolant bypass flow. The core shroud forms the perimeter of the core and acts as the transition structure between the rectilinear polygon core cross section and the cylindrical core support barrel. The shroud consists of two Type 304 stainless steel ring sections welded to each other and to the core support plate. The AN0-2 core shroud assembly is welded; there are no core shroud bolts.

A small gap is provided between the core shroud outer perimeter and the core support barrel in order to provide upward coolant flow between the core shroud and the core support barrel, thereby minimizing thermal stresses in the core shroud and eliminating stagnant pockets. The Report No. 1500227.401.RO 3-3 i}Structural Integrity Associates, Inc.

AN0-2 core shroud assembly is shown in Figure 3-5. Examples of the AN0-2 welded core shroud configuration from MRP-227-A [4] are shown in Figure 3-6 and Figure 3-7. Four equally spaced lugs are furnished on the top of the core shroud to provide alignment of the shroud with the fuel alignment plate.

3.3 Flow Skirt The Inconel flow skirt is a right circular cylinder perforated with flow holes. The flow skirt is used to reduce inequalities in core inlet flow distributions and to prevent formation of large vortices in the lower plenum. The skirt provides a nearly equalized pressure distribution across the bottom of the core support barrel. The skirt is supported by nine equally spaced, machined sections which are welded to the bottom head of the pressure vessel.

3.4 Upper Guide Structure Assembly This assembly consists of the upper guide structure support plate assembly, CEA shrouds and a fuel assembly alignment plate (Figure 3-8). The upper guide structure assembly aligns and laterally supports the upper end of the fuel assemblies, maintains the CEA spacing, holds down the fuel assemblies during operation, prevents fuel assemblies from being lifted out of position during a severe accident condition, protects the CEAs (Figure 3-9) from the effect of coolant crossflow in the upper plenum and supports the in-core instrumentation plate assembly. The

' c upper guide structure assembly is handled as one unit during installation and refueling.

The upper end of the assembly is a structure consisting of a support flange welded to the top of a cylinder. A support plate is welded to the inside of the cylinder approximately in the middle. The support plate is welded to a grid array of deep beams, the ends of which are welded to the cylinder. The support flange contains four accurately machined and located alignment keyways, equally spaced at 90 degree intervals, which engage the core barrel alignment keys. This system of keys and slots provides an accurate means of aligning the core with the closure head and thereby with the CEA drive mechanisms. The support plate aligns and supports the upper end of

\

the CEA shrouds. The shrouds extend from the fuel assembly alignment plate to an elevation Report No. 1500227.401.RO 3-4 e Structural Integrity Associates, Inc.

above the upper guide structure support plate. The CEA shroud consists of a cylindrical upper section welded to a base and a flow channel structure shaped to provide flow passage for the coolant through the alignment plate, while isolating the CEAs from crossflow. The shrouds are bolted and lockwelded to the fuel assembly alignment plate. At the upper guide structure support plate, the shrouds are connected to the plate by spanner nuts. The spanner nuts are tightened to proper torque to assure a rigid connection and lockwelded.

The fuel assembly alignment plate is designed to align the upper ends of the fuel assemblies and to support and align the lower ends of the CEA shrouds. Precision machined and located holes in the fuel assembly alignment plate engage machined posts on fuel assembly upper end fittings to provide accurate alignment. The fuel assembly alignment plate also has four equally spaced slots on its outer edge which engage with Stellite hardfaced pins protruding from the core shroud to limit lateral motion of the upper guide structure assembly during operation. The fuel alignment plate bears the upward force of the fuel assembly holddown devices. This force is transmitted from the alignment plate through the CEA shrouds to the upper guide structure support plate.

The flange of the upper guide structure support plate is designed to resist axial upward movement of the upper guide structure assembly and to accommodate axial differential thermal expansion between the core barrel flange, upper guide structure and pressure vessel flange support ledge and head flange recess.

3.5 In-Core Instrumentation Support System The complete in-core neutron flux monitoring system includes self-powered in-core detector assemblies, supporting structures and guide paths, and an amplifier system to process detector signals. The instrumentation supporting structures and guide paths are described in this section.

The support system begins outside the pressure vessel, penetrates the vessel boundary and terminates at the lower end of the fuel assembly. Each instrument is guided over its full length by the*extemal guidance conduit, the instrument plate structure guide tubes and the thimbles that extend downward into selected fuel bundles. The in-core instrumentation guide tubes route the instruments so that the detectors are located and spaced throughout the core. The guide tubes and Report No. 1500227.401.RO 3 _5 estructural Integrity Associates, Inc.

the in-core thimbles are attached to and supported by the instrument plate assembly shown in Figure 3-10.

The instrumentation plate assembly fits within the confines of the reactor vessel head and rests in the recessed section of the upper guide structure assembly. Its weight is supported by four bearing pins. The upper guide structure CEA shrouds extend through the instrumentation plate clearance holes. Above the instrumentation plate, the guide tubes bend and are gathered to form stalks which extend into the reactor vessel head instrumentation nozzles. The instrumentation plate assembly is raised and lowered during refueling to insert or withdraw all instruments and their thimbles simultaneously. The pressure boundaries for the individual instruments are at the instrumentation nozzle flange where the external electrical connections to the in-core instruments are also made.

The supporting structures for the in-core instruments are designed such that the temperature of the coolant surrounding the thermocouples in the in-core instruments is representative of fuel assembly outlet temperatures . The in-core instrument lengths and thimbles are designed to locate individual neutron detectors within a tolerance of +/-2 inches.

The assemblies have an integral seal plug which forms a seal at the instrument flange and through which the signal cables pass. Carbon packing rings fitted in a recess in the instrument flange are used to seal against operating pressure.

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- IN-CORE INSTRUMENTATION I SUPPORT PLATE

__ - - CEOM NOZZLE INSTRUMENTATION NOZZLE IN-CORE INSTRU ENT GUIDE TUBE CONTROL _ ALIGNMENT PIN ELEMENT ASSEMBLY FULLY UPPER GUIDE WITHDRAWN STRUCTURE 42"10 / * - 30* 10 INLET OUTLET NOZZLE NOZZLE CORE SUPPORT SURVEILLANCE BARREL HOLDER - --

150" ACTIVE

- CORE SHROUD CORE LENGTH FUEL ASSEMBLY SNUBBER - LOWER SUPPORT STRUCTURE CORE STOP FLOW SKIRT Figure 3-1. Illustration of the AN0-2 Vessel and Internals [18 , Figure 4.1-1]

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FLANGE ~

ALIGNME KEYS (4)

OUTLET -_.,...

NOZZLES CORE BARREL CORE SHROUD CORE SUPPORT PLATE SNUBBER\

LOWER SUPPORT - -**

ASSEMB LY Figure 3-2. AN0-2 Core Support Barrel Assembly [12, Figure 4.1-3]

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Core Support Barrel Core Support Plate

0

~----------

~

Lower Support Assembly Snubbers~

i.~

Figure 3-3. AN0-2 Lower Support Structure [12, Figure 4.1-5]

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Core Sta biliz ing Lug Hex Head Bolt (4 Req'd per Assy~ ~

Pin ------- ~

(4 Req'd per Assy)

Shim

. - (2 Req'd per A ssy)

YC SB Snubbers Figure 3-4. AN0-2 Snubber Assembly [18 , 4.2-1 O]

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90* /

Figure 3-5 . AN0-2 Core Shroud Assembly [18 , Figure 4.2-11]

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Weld locations potentiall

  • affected by swelling in horizontal stiffenef'i Core shroud plate-former plate weld locations with stresses potentially abo\*e LASCC threshold.

Weld locations potentially afft>cted by swelling in horizontal stiffener Figure 3-6. Potential Crack Locations for CE Welded Core Shroud Assembled in Stacked Sections

[4, Figure 4-12]

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Inspect for Separation of Upper and Lower Shroud Sections Figure 3-7. Locations of Potential Separation Between Core Shroud Sections Caused by Swelling Induced Warping of Thick Flange Plates in CE Welded Core Shroud Assembled in Stacked Sections [4, Figure 4-14]

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EXTENSION SHAFT GUIDE HOLD-DOWN RING UGS SUPPORT PLATE ASSY.

CEA SHROUD FUEL ALIGNMENT PLATE Figure 3-8 . AN0-2 Upper Guide Structure Assembly [12, Figure 4.1-2]

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END FITTING PISTON HOLDOOVN SPRI G

  • PLENUM
  • ~- SST SPACER B4C POISON PELLETS SST SPACER END CAP Figure 3-9. Control Element Assembly [31 , Figure 42]

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GUIDE TUBE CLUSTER ASSY GU IDE TUBE SUPPORT PLATE ASSY Figure 3-10. In-Core Support Assembly [12, Figure 4.1-8a]

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3.6 AN0-2 Design Distinctions Per Section 4.1 of the AN0-2 UFSAR [18]:

Arkansas Nuclear One - Unit 2 incorporates a Pressurized Water Reactor (PWR) with two reactor coolant loops. A vertical cross section of the reactor is shown in Figure 4.1-1. The reactor core is composed of 177 fuel assemblies and 81 Control Element Assemblies (CEAs). Th e fuel assemblies are arranged to approximate a right circular cylinder with an equivalent diameter of 123 inches and an active length of 150. 0 inches for fuel Batches A through Hand 149. 610 inches for fuel Batches J through N Th e active fuel length for Batch P (Cycle 12 reload batch) and subsequent reload batches is 15 0 inches. Each fuel rod shall contain a maximum total weight of 2114 grams of uranium.

The fuel assembly, which provides for 23 6 fuel rod positions, consists ofjive guide tubes welded or bulged to spacer grids and is closed at the top and bottom by end fittings. The welded construction was used for fuel Batches A through Y. A bulged construction was introduced in Batch Z with the implementation of Next Generation Fuel (NGF) . The guide tubes each displace four fuel rod positions and provide channels which guide the CEAs over their entire length of travel. In selected fuel assemblies, the central guide tube houses incore instrumentation.

3.7 AN0-2 Unit Operating Experience In the CE-designed plants, zirconium-base alloy thimbles exhibited growth due to irradiation.

This thimble growth was a major aging management issue, and the thimbles were subsequently replaced. AN0-2 has monitored the growth of the Zircaloy section of the thimble tube due to the high level of neutron radiation exposure and replaced ICI thimble tubes [32].

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4.0 EXAMINATION AND ACCEPTANCE AND EXPANSION CRITERIA 4.1 Examination Acceptance Criteria 4.1.1 Visual (VT-3) Examination Visual (VT-3) examination has been determined to be an appropriate NDE method for the detection of general degradation conditions in many of the susceptible components. The ASME Code Section XI, Examination Category B-N-3 [9], provides a set of relevant conditions for the visual (VT-3) examination ofremovable 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 in the Existing Programs group, these general relevant conditions are sufficient.

However, for components where visual (VT-3) is specified in the Primary or the Expansion group, more specific descriptions of the relevant conditions are provided in Table 5-1 through Table 5-4 of this document. Typical examples are "fractured material" and "completely separated material." One or more of these specific relevant condition descriptions may be applicable to the Primary and Expansion components listed in Table 5-1 and Table 5-2 of this document. The examination acceptance criteria for components requiring visual (VT-3) examinations is thus the absence of the relevant condition(s) specified in Table 5-4 of this document. The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued Report No. 1500227.401.RO 4-1

~Structural Integrity Associates, Inc.

operation with a known relevant condition, or repair/replacement to remediate the relevant condition.

Relevant conditions are defined in ASME Section XI, IWA-9000 [9]; 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 components, including such conditions as cracks, wear, corrosion, or erosion." For these guidelines VT-1 has only been selected to detect distortions as evidenced by small gaps between the upper-to-lower mating surfaces of CE welded core shroud assembled in two vertical sections. The examination acceptance criterion is thus the absence of the relevant condition of gaps that would be indicative of distortion from void swelling.

4.1.3 Enhanced Visual (EVT-1) Examination Enhanced visual (EVT-1) examination has the same requirements as the ASME Code Section XI visual (VT-1) examination, with additional requirements given in MRP-228 [14]. These enhancements are intended to improve the detection and characterization of discontinuities taking into account the remote visual aspect of reactor internals examinations. As a result, EVT-1 examinations are capable of detecting small surface-breaking cracks and sizing surface crack length when used in conjunction with sizing aides (e.g. landmarks, ruler, and tape measure). EVT-1 examination is the appropriate NDE method for detection of cracking in plates or their welded joints. Thus the relevant condition applied for EVT-1 examination is the same as for cracking in Section XI which is crack-like surface-breaking indications. The examination acceptance criterion for EVT-1 examination is the absence of any detectable surface-breaking indication.

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4.1.4 Surface Examination Surface ET (eddy current) examinations are specified as an alternative or as a supplement to visual examinations. No specific acceptance criteria for surface ET examination of PWR internals locations are provided in the ASME Code Section XI. Since surface ET is employed as a signal-based examination, a technical justification is documented in MRP-228 [14]. MRP-228 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 tpe same as the relevant condition for enhanced visual (EVT-1) examination. The acceptance criteria for enhanced visual (EVT-1) examinations are therefore applied when this method is used as an alternative or supplement to visual examination.

4.1.5 Volumetric Examination The intent of volumetric examinations specified for bolts and pins is to detect planar defects. No flaw sizing measurements are recorded or assumed in the acceptance or rejection of individual bolts or pins. Individual bolts or pins are accepted based on the lack of detection of any relevant indications established as part of the examination technical justification. When a relevant indication is detected in the cross-sectional area of the bolt or pin, that bolt or pin is assumed to be non-functional and the indication is recorded. A bolt or pin that passes the criterion of the examination is assumed to be functional.

Because there are no baffle-former bolts in the AN0-2 design, no volumetric examinations of the internals are needed to meet MRP-227-A requirements.

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. For CE designs, no physical measurements are specified; however, AN0-2 has a core barrel shroud assembled in Report No. 1500227.401.RO 4-3 e Structural Integrity Associates, Inc.

two vertical sections. If gaps between the two vertical sections of the core barrel shroud are identified during the required VT-1 examination required by MRP-227-A (see Table 5-1),

physical measurements must be performed for distortion in the gap between the top and bottom core shroud segments. See Section 5.5 for more details.

4.2 Expansion Criteria The criteria for expanding the scope of examination from the Primary components to their linked Expansion components are contained in Table 5-4.

4.3 Evaluation, Repair, and Replacement Strategy Any condition detected during examinations that do not satisfy the examination acceptance criteria of Section 4.1 shall be entered and dispositioned in the corrective action program.

The options listed below will be considered for disposition of such conditions. Selection of the most appropriate option(s) will be dependent on the nature and location of the indication detected.

1. Supplemental examinations will be used in order to further characterize and disposition a detected condition
2. Engineering evaluations that demonstrate the acceptability of detected conditions
3. Repair to restore a component with a detected condition to acceptable status
4. Replacement of a component The methodology used to perform engineering evaluations to determine the acceptability of a detected condition (item 2 above) shall be conducted in accordance with an NRC approved evaluation methodology. WCAP-17096-NP [16] and other NRC approved methodologies will be used to provide acceptance criteria for Primary and Expansion category items.

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4.3.1 Reporting Reporting and documentation of relevant conditions and disposition of indications that do not meet the examination acceptance criteria will be performed consistent with MRP-227-A and the AN0-2 Corrective Action Program. Entergy 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 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 ofreporting under the NEI 03-08, Materials Initiative Protocol [3].

4.4 Implementation Schedule The Program Enhancement and Implementation Schedule for AN0-2 is provided in Table 5-6.

4.5 Commitment Tracking A summary of actions related to the Aging Management of Reactor Vessel Internals for AN0-2 is provided in Table 5-7.

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5.0 RESPONSES TO NRC SAFETY EVALUATION APPLICANT/LICENSEE ACTION ITEMS As part of the NRC Revision 1 of the Final Safety Evaluation ofMRP-227 [5], a number of action items and conditions were specified by the staff. Table 5-5 documents AN0-2's conformance to the Topical Report Conditions and the Applicant/Licensee Action Items in the NRC Safety Evaluation of MRP-227 [5]. 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 andfunctionality analyses for reactors of their design (i.e., Westinghouse, CE or B&W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RV! 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-227.

The assumptions regarding plant design and operating history made in MRP-191 [13] are appropriate for AN0-2. The FMECA and functionality analyses were based on the assumption of 30 years of operation with high leakage core loading patterns followed by 30 years of low-leakage core fuel management strategy [27, Section 3.0]; therefore, AN0-2 is bounded by the assumption in MRP-191 [13].

As discussed in Section 1.8.4.1 of this document, operations at AN0-2 conform to the assumptions in Section 2.4 ofMRP-227-A [4].

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  • AN0-2 historic core management practices [27] meet the requirements of MRP-227-A [4]
  • AN0-2 operates as a base load unit [18, Section 10.2.1]
  • No design changes were implemented beyond those identified in general industry guidance or recommended by the vendor (CE or Westinghouse)

AN0-2 is actively participating in a joint industry program under the PWROG aimed at addressing the 20% cold work issue for non-weld or bolting austenitic stainless steel components on a generic rather than plant-specific basis. A discussion of this ongoing program (PA) follows:

PA-MSC-1288, PWR Materials Assessment, was discussed with the NRC at the June 2-4, 2015 Annual Materials Programs Technical Information Exchange Public Meeting (Ref.

ML15155B431). This PA utilizes a statistical approach for determining and assessing material or fabrication factors for PWR internals components. To date, plant-specific component manufacturing records have been gathered for over 50% of the domestic PWRs.

A review of these records in accordance with the guidance provided in MRP 2013-025 (ML1322A454) has revealed the following:

  • 20% cold work limitation was already recognized at the time of plant construction, i.e.

from 1970's

  • Plant fabricators quality programs were in place to adhere to limitations in cold work in austenitic stainless steels in these times
  • Plant specific assessments conducted to date confirm that no non-fastener materials contain cold work greater than 20%
  • Correlation of data based on searches to date demonstrates consistency across the PWR fleet - B&W, CE and, W show no cold worked non-fastener materials used in reactor vessel internals A final report (PWROG-15105-NP [34]) for this PA was issued in April 2016. AN0-2 will continue to participate in and follow the progress of PA-MSC-1288, including interactions with the NRC. If necessary, plant specific information will be provided to the NRC to supplement this joint industry program, if requested.

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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 RV! components are within the scope ofLR/or its facility. Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189, Revision 1, and Tables 4-4 and 4-5 in MRP-191 and identify whether these tables contain all of the R VI components that are within the scope of the LR for their facilities in accordance with JO CFR 54.4. If the tables do not identify all the RV! components that are within the scope ofLR/or its facility, the applicant or licensee.shall identify the missing component(s) and propose necessary modifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specific AMP. The AMP shall provide assurance that the effects of aging on the missing component(s) will be managedfor the period of extended operation.

The information contained in Table 4-5 ofMRP-191 [13] was reviewed and that review determined that this table contained all of the RVI components that are within the scope of license renewal for AN0-2. The aging management review performed as part of the AN0-2 LRA is described in Section 1.7.1 and summarized in Table 3.1.2-2 of the LRA.

5.3 SE Section 4.2.3, Applicant/Licensee Action Item 3 (Evaluation of the Adequacy of Plant-Specific Existing Programs):

As addressed in Section 3.2.5.3 in this SE, applicants/licensees of CE and Westinghouse are required to perform plant-specific analysis either to justify the acceptability of an applicant's/licensee's existing programs, or to identify changes to programs that should be implemented to manage the aging of these components for the period of extended operation. The results of this plant-specific analyses and a description of the plant-specific programs being relied on to manage aging of these components shall be submitted as part of the applicant's/licensee's AMP application. The CE and Westinghouse components identifiedfor this type ofplant-specific evaluation include: CE thermal shield positioning pins and CE in-core Report No. 1500227.401.RO 5-3 S}Structural Integrity Associates, Inc.

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).

The SE for MRP-227 [5] requires CE plants to evaluate whether existing plant-specific programs are adequate to manage the aging effects of thermal shield positioning pins and in-core instrument thimble tubes. AN0-2 complies with Applicant/Licensee Action 3 through management and replacement of in-core instrumentation thimble tubes as described in [32].

Thermal shields are not present in the AN0-2 reactor vessel internals.

5.4 SE Section 4.2.4, Applicant/Licensee Action Item 4 (B&W Core Support Structure Upper Flange Stress Relief):

As discussed in Section 3.2.5.4 of this SE, the B&W applicants/licensees shall confirm that the core support structure upper flange weld was stress relieved during the original fabrication of Reactor Pressure Vessel in order to confirm the applicability of MRP-227, as approved by the NRC, to their facility. If the upper flange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component. If necessary, the examination methods and frequency for non-stress relieved B&W core support structure upper flange welds shall be consistent with the recommendations in MRP-227, 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 3.3.1and4.3.1 of this SE. The applicant's/licensee's resolution of this plant-specific action item shall be submitted to the NRC for review and approval.

This action does not apply to CE designed units.

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5.5 SE Section 4.2.5, Applicant/Licensee Action Item 5 (Application of Physical Measurements as part ofl&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-227 for loss of compressibility for Westinghouse hold down springs, and for distortion in the gap between the top and bottom core shroud segments in CE units with core barrel shrouds assembled in two vertical sections. The applicant/licensee shall include its proposed acceptance .criteria and an explanation of how the proposed acceptance criteria are consistent with the plants ' licensing basis and the need to maintain the functionality of the component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227.

AN0-2 has a core barrel shroud assembled in two vertical sections. Per the examination coverage criteria defined in Table 4-2 ofMRP-227-A (see Table 5-1), if a gap between the two core barrel shroud vertical sections are identified during the required VT- I examination, three to five measurements of the gap opening from the core side at the core shroud re-entrant comers is required and an evaluation of the gap shall be performed to determine the frequency and method for additional examinations. Prior to performing the VT-I examination AN0-2 will develop acceptance criteria that are consistent with the licensing basis to ensure that the core shroud remains capable of performing its required functions.

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 baffle-to-baffle bolts and their locking devices, Report No. 1500227.401.RO 5-5 SJ Structural Integrity Associates, Inc.

B&W core barrel-to-former bolts and their locking devices, and B&W core barrel assembly internal baffle-to-baffle bolts. The MRP also identified that although the B&W core barrel assembly internal baffle-to-baffle bolts are accessible, the bolts are non-inspectable using currently available examination techniques.

Applicants/licensees shall justify the acceptability of these components for continued operation through the period of extended operation by performing an evaluation, or oy proposing a scheduled replacement of the components. As part of their application to implement the approved version of MRP-227, applicants/licensees shall provide their justification for the continued operability of each of the inaccessible components and, if necessary, provide their plan for the replacement of the components for NRC review and approval.

This action does not apply to CE designed units.

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

As discussed in Section 3.3.7 of this SE, the applicants/licensees ofB&W, CE, and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B&W !MI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operation or for additional RV! components that may be fabricated from 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 apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be Report No. 1500227 .40l.RO 5_6 i)structural Integrity Associates, Inc.

consistent with the plant's licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation. The applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227.

The SE for MRP-227 [5] requires the applicants/licensees of CE reactors to develop plant-specific analyses to be applied for their facilities to demonstrate that the CE lower support columns will maintain their functionality during the period of extended operation. It also requires that the licensee provide technical justification that other RVI components that may be fabricated from CASS, martensitic stainless steel, or precipitation hardened stainless steel materials will maintain their functionality during the period of extended operation. AN0-2 does not have lower support columns fabricated from CASS, martensitic stainless steel, or precipitation hardened stainless steel as part of the reactor vessel internals. The AN0-2 lower support columns are 304 stainless steel.

The CEA shroud tube is the only component fabricated from CASS material listed in Table 3.1.2-2 ofLRA [1], which is in Category A by MRP-191 [13]. This component item initially screened in for SCC (welds) and TE but the FMECA determined that these age-related degradation mechanisms have minimal likelihood to cause failure. Thus, this component was assigned to Category A per MRP-191. The *CEA s)lroud tubes are considered a no additional measures component per MRP-227-A and the existing inservice illspection program [2] is adequate to manage this CASS RVI component during the period of extended operation. The CEA shroud tubes are not considered a primary or expansion component for CE plants in Table 4-2 and Table 4-5 ofMRP-227-A [4].

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.J in this SE, applicants/licensee shall make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by this SE, as an Report No. 1500227.401.RO 5-7 SJ Structural Integrity Associates, Inc.

AMP for the R VI 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 ofInformation for Staff Review and Approval):

In addition to the implementation of MRP-227 in accordance with NE! 03-08, applicants/licensees whose licensing basis contains a commitment to submit a PWR RV! AMP and/or inspection program shall also make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by this SE. An applicant's/licensee's application to implement MRP-227, as amended by this SE shall include the following items (1) and (2).

Applicants who submit,applicationsfor LR after issuance of this SE shall, in accordance with the NUREG-1801, Revision 2, submit the information provided in the following items (1) through (5) for staff review and approval.

1. An AMP for the facility that addresses the 10 program elements as defined in NUREG-1801, Revision 2, AMP XI.M16A.

The attributes of the AN0-2 RVI AMP and their compliance with the ten elements of NUREG-1801 (GALL Report), Revision 2, Chapter Xl.M16A, "PWR Vessel Internals" [10]

that are essential for successful management of component aging are described in Section 2.3 of this document.

2. To ensure the MRP-227 program and plant-specific action items will be carried out by applicants/licensees, applicants/licensees are to submit an inspection plan which addresses the identified plant-specific action items for staff review and approval consistent with the licensing basis for the plant. If an applicant/licensee plans to implement an AMP which deviates.from the guidance provided in MRP-227, as approved by the NRC, the applicant/licensee shall identify where their program deviates from the recommendations of MRP-227, as approved by the NRC, and shall provide a justification Report No. 1500227.401.RO 5-8 e Structural Integrity Associates, Inc.

for any deviation which includes a consideration ofhow the deviation affects both "Primary" and "Expansion " inspection category components.

The aging management program plan for the AN0-2 will not deviate from the recommendations ofMRP-227-A. Inspection of Primary, Expansion, and components credited as part of plant specific existing programs provided in Table 5-1 through Table 5-4 of this document will be performed in accordance with the requirements ofMRP-227-A.

AN0-2 qualifies as a Category B plant according to the NRC Regulatory Issue Summary (RIS)

[25]. This AMP fulfills the license renewal commitment to submit a description of this program, including the inspection plan, to the NRC for review and approval.

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Table 5-1. CE Plants Primary Category Components from Table 4-2 ofMRP-227-A [4]

Expansion Examination Item Applicability Effect (Mechanism) Examination Coverage Comments Link<1l Method/Frequencyl1l Core Shroud Bolted plant Cracking (IASCC, Fatigue) Core support Baseline volumetric (UT) 100% of accessible bolts<3l. N/A Assembly designs column bolts, examination between 25 and Heads are accessible from the (Bolted) Aging Management barrel-shroud 35 EFPY, with subsequent core side. UT accessibility may Core shroud bolts (IE and ISR) <2l bolts examination on a ten-year be affected by complexity of interval. head and locking device (Not applicable for designs.

AN0-2)

See Figure 4-24 of MRP-227-A Core Shroud Plant designs Cracking (IASCC) Remaining Enhanced visual (EVT-1) Axial and horizontal weld seams AN0-2 must Assembly with core axial welds examination no later than 2 at the core shroud re-entrant perform the (Welded) shrouds Aging Management (IE) <2> refueling outages from the corners as visible from the core EVT-1 initial Core shroud assembled in beginning of the license side of the shroud, within six augmented plate-former plate two vertical renewal period and inches of central flange and inspection by weld sections subsequent examination on a horizontal stiffeners. refueling ten-year interval. outage 27 See Figure 3-6 and Figure 3-7 (Figures 4-12 and 4-14 of MRP-227-A)

Core Shroud Plant designs Cracking (IASCC) Remaining Enhanced visual (EVT-1) Axial weld seams at the core N/A Assembly with core axial welds, ribs examination no later than 2 shroud re-entrant corners, at the (Welded) shrouds Aging Management (IE) <2l and rings refueling outages from the core mid-plane (+/- three feet in Shroud plates assembled with beginning of the license height) as visible from the core full-height renewal period and side of the shroud.

(Not applicable for shroud plates. subsequent examination on a AN0-2) ten-year interval. See Figure 4-13 of MRP-227-A Core Shroud Bolted plant Distortion (Void Swelling), None Visual (VT-3) examination no Core side surfaces as indicated. N/A Assembly designs including: later than 2 refueling outages (Bolted)

  • Abnormal interaction with from the beginning of the See Figures 4-25 and 4-26 of Assembly fuel assemblies license renewal period. MRP-227-A
  • Gaps along high fluence Subsequent examinations on a (Not Applicable shroud plate joints ten-year interval.

for AN0-2)

  • Vertical displacement of '

shroud plates near high fluence joint Aging Management (IE)

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Table 5-1. CE Plants Primary Category Components from Table 4-2 ofMRP-227-A [4] (continued)

Expansion Examination Item Applicability Effect (Mechanism) Examination Coverage Comments Link<1l Method/Frequency(1)

Core Shroud Plant designs Distortion None Visual (VT-1) examination no If a gap exists, make three to AN0-2 must Assembly with core later than 2 refueling outages five measurements of gap perform the VT-1 (Welded) shrouds (Void Swelling), as evidenced from the beginning of the opening from the core side at initial augmented Assembly assembled in by separation between the license renewal period. the core shroud re-entrant inspection by two vertical upper and lower core shroud Subsequent examinations on a corners. Then, evaluate the refueling outage sections segments ten-year interval. swelling on a plant-specific 27 \

basis to determine frequency Aging Management (IE) and method for additional examinations.

See Figures 4-12 and 4-14 of MRP-227-A Core Support All plants Cracking (SCC) Lower core Enhanced visual (EVT-1) 100% of the accessible AN0-2 must Barrel Assembly support beams examination no later than 2 surfaces of the upper flange perform the Upper (core Core support refueling outages from the weld.<4> EVT-1 initial support barrel) barrel assembly beginning of the license augmented flange weld upper cylinder renewal period. Subsequent inspection by Upper core examinations on a ten-year See Figure 4-15 of refueling outage barrel flange interval. MRP-227-A 27 Core Support All plants Cracking (SCC, IASCC) Lower cylinder Enhanced visual (EVT-1) 100% of the accessible AN0-2 must Barrel Assembly axial welds examination no later than 2 surfaces of the lower cylinder perform the Lower cylinder Aging Management (IE) refueling outages from the welds <4> EVT-1 initial girth welds beginning of the license augmented renewal period. Subsequent inspection by examinations on a ten-year See Figure 4-15 of refueling outage interval. MRP-227-A 27 Lower Support All plants Cracking (SCC, IASCC, None Visual (VT-3) examination no 100% of the accessible AN0-2 must Structure Fatigue including damaged or later than 2 refueling outages surfaces of the core support perform the VT-3 Core support fractured material) from the beginning of the column welds<5 l initial augmented column welds license renewal period. inspection by Aging Management (IE, TE) See Figures 4-16 and 4-31 of refueling outage Subsequent examinations on a MRP-227-A 27 ten-year interval.

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Table 5-1. CE Plants Primary Category Components from Table 4-2 ofMRP-227-A [4] (continued)

Expansion Examination Item Applicability Effect (Mechanism) Examination Coverage Comments Linkl1> Method/Frequency(1)

Core Support All plants Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be AN0-2 must Barrel Assembly demonstrated by time limited defined by evaluation to perform the Lower flange weld aging analysis (TLAA), determine the potential EVT-1 initial enhanced visual (EVT-1) location and extent of fatigue augmented examination, no later than 2 cracking. inspection by refueling outages from the refueling outage beginning of the license See Figures 4-15 and 4-16 of 27 renewal period. Subsequent MRP-227-A examination on a ten-year interval.

Lower Support All plants with a Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be AN0-2 must Structure core support demonstrated by time limited defined by evaluation to perform the Core support plate Aging Management (IE) aging analysis (TLAA), determine the potential EVT-1 initial plate enhanced visual (EVT-1) location and extent of fatigue augmented examination, no later than 2 cracking. inspection by refueling outages from the refueling outage beginning of the license See Figure 4-16 of 27 renewal period. Subsequent MRP-227-A examination on a ten-year interval.

Upper Internals All plants with Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be N/A Assembly core shrouds demonstrated by time limited defined by evaluation to Fuel alignment assembled with aging analysis (TLAA), determine the potential plate full-height enhanced visual (EVT-1) location and extent of fatigue shroud plates examination, no later than 2 cracking.

(Not applicable for refueling outages from the AN0-2) beginning of the license See Figure 4-17 of renewal period. *Subsequent MRP-227-A examination on a ten-year interval.

Control Element All plants with Cracking (SCC, Fatigue) that Remaining Visual (VT-3) examination, no 100% of tubes in peripheral AN0-2 must Assembly instrument results in missing supports or instrument later than 2 refueling outages CEA shroud assemblies (i.e., perform the VT-3 Instrument guide guide tubes in separation at the welded joint guide tubes from the beginning of the those adjacent to the initial augmented tubes the CEA shroud between the tubes and within the CEA license renewal period. perimeter of the fuel alignment inspection by assembly supports shroud Subsequent examination on a plate). refueling outage assemblies. ten-year interval. 27 See Figure 4-18 of Plant-specific component MRP-227-A integrity assessments may be required if degradation is detected and remedial action is needed.

Report No. 1500227.401.RO 5-12 iJ Structural Integrity Associates, Inc.

Table 5-1. CE Plants Primary Category Components from Table 4-2 ofMRP-227-A [4] (continued)

Expansion Examination Comments Item Applicability Effect (Mechanism) Link<1l Examination Coverage Method/Frequency(1)

Lower Support All plants with Cracking (Fatigue) that results None Enhanced visual (EVT-1) Examine beam-to~beam N/A Structure core shrouds in a detectable surface- examination, no later than 2 welds, in the axial elevation Deep beams assembled with breaking indication in the refueling outages from the from the beam top surface to full-height welds or beams beginning of the license four inches below.

(Not applicable to shroud plates. renewal period. Subsequent AN0-2) Aging Management (IE) examination on a ten-year See Figure 4-19 of interval, if adequacy of MRP-227-A remaining fatigue life cannot be demonstrated.

Notes:

1. Examination acceptance criteria and expansion criteria for the CE components are in Table 5-4 (MRP-227-A Table 5-2)
2. Void swelling effects on this component is managed through management of void swelling on the entire core shroud assembly.
3. A minimum of75% of the total population (examined+ unexamined), including coverage consistent with the Expansion criteria in Table 5-4, must be examined for inspection credit.
4. A minimum of 75% of the total weld length (examined +unexamined), including coverage consistent with the Expansion criteria in Table 5-4, must be examined from either the inner or outer diameter for inspection credit.
5. A minimum of75% of the total population of core support column welds.

Report No. 1500227.401.RO 5-13 e Structural Integrity Associates, Inc.

Table 5-2. CE Plants Expansion Category Components from Table 4-5 ofMRP-227-A [4]

Examination Examination Item Applicability Effect (Mechanism) Primary Link Comments Method/Frequency <1> Coverage/Frequency <1l Core Shroud Bolted plant Cracking (IASCC, Core shroud Volumetric (UT) 100 % (or as supported by NIA Assembly designs Fatigue) bolts examination. plant-specific justification) <2>

(Bolted) of barrel-shroud and guide Barrel-shroud bolts Aging Management Re-inspection every 1O lug insert bolts with neutron (IE and ISR) years following initial fluence exposures > 3 (Not applicable for inspection. displacements per atom AN0-2) (dpa).

See Figure 4-23 of MRP-227-A Core Support Barrel All plants Cracking (SCC, Upper (core Enhanced visual (EVT-1) 100% of accessible welds Contingency if Assembly Fatigue) support barrel) examination. and adjacent base metal <2l. indications are found Lower core barrel flange flange weld. in EVT-1 exam of Re-inspection every 1O See Figure 4-15 of Upper (core support years following initial MRP-227-A barrel) flange weld inspection.

Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1) 100% of accessible surfaces Contingency if Assembly support barrel) examination. of the welds and adjacent indications are found Aging Management in EVT-1 exam of Upper cylinder flange weld. base metal <2 l.

(IE) Re-inspection every 1O Upper (core support (including welds) years following initial barrel) flange weld inspection. See Figure 4-15 of MRP-227-A Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1)

  • 100% of accessible bottom Contingency if Assembly support barrel) examination. surface of the flange <2l indications are found Upper Core Barrel flange weld in EVT-1 exam of Upper (core support Flange See Figure 4-15 of barrel) flange weld Re-inspection every 10 MRP-227-A years following initial inspection Report No. 1500227.401.RO 5-14 e Structural Integrity Associates, Inc.

Table 5-2. CE Plants Expansion Category Components from Table 4-5 ofMRP-227-A [4] (continued)

Examination Examination Item Applicability Effect (Mechanism) Primary Link Comments Method/Frequency (1) Coverage/Frequency <1>

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

See Figures 4-15 of MRP-227-A.

Lower Support All plants except Cracking (SCC, Upper (core Visual (EVT-1) examination. 100% of accessible Contingency if Structure those with core Fatigue) including support barrel) surface. (2 ) indications are found Lower core support shrouds assembled damaged or fractured flange weld Re-inspection every 10 in EVT-1 exam of beams years following initial See Figures 4-16 and 4-31 of Upper (core support with full-height material inspection. MRP-227-A. barrel) flange weld shroud plates Core Shroud Bolted plant Cracking (IASCC, Core shroud Ultrasonic (UT) 100 % (or as supported by N/A Assembly designs Fatigue) bolts examination. plant-specific analysis) of (Bolted) core support column bolts Core support column Aging Management Re-inspection every 10 with neutron fluence bolts (IE) years following initial exposures > 3 dpa. (2) inspection.

(Not applicable for See Figures 4-16 and 4-33 C?f AN0-2) MRP-227-A Core Shroud Plant designs with Cracking (IASCC) Shroud plates of Enhanced visual (EVT-1) Axial weld seams other than N/A Assembly core shrouds welded core examination. the core shroud re-entrant (Welded) assembled with Aging Management shroud corner welds at the core mid-Remaining axial welds, fUll-height shroud (IE) assemblies Re-inspection every 10 plane, plus ribs and rings.

Ribs and rings plates. years following initial inspection. See Figure 4-13 of (Not applicable for MRP-227-A AN0-2)

Report No. 1500227.401.RO 5-15 SJ Structural Integrity Associates, Inc.

Table 5-2. CE Plants Expansion Category Components from Table 4-5 ofMRP-227-A [4] (continued)

Examination Examination Item Applicability Effect (Mechanism) Primary Link Comments

- Method/Frequency <1> Coverage/Frequency <1>

Control Element All plants with Cracking (SCC, Peripheral Visual (VT-3) examination. 100% of tubes in CEA shroud Contingency if Assembly instrument guide Fatigue) that results in instrument guide assemblies. <2 > indications are found Remaining instrument tubes in the CEA missing supports or tubes within the Re-inspection every 1O in VT-3 exam of the guide tubes shroud assembly. separation at the CEA shroud years following initial See Figure 4-18 of peripheral instrument welded joint between assemblies. inspection. MRP-227-A guide tubes within the the tubes and supports. CEA shroud assemblies Notes:

1. Examination acceptance criteria and expansion criteria for the CE components are in Table 5-4 (MRP-227-A Table 5-2).
2. A minimum of75% coverage of the entire examination area or volume, or a minimum sample size of75% of the total population oflike components of the examination is required (including both accessible and inaccessible portions).

ReportNo. 1500227.401.RO 5-16 e Structural Integrity Associates, Inc.

Table 5-3. CE Plants Existing Program Components Credited in Table 4-8 ofMRP-227-A [4]

Effect Primary Examination Coverage and Item Applicability Examination Method Comments (Mechanism) Link Schedule Core Shroud All plants Loss of material ASME Visual (VT-3) examination, general First 10-year IS I after 40 years To be inspected Assembly (Wear) Code condition examination for detection of of operation, and at each Guide lugs Section XI excessive or asymmetrical wear. subsequent inspection Guide lug Aging interval.

inserts and Management bolts (ISR) Accessible surfaces at specified frequency Lower All plants with core Cracking (SCC, ASME Visual (VT-3) examination to detect Accessible surfaces at N/A Support shrouds assembled with IASCC, Fatigue) Code severed fuel alignment pins, missing specified frequency Structure full-height shroud plates Section XI locking tabs, or excessive wear on the fuel Fuel Aging alignment pin nose or flange.

alignment Management pins (IE and ISR)

(Not applicable for AN0"2)

Lower All plants with core Loss of Material ASME Visual.(VT-3) examination Accessible surfaces at To be inspected Support shrouds assembled in (Wear) Code specified frequency Structure two vertical sections Section XI Fuel Aging alignment . Management pins (IE and ISR)

Core Barrel All plants Loss of Material ASME Visual (VT-3) examination Area of the upper flange To be inspected Assembly (Wear) Code potentially susceptible to wear Upper flange Section XI Report No. 1500227.401.RO 5-17 e Structural Integrity Associates, Inc.

Table 5-4. CE Plants Examination Acceptance and Expansion Criteria from Table 5-2 ofMRP-227-A [4]

Examination Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Acceptance Criteria <1> Acceptance Criteria Core Shroud Assembly Bolted plant designs Volumetric (UT) a. Core support a. Confirmation that >5% of the a and b. The examination (Bolted) examination. column bolts core shroud bolts in the four acceptance criteria for the UT of plates at the largest distance the core support column bolts Core shroud bolts The examination from the core contain and barrel-shroud bolts shall be acceptance criteria for the b. Barrel-shroud unacceptable indications shall established as part of the (Not applicable for AN0-2) UT of the core shroud bolts bolts

/ shall be established as part require UT examination of the examination technical of the examination technical lower support column bolts justification.

justification. barrel within the next 3 refueling cycles.

b. Confirmation that > 5% of the core support column bolts

) contain unacceptable indications shall require UT examination of the barrel-shroud bolts within the next 3 refueling cycles.

Core Shroud Assembly Plant designs with core Visual (EVT-1) examination. Remaining axial Confirmation that a surface- The specific relevant condition is (Welded) shrouds assembled in two welds breaking indication > 2 inches in a detectable crack-like surface vertical sections length has been detected and indication.

Core shroud plate-former The specific relevant sized in the core shroud plate-plate weld condition is a detectable former plate weld at the core crack-like surface indication shroud re-entrant corners (as visible from the core side of the shroud), within 6 inches of the central flange and horizontal stiffeners, shall require EVT-1 examination of all remaining axial welds by the completion of the next refueling outage.

ReportNo. 1500227.401.RO 5-18 e Structural Integrity Associates, Inc.

Table 5-4. CE Plants Examination Acceptance Criteria and Expansion Criteria from Table 5-2 ofMRP-227-A [4] (continued)

Examination Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Acceptance Criteria <1> Acceptance Criteria Core Shroud Assembly Plant designs with core Visual (EVT-1) examination. a. Remaining axial a. Confirmation that a surface The specific relevant condition is (Welded) shrouds assembled with full- welds. breaking indication > 2 inches in a detectable crack-like surface height shroud plates length has been detected and indication.

Shroud plates The specific relevant b. Ribs and rings sized in the axial weld seams at condition is a detectable (Not applicable for AN0-2) the core shroud re-entrant crack-like surface indication corners at the core mid-plane shall require EVT-1 or UT examination of all remaining axial welds by the completion of the next refueling outage.

b. If extensive cracking is detected in the remaining axial welds, an EVT-1 examination shall be required of all accessible rib and ring welds by the completion of the next refueling outage.

Core Shroud Assembly Bolted plant designs Visual (VT-3) examination. None N/A N/A (Bolted)

The specific relevant Assembly conditions are evidence of abnormal interaction with (Not applicable for AN0-2) fuel assemblies, gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near. high fluence joints.

Report No. 1500227.401.RO 5-19 e Structural Integrity Associates, Inc.

Table 5-4. CE Plants Examination Acceptance Criteria and Expansion Criteria from Table 5-2 ofMRP-227-A [4] (continued)

Examination Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Acceptance Criteria <1> Acceptance Criteria Core Shroud Assembly Plant designs with core Visual (VT-1) examination. None N/A N/A (Welded) shrouds assembled in two vertical sections The specific relevant Assembly condition is evidence of physical separation between the upper and lower core shroud sections.

Core Support Barrel All plants Visual (EVT-1) examination. Lower core support Confirmation that a surface The specific relevant condition is Assembly beams breaking indication >2 inches in a detectable crack-like surface The specific relevant length has been detected and indication.

Upper (core support barrel) condition is a detectable Upper core barrel sized in the upper flange weld flange weld crack-like surface indication. cylinder (including - shall require that an EVT-1 welds) examination of the lower core support beams, upper core Upper core barrel barrel cylinder and upper core flange barrel flange be performed by the completion of the next refueling outage Core Support Barrel All plants Visual (EVT-1) examination. Lower cylinder axial Confirmation that a surface The specific relevant condition Assembly welds breaking indication >2 inches in for the expansion lower cylinder The specific relevant the length has been detected axial welds is a detectable Lower cylinder girth welds condition is a detectable and sized in the lower cylinder crack-like surface indication crack-like surface indication. girth weld shall require an EVT-1 examination of all accessible lower cylinder axial welds by the completion of the next refueling outage.

Lower Support Structure All plants Visual (VT-3) examination. None None N/A Core support column welds The specific relevant condition is missing or separated welds. J Report No. 1500227.401.RO 5-20 e Structural Integrity Associates, Inc.

Table 5-4. CE Plants Examination Acceptance Criteria and Expansion Criteria from Table 5-2 ofMRP-227-A [4] (continued)

Examination Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Acceptance Criteria <1> Acceptance Criteria Core Support Barrel All plants Visual (EVT-1) examination. None N/A N/A Assembly The specific relevant Lower flange weld condition is a detectable crack-like indication.

Lower Support Structure All plants with a core support Visual (EVT-1) examination. None N/A N/A plate Core support plate The specific relevant condition is a detectable crack-like surface indication.

Upper Internals Assembly All plants with core shrouds Visual (EVT-1) examination. None N/A N/A assembled with full-height Fuel alignment plate shroud plates The specific relevant condition is a detectable (Not applicable for AN0-2) crack-like surface indication.

Control Element All plants with instrument tubes Visual (VT-3) examination. Remaining Confirmed evidence of missing The specific relevant conditions Assembly in the CEA shroud assembly instrument tubes supports or separation at the are missing supports and The specific relevant within CEA shroud welded joint between the tubes separation at the welded joint Instrument guide tubes conditions are missing between the tubes and the assemblies and supports shall require the supports and separation at visual (VT-3) examination to be supports.

the welded joint between the expanded to the remaining tubes and the supports. instrument tubes within the CEA shroud assemblies by completion of the next refueling outage.

Report No. 1500227.401.RO 5-21 l)Structural Integrity Associates, Inc.

Table 5-4. CE Plants Examination Acceptance Criteria and Expansion Criteria from Table 5-2 ofMRP-227-A [4] (continued)

Examination Additional Examination Item Applicability Expansion Link(s) Expansion Criteria Acceptance Criteria <1> Acceptance Criteria Lower Support Structure All plants with core shrouds Visual (EVT-1) examination. None N/A N/A assembled with full-height Deep Beams shroud plates The specific relevant condition is a detectable (Not applicable for AN0-2) crack-like indication Note:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

Report No. 1500227.401.RO 5-22

!iJ Structural Integrity Associates, Inc.

Table 5-5. AN0-2 Response to the NRC Final Safety Evaluation of MRP-227-A [5]

MRP-227 SE Item AN0-2 Response SE Section 4.1.1, Topical Report In accordance with SE Section 4.1.1, the Lower Core Support Beams, Core Condition I: High consequence Support Barrel Assembly Upper Cylinder and Upper Core Barrel Flange components in the "No Additional have been added to the AN0-2 "Expansion" inspection category and are Measures" inspection category. contained in Table 5-2. The components are linked to the "Primary" components Core Support Barrel Upper (core support barrel) flange weld.

SE Section 4.1.2, Topical Report In accordance with SE Section 4.1.2, the Core Support Barrel Assembly Condition 2: Inspection of Lower Cylinder Girth Welds have been added to the AN0-2 "Primary" components subject to irradiation- inspection category and are contained in Table 5-1. The examination assisted stress corrosion cracking. method is consistent with the MRP recommendations for these components, the examination coverage conforms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination frequency is on a IO-year interval consistent with other "Primary" inspection category components.

SE Section 4.1.3, Topical Report In accordance with SER Section 4.1.3, the Core Support Column (casting Condition 3: Inspection of high or wrought) welds in the lower support structure have been added to the consequence components subject AN0-2 "Primary" inspection category and are contained in Table 5-1. The to multiple degradation examination method is consistent with MRP recommendations for these mechanisms components. The coverage confirms to the criteria described in Section 3.3.l of the NRC SE, and the re-examination frequency is on a IO-year interval consistent with the other "Primary" inspection category components.

SE Section 4.1.4, Topical Report In accordance with SE Section 4.1.4, AN0-2 will meet the minimum Condition 4: Imposition of inspection coverage specified in the SE. The appropriate wording has been minimum examination coverage added to Table 5-2 (Note 2) examination coverage.

criteria for "expansion" inspection category components SE 4.1.5, Topical Report Not applicable for AN0-2.

Condition 5: Examination frequencies for baffle former bolts and core shroud bolts Report No. 1500227.401.RO 5-23 i)Structural Integrity Associates, Inc.

Table 5-5. AN0-2 Response to NRC Final Safety Evaluation ofMRP-227-A [5] (continued)

MRP-227 SE Item AN0-2 Response SE 4.1.6, Topical Report In accordance with SE Section 4.1.6, Table 5-2 requires a 10-year re-Condition 6: Periodicity of the re- examination interval for 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. "Re-inspection every 10 years following initial inspection" is added to every component under the Examination Method/Frequency column in Table 5-2.

SE Section 4.1.7, Topical Report This condition applies to update of the industry guidelines. No plant-Condition 7: Updating ofMRP- specific actions are required.

227, Revision 0, Appendix A SE Section 4.2.1, The evaluation of design and operating history demonstrating that Applicant/Licensee Action Item 1: MRP-227-A is applicable to AN0-2 is contained in Section 1.8.4.1 and Applicability ofFMECA and Section 5.1 of this document.

Functionality Analysis Assumptions SE Section 4.2.2, The AN0-2 review of components within the scope of license renewal was Applicant/Licensee Action Item 2: compared against the information contained in Table 4-5 of MRP-191. The PWR Vessel Internals Aging Management Review performed as part of the AN0-2 LRA is Components Within the Scope of described in Section 1.7.1 of this document and summarized as part of License Renewal Applicant/Licensee Action Item 2 in Section 5.2.

SE Section 4.2.3, AN0-2 complies through management and replacement of in-core Applicant/Licensee Action Item 3: instrumentation thimble tubes as described in [32]. Thermal shields are not Evaluation of the Adequacy of present in the AN0-2 reactor vessel internals.

Plant-Specific Existing Programs SE Section 4.2.4, No action required. This action does not apply to CE designed units.

Applicant/Licensee Action Item 4:

B&W Core Support Structure Upper Flange Stress Relief

, Report No. 1500227.401.RO 5-24 e Structural Integrity Associates, Inc.

Table 5-5. AN0-2 Response to NRC Final Safety Evaluation ofMRP-227-A [5] (continued)

MRP-227 SE Item AN0-2 Response SE Section 4.2.5, AN0-2 has a core barrel shroud assembled in two vertical sections. Per the Applicant/Licensee Action Item 5: examination coverage criteria defined in Table 4-2 ofMRP-227-A (see Application of Physical Table 5-1 ), if a gap between the two core barrel shroud vertical sections are Measurements as part of I&E identified during the required VT-1 examination, three to five Guidelines for B&W, CE and measurements of the gap opening from the core side at the core shroud re-Westinghouse RVI Components entrant comers is required. Prior to performing the VT-1 examination AN0-2 will develop acceptance criteria that are consistent with the licensing basis to ensure that the core shroud remains capable of performing its required functions.

SE Section 4.2.6, No action required. This action does not apply to CE designed units.

Applicant/Licensee Action Item 6:

Evaluation oflnaccessible B&W Components SE Section 4.2.7, The CEA shroud tube is the only component fabricated from CASS Applicant/Licensee Action Item 7: material listed in Table 3 .1.2-2 of LRA [l ], which is in Category A by Plant Specific Evaluation of CASS MRP-191 [13]. This component item initially screened in for SCC (welds)

Materials and TE but the FMECA determined that these age-related degradation mechanisms have minimal likelihood to cause failure. Thus, this component was assigned to Category A per MRP-191. The CEA shroud tubes are considered a no additional measures component per MRP-227-A and the existing inservice inspection program is adequate to manage this CASS RVI component during the period of extended operation. The CEA shroud tubes are not considered a primary or expansion component for CE plants in Table 4-2 and Table 4-5 ofMRP-227-A [4].

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.

Report No. 1500227.401.RO 5-25 tJ Structural Integrity Associates, Inc.

Table 5-6. AN0-2 Program Enhancement and Implementation Schedule Refueling Cycle End AMP-Related Scope Inspection Method and Criteria Comments Outage QuarterNear 2R25 Spring 2017 Not applicable Not applicable Extended Operation Period begins Midnight on July 17, 2018 2R26 Fall 2018 ASME Section XI 10 Year ISi Inspections in accordance with Not applicable inspections of core shroud assembly AN0-2 ISi Program (guide lugs, guide lug inserts and bolts),

core barrel assembly (upper flange), and lower support structure (fuel alignment pins) 2R27 Spring 2020 Initial MRP-227-A augmented MRP-227-A inspections in AN0-2 plans to begin extended inspections for core shroud assembly accordance with MRP-228. operation during Cycle 26.

(core shroud plate-former plate weld, AN0-2 has the option to perform assembly), core support barrel assembly these inspections until 2R27, (upper core support barrel flange weld, which is no later than 2 refueling lower cylinder girth welds, lower flange outages from the beginning of the weld), lower support structure (core license renewal period.

support column welds, core support plate), and control element assembly (instrument guide tubes).

2R28 Fall 2021 Not applicable Not applicable Not applicable 2R29 Spring 2023 Not applicable Not applicable Not applicable 2R30 Fall 2024 Not applicable Not applicable Not applicable 2R31 Spring 2026 Not applicable Not applicable Not applicable 2R32 Fall 2027 Not applicable Not applicable Not applicable Report No. 1500227.401.RO 5-26 e Structural Integrity Associates, Inc.

Table 5-6. AN0-2 Program Enhancement and Implementation Schedule (continued)

Refueling Cycle End AMP-Related Scope Inspection Method and Criteria Comments Outage Quarter/Year 2R33 Spring 2029 ASME Section XI 10 Year ISi Inspections in accordance with Not applicable inspections of core shroud assembly AN0-2 ISi Program (guide lugs, guide lug inserts and bolts),

core barrel assembly (upper flange), and lower support structure (fuel alignment pins) 2R34 Fall 2030 Subsequent MRP-227-A augmented MRP-227-A inspections in Not applicable inspections for core shroud assembly accordance with MRP-228.

(core shroud plate-former plate weld, assembly), core support barrel assembly (upper core support barrel flange weld, lower cylinder girth welds, lower flange weld), lower support structure (core support column welds, core support plate), and control element assembly (instrument guide tubes).

2R35 Spring 2032 Not applicable Not applicable Not applicable 2R36 Fall 2033 Not applicable Not applicable Not applicable 2R37 Spring 2035 Not applicable Not applicable Not applicable 2R38 Fall 2036 Not applicable Not appJicable Not applicable NIA July, 2038 Not applicable Not applicable Renewed Operating License expires Midnight on July 17, 2038.

Report No. 1500227.401.RO 5-27 IJ Structural Integrity Associates, Inc.

Table 5-7. Summary of Actions Related to Aging Management ofRVI for AN0-2 Item AN0-2 Action Program/Action Description No.

1 AN0-2 Update to the Reactor Vessel Internals Inspection Program Commitments Submit the AN0-2 reactor vessel internals aging management program and inspection plans in accordance with MRP-227-A no later than two years prior to the period of extended operation (July 17, 2016).

2 AN0-2 will review plant specific and fleet operating experience based on updates to Appendix A of Reactor Vessel Internals Aging Management Program MRP-227-A and make updates to the RVI AMP as necessary.

3 Participation in Industry Groups (e.g. PWR Owners Group Materials Subcommittee, EPRI MRP) Reactor Vessel Internals Aging Management Program 4 Entergy Personnel Responsibilities: Ensure department specific actions are performed as it

  • Engineering Internals 5 Plant Specific Programs: The RVI AMP takes credit for plant specific
  • Primary Chemistry Monitoring Program specific programs (e.g. ASME Section XI ISI program components credited as part ofMRP-227-A inspections) 6 Examinations specified in the MRP-227-A guidelines shall be conducted in accordance with Reactor Vessel Internals Aging Management Inspection Standard MRP-228.

Program/ASME Section XI ISI Program 7 Examination results that do not meet the examination acceptance criteria defined in Section 5 of AN0-2 Procedure, EN-LI-102, "Correctiye Action MRP-227-A guidelines shall be recorded and entered in the plant corrective action program and Process."

dispositioned.

ReportNo. 1500227.401.RO 5-28 e Structural Integrity Associates, Inc.

Table 5-7. Summary of Actions Related to Aging Management ofRVI for AN0-2 (continued)

Item AN0-2 Action Program/Action Description No.

8 Each commercial U.S. PWR unit shall provide a summary report of all inspections and monitoring, Reactor Vessel Internals Aging Management Program items requiring evaluation, and new repairs to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals within the scope ofMRP-227-A are examined 9 If an engineering evaluation is used to disposition an examination result that does not meet the AN0-2 will comply with this requirement by using examination acceptance criteria in Section 5 ofMRP-227-A, this engineering evaluation shall be NRC-approved evaluation methodology (e.g.

conducted in accordance with an NRC-approved evaluation methodology. WCAP-17096)

IO Inspection acceptance and expansion criteria will be reviewed whenever new versions of the NRC Reactor Vessel Internals Aging Management Program approved versions of MRP-227 and WCAP-17096 are published, as the industry continues to develop and refine the information. Relevant changes based on the review of these NRC approved documents will be included as updates to the RVI AMP.

Report No. 1500227.401.RO 5-29 tJ Structural Integrity Associates, Inc.

6.0 REFERENCES

1. "Arkansas Nuclear One - Unit 2 License Renewal Application," October 15, 2003.
2. Entergy Nuclear Engineering Program, SEP-ISI-AN02-105, Revision 1, "Program Section for ASME Section XI, Division 1 ANO 2 Inservice Inspection Program,

February 13, 2014 (SI File No. 1500227.204).

3. NEI 03-08, Revision 2, "Guideline for the Management of Materials Issues," dated January, 2010.
4. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863.
5. Letter from Robert A. Nelson (NRC) to Neil Wilmshurst (EPRI) dated December 16, 2011, "Revision 1 to the Final Safety Evaluation ofEPRI 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. ML11308A770.

6. U.S. Code of Federal Regulations, Title 10, "Energy," Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."
7. Westinghouse Report No., WCAP-17432-NP, Revision 0, "Reactor Vessel Internals Program Plan for Aging Management of Reactor Internals at Arkansas Nuclear One, Unit 2," November 2011 (SI File No. 1500227.202).
8. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0). EPRI, Palo Alto, CA: 2008. 1016596.
9. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components, 2001 Edition through 2003 Addenda.
10. NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report, U.S.

Nuclear Regulatory Commission, December 2010.

6-1

(}Structural Integrity Associates, Inc.

Report No. 1500227.401.RO

11. NUREG-1800, Revision 2, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, December 2010.

1-2. Combustion Engineering Owners Group Report No., CE NPSD-1216, Revision 0, "Generic Aging Management Review Report for the Reactor Vessel Internals," CEOG Task 1185, March 2001 (SI File No. 1500227.203).

13. Materials Reliability Program: "Screening, Categorization, and Ranking ofReactor Internals Components of Westinghouse and Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.
14. Materials Reliability Program: Inspection Standard for Reactor Internals (MRP-228).

EPRI, Palo Alto, CA: 2009. 1016609. EPRI Proprietary

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