PNP 2012-080, Revised Program Plan for Aging Management of Reactor Vessel Internals

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Revised Program Plan for Aging Management of Reactor Vessel Internals
ML12257A352
Person / Time
Site: Palisades Entergy icon.png
Issue date: 09/13/2012
From: Gustafson O
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
PNP 2012-080 1101403.401, Rev 1
Download: ML12257A352 (124)
v • d • e


Text

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant t 27780 Blue Star Memorial Highway Covert, Ml 49043 Tel 269 764 2000 Otto W Gustafson Licensing Manager PNP 201 2-080 September 13, 2012 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Revised Program Plan for Aging Management of Reactor Vessel Internals Palisades Nuclear Plant Docket 50-255 License No. DPR-20

References:

1. Entergy Nuclear Operations, Inc. letter dated March 10, 2010, Program Plan for Aging Management of Reactor Vessel Internals (ADAMS Accession Number ML100710083)
2. Entergy Nuclear Operations, Inc. letter dated March 31, 2011, Withdrawal and New Commitment for Program Plan for Aging Management of Reactor Vessel Internals (ADAMS Accession Number ML110910461)
3. NRC Regulatory Issue Summary 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management, dated July 21, 2011 (ADAMS Accession Number ML111990086)
4. Entergy Nuclear Operations, Inc. letter dated June 20, 2012, Revised Commitment Date for Program Plan for Aging Management of Reactor Vessel Internals Submittal (ADAMS Accession Number ML12173A219)

Dear Sir or Madam:

In Reference 1, Energy Nuclear Operations, Inc. (ENO) submitted a program plan for aging management of reactor vessel internals for Palisades Nuclear Plant (PNP).

In Reference 2, ENO withdrew the program plan for aging management of reactor vessel internals and made a new commitment to submit a revised program plan. The revised program plan would be submitted within a year of the issuance of the final Nuclear Regulatory Commission (NRC) safety evaluation for the Materials Reliability Program report MRP-227, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines.

PNP 2011-080 Page 2 In Reference 3, the NRC stated plants with renewed licenses that have already submitted an inspection plan based on MRP-227, Revision 0, may withdraw their inspection plan (which had already been done for PNP) and resubmit an inspection plan in accordance with the approved MRP-227-A no later than October 1, 2012.

In Reference 4, ENO changed its commitment date to submit the revised program plan for aging management of reactor vessel internals, for PNP, in accordance with MRP 227-A, to no later than October 1,2012.

To complete its commitment, ENO is providing a revised reactor vessel internals aging management program plan for PNP in Attachment 1.

Attachment:

1. Palisades Reactor Vessel Internals Aging Management Program cc: Administrator, Region III, USNRC Project Manager, Palisades, USN RC Resident Inspector, Palisades, USNRC

ATTACHMENT 1 PALISADES REACTOR VESSEL INTERNALS AGING MANAGEMENT PROGRAM 121 pages follow

Report No. 1101403.401 Revision 1 Project No. 1101403 July 2012 Palisades Reactor Vessel Internals Aging Management Program Preparedfor:

Entergy Nuclear Operations, Inc.

Contract Order No. 10345958 Prepared by:

Structural Integrity Associates, Inc.

San Jose, California Prepared by: Date: 7/25/2012 Vikram Marthandam Reviewed by:

,j Date: 7/25/2012 Timothy J. Griesbach Approved by: Date: 7/26/2012 Timothy J. Griesbach jStrwturaI Integrity Associates, mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 2 of 121 Record of Revisions Rev. Date Description/Affected Pages A 6/1/12 Initial Draft.

0 6/23/12 Initial Issue.

1 7/26/12 Comment Resolution.

StructuraI Integrity Associates. kic

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 3 of 121 Table of Contents SECTION PAGE List of Acronyms 7 1.0 Introduction 9 1.1 Objective 9 1.2 Palisades License Renewal Background 9 1.3 Palisades Reactor Vessel Internals Aging Management Program Background... 10 1.4 Palisades Aging Management Program Intent 11 1.5 Palisades Reactor Vessel Internals Program Elements 11 1.6 Responsibilities 14

1.7 Purpose and Scope

15 2.0 Discussion 17 2.1 Mechanisms of Age-Related Degradation in PWR Internals 17 2.1.1 Stress Corrosion Cracking 17 2.1.2 Irradiation-Assisted Stress Corrosion Cracking 17 2.1.3 Wear 17 2.1.4 Fatigue 18 2.1.5 Thermal Aging Embrittlement 18 2.1.6 Irradiation Embrittlement 18 2.1.7 Void Swelling and Irradiation Growth 19 2.1.8 Thermal and Irradiation-Enhanced Stress Relaxation or Creep 19 2.2 Aging Management Strategy 20 2.3 Reactor Vessel Internals Aging Management Program Attributes 22 2.3.1 NUREG-1801/AMP Program Element 1: Scope of Program 22 2.3.2 NUREG-180 1/AMP Program Element 2: Preventive Actions 24 2.3.3 NUREG-180 1/AMP Program Element 3: Parameters Monitored/Inspected 25 2.3.4 NUREG-1801/AMP Program Element 4: Detection of Aging Effects.. ..25 2.3.5 NUREG-180 1/AMP Program Element 5: Monitoring and Trending 27 StnictwaIInteg,#JrAssociates,Inc?

Report No. 1101403.401R1 3

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 4 of 121 2.3.6 NUREG-180 1/AMP Program Element 6: Acceptance Criteria 28 2.3.7 NUREG-180 1/AMP Program Element 7: Corrective Actions 29 2.3.8 NUREG-1801/AMP Program Element 8: Confirmation Process 30 2.3.9 NUREG-1801/AMP Program Element 9: Administrative Controls 31 2.3.10 NUREG-1 801/AMP Program Element 10: Operating Experience 31 3.0 Palisades Reactor Vessel Internals Design and Operating Experience 34 3.1 Upper Guide Structure 34 3.2 Core Support Barrel 35 3.3 Lower Support Assembly 35 3.4 Core Shroud Assembly 35 3.5 Control Rod Shroud Assemblies 35 3.6 In-Core Instrumentation Support System 36 3.7 C-E Design Plants/Non-Relevant Operating Experience 50 3.8 Changes in Plant Operation 50 3.9 Conformance with MRP-227-A Assumptions 51 4.0 Program Description 52 4.1 Preventive Actions 52 4.2 Operational Experience 52 4.3 Component Inspection and Evaluation Overview 52 4.4 Examination of Reactor Vessel Internals 54 4.5 Inspection and Evaluation Requirements for Primary Components 56 4.6 Inspection and Evaluation Requirements for Expansion Components 56 4.7 Inspection of Existing Plant Components 56 4.8 Examination Systems (per MRP-228) 57 4.9 List of Vessel Internals Components for Examination 57 4.10 Inspection Schedule 57 4.11 Description of Existing Aging Management Programs 57 4.11.1 ASME Section XI Inservice Inspection Program of Vessel Internals...58 4.11.2 Water Chemistry Program 58 4.11.3 Industry Programs 59 4.12 Revised Commitments 59 5.0 Examination Acceptance and Expansion Criteria 60 5.1 Examination Acceptance Criteria 60 5.1.1 Visual (VT-3) Examination 60 Report No. 1101403.401 Ri 4 StnicIura.f Integrity Associates, Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 5 of 121 5.1.2 Visual (VT-i) Examination 60 5.1.3 Enhanced Visual (EVT-1) Examination 61 5.1.4 Surface Examination 61 5.1.5 Volumetric Examination 61 5.1.6 Physical Measurement Examination 62 5.2 Expansion Criteria 62 5.3 Evaluation, Repair and Replacement Strategy 63 5.3.1 Reporting 63 6.0 Operating Experience and Additional Considerations 64 6.1 Internal and External 64 7.0 Responses to the NRC Safety Evaluation Report Applicant/Licensee Action Items 65 7.1 SER Section 4.2.1, Applicant/Licensee Action Item 1 65 7.2 SER Section 4.2.2, Applicant/Licensee Action Item 2 65 7.3 SER Section 4.2.3, Applicant/Licensee Action Item 3 65 7.4 SER Section 4.2.4, Applicant/Licensee Action Item 4 66 7.5 SER Section 4.2.5, Applicant/Licensee Action Item 5 66 7.6 SER Section 4.2.6, Applicant/Licensee Action Item 6 66 7.7 SER Section 4.2.7, Applicant/Licensee Action Item 7 66 7.8 SER Section 4.2.8, Applicant/Licensee Action Item 8 66 8.0 References 68 Attachment A Section XI 10 Year ISI Examinations of B-N-2 and B-N-3 Internals Components for Palisades Nuclear Plant 89 Attachment B Palisades Reactor Vessel Internals Components in LRA and Aging Management Evaluation 93 Attachment C Planned Scope of Existing and Augmented Reactor Vessel Internals Examinations 110 Report No. 1101403.401R1 5 Sinjclural IaterIty Associates, Ilic?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 6 of 121 List of Tables TABLE NO. PAGE

1. Key Elements of the Reactor Internals Aging Management Program 13
2. C-E Plants Primary Components from MRP-227-A 72
3. C-E Plants Expansion Category Components from MRP-227-A 77
4. C-E Plants Existing Program Components Credited in MRP-227-A 80
5. C-E Plants Examination Acceptance and Expansion Criteria Applicable to Palisades Nuclear Plant 81
6. Palisades Response to the NRC Final Safety Evaluation of MRP-227-A 84
7. Program Enhancement and Implementation Schedule 86 List of Figures FIGURE NO. PAGE
1. Combustion Engineering Vessel and Internals Arrangement 37
2. Palisades Reactor Vessel Internals Arrangement 38
3. Palisades Bolted Core Shroud Assembly 39
4. Palisades Core Shroud Plate with Anchor Bolts 40
5. High Fluence Seam Locations in Baffle Former Assembly 41
6. Exaggerated View of Void Swelling Induced Distortion in Baffle Former Assembly 42
7. Palisades Fuel Bundle Alignment Plate and Upper Guide Structure 43
8. Typical C-E Core Support Barrel Structure 44
9. Palisades Core Support Barrel, Core Shroud Assembly, and Lower Support Structure 45
10. Palisades Lower Core Support Structure Assembly 46
11. Palisades Core Support Column 47
12. Palisades Control Rod Shroud Assembly and ICI Instrument Guide Tubes 48
13. Core Shroud Plate Showing Core Shroud Bolts 49
14. Palisades Reactor Vessel Internals Inspection Plan 88 Report No. 1 101403.401R1 6 AssovJates mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 7 of 121 LIST OF ACRONYMS AMD Aging Management Document AMP Aging Management Program AMR Aging Management Review ARDM Age-related degradation mechanism ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel CASS Cast austenitic stainless steel C-E Combustion Engineering CEOG Combustion Engineering Owners Group CFR Code of Federal Regulations CLB Current licensing basis EFPY Effective full power years ENO Entergy Nuclear Operations EPRI Electric Power Research Institute EVT Enhanced visual testing (visual NDE method indicated as EVT-1)

FMECA Failure modes, effects, and criticality analysis GALL Generic Aging Lessons Learned I&E Inspection and Evaluation IASCC Irradiation Assisted Stress Corrosion Cracking ICI In-Core Instrumentation IE Irradiation Embrittlement 151 Inservice Inspection ISR Irradiation-Enhanced Stress Relaxation LRA License Renewal Application MRP Materials Reliability Program NDE Nondestructive Examination NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OE Operating Experience PCS Primary Coolant System PH Precipitation Hardened PNP Palisades Nuclear Plant PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group RFO Refueling Outage Report No. 1101403 .401R1 7 ,SirucfumIhstegrIty Associates, mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 8 of 121 RPV Reactor Pressure Vessel RVI Reactor Vessel Internals SCC Stress Corrosion Cracking SER Safety Evaluation Report SS Stainless Steel TLAA Time-limited Aging Analysis TS Technical Specifications UT Ultrasonic Testing UGS Upper Guide Structure VT Visual Testing Report No. 1101403 .401R1 8 Associates, mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 9 of 121

1.0 INTRODUCTION

1.1 Objective This program document describes the potential aging concerns in the reactor vessel internals (RVT) and implements the industry recommended guidance for managing these aging concerns at the Palisades Nuclear Plant (PNP). This program document coordinates with the existing ASME Section XI inservice inspection (151) program and supplements that program with augmented examinations for managing the potential aging effects. This program document establishes appropriate monitoring and inspection programs to maintain the reactor vessel internals functionality; the strategy is to assure nuclear safety, plant reliability and to manage aging effects of reactor vessel internals. This document will provide assurance that PNP operations will continue to be conducted in accordance with the current licensing bases for the reactor vessel internals, and it will provide the technical basis for managing the time-limited aging concerns for the duration of plant life. This document identifies the internals components that must be considered for aging management review and identifies the augmented inspection plan for the PNP RVI. The program plan supports the NET 03-08 Materials Initiative Process [1], the NET 03-08 Guideline for the Management of Materials Issues [2], the EPRI Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A [3b]), and the Applicant/Licensee Action Items in the SE [4]. This program document meets the license renewal commitment for PNP for aging management of reactor vessel internals.

1.2 Palisades License Renewal Background In order to meet the license renewal commitment, Entergy Nuclear Operations (ENO) had planned to submit the aging management program document for PNP in March 2009. The original Commitment Item 33 specified that ENO would participate in industry initiatives that would generate additional data on aging mechanisms relevant to the RVIs, including void swelling, and would develop appropriate inspection techniques that would permit detection and characterization of features resulting from this effort and would incorporate them into the Reactor Vessel Internals Program as applicable.

However, one of the industry guidance documents needed to develop the program was the Electric Power Research Institute Material Reliability Program (EPRI/MRP) document, MRP 227-Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines

[3a]. However, in order to evaluate the applicability of MRP-227-A, PNP was required to show compliance with MRP-227-A assumptions. Section 3.9 provides a discussion on the applicability of MRP-227-A to the PNP RVI. In order to provide time for development of an aging management program that would include EPRI/MRP guidelines incorporated in the program basis document WCAP-17 133-NP [5], ENO requested a change in the submittal date to March 24, 2010. Subsequently, NRC inspectors reviewed the program basis document and Report No. 1 101403.401R1 9 ASfruC*11aIInteUrItY Associates. mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 10 of 121 implementing procedures and determined that the program was being implemented as described in the SER [6]. Based on the timeliness and adequacy of the licensee actions at that time, the inspectors determined that Commitment Item 33 had been met.

A new commitment not documented in Appendix A to the SER was initiated by ENO related to the program plan for aging management of the reactor vessel internals. WCAP-17133-NP [5]

was submitted to NRC on March 10, 2010. In a letter dated September 17, 2010, ENO received a request for additional information (RAI) [7] noting that the aging management report used MRP-227, Rev. 0 [3a] as the technical basis for developing the aging management program, and that document was still under NRC review. In order to resolve the RAI, ENO agreed to review the SER for MRP-227, Rev. 0, modify the aging management program, and resubmit the program plan. Subsequently, on July 20, 2012, ENO revised the commitment date [9] originally specified in Reference 8. This commitment is:

ENO will submit a revised program for aging management of reactor vessel internals in accordance with MRP-227, by October 1, 2012.

1.3 Palisades Reactor Vessel Internals Aging Management Program Background The PNP Reactor Vessel Internals (RVI) Aging Management Program (AMP) includes the 10 attributes ofNUREG-1801, Generic Aging Lessons Learned (GALL) Report, Rev. 2 [10],

Section XI.M16A, PWR Vessel Internals. The AMP credits inspections from the existing Section XI inspection program, credits industry programs issued under the guidance of NEI 08 as implemented by ENO Procedure [1], and credits previous industry operating experience.

In addition, the AMP incorporates recommendations for augmented inspections provided by Industry Guidelines contained in MRP-227-A [3b]. Existing Program aspects of this AMP, such as the ASME Section XI program and primary system chemistry monitoring, have been and continue to be on-going programs at PNP. These programs are supplemented as needed to form the basis for the PNP RVI AMP. The plant specific program elements that are credited for effective aging management of the PNP RVI are provided as part of this AMP.

This AMP for reactor internals takes effect in 2012. Augmented inspections for aging degradation mechanisms have been scheduled to coincide with the ASME Section XI, 10 year In-service Inspections (ISI) program requirements in the Fall of 2013 (RFO 23).

Report No. 1101403.401R1 10 Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 11 of 121 1.4 Palisades Aging Management Program Intent The PNP RVI AMP is developed to satisfy the Mandatory and Needed requirements in MRP-227-A. The reactor internals AMP utilizes a combination of prevention, mitigation and condition monitoring. Where applicable, credit is taken for existing programs such as inspections prescribed by the ASME Section XI In-Service Inspection Program [11], other existing plant programs such as water chemistry [12], and combined with augmented inspections or evaluations as specified in MRP-227-A [3b].

This document addresses the concerns for the following aging mechanisms:

  • Wear
  • Fatigue
  • Thermal Aging Embrittlement
  • Irradiation Embrittlement
  • Void Swelling and Irradiation Growth
  • Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation Enhanced Creep The degradation mechanisms used for the screening of PWR internals for susceptibility are discussed in Section 2.0.

Section 2.3, provides a description of the ten attributes of the GALL Report [10] and incorporates all available programs and activities that are credited for managing the aging effects produced by the mechanisms listed above.

1.5 Palisades Reactor Vessel Internals Program Elements Industry experience and research has shown that active degradation mechanisms may be present that could affect the ability of the internals components to perform their design functions.

Because of this, industry groups such as EPRI and other PWR Owners Groups began an effort to investigate these aging mechanisms, examine the materials of construction, consider the individual plant designs and operating conditions, and determine the internals components that may be susceptible to degradation and could potentially lead to loss of function.

To manage these aging concerns, the EPRI Materials Reliability Program (MRP) published the MRP-227, Rev. 0 guidelines document in December 2008 which contained Mandatory and Needed actions under the NET 03-08 Materials Initiative [2]. These guidelines were subsequently approved by NRC as MRP-227-A. Implementation of this Reactor Vessel Internals Aging Management Program fulfills MRP-227-A Mandatory and Needed requirements for Report No. 1 101403.401R1 11 StrucfuraIIntegrllyAssociaWs. klcL

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 12 of 121 PNP. Table 1 outlines the key elements of the PNP RVI AMP. The Program Attributes are described in more detail in Section 2.3. In addition, ENO actively participates in the PWR Owners Group Materials Subcommittee and the EPRI MRP. These industry groups actively manage generic work with a focus on improving plant performance and providing an effective interface with the NRC. Best practices and lessons learned are shared and discussed among members. ENO actively participates in these industry groups.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 13 of 121 Table 1. Key Elements of the Reactor Vessel Internals Aging Management Program:

Plan Attribute Approach and supplemental information 1 Scope of Program The scope of this AMP is MRP-227-A [3b] and SER [4]. Supplemental inspections of RVI are described in MRP-227-A [3b]. Additional actions and long range plans for aging management of internals are defined within this document. The scope of the program is described in more detail in Section 2.3.1.

2 Preventive Actions Preventive measures are described in Section 2.3.2.

3 Parameters ENO 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 through inspection and condition monitoring as described in Section 2.3.3.

4 Detection of Aging The PNP ASME Section XI [13] 151 program for B-N-2 and B-N-3 internals Effects components, and the additional locations identified in 1VIRP-227-A [3b], form the inspection plan for detection and monitoring of aging effects in the RVI as described in Section 2.3.4.

5 Inspection Program for This program, in combination with the ASME Section XI [13] ISI program, Monitoring and Trending provides direction for inspections required to support continued RV internals component reliability as described in Section 2.3.5 6 Acceptance Criteria Acceptance criteria used in the RVI Aging Management Document are based on the most appropriate ASME Section XI [13] and WCAP-17096 [14] criteria as described in Section 2.3.6.

7 Corrective Actions Components with identified relevant conditions shall be dispositioned as described in Section 2.3.7. 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 RVI Program is described in Section 2.3.8.

9 Administrative Controls This program is a support program of EN-DC-202, Rev. 5 [1] as described in Section 2.3.9.

10 Operating Experience Operating experience related to the PNP RVI is described in Section 2.3.10.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 14 of 121 1.6 Responsibilities The RVI Program Manager has overall responsibility for the development and implementation of the RVI aging management plan. The responsibilities for implementing the NET 03-08, Materials Initiative Process, are described in Reference 1. ENO actively participates in industry programs related to materials initiatives such as PWROG, EPRI MRP and other programs related to the 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 RVJ 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 IP guidelines in accordance with NET 03-08

[2] requirements,

  • Ensure prompt notification of the PCS Materials Degradation Management Program Manager whenever an issue or indication of potential generic industry significance is identified,
  • Participating in the planning and implementation of inspections of the internals.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 15 of 121 The 1ST Engineer is responsible for:

  • Planning and implementing inspections required by Section XI for category B-N-3 components [13], the supplemental inspections identified in this program plan, and any other plant-specific commitments for inspection for managing aging of the internals,
  • Participating in industry groups such as PDI, EPRI MRP TAC Inspection Subcommittee, etc.

The 151 Engineer and the Level III NDE Coordinator are responsible for:

  • Providing the NDE services,
  • Reviewing and approving the vendor NDE procedures and personnel qualifications,
  • Providing direction and oversight of contracted NDE activities,

1.7 Purpose and Scope

The purpose of the PNP RVI AMP is to manage the aging effects of reactor vessel internals through the license renewal period. This program coordinates with the existing ASME Section XI inservice inspection (1ST) program [13] and supplements that program with augmented examinations for managing the potential aging effects. The program establishes appropriate monitoring and inspection programs to maintain the reactor vessel internals functionality through the period of extended operation. This program is intended to provide assurance that PNP operations will continue to be conducted in accordance with the current licensing bases for the reactor vessel internals, including the incorporation of the license renewal commitments and the EPRT MRP-227-A Inspection and Evaluation (I&E) Guidelines [3b] for PWR Internals. This program plan supports the NE! 03-08 Materials Initiative Process [1].

In the PNP license renewal application [15] and in the SER [16] it is specified that these components of the RV internals provide structural and/or functional support to safety related equipment, are within the scope of license renewal (per GALL Rev. 0 [17]), and are subject to an aging management review:

  • IV.B3.2 Control element assembly shroud assemblies, including: control rod shroud, shroud support lug, fuel guide pin, fuel guide pin nuts, shroud top support, control rod support Lug, fuel plate cap screw
  • IV.B3.4 Core shroud assembly: anchor block, centering plate, core shroud plate, anchor screw & pin, centering screw & pin, positioning screw, shroud bolt & pin ReportNo. 1101403.401R1 15 SLnJcLul8IIutegrItyAssoc?ates,Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 16 of 121

  • IV.B3.3 Core support barrel assembly core support barrel, core support barrel integral upper flange
  • Incore instrument guide tube (not addressed in GALL IV.B3), instrument guide tube, guide tube bracket, guide tube plugs, guide tube plug screw, guide tube support
  • IV.B3.5 Lower internal assembly core support plate, core support column, core support barrel snubber lug, core support barrel cap screws, fuel alignment pins, core support beams and ties rods Note: The PNP design does not have core shroud tie rods or fuel alignment pins.
  • Upper guide structure (not addressed in GALL) spacer shim, instrument sleeve
  • IV.B3.1 Upper internal assembly fuel alignment plate, fuel plate align nut, fuel plate cap screw, fuel plate guide pin, hoiddown ring plunger, holddown ring strap, hoiddown ring, brace grid beam, cross brace screw, shroud grid ring These internals components were also identified in WCAP-17133-NP, Appendix B [5] as components to be included in the PNP RVI AMP; however, there was no further disposition provided in that document to specify the method(s) by which the component aging effects were being managed. Therefore, an additional aging management review of these internals components was performed to determine which of the existing programs were managing the aging effects and, if not covered by those programs, to identify any remaining components for inclusion in the RVI Inspection Program based on the latest industry guidelines for managing internals (Attachment B). The details of the additional aging management review are provided in Section 3.0. This review is subject to the confirmation and verification that the programs being credited for aging management of these components are in place and are being implemented properly.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 17 of 121 2.0 Discussion The reactor internals assembly is a part of the primary coolant system (PCS). The reactor internals are long-lived passive structural components designed to support the functions of PCS core cooling, control rod drive (CRD) insertion, and integrity of the fuel and pressure vessel boundary.

The core support structures provide support and restraint of the core. Static (deadweight and mechanical) loads from the assembled components, fuel assemblies, and dynamic loads (hydraulic flow, flow-induced vibration, thermal expansion, and seismic) and LOCA loads are all carried by the core support structures. In addition to core support, the various internals assemblies provide a flow boundary to direct coolant flow from the cold leg inlet nozzles, down the annulus to the lower plenum, past the core, into the upper plenum region above the core, and out the outlet nozzles to the hot leg piping.

2.1 Mechanisms of Age-Related Degradation in PWR Internals The EPRI MRP program considered the potential aging mechanisms that could affect PWR internals for the long term operation. Of particular concern are those aging mechanisms that could have an impact on component functionality. The age-related degradation mechanisms used for the screening of the PWR internals for susceptibility were as follows:

2.1.1 Stress Corrosion Cracking Stress Corrosion Cracking (SCC) refers to local, non-ductile cracking of a material due to a combination of tensile stress, environment, and metallurgical properties. The actual mechanism that causes SCC involves a complex interaction of environmental and metallurgical factors. The aging effect is cracking. If components are susceptible to SCC and require aging management, EVT-1 exams would be performed per MRP-227-A.

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. For those components identified as being susceptible to IASCC and would require aging management, EVT- 1 or UT exams would be performed in accordance with MRP-227-A.

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. Components that are susceptible to wear are to be inspected by visual (VT-3) exams.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 18 of 121 2.1.4 Fatigue Fatigue is defined as the structural deterioration that can occur as the result of repeated stress/strain cycles caused by fluctuating loads and temperatures. After repeated cyclic loading of sufficient magnitude, microstructural damage can accumulate, leading eventually to macroscopic crack initiation at the most highly affected locations. Subsequent mechanical or thermal cyclic loading can lead to growth of the initiated crack.

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

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

2.1.5 Thermal Aging Embrittlement Thermal aging embrittlement is the exposure of delta ferrite within cast austenitic stainless steel (CASS) and precipitation-hardenable (PH) stainless steel to high inservice temperatures, which 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. It is noted that there are no CASS or PH components in the PNP lower core support structures [28].

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 ReportNo. 1101403.401R1 18 StnjcfumIhiteIIyAssociates,Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 19 of 121 effect if a crack is present and the local applied stress intensity factor exceeds the reduced fracture toughness.

2.1.7 Void Swelling and Irradiation Growth Void swelling is defined as a gradual increase in the volume of a component caused by formation of microscopic cavities in the material. These cavities result from the nucleation and growth of clusters of irradiation produced vacancies. Helium produced by nuclear transmutations can have a significant impact on the nucleation and growth of cavities in the material. Void swelling may produce dimensional changes that exceed the tolerances on a component. Strain gradients produced by differential swelling in the system may produce significant stresses. Severe swelling

(>5% by volume) has been correlated with extremely low fracture toughness values. Also included in this mechanism is irradiation growth of anisotropic materials, which is known to cause significant dimensional changes in in-core instrumentation tubes fabricated from zirconium alloys.

(Note: The PNP RVI do not include any zirconium alloys.) While the initial aging effect is dimensional change and distortion, severe void swelling may eventually result in cracking under stress. Aging management of void swelling is by visual inspection targeted at locations where swelling is most likely to occur.

2.1.8 Thermal and Irradiation-Enhanced Stress Relaxation or Creep The loss of preload aging effect can be caused by the aging mechanisms of stress relaxation or creep. Thermal stress relaxation (or, primary creep) is defined as the unloading of preloaded components due to long-term exposure to elevated temperatures, such as seen in PWR internals.

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

Creep (or more precisely, secondary creep) is a slow, time and temperature dependent, plastic deformation of materials that can occur when subjected to stress levels below the yield strength (elastic limit). Creep occurs at elevated temperatures where continuous deformation takes place under constant strain. Secondary creep in austenitic stainless steels is associated with temperatures higher than those relevant to PWR internals even after taking into account gamma heating.

However, irradiation-enhanced creep (or more simply, irradiation creep) or irradiation-enhanced stress relaxation (ISR) is a thermal process that depends on the neutron fluence and stress; and, it can also be affected by void swelling, should it occur. The aging effect is a loss of mechanical closure integrity (or, preload) that can lead to unanticipated loading which, in turn, may eventually cause subsequent degradation by fatigue or wear and result in cracking.

Report No. 11 01403.401R1 19 117 1181 I1JIeVIIY Associates, Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 20 of 121 2.2 Aging Management Strategy The MRP-227-A [3b] Guidelines define a supplemental inspection program for managing aging effects and provide generic guidance to help develop this AMP for PNP. The EPRI MRP Reactor Internals Focus Group developed the Guidelines to support the demonstration of continued functionality, with requirements for inspections to detect the effects of aging degradation with requirements for the evaluation of those aging effects. 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 to determine the appropriate aging management methodology, baseline examination timing, and the need and timing of subsequent inspections and identified the components and locations for supplemental examination by categorization.

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 PNP RVI AMP.

In accordance with the MRP-227-A I&E Guidelines [3b], this inspection strategy consists of the following:

  • Selection of the type of examination appropriate for each degradation mechanism,
  • Specification of the required level of examination qualification,
  • Schedule of first inspection and frequency of any subsequent inspections,
  • Requirements for sampling and coverage,
  • Requirements for expansion of scope if unanticipated indications are found,
  • Inspection acceptance criteria,
  • Methods for evaluating examination results not meeting the acceptance criteria,
  • Updating the program based on industry-wide results; and
  • 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 Report No. 1101403 .401R1 20 Associates, mc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 21 of 121 communication. Based on the framework and strategy and on the accumulated industry research data, the following elements of an AMP were further developed [19, 20]:

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

Components for which the effects of the postulated aging mechanisms are insignificant, Components that are moderately susceptible to the aging effects, and Components that are significantly susceptible to the aging effects.

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

Aging management strategies were developed combining the results of 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.

The industry effort, as coordinated by the EPRI MRP, has produced the Inspection and Evaluation (I&E) Guidelines for reactor internals document (MRP-227, Rev. 0 [3a]) that was submitted to the NRC with a request for a formal Safety Evaluation Report (SER). A supporting document addressing inspection requirements (MRP-228 [22]) was also completed. The industry guidance is contained within these two separate EPRI MRP Documents:

  • MRP-227-A [3a] provides industry background for the guidelines, lists of reactor internals components requiring inspection, and the timing for initial inspections of those components. For each component, the guidelines require a specific type of non-destructive examination (NDE) and give 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 C-E). The document has been reviewed by the NRC and has received a SER [4].

Report No. 1101403.401R1 21

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 22 of 121 MRP-228 provides guidance on the qualification/demonstration of the required NDE techniques and other criteria pertaining to the actual performance of the inspections.

The PWR Owners Group (PWROG) has also developed and submitted for NRC review and approval, WCAP- 17096 (Reactor Internals Acceptance Criteria Methodology and Data Requirements) [14] for the MRP-227-A inspections, where feasible. Plant specific acceptance criteria can also be developed for some internals components if a generic approach is not practical.

PNP is not requesting any deviations from the guidance provided in MRP-227-A [3b], as approved by the NRC.

2.3 Reactor Vessel Internals Aging Management Program Attributes The attributes of the PNP RVI AMP and their compliance with the 10 elements of NUREG- 1801 (GALL Report),Section XI.M16A, PWR Vessel Internals [10] are essential for successful management of component aging are described in this section.

This AMP is consistent with the GALL process and includes consideration of the augmented inspections identified in MRP-227-A [3bJ. Specific details of the PNP RVI AMP are summarized in the following subsections.

2.3.1 NUREG-1801/AMP Program Element 1: Scope of Program The scope of the program includes all R VI components as the Palisades Nuclear Plant, which is built to a C-E NSSS design. The scope of the program applies to the methodology and guidance in the most recently NRC-endorsed version ofMRP-227-A, which provides augmented inspection andflaw evaluation methodologyfor assuring the functional integrity ofsafety-related internals in commercial operating US. PWR nuclear power plants designed by B& W, CE, and Westinghouse. The scope of components consideredfor inspection under MRP-227-A guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASME Code,Section XI), those R VI components that serve an intended license renewal safetyfunction pursuant to criteria in 10 CFR 54.4 (a) (1), and other RVI components whose failure could prevent satisfactory accomplishment ofany of the functions identified in 10 CFR 54.4(a) (1) (1), (ii), or (iii). The scope of the program does not include consumable items, such asfuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope ofthe components that are required to be subject to an aging management review (AMR), as defined by the acceptance criteria set in 10 CFR 54.21(a) (1). The scope of the program also does not include welded attachments to Report No. I 101403.401R1 22 SfruCLuraI hitegrity Associates, h7c.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 23 of 121 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 applicants AMP that corresponds to GALL AMP XI.M1, ASME Code,Section XI Inservice Inspections, Subsections IWB, IWC, and IWD.

The scope of the program includes the response bases to applicable license renewal applicant action items (LRAAIs) on the MRP-227-A methodology, and any additional programs, actions, or activities that are discussed in these LRAAI responses and creditedfor aging management of the applicants R VI components. The LRAAIs are identjfIed in the staffs safety evaluation on MRP-227-A and include applicable action items on meeting those assumptions thatformed the basis ofthe MRPs augmented inspection andflaw evaluation methodology (as discussed in Section 2.4 ofMRP227-A). The guidance in MRP-227-A spec/Ies 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 ofapplicability by the applicantfor each reactor and are covered in Section 2.4 ofMRP-227-A.

Program Scope A description of the PNP RVI design is provided in Section 3.0 of this AMP. The PNP RVIs that required aging management review are indicated in the PNP License Renewal Application

[15]. A summary of the results of the AMR is provided in Attachment B. This table identifies aging effects that require management (for those components requiring AMRs). A column in the table lists the programs and activities at PNP that are credited to address the aging effects for each component during the period of extended operation.

The results of the industry research provided by MRP-227-A provide the basis for the required augmented inspections, inspection techniques to permit detection and characterization of the aging effects of interest (cracks, loss of material, loss of preload, etc.), prescribed frequency of inspection, and examination acceptance criteria. The PNP RVI AMP scope is based on previously established and approved GALL Report approaches through application of approved methodologies to determine those components that require aging management. Likewise, the additional information provided in the industry guidelines document, MRP-227-A, is rooted in the GALL methodology and provides a basis for augmented inspections that were required to complete the PNP RVI AMP by providing the inspection method, frequency of inspection, and examination acceptance criteria. This program addresses aging management of the PNP vessel internals and identifies the components for augmented inspections under MRP-227-A as further clarified by the MRP-227 SER [4].

Report No. 1 101403.401R1 23 Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 24 of 121 Conclusion This element complies with the corresponding aging management attribute in Revision 2 of NUREG-1 801 [10], Section Xl.M1 6A and the list of commitments in the PNP License Renewal SER.

2.3.2 NUREG-1801/AMP Program Element 2: Preventive Actions The guidance in MRP-227-A relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss ofmaterial induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms [SCC, P WSCC; or IASCCJ). 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 AMP X1.M2, Water Chemistry.

Preventive Action The PNP RVI AMP includes the following existing program that complies with the requirement of this element. A description and applicability to the PNP RVI AMP is provided in the following subsection.

Primary Water Chemistry Program The primary goal of this program is to mitigate loss of material due to general, pitting, and crevice corrosion, cracking due to Stress Corrosion Cracking (SCC) by controlling the internal environment of systems and components. This program relies on monitoring and control of water chemistry to keep peak levels of various contaminants below the system-specific limits.

The PNP Primary Water Chemistry Program [12] is based on the current, approved revisions of EPRI PWR Primary Water Chemistry Guidelines.

This program is consistent with the corresponding program described in Revision 1 for GALL Report [21]. 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 [19].

Conclusion This element complies with the corresponding aging management attribute in Revision 2 of NUREG-1801 [10], Section Xl.M16A and the PNP License Renewal SER.

Report No. 1101403.401R1 24 SIructuraIIntegrItyASsocetes,h7C.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 25 of 121 2.3.3 NUREG-180 1/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 R VI components at the facility: (a) cracking induced by SCC, P WSCC, JASCC, orfatigue/cyclical loading, (b) loss of material induced by wear; (c) loss offracture toughness induced by either thermal aging or neutron irradiation embrittlement, (d) changes in dimension due to void swelling and irradiation growth, distortion, or deflection, and (e) loss ofpreload caused by thermal and irradiation-enhanced stress relaxation or creep.

Parameters Monitored/Inspected ENO 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 [3bj and ASME Section XI [13].

In accordance with the PNP license renewal SER, the ASME Section XI Program consists of periodic volumetric, surface andlor visual examination of components for assessment, signs of degradation, and corrective actions. This program is consistent with the program described in Revision 2 ofNUREG-1801 [10].

Conclusion This element complies with or exceeds the corresponding aging management attribute in NUREG-1801,Section XI.M16A and the PNP License Renewal SER.

2.3.4 NUREG-1801/AMP Program Element 4: Detection of Aging Effects:

The detection of aging effects is covered in two places: (a) the guidance in Section 4 ofMRP-227-A provides and introductory discussion andjustification of the examination methods selectedfor detecting the aging effects of interest; and (b) standardsfor examination methods, procedures, andpersonnel are provided in a companion document, MRP-228. In all cases, well established methods were selected. These methods include volumetric UT examination methodsfor detecting flaws in bolting, physical measurementsfor detecting changes is dimensions, and various visual (VT-3, VT-i, and EVT-i) examinationsfor detecting effects ranging from 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 ofsurface breaking discontinuities.

Report No. 1101403.401R1 25 Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 26 of 121 Cracking caused by SCC, L4SCC, andfatigue is monitored/inspected by either VT-i or EVT-i examination (for internals other than bolting) or by volumetric UT examination (bolting). The VT-3 visual methods may be appliedfor the detection of cracking only when the flaw tolerance of the component or affected assembly, as evaluatedfor reducedfracture toughness properties, is known and has been shown to be tolerant ofeasily detected largeflaws, even under reducedfracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspectfor loss of material induced by wear andfor general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects ofloss ofpreload caused by thermal and irradiation-enhanced stress relaxation and creep.

In addition, the program adopts the recommended guidance in MRP-227-A for defining the Expansion criteria that need to be applied to inspections ofPrimary Components and Existing Requirement Components andfor expanding the examinations to include additional Expansion Components. As a result, inspections performed on the R VI components are performed consistent with the inspection frequency and sampling basesfor Primary Components, Existing Requirement Components, and Expansion Components in MRP-227-A, 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.

Detection of Aging Effects Detection of indications required by the ASME Section XI ISI Program is well-established and field-proven through application of the Section XI ISI Program. Those augmented inspections that are taken from the MRP-227-A recommendations would be applied through use of the MRP-228 [22] Inspection Standard.

Inspection can be used to detect physical effects of degradation including cracking, fracture, wear, and distortion. The choice of an inspection technique depends on the nature and extent of the expected damage. The recommendations supporting aging management of the RVI, as contained in this program, are built around three basic inspection techniques: (1) visual, (2) ultrasonic, and (3) physical measurement. The visual techniques include VT-3, VT-i, and EVT- 1. Inspection standards developed by the industry for the application of these techniques in augmented RVI inspections are documented in MRP-228 [22]. 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.

Report No. 1 101403.401R1 26 Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 27 of 121 Conclusion This element complies with or exceeds the corresponding aging management attribute in Revision 2 in NUREG-1801 [10], Section XLM16 and the PNP License Renewal SER.

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

The methodsfor monitoring, recording, evaluating, and trending the data that result from the program s inspections are given in Section 6 ofMRP-227 and its subsections. The evaluation methods include recommendationsfor flaw depth sizing andfor crack growth determinations as wellfor performing applicable limit load, linear elastic and elastic-plasticfracture analyses ofrelevantflaw indications. The examinations and re-examinations required by the MRP-22 7 guidance, together with the requirements specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects ofthe age-related degradation mechanisms within the scope of the program. The extent of the examinations, beginning with the sample ofsusceptible PWR internals component locations identified as Primary Component locations, with the potentialfor inclusion of Expansion Component locations f 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 examinationsfor core support structures, provides a high degree of confidence in the total program.

Monitoring and Trending 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. ENO maintains cognizance of industry activities and shares operating experience information related to PWR internals inspection and aging management.

Inspections credited as part of the existing programs are based on utilizing the PNP 10-year 151 program and the augmented inspections derived from the industry programs. These inspections, where practical, are scheduled to be conducted in conjunction with typical 10-year ISI examinations.

Inspections performed at PNP as part of the 1ST program are provided in Attachment A. Tables 2 and 3 identify the inspection requirements for Primary and Expansion category components credited for aging management of RVI. As discussed in MRP-227-A [3b], the sampling Report No. 1 101403.401R1 27 SJIuCLutaI hitegrit, Associates, h7c?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 28 of 121 inspections of the Primary components, with the potential for expanding the sampling program if unexpected effects are found, provides reasonable assurance for demonstrating the ability of the reactor vessel internal components to perform the intended functions.

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

Conclusion This element complies with or exceeds the corresponding aging management attribute in Revision 2 ofNUREG-1801 [10],Section XI.M16A and the PNP License Renewal SER.

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

Section 5 ofMRP-227 provides specfIc examination acceptance criteriafor the Primary and Expansion Component examinations. For components addressed by examinations referenced to ASME Code,Section XI, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the examination acceptance criteria are described within the Existing Program reference document.

The guidance provided in MRP-227 contains three types ofexamination 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 specflc, descrztive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sizedfor length by VT-i/E VT-i examinations;
  • For volumetric examination, the examination acceptance criterion is the capabilityfor reliable detection of indications in bolting, as demonstrated in the examination Technical Justflcation; in addition, there are requiements for system-level assessmentfor bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specJIed limits; and
  • Physical measurements: Per Revision 2 ofNUREG-]801, this is not applicable for C-E designedplants.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 29 of 121 Acceptance Criteria Recordable indications that are the result of inspections required by the existing PNP 1ST program [11] are evaluated in accordance with the applicable requirements of the ASME Code through the ENO Corrective Action Program.

Inspection acceptance and expansion criteria are provided in Table 5. These criteria would be reviewed periodically as the industry continues to develop and refine the information and would be included as part of updates to the PNP RVI AMP. Updates would be based on the availability of state-of-the-art information and techniques.

Augmented inspections, as defined by the MRP-227-A requirements that result in recordable relevant conditions, would be entered into the plant Corrective Action Program and addressed by appropriate actions that may include enhanced inspection, repair, replacement, mitigation, or analytical evaluations. Additional analysis to establish acceptable evaluation criteria for components is also considered in determining the acceptance of the inspection results to support continued component or assembly functionality. Industry groups are working to develop a consistent set of tools inline with approved methodologies to support this element. Additional analysis to establish evaluation acceptance criteria for expansion category components has been developed by the PWROG (WCAP-17096-NP). The status of these ongoing processes is monitored via participation in various industry programs related to aging management of PWR internals.

Conclusion This element complies with or exceeds the corresponding aging management attribute in Revision 2 ofNUREG-1801 [10],Section XI.M16A and the PNP License Renewal SER.

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

Corrective actions following the detection of unacceptable conditions are fundamentally providedfor in each plants 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 evaluationfor continued service until the next inspection.

The disposition will ensure that design basis functions of the reactor internals components will continue to befulJIlledfor all licensing basis loads and events.

Examples ofmethodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code,Section XI or in Section 6 of MRF-22 7.

Section 6 ofMRP-22 7 describes the options that are available for disposition of detected conditions that exceed the examination acceptance criteria ofSection 5 of Report No. 1 101403.401R1 29 1 meL Associates

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 30 of 121 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 Xl. The implementation of the guidance in MRP-227, plus the implementation ofany ASME Code requirements, provides an acceptable level ofaging management ofsafety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Corrective Actions The existing ENO procedure for Corrective Actions Program [23] is credited to address this element of the GALL attributes. Indications that require repair and replacement would be addressed through the ENO Corrective Action Program. The repair and replacement activities would be performed in accordance with methodologies provided in Section 6 of MRP-227-A

[3b] and ASME Code Section XI [13].The corrective actions for existing Sections XI (B-N-3) examinations would include the identification of a repair plan and verification of acceptability of replacements. The corrective actions for augmented inspections and any associated relevant indications would be developed using evaluation methods or repair and replacement procedures in accordance with or equivalent to the requirements in ASME Code Section XI. Actions to evaluate and monitor flaws/indications would be a part of the corrective action process. This evaluation guidance is included in MRP-227-A and WCAP-17096-NP. For example, the guidance provided in WCAP- 17096-NP maybe used to evaluate component degradation that exceeds acceptance criteria in Section 5 of MRP-227-A when it is observed during required inspections. Other methods may also be used if approved by NRC.

Conclusion This element complies with the corresponding aging management attribute in Revision 2 of NUREG-1 801 [10],Section XI.M16A and the PNP License Renewal SER.

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

Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR Part 50, Appendix B, or equivalent, as applicable. It is expected that the implementation of the guidance in MRP-227-A will provide an acceptable level of qualityfor inspection, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with 10 CFR Part 50, Appendix B or equivalent, as applicable, confirmation process, and administrative controls.

Report No. 1101403.401R1 30

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 31 of 121 Confirmation Process The PNP RVI AMP meets the Mandatory and Needed requirements under NET 03-08. This program conforms to the NEI 03-08 Materials Initiative Process for ENO plants [1], and it ensures that deviations, self-assessments and benchmarks are conducted as necessary to support the NET Materials Initiative. The PalisadesSection XI Inspection Program and the ENO Corrective Action Process meet the requirements for QA programs. In particular, all QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B [24].

Conclusion This element complies with or exceeds the corresponding aging management attribute in Revision 2 ofNUREG-1801 [10J, SectionXl.M16A and the PNP License Renewal SER.

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

The administrative controlsfor such programs, including their implementing procedures and review and approval processes, are under the existing site 10 CFR 50 Appendix B Quality Assurance Programs, or their equivalent, as applicable. Such a program is thus expected to be established with a sufficient level of documentation and administrative controls to ensure effective long-term implementation.

Administrative Controls ENO QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50, Appendix B. This element complies with or exceeds the corresponding aging management attribute in Revision 2 of NUREG- 1801 [10], SectionXl .M 1 6A and the PNP License Renewal SER.

2.3.10 NUREG-1801/A11P Program Element 10: Operating Experience:

Relativelyfew incidents ofPWR internals aging degradation have been reported in operating US. commercial PWR plants. A summary ofobservations to date is provided in Appendix A ofMRP-227-A. The applicant is expected to review subsequent operating experience for impact on its program or to participate in industry initiatives that perform thisfunction.

The application ofMRP-227 guidance will establish a considerable amount of operating experience over the nextfew years. Section 7 ofMRP-22 7 describes the reporting requirementsfor these applications, and the plan for evaluating the accumulated additional operating experience.

ReportNo. 1101403.401R1 31 1711 aaIIiuIeur1t, Assowates,h7c

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 32 of 121 Operating Experience Extensive industry and PNP operating experience (OE) has been reviewed during the development of the PNP RVI AMP.

Industry and PNP specific information relevant to aging has been compiled into a PNP OE database [251. Industry operating experience sources in the database 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 (LER5), Corrective Action Process documents (CAPs, CRs, DR, ERs), etc.

A review of the industry Operating Experience (OE) revealed several instances of cracked bolts or loose bolts in reactor vessel internals. Due to the differences in design, none of these experiences are applicable to the PNP RVI [26].

Although there were no PNP specific operating experiences in the OE database, PNP FSAR indicates wear was found at several RVI locations, including core barrel keys and core barrel ledge in the reactor vessel flange due to insufficient hold-down force of original design. It was discovered during a 1973 outage inspection as part of an investigation of core neutron flux oscillations observed on the excore detectors. The insufficient force allowed the core barrel and upper guide structure to move, which allowed wear to occur and further reduced the force, which compounded the wear. The original design of the compensating ring shim was replaced with a hold-down device that provides a much greater hold-down force [27]. The NRC staff reviewed the improvements made to the hold-down devices and agreed that improved hold-down device will reduce the probability of flow-induced vibrations that could compromise the integrity of the core internals during future operations. The initiation of the surveillance program with limiting conditions was found to be similar to that being incorporated in new operating licenses and that it is adequate to provide early detection in the unlikely event of an unacceptable degree of flow-induced vibration [27]. The core barrel keys were also replaced to ensure proper vessel internals alignment.

The review of operating experiences did not identify any new aging issues related to the PNP RVI [28].

Inspections performed as part of the 10-year ISI program have been conducted as designed by existing commitments and would be expected to discover general internals structure degradation.

Industry operating experience is routinely reviewed by Institute of Nuclear Power Operations (INPO) and other informational sources, as directed under the applicable procedure for the determination of additional actions and lessons learned. These insights, as applicable, can be Report No. 1 101403.401R1 32 Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 33 of 121 incorporated into the PNP System and Program health reports and further evaluated for incorporation into plant programs. Any potential indications can be entered into the Corrective Action Program.

A review of industry and plant-specific experience with RVI reveals that the U.S. Nuclear fleet (including PNP) has responded proactively to issues related to degradation of RVI by participation in PWROG, MRP, etc.

Conclusion This element complies with or exceeds the corresponding aging management attribute in Revision 2 ofNUREG-1801 [10], SectionXl.M16A and the PNP License Renewal SER.

The specifics of the PNP RVI design are described in Section 3.0 Report No. 1101403.401R1 33 1 Inc.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 34 of 121 3.0 Palisades Reactor Vessel Internals Design and Operating Experience The PNP RPV and vessel internals were designed by Combustion Engineering (C-E). The components of the reactor vessel internals are divided into smaller sub-assemblies consisting of the upper guide structure, control rod shroud assemblies, core support barrel assembly, core shroud assembly, lower internals assembly, and in-core instrumentation system. The arrangement of a typical C-E design vessel and internals package is shown in Figure 1. The PNP design Code is the 1965 version of the ASME Section III with Addenda through Winter of 1965 [18]. The PNP RVI were designed before the incorporation of the 1974 addition of Subsection NG into ASME Section III, and therefore the internals structures do not have an ASME Stress Report or calculated fatigue usage factors.

The C-E designed PWR internals consist of three major structural assemblies: an upper internals assembly (also known as an upper guide structure [UGS] at PNP) that is removed during refueling as a single component to provide access to the fuel assemblies, a core support barrel, and a lower support structure. In addition, there are three other RVI assemblies in C-E designed plants: the control element assemblies (CEA) shroud assembly, the core shroud assembly, and the in-core instrumentation support system. Since PNP has a unique design that is different from other C-E designed plants, some of the components have different nomenclature. One of these is the upper guide structure, which is similar to the upper internals assembly in other C-E designed plants. Additionally, PNP has a control rod shroud assembly instead of a CEA shroud assembly and an in-core instrumentation guide system rather than an in-core instrumentation support system. These components have some differences in design, but they serve similar functions. It should be noted that the MRP-227 augmented inspection requirements are still applicable for these components. A description of the design characteristics for these internals specific to PNP is provided in the following subsections. The general arrangement of the C-E designed PWR internals is shown in Figure 1. Schematic representations of other reactor vessel internals components are provided in Figures 2 through 13.

3.1 Upper Guide Structure The upper guide structure (UGS) is located above the reactor core, within the core support barrel assembly, and is removed during refueling as a single component in order to provide acces to the fuel assemblies. The UGS shown in Figure 7 is an integral assembly that includes the upper fuel assembly alignment plate, the control rod shroud assemblies, the in-core instrumentation guide systems, and the hold-down assembly. At PNP, the hold-down ring present in other C-E designs has been replaced by a system of Belleville springs. The functions of the UGS are to provide alignment and support to the fuel assemblies, to prevent movement of the fuel assemblies in the case of a severe accident condition, and to protect the control rods from cross-flow effects in the upper plenum. The flange on the upper end of the UGS rests on the core support barrel.

Report No. 1101403.401R1 34 Sfr7Jc1urW Integrity Associates, mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 35 of 121 3.2 Core Support Barrel The core support barrel assembly consists of the core support barrel, the core support barrel upper flange, core support barrel alignment keys, and the core support barrel snubbers. The core support barrel shown in Figure 8 is a cylinder which contains the core and other internals. Its function is to resist static loads from the fuel assemblies and other internals, and dynamic loads from normal operating hydraulic flaw, seismic events, and loss-of-coolant-accident (LOCA) events. The core support barrel also supports the lower internals assembly and its core support plate, upon which the fuel assemblies rest. The core support barrel upper flange is a thick ring that supports and suspends the core support barrel from a ledge on the reactor vessel.

3.3 Lower Support Structure The lower support structure shown in Figures 9 and 10 consists of the core support plate, the core support columns and the lower support structure beam assemblies. The core support plate functions are to position and support the reactor core, and to provide control of reactor coolant flow into each fuel assembly. The weight of the core is transmitted to the core support barrel through the lower support structure via the core support columns (Figures 10 and 11). Fuel alignment holes in the core support plate engage lower fuel assembly alignment pins to provide guidance and limit lateral movement of the individual fuel assemblies.

3.4 Core Shroud Assembly The core shroud assembly shown in Figures 3 and 9 is located within the core support barrel and directly below the UGS. PNP has a bolted core shroud assembly where the core shroud plates are fastened to the former plates with structural bolts, as shown in Figure 13. The former plates, and thus the assembly, are attached to the core support barrel by structural bolts as well. The core shroud assembly provides a boundary between the reactor coolant bypass flow on the inside of the core support barrel and the reactor coolant flow through the fuel assemblies. It also limits the amount of coolant bypass flow and reduces the lateral motion of the fuel assemblies.

3.5 Control Rod Shroud Assemblies The control rod shroud assemblies shown in Figures 2 and 12 consist of control rod shrouds, the control rod shroud bolts, and the extension shaft guide tubes. The control rod shrouds protect the cruciform control rods from cross-flow effects in the upper plenum. The extension shaft guides also protect the control rods from cross-flow effects in the upper plenum and provide lateral support and alignment of the control rod extension shafts during refueling operations. The control rod drive mechanisms are positioned on the reactor vessel closure head and are coupled to the control rods by the control rod extension shafts. The control rod shroud assemblies were Report No. 1101403.401R1 35 SfruCflraIIfltegrItyAssocIates,InC.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 36 of 121 considered in the development of this AMP. The control rod drive mechanisms, extension shafts, and control rods were not included in the list of internals components.

3.6 In-Core Instrumentation Support System The in-core instrumentation support system consists of in-core instrumentation guide tubes and components which provide support to the in-core instrumentation as shown in Figure 12. For plants with top-entry in-core instrumentation assemblies, such as PNP, the in-core instrumentation is inserted through the reactor vessel head via a nozzle and into a guide tube. The guide tubes interface with the upper fuel alignment plate to align with holes in the fuel assemblies. ICI elements are inserted from the guide tubes directly into the fuel assemblies.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 37 of 121 instrumentation Nozzle ipper Structure Figure 1. Combustion Engineering Vessel and Internals Arrangement ReportNo. 1101403.401R1 37 S1njctural Integrity Associates, Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 38 of 121 ContioL Ro1 ttv

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 41 of 121 Hah-Fhzece Se3rn Figure 5. High-Fluence Seam Locations in Baffle-Former Assembly Report No. 1101403.401R1 41 SfructuraI hitegrity Associates, h70.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 42 of 121 1

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 43 of 121

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 44 of 121 I-lange VVeIcl Axial Weld Upper Core Barrel to Lower Core Barrel Circumferential Weld Loer Barrel Axial Weld Lower Barrel Circumferential Weld Lower Barrel Axial Weld Core Darrel to Support Plate Weld Figure 8. Typical C-E Core Support Barrel Structure ReportNo. 1101403.401R1 44 5SbucturaI Integrity Associates mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 45 of 121

/%j Flange V eld Upper Core Barrel Flange Core Banel Core Shroud Assembly Core Support Plate Lower Support Structure Figure 9. Palisades Core Support Barrel, Core Shroud Assembly, and Lower Support Structure Report No. 1101403.401R1 45 SkucturaI Ililegrity Associates, mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 46 of 121 Core Barrel Core Shroud Assembly Core Support Core Plate Support Colunms Support Beams Figure 10. Palisades Lower Core Support Structure Assembly Report No. 1101403.401R1 46 SIrijcturaIIntegMty Associates, Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERI1ALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 47 of 121 Figure 11. Palisades Core Support Column Report No. 1101403.401R1 47 Sb7JcJur8IIutegrIty Associates, Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 48 of 121 CONTROL SHROUD ulo AS$4SLY FUEL IWIDLE

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 50 of 121 3.7 C-E Design Plants/Non-Relevant Operating Experience No design modifications beyond those identified in general industry guidance or recommended by the original vendors have been made for the PNP RVI. Other components with a history of degradation in PWR Reactor Vessel Internals include:

  • Thermal shield fasteners and thermal shields have experienced fatigue and wear resulting from flow-induced vibrations. SCC has also affected thermal shields. The thermal shield was removed from the PNP reactor vessel during early years of operation [29].
  • Incore instrumentation tubes (thimble tubes) have experienced fatigue and wear damage caused by flow-induced vibrations. PNP uses top-mounted instrument guide tubes, which are not susceptible to these aging mechanisms [29].
  • Hold-down springs have become permanently deformed, resulting in a Loss of spring force, in some cases causing damage to the vessel flange and mating internals components. PNPs new hold-down device does not use Hold-down Springs, and is not susceptible to this aging mechanism [29]. Surveillance after plant modification has shown that vibration fatigue does not occur in the core barrel.
  • PNP is one of two C-E plants without susceptibility to cracking of the baffle-to-former plate (i.e., core shroud) bolts. In response to NRCs response to request for additional information (RAT No. 14) dated August 27, 2005 [26], PNP clarified that it is one of only two C-E designed plants that uses bolts to attach the core shroud panels (i.e., the baffle or core shroud plates) to the former plates. It was determined that these bolts are less susceptible to IASCC because: (1) the material used in these bolts is annealed Type 316 stainless steel, which is not cold worked; (2) the bolt stress from preload, as percentage of yield strength, is much less than that of the susceptible plants; (3) the differential pressure across the core shroud panels does not result in tensile loads on the panel (i.e., the baffle or core shroud); and (4) the core shroud panel design allows for some flexing of the former plate relative to the core barrel, thus reducing the load on the panel (i.e., core shroud) bolts.

3.8 Changes in Plant Operation ENO Operations, Inc. submitted a request to the US NRC, a request for changes to the PNP, Operating License and Technical Specifications (TSs). The requested changes were pertaining to an increase in the power level from 2530 Megawatts thermal (MWt) to 2565.4 MWt, an approximately 1.4% increase. This increase was justified based on reduced core power measurement uncertainty resulting from the use of more accurate feedwater flow measurement Report No. 1101403.401R1 50 SJflJcáuraI h$grIty Associates, Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 51 of 121 instrumentation. In June 2000, the Nuclear Regulatory Commission approved a change to 10 CFR 50 Appendix K, providing licensees the option of maintaining the 2% power margin between the licensed core power level and the assumed core power level for emergency core cooling system (ECCS) evaluations, or applying a reduced margin to the ECCS evaluations. The Crossflow ultrasonic flow measurement system has been in use at PNP for feedwater measurement since 1997. The core power measurement uncertainty using Cross flow was determined to be less than 0.59%. Therefore, it was proposed to reduce the power measurement uncertainty required by 10 CFR 50, Appendix K, from 2% to 0.5925% to permit an increase in the licensed power level by 1.4% to 2565.4 MWt [30].

In a letter dated July 15, 2004, the Staff approved the amendment changes to the PNP operating license and TS associated with the increase in the power level from 2530 MWt to 2565.4 MWt

[31].

3.9 Conformance with MRP-227-A Assumptions Operations at the PNP conform to the assumptions in Section 2.4 of MRP-227 [3b].

  • PNP operated for 18 effective full power years (EFPY) with high-leakage core patterns, followed by implementation (Cycle 8) of a low-leakage fuel management strategy for the remaining years of operation [32];
  • PNP operates as a base load unit, and
  • No design changes were implemented beyond those identified in general industry guidance or recommended by the vendor (C-E or Westinghouse)

The MRP-227-A methodology used to developing the I&E guidelines were based on three important precursor elements, namely, screening criteria, categorization, and, functionality assessments. Categorization of PWR internals, based on screening criteria and the likelihood and severity of safety consequences, into categories that range from those components for which these issues were insignificant (Category A) to those components that are potentially moderately significant (Category B) to those components that are potentially significantly accepted (Category C). Functionality assessment of components and assemblies of components based on representative plant designs using irradiated and aged material properties to determine the effects of the degradation mechanisms on functionality. Similar assessments were performed for the unique PNP RVI design. Plant specific screening and categorization was performed for the PNP RVI by comparing the components required to be managed as a commitment in the license renewal application to those evaluations performed in MRP-191 [20] and MRP-227-A [3b] as described in Section 4.4.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 52 of 121 4.0 Program Description Management of component aging effects includes actions to prevent or control degradation due to aging effects, review of operational experience to better understand the potential for degradation to occur, inspections to detect the onset of aging effects in susceptible components, protocols for evaluation and remediation of degradation due to aging, and procedures to ensure component aging is managed in a coordinated program.

4.1 Preventive Actions ENO is currently managing water chemistry to mitigate SCC initiation in nickel alloys. This is addressed by the ENO Water Chemistry Program [12].

4.2 Operational Experience Operational experience (OE) related to degradation of reactor internal components covered in this aging management document would be reviewed on a periodic basis. Worldwide operation experience through 2009 is summarized in Reference 33. Results of reactor internal components inspected in accordance with MRP-227-A would be summarized in the biannual MRP Inspection Data Survey, MRP-219 [34]. These reports would serve to assist in review of operating experience and required monitoring and trending for the RV internals aging management program.

4.3 Component Inspection and Evaluation Overview A description of Aging Management Document categorization and the steps used to develop this program document are given below.

This program summarizes the guidance of the MRP I&E guidelines necessary to understand implementation but does not duplicate the full discussion of the technical bases. MRP-227-A

[3b] and its supporting documents should be consulted for a complete description of the technical bases of the program.

MRP-227-A [3b] establishes four groups of reactor internals components with respect to inspection requirements: Primary, Expansion, Existing Programs, and No Additional Measures, as summarized below.

  • Primary: Those PWR internals components 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 are described in the I&E Guidelines and are needed to ensure functionality of Primary components. The Primary group also includes components which have low or moderate susceptibility to a specific aging degradation effect, but for ReportNo. 1101403.401R1 52

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 53 of 121 which no highly susceptible component exists or for which no highly susceptible component is accessible.

  • Expansion: Those PWR internals components that are highly or moderately susceptible to the effects of at least one of the eight aging mechanisms, but for which functionality assessment has shown a degree of tolerance to those effects, were placed in the Expansion group. The schedule for implementation of aging management requirements for Expansion components will depend on the findings from the examinations of the Primary components at individual plants.
  • Existing Programs: Those PWR internals components that are susceptible to the effects of at least one of the eight aging mechanisms and for which existing program elements are capable of managing those effects, were placed in the Existing Programs group.
  • No Additional Measures: Those PWR internals components for which the effects of all eight aging mechanisms are below the screening criteria, and which were placed in Category A by the initial screening step were placed in the No Additional Measures group. Through the functionality assessment process, some of the PWR internals components other than Category A components were also placed in No Additional Measures. No further action is required for managing the aging of the No Additional Measures components, other than the continuation of any existing plant requirements that apply to these components. Many of the No Additional Measures components are not core support structures, and therefore may not be covered by a program element such as the ASME B&PV Code, or Section XI periodic in-service examination [13].

The inspections required for Primary and Expansion components were selected from existing, visual, surface, and volumetric examination methodologies that are applicable and appropriate for the expected degradation effect (e.g., cracking caused by particular mechanisms, loss of material caused by wear). The inspection methodologies include: Visual (VT-3) examinations, Visual (VT-i) examinations, surface examinations, volumetric (specifically, UT) examinations, and physical measurements. MRP-227-A provides detailed justification for the components selected for inspection and the specific examination methodologies selected for each. The MRP 228 report, PWR Internals Inspection Standards [22], provides detailed examination requirements for the components listed. The PNP License Renewal SER [16] identified components requiring aging management review. AMRs were performed for each of the PNP RVI components in accordance with the screening and categorization process as described in Section 4.4.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 54 of 121 4.4 Examinations of Reactor Vessel Internals The PNP RVI components to be examined under the current Section XI inspection program are given in Attachment A [11]. Additional components and locations are to be examined in accordance with the current license renewal commitments and the industry guidance document, MRP-227-A [3b]. Initial susceptibility and categorization of reactor vessel internals components was considered by the methods in MRP-191 [20]. These methods formed the basis for the screening and categorization, in combination with the functionality analyses of MRP-230

[35] and the judgments reached in MRP-232 [36], and led to the final categorization in MRP 227-A. Some component designs are unique to PNP, and for this reason additional evaluations have been performed to evaluate these components, including the list of RVI components [28]

for aging management as described in the PNP LRA.

The evaluation for the LRA components is described below and the results are shown in Attachment B. This table (Attachment B) contains all of the components that were required to be managed as a commitment in the license renewal application. A number of the components are already covered under the current Section XI inspection program shown in Attachment A.

However, there are other components which do not currently receive any additional inspections.

In order to determine if these components required additional inspections, the component list was compared [37] to the evaluations performed in MRP-191 [20] and MRP-227-A {3bJ.

The components presented in Attachment B were compared against the components for C-E plants in MRP-191 [20]. The corresponding screening code given in MRP-191 was listed for each component if it was specifically listed in MRP- 191. MRP classified the components in three categories: A, B, or C. Category A components do not meet the screening criteria and have low safety significance. Category C components are leading items and have high safety significance. Category B components are not leading items and may have some safety significance, but can be re-classified as A due to other required inspections. Therefore, all A components listed in Attachment B do not require any additional inspections. The Category B and C components listed in table presented in Attachment B were addressed by comparing the components to the MRP-227-A list of Primary and Expansion components. The components listed in MRP-227-A were noted in the Attachment B with the associated inspection requirements. The table in Attachment B also contains additional components that were not listed in MRP- 191 or MRP-227 due to PNP specific plant designs that differ from the other C-E plants. These components required screening and judgment to determine if any additional inspections were required.

In order to determine if the additional components required any additional inspections, the components were compared to the other C-E designed plants to determine similarities. PNP has Report No. 1 101403.401R1 54 Sb7JC1u18I Integrity Associates, Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 55 of 121 control rod shrouds in the upper guide structure where other C-E plants have CEA shrouds.

MRP- 191 only screens the CEA shroud configurations to determine if additional exams are required. The control rod shroud components were included in the license renewal application as being susceptible to stress corrosion cracking (in welds), void swelling (changes in dimensions), and irradiation-induced stress relaxation (loss of preload). MRP-l 91 classified all the components susceptible to stress corrosion cracking as Category A for the CEA shrouds.

Since the control rod shroud assemblies are above the core they are not expected to see any effect due to void swelling or stress relaxation. Therefore, these components would fall into category A and not require any augmented examinations.

For the core shroud assembly and core support barrel assembly, the components are already listed in MRP-191 or MRP-227-A. Therefore, the inspection requirements are already defined for these components and no additional screening or justification was necessary.

The incore instrument (ICI) guide tube assemblies were not specifically listed in MRP-191 or MRP-227-A. These components were identified as being susceptible to loss of preload (irradiation-induced stress relaxation), changes in dimensions (void swelling), cracking (stress corrosion cracking), loss of material (wear), and reduction in fracture toughness (neutron irradiation embrittlement). The FMECA evaluation contained in MRP- 191 found every item in the ICI assembly to be classified as A. Therefore, there would be no additional examinations required for these components. MRP-227 specifies that certain components are not included in the lists as they are subject to plant specific designs. Two components are listed specifically in Section 4.4.2 of MRP-227-A [3b] are the zircaloy ICI thimble tubes and the thermal shield positioning pins. These are not an issue for PNP as neither ICI thimble tubes nor a thermal shield are present in the plant.

The lower internals assembly components are mostly listed in MRP-191 and MRP-227-A.

However, the core support beams include tie rods which are not explicitly listed in MRP- 191.

They are subject to changes in dimensions, cracking, and reduction in fracture toughness. This location is in a low fluence area and hence is rated as category A. Thus, no augmented examination would be required. Based on the AMR review of PNP reactor vessel internals components [28, 29] it was determined that no CASS or precipitation hardened materials are part of the core support structures.

For the upper internals and guide structure many of the components were listed in MRP-191 or MRP-227-A. However there were a few components that required additional evaluation. For example, PNP has a spacer shim, which was installed due to excessive wear noticed in the upper flange area of the core barrel. Because of the prior operating experience and the structure Report No. 1101403.401R1 55 Sb7JCturaI Integrity Associates, mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 56 of 121 functional support of the spacer shim, future component examination would ensure that no additional wear is present. For the other components in the PNP design subject to changes in dimensions, cracking, and loss of material, such as the control rod shroud plates, shroud top support, shroud support lug, control rod support lug, fuel guide pin nuts, and fuel plate cap screws, these are very similar to components on the CEA shroud, they are not expected to see significant fluence, do not provide structure functional support, and were classified as category A. Thus, no augmented examinations would be required. The remaining components are removable and subject only to wear. These components would be removed to get to the core barrel and evaluated as part of disassembly and reassembly.

Based on the evaluations of the LRA components and their comparisons and similarities with the components listed in MRP- 191 and MRP-227-A, the components requiring augmented inspections were determined. The components highlighted green in the Table (Attachment B) do not require any additional augmented inspections. The other components not highlighted in Attachment B require additional inspections or additional evaluations. No additional components were identified as a result of further AMR of PNP RVI components.

The list of primary components to be evaluated for PNP in accordance with MRP-227-A is given in Table 2. The list of expansion category components is given in Table 3. The list of existing program components from MRP-227-A is given in Table 4.

4.5 Inspection and Evaluation Requirements for Primary Components The inspection requirements for Primary Components of C-E designed plants applicable to PNP from MRP-227-A [3b] are provided in Table 2.

4.6 Inspection and Evaluation Requirements for Expansion Components The inspection requirements for Expansion Components of C-E designed plants applicable to PNP from MRP-227-A [3b] are provided in Table 3.

4.7 Inspection of Existing Plant Components The list of Existing Plant Components of C-E designed plants applicable to PNP from MRP-227-A [3bJ are provided in Table 4. This list of components in the current Section XI ISI program for PNP designated as B-N-2 and B-N-3 locations are shown in Attachment A [13]. The current ISI program considers existing inspections are implemented for each inspection interval [11].

Table 7 contains the planned implementation schedule for supplemental inspections in Report No. I 101403.401R1 56 mc?

PALISADES NUCLEAR PLANT CEP.-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 57 of 121 accordance with the requirements of MRP-227-A. These additional examinations, the methods to be used, and the acceptance and expansion criteria are described below.

The ASME Section XI ISI Inspections and additional augmented inspections [11] identified in this Reactor Vessel Internals (RVI) Aging Management Document would be performed in accordance with the required inspection interval.

4.8 Examination Systems (per MRP-228)

Equipment, techniques, procedures and personnel used to perform examinations required under this program shall be consistent with the requirements of MRP-228 Section 7.2 [22]. Indications detected by visual and ultrasonic examinations shall be measured and classified as necessary, and determined to be either relevant or non-relevant and reported in accordance with the applicable requirements of MRP-228. Examination results that do not meet the examination acceptance criteria (defined in MRP-227-A, Section 5) shall be recorded and entered into the ENO Corrective Action Program and dispositioned. This is Needed requirement 7.5 under MRP-227-A [3bJ.

4.9 List of Vessel Internals Components for Examination The schedule for performing the MRP-227-A augmented inspections is presented in Table 7.

This list contains the components for both existing and augmented examination. This list contains the components which have been elevated to primary or expansion status based on the NRC SE for MRP-227-A [4].

4.10 Inspection Schedule The schedule for the PNP RVI augmented primary component inspections is provided in Table

7. The program enhancement and implementation schedule provided in Table 7 is based on the requirements of MRP-227-A for an 18-month refueling cycle and considers cumulative operation. This information includes a description of the currently projected scope of inspection pertaining to the reactor internals AMP. Should a change occur in plant operational practices or operating experience result in changes to the projections, appropriate updates would be performed in accordance with approved procedures.

4.11 Description of Existing Aging Management Programs The overall strategy for managing the effects of aging in the reactor vessel internals components at PNP is supported by the following existing programs/activities:

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  • Water Chemistry Program [12] as described in Reference [1]
  • Industry Programs for Managing Aging of Internals These are established programs/activities that support the aging management of PCS components in addition to the RVI components. Although affiliated with and supporting the aging management of reactor vessel internals, these programs would continue to be managed under the program management structure for ENO.

4.11.1 ASME Section XI Inservice Inspection Program of Vessel Internals The ASME Section XI [13] Inservice Inspection Program is an existing program that facilitates inspections to identify and correct degradation in Class 1, 2 and 3 piping components, their supports and integral attachments, including welds, pump casings, valve bodies, pressure retaining bolting, piping/component supports and reactor head closure studs. These are identified in ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, [13] or commitments requiring augmented inservice inspections. This program is in accordance with 10CFR5O.55a [38]. The original Section XI inspection plan was based on knowledge at the time of original license and an expected service life of 40-years. MRP-227-A

[3b] is designed to supplement original inspection requirements to address aging beyond the original design life.

The categories that apply to the vessel internals include: (1) the interior attachments beyond the beltline (B-N-2) and (2) core support structures (B-N-3). The core support structures would be removed from the reactor vessel for examination during the vessel IS! examination.

4.11.2 Water Chemistry Program The water chemistry program is credited for managing aging effects by controlling the environment to which internal surfaces of systems and components are exposed. Such effects include the following:

  • Loss of material due to general, pitting and crevice corrosion,
  • Cracking due to SCC,

The aging effects are minimized by controlling the chemical species that cause the underlying mechanisms that produce them. The water chemistry program provides assurance that an Report No. 1101403.401R1 58 hic?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 59 of 121 elevated level of contaminants and, where applicable, oxygen does not exist in the system and components covered by the program, thus minimizing the occurrence of aging effects, and maintaining each components ability to perform the intended functions.

PNP has installed zinc injection as part of the PCS water chemistry program in order to achieve dose rate reduction [39, 40]. This is monitored by the ENO Water Chemistry Program [12] in accordance to the EPRI PWR Primary Water Chemistry Guidelines [19].

4.11.3 Industry Programs ENO actively participates in the EPRI Materials Reliability Program and the PWR Owners Group that provides information on specific issues related to degradation of C-E designed reactor vessel internals.

4.12 Revised Commitments A new commitment not documented in Appendix A to the Safety Evaluation Report (SER) has been initiated by ENO Nuclear Operations (ENO) related to the program plan for aging management of the reactor vessel internals [41]. ENO submitted WCAP-17133-NP [5] to the NRC on March 10, 2010. Tn an electronic letter dated September 17, 2010, a request for additional information (RAT) was issued by the NRC. The RAT noted that the aging management report submitted by the licensee used MRP-227, Rev. 0 [3a] as a technical basis for developing the aging management program which was under NRC review. As part of the RAT resolution, ENO was to review the SER for MRP-227, Rev. 0 [4], modify its aging management program accordingly, and re-submit the program plan within a year of issuance of the final NRC SER for MRP-227, Rev. 0. The date for submittal of the modified RV internals AMP document has since been revised to October 1, 2012 [9]. This aging management program has been reviewed and changes incorporated using the guidance provided in the SER to MRP-227, Rev. 0 (which now becomes MRP-227-A [3b]) in order to meet this commitment.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 60 of 121 5.0 Examination Acceptance and Expansion Criteria 5.1 Examination Acceptance Criteria 5.1.1 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 [13], provides a set of relevant conditions for the visual (VT-3) examination of removable core support structures in IWB-3520.2. These are:

1. Structural distortion or displacement of parts to the extent that component function may be impaired;
2. Loose, missing, cracked, or fractured parts, bolting, or fasteners;
3. Corrosion or erosion that reduces the nominal section thickness by more than 5%;
4. Wear of mating surface that may lead to loss of functionality; and
5. Structural degradation of interior attachments such that the original cross-sectional area is reduced more than 5%.

For components in the Existing Programs group, these general relevant conditions are sufficient. However, for components where visual (VT-3) is specified in the Primary or the Expansion group, more specific descriptions of the relevant conditions are provided as part of the table in Attachment C. Typical examples are fractured material and completely separated material. One or more of these specific relevant condition descriptions may be applicable to the Primary and Expansion components listed in Tables 2 and 3. The examination acceptance criteria for components requiring visual (VT-3) examinations is thus the absence of the relevant condition(s) specified in Table 5. The disposition can include a supplementary examination to further characterize the relevant condition, an engineering evaluation to show that the component is capable of continued operation with a known relevant condition, or repair/replacement to remediate the relevant condition.

5.1.2 Visual (VT-i) Examination Visual (VT-i) examination is defined in the ASME Code Section XI as an examination conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion. For these guidelines VT-i has only been selected to detect distortions as evidenced by small gaps between the upper-to-lower mating Report No. 1 i01403.4OiRl 60 mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 61 of 121 surfaces of C-E welded core shroud assembled in two vertical sections. The examination acceptance criterion is thus the absence of the relevant condition of gaps that would be indicative of distortion from void swelling. Visual (VT-i) examinations do not apply to PNP RVI.

5.1.3 Enhanced Visual (EVT-1) Examination Enhanced visual (EVT-1) examination has the same requirements as the ASME Code Section XI visual (VT-i) examination, with additional requirements given in the Inspection Standard, MRP 228 [22]. 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-i examination is the same as for cracking in Section Xl which is crack-like surface-breaking indications.

Therefore, until such time as engineering studies provide a basis by which a quantitative amount of degradation can be shown acceptable for the specific component, any observed relevant condition must be dispositioned. In the interim, the examination acceptance criterion is the absence of any detectable surface-breaking indication.

5.1.4 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 per the Inspection Standard, MRP-228 [22]

provides the basis for detection and length sizing of surface-breaking or near-surface cracks.

The signal-based relevant indication for surface (ET) is thus the same as the relevant condition for enhanced visual (EVT-i) 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.

5.1.5 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 detection of relevant indications established as part of the examination technical justification. When a relevant indication is detected in the cross-sectional area of the bolt or pin, it is assumed to be non-functional and the Report No. 1101403.401R1 61

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 62 of 121 indication is recorded. A bolt or pin that passes the criterion of the examination is assumed to be functional.

Because of this pass/fail acceptance of individual bolts or pins, the examination acceptance criterion for volumetric (UT) examination of bolts and pins is based on a reliable detection of indications as established by the individual technical justification for the proposed examination.

This is keeping with current industry practice. For example, planar flaws on the order of 30% of the cross-sectional area have been demonstrated to be reliably detectable in previous bolt NDE technical justifications for baffle-former bolting.

Bolted and pinned assemblies are evaluated for acceptance based on meeting a specified number and distribution of functional bolts and pins. Criteria for this evaluation can be: 1) found in previous Owners Group reports, 2) developed for use by the PWROG at some future time, or 3) developed on a plant-specific basis.

Locations for augmented MRP-227-A inspections for the PNP RVI are identified in Figure 14.

5.1.6 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. An exaggerated view of void swelling in baffle-former plates is shown in Figure 6 as distortion of the plates. Analysis of void swelling for C-E and Westinghouse designed plants is summarized in MRP-230 [35].

Those conservative analyses predicted that the volume of material exceeding 5% swelling would be small. However, it is an aging effect that must be managed by inspections. The combination of EVT-1 and VT-3 visual inspections recommended for the C-E core shroud structure are designed to provide the information required to manage degradation due to void swelling, embrittlement and IASCC. The visual inspections are targeted at the locations where displacement or separation of plates is most likely to be noted. The extent and character of the distortion at these locations is discussed in MRP-230. These inspections are included to provide physical validation of the swelling calculations. If distortion at these locations is not observed, it is reasonable to assume that the MRP-230 analysis continues to bound the behavior of the structure.

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

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 63 of 121 5.3 Evaluation, Repair and Replacement Strategy Any condition detected during examinations that do not satisfy the examination acceptance criteria of Section 5.1 would be entered and dispositioned in the Corrective Action Program.

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

1. Supplemental examinations, such as surface examination to supplement a visual (VT-i) examination to further characterize and potentially dispose of a detected condition
2. Engineering evaluations that demonstrate the acceptability of detected conditions;
3. Repair to restore a component with a detected condition to acceptable status; or
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 [14] and other NRC approved methodologies would be used to provide acceptance criteria for primary and expansion category items 5.3.1 Reporting Reporting and documentation of relevant conditions and disposition of indications that do not meet the examination acceptance criteria would be performed consistent with MRP-227-A and the ENO Corrective Action Program. ENO will provide a summary report to the EPRI MRP Program Manager of all inspections and monitoring, items requiring evaluation, and new repairs 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 would be expeditiously reported to the PCS Materials Degradation Program manager for consideration of reporting under the NEI 03-08, Materials Initiative Protocol [2].

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 64 of 121 6.0 Operating Experience and Additional Considerations 6.1 Internal and External Operating Experience related to the vessel internals would be periodically reviewed and evaluated by ENO for applicability to this program. Evaluation of all inspections and significant external events would be periodically documented as to their significance for the PNP RVI and the results compiled into an overall industry report to track industry progress, aid in evaluation of significant issues, and identif fleet trends. This report, to be provided by the MRP, will serve to assist ENO in review of operating experience and required monitoring and trending for aging management programs established by the industry. This meets the Needed requirement 7.6 under MRP-227-A [3b].

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 65 of 121 7.0 Responses to the NRC Safety Evaluation Report Applicant/Licensee Action Items As part of the NRC Final Safety Evaluation of MRP-227 [4], a number of action items and conditions were specified by the staff. Table 6 documents PNP responses to the NRC Final Safety Evaluation of MRP-227 [4]. Wherever possible, these items have been addressed in the appropriate sections of this document. All NRC action items and conditions not addressed elsewhere in this document are discussed in this section.

7.1 SER Section 4.2.1. Applicant/Licensee Action Item 1:

ENO has assessed its plant design and operating history and has determined that MRP-227-A

[3b] is applicable to the facility. The assumptions regarding plant design and operating history made in MRP-191 [20] are appropriate for PNP and there are no differences in component inspection at PNP. PNP operated the first 7 cycles of operation with a high leakage core loading pattern. The FMECA and functionality analyses were based on the assumption of 30 years of operation with high leakage core loading patterns; therefore, PNP is bounded by the assumption in MRP-191 [20].

Operations at the PNP conform to the assumptions in Section 2.4 of MRP-227-A [3b].

  • PNP operated for 18 effective full power years (EFPY) with high-leakage core patterns, followed by implementation (Cycle 8) of a low-leakage fuel management strategy for the remaining years of operation [32];
  • PNP operates as a base load unit, and
  • No design changes were implemented beyond those identified in general industry guidance or recommended by the vendor (C-E or Westinghouse) 7.2 SER Section 4.2.2. Applicant/Licensee Action Item 2:

ENO reviewed the information in Table 4-5 of MRP- 191 [20] and determined that this table contains all of the RVI components that are within the scope of license renewal. This is shown in the Table presented in Attachment B.

7.3 SER Section 4.2.3, Applicant/Licensee Action Item 3:

The SE for MRP-227 [4] requires C-E plants to evaluate whether existing plant-specific programs are adequate to manage the aging effects of (1) thermal shield positioning pins and (2) in-core instrument thimble tubes. These actions are not applicable because PNP does not have ICI thimble tubes or a thermal shield.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 66 of 121 7.4 SER Section 4.2.4, Applicant/Licensee Action Item 4:

This action does not apply to C-E designed units.

7.5 SER Section 4.2.5, Applicant/Licensee Action Item 5:

Per the SE for MRP-227 [4], C-E designed plants are required to provide plant-specific acceptance criteria to be applied when performing physical measurements for measuring distortion in the gap between the top and bottom core shroud segments in units with core barrel shrouds assembled in two vertical sections. This requirement is not applicable as the PNP core shroud is bolted and is not assembled in two vertical sections.

7.6 SER Section 4.2.6, Applicant/Licensee Action Item 6:

This action does not apply to the C-E designed units.

7.7 SER Section 4.2.7, Applicant/Licensee Action Item 7:

The SE for MRP-227 [4] requires the applicants/licensees of C-E reactors to develop plant-specific analyses to be applied for their facilities to demonstrate that the C-E lower support columns will maintain their functionality during the period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel or precipitation hardened stainless steel materials. This requirement is not applicable to PNP as CASS, martensitic stainless steel or precipitation hardenened stainless steel materials are not present in the reactor vessel internal lower support structures [28, 29].

7.8 SER Section 4.2.8, Applicant/Licensee Action Item 8:

This document includes the ten attributes in the GALL Report and an inspection plan which addresses the identified plant-specific action items contained in the NRC Final Safety Evaluation for MRP-227 [4]. Regarding fatigue of internals, as stipulated in the SER for MRP-227, PNP reviewed the design bases for the plants core support structure for license renewal, and no time-limited aging analyses (TLAAs) were identified. The additional review per MRP-227-A revealed no new TLAAs. However, fatigue damage for primary and (potentially) expansion internals components such as the core support plate will be managed by performing visual examinations, including any required periodic enhanced visual (EVT-1) examinations. No reduction in examination coverage by plant-specific analysis will be requested. Therefore, this approach is essentially equivalent to managing the effects of fatigue on reactor internals components with fatigue analyses during the period of extended operation through 10 CFR 24.21 (c)(1)(iii). As an example, for the 2013 outage, the core support plate has been identified as a Report No. 1101403 .401R1 66 Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 67 of 121 Primary component requiring enhanced visual inspection (EVT-1) to manage the effects of fatigue, as shown in Table 2. PNP is not requesting any deviations from the guidance provided in MRP-227-A [3b], as approved by the NRC.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 68 of 121 8.0 References

1. EN-DC-202, Rev. 5, NE! 03-08 Materials Initiative Process, Entergy Nuclear Management Manual, 5/18/11. (SI File No. 1101403.201).
2. Nuclear Energy Institute, Revision 2 to NEI 03-08, Guideline for the Management of Materials Issues, dated January, 2010. (SI File No. 1101403.202).

3.a Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 0), EPRI, Palo Alto, CA: 2008. 1016596.

(SI File No. 1 101403.203P). EPRIPROPRIETARY.

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

1 101403.204P). EPRJ PROPRIETARY.

4. Letter from Robert A. Nelson (NRC) to Neil Wilmshurst (EPRI) dated December 16, 2011, Revision 1 to the Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR)

Internals Inspection and Evaluation Guidelines (TAC No. ME0680), NRC ADAMS Accession No. MLI 1308A770. (SI File No. 1101403.206).

5. WCAP-17 133-NP, PWR Vessel Internals Program Plan for Aging Management of Reactor Internals at Palisades Nuclear Plant, Revision 0, November 2009. (51 File No.

1101403.205).

6. NUREG-1 871, Safety Evaluation Report Related to the License Renewal of Palisades Nuclear Plant, Docket No. 50-255, ADAMS Accession No. ML070600578.
7. Email Transmittal from US NRC to ENO dated September 17, 2010, Palisades-Program Plan for Aging Management of Reactor Vessel Internals ME4084, ADAMS Accession No. ML102670721, (SI File No. 1101403.227).
8. Letter from Entergy Nuclear Operations to the U.S. Nuclear Regulatory Commission dated March 31, 2011, Withdrawal and New Commitment for Program Plan for Aging Management of Reactor Vessel Internals, ADAMS Accession No. ML! 10910461, (SI File No 1101403.226).
9. Letter from Entergy Nuclear Operations to US NRC dated June 20, 2012, Revised Commitment Date for Program Plan for Aging Management of Reactor Vessel Internals Submittal, (SI File No.1101403.233).

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10. NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Rev. 2, U. S.

Nuclear Regulatory Commission, December 2010.

11. Palisades Nuclear Plant Fourth 10-Year Interval Master Inservice Inspection Plan, Latest Revision (SI File No. 1101403.228).
12. LR-AMPBD-26-CHEMISTRY, Rev. 3, Water Chemistry Program, License Renewal Aging Management Program Basis Document. (SI File No. 1101403.207).
13. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Rules for Inservice Inspection of Nuclear Power Plant Components, 2001 Edition, 2003 Addenda.
14. WCAP-17096-NP, Reactor Internals Acceptance Criteria Methodology and Data Requirements, Revision 2, December 2009. (SI File No. 1101403.222).
15. Palisades Nuclear Plant Application for Renewed Operating License, March 22, 2005, (SI File No. 1101403.208).
16. Safety Evaluation Report with a Confirmatory Item Related to the License Renewal of the Palisades Nuclear Plant, Docket No. 50-255, ADAMS Accession No. ML061530042. (SI File No. 1101403.209).
17. NUREG- 1801, Generic Aging Lessons Learned (GALL) Report, Rev. 0, U. S.

Nuclear Regulatory Commission, July 2001.

18. ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition including Addenda through Winter 1965.
19. Pressurized Water Reactor Primary Water Chemistry Guidelines, Volumes 1 and 2, Revision 6, Electric Power Research Institute, Palo Alto, CA: 2007, 1014986.
20. Materials Reliability Program: Screening Categorization and Ranking of Reactor Internals Components of Westinghouse and Combustion Engineering PWR Design (MRP-191), Electric Power Research Institute, Palo Alto, CA: 2007. 1013234. (SI File No. 1101403.221P). EPRI PROPRIETARY.
21. NUREG-1801, Generic Aging Lessons Learned (GALL) Report, Rev. 1, U. S.

Nuclear Regulatory Commission, September 2005.

22. EPRI Report MRP-228, Materials Reliability Program Inspection Standard for Reactor Internals, Latest Revision. (SI File No. 1101403.210). EPRI PROPRIETARY.
23. NMM LI-102, Corrective Action Process Report No. 1 101403.401R1 69 mc?

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24. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
25. LRPG 12, LRP Operation Experience Data Collection, Palisades Nuclear Plant.
26. Nuclear Management Company response to NRC RAI Dated August 27, 2005, Supplementary Information for the Palisades Application for Renewed Operating License Resulting from Aging Management Review Audit, ADAMS Accession No. ML052440392. (SI File No. 1101403.212).
27. Letter from Karl R. Goller, (Directorate of Licensing) to R. C. Youngdahl (Consumers Power Company) dated August 30, 1974 (Docket No. 50-255), ADAMS Accession No. ML02078490, (SI File No. 1101403.229).
28. LR-AMR-RVI, Rev. 4, Aging Management Review Reactor Vessel Internals Palisades Nuclear Plant License Renewal Project, October 20, 2005 (SI File No. 1101403.223).
29. License Renewal Aging Management Program Basis Document, LR-AMPBD VSLINTERNALS, Reactor Vessel Internals Inspection Program. (SI File No.

1101403.211).

30. Letter from Nuclear Management Company to U.S. Nuclear Regulatory Commission Dated June 3, 2003, License Amendment Request: Increased Rated Thermal Power, ADAMS Accession No. MLO3 1611053. (SI File No. 1101403.215).
31. U.S. Nuclear Regulatory Commission Notice No: 04-087 Dated July 15, 2004, NRC Approves Power Uprate for Palisades. (SI File No. 110 1403.230).
32. Letter from Darl S Hood (US NRC) to Nathan Haskell (Palisades Nuclear Plant) Dated November 14, 2000, Palisades Plant Reactor Vessel Neutron Fluence Evaluation and Revised Schedule for Reaching Pressurized Thermal Shock Screening (TAC No.

MA8250), ADAMS Accession No. ML003768794. (SI File No. 1101403.217).

33. EPRI Letter MRP 20 10-025, Summary of Operating Experience with Pressurized Water Reactor Internals through 2009, March 30, 2010. EPRIPROPRIETARY.
34. EPRI Report MRP-219, Materials Reliability Program: Pressurized Water Reactor Inspection Data Survey, Latest Revision. EPRIPROPRIETARY.

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35. Materials Reliability Program: Functionality Analysis for Westinghouse and Combustion Engineering Representative PWR Internals (MRP-230-Rev.1). Electric Power Research Institute, Palo Alto, CA: 2009. 1019091. (SI File No. 1 101403.231P)

EPRI PROPRIETARY.

36. Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals (MRP-232). Electric Power Research Institute, Palo Alto, CA: 2008. 1016593. (SI File No. 1 101403.232P) EPRIPROPRIETARY.
37. SI Report, 1000434.401, Revision 0, Independent Review of Palisades Reactor Vessel Internals Program and Inspection Plan, September 2011. (SI File No.

1000434.40 1)

38. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities, 50.55a, Codes and Standards.
39. Letter to US NRC Dated October 19, 2005, Report of Facility Changes, Tests and Experiments and Summary of Commitment Changes, Docket No. 50-255, ADAMS Accession No. ML053060340. (SI File No. 1101403.213).
40. Inside WANO, Reducing Radiation Dose Rates, Vol. 14, No. 2, 2006. (SI File No.

1101403.214).

41. Letter from US Nuclear Regulatory Commission to Entergy Nuclear Operations Dated May 6, 2011, Palisades Nuclear Plant Post-Approval Site Inspection for License Renewal, Inspection Report 05000255/201 1008(DRS). (SI File No. 1101403.218).

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 72 of 121 Table 2. C-E Plants Primary Category Components from MRP-227-A L3b]

Effect Expansion Examination Item Applicabilitj Examination Coverage Comments (Mechanism) Link MethodIFrequency Core Shroud Bolted plant Cracking (IASCC, Core support Baseline volumetric (UT) 100% of accessible bolts (3)

Inspections to be performed

, or as Assembly designs Fatigue) column bolts, examination between 25 and supported by pant-specwic in 2022 (at approximately (Bolted) barrel-shroud 35 EFPY, with subsequent justification. Heads are 34.3 EFPY)

Core shroud Aging bolts examination after 10 to 15 accessible from the core side. UT bolts Management additional EFPY to confirm accessibility may be affected by (IE and ISR) (2) stability of bolting pattern, complexity of head and locking Re-examination for high- device designs.

leakage core designs requires continuing inspections on a ten-year See Figure 4-24 of MRP-227-A interval.

Core Shroud Plant designs Cracking (IASCC) Remaining Enhanced visual (EVT-1) Axial and horizontal weld seams Assembly with core axial welds examination no later than 2 at the core shroud re-entrant (Welded) shrouds Aging refueling outages from the corners as visible from the core Core shroud assembled in Management (IE) beginning of the license side of the shroud, within six plate-former plate two vertical (2) renewal period and inches of central flange and weld sections subsequent examination on horizontal stiffeners.

a ten-year interval.

(Not applicable See Figures 4-12 and 4-14 of for Palisades) NIA MRP-227-A Stn,ctu,wI Integrity Associates, h7c ReportNo. 1101403.401R1 72

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 73 of 121 Item Applicability Exannon Examination Coverage Comments (Mh:nm) MethodIFrequency Core Shroud Plant designs Cracking (IASCC) Remaining Enhanced visual (EVT-1) Adal weld seams at the core Assembly with core axial welds, examination no later than 2 shroud re-entrant corners, at the (Welded) shrouds Aging ribs and rings refueling outages from the core mid-plane (+/- three feet in Shroud plates assembled with Management (IE) beginning of the license height) as visible from the core (2) full-height renewal period and side of the shroud.

(Not applicable shroud plates. subsequent examination on N/A for Palisades) a ten-year interval. See Figure 4-13 of MRP-227-A Core Shroud Bolted plant Distortion None Visual (VT-3) examination no Core side surfaces as indicated. Visual inspections planned Assembly designs (Void Swelling), later than 2 refueling outages to be performed in (Bolted) including: from the beginning of the See Figures 4-25 and 4-26 of augmented program in 2013.

Assembly

  • Abnormal license renewal period. MRP-227-A interaction with Subsequent examinations on fuel assemblies a ten-year interval.
  • Gaps along high fluence shroud plate joints

Stnictura! Int.qrlly Assoclatos, liw?

Report No. 1101403.401R1 73

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 74 of 121 Item Expanon Applicability Examination Coverage Comments (Mhanism) MethodIFrequency Core Shroud Plant designs Distortion None Visual (VT-i) examination no If a gap exists, make three to five Assembly with core later than 2 refueling outages measurements of gap opening (Welded) shrouds (Void Swelling), as from the beginning of the from the core side at the core Assembly assembled in evidenced by license renewal period, shroud re-entrant comers. Then, two vertical separation Subsequent examinations on evaluate the swelling on a plant-(Not Applicable sections between the upper a ten-year interval, specific basis to determine for Palisades) and lower core frequency and method for N/A shroud segments additional examinations.

Aging See Figures 4-12 and 4-14 of Management (IE) MRP-227-A Core Support All plants Cracking (SCC) Lower core Enhanced visual (EVT-1) 100% of the accessible surfaces Enhanced visual Inspections Barrel support examination no later than 2 (4) of the upper flange weld. planned to be performed in Assembly beams refueling outages from the augmented program in 2013.

Upper(core Core support beginning of the license support barrel) barrel renewal period. Subsequent See Figure 4-15 of MRP-227-A flange weld assembly examinations on a ten-year upper cylinder interval.

Upper core barrel flange Core Support All plants Cracking (SCC, Lower cylinder Enhanced visual (EVT-1) 100% of the accessible surfaces Enhanced visual Inspections Barrel IASCC) axial welds examination no later than 2 (4) of the lower cylinder welds planned to be performed in Assembly refueling outages from the augmented program in 2013.

Lower cylinder Aging beginning of the license girth welds Management (lE) renewal period. Subsequent See Figure 4-15 of MRP-227-A examinations on a ten-year interval Lower Support All plants Cracking (SCC, None Visual (VT-3) examination no 100% of the accessible surfaces Visual inspections planned Structure IASCC, Fatigue later than 2 refueling outages of the core support column to be performed in Core support including damaged from the beginning of the 5 welds augmented program in 2013.

column welds or fractured license renewal period.

material) See Figure 11 for Palisades Subsequent examinations on design.

Aging a ten-year interval.

Management

. (lE,TE)

Stn,ctur,I Integrity Associates, ft Report No. 1101403.401R1 74

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 75 of 121 Item Applicability Examination Coverage Comments (Mechanism) Unk Core Support All plants Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Barrel demonstrated by time limited defined by evaluation to determine Assembly aging analysis (TLA), the potential location and extent of Lower flange enhanced visual (EVT-1) fatigue cracking.

weld examination, no later than 2 refueling outages from the See Figures 4-15 and 4-16 of N/A (Not applicable beginning of the license MRP-227-A for Palisades) renewal period. Subsequent examination on a ten-year interval.

Lower Support All plants with a Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Enhanced visual Inspections Structure core support demonstrated by time limited defined by evaluation to determine planned to be performed in Core support plate Aging aging analysis (TLAA.), the potential location and extent of augmented program in 2013.

plate Management (IE) enhanced visual (EVT-1) fatigue cracking.

examination, no later than 2 refueling outages from the See Figure 4-16 of MRP-227-A beginning of the license renewal period. Subsequent examination on a ten-year interval.

Upper Internals All plants with Cracking (Fatigue) None If fatigue life cannot be Examination coverage to be Assembly core shrouds demonstrated by time limited defined by evaluation to determine Fuel alignment assembled with aging analysis (TLAA), the potential location and extent of plate full-height enhanced visual (EVT-1) fatigue cracking.

shroud plates examination, no later than-2 (Not applicable to refueling outages from the See Figure 4-17 of MRP-227-A Palisades) beginning of the license renewal period. Subsequent examination on a ten-year N/A interval.

Sinictuiwl Int.gdty Associalos, mc?

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 76 of 121

. .. Effect Expansion Examination Item Applicability Examination Coverage Comments (Mechanism) 1 Link° MethodIFrequency 11 Control Element All plants with Cracking (SCC, Remaining Visual (VT-3) examination, 100% of tubes in peripheral CEA Visual VT-3 exam planned to 161 Assembly instrument Fatigue) that instrument no later than 2 refueling shroud assemblies (i.e., those be performed in augmented Instrument guide guide tubes in results in missing guide tubes outages from the beginning adjacent to the perimeter of the program in 2013.

tubes the CEA shroud supports or within the of the license renewal period, fuel alignment plate).

assembly separation at the CEA shroud Subsequent examination on welded joint assemblies, a ten-year interval.

Note: Palisades between the tubes See Figure 12 for Palisades has a unique and supports Plant-specific component design showing Control Rod configuration. integrity assessments may Shroud Assembly and ICI be required if degradation is Instrument Guide Tubes detected and remedial action is needed.

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

full-height surface-breaking beginning of the license (Not applicable to shroud plates. indication in the renewal period. Subsequent See Figure 4-19 of MRP-227-A Palisades) welds or beams examination on a ten-year N/A interval, if adequacy of Aging remaining fatigue life cannot Management (IE) be demonstrated.

Notes:

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

SL,vctwiI h,LeqrltJr Associates, hw.

ReportNo. 1101403.401R1 76

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 77 of 121 Table 3. C-E Plants Expansion Category Components from MRP-227 13b1 Examination Examination Comments Item (1)

Applicability Effect (Mechanism) Primary Link Method/Frequency CoveragelFreguency 1

Core Shroud Bolted plant Cracking (IASCC, Core shroud Volumetric (UT) 100 % (or as supported by Contingency if Assembly designs Fatigue) bolts examination, plant-specific justification) (2) indications are found (Bolted) of barrel-shroud and guide in UT exam of core Barrel-shroud bolts Aging Management Re-inspection every 10 lug insert bolts with neutron shroud bolts (IE and ISR) years following initial fluence exposures > 3 inspection, displacements per atom (N/A for 2013)

(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 Assembly Fatigue) support barrel) examination, and adjacent base metal (2)

Lower core barrel flange flange weld.

Re-inspection every 10 See Figure 4-15 of N/A (Not applicable to years following initial MRP-227-A Palisades) inspection.

Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1) 100% of accessible surface Contingency if Assembly support barrel) examination, of the weld and adjacent Upper Cylinder indications are found Aging Management flange weld 2 base.

(including welds) (IE) Re-inspection every 10 in EVT 1 exam o f years following initial upper flange weld in See Figure 4-15 of inspection MRP-227-A 2013 Core Support Barrel All plants Cracking (SCC) Upper (core Enhanced visual (EVT-1) 100% of accessible bottom Contingency if Assembly support barrel) examination, surface of the flange (2) indications are found Upper Core Barrel flange weld Re-inspection every 10 in EVT-1 exam of Flange years following initial See Figure 4-15 of upper flange weld in inspection MRP-227-A 2013 STh1cturBI Integrity Associates, mc?

Report No. 1101403.401R1 77

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGNG MANAGEMENT PROGRAM July 26, 2012 Page 78 of 121 Examination Examination Comments Item Applicability Effect (Mechanism) Primary Link MethodlFreguency ii) Coverage/Frequency 111 Core Support Barrel All plants Cracking (SCC) Core barrel Enhanced visual (EVT-1) 100% of one side of the Assembly assembly girth examination, with initial and accessible weld and adjacent Core barrel assembly welds subsequent examinations base metal surfaces for the axial welds dependent on the results of weld with the highest core barrel assembly girth calculated operating stress. N/A 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 shroud plates inspection. MRP-227-A. barrel) flange weld in Core Shroud Bolted plant Cracking (IASCC, Core shroud Ultrasonic (UT) 100 % (or as supported by Contingency if Assembly designs Fatigue) bolts examination, plant-specific analysis) of indications are found (Bolted) core support column bolts in UT exam of core Core support column Aging Management Re-inspection every 10 with neutron fluence shroud bolts bolts (IE) years following initial exposures> 3 dpa. (2) inspection. (N/A for 2013)

See Figures 4-16 and 4-33 of MRP-227-A StructurtI Integrity Associates, Inc Report No. 1 101403.401R1 78

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 79 of 121 Examination Examination Comments Item Applicability Effect (Mechanism) Primary Link MethodlFrequency (1) Coverage/Frequency I Core Shroud Plant designs with Cracking (IASCC) Shroud plates of Enhanced visual (EVT-1) Axial weld seams other than 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. N/A Ribs and rings plates. years following initial inspection. See Figure 4-13 of (Not applicable for MRP-227-A Palisades)

Control Element All plants with Cracking (SCC, Peripheral Visual (VT-3) examination. 100% of tubes in CEA shroud Contingency if Assembly 3 instwment guide Fatigue) that results in instwment guide (2) assemblies. indications are found Remaining instrument tubes in the CEA missing supports or tubes within the Re-inspection every 10 in VT-3 exam of core guide tubes shroud assembly. separation at the CEA shroud years following initial instrument guide See Figure 12 welded joint between assemblies, inspection, tubes Note: Palisades the tubes and supports.

has a unique configuration.

Notes:

1. Examination acceptance criteria for the C-E components are in Table 5 (MRP-227-A Table 5-2).
2. A minimum of 75% coverage of the entire examination area or volume, or a minimum sample size of 75% of the total population of like components of the examination is required (including both accessible and inaccessible portions).
3. Designated as Control Rod Shroud Assembly for Palisades.

Stnictur& Int9grily Associates, h Report No. 1101403.401R1 79

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 80 of 121 Table 4. C-E Plants Existing Program Components Credited in MRP-227-A [3b]

Item Applicability Examination Method Comments

( Mech:ni S m ) Examination Coverage Core Shroud All plants Loss of material ASME Visual (VT-3) examination, general First 1 0-year ISI after 40 To be inspected Assembly (Near) Code condition examination for detection of years of operation, and at Guide lugs Section Xl excessive or asymmetrical wear. each subsequent inspection Guide lug inserts Aging interval.

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

(Not applicable to Management Palisades) (IE and ISR) N/A Lower Support All plants with core Loss of Material ASME Visual (JT-3) examination Accessible surfaces at Structure shrouds assembled in (Wear) Code specified frequency Fuel alignment pins two vertical sections Section Xl (Not applicable to Aging N/A Palisades) Management (IE and ISR)

Core Barrel All plants Loss of Material ASME Visual T-3) examination Area of the upper flange To be inspected Assembly (Wear) Code potentially susceptible to Upper flange Section Xl wear Stn,cturW Integrity Assoclales, ftu Report No. 1101403.401R1 80

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 27, 2012 Page 81 of 121 Table 5. C-E Plants Examination Acceptance and Expansion Criteria Applicable to Palisades Nuclear Plant [3b]

Examination Item Applicability Acceptance Criteria ExpaflSiOfl Criteria Additional Examination Expansion Link(s)

(Note 1) Acceptance Criteria Core Shroud Assembly Bolted plant designs Volumetric (UT) a. Core support a. Confirmation that >5% of the a and b. The examination (Bolted) examination, column bolts core shroud bolts in the four acceptance criteria for the UT of Core shroud bolts plates at the largest distance the core support column bolts The examination b. Barrel-shroud from the core contain and barrel-shroud bolts shall be acceptance criteria for the bolts unacceptable indications shall established as part of the UT of the core shroud bolts require UT examination of the examination technical shall be established as part lower support column bolts justification.

of the examination technical barrel within the next 3 refueling justification, cycles.

b. Confirmation that> 5% of the core support column bolts contain unacceptable indications shall require UT examination of the barrel-shroud bolts within the next 3 refueling cycles.

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

Assembly The specific relevant conditions are evidence of abnormal interaction with fuel assemblies, gaps along high fluence shroud plate joints, and vertical displacement of shroud plates near high fluence joints Structuil Integrity Associates, h7c?

Report No. 1101403.401R1 81

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 27, 2012 Page 82 of 121 Examination Item Applicability Acceptance Criteria Expansion Criteria Additional Examination Exjnsion Link(s)

(Note 1) Acceptance Cntena 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 Upper (core support barrel) The specific relevant length has been detected and indication.

flange weld condition is a detectable Upper core barrel sized in the upper 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 Lower cylinder girth welds The specific relevant the length has been detected axial welds is a detectable 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 Core support column welds The specific relevant condition is missing or separated welds.

Core Support Barrel All plants Visual (EVT-1) examination. None N/A N/A Assembly Lower flange weld The specific relevant condition is a detectable (Not applicable to crack-like indication.

Palisades)

Sbvctuml hJt5grily AssocIws, hic.

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PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 27, 2012 Page 83 of 121 Examination Item Applicability Acceptance Criteria Expansion Criteria Additional Examination Expansion Link(s)

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

Upper Internals All plants with core shrouds Visual (EVT-1) examination. None N/A N/A Assembly t2 assembled with full-height Fuel alignment plate shroud plates The specific relevant condition is a detectable (Not applicable to crack-like surface indication.

Palisades)

Control Element All plants with instrument tubes Visual (VT-3) examination. Remaining Confirmed evidence of missing The specific relevant conditions Assembly 131 in the CEA shroud assembly instrument tubes supports or separation at the are missing supports and Instrument guide tubes The specific relevant within CEA shroud welded joint between the tubes separation at the welded joint conditions are missing assemblies and supports shall require the between the tubes and 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.

Note:

1. The examination acceptance criteria for visual examination is the absence of the specified relevant condition(s).
2. Listed as Upper Internals Assembly in MRP-227 Table 5-2.
3. Designated as Control Rod Shroud Assembly for Palisades.

SIrUctU,aI Integrity Associates, h,c ReportNo. 1101403.401R1 83

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 84 of 121 Table 6. Palisades Response to the NRC Final Safety Evaluation of MRP-227-A [3bJ MRP-227 SER Item Palisades Response SER Section 4.1.1, Topical Report In accordance with SER Section 4.1.1, the Lower Core Support Beams, Core Condition 1: Moving components Support Barrel Assembly Upper Cylinder and Upper Core Barrel Flange have from No Additional Measures to been added to the Palisades Expansion inspection category and are Expansion category. contained in Table 3. The components are linked to the Primary components Lower Support Structure Deep Beams and Core Support Barrel Upper (core support barrel) flange weld.

SER Section 4.1.2, Topical Report In accordance with SER Section 4.1.2, the Core Support Barrel Assembly Condition 2: Inspection of Lower Cylinder Girth Welds have been added to the Palisades Primary components subject to irradiation- inspection category and are contained in Table 2. The examination method is assisted stress corrosion cracking. consistent with the IVERP recommendations for these components, the examination coverage conforms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination frequency is on a 10-year interval consistent with other Primary inspection category components.

SER Section 4.1,3, Topical Report In accordance with SER Section 4.1.3, the Core Support Column (casting or Condition 3: Inspection of high wrought) welds in the lower support structure have been added to the consequence components subject to Palisades Primary inspection category and are contained in Table 2. The multiple degradation mechanisms examination method is consistent with MRP recommendations for these components. The coverage confirms to the criteria described in Section 3.3.1 of the NRC SE, and the re-examination frequency is on a 10-year interval consistent with the other Primary inspection category components.

SER Section 4.1.4, Topical Report In accordance with SER Section 4.1.4, Palisades would meet the minimum Condition 4: Minimum inspection coverage specified in the SER. The appropriate wording has been examination coverage criteria for added to Table 3 examination coverage.

expansion inspection category components SER 4.1.5, Topical Report In accordance with SER Section 4.1.5, the examination frequency for baffle-Condition 5: Examination former bolts (i.e., core shroud bolts) specifies a 10-year inspection frequency frequencies for baffle former bolts following the baseline inspection in Table 2.

and core shroud bolts SER 4.1.6, Topical Report In accordance with SER Section 4.1.6, Table 3 requires a 10-year re Condition 6: Periodicity of the re- examination interval for all Expansion: inspection category components once examination of expansion degradation is identified in the associated Primary inspection category inspection category components component and examination of the expansion category component commences.

Sinicturel Intagrity Associates, Inc Report No. 1101403.401R1 84

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 85 of 121 Table 6. Palisades Response to the NRC Final Safety Evaluation of MRP-227-A [3bj (contd)

MRP-227 SER Item Palisades Response SER Section 4.1.7, Topical Report This condition applies to update of the industry guidelines. No plant-specific Condition 7: Updating of industry actions are required.

guideline SER Section 4.2.1, The evaluation of design and operating history demonstrating that IVERP-227 Applicant/Licensee Action Item 1 is applicable to Palisades is contained in Section 4.10.

SER Section 4.2.2, The Palisades review of components within the scope of license renewal Applicant/Licensee Action Item 2 against the information contained in MRP- 191 Table 4-5 is discussed in Section 4.10.

SER Section 4.2.3, No action required. Neither ICI thimble tubes nor thermal shields are present Applicant/Licensee Action Item 3 in the Palisades reactor vessel.

SER Section 4.2.4, No action required. This action does not apply to C-E designed units.

Applicant/Licensee Action Item 4 SER Section 4.2.5, No action required. Palisades has a bolted core shroud configuration.

Applicant/Licensee Action Item 5 SER Section 4.2.6, No action required. This action does not apply to C-E designed units.

Applicant/Licensee Action Item 6 SER Section 4.2.7, No action required. Palisades does not have CASS, martensitic stainless steel Applicant/Licensee Action Item 7 or precipitation hardened stainless steel materials in the reactor vessel internals.

SER Section 4.2.8, The submittal of responses to meet A/LAI No. 8 are contained in Sections 2.3, Applicant/Licensee Action Item 8 4.4, 4.5, 4.6, 4.10, 7.8 and Attachment C.

Sinictura! Integrity Associates, Inc.

Report No. 1101403.401R1 85

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 86 of 121 Table 7. Program Enhancement and Implementation Schedule RFO Cycle End Estimated Inspection Method and AMP-Related Scope Comments fumber Quarter/Year EFPY Criteria 23 Fall 2013 26.1

  • MRP-227-A Inspections Contingency exams of relevant shroud assembly (assembly), Core support barrel in accordance with expansion components if indications assembly (upper core support barrel flange weld), MRP-228. are found in examinations of upper Lower support structure (core support plate), flange weld.

Control rod shroud assembly (ICI instrument guide tubes)

  • Inspections in

. ASME Section XI 10 Year ISI inspections of accordance with Core shroud assembly (Guide lugs, guide lug Palisades ISI Program inserts and bolts) and Core barrel assembly (upper flange) 24 Spring2ol5 274 Notapplicable Notapplicable Notapplicable 25 Fall 2016 28.8 Not applicable Not applicable Not applicable 26 Spring2OtS 30.2 Notapplicable Notapplicable Notapplicable 27 Fall 2019 31.5 Not applicable Not applicable Not applicable 28 Spring 2021 32.9 Not applicable Not applicable Not applicable 29 Fall 2022 34.3

  • MRP-227-A inspections Contingency exams of relevant shroud assembly (assembly and core shroud in accordance with expansion components if indications bolts), Core support barrel assembly (upper core MRP-228. are found in examinations of upper support barrel flange weld), Lower support flange weld.

strupture (core support plate), Control rod shroud assembly (instrument guide tubes)

  • Inspections in accordance with Core shroud assembly (Guide Jugs, guide lug Palisades ISI Program inserts and bolts) and Core barrel assembly (upper flange)

Sinicturil InMqrtty Assoclatos, Inc Report No. 1101403.401R1 86

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 87 of 121 Table 7. Program Enhancement and Implementation Schedule (contd)

RFO Cycle End Estimated Inspection Method and AMP-Related Scope Comments Number Quarter/Year EFPY Criteria 30 Spring 2024 35.7 Not applicable Not applicable Not applicable 31 Fall 2025 37.1 Not applicable Not applicable Not applicable 32 Spring 2027 38 4 Not applicable Not applicable Not applicable 33 Fall 2028 39.8 Not applicable Not applicable Not applicable 34 Spring 2030 412 Not applicable Not applicable Renewed Operating License Expires 3/24/2031 fSbvctu8l Integrity Associates, ft1c ReportNo. 1101403.401R1 87

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 88 of 121

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Espinsien: EVIl eranisaitha cub icriiii and cab a cccl anaccinatiane dvpecdastan rate barrel ->tblyup-rre>-ld e.xazinaticn Caverale: 103 at arblecvald and adjaceni lace natal elr&cee Ott lb Idivith the Piimar Enhanced senal çEVT-l2 Priiai: Eai;o a(UT iatai afc aSa4 b;ba exatrinrean of care aappart pIta.

2535 £FT>1 ub3eq 15daaiE1Y no latea:han2 ,sfnchszouaces fromthebeitnningofihelccecaa 1ac apeio1 ana:n-vaaia;a1 renewal period Sabaoaexi uia Va)ma LT1 ,:aiaai vi0iiatai andaab ua 0 onasrixaliaxoa atenyearintclsal axaarauhz,aa fr *:ia ye utonthensabe at :1E co,e thoadbal

- L t*.

Cev.jae(i) 100 a! c ble aaaa by]ea aia aaaaccas blefram aaaaida UTac:asthil zv bffara3 bcaa3axifbeaiad1o:kdeadaea cl R-27 I apethoeis kr Paheede Priner (Expeeieub: - an Expansion Coinponenta dcl a1 bali th .acae,>3 dcc Figure 14. Palisades Reactor Vessel Internals Inspection Plan SL,wcturW Integrity Associates, inc Report No. 1101403.401R1 88

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 89 of 121 ATTACHMENT A SECTION XI 10 YEAR ISI EXAMINATIONS OF B-N-2 AND B-N-3 INTERNALS COMPONENTS FOR PALISADES NUCLEAR PLANT [11]

Structural InteqrIty Associates, Thc?

ReportNo. 1101403.401R1 89

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 90 of 121 Fabrication ID Examination Area Code Exam Extent of Exam Description Category Method Lower Internals Exterior (Core Barrel)

Alignment Keys and Fastening B-N-3 VT-3 All accessible surfaces Devices @ 00 Alignment Keys and Fastening B-N-3 VT-3 All accessible surfaces Devices @ 90° Alignment Keys and Fastening B-N-3 VT-3 All accessible surfaces Devices @ 180° Alignment Keys and Fastening B-N-3 VT-3 All accessible surfaces Devices @ 270° Lifting Rig Hookup and Pin B-N-3 VT-3 All accessible surfaces Attachment Upper Barrel to Flange Weld and B-N-3 VT-3 Accessible weld surface (OD.)

Welded Gussets, 0° through 360° Outlet Nozzle and Mating Interface Surfaces, B-N-3 VT-3 All accessible surfaces

@0° & 180° Upper Barrel to Lower Barrel Shell Weld, B-N-3 VT-3 Accessible weld surface (O.D.)

0° through 360° Exterior Shroud and Fastening Devices, B-N-3 VT-3 All exterior accessible surfaces 0° through 360° Core Stabilizer Keys on Bottom of Core Barrel, B-N-3 VT-3 All accessible surfaces 0° through 360° Alignment Pins (located beneath B-N-3 VT-3 All accessible surfaces nozzles) @ 0° Alignment Pins (located beneath B-N-3 VT-3 All accessible surfaces nozzles) @ 90° Alignment Pins (located beneath B-N-3 VT-3 All accessible surfaces nozzles) @ 180° Alignment Pins (located beneath B-N-3 VT-3 All accessible surfaces nozzles) @ 270° Sinwiural Inteqdty Assoclatos. Inc.

Report No. 1101403.401R1 90

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 91 of 121 Lower Internals Interior (Core Barrel)

Alignment Keys and Top Ledge of B-N-3 VT-3 All accessible surfaces Core Shroud 0° through 360° Core Shroud Boltingo° through 3600 B-N-3 VT-3 All accessible bolts Top Ledge of Core Shroud 0° through B-N-3 VT-3 All accessible surfaces 360° Lower Core Plate, Locking Devices, Access Port and Bottom Row of Core B-N-3 VT-3 All accessible surfaces Shroud Bolting 0° through 3600 Top Section of Core Shroud, Keys, B-N-3 VT-3 All accessible surfaces and Ledge 0° through 3600 Upper Guide Structure Exterior Exterior Surfaces of the Upper Guide Structure, Thermocouples, Connectors, Guide Tubes, Fastening All accessible areas and Devices, Flow Holes, Keyways and B-N-3 Peripheral Scan. Fuel Alignment surfaces Guide Pins on Bottom. 0° through 360° Reactor Vessel Interior Reactor Vessel to Head Seal Surface B-N-2 VT-3 All accessible surfaces Area 0° through 360° Reactor Vessel to Core Barrel Seal Surface Area and Keyways 0° through B-N-2 VT-3 All accessible surfaces 360° Outlet and Inlet Nozzle Surfaces and Outlet Nozzle Mating Interface B-N-2 VT-3 Surfaces 0°, 60°, 120°, 180°, 240°,

and 300° Surveillance Capsule Holder @ 80° B-N-2 VT-3 Surveillance Capsule Holder @ 1000 B-N-2 VT-3 Surveillance Capsule Holder @ 110° B-N-2 VT-3 Skuctural Integrity ASSOCIateS, 1nc Report No. 1101403.401R1 91

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 92 of 121 Surveillance Capsule Holder @ 2600 B-N-2 VT-3 Surveillance Capsule Holder @ 2800 B-N-2 Surveillance Capsule Holder @ 290° B-N-2 VT-3 Core Stabilizer Lugs 00 B-N-2 VT-3 All accessible surfaces Core Stabilizer Lugs 60° B-N-2 VT-3 All accessible surfaces Core Stabilizer Lugs 120° B-N-2 VT-3 All accessible surfaces Core Stabilizer Lugs 1800 B-N-2 VT-3 All accessible surfaces Core Stabilizer Lugs 240° B-N-2 VT-3 All accessible surfaces Core Stabilizer Lugs 300° B-N-2 VT-3 All accessible surfaces Core Stop Lugs and Flow Skirt 0° B-N-2 VT-3 All accessible surfaces through 360 General Scan of Bottom Head for VT-3 Debris 0° through 360° Stnrntuml Integrity Assodatos, Ix.

Report No. 1101403.401R1 92

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGTNG MANAGEMENT PROGRAM July 26, 2012 Page 93 of 121 ATTACHMENT B PALISADES REACTOR VESSEL INTERNALS COMPONENTS IN LRA [15] AND AGING MANAGEMENT EVALUATION [281 Report No. 11 01403.401R1 93 jStntcturaI Inleurity Associates, mc?

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 94 of 121 Componenti Additional Aging Effect Aging Augmented Current Classification Classification Final Commodity Description Maamnt Similarity SoXI Comments in MRP-191 in MRP-227 Classification Management Requirements Control Rod Shroud Control Rod Changes in Reactor Not specifically Not specifically Similar to A No

- NONE Outer surface Only outer Assembly Shroud Plates Dimensions, Vessels listed listed CEA Inspection would surfaces are Control Rod Shroud Cracking Internals shrouds and covered easily Inspection therefore under tJGS accessible for Program would exterior exam exam Water receive Chemistry similar Program dassificatlo n

Control Rod Shroud Top Support Cracking Reactor Not specifically Not specifically Similar to A No

- NONE Would be These are Assembly Adjustable Vessels listed listed CEA shroud Inspection partly inspectable Shroud Top Support Bushings and Internals bolts and covered as they are in Screws Inspection therefore under UGS the top of the Program would exterior exam IJGS Water receive Chemistry similar Program dassificatio n

Control Rod Shroud Top and Cracking Reactor Not specifically Not specifically This A No

- NONE Part of these Assembly Bottom Shroud Vessels listed Fisted component Inspection would be Shroud Support Lug Support Lugs Internals would be seen dunng Inspection evaluated in the UGS Program the same exterior exam Water manner as Chemistry additional Program components on the CEA shroud assemblies and would recewe

- simfar

- dassificatio ns -

Report No. 1101403.401R1 94 Stnwtur8lInt9gri!yAssociaJsft1c.*

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 95 of 121 CmOd Management Similarity SeconXt Comments M Clason Requements Control Rod Shroud Changes in Reactor Not specifically Not specifically This A No

- NONE Part of these These could Assembly Dimensions Vessels listed listed component Inspection would be be difficult to Control Rod Support Internals would be seen during see as they Lug Inspection evaluated in the IJGS are in the Program the same exterior exam UGS. Only manner as the outer additional surfaces components - would be on the CEA visible.

shroud assemblies and would receive similar ctassificatio ns Control Rod Shroud Fuel Alignment Changes in Reactor A NIA NIA A No

- NONE Covered In Section XI Assembly Plate Fuel Dimensions, Vessels Inspection under UGS program Fuel Guide Pin Guide Pins Cracking, Internals exterior exam Loss of Inspection Preload Program Water Chemistry Program ASME Section XIIWB,IWC, -

IWDJWF Iriservice Inspection Program Control Rod Shroud Fuel Alignment Changes in Reactor A NIA NIA A No

- NONE Covered In Section XI Assembly Plate Fuel Dimensions. Vessels . Inspection under UGS program Fuel Guide Pin Nuts Guide Pin Nuts Cracking, Internals extenor exam (Hex) Loss of Inspection Preload Program Water Chemistry Program ASME Section - - -

Xl IWB IWC .

IWD,IWF . - .,.

. Inservice - - ,, .-

4 . - .,.,

Inspection - ..

--. Program .--

Report No. 11 01403.401R1 95 PJctur&ThtegI14f Associates Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 96 of 121 Control Rod Shroud DescnpUon Fuel Plate Loss of nt ASME Section Not specifically Classmcation Not specifically Similarity Classification Requements si Comments This A No

- NONE Covered In Section Xl Assembly Alignment Fuel Preload XI IWB, IWC, listed listed component Inspection under IJGS program Fuel Plate Cap Screw Cap Screws IWD, IWF - would be exterior exam Inservice evaluated in Inspection the same Program manner as J other components -

Iikethefuel guide pins ,-*

which were dassifledas Incore Instrument lCl Instrument Changes in Reactor A Possibly These are Primary per VT-3 This Guide Tubes Guide Tubes Dimensions, Vessels Primary - primary for MRP-227-A for examination no component is lCI Instrument Guide including Cracking, Internals CEA shroud plants with later than 2 accessible Tube Extension Loss of Inspection assemblies CEA shrouds. refueling during the Assembly Material, Program where the outages from visual exam Reduction in Water guide tube the beginning of the exterior Fracture Chemistry is welded to of the license of the UGS Toughness Program the CEA renewal ASME Section shroud penod XI IWB, IWC, Palisades Subsequent IWO, Wi/F has ICI exams on ten Inservice guide tubes year interval Inspection that are also Program welded to 100% of tubes the shrouds in peripheral of These UGS assembly would shall be therefore be examined primary per MRP-227 Report No. 1101403.401R1 96

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 97 of 121 Core Shroud

=

Anchor Blocks Changes in Reactor CbSSrflCatlOfl Not specifically CsSrilCatIOfl Not specifically Similarity The anchor Classification RuementsSection XI Comments B Screened

- NONE Assembly - Tie to Core Dimensions, Vessels listed listed screw and out. The only Anchor Block Support Plate Cracking, Internals pins are plates Reduction in Inspection - screened examined are Fracture Program out per the shroud Toughness Water . MRP-191. plates ID Chemistry However, surface.

Program the anchor blockis more simfar tothe centenng plate (core shroud former plate) which is classified as B.

Core Shroud Core Shroud Changes in Reactor B NIA The former B -Screened Assembly Former Plates Dimensions, Vessels plates are out.

Centering Plate Cracking, Internals not Reduction in Inspection examined Fracture Program as the ID - -

Toughness Water surface of Chemistry the shroud Program plates are examined. -

Core Shroud Core Shroud Changes in Reactor C Primary N/A The core VT-3 exam no The current Assembly Plates Dimensions, Vessels shroud later than 2 Section Xl Core Shroud Plate Cracking, Internals assembly must refueling visual of the Reduction in Inspection be examined, outages from interior of the Fracture Program the beginning core barrel Toughness Water of the license would Chemistry renewal examine this Program period. Exam surface is to cover core side surfaces as indicated in Figures 4-25 and 4-26 of MRP-227.

Report No. 1101403.401R1 97 StnwturøJ Integrity AssoclaMs, Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 98 of 121 Component! Additional Aging Effect Aging Augmented Current Requiring Management Classification Classification Final Commodity Description Similarity Inspection Section XI Comments Management in MRP-191 in MRP-227 Classification Programs Requirements Exams Core Shroud Anchor Block Cracking, Reactor A N/A A No

- NONE Stated that Assembly to Core Plate Changes in Vessels Inspection its covered in Anchor Screw & Pin Dimensions. Internals the Section Reduction in Inspection xi program, Fracture Program but this bolt is Toughness, Water inaccessible Loss of Chemistry due to its Preload Program location ASME Section Xl IWB, IWC, IWD, IWF Inservice Inspection Program Core Shroud Barrel to Cracking, Reactor B Expansion N/A Barrel to Volumetric UT Visual exams Assembly Former Bolt Changes in Vessels Former Bolts with initial and currently Centering Screw & Dimensions, Internals are an subsequent being done Pin Reduction in Inspection expansion exams based during Fracture Program exam based on results of exterior of Toughness, Water on the results shroud bolt core barrel Loss of Chemistry of the shroud exams, exam. This Preload Program bolt exams would have to ASME Section 100% of bolts be expanded Xl IWB, IWC, with fluence to UT of all IWD, IWF greater than 3 the bolts Inservice dpa to be meeting the Inspection examined. fluence Program threshold.

Report No. 1101403.401 Ri 98 integrny Associates, 1nc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 99 of 121 Component! Aging Effect Aging Augmented Current Additional Classification Classification Final Commodit, Description Requiring Management Similarity Inspection Section XI Comments Management MRP.191 in MRP-227 Classification Programs Requirements Exams Core Shroud Shroud Bolts Cracking, Reactor B Primary N!A Shroud bolts Baseline liT Visual exams Assembly Changes in Vessels are Primary between 25 currently Positioning Screw Dimensions, Internals based on and 35 EFPY being done Reduction in Inspection MRP-227. with during interior Fracture Program subsequent of core barrel Toughness, Water exams after 10 exam. This Loss of Chemistry to 15 additional would have to Preload Program EFPY. Re- be expanded ASME Section examination to include UT Xl IWB, IWC, for high- of all bolts in IVtiD, IWF leakage core the interior of Inservice designs the core Inspection requires barrel. This Program continuing would also inspections on include the a ten-year anchor block interval, to core shroud bolts.

100% of accessible bolts to be examined in 2022 at approximately.

34.3 EFPY.

Report No. 1101403.401 Ri 99 ictur8 Integrity AssodaLos, Thc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 100 of 121 Component! Aging Effect Aging Augmented Current Additional Classification Classification Final Commodity Description Requiring Management Similarity Inspection Section XI Comments Management in MRP-191 in MRP-227 Classification Programs Requirements Exams Core Shroud Anchor Block Cracking, Reactor Not specifically NIA This bolt Primary based Baseline UT Visual exams Assembly to Core Shroud Changes in Vessels listed would be on the between 25 currently Shroud Bolt & Pin Dimensions, Internals similar to similarity to the and 35 EFPY being done Reduction in Inspection the positioning with during interior Fracture Program positioning screws (shroud subsequent of core barrel Toughness, Water screws bolts) exams after 10 exam. This Loss of Chemistry (shroud to 15 additional would have to Preload Program bolts). EFPY. Re- be expanded ASME Section examination to include UT Xl IWB, IWC, for high- of all bolts in IWO, IWF leakage core the interior of Inservice designs the core Inspection requires barrel. This Program continuing would also inspections on include the a ten-year anchor block interval, to core shroud bolts.

100% of accessible bolts to be examined.

Core Support Barrel Core Support Cracking Reactor B for Upper Expansion for N/A Additional core VT-3 exam These welds In Section Xl Assembly Barrel Vessels Cylinder, C for remaining core support barrel with initial and are examined program Core Support Barrel Internals Lower Cylinder support barrel welds to be subsequent during the Inspection welds examined exams based exterior core Program based on on plant barrel exam Water exam of CSB evaluation of in the current Chemistry upper flange SCC Section Xl Program weld, susceptibility exam and demonstration of remaining fatigue life.

Coverage determined by plant specific analysis.

Report No. 1101403.401R1 100

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 101 of 121 Componentl Additional Aging Effect Aging Augmented Current Requiring Management Classification Classification Final Commodity Description in MRP-191 Similarity Inspection Section Xl Comments Management Programs in MRP-227 Classification Requirements Exams Core Support Barrel Core Support Cracking, Reactor B Primary N/A The upper EVT-1 exam This weld is In Section Xl Assembly Barrel Upper Changes in Vessels flange weld is no later than 2 already program.

Core Support Barrel Flange Weld Dimensions Internals classified as refueling examined The exam Integral Upper Flange Inspection primary per outages from during the would have to Program MRP-227. the beginning exterior core be updated to Water of the license barrel exam, a EVT-1 Chemistry renewal It would have exam Program period, to be updated Subsequent to an EVT-1 exams on 10- exam.

year interval.

100% of the accessible surfaces of the upper flange weld.

Incore Instrument Guide Tube Loss of Reactor Not specifically NIA These A No

- NONE This is In Section Xl Guide Tubes Extension Plug Preload, Vessels listed components Inspection covered program Guide Tube Plug Cap Screw Changes in Internals are simtar during the Screw Dimensions, Inspection to visual exam Cracldng, Program components of the exterior Loss of Water in the UGS of the UGS.

Material, Chemistry of CEA Reduction in Program shroud Fracture ASME Section plants and Toughness XI IWB, IWC, also plants IWD, IWF with thimble lnsecvice tubes.

Inspection . . These types Program of components screened out in similar

.. - locations in

. wt. other plants Report No. 1101403.401R1 101 Stn,cture1 Integnly Associates, lnc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 102 of 121 Component! Aging Effect Aging Augmented Current Additional Classification Classification Final Commodity Description Mafl::ernent in MRP-191 Similarity Comments Mnt in MRP-227 Classification seo:xl Requirements Incore Instrument Guide Tube Changes in Reactor Not specifically NIA These A No

- NONE This is In Section Xl Guide Tubes Support Dimensions, Vessels listed components Inspection - covered program Guide Tube Bracket Bracket Cracking. Internals - are similar during the Loss of Inspection - to - visual exam Material, Program - components of the exterior Reduction in Water in the UGS of the UGS.

Fracture Chemistry of CEA Toughness Program shroud ASME Section plants and Xl IWB, IWC, also plants IWD, IWF with thimble Inservice tubes.

Inspection These types -

Program of components screened out in similar locations in other plants Incore Instrument Guide Tube Changes in Reactor Not specifically N/A These A- No NONE This is In Section Xl Guide Tubes Extension Dimensions, Vessels listed components Inspection covered program Guide Tube Plugs Assembly Cracking, Internals are similar during the Upper and Loss of Inspection to visual exam Lower Plugs Material, Program components of the exterior Reduction in Water in the UGS of the UGS.

Fracture Chemistry of CEA Toughness Program shroud ASME Section plants and Xl IWB. IWC, also plants IWD, IWF with thimble Inservice tubes.

Inspection These types Program of

. components screened outin

. similar locations in other plants -.

ReportNo. 1101403.401R1 102 Stn,cturaI Integrity Associates, lnc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 103 of 121 Componentl Aging Effect Aging Augmented Current Additional Classification Classification Final Commodity Description M:gemnt . .

Similarity Section Xl Comments Management in MRP-191 in MRP-227 Classification Re ire Incore Instrument Guide Tube Changes in Reactor Not specifically NIA These A No

- NONE This is In Section Xl Guide Tubes Support Dimensions, Vessels listed components Inspection covered program Guide Tube Support Brackets, Cracking, Internals are similar during the Spreaders and Loss of Inspection to visual exam Tubesheets Material, Program components of the extenor Reduction in Water in the UGS of the UGS.

Fracture Chemist,y of CEA Toughness Program shroud ASME Section plants and XI IWB, IWC, also plants IWO, IWF with thimble Inservice tubes.

Inspection These types Program of components -

screened out in similar

  • locations in

- other plants -2 Lower Internal Cap Screws Changes ri Reactor B Expansion Expansion per UT exam with Assembly for Core Dimensions, Vessels MRP-227 initial and Core Support Barrel Support Cracking, Internals subsequent Cap Screws Columns and Reduction in Inspection exam Manhole Cover Fracture Program frequencies Toughness Water dependent on Chemistry the results of Program core shroud bolt examinations 100% of core support column bolts with neutron fluence exposures > 3 dpa Report No. 1101403.401R1 103

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGiNG MANAGEMENT PROGRAM July 26, 2012 Page 104 of 121 Componenti Aging Effect Additional Classification Classification Final Commodity Description Management Similarity Inspection Section Xl Comments Management Programs in MRP191 in MRP.227 Classification Requirements Exams Lower Internal Snubber Lugs Changes in Reactor A N/A A No

- NONE This would be In Section XI Assembly on Core Dimensions, Vessels Inspection covered by program Core Support Barrel Support Barrel Cracking, Internals I the visual Snubber Lug and Vessel Loss of Inspection .

exam of the Wall Material, Program -

extenorof the Reduction in Water core barrel Fracture Chemistry Toughness Program ASME Section - .,.

XI IWB IWC

- IWD, IWF

- Inservice I

Inspection -

Program -

Lower Internal Core Support Changes in Reactor B for Columns, E. Core support VT-3 exam Assembly Column Dimensions, Vessels A for Beams column welds with initial and Core Support Column including Cracking Internals are expansion subsequent Column Base, Inspection per MRP-227- examinations Beams, Tie Program A based on plant Rods, and Water evaluation of Adjustable Chemistry SCC and Nuts Program susceptibility and demonstration of remaining fatigue life Exam coverage determined by plant-specific analysis Lower Internal See Above Changes in Reactor A for Beams A No

- NONE , No Exam Assembly Core Support Column Dimensions, Cracking, Vessels Internals Inspection  : - uired Support Beams and Reduction in Inspection Tie Rods Fracture Program -

ft Toughness Water Chemistry - -

Program -

Report No. 1101403.401R1 104 jstnwturai Integrity AssocIal8s, Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 105 of 121 o

Lower Internal

=1 Core Support

z: Management CIaSfiC5U0fl Similarity Classification Requements Soxl Comments Changes in Reactor C Primary Primary per If fatigue life EVT-1 exam Assembly Plate Dimensions, Vessels MRP-227-A. cannot be to be Core Support Plate Cracking, Internals demonstrated peiformed.

Reduction in Inspection by TLAA then Fracture Program EVT-1 exam Toughness Water no later than 2 Chemistry refueling Program outages from the beginning of the license renewal period.

Subsequent exam on 10-year interval.

Exam coverage to be defined by plant-specific fatigue analysis.

Upper Guide Fuel Alignment Changes in Reactor Not specifically Not specifically The A No

- NONE This would be In Section Xl Structure Plate Dimensions, Vessels listed listed instrument Inspection covered program Instrument Sleeve Instrument and Reduction in Internals sleeve during the Connector Fracture Inspection would be exterior exam Toughness, Program evaluated of the UGS Cracking, ASME Section similarly to Loss of Xl IWB, IWC. the other Material lVvO, IWF components Inservice in the UGS.

Inspection Therefore, it Program would have received an ,. ....

AperMRP 191 Report No. 1 101403.401R1 105

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 106 of 121 Component! Additional Aging Effect Aging Augmented Current Classification Classification Final Commodity Description Mamnt Similanty StXI Comments Management in MRP-191 in MRP-227 Classification Requirements Upper Guide UGS Spacer Cracking, Reactor Not specifically Not specifically Palisades Perform visual Examine by This Structure Shim Loss of Vessels listed listed has exam to check VT-3 to component Spacer Shim Material due Internals installed a for wear, determine if would be to wear Inspection spacer shim additional wear removed with Program due to wear has occurred, the UGS and ASME Section discovered examined Xl IWB, IWC, in the plant along with the IWO, IWF in the mid- Section Xl ISI Inservice 90s. program Inspection Program Upper Internal Shroud Grid Changes in Reactor A NIA NIA A No

- NONE This location Assembly Assembly Dimensions. Vessels Inspection would be Brace Grid Beam Cross Brace Cracking Intemals seen as part Grid Beams Inspection i) the existing Program .

exam of the Water exterior of the Chemistry UGS.

Program -

Upper Internal Cross Brace Changes in Reactor A NIA N!A A No

- NONE This location Assembly Socket Head Dimensions, Vessels Inspection would be Cross Brace Screw Cap Screws Cracking lntemals seen as part Inspection of the existing Program exam of the Water exteriorof the Chemistry UGS.

Program Upper Internal Gnd Ring Changes in Reactor A NIA NIA A No NONE This location Assembly Dimensions. Vessels - Inspection would be Shroud Grid Ring Cracking Internals seen as part

  • Inspection of the existing Program exam of the Water - extenor of the Chemistry ,. UGS.

Program Report No. 1 101403.401R1 106

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 107 of 121 Component! Additional Aging Effect Aging Augmented Current Classification Classification Final Commodity Maamnt .

Similarity SectionXl Comments Description in MRP-191 in MRP-227 Classification Management Requirements Upper Internal Fuel Alignment Changes in Reactor B Not Primary for N/A No Inspection NONE Wauld be In Section XI Assembly Plate Dimensions, Vessels Palisades as examined program Fuel Alignment Plate Cracking. Internals they do not during exam Loss of Inspection have full height for exterior of Material Program shroud plates UGS Water Chemistry Program ASME Section XI IWB, IWC, IWD. IWF Inservice Inspection Program Upper Internal Fuel Plate Changes in Reactor Not specifically Not specifically This is a A No

- NONE This would be In Section Xl Assembly Alignment Lug Dimensions, Vessels listed listed part of the Inspection examined program Fuel Plate Align Lug Cracking, Internals upper visually when Loss of Inspection internals the keyways Material Program and would are examined Water be in the for the UGS.

Chemistry similar area Program as the fuel ASME Section guide pin.

Xl IWB IWC Therefore it IWO IWF should be Inservice evaluated -

Inspection similarly.

Program Upper Internal Fuel Plate Changes in Reactor Not specifically Not specifically This is a A No

- NONE This would be In Section XI Assembly AlignmentCap Dimensions. Vessels listed listed partofthe Inspection examined program Fuel Plate Cap Screw Screws Cracking, Internals upper visually when Loss of Inspection internals the keyways Material Program and would are examined Water be in the for the UGS.

Chemistry similar area Program as the fuel ASME Section guide pin.

XI IWB, IWC, Therefore, it IWD, IWF should be Inservice evaluated -

Inspection similarly.

Program --.

Report No. 1101403.401R1 107

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 108 of 121 Componentl Additional Aging Effect Aging Augmented Current Classification Classification Final Commodity Description M:gemnt Similarity So X

5 l Comments Management in MRP-191 in MRP-227 Classification ReqUiments Upper Internal Fuel Plate Changes in Reactor A N!A A No

- NONE This would be In Section Xl Assembly Guide Pins Dimensions, Vessels Inspection examined program Fuel Plate Guide Pin Cracking, Internals visually when Loss of Inspection the exterior of Material Program theUGSis Water -

examined.

Chemistry Program ASME Section XI IWB. IWC, IWD. IWF Inservice Inspection Program Upper Internal Hold Down Loss of ASME Section Not specifically Not specifically These A No

- NONE In Section Xl Assembly Ring Plungers Material Xl IWB, IWC, listed listed components Inspection program.

Hoiddown Ring including IWO. IWF are similar However they Plunger washers and lnsecvice to the hold would be jack screws Inspection down ring removed Program and are during the located in removal of the same the UGS and area, subsequently Therefore, examined they would accordingly be similarly evaluated.

Upper Internal Hold Down Loss of ASME Section Not specifically Not specifically These A No

- NONE In Section Xl Assembly Ring Straps Material Xl IWB, IWC, listed listed components Inspection program.

Holddown Ring Strap and Jack IWO, IWF are simar However they Blocks lnservice to the hold would be Inspection down nng removed Program and are during the located in removal of the same the UGS and area subsequently Therefore, - examined they would accordingly.

be similarly

,,, evaluated.

Report No. 1101403.401R1 108

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 109 of 121 Component!

.. Aging Effect Aging Augmented Current Additional Classification Classification F!nal Commodity Requiring Management Similarity Inspection Section Xl Comments Descnption in MRP-191 in MRP-227 Classification Management Programs Requirements Exams Upper Internal Hold Down Loss of ASME Section A NIA N!A A No

- NONE In Section Xl Assembly Ring Matenal XI IWB, IWC, Inspection program.

Hoiddown Ring IWO, IWF However they Inservice would be Inspection removed Program - -

during the removal of the UGSand sUbseqUeflhly

- examined accordingly Report No. 1101403.401R1 109

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 110 of 121 ATTACHMENT C PLANNED SCOPE OF EXISTING AND AUGMENTED REACTOR VESSEL INTERNALS EXAMINATIONS Report No. 1101403.401 Ri 110 Associates, Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 111 of 121 Component Item Program Start Frequency Expansion Effect Method Exam Extent Items Lower Internals Alignment Keys and Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core Fastening Devices XI surfaces Barrel) 00 Lower Internals Alignment Keys and Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core Fastening Devices XI surfaces Barrel) 90° Lower Internals Alignment Keys and Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core Fastening Devices Xl surfaces Barrel) 180° Lower Internals Alignment Keys and Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core Fastening Devices XI surfaces Barrel) 270° Lower Internals Lifting Rig Hookup Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core and Pin Attachment XI surfaces Barrel)

Lower Internals Upper Barrel to Section Existing 10 years N/A VT-3 Accessible N/A

- Exterior (Core Flange Weld and XI weld surface Barrel) Welded Gussets, 00 (O.D.)

through 360° Lower Internals Outlet Nozzle and Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core Mating Interface XI surfaces Barrel) Surfaces, @ 00 &

180° Report No. 1101403.401R1 111 StnicturafhitegrItyAssooiates,Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL iNTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 112 of 121 Component Item Expa:sion Program Start Frequency Effect Method Exam Extent Lower Internals Upper Barrel to Section Existing 10 years N/A VT-3 Accessible N/A

- Exterior (Core Lower Barrel Shell XI weld surface Barrel) Weld, 0° through (O.D.)

360° Lower Internals Exterior Shroud and Section Existing 10 years N/A VT-3 All exterior N/A

- Exterior (Core Fastening Devices, 0° XI accessible Barrel) through 360° surfaces Lower Internals Core Stabilizer Keys Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core on Bottom of Core XI surfaces Barrel) Barrel, 0° through 360° Lower Internals Alignment Pins Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core (located beneath XI surfaces Barrel) nozzles) @ 0° Lower Internals Alignment Pins Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core (located beneath XI surfaces Barrel) nozzles) @ 90° Lower Internals Alignment Pins Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core (located beneath XI surfaces Barrel) nozzles) @ 180° Lower Internals Alignment Pins Section Existing 10 years N/A VT-3 All accessible N/A

- Exterior (Core (located beneath XI surfaces ReportNo. l101403.401R1 112

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 113 of 121 Component Item Program Start Expansion Frequency Effect Method Exam Extent Items Barrel) nozzles) @ 2700 Lower Internals Alignment Keys and Section Existing 10 years N/A VT-3 All accessible N/A

- Interior (Core Top Ledge of Core XI surfaces Barrel) Shroud 0° through 360° Lower Internals Core Shroud Bolting Section Existing 10 years N/A VT-3 All accessible N/A

- Interior (Core 0° through 360° Xl bolts Barrel)

Lower Internals Top Ledge of Core Section Existing 10 years N/A VT-3 All accessible N/A

- Interior (Core Shroud 0° through XI surfaces Barrel) 360° Lower Internals Lower Core Plate, Section Existing 10 years N/A VT-3 All accessible N/A

- Interior (Core Locking Devices, XI surfaces Barrel) Access Port and Bottom Row of Core Shroud Bolting Lower Internals Top Section of Core Section Existing 10 years N/A VT-3 All accessible N/A

- Interior (Core Shroud, Keys, and XI surfaces Barrel) Ledge Report No. 1101403.401R1 113

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 114 of 121 Component Item Program Expansion Start Frequency Effect Method Exam Extent Upper Guide Exterior Surfaces of Section Existing 10 years N/A VT-3 All accessible N/A Structure - the Upper Guide XI areas and Exterior Structure, surfaces Thermocouples, Connectors, Guide Tubes, Fastening Devices, Flow Holes, Keyways and Peripheral Scan. Fuel Alignment Guide Pins on Bottom. 0° through 3600 Reactor Vessel - Reactor Vessel to Section Existing 10 years N/A VT-3 All accessible N/A Interior Head Seal Surface XI surfaces Area 0° through 360° Reactor Vessel - Reactor Vessel to Section Existing 10 years N/A VT-3 All accessible N/A Interior Core Barrel Seal XI surfaces Surface Area and Keyways 0° through 360° Reactor Vessel - Outlet and Inlet Section Existing 10 years N/A VT-3 N/A Interior Nozzle Surfaces and XI Outlet Nozzle Mating Interface Surfaces 0°,

60°, 120°, 180°, 240°,

and 300° Report No. 1101403.401R1 114

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 115 of 121 Component Item Program Expansion Start Frequency Effect Method Exam Extent Items Reactor Vessel - Surveillance Capsule Section Existing 10 years N/A VT-3 N/A Interior Holder @ 80° XI Reactor Vessel - Surveillance Capsule Section Existing 10 years N/A VT-3 N/A Interior Holder @ 100° XI Reactor Vessel - Surveillance Capsule Section Existing 10 years N/A VT-3 N/A Interior Holder @ 1100 XI Reactor Vessel Surveillance Capsule Section Existing 10 years N/A VT-3 N/A Interior Holder @ 260° XI Reactor Vessel - Surveillance Capsule Section Existing 10 years N/A VT-3 N/A Interior Holder @ 280° XI Reactor Vessel - Surveillance Capsule Section Existing 10 years N/A VT-3 N/A Interior Holder @ 290° XI Reactor Vessel - Core Stabilizer Lugs Section Existing 10 years N/A VT-3 All accessible N/A Interior 0° XI surfaces Reactor Vessel - Core Stabilizer Lugs Section Existing 10 years N/A VT-3 All accessible N/A Interior 60° XI surfaces Reactor Vessel - Core Stabilizer Lugs Section Existing 10 years N/A VT-3 All accessible N/A Interior 120° XI surfaces Report No. 1 101403.401R1 115 Inc.

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 116 of 121 Component Item Program Start Expansion Frequency Effect Method Exam Extent Items Reactor Vessel - Core Stabilizer Lugs Section Existing 10 years N/A VT-3 All accessible N/A Interior 180° XI surfaces Reactor Vessel - Core Stabilizer Lugs Section Existing 10 years N/A VT-3 All accessible N/A Interior 240° XI surfaces Reactor Vessel - Core Stabilizer Lugs Section Existing 10 years N/A VT-3 All accessible N/A Interior 300° XI surfaces Reactor Vessel - Core Stop Lugs and Section Existing 10 years N/A VT-3 All accessible N/A Interior Flow Skirt 0° through XI surfaces 360° Reactor Vessel - General Scan of Section Existing 10 years N/A VT-3 N/A Interior Bottom Head for XI Debris 0° through 360° Report No. 1101403.401R1 116

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. I AGING MANAGEMENT PROGRAM July 26, 2012 Page 117 of 121 Component Item Program Start Expansion Frequency Effect Method Exam Extent Items Core Shroud Core Shroud Bolts MRP- Baseline examination Subsequent Cracking Volumetric 100% of Core Assembly 227-A between 25 and 35 EFPY. examination (IASCC, UT accessible support (Bolted) Currently planned for after 10 to 15 Fatigue) bolts, or as column 2022 @ 34.3 EFPY additional supported by bolts, EFPY to plant-specific barrel-confirm justification. shroud bolts stability of Heads are bolting pattern, accessible Re-examination from the core for high- side. UT leakage core accessibility designs may be requires affected by continuing complexity of inspections on head and a ten-year locking device interval, designs.

Report No. 1101403.401 Ri 5IFJILIIaI IntegrIIr Associates, Inc.

117

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 118 of 121 Component Item Program Expansion Start Frequency Effect Method Exam Extent Items Core Shroud Assembly MRP- Examination no later than Subsequent Distortion VT-3 Core side None Assembly 227-A 2 refueling outages from examinations (Void surfaces as (Bolted) the beginning of the on a ten-year Swelling), indicated.

license renewal period, interval, including:

Currently planned for

  • Abnormal 2013. interaction with fuel assemblies
  • Gaps along high fluence shroud plate joints
  • Vertical displacement of shroud plates near high fluence joint Core Support Upper (core support MRP- Examination no later than Subsequent Cracking EVT-1 100% of the Remaining Barrel barrel) flange weld 227-A 2 refueling outages from examinations (SCC) accessible core barrel Assembly the beginning of the on a ten-year surfaces of the assembly license renewal period, interval, upper flange welds, core Currently planned for weld. support 2013. column welds Report No. 1101403.401R1 118

PALISADES NUCLEAR PLANT CEP-RVI-.PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 119 of 121 Component Item Program Start Expansion Frequency Effect Method Exam Extent Core Support Lower cylinder girth MRP- Examination no later than Subsequent Cracking EVT-1 100% of the Lower Barrel welds 227-A 2 refueling outages from examinations (SCC, accessible cylinder Assembly the beginning of the on a ten-year IASCC), and surfaces of the axial welds license renewal period. interval, neutron upper flange Currently planned for embrittlement weld.

2013.

Lower Support Core Support Plate MRP- Examination no later than Subsequent Cracking EVT-1 100% of the None Structure 227-A 2 refueling outages from examinations (Fatigue) accessible the beginning of the on a ten-year surface of the license renewal period. interval, core support Currently planned for plate, or 2013. future exam coverage to be defined by plant-specific fatigue analysis.

Report No. 1 101403.401R1 119 YJII9T1Iy Associates, Inc

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 120 of 121 Component Item Expansion Program Start Frequency Effect Method Exam Extent Items Lower Support Core Support Column IvERP- Examination no later than Subsequent Cracking EVT-1 100% of the Structure Welds 227-A 2 refueling outages from examinations (SCC, accessible the beginning of the on a ten-year IASCC, surfaces of the license renewal period, interval. Fatigue core support Currently planned for included column welds.

2013. damaged or fractured material)

Control Element ICI Instrument Guide MRP- Examination no later than Subsequent Cracking VT-3 100% of tubes Remaining Assembly Tubes 227-A 2 refueling outages from examinations (SCC, in peripheral ICI the beginning of the on a ten-year Fatigue) that control rod instrument (Designated as license renewal period, interval, results in shroud guide tubes Control Rod Currently planned for missing assemblies within the Shroud 2013. supports or (i.e., those control rod Assembly for separation at adjacent to the shroud Palisades) the welded perimeter of assemblies joint between the fuel the tubes and alignment supports plate).

ReportNo. 1101403.401R1 120

PALISADES NUCLEAR PLANT CEP-RVI-PNP REACTOR VESSEL INTERNALS Rev. 1 AGING MANAGEMENT PROGRAM July 26, 2012 Page 121 of 121 Component Item Program Start Expansion Frequency Effect Method Exam Extent Core Shroud Guide Lugs and Section Existing Subsequent Loss of VT-3 General N/A Assembly Guide Lug Inserts XI examinations Material condition and Bolts on a ten-year (wear) exam for interval, detection of excessive or asymmetrical wear.

Core Barrel Upper Flange Section Existing Subsequent Loss of VT-3 Area of the N/A Assembly XI examinations Material upper flange on a ten-year (wear) potentially interval, susceptible to wear.

Upper Guide Spacer Shim Plant This component is plant Subsequent Loss of VT-3 Area of the N/A Structure Specific specific due to previously examinations Material spacer shim observed wear on a ten-year (wear) susceptible to interval, wear.

Report No. 1101403.401 Ri 121 ASSOWWeS, h7c.

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