Text

Submittal of Inspection Plan for Reactor Internals
	ML12108A029
Person / Time
Site:	Crane
Issue date:	04/16/2012
From:	Jesse M; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	TMI-12-069
	Download: ML12108A029 (91)
	v • d • e

10 CFR Part 54 TMI-12-069 April 16, 2012 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Submittal of Inspection Plan for Reactor Internals

Reference:

NUREG-1928, "Safety Evaluation Report Related to the License Renewal of Three Mile Island Nuclear Station, Unit 1," dated October 2009 Appendix A ("Long Term Commitments for License Renewal of TMI-1") of the referenced U.S. Nuclear Regulatory Commission Safety Evaluation Report contains commitments associated with the license renewal of the Three Mile Island Nuclear Station (TMI), Unit 1.

Commitment No. 36 commits to the following:

"The PWR Vessel Internals Program will commit to the following activities: 1. Participate in the industry programs for investigating and managing aging effects on reactor internals.

2. Evaluate and implement the results of the industry programs as applicable to the reactor internals. 3. Upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval."

The period of extended operation begins April 20, 2014. Accordingly, attached is the requested inspection plan which is being submitted not less than 24 months before entering the period of extended operation.

The "Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals" (Attachment 1) is based upon the above commitment and MRP-227-A, "Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A)," dated December 2011. MRP-227-A includes the United States Nuclear Regulatory Commission (NRC)

Safety Evaluation Report accepting MRP-227 for use as an acceptable Aging Management Program (AMP) for PWR Reactor Vessel Internals. The NRC SER includes licensee specific actions which shall be addressed by licensees choosing to implement MRP-227-A. The inspection plan discussed below is complete. However, changes to the plan may result from

Submittal of Inspection Plan for Reactor Internals April 16, 2012 Page 2 ongoing discussions between the Materials Reliability Program (MRP) and the NRC in resolving final MRP-227-A inspection and evaluation methodologies. The final evaluation methodologies are scheduled to be reviewed and accepted by the NRC in 2012 and will allow final definition of work scope and schedule for presentation to the NRC for the three ApplicanVLicensee Action Items defined below. A brief summary of the ongoing actions is provided below with additional detail provided in Appendix 0 of the enclosed inspection plan:

1.

ApplicanVLicensee Action Item 2 (Table 0-2), based upon Section 4.2.2 of the NRC Safety Evaluation Report (SER), requires a licensee to identify which Reactor Vessel Internals (RVI) components are within the scope of license renewal. If a licensee identifies additional components within the scope of license renewal that are not currently included in MRP-227-A then the licensee shall propose necessary inspection requirements beyond the requirements of MRP-227-A. Exelon has identified that the RVI vent valve locking device should be further reviewed with MRP-227-A methodology to determine the impact on the RVI Aging Management Program (AMP) and the enclosed plan. TMI is currently working with the Pressurized Water Reactor Owners Group (PWROG) to address the action item. Exelon will submit an update of the PWROG progress in evaluating this item and a schedule for showing when Exelon will submit the results of an evaluation by April 19, 2013.

2.

ApplicanVLicensee Action Item 6 (Table 0-2), based upon Section 4.2.6 of the SER, requires a licensee to justify the acceptability of certain inaccessible components that Table 4-4 of MRP-227-A identifies as expansion components. The SER action requires that a licensee submit an analysis of the acceptability of these components for continued service or a schedule for replacement of the subject components with the NRC submittal documenting the intent to implement the requirements of MRP-227-A. TMI is currently working with the Pressurized Water Reactor Owners Group (PWROG) to address the action item. Exelon will submit an update of the PWROG progress in evaluating these components and a schedule for showing when Exelon will submit an evaluation for continued service or a schedule for replacement by April 19, 2013.

3.

ApplicanVLicensee Action Item 7 (Table 0-2), based upon Section 4.2.7 of the SER, requires that the licensee develop plant-specific analyses to demonstrate that there is not a loss of functionality of the Incore Monitoring Instrumentation (IMI) guide tube assembly spiders and Control Rod Guide Tube (CRGT) spacer castings due to loss of fracture toughness. TMI is currently working with the PWROG to address the action item. Exelon will submit an update of the PWROG progress in evaluating these components and a schedule for showing when Exelon will submit an evaluation for continued service or a schedule for replacement by April 19, 2013. contains three new commitments that correspond to the open actions described above. Exelon considers that license renewal commitment No. 36 is satisfied by virtue of the inspection plan and the new commitments made in Attachment 2. There are no other regulatory commitments contained in this letter.

Section 2.4 of MRP-227-A identifies certain general assumptions for plant operation that are based on core design, operation as a base load plant, and design changes that could have an impact on the reactor vessel internals. MRP-229, "Materials Reliability Program: Functionality Analysis for B&W Representative PWR Internals," Revision 3, was a base document used in

Submittal of Inspection Plan for Reactor Internals April 16, 2012 Page 3 development of MRp*227-A. MRp*229 identified that the effect of power uprates has not been considered in the functionality analysis. The impact of any future power uprates on this RVI Inspection Plan will be addressed, as appropriate, in the LAR application to implement the proposed uprate.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765*5510.

Respectfully, Michael D. Jesse Director - Licensing RegUlatory Affairs Exelon Generation Company, LLC Attachments: 1)

Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals 2)

Summary of Commitments cc: Regional Administrator, Region I, USNRC USNRC Senior Resident Inspector, TMI USNRC Project Manager, [TMIJ USNRC Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals

Inspection Plan for the Three Mile Island Unit 1 Reactor Vessel Internals (AREVA document 77-2952-001)

By Ron Finnin, AREVA NP, Inc.

Sarah Davidsaver, AREVA NP, Inc.

And Stephen Leshnoff, Exelon Nuclear Gene Navratil, Exelon Nuclear Prepared for Exelon Nuclear Three Mile Island Unit 1 Nuclear Station 200 Exelon Way Kennett Square, PA 19348 by AREVA NP, Inc.

3315 Old Forest Road P.O. Box 10935 Lynchburg, VA 24506-0935 ANP-2952 Revision 001 March 2012

Revis ion Date Section AI\\lI-"-,,~~:JL. Rev. 001 Nature of Changes Changes 000 November 20 I0 001 March 2012 General TOC LOAA 1.0 2.0 4.1 4.2 4.3 5.0 6.0 7.0 Appen dices Original Release the elements ofthe NRC SER for MRP-227. and ch,m~~es made to MRP-227-

~

A. Details below.

Word searched and added "-A" to most occurrences ofMRP-227. Historical reference to the OJ submitted MRP-227 were left without the "-A".

Updated to retlect other in the document.

List of Acronyms and Abbreviations to retlect other in the document.

.M 16A of NUREG-180 I Rev. 2 for the program comparison as this is required R for MRP-227-A. Minor wording h,maN "current" to "original" as the currcnt license expiration date is now 2034.

Changed AMP to management review" to correctly retlect the subject.

Added discussion ofTLAA in the TMI-l LRA for completeness.

Changed Table 2-1 to reHect the LR commitment in Appendix A of the LRA (UFSAR Supplement) rather than Appendix B.

Revised MRP-227 history to include issuance of the SER and submittal ofMRP-227-A.

Modified paragraph about Appendix A, now App E, to better retlect what's in App E.

Added MRP-231 and MRP-189 as references.

Used MRP-227-A intended functions ofthe internals. Changed title of figure trom B&W internals to TMI-I internals.

Reworded slightly. Brought some info from 4.2 to 4.1 and combined these sections.

Deleted info about the reactor vessel inspections in the ASME code. this is not pertinent to the internals inspections.

All TMI-l recategorizations have been incorporated (or eliminated) in MRP-227-A, this section has been deleted in its entirety. Renumbered sections as some sections were deleted.

Added discussion of fabrication records search to meet Applicant/Licensee Action Item #4.

Deleted the discussion of aging etfects for the UCB and LCB bolts and bolt locking devices. This was a TMI re-categorization but has been incorporated into MRP-227-A and thus is no longer are-categorization.

Note this used to be 4.3. Added sentence that was in 4.3.2 as it seemed out of place there.

Revised to say current guidance is in MRP-227-A, rather than saying it culminated in the submittalofMRP-227.

Under Mandatory and Needed: reworded to reHect current status of MRP-227-A.

Changed most recent TMI-I fuel cycle from cycle 18 to cycle 19. Updated reference to the cycle 19 report. Info provided by Exelon.

Added assumptions trom MRP-190 and MRP-229 to those trom MRP-227 to complete Applicant/Licensee Action Item # I.

Note this used to be 4.4. Simplified discussion, deleted redundant information.

This entire section was re-written to include a typical License Renewal 10-element program comparison as requested by the NRC SER for MRP-227-A and comparison to NUREG-180 I. Revision 2.

Updated summary to retlect the other changes throughout this document.

Updated References Moved Appendix A to Appendix E. The TMI-I Primary components, Expansion components, and Examination Acceptance/Expansion Criteria are now Appendices A, B and C.

Page iii

Revis ion Date Section Changes Word searched the document fc)r..

and revised references as needed.

/\\pp A App B AppC App D App E New AppD App E The previous App A has been moved to App E.

Previous App B is now App A.

Revised Appendix number and title; this is now the TMI-l Primary Components rather than a reprint of the MRP-227-A Primary Components.

Made the changes in MRP-227-A Deleted the items that were not applicable to TMI-I Previous App C is now App B. Revised Appendix number and the title; this is now the TMI-I Expansion Components rather than a reprint of the MRP-227-A Expansion Components Made the G

in MRP-227-A Deleted the items that were not applicable to TMI-l Previous App D is now App C.

Revised Appendix number and the title; this is now the I'M I-1 Examination Acceptance and Expansion Criteria rather than a reprint ofthe MRP-227-A Examination Acceptance and Expansion Criteria Table Made the changes in MRP-227-A There are no TMI-l specific changes.

Previous App E was a discussion ofI'MI-I SpecifIC re-categorizations. MRP-227-A has incorporated all of the TMI-I re-categorizations, and this appendix is no longer necessary.

It has been deleted.

Added new appendix D to document the disposition of the Topical Report Conditions and Applicant/Licensee Action Items from the NRC SER contained in MRP-227-A.

The previous Appendix A is now Appendix E. Revised introduction to reflect what was requested by the NRC SER for MRP-227-A. Deleted the GALL comparison (the last three columns in Rev. 0). Made the location and disposition their own columns.

Pageiv

1.0 2.0 2.1 2.2 2.3 2.4 3.0 4.0 4.1 ANI~-L~~~L. Rev. 001 Table of Contents Page NATURE OF CHANGES ii LIST OF ACRONYMS AND ABBREViATIONS VII INTRODUCTION 1-1 BACKGROUND 2-1 TMI-1 License Renewal Background 2-1 TMI-1 RV Internals AMR/lndustry Program Background 2-2 TMI-1 RV Internals AMP Intent 2-3 TMI-1 RV Internals Background 2-3 PROGRAM OWNER 3-1 INDUSTRY AND TMI-1 PROGRAMS AND ACTIVITIES 4-1 TMI-1 Programs and Activities 4-1 4.1.1 ASME B&PV Code Section XI lSI Program 4-1 4.1.2 Primary Water Chemistry Program 4-1 4.1.3 Plant Technical Specifications 4-1 4.1.4 Time Limited Aging Analyses 4-1 4.1.5 Fabrication Records Searches 4-2 4.1.6 UCB and LCB Bolt Analysis 4-2 4.1.7 Past RV Internals Inspections 4-2 4.1.8 Fuel/Baffle Interaction Investigation 4-3 4.2 Other Industry Programs and Activities 4-3 4.2.1 B&WOG Generic License Renewal Project Report (BAW-2248A) 4-3 4.2.2 MRP-227-A 4-3 4.2.3 Joint Owner's Baffle Bolt Program 4-6 4.2.4 PWROG and EPRIIMRP 4-6 4.3 Conclusions of Section 4.0 4-6 5.0 TMI-1 RV INTERNALS AMP ATTRIBUTE EVALUATION 5-1 5.1 AMP Element 1 - Scope of Program 5-1 5.1.1 NUREG-1801, XI.M16A Scope of Program 5-1 5.1.2 TMI-1 Scope of Program 5-2 5.1.3 Conclusion 5-2 5.2 AMP Element 2 - Preventative Actions 5-2 5.2.1 NUREG-1801, XI.M16A Preventive Actions 5-2 5.2.2 TMI-1 Preventive Actions 5-2 5.2.3 Conclusion 5-2 5.3 AMP Element 3 - Parameters Monitoredllnspected 5-2 5.3.1 NUREG-1801, XI.M16A Parameters Monitored/Inspected 5-2 5.3.2 TMI-J Parameters Monitoredllnspected 5-3 5.3.3 Conclusion 5-3 5.4 AMP Element 4 - Detection of Aging Effects 5-3 Page v

M'--.'07.JL Rev. 001 1

NUREG-1801, XI.M16A Detection of Aging Effects TMI-1 Detection of Aging 5.4.3 Conclusion 5-4 5.5 AMP Element 5 Monitoring and Trending 5-5 1

NUREG-1801, XI.M16A Monitoring and Trending 5-5 TMI-1 Monitoring and Trending 5-5 5.5.3 Conclusion 5.6 AMP Element 6 - Acceptance Criteria 5-5 5.6.1 NUREG-1801, XI.M16A Acceptance Criteria 5-5 5.6.2 TMI-1 Acceptance Criteria 5-6 5.6.3 Conclusion 5-6 5.7 AMP Element 7 - Corrective Actions 5-6 5.7.1 NUREG-1801 XI.M16A Corrective Actions 5-6 5.7.2 TMI-1 Corrective Actions 5-6 5.7.3 Conclusion 5-6 5.8 AMP Element 8 Confirmation Process 5-7 5.8.1 NUREG-1801, XI.M16A Confirmation Process 5-7 5.8.2 TMI-1 Confirmation Process 5-7 5.8.3 Conclusion 5-7 5.9 AMP Element 9 - Administrative Controls 5-7 5.9.1 NUREG-1801 XI.M16A Administrative Controls 5-7 5.9.2 TMI-1 Administrative Controls 5-7 5.9.3 Conclusion 5-7 5.10 AMP Element 10 - Operating Experience 5-7 5.10.1 NUREG-1801 XI.M16A Operating Experience 5-7 5.10.2 TMI-1 Operating Experience 5-8 5.10.3 Conclusion 5-8 6.0

SUMMARY

AND CONCLUSIONS 6-1

7.0 REFERENCES

7-1 APPENDIX A: TMI-1 PRIMARY COMPONENTS A-1 APPENDIX B: TMI-1 EXPANSION COMPONENTS B-1 APPENDIX C: TMI-1 EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA C-1 APPENDIX 0: TMI-1 RESPONSE TO THE TOPICAL REPORT CONDITIONS AND THE APPLICANT/LICENSEE ACTION ITEMS FROM THE NRC SER, REVISION 1, FOR MRP-227, REVISION 0 0-1 APPENDIX E: LRA AMR AND INDUSTRY PROGRAM COMPARISON E-1 Page vi

List of Acronyms and Abbreviations AMP Management Program AMR Review ASME of Mechanical Engirleers B&PV Boiler and Pressure Vessel B&W Babcock and Wilcox B&WOG B&W Owner's Group CAP Corrective Action Program CASS Cast Austenitic Stainless Steel CFR ~ Code of Federal Regulations CLB Current Licensing Basis CRGT Control Rod Guide Tube CSA Core Support Assembly CSS Core Support Shield EPRI Electric Power Research Institute Exelon - Exelon Nuclear FD - Flow Distributor FMECA - Failure, Modes, Effects, and Criticality Analysis FSER - Final Safety Evaluation Report GALL - Generic Aging Lessons Learned (NUREG-180 I)

GLRP Generic License Renewal Program I&E Guidelines - Inspection and Evaluation Guidelines (MRP-227-A)

IASCC - Irradiation-Assisted Stress Corrosion Cracking IE Irradiation Embrittlement IMI - Incore Monitoring Instrumentation lSI - In-Service Inspection JOBB - Joint Owners' Baffle Bolt (Program)

LCB - Lower Core Barrel LR - License Renewal LRA License Renewal Application LRAAI - License Renewal Applicant Action Item LTS - Lower Thermal Shield MRP - Materials Reliability Program MUR - Measurement Uncertainty Recapture NDE Non-Destructive Examination NRC - Nuclear Regulatory Commission Rev, 001 Page vii

OE PWR Pressurized Water Reactor PWROG Pressurized Water Reactor Owners Group QA Quality Assurance QATR Quality Assurance Topical Report RFO RV Reactor Vessel SCC.~ Stress Corrosion Cracking SER Safety Evaluation Report TLAA Time-Limited Analysis TE - Thermal Embrittlcment TMI-I Three Mile Island Unit I UCB Upper Core Barrel UFSAR - Updated Final Safety Analysis Report U.S.

United States UT Ultrasonic Testing (Nondestructive Examination Technique)

UTS - Upper Thermal Shield VT Visual Examination Rev. 001 Page viii

Rev. 001

1.0 INTRODUCTION

The purpose of this is to document the Three Mile Island Unit I (TMI-I) Reactor Vessel (RV) Internals Inspection Plan for submittal to the United States (U.S.) Nuclear Regulatory Commission (NRC). This report provides a of the TMI-I RV Internals Inspection Plan as it relates to the management of effects consistent with previous commitments. The TMI-I RV Internals Inspection Plan is based onMRP-227-A "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines"lli.

The Reactor Vessel Internals Inspection Plan is a part of the TMI-I Reactor Vesscl Internals Management Program (AMP). The overall program consists of not only the RV Internals inspections described in this document, but includes other TMI-I and Industry Programs and Activities as discussed in Section 4.0 of this inspection plan. Section 5.0 of this plan compares the TMI-I AMP to the 10-element program in Section XIMl6A ofNUREG-1801, Revision 2 "Generic Aging Lessons Learned (GALL)" reportl21 as required by Applicant/Licensee Action item #8 of the NRC Safety Evaluation Report (SER) for MRP-227. Although this comparison is made to the NUREG-1801, Revision 2 program as required by the SER, TMI-I remains committed to NUREG-180 I, Revision I as addressed in their License Renewal (LR) Application (LRA) and SER.

This TMI-I RV Internals Inspection Plan contains a discussion of the background of the Babcock and Wilcox (B&W)-designed plant RV Internals programs, first sponsored by the utilities through the B&W Owner's Group (B&WOG) and later by the PWR Owners group (PWROG), and submitted to the NRC through the Electric Power Research Institute (EPRI) PWR Materials Reliability Program (MRP). The TMI-I RV Internals Inspection Plan also contains a discussion of operational experience, and relevant TMI-I and industry programs and activities.

The TMI-I RV Internals AMP will include this TMI-I RV Internals Inspection Plan and wiII demonstrate that the program adequately manages the etfects of aging for RV Internals components and establishes the basis for providing reasonable assurance that the RV Internals componcnts remain functional through the TMI-I LR period of extended operation.

Page 1-1

AI\\lI-'-':::l::I:J.,t.. Rev. 001

2.0 BACKGROUND

2.1 TMI*1 License Renewal Background Bv letter dated 8.

, LLC submitted the LRA for TMI-l in accordance with Title 10. Part of the Code of (10 CFR IJ] Through the LRA, AmerGen Company, LLC (Exelon Nuclear, herein Exelon) requested the NRC renew the operating license for TMI-I number DPR-50) for a period 01'20 years beyond the original expiration of midnight April 19,2014. The renewed license was issued by the NRC on October 2009. 141 The SER NUREG-192815]documented the technical review of the TMI-l LRA by the NRC Staff.

Section 3.1 ofTMI-1 's LRAI61discusses the RV Internals management review (AMR) for LR. The components that are subject to AMR have been identified in accordance with the requirements of 10 CFR 54.4.

The AMPs selected to manage eHects for the RV Internals are identified in Section 3.1.2.1.3 and Table 3.1.2-3 of the LRA and include time-limited analyses (TLAAs) and Water Chemistry. A description of the water chemistry program is provided in Appendix B of the LRA; while the TLAAs are described in Section 4 and Appendix A ofthe LRA. In the TMI-I LRA, Exelon committed to (1) participate in industry programs for investigating and managing effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3) upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

Section 3.1.3 ofNUREG-1928[5 1 concludes that Exelon has provided suHicient information to demonstrate that the efIects of aging for the RV Internals components, within the scope of LR and subject to an AMR, will be adequately managed so that the intended functions will be maintained consistent with the current licensing basis (CLB) for the period of extended operation, as required by 10 CFR 54.21 (a)(3).

Section 4.9 ofNUREG-I928[5] identifies that the TLAAs from the LRA were reviewed. The SER concludes the identified TLAAs associated with RV Internals for TMI-I comply with the requirements of 10 CFR 54.21(c).

The three TLAAs for the RV internals in the TMI-I LRA include low cycle fatigue, high cycle fatigue, and neutron embrittlement. Low cycle fatigue is managed by the Fatigue Monitoring Program (LRA Section 4.3.4).

High cycle fatigue (I1ow induced vibration) is projected through the period of extended operation (LRA Section 4.3.5). Neutron embrittlement is managed by the PWR Vessel Internals Program and neutron embrittlement is one of the aging effects evaluated in MRP-227-A (see Section 3.2.6 ofMRP-227-A).

Section 6 of NUREG-I92815] concludes the staff determined that the requirements of 10 CFR 54.29(£1) were met by the TMI-I LRA.

Table 2-1 demonstrates how the TMI-I RV Internals LR commitments have been fulfilled.

Page 2-1

AI\\I/-'-L:::1::l,t.. Rev. 001 Table 2-1: TMI-1 RV Internals LRA Commitment Resolutions Commitment reference location TlVlI-l

,1 61 Appendix A, License Renewal Commitment 36 Commitment/Action Items TMI-l commits to the fiJllowing activities for the PWR Vessel Internals program:

Par*ticipate in the industry programs for andmaJaagrng effects on reactor internals.

Evaluate and implement the results of the industry programs as applicable to the reactor internals.

Upon completion of these programs, but not less than 24 months before entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval.

Fulfilled, as described in reference location Exelon currently participates and will continue to participate in the industry programs for investigating and effects on reactor internals during TMI-l current and extended license periods.

The second and third bullets are being fulfilled by the development and submittal of this TMI-I RV Internals Inspection Plan (based on MRP-227-A,) to the NRC for review and approval.

2.2 TMI-1 RV Internals AMRllndustry Program Background In the TMI-I LRA, ExeIon committed to submitting an inspection plan for the RV Internals components; this document is that inspection plan.

The initial industry work performed, which supports the TMI-l RV Internals Inspection Plan, included an AMR documented in BAW-2248171 that was directed by the B&WaG Generic License Renewal Program (GLRP). The NRC final safety evaluation report (FSER) of BAW-2248 was attached to the NRC's letter to the B&WOG dated December 9, 1999.181 The NRC's letter and FSER are included in the updated BAW-2248A report.19] Per the NRC FSER, the BAW-2248 approach was found acceptable.

As presented in BAW-2248A, Table 4-1, a combination of existing inspections and additional work to be identified by the "RV Internals Aging Management Program" was credited for aging management of the B&W operating plant RV Internals, including the TMI-l RV Intcrnals.

The additional industry work on the aging of the RV Internals, begun by the submittal ofBAW-2248 for B&W plants resulted in the submittal of "PWR Internals Inspection and Evaluation Guidelines," MRP-227, Rev. 0 in January of 2009 for NRC review. IIOI The NRC issued a Safety Evaluationllli accepting MRP-227 with certain conditions and applicant action items. MRP-227-A was then completed in December 2011 with the changes requested by the NRC SER. MRP-227-AII ] is considered the industry program for TMI-l RV Internals.

Components recommended for augmented examinations are categorized as "Primary", "Expansion", or "Existing Programs". Components not recommended for augmented examinations are categorized as "No Additional Measures". The industry program is intended to provide a consistent approach to the aging management ofPWR RV Internals components across the PWR fleet. For additional information about MRP-227-A, see Section 4.2.2 of this report.

An AMR for LR was performed and the results are documented in Table 3.1.2-3 of the TMI-l LRA. A comparison of the results of the TMI-l LRA AMR and the results of the industry activity are provided in Appendix E of this report.

Initial augmented examinations for aging degradation mechanisms not yet completed are currently scheduled to be completed no later than the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel (B&PV) Code Section XI, 10-year In-service (lSI) inspection. A relief request for the 10-year lSI for TMI-l RV Internals, previously scheduled for the Fall 2011 outage, was submitted on October 29,2009 to allow certain ASME Code inspections, including the visual examinations of category B-N-2 and B-N-3 RV Internals, to be deferred to the Fa112015 outage. 112] By letter dated September 21,2010 containing the safety evaluation for this Page 2-2

ANI-'-L'H~L Rev. 001 the NRC authorized the deferral of the visual examination of the RV Internals until the Fall 2015 2.3 TMI-1 RV Internals AMP Intent The TMI-I RV Internals A.MP, which will include the TMI-I RV Internals Inspection Plan described in this document after review and approval by the NRC, utilizes a combination of prevention and condition monitoring.

Where applicable, credit is taken for programs primary water chemistry program and ASME B&PV Code Section XI inspection program). The TMI-I RV Internals Inspeetion Plan incorporates the requirements for augmented inspections provided by industry guidelines in MRP-227-A. Augmented inspections are in addition to the requirements of ASME B&PV Code Section

, the augmented inspections do not reduce, alter, or otherwise affect current ASME B&PV Code Section XI inspections at TMI-I.

degradation mechanisms that impact the RV Internals have been identified in MRP-227-A. MRP-227-A provides augmented examination requirements for detection of the effects of degradation meehanisms as listed in Table 2-2.

Table 2-2 RV Internals Aging Degradation Mechanisms and Their Aging Effects Aging Degradation Mechanism Stress Corrosion Cracking (SCC)

Irradiation-Assisted Stress Corrosion Cracking (IASCC)

Wear Fatigue Thermal Aging Embrittiement Irradiation Embrittlement (IE)

Void Swelling and Irradiation Grmv1h Thermal and Irradiation-Enhanced Stress Relaxation or Irradiation Enhanced Creep Aging Effect Cracking Cracking Loss of Material ghness, Unstable Crack Extension ness, Unstable Crack Extension Dimension Change, Distortion, and Cracking Loss of Mechanical Closu Section 5.0 of this report compares the TMI-I RV Internals AMP (including this inspection plan) to the ten clements in AMP XLMI6A ofNUREG-1801, Revision 2. The TMI-I RV Internals AMP incorporates programs and activities that are credited for managing the aging effects produced by the mechanisms listed in Table 2-2.

TMI-I RV Internals components within the scope of the LRA and NUREG-1928 have been considered in this TMI-I RV Internals Inspection Plan.

Appendix D to this inspection plan identifies the TMI-I response to each ofthe Topical Report Conditions and Applicant/Licensee Action Items in the NRC Safety Evaluation for MRP-227 Rev O. While disposition to several items are included in this plan, disposition of certain Applicant/Licensee Action Items will subsequently be addressed as listed in Appendix D.

A comparison of the results of the TMI-I LRA AMR (LRA Table 3.1.2-3) and the results of the industry activity (MRP-227-A and supporting documents MRP-189l161 and MRP-23I is provided in Appendix E of this report.

As discussed in response to Applicant/Licensee Action Item #2 in Appendix D, the only component identified in the LRA that is not included in the MRP-227-A development is the vent valve locking device. A review of this component using the MRP-227-A development methodolo!:,'Y will be performed and will be submitted to the NRC upon completion.

2.4 TMI-1 RV Internals Background The functions ofPWR reactor internals, as stated in Section 3.1 ofMRP-227-A, are to:

I.

provide support, guidance, and protection for the reactor core; 2.

provide a passageway for the distribution of the reactor coolant flow to the reactor core; Page 2-3

Rev. 001 3.

provide a passageway for support, gU:ldanCie, and protection for control elements and invessel/core instrUlme.t1tatiOll; and 4.

provide gamma and neutron shielding for the RV.

The TMI-I RV Internals consists of two structural subassemblies that are located within the RV: the plenum assembly and the core support assembly (CSA). The general arrangement of the RV Internals is shown in Figure 2-1.

Note that the TMI-I LRA includes the fuel and the control rods as RV Internals; however, these components are short-lived and they require no management.

Page 2-4

ANP-2952, Rev. 001 Figure 2-1: General Arrangement of the TMI-1 RV Internals

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,-: Vent Valve J

Ie Core Support Shield Ii++-I--Thermal Shield I.-+-+-Core Barrel

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AI~~-,/Y~\\/ Rev. 001 3.0 PROGRAM OWNER The TMI-I En,,\\ll1'Cerll1,1!; Programs Group is responsible for erc:atlng, maintaining, and implementing the TMI-I RV Internals Inspection Plan.

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AI\\JI-'-':ll:l:J.L. Rev. 001 4.0 INDUSTRY AND TMI-1 PROGRAMS AND ACTIVITIES The TMI-I RV Internals Inspection Plan is based on the TMI-I LRA, NUREG-I928, andMRP-227-A. The TMI-I RV Internals AMP implements MRP-227-A through this RV internals inspection plan. The TMI-I RV Internals AMP also includes the continuation of some of the programs and activities discussed in this section.

4.1 TMI-1 Programs and Activities The TMI-I RV Internals AMP consists of a number of programs and activities that support aging managellle:nt of the RV Internals; these include the ASME B&PV Code Section XI lSI Program, primary water chemistry program, plant technical specifications (vent valve inspection and exercise program),

TLAAs, fabrication records searches, upper core barrel (UCB) and lower core barrel (LCB) bolt analysis, past RV Internals inspections, and a fuel/baffle interaction investigation.

4.1.1 ASME B&PV Code Section XI lSI Program The ASME B&PV Code Section XI lSI program is an existing program developed under ASME B&PV Code Section XI, Subsection IWB-2500.1151 ASME B&PV Code Section XI, Table IWB-2500-1, Examination Category B-N-3 applies to core support struetures (i.e., RV internals). Examination of these B-N-3 RV Internals is required once every 10-year lSI interval. The VT-3 examinations are perfonlled with the aid of visual examination tools, in accordance with a code compliant VT-3 procedure.

The next TMI-I ASME B&PV Code Section XI IO-year lSI inspection requiring CSA removal is currently scheduled for the Fall 2015 refueling outage (RFO).1131 4.1.2 Primary Water Chemistry Program The TMI-I reactor coolant chemistry program, as implemented by the TMI-I Reactor Coolant System Chemistry Program, limits the concentration of oxygen, halogens, and sulfate species in the primary water.

The intent of these limits is to prevent the reactor coolant fl'om becoming an environment favorable to SCC or IASCC, therefore greatly reducing the probability of SCC and IASCC.

4.1.3 Plant Technical Specifications As described in thc TMI-I Plant Technical Spccifications, Section 4.16 "Rcactor Intcrnals Vent Valvc Survcillancc", vcnt valve tcsting and inspections are requircd to bc performcd each RFO. This requirement is fulfilled by the TMI-I Reactor Internals Vent Valvc Inspection and Exercisc procedure. The acccssible areas of the vcnt valve are typically inspected, including thc locking deviccs. Additionally, vent valve opcration is tested through manual actuation.

4.1.4 Time Limited Aging Analyses As dcscribcd briefly in Scction 2.0, the TMI-I RV Intcrnals Aging Managemcnt Program includes thrce Time Limited Aging Analyscs that were cvaluated and dispositioncd in the License Renewal Application.

The three TLAAs for thc RV Intcrnals include low cycle fatiguc, high cycle fatigue, and neutron cmbrittlement.

Low cycle fatigue is managed by thc Fatigue Monitoring Program (LRA Section 4.3.4).

High cycle fatigue (flow induced vibration) is projected through the period of extended operation (LRA Scction 4.3.5).

The LRA committed to manage neutron cmbrittlemcnt by participating in, evaluating, and implementing thc results of RV Internals industry programs that were then in devclopmcnt. Those programs (i.e. MRP-227-A) are now approved and are implementcd via this inspcction plan as part of the TMI-I RV Intcrnals AMP. Ncutron embrittlcment is onc of the aging effects considercd in the development ofMRP-227-A (see Section 3.2.6 ofMRP-227-A).

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AI\\lt-'-.<:l:1:J,L.. Rev. 001 The responses to Ap'plicalnt/ILic:ensee Action Items 6 and 7 from the NRC SER for MRP-227-A may result in additional the use of the TMI-I RV Internals. As these are developed, any time-dependent will be identified and addressed through the period of extended operation.

4.1.5 Fabrication Records Searches Fabrication records searches for TMI-I RV Internals components listed as "Primary" and "Expansion" in MRP-227, Rev. 0 were conducted by AREVA NP.

These fabrication records searches included a search of original fabrication records for several east austenitie stainless steel (CASS) RV Internals items, ineluding the core support shield (CSS) vent valve discs and control rod tube (CRGT) spacer castings.

In 2012, a search of the original fabrication records was performed to eonflrm that the CSS upper t1ange weld was stress relieved, as required by Applicant/Licensee Action Item#4 in the NRC SER for MRP-227. This search confirmed that the TMI-I CSS upper flange weld was stress relieved.

4.1.6 UCB and lCB Bolt Analysis An analysis was performed for the TMI-l UCB and LCB bolts. The stress limits for threaded structural t1:1steners in Subsection NG of the ASME B&PV Codell8J were used to develop the acceptance criteria for the integrity of the bolted joint. The analytical model is intended to be used to assess inspection results for the UCB and LCB bolts at TMI-I.

Five cases representing hypothetical UCB and LCB degraded bolt configurations were analyzed to provide insight to the number and pattern of degraded bolts that can be tolerated without exceeding the analytical criteria. The results indicate a large number ofdegraded UCB and LCB bolts can be tolerated without exceeding the analytical criteria for the joint if the degraded bolts are not adjacent. If degraded bolts are adjacent to each other, and located in the worst location, the number ofdegraded UCB and LCB bolts that can be tolerated is lowcr.

4.1.7 Past RV Internals Inspections Past inspections of the RV Internals include the vent valve inspections, UCB inspections, and core clamping measurements at TMI-I.

4.1.7.1 Vent Valve Inspection As discussed in Section 4.1.3 of this report, vent valve testing and inspeetions are required to be performed each RFO. Results from the vent valve inspection and exercise test performed in the 2007 and 2009 RFOs, were deemed acceptable. During the vent valve inspection and exercise test performed during the 20 II RFO, one vent valve exceeded the stay open force acceptance criteria. The valve was cycled five times, after whieh the holding force was found to be acceptable. The high as-found loading force required to maintain the valve open was attributed to erud from the reactor coolant system oxidation process. No other notable conditions were observed during the 2011 inspection.

4.1.7.2 UCB UT Examination at TMI-1 During the TMI-I 2009 RFO, AREVA NP performed an UT examination of 100% of the UCB bolts. The examinations were performed under direction of the AREVA NP non-destructive examination (NDE) group.

No recordable indications were detected in any of the UCB bolts. UT examinations were also performed at TMI-I on the UCB bolts in 1982 (partial) and 1991 (100%). No recordable indications were detected by these examinations.

4.1.7.3 Core Clamping Measurements Core clamping measurements were obtained by AREVA NP at TMI-I in 2010 to satisfy the MRP-227-A examination requirements for a one-time physical measurement of the differential height of top of the plenum rib pads to the reactor vessel seating surface. This measurement was taken with the plenum cover weldment Page 4-2

r\\1"r-LC:7.JL. Rev. 001 rib plennm cover support and CSS top inside the RV, but with the fuel assemblies removed per Section 4.3.1 ofMRP-227-A. The conclusions of the core clamping summary document are that there was no evidence of wear occurring the service period and the measurements were acceptable.

core clamping measurements at TMI-I meet the one-time physical measurement requirement 4.1.8 Fuel/Baffle Interaction Investigation An was conducted on the interaction between the baffle plates and fuel assembly grid straps in similarly-designed units between 2004 and 2010. Exelon has participated and will continue to participate in this project.

4.2 Other Industry Programs and Activities Exelon participates in industry activities which support the management of aging of the RV Internals. These activities are described in the sections below. These industry programs and activities have helped to define the required examinations and examination techniques for the components covered by this TMI-I RV Internals Inspection Plan.

4.2.1 B&WOG Generic License Renewal Project Report (BAW-2248A)

As described previously in Section 2.2 of this report, the B&WOG GLRP report (BAW-2248A) contains a technical evaluation of effects for B&W RV Internals components. The B&WOG sent the report to the NRC staff to demonstrate that participating member plant owners subscribing to the report eould adequately manage effects of aging on RV Internals during the period ofextended operation.

BAW-2248A was a predecessor to ongoing industry work through the EPRI PWR MRP for the B&W units, which resulted in the current guidance in MRP-227-A.

4.2.2 MRP-227-A The MRP-227-A "Pressurized Water Reactor Internals Inspections and Evaluation (I&E) Guidelines" were developed by a team of industry representatives who reviewed available data and industry experience to identify and prioritize I&E requirements for RV Internals. MRP-227-A is the result of the industry work and NRC review that began with BAW-2248A for the B&W plants. The key sequential steps in the process included the following:

The development of screening criteria, with susceptibility for the eight postulated aging mechanisms relevant to reaetor internals and their effects; An initial component screening and categorization, using the susceptibility levels and FMECA (failure modes, effects, and criticality analysis) to identify the relative ranking of the components; Functionality assessment of degradation for components and assemblies of components; and Aging management strategy development combining the 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 for and the timing of subsequent inspections.

Through this process, the RV Internals for all three PWR designs in the U.S. were evaluated, and appropriate requirements for aging management actions specific to each component were provided.

MRP-227-A utilized the screening and ranking process to aid in the identification of required examinations for "Primary" and "Expansion" components and credits "Existing Programs" when they were deemed adequate.

The basic description of each classification is as follows:

"Primary" Page 4-3

Rev. 001 Those PWR internals that are to the effects of at least one of the mechanisms wcre in the group. The needed to ensure functionality of components are in these I&E

[MRP-227-AJ. The "Primary" group also includes components which have shown a of tolerance to a degradation effect, bnt for which no highly susceptible component exists or for which no highly susceptible cornponent is accessible.

'Expansion" Those PWR internals that are highly or moderately susceptible to the effeets of at least one of the eight mechanisms, but for which functionality assessment has shown a of tolerance to those v<Iv'-"" were placed in the "Expansion" group. The schedule for implementation of aging management examinations for "Expansion" components will depend on the findings from the examinations of the "Pr,n"lfV" components at individual plants.

"Existing Programs" Those PWR internals that are susceptible to the efTects of at least one of the eight aging mechanisms and for which and plant-specific existing AMP elements are capable of managing those effects, were placed in the "Existing Programs" group. There are no B&W plant internals components in this group.

"No Additional Measures" Those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria were placed in the "No Additional Measures" group. Additional components were placcd in the "No Additional Measures" group as a result ofFMECA and the functionality assessment. No further action is recommended by these guidelines for managing the aging of the "No Additional Measures" components.

The categorization and analysis processes used in the MRP-227-A approach are not intended to supersede ASME B&PV Code Section XI requirements.

The requirements ofMRP-227-A arc classified in accordance with NEI 03_08[19] guidelines. For the MRP-227-A guidelines there arc one "Mandatory", five "Needed", and zero "Good Practice" requirements as follows:

"Mandatory" Each commercial U. S. PWR unit shall develop and document a program for management of aging of reactor internal components within thirty-six months following issuance of MRP-227, Rev 0 (that is no later than December 31,2011).

Exclon has established procedures to develop and document an AMP for the Reactor Vessel Internals.

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

The applicable B&W tables arc MRP-227-A, Table 4-1 ("Primary"), Table 4-4 ("Expansion"), and Table 5-1 (Examination Acceptance and Expansion Criteria). There arc no "Existing Program" components in MRP-227-A for B&W-designed PWRs. TMI-I has followed the MRP-227-A requirements by the inspection activities already planned or performed as described in this TMI-I RV Internals Inspection Plan. Appendix A is MRP-227-A Table 4-1 modified to reflect only the TMI-I primary components.

Appendix B is MRP-227-A Table 4-4 modified to reflect only TMI-I expansion components. Appendix Cis MRP-227-A Table 5-1 modified to reflect only TMI-I acceptance/expansion criteria. Therefore, implementation of this inspection plan will fulfill this "Needed" requirement for TMI-I.

"Needed" Examinations specified in these guidelines shall be conducted in accordance with the Inspection Standard rMRP-228J.

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AI\\Jt-'-LH:JL Rev. 001 Inspec:tlOln sltanclalrds developed under will be used by Exelon for the augmented examinations described in this TMI-l RV Internals Inspection Plan developed in accordance with MRP-227-A. Implementation of this RV Internals Inspection Piau will fulfill this "Needed" requirement for TMI-l.

"Needed" Examination results that do not meet the examination acceptance criteria defined in Section 5 of the MRP~227~Aguidelines shall be recorded and entered in the plant corrective action program and dispositioned.

The TMI-l Correetive Action Program (CAP) will be used, as discussed in Section 5.7 of this report.

Implementation of this TMI-I RV Internals Inspection Plan will fulfill this "Needed" requirement for TMI-I.

"Needed" Each commercial U. S. PWR unit shall provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MRP Program Manager within 120 days of the completion of an outage during which PWR internals within the scope of MRP~227~A are examined.

Exelon will provide a summary report of all inspections and monitoring, items requiring evaluation, and new repairs to the MRP program manager for future TMI-I inspection activities within 120 days of the completion of an outage during which PWR RV Internals within the scope ofMRP-227-A are examined.

Implementation of this TMI-I RV Internals Inspection Plan wi II fulfill this "Needed" requirement for TMI-l.

"Needed" If an engineering evaluation is used to disposition an examination result that does not meet the examination acceptance criteria in Section 5 [ofMRP-227-A), this engineering evaluation shall be conducted in accordance with an NRC~approvedevaluation methodology.

Exelon will disposition all examination results that do not meet the aeeeptance criteria [in Section 5 of MRP-227-AJ in accordance with NRC approved evaluation methodology. Implementation of this TMI-I RV Internals Inspection Plan will fulfill this "Needed" requirement for TMI-l.

4.2.2.1 MRP-227~A Applicability to TMI-1 The applicability ofMRP-227-A guidelines are based on several general assumptions that were used for the analysis in the development ofMRP-227-A. Additional assumptions in the Failure Modes, Effects, and Criticality Analysis (FMECA - MRP-190(21 1) and the functionality analysis report (MRP_229[22J) are also included as required by Applicant/Licensee Action Item #1 from the NRC SER for MRP-227, Rev O.

MRP-227-A:

The assumptions found in Section 2.4 ofMRP-227-A and their applicability to TMI-I are listed below:

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

The fuel management program for TMI-l changed from a high to a low leakage core loading pattern prior to 30 years of plant operation. This change was started in TMI-I Cycle 6 (1987) and has been continually implemented through the most recent fuel cycle, TMI-I Cycle 19 (20 II). This change is considered to be a preventative action to lessen the effects of aging on the TMI-I RV Internals. TMI-1 will continue to use low-leakage core loading pattern.

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

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Rev. 001 TMI-I opc:rat\\~sas a base load unit.

ch.an!~cs beyond those identitied in.",'-"'-,.,,, industry gUldaJrlce or recommended by the MRP-227-A states that the requirements arc applicable to all U. S. PWR operating plants as ofMay 2007 for the three (i.e., B&W, Westinghouse, and CE) considered. No modifications have been made to the TMI-I RV Internals since May 2007.

iVIRP-190:

Section 4 of MRP-190 (the FMECA) contains 6 assumptions and observations. As stated in Section 4 of MRP-190, these assumptions are either bounding or methodological, and do not require plant-specitic verification for each of the B&W-designed operating units.

MRP-229, Rev. 3:

Section 2.4.1 of MRP-229 (the functionality analysis report) identifies eight bulleted limitations and assumptions. Seven ofthose are programmatic rather than site specitic. Only one pertains to the plant design and operating history, and that one is discussed below:

The crfect of power uprates has not been considered by the functionality analysis.

TMI-I intends to apply for a measurement uncertainty recapture (MUR) uprate in the future. The impact of that uprate and any other uprates on MRP-227-A will be formally reviewed and reported as part of the LAR for the power uprate.

Based on the above review, MRP-227-A is applicable to TMI-I and Applicant/Licensee Action Item # I from the NRC SER for MRP-227-A is complete. See Appendix 0, Table 0-2 for more information.

4.2.3 Joint Owner's Baffle Bolt Program The Joint Owners' Baffle Bolt (JOBB) Program stemmed from UT examinations ofbaffle-to-tormer bolts at several European plants. Indieations were noted under the bolt head in the head-to-shank fillet radius. The bolt failures were attributed to IASCC. Various tasks, including NOE examinations, irradiation and mechanical testing, corrosion testing, and microstructural evaluation were used to characterize the crfect of irradiation on bolting materials under the JOBB program.

The JOBB program is now being managed by EPRI with additional research on RV Internals material being performed under EPRI programs. The results of the JOBB program have been incorporated into EPRI PWR MRP documents, and specitieally referenced in MRP-227-A. The EPRI PWR MRP provides results to the NRC during meetings; an example of the results presented in these meetings is given in Reference 23.

4.2.4 PWROG and EPRI/MRP The utilities sponsor activities related to PWR RV Internals aging management through both the PWROG and EPRI PWR MRP. Exelon's participation in current PWROG and EPRI PWR MRP activities will continue.

4.3 Conclusions of Section 4.0 This section contains a description of the TMI-I Reactor Vessel Internals Aging Management Program, including this RV Internals Inspection Plan. The TMI-I RV Internals Inspection Plan is based on the TMI-I LRA, NUREG-1928, and evaluations supporting MRP-227-A. Inspections will consist of the ASME B&PV Code Section XI inspections and the augmented examinations from MRP-227-A. Changes resulting from the NRC's review of this report will be incorporated as appropriate.

As part of the TMI-I License Renewal Program, Exelon agreed, by updated final safety analysis (UFSAR) commitment, to participate in industry programs for investigating and managing aging etfects on RV Internals and to evaluate and implement the results of the industry programs as applicable to the RV Internals. Once these programs have been completed, but not less than twenty-four months before entering the period of Page 4-6

Rev. 001 extended opieratIOI1, Exelon has committed to submitting a TM]~I RV Internals Inspection Plan to the NRC f<lr review and which is the and submission of this TMI-l RV Internals Inspection Plan.

Appendix E of this report a comparison of the results of the TMI-I LRA AMR and the results of the industry that lead up to MRP~227 ~A.

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Rev. 001 5.0 TMI-1 RV INTERNALS AMP ATTRIBUTE EVALUATION The TM]-I RV Internals AMI" which includes this TMI-I RV Internals Inspection Plan, utilizes a combination of prevention and condition monitoring to manage the effects of the eight age-related degradation mechanisms given in Section 2.3 of this thereby providing reasonable assurance the RV Internals continue to perform their LR function the period of extended operation. Where applicable, eredit is taken for existing programs and activities primary water chemistry program, vent valve testing, TLAAs, and ASME B&PV Code Section XI inspections). This TMI-I RV Intcrnals Inspcction Plan incorporates the industry guidancc in MRP-227-A.

This section compares the TMI-I RV Internals AMP, including this inspection plan, to the ten element AMP XLM 16A from NUREG-180 I, Revision 2. The conclusion of this comparison is that the TMI-I Reactor Internals AMP is consistent with the NUREG-1801 XLMl6A program with no exceptions and no enhancements.

Aging Management Program Description (from NUREG-1801, Revision 2, XLM16A):

frp**rp!n/pd degradation in the reactor internals is managed through an integratedprogram. Specific features ol the integratedprogram are listed in thefollowing ten program elements. Degradation due to changes in material properties (e.g., loss of/racture toughness) was con...;idered in the determination olinspection recommendations and is managed by the requirement to use appropriatezy degraded properties in the evaluation olidentified delects. The integratedprogram is implemented by the applicant through an inspection plan that is submitted to the NRClor review and approval with the applicationfc)r license renewal.

5.1 AMP Element 1 - Scope of Program 5.1.1 NUREG-1801, XI.M16A Scope of Program The scope olthe program includes all RVI components at Three Mile Is'land Unit I, which is built to a B&W NSSS design. The scope olthe program applies the methodology and guidance in the most recently NRC-endorsed version olA4RP-227, which provides augmented inspection andflaw evaluation methodologyfbr assuring the fimctional integri~y olsale~-relatedinternals in commercial operating u.s. PWR nuclear power plants designed by B&W, CE, and Westinghouse. The scope ofcomponents consideredfc)r inspection under AlRP-227 guidance includes core support structures (typically denoted as Examination Category B-N-3 by the ASAIE Code,Section XI), those RVI components that serve an intended license renewal safe~fimction pursuant to criteria in 10 CFR 54.4(a)(I), and other RVI components whosefailure could prevent satisfactory accomplishment ofany ofthe fimctions identified in 10 CFR 54.4(a)(l)(i), (ii), or (iii). The scope ofthe program does not include consumable items, such asfitel assemblies, reactivi~ control assemblies, and nuclear instrumentation, because these components are not typicalzy within the scope ofthe components that are required to be subject to an aging management review (AlvIR), as defined by the criteria set in 10 CFR 54.21(a)(1), The scope ofthe program also does not include welded attachments to the internal surface olthe reactor vessel because these components are considered to be ASj\\;IE Code Class 1 appurtenances to the reactor vessel and are adequatezy managed in accordance with an applicant's Alli!P that corresponds' to GALL AMP XI.M1, "AS!'vIE Code, Section XlInservice Inspection, Subsections IWB, avc, and IWD.,.

The scope ofthe program includes the response bases to applicable license renewal applicant action items (LRAAIs) on the,'-'!RP-227 methodolo,t,'Y, and any additional programs. actions, or activities that are discussed in these LRAAI responses and creditedfor aging management ofthe applicant's RVI components. The LRAAls are identified in the statts sale~ evaluation on.~IRP-227 and include applicable action items on meeting those assumptions thatfbrmed the basis olthe MRP's augmented inspection aneiflaw evaluation methodology (as discussed in Section 2.4 01lliIRP-227), and NSSS vendor-specific or plant-specific LRAAIs as well. The responses to the LRAAIs on iVtRP-227 are provided in Appendix C ofthe LRA.

The guidance in MRP-227 specifies applicability limitations to base-loadedplants and the.lite/loading management assumptions upon which thefimctionali~y analyses were based. These limitations and assumptions require a determination ofapplicabili~ by the applicantfbr each reactor and are covered in Section 2.4 ofMRP-227.

Page 5-1

5.1.2 TMI-1 Scope of Program AN~-/'~hL. Rev. 001 A description of the TMI-I RV lntemals is provided in Section 2.4 of this report. Additional RV Intemals details are provided in the TMI-l UFSAR.

The TMI-l RV Intemals includes the components identified in MRP-227-A for B&W-designed RV Intemals applieable to TMI-l.

The TMI-l RV Intemals includes responses to the applicant/licensee action items in the MRP-227-A SER Appendix D to this document). Those responses are in this inspection plan rather than in the TMI-l LRA, because the NRC SER for Rev. 0 was not available when the TMI-l LRA was written.

The scope of the TMI-I RV Intemals Program includes those components identified in the results of the TMI-l LRA AMR. Appendix E compares the results of the TMI-l LRA AMR with the results of the industry aetivity for all TMI-l components in scope for license renewal.

The limitations and assumptions in as well as in the FMECA (MRP-190) and functionality analysis (MRP-229), are applicable to TMI-I, and are discussed in detail in Section 4.2.2.1 of this report.

5.1.3 Conclusion The seope of the TMI-l RV Intemals Program is consistent with the scope ofNUREG-180l program XLM16A.

5.2 AMP Element 2 -- Preventative Actions 5.2.1 NUREG-1801, XI.M16A Preventive Actions The guidance in AlRP-227 relies on PWR water chemistry control to prevent or mitigate aging eflects that can be induced corrosive mechanisms loss olmaterial induced by general. pitting corrosion, crevice corrosion, or stress corrosion cracking or any olitsforms [SCC. PWSCC, or IASCC]). Reactor coolant >vater chemistlY is monitored and maintained in accordance with the Water Chemistry Program. The program description. evaluation, and technical basis olwater chemistlY are presented in GALL AA1P XJ.lvf2,..Water Chemistry. "

5.2.2 TMI-1 Preventive Actions The TMI-I RV Intemals AMP credits the Reactor Coolant Chemistry Program for maintaining high water purity (See Section 4.1.2 of this report).

Additionally, TMI-I has implemented low-leakage core loading pattems as a preventative action (See Section 4.2.2.1 of this report).

5.2.3 Conclusion The preventative actions for the TMI-I RV Intemals AMP are consistent with the preventive actions in NUREG-180I program XLM 16A.

5.3 AMP Element 3 -- Parameters Monitored/Inspected 5.3.1 NUREG-1801, XI.M16A Parameters Monitored/Inspected The program manages thef()llowing age-related degradation effects and mechanisms that are applicable in general to the RVI components at thefacility: (a) cracking induced by SCC, PWSCC, IASCC, orfatigue/cyclica/

loading: (b) loss olmateria/ induced by wear: (c) loss ofFacture 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 olpreload caused by thermal and irradiation-enhanced stress relaxation or creep. For the management ofcracking, the program monitorsfor evidence olsurface breaking linear discontinuities ila visual inspection technique is used as the non-destruction examination (NDE) method, orlor relevantflaw presentation signals ifa volumetric UT method is used as the NDE method. For the management of loss olmaterial, the program monitorsfbr gross or abnormal sllrlace conditions that may be indicative ofloss ol Page 5-2

Rev. 001 loss material in the For the preload, the program rnonitorsj(Jr gross su,rjLJ!ce conditions that may indicative in

bolted, or pinned connections. The progrwn does not monitorlor loss that is induced thermal or neutron irradiation embrittlement, or bJ' and irradiation growth; instead the impact fro'{'fllrp tOlfgi'1i1I?SS on is indirectlv managed by using visual or volumetric examination tec'hn'iql'Ies to monitorfiJr in the components and by applying applicable reducedfracture fOlIf<!'1i1I?SS fWI'lf)l,wtip<: in evaluations il is detected in the components and is extensive enough to warrant a orflaw tolerance evaluation under the AlRP-227 or AS/,\\;fE Code, Section Xl The program uses physical measurements to monitorfor any dimensional changes due to void sH',?llinf<, irradiation growth, distortion, or deflection.

the program implements the parameters monitored/inspected criteriaj(n< B&W designed Primwy Components in Table 4-1

. Additionally, the program implements the parameters monitored/inspected criteria/iJr B&W designed Er:pansion Components in Table 4-4 olMRP-227. The parameters monitored/inspectedfilr Program Components/ill/ow the basesfor relerenced Exi,yting Program.';, such as the requirementsfiJr ASAlE Code Class R VI components in ASAlE Code,Section XI, Table 1WB-2500-1, Examination Categories B-N-3, as implemented through the applicant's ASAlE Code,Section XI program, or the recommended programfbr inspecting Westinghouse-designedflux thimble tubes in GALL Alv!P X1.J\\437, "Flux Thimble Tube 1nspection." No inspections, exceptj(Jr those specilied in ASME Code,Section XI, are requiredfiJr components that are identifIed as requiring "No Additional Aleasures, " in accordance >vith the analyses reported in lv!RP-227.

5.3.2 TMI-1 Parameters Monitoredllnspected The TMI-I RV Internals Inspection Plan monitors for the detcctable effects of the eight aging def,'Tadation mechanisms outlined in Seetion 2.3 of this report. The TMI-I RV Internals AMP credits, and further augments, the ASME Section XI Inservice Inspection Program with the examinations in MRP 227-A, Tables 4-1 and 4-4 as applicable to TMI-l. See Appendiees A and B of this report.

The TMI-I RV Internals Inspeetion Plan uses Ultrasonic Testing (UT), Visual Examination (VT-3), and physical measurement to monitor for the deteetable effects of the eight degradation mechanisms outlined in Section 2.3 of this report.

5.3.3 Conclusion The parameters monitored/inspected in the TMI-I RV Internals AMP are consistent with the parameters monitored/inspected in the NUREG-180 I program XI.M 16A.

5.4 AMP Element 4 - Detection of Aging Effects 5.4.1 NUREG-1801, XI.M16A Detection of Aging Effects The detection olaging effects is covered in two places: (a) the guidance in Section 4 of}...fRP-227 provides an introductory discussion andjustilication olthe exarnination methodl' selectedj(Jr detecting the aging effects ol interest; and (b) standardsj(Jr examination methods, procedures, and personnel are provided in a companion document. lv!RP-228. In aI/ cases, well-established methodl' were selected. These methods include volumetric UT examination methody/or detectingpaws in bolting, physical measurements fbI' detecting changes in dimension.

and various visual (VT-3, VT-1, and EVT-1) examinationsjiJr detecting effects rangingfi'om general conditions to detection and sizing olswface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations/or detection and sizing olswjace-breaking discontinuities, Cracking caused by SCC, IASCC, and/atigue is monitored/inspected by either VT-l or EVT-l examination (fbI' internals other than bolting) or by volumetric UT examination (bolting). The VT-3 visual methods may be applied

/iJr the detection olcracking only when theflaw tolerance ofthe component or afJected assembly, as evaluatedfor reducedfracture toughness properties, is known and has been shown to be tolerant oleasily detected large flaws, even under reducedfracture toughness conditions. In addition, VT-3 examinations are used to monitor/inspect jiJr loss olmaterial induced by wear andfor general aging conditions, such as gross distortion caused by void Page 5-3

Rev. 001 o/preload caused thermal and irradiation-enhanced or gross Sw,(?fI J'nrr and irradiation stress relaxation and creep.

In addition, the program adopts the recommended in AlRP-227for defining the Expansion criteria that need to be applied to inspections Components and E,isting Requirement Components andf;)r examinations to additional As a result, inspections perfbrmed on co,m,;'on'ents are consistent vvith the and sampling bases j;)r Primary Components, Requirernent Components, and Components in MRP-227, which have been demonstrated to be in confbrmance with the inspection sampling basis criteria, and sample Expansion criteria in Section A.1.2.3.4 olNRC Branch Position RLSB-1.

Specifically, the program implements the parameters monitored/inspected criteria and basesfbr inspecting the relevant parameter conditionsfbr B&W Primary Components in Table 4-1 of},JRP-227; andfbr B&W desi.f,~ne'dExpansion Components in Table 4-4 The program is supplemented thefbllowing planHpecific Primary Component and Expansion Component inspectionsfor the program (as applicable): For TMI-I, the reactor vessel internals vent valve locking devices will be evaluated as possible planh)lJecific RV Internals components to be inspected.

In addition, in some cases (as defined in ;~,4RP-227), physical measurements are used as supplemental techniques to managefor the gross effects olwear, loss olpreload due to stress relawtion, orjc)r changes in dimension due to void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program include the physical measurements neededf;)r the B&W RV internals core clamping items already identified in AtRP-227-A, Table 4-1.

5.4.2 TMI*1 Detection of Aging Effects The methods for detection of aging effects in the TMI-I RV Internals Inspection Plan include UT examination methods for detecting flaws in bolting, physical measurements for detecting changes in dimension, and visual (VT-3) examinations.

Visual (VT-3) examinations are conducted to determine the general mechanical and structural condition of components by detecting discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion; and by identifying conditions that could affect operational or functional adequacy of components. In accordance with Tables 4-I and 4-4 of MRP-227-A, VT-3 is scheduled to be used to detect cracking and other aging effects in various components. VT-3 examinations will be used to detect cracking only after evaluation of the flaw tolerance of the component or affected assembly, under reduced fracture toughness conditions, has been shown to be tolerant ofeasily detected flaws.

The TMI-I Reactor Vessel Internals Aging Management Program requires inspections of the "Expansion" components in MRP-227-A in accordance with the expansion criteria in MRP-227-A.

Applicant/licensee action item #2 requires each licensee to identify whether additional RV Internals components need to be inspected based upon review of the licensee's License Renewal Application. As summarized in Appendix D, Table D-2 of this report and tabulated in Appendix E of this report, the vent valve locking devices were the only additional components in the TMI-I LRA that were not included in the RV Internals industry program.

The one-time physical measurements needed for the B&W RV Internals core clamping items has already been completed at TMI-I and are discussed in Section 4. 1.7.3.

5.4.3 Conclusion The detection of aging efTects in the TMI-I RV Internals AMP is consistent with the detection of aging efTects in NUREG-1801 program XLMI6A.

Page 5-4

MI\\J.....-/';.l:")/ Rev. 001 5.5 AMP Element 5 - Monitoring and Trending 5.5.1 NUREG-1801, XI.M16A Monitoring and Trending The methods'/br monitoring, evaluating, and the data that resultfrom the program :s' in.\\:pf:'ctl'oliis are in Section 6 i\\lIRP-227 and its subsections. The evaluation methods include recommendations flaw andjbr crack detcrminations as,vellfor perflmning applicable limit load, linear elastic and elastic-plasticfracture analyses olrelevantflaw indications. The examinations and re-examinations the iVlRP-227 guidance, with the requirements specified in lvlRP-228fbr inspection methodologies, inspection procedures, and inspection personnel, provide time~v detection, reporting, and corrective actions with to the olthe degradation mechanisms within the scope ol the program. The extent olthe examinations, beginning,vith the sample olsusceptible PWR internals component locations as PrimaJy Component locations, with the potentialfl)r inclusion olExpansion Component locations ilthe are than anticipated, plus the continuation olthe Existing Programs activities, such as the ASA1E Section Xl, Examination Categ01y B-N-3 examinationsfor core support structures, provides a high olconfidence in the total program.

5.5.2 TMI-1 Monitoring and Trending The TMI-l RV Internals AMP, including one-time, periodic, and conditional examinations and other aging management methodologies, scheduled in accordance with the ASME B&PV Code,Section XI and MRP-227-A, provide timely detection of aging effects. The TMI-I program incl udes both the "Primary" components, and the "Expansion" components identified in MRP-227-A. The "Expansion" components will be inspected as required by the results of the examinations of the "Primary" components.

The TMI-I RV Internals AMP will follow the reporting requirements in MRP-227-A which allow the industry to monitor and trend results, thus driving preemptive industry action through notifications and updating of the MRP-227-A guidelines.

5.5.3 Conclusion The monitoring and trending in the TMI-I RV Internals AMP is consistent with the monitoring and trending in NUREG-180 I program XLM 16A.

5.6 AMP Element 6 - Acceptance Criteria 5.6.1 NUREG-1801, XI.M16A Acceptance Criteria Section 5 ollvlRP-227 provides specific examination acceptance criteriafbr the Prinu1!y and Expansion Component examinations. For components addressed by examinations relerenced to ASlvlE Code, Section Xl, the IWB-3500 acceptance criteria apply. For other components covered by Existing Programs, the exarnination acceptance criteria are described within the Existing Program relerence document. The guidance in MRP-227 contains three types olexamination acceptance criteria:

For visual examination (and sw1ace examination as an alternative to visual examination), the examination acceptance criterion is the absence olany olthe specific, descriptive relevant conditions; in addition, there are requirements to record and disposition sw1ace breaking indications that are detected and sizedfor length by VT-J/EVT-J examinations; For volumetric examination, the examination acceptance criterion is the capability for reliable detection olindications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for sTstem-level assessment olbolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and For physical measurements, the examination acceptance criterion for the acceptable tolerance in the measured differential heightlrom the top olthe plenum rib pads' to the vessel seating swjace in B&W plants are given in Table 5-J ofMRP-227.

Page 5-5

5.6.2 TMI-1 Acceptance Criteria AI\\It-'-L~::JL. Rev. 001 The TMI-I RV Internals AMP uses the examination criteria in Section 5 ofMRP-227-A for the "Primaty" and "Expansion" components. This includes examination acceptance criteria for visual examinations, volumetric examinations (as demonstrated in the examination Technical Justification), and physical measuremcnts, as well as Criteria for expanding the examinations from the "Primary" components to include the TMI-l NDE examination techniques will be qualified to the extent required by MRP-227-A and MRP-228. Component degradation that exceeds the examination acceptance criteria will be evaluated per MRP-227-A, but will also be evaluated considering the supplemental guidance in WCAP-17096, including any additional guidance resulting from the ongoing NRC review of that document.

Relevant conditions requiring corrective action for ASME B&PV Code Section XI Category B-N-3 VT-3 examinations of RV Internals eomponents are detailed in ASME B&PV Code Section XI, IWB-3000.

5.6.3 Conclusion Acceptance criteria for the TMI-I RV Internals AMP are consistent with the acceptance criteria in NUREG-180 I program XLM 16A.

5.7 AMP Element 7 - Corrective Actions 5.7.1 NUREG-1801 XI.M16A Corrective Actions Corrective actionsfiJllowing the detection o{unacceptable conditions arefimdamentally providedfor in each plant 's corrective action program. Any detected conditions that do not satisfY the examination acceptance criteria are required to be dispositioned through the plant corrective action program. which may require repair.

replacement, or analvtical evaluationfor continued service until the next inspection. The disposition will ensure that design basisfimctions o{the reactor internals components will continue to befidfilledfiJr all licensing basis load\\' and events. Examples o{methodologies that can be used to analytically disposition unacceptable conditions arefinmd in the ASME Code. Section Xl or in Section 6ofAlRP-227. Section 6 of-:A.4RP-227 describes the options that are availablefor disposition o{detected conditions that exceed the examination acceptance criteria o{Section 5 o{the report. These include engineering evaluation methods. as well as supplementOlY examinations tofitrther characterize the detected condition. or the alternative ofcornponent repair and replacement procedures. The latter are subject to the requirements o{the ASME Code. Section Xl. The implementation ofthe guidance in MRP-227. plus the implementation ofany ASME Code requirements. provides an acceptable level o{aging management ofsa{ety-related components addressed in accordance with the corrective actions of I0 CFR Part 50. Appendix B or its equivalent. as applicable.

Other alternative corrective action bases may be used to disposition relevant conditions ij'they have been previously approved or endorsed by the NRC. Examples ofpreviously NRC-endorsed alternative corrective actions bases include those corrective actions basesfor B& W-designed RVI components in B&W Report No.

BAW-2248.

B&WReport No. BAW-2248 was endorsedjor use in an SE to Frarnatome Technologies on behalfof the B&W Owners Group. dated December 9. 1999. Alternative corrective action bases not approved or endorsed by the NRC will be submittedfor NRC approval prior to their implementation.

5.7.2 TMI-1 Corrective Actions In accordance with 10 CFR 50, Appendix B, Exelon has established a CAP for TMI-I. The CAP at TMI-I is implemented through the Exclon CAP Procedure. The purpose of the CAP is to promote continuous improvement through organizational learning and provide direction on the resolution and documentation of undesirable conditions. The CAP procedure encompasses investigation, corrective action determination, investigation report review and approval, action tracking, and issue analysis.

TMI-I will evaluate any RV Internals examination that fails to meet established acceptance criteria via the CAP.

5.7.3 Conclusion Exelon's corrective actions are consistent with the cOiTective actions in NUREG-1801 program XI.MI6A.

Page 5-6

ANI-'-L:~:1.L Rev. 001 5.8 AMP Element 8 - Confirmation Process 5.8.1 NUREG-1801, XI.M16A Confirmation Process Site assurance review and approval processes, and administrative controls are implemented in accordance with the 10 CFR Part 50, B, or their equivalent, as applicable. It is eXl/eeted that the in it4RP-227 will provide an acceptable level olqualityfiJr inl:np,ction flaw and other olaging management olthe PWR internals that are addressed in accordance with the 10 CFR Part 50, Appendix B, or their equivalent (as applicable), confirmation process, and administrative controls.

5.8,2 TMI-1 Confirmation Process Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with the requirements of 10 CFR 50. Appendix B. Effectiveness reviews are performed after implementation of the final corrective action to prevent recurrence or action; the reviews are perfonned after sufficient time has elapsed to challenge the corrective action to prevent recurrence or action. The effectiveness reviews are performed to detennine whether the associated corrective action or corrective action to prevent recurrence or other actions have eliminated the cause or reduced the recurrence to an acceptable level.

5.8.3 Conclusion Exelon's confirmation process is consistent with the confirmation process in NUREG-1801 program XI.M 16A.

5.9 AMP Element 9 - Administrative Controls 5.9.1 NUREG-1801 XI.M16A Administrative Controls The administrative controlsj(lr such programs, including their implementing procedures and review and approval processes, are under 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 ol documentation and administrative controls to ensure effective long-term implementation.

5.9.2 TMI-1 Administrative Controls Administrative controls, including Exelon-specific and TMI-l-specific documents used to implement this TMI-l RV Internals Inspection Plan, provide for a formal review and approval process. The Exelon Quality Assurance (QA) Topical Report (QATR) provides administrative controls for the review and approval process for TMI-I.

The QATR is based on 10 CFR 50, Appendix B.

5.9.3 Conclusion Exelon's administrative controls are consistent with the administrative controls in NUREG-1801 program XLMI6A.

5.10 AMP Element 10 - Operating Experience 5.10.1 NUREG-1801 XI.M16A Operating Experience Relativelyfew incidents ofPWR internals aging degradation have been reported in operating Us. commercial PWR plants. A summary olobservations to date is provided in Appendix A olAlRP-227-A. The applicant is expected to review subsequent operating experiencefiJr impact on its program or to participate in industry initiatives that perfiJrm thisfimction.

The application ofthe 1'vfRP-227 guidance will establish a considerable amount ofoperating experience over the nextfew years. Section 7 ofMRP-227 describes the reporting requirementsfiJr these applications, and the plan fill' evaluating the accumulated additional operating experience.

Page 5-7

5.10.2 TMI-1 Operating Experience AI~~-L~~~L. Rev. 001

('lfltlVif~lv few incidents of PWR intel11als degradation have been reported in operating U.S. commercial PWR plants. In B&W units, the only incidents have been in the baffle-to-baffle bolts, RV Intel11als high-strength bolting, and devices for the vent valve This operating (OE) was considered in the development of the MRP-227-A examination recIUIl:enlerlts.

The Exelon Operating Program provides direction for the receipt, processing, evaluation, implementation, and distribution of OE information to prevent or the consequences of similar events reported by both external and intel11al sources.

The TMI-I Reactor Internals AMP will report findings in accordance with Section 7 of MRP-227-A.

5.10.3 Conclusion TMI-I 's OE is consistent with the OE in NUREG-1801 program XLMI6A.

Page 5-8

MI~r-F~"L. Rev. 001 6.0

SUMMARY

AND CONCLUSIONS This report documents and provides a description of the TMI-I RV Internals Inspection Plan and how it relates to the AMP at TMI-l for the of effects consistent with previous commitments. This TMI-I RV Internals Plan is based on as applicable to TMI-I. Section 5.0 of this report demonstrates the TMI-I RV Internals AMP, which includes the TMI-I RV Internals Inspection Plan, is consistent with the ten program elements of NUREG-180 I, Revision 2 AMP XLM 16A.

This TMI-l RV Internals Inspection Plan contains a discussion of the background of the B&W-designed plant RV Internals programs, including operational experience, TLAAs, and TMI-l programs and activities.

The next removal of the RV Internals from the TMI-l RV is scheduled during the 2015 RFO. The appropriate examinations required by ASME B&PV Code Section XI, and the remainder of the initial examinations required by MRP-227-A are scheduled to be performed at that time. Any rclevant conditions will be documented and dispositioned in Exelon's CAP and reported to the industry.

The TMI-l RV Internals AMP will include this TMI-l RV Internals Inspection Plan and will manage the effects of for RV Internals components. This plan provides reasonable assurance that the RV Internals components will remain functional through the TMI-l LR period of extended operation.

Appendix D to this RV Internals Inspection Plan provides the Exclon responses to the Topical Report Conditions and Applicant/Licensee Action Items found in the NRC SER for MRP-227.

Appendix E of this report provides a comparison of the TMI-l LRA AMR and the results of the industry activity that lead up to the development ofMRP-227-A.

Page 6-1

AI\\I!-'-L:::I:1,L. Rev. 001

7.0 REFERENCES

Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 201 L 1022863.

NUREG-180 I, Revision 2, "Generic Lessons Learned (GALL) Report," Dated December, 2010.

3 Letter Transmitting Three Mile Island Nuclear Station, Unit I Application for Renewed Operating License.

January 8, 2008 (Accession No. ML0802202 I9) 4 Issuance of Renewed Facility Operating License No. DPR 50 for Three Mile Island, Unit I, October 22, 2009 (Accession No. ML0927 1040I) 5 U. S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the License Renewal of Three Mile Island Nuclear Station, Unit I," NUREG-I928, October 31,2009 (Accession Nos. ML092950449 and ML092950450) 6 Three Mile Island Nuclear Station, Unit I, License Renewal Application, January 8, 2008 (Accession No. ML0802202 I9) 7 BAW-2248, "Demonstration of the Management of Aging EfTects for the Reactor Vessel Internals," July 1997 8

U. S. Nuclear Regulatory Commission Letter, "Acceptance tor Referencing ofGeneric License Renewal Program Topical Report Entitled, 'Demonstration of the Management of Aging Effects for the Reactor Vessel Internals'," BAW-2248, July 1997. Accession No. ML993490288 9

BAW-2248A, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," April 2000. Accession No. ML003708443 10 Letter (from Christian Larsen) Report Transmittal: Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), EPRI Palo Alto, CA, 2008, 1016596, January 2009 (NRC Accession No. ML090I60204)

II Letter from Robert A Nelson (NRC) to Niel Wilmshurst (EPRI), "Revision I to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines" (TAC NO. ME0680)," December 16, 20 I I, NRC Accession No. ML I 1308A770 12 Letter from Pamela B. Cowen to U. S. Nuclear Regulatory Commission, "Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations Relief Request RR-09-0 I and RR-09-02," October 29, 2009, NRC Accession No. ML093020523 13 Letter from Nuclear Regulatory Commission to Michael J. PaciIio, "Three Mile Island Station, Unit I (TMI-I) ~ Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld and Internal Examinations, Proposed Alternative Request Nos. RR-09-0I and RR-09-02 (TAC Nos. ME2483 and ME2484)," September 21, 20 I0, NRC Accession No. ML I023900 I8 14 Letter from Peter J. Bamford, US NRC, to Michael J. Pacilio, "Three Mile Island Nuclear Station, Unit I (TMI-I) ~ Correction Letter Regarding Request to extend the Inservice Inspection Interval for Reactor Vessel Internal Examinations, RR-09-02 (TAC No. M2484)," November 5,2010, NRC Accession No. MLI029506I2 15 Ameriean Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, American Society of Mechanical Engineers, New York, NY 16 Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals Component Items (MRP-189-Rev. 1). EPRI, Palo Alto, CA: 2009. 1018292.

Page 7-1

B&W PWR Intemals (MRP-23 I-Rev. 2).

M,'<r-LnOJ.L.. Rev. 001 17 Materials Reliability l"rrlurcH1,,"

EPRI, Palo Alto, CA: 20 IO. 1021028.

18 Ameriean Soeiety of Mechanical Boiler and Pressure Vessel Code,Section III, Division I, "Core Support

" 1998 Edition through 2000 Addenda 19 Nuclear Institute, "Guideline for the of Materials

" NEI 03-08, Revision 2, January 2010, (NRC Accession No. ML I0288(028) 20 Materials Reliability Program: Inspeetion Standard tor PWR Internals (MRP-228). EPRI, Palo Alto, CA:

2009.1016609.

21 Materials Reliability Program: Failure Modes, Effects, and Criticality Analysis ofB&W-Designed PWR Internals (MRP-190). EPRI, Palo Alto, CA: 2006.1013233.

22 Materials Reliability Program: Functionality Analysis for Babcock & Wilcox Representative PWR Internals (MRP-229, Rev. 3). EPRI, Palo Alto, CA: 2010. 1022402.

23 "Briefing Material for June 14,2005 Mtg on Rx Vessel Internals from the EPRI MRP ~ HT Tang", June 14, 2005. NRC Aceession No. ML051710058 Page 7-2

Rev. 001 APPENDIX A: TMI-1 PRIMARY COMPONENTS Appendix A contains selected columns from Table 4-1, "B&W Plants Primary Components" from MRP-227-A modified to be TMI-I unit-specific. Items not to TMI-I are not induded in this Appendix.

Table A-1, TMI-1 Primary Components (based on Table 4-1 from MRP-227-A)

Item Plenum Cover Assembly &

Core Support Shield Assembly Plenum cover weldment rib pads Plenum cover support flange CSS top tlange Control Rod Guide Tube Assembly CRGT spacer casting Core Support Shield Assembly C55 vent valve top retaining ring CSS vent valve bottom retaining ring (Note I)

Effect (Mechanism)

Loss of material and associated loss of core clamping pre-load (Wear)

Cracking (TE), inclnding the detection of fractured spacers or missmg screws Cracking (TE), including i the detection of surf'lce irregularities. such as damaged, fractured materiaL or missing items Expansion Lin (Note 2)

None None None xamination Method/Frequency (Note 2)

One-time physical measurement no later than two refueling ontages from the beginning of the license renewal period.

Perf()rm subsequent visual (VT-3) examination on the 10-year lSI interval.

Visual (VT-3) examination during the next 10-year lSI.

Subseqnent examination on the lO-year ISf interval.

Visual (VT-3) examination during the next 10-year lSI.

Subsequent examinations on the 10-year lSI interval.

Examination Coverage Determination of dit1erential height of top of plenum rib pads to reactor vesscl seating surtrlce, with plenum in reactor vessel.

See Figure 4-1 ofMRP-227-A.

Accessible surfaces at eaeh of the 4 screw locations (at every 90°) of 100'% of the CRGT spacer castings (limited acceSSibility)

See Figure 4-5 ofi'v1RP-227-A.

100'x, of accessible surfaces.

See Page 4-3 and Table 4-1 of BAW-2248A See Figure 4-11 ofMRP-227-A PageA-1

A A;'.A-TMI-1

~",,,a, (based on Table 4-1 from

,m~ ~~.,

Item Core Support Shield Assembly Upper core barrel (UCB) bolts and their locking Core Barrel Assembly Lower core barrel (LCB) bolts and their locking devices Core Barrel Assembly Bame-to-Ic)rmer bolts Core Barrel Assembly Bame plates Effect (Mechanism)

Bolts: Cracking (SCC)

Locking Devices:

of material, damaged, distorted or locking devices (\\Vear or Fatigue damage by t"iled bolts).

Bolts: Cracking (SCC)

Locking Devices: Loss of material. damaged, distorted, or missing locking devices (Wear or Fatigue damage by I"i led bolts)

Cracking (IASCC, IE, Overload) (Note 4)

Cracking (IE), including the detection of readi Iy deteetable eraeking in the battle plates Expansion link (Note 2)

UTS bolts and LTS bolts and their locking devices Lower grid shock pad bolts and their locking devices UTS bolts and LTS bolts and their locking devices Lower grid shock bolts aud their locking Baft1e-to-baft1e bolts, Core barrel-to-Iorrner bolts Core barrel cyl inder (including vertieal and circumferential seam welds)

Former plates Examination Method/Frequency (Note 2)

Volumetric examination (UT) of the bolts within two refueling outages from 1/1/2006 or next 10-year lSI interval, whichever is first.

Subsequent examination on the 10-year lSI interval unless an evaluation of the baseline results submitted for NRC staff justifies a longer interval examinations.

Visual (VT-3) examination of bolt locking devices on the 10-year lSI interval.

Volumetric examination (UT) of the bolts during the next 10-year lSI interval Irom 1/1/2006, Subseqnent examination on the 10-year lSI interval unless an evaluation of the baseline results submitted lor NRC stall' approval justifies a longer interval between examinations.

Visual (VT-3) examination of bolt loeking deviees on the IO-year lSI interval.

Baseline volumetric examination (UT) no later than two refueling outages rrom the beginning of the lieense renewal period Iwith subsequent examination after 10 I additional years, I Visual (VT-3) examination during the I next 10-year lSI.

Subsequent examinations on the IO-year lSI interval.

Examination Coverage 100% of accessible bolts and their locking deviccs (Note 3)

See Figure 4-7 of rvtRP-227-A.

100% of accessible bolts and their loeking devices (Note 3).

See Figure 4-8 of rvtRP-227-A.

100':;', of accessible bolts (Note 3).

See Figure 4-2 of rvtRP-227-A, 100% of the accessible surf"ce witbin I ineh around eaeh Ilow and bolt hole, See Figure 4-2 ofrvtRP-227-A.

Page A-2

A-TMI-1 Item Core Barrel Assembly devices, including welds, of baffle-to, li,rmer and internal baffle-to,baffle bolts Flow Distributor Assembly Flow distributor (FD) bolts and their locking devices Lower Grid Assembly Alloy X,750 dowel-to-gnide block welds Ineore Monitoring Instrumentation (IMI)

Guide Tube Assembly IMI guide tube spiders IMI guide tube spider,to, lower grid rib section welds Notes:

Bolts: Cracking (SCC)

Locking Devices: Loss of ma.terial, damaged or distorted or missing locking devices (Wear or damage by bolts),

Cracking (SCC),

inclnding the detection of or welds, or dowels Cracking (TE/IE),

including tbe detection of Iractured or missing spider am1S Of, Cracking (IE), including separation ofspider arms from the lower grid rib section at the weld including welds, li,r the haf'flc,to,baffle holts and Core barrel,to, former bolts UTS bolts and LTS bolts and their locking devices.

Lower grid shock bolts and their locking Alloy X*750 dowel locking welds to the upper and lower fuel assembly support pads Lower grid fuel assembly support pad items: pad, pad*to-rib section welds, Alloy X,750 dowel, cap screw, and their locking welds (Note: the pads, dowels, and cap screws are included because of IE ofthe welds) ethod/Frequency (Note 2)

Visnal (VT,3) examination dnring the next IO,year lSI, Subseqnent examinations on the II),year lSI interval.

Volumetrie examiuatiou (UT) of the bolts the next ((),year lSI interval Irom Subsequent examination on the lSI interval unless an evalnation baseliue results, submitted fi,r NRC staff approval, justifies a louger interval bet\\vecn examinations.

Visual (VT,3) examination of bolt locking devices ou the IO,year lSI interval.

Initial visual (VT,3) examination no later than two refueling outages from the beginning of the Iieense renewal period.

Subsequent examination on the IO-year lSI interval.

Initial visual (VT-3) examination no later than two refueling outages ti'om the beginning of the license renewal period, Subsequent examinations on IO*year lSI

interval, Rev, 001 Examination Coverage i00% of accessible baftle,to, fimner and internal baffle-to-baffle bolt locking devices (Note 3),

See Figure 4,2 of MRP,227,k IOO')/" of accessible bolts and their locking devices (Note 3).

See Figure 4,8 of MRP,227,A.

Accessible snrfaces of IOO';!"

of the 24 dowel,to,guide block welds, See Fignre 4-4 ofMRP-227,;\\'

100% of top surt;!ces of 52 spider castings and we Ids to the adjacent lower grid rib section.

See Figures 4,3 and 4*6 of MRp*227-A.

A verification of the operation ofeaeh vent valve shall also be performed through manual aetuation of the valve. Verify that the valves are not slUck in the open position and that no abnormal degradation has oecurred. Examine the valves for evidence of scratches, pitting, embedded particles, leakage of the seating surfaces, craeking of lock welds and locking eups, jaek serews for proper position, and wear. The frequeney is defined in TMI-I 's teehnical speci fications.

Page A-3

A-TMI-1 on Table 4-1 from Examination acceptance criteria and expansion criteria arc in Appendix C of this report.

A minimum of 75'% of the total population fexamined + unexamined), including coverage consistent with the Expansion criteria in Appendix C, must be examined for inspection 4.

The primary aging mechanisms for loss ofioint tightncss for this item arc IC and ISR. Fatigue and Wear, which can also lead to cracking, arc secondary mechanisms after stress relaxation and loss of preload has occurred due to IC/ISR. Bolt stress relaxation cannot readily inspected by NDE. Only bolt is inspected by UT inspection. The effect of loss ofjoint tightness on the functionality will be addressed by analysis of the core barrel assembly, which will be performed to address Applicant/Licensee Action Item #6 of the NRC Safety Evaluation for MRP-227 (Sec Appendix D of this reportt.

Page A-4

Rev. 001 APPENDIX B: TMI-1 EXPANSION COMPONENTS Appendix B eontains selected columns from Table 4-4, "B&W Plants Expansion Components" from l'vlRP-227-A, modified to be Tl'vll-l unit-spceific.

Items not applicable to Tl'vll-]

not included in this /\\ppendix.

Table 8-1, TMI-1 Expansion Components (based on Table 4-4 from MRP-227-A)

Item Upper Grid Assembly Alloy x-7S0 dowel-to-upper grid fuel assembly support pad welds Core Barrel Assembly Upper themJaI shield (UTS) bolts and their locking devices Core Barrel Assembly Core barrel cylinder (inclnding vertical and circUlllti~rential seam welds)

Former plates Core Barrel Assembly Baftle-to-baffle bolts Core barrel-to-t()rmer bolts Effect (Mechanism)

Cracking (SCC) including the detection of or missing welds. or missing Bolts: Cracking (SCC)

Locking Devices: Loss of

material, distorted or locking devices (Wear or Fatigue damage by illiled bolts).

Cracking (IE) including readily detectable cracking Cracking (lASCC, IE, Overload)

(Note 3)

Primary Link (Note 1)

Alloy X-7S0 dowel-to-guide block welds UCB. LCB or FD bolts and their locking devices Baffle plates Baffle-to-t()rmer bolts Examination Method/Frequency (Note 1)

Visual (VT-3) examination.

Subsequent examinations on the 10-year lSI interval unless an applicant/licensee provides an evaluation for NRC statf approval that Justities a longer interval between inspections.

Bolt: Volumetric examination (UT)

Locking Devices: Visual (VT-3) examination Subsequent examinations on the 10-year 1St interval unless an applicant/licensee provides an evaluation for NRC statf approval that justifies a longer interval between inspections.

No examination requirements.

Justify by evaluation or by replacement Internal battle-to-baffle bolts:

No examination requirements, I Justify by evaluation or by I replacement Examination Coverage Accessible surfilees of 1OO')!" of dowel locking welds.

See Figure 4-6 of MRP-227 A, (i.e.* these are 5i mi lar to the lower grid fuel assembly support pads).

IOO'X) of aecessible bolts and their locking devices (Note 2)

See Figure 4-7 of MRP-227-A.

1naccessible See Figure 4-2 ofMRP-227-A.

An acceptable examination technique currently not available See Figure 4-2 ofMRP-227-A.

Page B-1

x B-TMI.1 F.

(based on Table 4-4 from Ohn ~~~

,n ~hrh Rev. 001 Item Core Barrel Assembly Locking devices, inclnding locking welds, f(Jr the external batl1e-to-batl1e bolts and cnre barrel-to-tt)flner bolts Lower Grid Assembly Lower grid fuel assembly support pad items: pad, pad-to-rib sections welds, Alloy dowel, eap screw, and tbeir locking welds (Note: the pads, dowels, and screws are included because IE ofthe welds)

Lower Grid Assembly Alloy X-750 dowel-to-Iower grid fuel assembly support pad welds Effect (Mechanism)

Cracking (IASCC, IE)

Cracking (IE), including the detection of separated i or missing welds, missing support pads, dowels, cap screWS and locking welds, or misalignment or the support pads Cracking (SCC),

including the detectiou of separated or missing locking welds, or missing dowels I

I Primary Link (Note 1)

Lncking devices.

including locking welds, ofbattle-to-ttJrmer bolts or internal baffle-to-baffle bolts IMI b'llide tube spiders and spider-to-lower grid rib section welds X-750 dowel-to-block welds Examination Method/Frequency (Note 1)

External batl1tHo-battle bolts, core barrel-to-Itmner bolts:

No examination requirements, Justify by evaluation or by replacement No examination requirements.

Justify by evalnation or by replacement Visual (VT-3) examination Subsequent examinations on the 10-year lSI interval unless an applicant/licensee provides an evaluation for NRC statfapproval that justities a longer interval between inspections.

Visual (VT-3) examination Snbsequent examinations on the 10-year lSI interval nnless an applicant/licensee provides an evaluation for NRC staffapproval thatJustities a longer interval between inspections.

Examination Coverage Inaccessible See Figure 4-2 ofMRP-227-A.

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

Accessible surfaces of the pads, dowels, and cap screws and associated welds in 100% of tile lower grid fuel assembly support pads.

See Figure 4-6 of MRP-227-A.

Accessible snrfaces of IOO'y" of the support pad dowel locking welds.

See Fignre 4-6 ofMRP-227-A.

Page B-2

A B~ TMI-1 Item (based on Table 4-4 from.mn """'

A Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

,n "he" Rev. 001 Examination Coverage Lower (;rid Assembly Lower grid shock pad bolts and their locking devices Lower Grid Assembly Lower thermal shield (LTS) bolts and their locking devices Notes:

Bolts; Crackmg (SCC) liCB, LCB or FD bolts and their locking devices Locking Devices; Loss of material.

distorted or locking devices (Wear or Fatigue damage by f'liled holts).

Bolts: Cracking (SCC) liCB. LCB or FD bolts and their locking devices Locking Devices: Loss of material.

distorted or locking devices (Wear or Fatigue damage by f'liled bolts).

Bolt: Volumetric examination (UT)

Locking Devices: Visual (VT-})

examination Subsequent examinations on the 10-year lSI interval unless an aPl,Ii,:antilicelnsc:c provides an ev,liwlticm for NRC staffapproval thatJustif!es a longer interval between Bolt: Volumetric examination (UT)

Locking Devices: Visual (VT*})

examination Subsequent examinations on the 10-year lSI interval unless an applicant/licensee provides an evaluation for NRC stafTapproval that justifies a longer interval between inspections.

100%, of accessible bolrs and their loeking devices (Note 2)

See Figure 4*4 ofMRp*227*A.

100% of accessible bolts and their locking devices (Note 2)

Figure 4*8 ofMRp*227*A 1.

Examination acceptance criteria and expansion criteria are in Appendix C of this report.

2.

A minimum of 75'% of the total population (examined unexamined) must be examined for inspection credit.

3.

The primary aging degradation mechanisms for loss ofjoint tightness for these items are IC and ISR. Fatigue and Wear, which can also lead to craeking, are secondary aging degradation meehanisms after signiticant stress relaxation and loss of preload has occurred due to IC/ISR. Bolt stress relaxation cannot readily be inspeeted by NDE. Only bolf cracking could be inspected by UT inspection if it were possible for these bolts. Therefore, the effects of loss ofjoint tigbtness and/or cracking on the functionality of these bolts relative to the entire core barrel assembly will be addressed by analysis of the core barrel assembly, which will be performed to address Applicant/Licensee Action Item #6 of the NRC Safety Evaluation for MRP-227 (See Appendix D of this report).

Page B-3

001 APPENDIX C: TMI*1 EXAMINATION ACCEPTANCE AND EXPANSION CRITERIA A[lp"nCllX C contains selected columns from Table 5~ I, "13&W Plants Examination Acceptance and Expansion Criteria" from MRP-227 -A. Items not aPlllic:able to TMI-l are not included in this Appendix.

Table C-1, TMI-1 Examination Acceptance and Expansion Criteria (based on Table 5-1 from MRP-227-A)

[tern Plenum Cover Assembly

& Core Support Sbield Assembly Plenum cover weldment rib pads Plenum cover support tlange CSS top flange Examination Acceptance Criteria (Note 1)

One-time physical measurement. In addition. a visual (VT-3) examination is condncted ftlr these items.

The measured differential height from the top of the plenum rib pads to the vessel sealing surface shall average less than 0.004 inches compared to the built condition.

The relevant condition ftlr these items WC,ir that may lead to a loss of function.

Expansion Link(s)

None NIl'.

Expansion Criteria Additional Examination Acceptance Criteria NIl'.

Control Rod Guide Tube Assembly CROT spacer castings Core Support Shield Assemhly CSS vent valve top retaining ring CSS vent valve bottom retaining ring The specific relevant condition lor the None VT-3 of the CROT spacer castings is evidence of fractured spacers or missing screws.

Visual (VT-3) examination.

None The specific relevant condition is evidence of damaged or fractured retaining ring material, and missing items.

NIl'.

NIA N/A NIA Page C-1

C-TMI-1 Examination and Criteria (based on Table 5-1 from

,,~~ MM~

Item Core Support Shield Assemhly core barrel (UCB) and their locking devices Examination Acceptance Criteria (Note 1)

I) Volumetric (UT) exammation of the UCI3 bolts.

The examination acceptance criteria tor the UT of the UCB bolts shall be established as a part of the examination technical Justificatiou.

Visual (VT-3) examination of the UCB bolt locking devices.

specific relevant condition f()[ the VT-3 of the UCB bolt locking devices is evidence of broken or missing bolt devices.

Expansion Link(s)

UTS bolts and, LTS bolts and their locking devices Lower grid shock pad bolts and tbeir locking devices Expansion Criteria I) Confirmed nnaeceptable indications ex(;eedinlg 10')1, of the UCB bolts shall require the UT examination be expanded by the completion of the next rctueling outage to include:

100% of the accessible UTS bolts and 1000;" of the accessible LTS bolts and.

100% of the accessible lower grid shock pad bolts.

2) Confirmed evidence of relevant conditions exceeding 10% of the UCB bolt locking devices shall require that the VT-3 examination be expanded by the completion of the next refueling outage to include:

100% of the accessible UTS bolt locking devices and 100% of the accessible LTS bolt locking devices and 100% of the accessible lower grid shock pad bolt locking devices, Additional Examination Acceptance Criteria I) The examination ac,;er.tallce criteria for the UT expansIOn bolting shall be established as part oftbe examination technical justification.

2) The specific relevant condition tor the VT-3 of the expansion locking devices is evidence of broken or missing bolt locking devices.

Page C-2

Item Examination Acceptance Criteria (Note 1)

Expansion L1nk(s) from Expansion Criteria Additional Examination Acceptance Criteria

2) Visual (VT*}) examination of the LCn bolt locking devices.

The specific relevant condition for the VT*} of the LCB bolt locking devices is eVIdence of broken or missing holt devices.

The examination acceptance criteria fbr the UT of the LCB bolts shall be establisbed of the examination technical ju:,tiJ*iC'lli'>n.

Core Barrel Assembly Lower core barrel (LCn) bolts and their locking devices I) Volnmetric (UT)

LCB bolts.

of the bolts and LTS bolts and their locking devices Lower grid shock pad bolts and their locking devices I) Confirmed nnacceptable indications ex,:eedinlg 10% of the LCn bolts shall reqUire the UT examination be expanded by the completion of the next refueling ontage to inclnde:

100°!., of the accessible UTS bolts and 100'Yo of the accessible LTS bolts and 100%, of the accessible lower grid shock pad bolts

2) Confirmed evidence of relevant conditions exceeding 10%, of the Lcn bolt locking devices shall require that the VT*} examination be expanded by the completion of the next refueling outage to include:

100% ofthe accessible UTS bolt devices and 100%, of the ac,,,",itlle LTS bolt locking devices and IOO'!I" of the accessible lower grid shock pad bolt locking devices.

I) The examination acceptance criteria for the UT of the expansion holting shall be estabhshed as part of the examination technical justification.

2) The specific relevant condition for the VT*} of the expansion locking deviccs is evidence of broken or missing bolt locking devices>

Page C-3

C-TMI-1 and Criteria (based on Table 5-1 from Item Core Barrel Assembly Bafl1e-to-flmner bolts Core Barrel Assembly Baffle plates Core Barrel Assembly Locking devices, including locking welds, of baffle-to-former bolts and internal baffle-to-baffle bolts Examination Acceptance Criteria (Note 1)

Baseline volnmetric (UT) examination oftlie bafl1e-to-former bolts.

The examination acceptance criteria for the liT of the baffle-to-f(mner bolts shall be estahlished as part of the examination technical justification, Visual (VT-3) examination.

The relevant condition is detectable cracking in tlie baffle plates.

Visual (VT-3) examinatiou.

The specifie relevant condition is missing, non-functional, or removed locking devices including locking welds.

Expansion Link(s)

Baffle-to-baffle bolts.

Core barrel-to-former bolts

a. Former plates

b. Core barrel (including and circumferential seam welds)

Locking devices, inclnding locking welds. for the external baffle-to-baffle bolts and core barrel-to-fornler bolts Expansion Criteria Confirmed unacceptable indications in greater than or equal to (or 43) of the bafl1e-to-former bolts. provided that none of the unacceptahle bolts are on f"t)fmer elevations 3. 4, and 5, or greater than 25%1 of the bolts on a single baffle plate, shall require an evaluation of the internal baffle,to-baffle bolts for the purpose of determining whether to examine or replace the internal baffle,to,baffle bolts. The evaluation may include external baffle-to-baffle bolts and core barrel-to-former holts for the purpose ofdetermining whether to replace them.

a and b. Confirmed cracking in multiple (2 or more) locations in the baffle plates shall require expansion, with continned operation of former plates and the core barrel eyhnder justified by evalnation or by replacement by the eompletion of the next refueling outage.

Confirn1ed relevant conditions in greater than or equal to 1% (or I I) of the baffle-to-fornler or intemal baf11e-to-batTle bolt locking devices, including locking welds, shall require and evaluation of the external batTle-to-baffle and core barrel-to-fonner bolt locking devices for the purpose of determining continued operation or replacement.

Additional Examination Acceptance Criteria N/A a and b: NIA N/A Page CA

C~TMI*1 and Criteria (based on Table 5*1 from

.. ~ ~,'~~ Rev. 001 Item Lower Grid Assembly Alloy X-750 dowel*to*

guide block welds I Examination Acceptance Criteria I

(Note 1)

Initial visual (VT-3) examination The specific relevant condition is separated or missing lockil1gweld~ or missing doweL Expansion Unk(s)

Alloy X-750 dowel locking welds to the upper and lower grid fuel assembly support pads Expansion Criteria Confirmed evidence of relevant conditions at two or more locations shall require that the VT-3 examination be expanded to include the Alloy X-750 dowel locking welds to the upper and lower grid fuel assembly support pads by the completion of the next refueling outage.

Additional Examination Acceptance Criteria The specific relevant condition for the VT-3 of the expansion dowel weld is separated or locking weld, or missing dowel.

Flow Oistributor Assembly Flow distributor (FD) bolts and their locking devices I

() Volumetric (UTi examination of the FDbolls.

The examination acceptance criteria fbr the UT of the FD bolts shall he established as part of the examination technical Justification,

2) Visual (VT-3) examination of the FD bolt locking devices.

The specific relevant condition for the VT-3 of the FD boll locking devices is evidence of broken or missing bolt locking devices.

UTS bolts and LTS bolts and their locking devices Lower grid shock pad bolts and their locking devices I) Confirmed unacceptable indications excecding IO'y" of the FD bolls shall require that the UT examination be expanded by the completion of the next refueling outage to include:

1000;" of the aceessible UTS bolts and 100% of the accessible LTS bolts and 100% of the accessible lower grid shock pad bolts.

2) Confirmed evidence of relevant conditions exceeding 100;" of the FD bolt locking devices shall require that the VT-3 examination be expanded by the completion ofthe next refueling outage to include:

100% of the accessible UTS bolt locking devices and 100% of the accessible LTS bolt locking devices, and 100% of the accessible lower grid shock pad bolt lorkin!! devices.

I) The examination acceptance criteria tor the liT of the expansion bolting shall be established as part of the examination technical jnstification.

2) The specific relevant condition tor the VT-3 of the expansion locking devices is evidence of broken or missing bolt locking devices.

Page C-5

C-TMI-1 and Criteria (based on Table 5-1 from

,~~ Rev. 001 Item Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Additional Examination Acceptance Criteria fhe specific relevant conditions for the IMI guide tube spiders arc fractured or spider arms.

T'he relevant conditions fl.)r the IMI grid rib section welds are separated or missing welds.

Incore Monitoring Instrumentation (IMI)

Guide Tube Assembly IMI guide tube spiders IMI guide tube spideHo~

lower grid rid section welds Notes:

Initial visual examination Lower filel gnd assembly support pad items: pad, pad~to~rib section welds, Alloy X~750 dowel, cap screw, and their locking welds Confirmed evidence of relevant conditions at two or more IMI guide tube spider locations or 1M I guide tube to lower grid rib section welds require that the VT ~3 examination be expanded to include lower fuel assembly support pad items by the completion of the next refueling outage.

The specific relevant conditions for the VT~3 of the lower grid fuel assembly support pad items (pads, pad*to~rib section welds, Alloy X~

750 dowels, cap screws, and their locking welds) are separated or missing welds, missing support pads, dowels, cap screws and locking welds, or misalignment of the support pads.

I.

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

Page C~6

Rev. 001 APPENDIX D: TMI-1 RESPONSE TO THE TOPICAL REPORT CONDITIONS AND THE APPLICANT/LICENSEE ACTION ITEMS FROM THE NRC SER, REVISION 1, FOR MRP-227, REVISION 0 Topical Report Conditions and Limitations:

These conditions were incorporated into MRP-227~A, which was in tum incorporated into this TMI-I RV Intemals Inspection Plan.

Table 0-1, Topical Report Condition/Limitation (from the NRC SER for MRP-227)

Topical Report Condition/Limitation SER Section I

4.1.1 2

4.1.2 3

4.1.3 Topical Report Condition/Limitation The examination coverage and re-examination frequency requirements for specific Westinghouse-designed and CE-designed RV Intemals components shall be as addressed when publishing MRP-227-A. Tables 4-5, and 4-6 shall be revised accordingly.

To ensure that the structural integrity and functionality of specific Westinghouse-designed and CE-designed RV Internals components are maintained under all licensing basis conditions during the period of extended operation, the staff has determined that each of these components shall be included in the "Primary" inspection category in the NRC-approved version ofMRP-227.

To ensure that the structural integrity and functionality of these RV Intemals components [t1ow distributor to shell forging bolts] are maintained under transient loading conditions during the period of extended operation, the staff has determined that the subject components shall be included in the "Primary" inspection category in the NRC-approved version ofMRP-227. The examination methods shall be consistent with the MRP's recommendations for these components, the examination coverage for the aforementioned components shall conform to the criteria as described in Section 3.3.1 of the SE for MRP-227, and the re-examination frequency shall be on a IO-year interval similar to other "Primary" inspection category components.

TlVH-l Response This item is not applicable to TMI-I; there are no B&W intemals components identified in this eondition/limitation, only Westinghouse and CE components.

This item is not applicable to TMI-I; there are no B&W intemals components identified in this condition/limitation, only Westinghouse and CE components.

MRP-227-A Table 3-1 has been modified to designate the flow distributor bolts as Primary category components.

MRP-227-A Table 4-1 (B&W Primary Components) has been modified to include the flow distributor bolts, along with the required examination method and coverage.

In addition, the B&W LCB bolts, flow distributor (FD) bolts, and their locking devices were deleted as Expansion Links for the DCB bolts. Note 3, about the "Examination Method/Frequency" column entry for the LCB bolts and FD bolts, has also been deleted.

MRP-227-A Table 4-2 is not applicable to TMI-1.

MRP-227-A Table 4-4 has been modified to delete the flow distributor bolts.

MRP-227-A Table 4-5 is not applicable to Page 0-1

""!J!J"'"UI?> 0: TMI-1

;>!JVII.;> to MRP-227 Safety Evaluation items Topical Report Condition/Limitation A~ln ~C5~, Rev. 001 SER Section Topical Report Condition/Limitation In addition, in MRP-227 Table 4-1, the TMI-I.

MRP included the B&W LCB bolts, flow distributor (FD) bolts, and their locking devices as applicable "Expansion I:

11 components for B&W upper core barrcl (DCB) bolts and their locking devices. Note 3 indicated this expansion was only applicable if the primary inspections of the LCB bolts or FD bolts had not yet been conducted. However, since the LCB and FD bolts are already included as "Primary" inspection category components in MRP-227 Table 4-1, it is inappropriate to include the LCB and FD bolts in the "Expansion Link" column references. Specifically, the staff determined that this note could be interpreted to mean that B&W licensees would not need to perform the primary inspeetions of the LCB or FD bolts at their next scheduled opportunity because, as "Expansion" components, the inspections of LCB or FD bolts would only be performed ifaging was detected from the "Primary" inspections of the DCB bolts. Therefore, the staff requires the MRP to move the reference for Note 3 to the "Examination Method/Frequency" column entry for the LCB bolts and FD bolts in MRP-227 Table 4-1 and should delete the "Expansion Category" reference for the LCB bolts and FD bolts from the "Expansion Link" column of the DCB bolt line item in that table.

In Table 4-1 ofMRP-227, the MRP identified that the pads, pad-to-rib section welds, and Alloy X-750 dowel cap screws and their locking devices in the lower grid fuel assembly were applicable "Expansion" category components for the Primary inspections that would be performed on the CRGT spacer castings.

The relevant mechanism for the spaeer castings is thermal embrittlement, while the relevant mechanism for the pads, pad-to-rib section welds, and Alloy X-750 dowel cap screws and their locking devices in the lower grid fuel assemblv is TMI-I Response Page 0-2

An'"I<:>nrliv D: TMI-1 Topical Report Condition/Limitation to MRP-227 Evaluation items AI'JI""'-L':-J:J.L, Rev. 001 SER Section 4

4.1.4 Topical Report Condition/Limitation irradiation embrittlement. Consistent with SE Sections 3.3.7 and the CRGT spacer are subject to the Applicant/Licensee Action Item No.7 on the MRP-227 methodology.

When publishing the approved version of MRP-227, Tables 3-1,4-1, 4-4, and 4-5 shall be revised accordingly.

As discussed in Section 3.3.1 and Section 3.3.2 of this SE, for "Primary" inspection category components,MRP-227 will require that 100 percent ofa "Primary" inspection category component's accessible inspection area or volume be examined and 75 percent of a "Primary" inspection category component's total (accessible inaccessible) inspection area or volume be examined or, when addressing a sct oflike components (e.g.,

bolting), that the inspection examine a minimum sample size 01'75 percent of the total population oflike components. For the inspection of a set of Iike components, 100 percent ofthe accessible volume/area of each accessible like component will be examined. This defines the minimum inspection required to meet the intent of MRP-227 provided that no defects are discovered during the inspection. If defects are discovered during the inspection, the licensee shall enter that information into the plant's corrective action program and evaluate whether the results of the examination ensure that the component (or set of like components) will continue to meet its intended function under all licensing basis conditions of operation until the next scheduled examination.

The maximum number of like components possible will be inspected (i.e., if95 percent of the population is accessible for inspection then 95 percent must be inspected). This condition does not apply to components having a predefined scope of inspection less than 100 percent of accessible components, area, or length. Examples of such TMI-l Response MRP-227-A Table 4-4 (including Note 2),

has been modified to include this requirement.

MRP-227-A Tables 4-5, and 4-6 are not applicable to TMI-I.

Page D-3

">IJIJCIIUl" 0: TMI-1

,C0IJVIIJC to MRP-227 Safety Evaluation items Topical Report Condition/Limitation ANP-2952, Rev. 001 SER Section 5

4.1.5 Topical Report Condition/Limitation ex<:eptto11s include welds for which the inspection requirements call for only a certain of weld above and below the core mid-plane to be inspected, or components where the inspection addresses a predefined sample portion of the population; for example the inspection requirements for the Westinghouse control rod plates (cards) which calls for a 20 percent sample inspection.

Another situation where the condition would not apply is a component for which 100 percent of the population or area must be inspected and any lesser percentage of coverage wonld be unacceptable. This condition also is understood not to apply where there is a known access limitation (generic to all plants of the NSSS type) such that the population of components or the area/volume accessible for inspection is known to be less than 75 percent of the total.

As discussed in Section 3.3.2 of this SE, an equivalent requirement shall be imposed for the inspection of components in the MRP-227 "Expansion" inspection category.

When the approved version ofMRP-227 is published, Tables 4-4, 4-5, and 4-6 shall be updated to include this requirement.

The staff has determined that the NRC-approved version ofMRP-227 shall specify a 10-year inspeetion frequency for these components [baffle-to-former bolts]

following the initial or baseline inspection unless an applicant/licensee provides an evaluation for NRC staff approval that justifies a longer interval between inspections. MRP-227 Tables 4-1, 4-2, and 4-3 shall be modified when the approved version of MRP-227 is published to reflect this change.

TlVlI-l Response MRP-227-A, Table 4-1 was modified to require re-inspection of the baffle-to-former bolts at 10 year intervals. The change was also made in Appendix A to the TMI RV I

Internals Inspection Plan.

MRP-227-A, Tables 4-2 and 4-3 are not applicable to TMI-l.

Page 0-4

D: TMI-1

~t:7"'fJUII"'t:7 to MRP-227 Safety Evaluation items Topical Report Condition/Limitation II I

SER Section 6

4.1.6 7

4.1.7 Topical Report Condition/Limitation Tables and 4-6 shall be modified when the approved version of MRP-227 is published to apply a baseline I re examination interval to all "Expansion"

~ ly components (once degradation is identified in the associated "Primary" inspection category component and examination of the "Expansion"

-0 'y component commences) unless an applicant/licensee provides an evaluation for NRC stafr approval that justifies a interval between inspeetions.

When the approved version ofMRP-227 is published, MRP-227, Appendix A shall be updated to include a reterence to AMP XLM I6A in NUREG-180 I, Revision 2 (or in subsequent revisions of the GALL report that follow) and the Operating Experience Summary.

TMI-l Response MRP-227-A, Table 4-4 was modified to require subsequent examination of expansion components on the 10-year lSI intervals unless an applicant/licensee provides an evaluation for NRC stafr approval that justifies a longer interval between inspections.

MRP-227-A, Tables 4-5 and 4-6 are not applicable to TMI-1.

MRP-227-A, Appendix A has been updated and now references AMP XLM I6A in NUREG-180 I, Revision 2 (or subsequent revisions) for guidance on preparing an aging Management Program. Appendix A of MRP-227-A now contains the Operating Experience Summary.

Page D-5

Am..",r,rliv 0: TMI-1 R",',nn,n",,,, to MRP-227 Evaluation items

~I\\J~-/~~/. Rev. 001 Table 0-2, Applicant/Licensee Action Items (from the NRC SER for MRP-227)

Applicant/Licensee Applicant/Licensee Action Item Action Item #

SER Section TlVH-I Response I

4.2.1 2

4.2.2 Each applicant/licensee is responsible for

~ its plant's design and operating history and demonstrating that the approved version of MRP-227 is applicable to the facility. Each applicant/licensee shall refer, in particular, to the assumptions regarding plant design and operating history made in the FMECA and functionality analyses for reactors of their design (i.e., Westinghouse, CE, or B&W) which support MRP-227 and describe the process used for determining plant-specific differences in the design of their RVI components or plant operating conditions, which result in different component inspection categories. The applicant/licensee shall submit this evaluation for NRC review and approval as part of its application to implement the approved version ofMRP-227.

Each applicant/licensee is responsible for identifying which RVI components arc within the scope of LR for its facility.

Applicants/licensees shall review the information in Tables 4-1 and 4-2 in MRP-189, Revision I, and Tables 4-4 and 4-5 inMRP-191 and identifY whether these tables contain all of the RV I components that arc within the scope of LR for their facilities in accordance with 10 CFR 54.4. If the tables do not identify all the RVI components that are within the scope of LR for its facility, the applicant or licensee shall identify the missing component(s) and propose any necessary modifications to the program defined in MRP-227, as modified by this SE, when submitting its plant-specific AMP. The AMP shall provide assurance that the effects of aging on the missing component(s) will be managed for the period of extended operation.

TMI-l is bounded by the plant design and operating history assumptions in MRP-227-A, the FMECA (MRP-190) and the B&W design functionality analysis (MRP-229) as addressed in Section 4.2.2.1 ofthis TMI RV Internals Inspection Plan.

A review of the TMI-l License Renewal Project documentation shows that all Reactor Vessel Internals sub-components in the scope ofLR were included in LRA Table 3.1.2-3; i.e. no components were screened out during the AMR. Appendix E of this inspection plan provides a Iisting of components in the TMI-l LRA Table 3.1.2-3 and where those components are addressed in the MRP-227-A supporting documents (MRP-189 and MRP-231 ).

The only component identified that is not included in the MRP-227-A development is the vent valve locking device. A review of this component using the MRP-227-A development methodology will be performed and will be submitted to the NRC upon completion.

Page 0-6

Ll.",,,,,,r,rli,, 0: TMI-1 R",',nn,n<::", to MRP-227 Evaluation items

,""""'-/':7,,£,, Rev. 001 Applicant/Licensee Applicant/Licensee Action Item Action Item #

SER Section TMI-1 Response 3

4.2.3 4

4.2.4 Appllcarlts/"llc(:ns1ccs of CE and IXI 0

are 1

to perform plant-specific

.1 either to the acceptability of an

.,/'

eXlstlJl og programs, or to identify ch:ln~~es to the programs that should be implemented to manage the of these components for the period of extended operation. The results of this plant-spe:cl1'-ic and a description of the plant-specific programs being relied on to manage of these components shall be submitted as part of the applicant's/licensee's AMP application.

The CE and Westinghouse components identified for this type of plant-specific evaluation include: CE thermal shield positioning pins and CE in-core instrumentation thimble tubes (Section 4.3.2 in MRP-227, Revision 0), and Westinghouse guide tube support pins (split pins) (Seetion 4.3.3 in MRP-227, Revision 0).

B&W applicants/licensees shall confirm that the core support structure upper flange weld was stress relieved during the original fabrication of the RPV in order to confirm the applicability of MRP-227, as approved by the NRC, to their facility. If the upper flange weld has not been stress relieved, then this component shall be inspected as a "Primary" inspection category component. If necessary, the examination methods and frequency for non-stress relieved B&W core support structure upper flange welds shall be consistent with the recommendations in MRP-227, as approved by the NRC, for the Westinghouse and CE upper core support barrel welds. The examination coverage for this B&W Hange weld shall conform to the staff's imposed criteria as described in Sections 3.3.1 and 4.3.1 of this SE. The applicant's/licensee's resolution of this plant-specific action item shall be submitted to the NRC for review and approval.

This item is not applicable to TMI-I; there are no actions for B&W internals identified in this action item, only Westinghouse and CE internals.

Original fabrication records have confinued that the TMI-l CSS upper Hange weld was stress relieved (See Section 4.1.5 of this report). The CSS upper f1ange weld does not need to be inspected as a "Primary" component.

Page 0-7

'"\\fJfJt:::IIUIA 0: TMI-1

~t:::"fJVII"t::: to MRP-227 Safety Evaluation items I

Applicant/Licensee Applicant/Licensee Action Item TMI-l Response Action Itern #

SER Section

~~IP-2Q~? Rev. 001 5

4.2.5 6

4.2.6 As addressed in Section 3.3.5 of this aplJIi<;ants/llicc~ns,ees shall identify plant-spc:cit'-ic acceptance criteria to be applied when fJC

'0 the physical measurements required by the NRC-approved version of MRP-227 for loss of compressibility for Westinghouse hold down "mine" and for distortion in the gap between the top and bottom core shroud in CE units with core barrel shrouds assembled in two vertical sections. The applicant/licensee shall include its proposed acceptance criteria and an explanation of how the proposed acceptance criteria are consistent with the plants' licensing basis and the need to maintain thc functionality of the component being inspected under all licensing basis conditions ofoperation as part of their submittal to apply the approved version ofMRP-227.

Applieants/licensees shall justify the aeceptability of these components [B&W core barrel cylinders (including vertical and circumferential seam welds), B&W former plates, B&W external baffle-to-baffle bolts and their locking devices, B&W core barrel-to-former bolts and their locking devices, and B&W core barrel assembly internal baffle-to-baffle bolts]

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

This item is not applicable to TMI-l; there are no actions for B&W internals identified in this action item, only Westinghouse and CE internals.

Exelon is working with the PWROG to evaluate the acceptability of the subject components for continued service without inspection.

The PWROG is currently concentrating analyses tasks for "Primary" components and is defining future analyses to support continued operation of the subject components. By April 19, 2013 Exelon will provide an update of the PWROG progress including a schedule showing when Exelon will submit for NRC review and approval either I) an analysis of the acceptability of the subject components for continued service without inspection, or 2) a schedule for replacement of the subject components. The submittal schedule is one year prior to entering the Period of Extended Operation and approximately 2Yz years prior to the majority of the currently planned TMI-l MRP-227-A examinations.

Page 0-8

~tJtJOI lUI" D: TMI-1....

to MRP-227 Safety Evaluation items I

Applicant/Licensee Applicant/Licensee Action Item Action Item #

SER Section TMI-l Response 7

4.2.7 8

4.2.8 The of B&W.

and W

reactors are required to

~

develop plant-specifie analyses to be applied for their f~lcilities to demonstrate that B&W IMI tube assembly and CRGT spaeer castings, CE lower support columns, and Westinghouse lower support column bodies 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. These analyses shall also consider the possible loss of fracture toughness in these eomponents due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiring aging management during development ofMRP-227. That is, the requirement would apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-speeific analysis shall be consistent with the plant's licensing basis and the need to maintain the funetionality of the components being evaluated under all licensing basis conditions of operation.

The applicants/lieensees shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227.

Applicants/licensees shall make a submittal for NRC review and approval to credit their implementation of MRP-227, as amended by this SE, as an AMP for the RVI components at their facility. This submittal shall include the information identified in Section 3.5.1, items I and 2, of the SE for MRP-227.

Exelon will develop either a plant-specific

'J or a PWROG /:5CIJCIIC analysis that bounds TMI-l, to evaluate the acceptability of the subject eomponents, and any additional RVI components that may be fabrieated from CASS, martensitic stainless or precipitation hardened stainless steel materials, for continued service.

The analysis will consider the possible loss of fraeture toughness in these components due to thermal embrittlement (TE) and/or IE, as well as limitations on aecessibility for inspection and the resolution/sensitivity of the inspection techniques. By April 19, 2013 Exelon will provide an update including a schedule showing when Exelon will submit for NRC review and approval either 1) analyses of the acceptability of the subject components for the maintenance of functionality during the period of extended operation, or 2) a schedule for replacement of the components. The proposed date of April 19, 20 I3 is one year prior to the Period of Extended Operation and approximately 2 Y2 years prior to the majority of the currently planned TMI-I MRP-227-A examinations.

See subsections 8. I and 8.2 below in this table.

Page 0-9

D.nt:>r,rli" D: TMI-1 R"",nnn"", to MRP-227 Evaluation items

""".-£:'1",<' Rev, 001 Applicant/Licensee Applicant/Licensee Action Item Action Item #

SER Section TMI-l Response 8.1 3.5.1.1 8.2 3.5.1.2 An AMP for the tacility that addresses the 10 program elements as defined in NUREG-180 I, Revision 2, AMP XLMI6A.

Applicants/licensees are to submit an

. r plan which addresses the identified plant-specific action items for staff review and approval consistent with the licensing basis tor the plant. If an applicant/licensee plans to implement an AMP which deviates from the guidance provided in MRP-227, as approved by the NRC, the applicant/licensee shall identify where their program deviates trom the reeommendations of MRP-227, as approved by the NRC, and shall provide a justitication tor any deviation which includes a consideration of how the deviation affeets both "Primary" and "Expansion" inspection category components.

The IO-element program for the AMP is addressed in Section 5.0 of this TMI RV Internals Inspection Plan.

The TMI-l RV Inspection Plan is consistent with MRP-227-A, contains no deviations, and addresses TMI-l specific action items (this table).

Page D-10

ANP-2952. Rev. 001 APPENDIX E: LRA AMR AND INDUSTRY PROGRAM COMPARISON table below shows how the eomponents identified in the results of the TM I-I LRA AMR are dispositioned by the industry program (summarized in MRP-227-A). Speeifically. Appendix E recreates the "Component Type." "Intended Function," "Aging Effect Requiring Management," and "Aging Management Programs" columns from Table I

of the TMI-I LRA. Those eolumns in the LRA that had a "Note I" for the Aging Management Program are to be managed by complianee with the industry RV Internals guidance under development at the time the LRA was submitted, i.e. by MRP-227-A and its supporting documents.

The two additional columns show how the components that previously said "Note I" in the LRA Aging Management Programs column have been dispositioned in MRP-227-A and/or its developmental references. The line items in the LRA that did not say "Note I" in the AMPs column remain managed by the program identified in the LRA. This completes Applicant/Licensee Action Item #2 in the NRC SER for MRP-227, Rev. O.

Table E-I, LRA AMR and Industry Program Comparison Line #

Component Type I.

Control Rod Assembly

,'ontrol rod guide tube 1;~:~l~;~bIY; CRGT pipe and Intended Funetion None Short Lived Stmctural Support to maintain core contiguration and flow distribution Aging Effect Requiring Management None

'hanges in Dimensions/Void Swelling Aging Management Programs l INone.

See Note I Supporting Document and Location N/A MRP-189 Table 4-1, Table 4-2 Disposition in Supporting Documene N/A Category A Iflange 4.

5.

6.

17 8.

'ontrol rod tube pipe and

'ontrol rod guide tube lim;en1bl:y; CRGT pipe and 111ange Control rod guide tube

~ssembly; cI<'GT pipe and flange

'ontrol rod guide tube assembly; CRGT pipe and tlange

ontrol rod guide tube lassembly; CRGT pipe and Iflange Control rod guide tube plSsembly; clWT rod guide isectors Structural Support to maintain core configuration and 110w distribntion IS(~~~~~l~ri~\\i~s~npportto maintain core Ic and 110w distribntion Stmctural Support to maintain core configuration and !low distribution Structural Support to maintain core configuration and flow distribution I

Stmctural Support to maintain core Iconfiguration and tlow distribution

'l'Stmctural Support to maintain core Iconfiguration and tlow distribution

'racking/Stress Corrosion Cracking, ISee Note I Irradiation-Assisted Stress Corrosion

'racking

'racking/Stress Corrosion Cracking, Vater Chemistry Ilrradiation-Assisted Stress Corrosion Cracking

umulative Fatigue Dama A

oss of Fracture Tonghness/Neutron

~ee Note I rradiation Embrittlement, Void welling Loss of Material/Pitting and Crevice Water Chemistry

'orrosion

'hanges in DimensionsiYoid See Note I Swelling MRP-189 Table 4-1.

Table 4-2 NA N/A MRP-189 Table 4-1.

Table 4-2 N/A MRP,231 Table 3-8 Category A N/A N/A Category A N/A Category A Page E-1

Appendix E-AMR and MRP"227"A, Rev. 0 Comparison ANP-2952, Rev. 001 Line #

Component Type Intended Function Aging Effect Requiring Management Aging Management Programs' Supporting Document and Location Disposition in Supportinlf Document 116.

'ontrol rod guide tnbe

'Issembly: CRGT rod guide tubes MRP-23I Category A Table 3,8 N/A N/A N/A NiA MRp*231 Category A Table 3-8 N/A N/A MRp*231 Category A Table 3-8 MRp*231 Category A Table 3-8 N/A N/A N/A N/A MRp*231 Category A Table 3-8 N/A N/A Water Chemistry ISee Note I I

oss of Material/Pitting and Crevice

'orroslon ass of Fracrure Toughness/Neutron rradiation Embrittlement, Void welling Loss of Material/Pitting and Crevice Water Chemistry

'orrosion Fatigue Damage/Fatigue

[rLAA

umulative Fatigne Damage/Fatigue [fLAA

~racking Loss of Fractnre Toughness/Neutron iSee Note I Irradiation Embrittlement. Void

'Swelling

'racking'Stress Corrosion Cracking.

~ec Note I Irradiation~Assisted Stress Corrosion Crackmg

~'racking/StressCorrosion Cracking.

IWater Chemistry Ilrradiation-Assisted Stress Corrosion

'ng

'racking/Stress Corrosion Cracking, See Note I Irradiation-Assisted Stress Corrosion

('racking

(/hanges in DimenSions;Void See Note I

Sweiling ICrackirlg/Stress Corrosion Cracking.

Iwater Chemistry IIrradiatiorh',ssisti:d Stress Corrosion Structural Support to maintain core contignration and flow distribution Structural Support to maintain core contiguration and tlow distribution Structural Support to maintain core contiguration and flow distribution Structural Support to maintain core contignration and tlow distribution Structural Support to maintain core contiguration and flow distribution Structural Support to maintain core contiguration and flow distribution IStructural Support to maintain core contiguration and flow distribution l<:t'",,,1 Support to maintain core lcontiguration and tlow distribution IStnn;tural Support to maintain core Icontiguration and tlow distribution Support to maintain core Icontiguration and flow distribution

!<:lmd,,",,1 Support to maintain core Icontiguration and tlow distribution tube rod guide

'ontrol rod guide tube assembly: CRGT rod guide tubes

'ontrol rod guide tube assembly; CRGT rod guide tubes "ontrol rod guide tube

'lSsembly: CRGT rod guide rs

ontrol rod guide tube assembly; CRGT rod guide sectors trol rod asocmbly:

sectors

'ontro1 rod guide tube

'!Ssembly; CRGT rod guide sectors

.'ontrol rod guide tube

~Is,;enlbl:y; CRGT rod guide Itubes

'ontrol rod kJUide tube

",cMnhhr CRGT rod guide

!sectors I~.

II.

17 19.

18.

13.

14.

10.

.'outrol rod guide tube

!tubes rod guide I-I-5~..--~P;;;;;;I rod gnide tube rr~~::nbIY; CRGT rod l,'11ide Page E-2

Appendix E-AMR and MRP*227-A, Rev 0 Comparison ANP-2952, Rev, 001 Line #

Component Type Intended Function Aging Effect Requiring Management Aging Management Programs l

Supporting Document and Location Disposition in Supporting Document" 26.

28, 29, 30,

'ontrol rod guide tube CRGT spacer castin

ontrol rod guide tube Support to maintain core sembly; CROT spacer casting ~ontlgu!rat!ion and t10w distribution tnbe

-Structural Support to maiutain core spacer casting configuratiou and flow distribution Structural Support to maintain core and now distributiou I,XlUl:tUI at Support to maintain core IconfigUiratlion and now distribution utml rod b'llide tube Structural Support to maintain core sembly; CRGT spacer casting contlguration and t10w distribution

'ontrol rod gnide tube Support to maintain core

'lssembly; CRGT spacer and now distribution

/ontrol rod guide tube IStructural Support to maintain core

'Issembly; CRGT spacer screws 'configuration and now distribution

'ontml rod b'llide tube IStructural Support to maintain core

'\\ssembly; CROT spacer screws contiguration and flow distribution

'ontrol rod guide CRGT Control rod b'llide tube assembly; CRGT spacer screws "hanges in DimensionslVoid See Note I Corrosion Cracking.

'ee Note I Iln'adiatiOlH',ssisted Stress Corrosion vater Chemistry LAA See Note I See Note I Loss of Material/Pitting and Crevice ter Chemistry

'orrosion IChanges in DimensionslVoid See Note I

'Swelling

racking/Stress Corrosion Cracking, See Note I Irradiation*Assisted Stress Corrosion

"'racking

'racking/Stress Corrosion Cracking, Water Chemistry IIrradiation.Assisted Stress Corrosion

[('racking

umulative Fatigue Damage/Fatigue LAA MRp*231 Table 3*8 MRp*231 Table 3*8 N/A N/A MRP*23I Table 3*8 MRp*23I Table 3*8 N/A MRp*189 Table 4*1 MRp*189 Table 4*1 N/A N/A Category A Category A N/A N/A Category A Expansion MRP*227*A changed this to a "Primary" component.

N/A Category A Category A N/A N/A Page E-3

Appendix E-AMR and MRP-227-A, Rev, 0 Comparison ANP-2952, Rev, 001 Line #

Component Type I

Intended Function Aging Effect Requiring Management Aging Management Programs l

Supporting Document and Location J)jsposition in Supportinli Docnment

'Structnral Support to maintaiu core-erscl'c,'vslconfiiglilraliio.n and flow distribution oss of Fracture Toughness/Nentron rradiation Embrittlement, Void

'welling See Note I MRP,189 Table 4-1 Category A 01 rod lide tube bly; CRGT spacer screWs Support to maiutain core and flow distribution oss of Preload/Stress Relaxation See Note I MRP-189 Table 4,1 Category A rol rod guide tube Structural Support to maintain core lbly; Flange,to,upper grid configuration and flow distribution

."crews See Note I MRP,189 Table 4-1 Category /\\

.'ontrol rod tube Structural Support to maintain core

'lssembly~ crll~'If"Bge"to-uIJPe,rgrid configuration and now distribution screws

'racking/Stress Corrosion Cracking, See Note I Irradiation,Assisted Stress Corrosion

racking MRP,189 Table 4-1 Category A 35.

,'ontrol rod guide tube Structural Support to maintain core

'lssembly; Flange,to,upper grid

'onfiguration and flow distribution screws

'racking/Stress Corrosion Cracking, rWater Chemistry Irradiation-Assisted Stress Corrosion

'raeking N/A N/A 36,

!='ontrol rod gnide tube Structnral Support to maintain core

,lS,sernnly; Flange,to,upper grid configuration and flow distribution

'umulative Fatigue Damage/Fatigue frLAA N/A N/A screws Category A Category A No Additional Measures MRP-231 Table 3-8 MRP,189 Table 4,1 MRP-189 Table 4,1 See Note I See Note I See Note I

'hanges in Dimensions/Void Swelling Loss of Fracture ToughnesS/Neutron Irradiation Embrittlement, Void Swelling Loss of Preload/Stress Relaxation Structural Support to maiutain core configuration and flow distribution Stmctural Support to maintain core configuration and flow distribution Structural Support to maintain core configuration and flow distribution 3C),

ore Barrel AssembIv; Baffle/former assembly 137,'ontrol rod guide tube I.

~ ::~:IY' Fl.",o<o.,~ gn,1

138, 01 rod !,'llide tube

/; Flange,to-upper gnd

-rew::;

40,

'ore Barrel Assembly; Battle/former assembly Structural Support to maintain core configuration and now distribution

---racking/Stress Corrosion Cracking, Irradiation,Assisted Stress Corrosion

~racking

~ee Note I MRP-231 Table 3-8 Category A (SCC),

No Additional Measures (lASCC) 41, Core Barrel Assembly; Baffle/fonner assembly Structural Support to maintain core configuration and flow distribution

'racking/Stress Corrosion Cracking, Irradiation-Assisted Stress Corrosion

'racking rWater Chemistry N/A N/A 42, lore BarreI AsscmbIy; Battle/fonner assembly Struetural Support to maiutain core configuration and flow distribution iumulative Fatigue Damage/Fatigue frLAA N/A N/A Page E,4

Appendix E-AMR and MRP-227-A, Rev. aComparison ANP-2952, Rev. 001 I

Line # I Componcnt Type I

Intended Fune!ion Aging Erree! Requiring Management Aging Management Programs!

Supporting Document and Location Disposition in Supporting Documcne 44.

45.

Core Barrel Assembly; IB"Jtle/f'Drrl1er assembly

'ore Barrel Assemblv:

Baffle/timner assembly

ore Barrel Assembly; IB"f11e/l')fme,r bolts and screws iF..dm.

Snpport to rmuntaln core Icontlgu.ration and flow distribution Stmctural Support to maintain core and flow distribution Stnrctural Support to maintain core confIguration and flow distribution Loss of Fracture ToughnessiNeutron Irradiation Embrittlement, Void Swelling Loss of Material/Pitting and Crevice

~~olTosion

~-hanges in Dimensions/Void ISwelling See Note I Vater Chemistry See Note I MRP-231 Table 3-8, and liable 3-9, and Table 3-10 N/A MRP-189 Table 4-1 and MRP-231 Table 3-8 Primary: baffle plates (IE),

and Expansion: former plates (IE)

N/A Category A, except No Additional 1V1easures (core-barrel-to-former bolts, baffle-to-f6rmer bolts, and baf11e-to-baf11e bolts)

'ore Barrel Structural Support to maintain core Baflle/former bolts aud screws configuration and flow distribution

'racking/Stress Corrosion Cracking~

Irradiation-Assisted Stress Corrosion

'racking

~ee Note I MRP-189 Table 4-1 and MRP-231 Table 3-8, and Table 3-9, and Table 3-10 Category A.

except (IASCC) Primary baf11e-to-former bolts, (lASCC)

Expansion core I

barrel-to-tonner bolts and external baffle-to-baffle bolts, and (lASCC)

No Additional ivleasures intenlal baf11e-to-baffle bolts 147

ore Barrel i\\ssernbty; Baf11e/t,)rmer bolts screws Support to maintain core Iconflguration and !low distribution

'racking/Stress Corrosion Cracking, Water Chemistry rradiation-Assisted Stress Corrosion

~racking N/A N/A 4x.

Core Barrel Assembly; Structural Support to maintain core iBaflle/former bolts and screws confIguration and tlow distribution

'umulative Fatigue Damage/Fatigue ITLAA N/A N/A Page E-5

Appendix AMR and MRP*227*A, Rev. 0 Comparison ANP-2952, Rev. 001 Line #

Component Type Intended Function I

I Aging Effect Requiring Management Aging Management Programs' Supporting Document and Location Disposition in Supporting Document2

~:ore Barrel Assembly; Ilk,ttle!fcllIner bolts and

ore Barrel Assembly; Baflk/former bolts and screws

'ore Barrel Assembly; Core barrel eylinder (top and bottom

,nange)

Structural Support to maintain core

'onfiguration and flow distribution Strnctural Support to maintam core Iconfiglll'atJOn and now distnbutiOn Strnctnral Support to maintain core configuration and flow distribution Loss of Fracture Toughness/Neutron Irradiation Embrittlement. Void iSwelling Loss of Preload/Stress Relaxation

~'hanges m Dunensions, Void ISwel" Sce Note I See Note I See Note I MRP-189 Table 4-1 and MRP-231 Table 3-8, and Table 3-9, and Table 3-10 MRP-189 Table 4-1 and MRP-231 Table 3-8. and Table 3-9. and Table 3-10 MRP-189 Table 4-1.

Table 4-2 Category A, except (IE) Primary battle-to-f(Jrmer bolts and (IE)

Expansion core barrel-to-f(lrmer bolts and battle-to-battle bolts Primary battle-to-

!(mner bolts, Expansion core barrel-to-former bolts and bafTk-to-battle bolts, and No Additional Measures thermal shield upper restraint cap screws Category A Core Barrel /\\sse,'nn'ly; Core Strnctural Support to maiutain core barrel cylinder (top bottom

'()J1figuration and flow distribution flange) 52, 53

'ore Barrcl Assembly; Core barrel cylinder (top and bottom flange)

Structural Support to maintain core configuration and flow distribution

~~racking/Stress Corrosion Cracking, See Note I Irradiation-Assisted Stress Corrosion

!,-'racking

racking/Stress Corrosion Cracking, Vater Chemistry Irradiation-Assisted Stress Corrosion

'racking MRP-189 Table 4-1, Table 4-2 N/A Category A NIA 54, 55 iCore Barrel Assembly; Core

~arrel cylinder (top and bottom Iflange) t;~~:}:~~~1 Assembly; Core

~'der(top a;,d bottom Structural Support to maintain core configuration and flow distribution Structural Support to maintain core configuration and flow distribution "umulative Fatigue Damage/Fatigue Loss of Fracture Toughness/Neutron Irradiation Embrittlement, Void

,Swelling

[rLAA See Note I NIA MRP-189 Table 4-1.

Table 4-2 N/A Category A Page E-6

Appendix E~ AMR and MRP-227-A, Rev. 0 Comparison ANP-2952, Rev. 001 Line #

56.

Component Type

'ore Barrel As:sel1nbly; Core barrel cylinder (top bottom flange)

Intended Funetion IStructural Support to mamtarn core

!configUirall,on and thnv distribution Aging Effect Requiring Management Loss of Material/Pitting and Crevice

"'orrosion Aging Management I>rograms' Water Chemistry Supporting Document and Location N/A Disposition in Supportin~

Document*

N/A 57.

ore Barrel Assemhly; Core

!barrel,to tbermal shield bolts

'ore Barrel Assembly; Core barrel,to,thermal shield bolts IStruetural Snpport to maintain core and flow distribution IStructural Support to maintain core contiguration and tlow distribution Changes in DimensionslVoid See Note I Swelling

'racking/Stress Corrosion Cracking, See Note I Irradiation,Assisted Stress Corrosion Cracking MRP-189 Category A Table 4-1 MRP-231 Expansion (SCC)

Table 3-8 and Table 3,10 59.

60.

61.

62.

64, 65.

66.

67.

.'ore Barrel Assembly; Core barrel,to-thermal shield bolts Core Barrel Core barrel,to,thermal shield bolts Core Barrel Assem barrel,to,thermal shi

ore Barrel Core barrel,to,thermal shield bolts

'ore Barrel Assembly; Lower Interual assembly,to-eore barrel bolts

ore Barrel Assembly; Lower Internals assembly-to,core barrel bolts Core Barrel Assembly: Lower Internals assembly-to,core barrel bolts Core Barrel Assembly; Lower Internals assembly,to,core barrel bolts

'ore Barrel Assembly: Lower Internals assemblv-to-core

!barrel bolts Structural Support to maintain core

~onfigu:rat:ionand tlow distribution

!'tnu,o,n,1 Support to maintain core figuration and flow distribution uctural Support to maintain core lfiguration and tlow distribution Structural Support to maintain core configuration and flow distribution Stmctural Support to maintain core

!eontiguratiou and now distribution Structural Support to maintain core configuration and now distribution Stmctural Support to maintain core configuration and tlow distribution

'Structural Support to maintain core configuration and flow distribution Structural Snpport to maintain core contiguration and flow distribution

'racking/Stress Corrosion Cracking,

~Vater Chemistry Irradiation,Assisted Stress Corrosion

'racking

'umulative Fatigue Damage/Fatigue rrLAA Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement, Void Swelling Loss of Preload/Stress Relaxation See Note I

'hanges in DimensionslVoid See Note I Swelling

~racking/StressCorrosion Cracking, See Note 1 Irradiation,Assisted Stress Corrosion

~racking

'racking/Stress Corrosion Cracking, Vater Chemistry Irradiation,Assisted Stress Corrosion

'racking

'umulative Fatigue Damage/Fatigue TLAA Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement, Void Swelling N/A N/A MRP,189 Table 4,1 MRP,189 Table 4~1 MRP-189 Table 4,1 MRP,231 Table 3-8 and Table 3,9 N/A N/A MRP-189 Table 4,1 N/A N/A Category A Category A Category A Primary (SCC)

N/A N/A Category A Page E-7

Appendix E-AMR and MRP-227-A, Rev, 0 Comparison ANP-2952. Rev. 001 Line Component Type Intended Function Aging I£l'f'ect Requiring Management Aging Management I>rograms l Supporting I Disposition in Document andl Supporting Location I

Documene Corrosion Cracking, See Note I IlolldiatiOll-l'lsslSt*ed Stress Corrosion k'racking 6~.

69.

70.

If',

Barrel Assembly; Lower Ilnternals assembly-to-core

~arrel bolts

~'ore support slucld assembiy; ICore support shIeld cylinder and bottom flange) pport shield assembly; shield cylinder Ilange)

Struetural Support to maintain core configuration and flow distribution Support to maintain core Ico.nlignration and flow distribution Support to maintain core Iccmfigllration and Ilow distribution Loss of Preload/Stress Relaxation

~'hanges in Dimensions}Void ISwel ISee Note I See Note I MRP-l~9 Table 4-1 MRP-I~9 Tahle4-I, Table 4-2 MRP-l~9 Table 4-1, Table 4-2 Category A Category A Category A 71 73.

ield cylinder flange) ore support shield assemhly;

_'ore shield cvlinder ktop and hottom flange)

'ore support shield assembly;

'ore shield evlinder I(top and bottom flange)

Support to maintain core Iccmfigllration and flow distribution

!Structural Support to maintain core

'onliguration and flow distribution Structural Support to maintain core configuration and flow distribution

'e' Corrosion Cracking, IWater Chemistry Ilo'adiatiorl-Plssist.ed Stress Corrosion k'racking

\\lmulative Fatigue Damage/Fatigue rrLAA Loss of Fractnre ToughnesslNeutron

[See Note I Ilrradiation Embrittlement, Voit!

ISwelling N/A N/A MRP-l~9 Table 4-1, Table 4-2 N/A N/A Category A 74.

75.

76.

7~

,I, 7~,

'ore support shield assembly; Core shield cylinder (top and bottom flange)

Core support shield assembly; "ore support shield cylinder (top flange)

'ore support shield assembly;

'ore support shield cylinder KIop flange)

~'ore support shield assembly;

~'ore support shield cylinder i(top Ilange)

'ore support shield assembly;

ore support shield cylinder (top flange) to maintain core

'onflguration and now distribution Structnral Support to maintain core n

'Stmetnral Support to maintain core configuration and flow distribution Stmctnral Snpport to maintain core configuration and flow distribution Structural Snpport to maintain core configuration and flow distribution jEoss of Material* Pitting and Cre"lce

!corrOSIon jLhanges in DimenSions. Void

[Swelling

'racking/Stress Corrosion Cracking, Irradiation-Assisted Stress Corrosion

'racking

raeking/Stress Corrosion Cracking, Irradiation-Assisted Stress Corrosion

'racking

'umulative Fatigue Damage/Fatigue Vater Chemistry See Note I See Note I Vater Chemistry LAA N/A MRP-189 Table 4-1, Table 4-2 MRP-l~9 Table 4-1, Table 4-2 N/A N/A N/A Category A Category A N/A N/A Page E-8

Appendix E~ AMR and MRP-227-A, Rev, 0 Comparison ANP-2952, Rev, 001 Line #

Component Type Intended Function Aging Effect Reqniring Management Aging Management I>rograms l SUPI)Orting Document and Location Disposition in Supporting Documene 79, support shield assembly;

{',

support shield cylinder (top flange)

StOictural Support to maintain core Iconfignration and flow distributiou Loss of Fracture Toughness/Neutron Irradiation Fmbrittlement. Void Swelling See Note I MRP-189 Table 4-1, fable 4-2 Category A 80,

{',

support shield assembly; Core support shield cylinder (lOP flange)

Structural Support to maintain core

~onfigUlcation and flow distribution Loss of Material/Pitting and Crevice

~orrosion Water Chemistry N/A 81 "ore support shield assembly; Core support shield-to-core barrel holts StOictural Support to maintain core Iconfiguration and flow distribution

'hanges in Dimensions! Void Swelling See Note I MRP-189 Table 4-1 Category A 82,

("ore snpport shield assembly;

'ore shield,to'eore barrel b~its StOictural Support to maintain core leoufiguration and flow distribution

'racking/Stress Corrosion Cracking, Irradiation*Assisted Stress Corrosion

~nlcking

'See Note I MRP*231 Table 3*8 aud Table 3*9 Primary (SCC) 83,

'ore support shield assembly;

'ore support silleld,to-core Jarrel bolts Structural Support to maintain core IconfigUl'ation and flow distribution Corrosiou Cracking, IIff'adiatiorr'PlSsist,ed Stress Corrosion

~'racking Water Chemistry N/A N/A, 84,

'ore support shield assembly;

'ore support shield-to*core barrel bolts Stfflctural Support 10 maintain core configuration and t10w distribution

'umulative Fatigue Damage/Fatigue fLAA I

Nil',

Category A Category A MRP-189 Table 4*1 MRp*189 Table 4*1 Loss of Preload/Stress Relaxation See Note I Loss of Fracture ToughnessiNeutron

[See Note I Irradiation Embrittlement, Void Swelling Stfflctural Support to maintain core configuration and tlow distribution

'ore support shield assembly;

'ore support sliield-to-core barrel bolts

~porttomaintaincore

~;ore support shield*to-core '

Iconfiguratim{ and flow distribution Jarrel bolts 85, 87.

Core support shield assembly;

)utlet and vent valve nozzles StOictural Support to maintain core configuration and flow distribution

~;hanges in DimensionsiVoid See Note I ISwe1l1 MRP*189 Table 4*1, Table 4*2 Category A 88

'ore support shield assembly; Outlet and vent valve nozzles IStructural Support to maintain core conflguration and flow distribution

'racking/Stress Corrosion Cracking, See Note I Irradiation,Assisted Stress Corrosion

'racking MRp*189 Table 4*1, Table 4,2 Category A 89.

ICore support shield assembly; i)utlet and vent valve nozzles Structural Support to maintain core IconflgUl'ation and flow distribution

'racking/Stress Corrosion Cracking, Water Chemistry Irradiation*Assisted Stress Corrosion

'racking N/A Nil',

Page E-9

Appendix E-AMR and MRP*227*A, Rev, 0 Comparison ANP-2952, Rev, 001 Disposition in Supportin~

Document' Supporting Document and Location Aging Management Programs' Aging Effect Requiring Management I,

Intended Function Line #

Component Type 9~-y-;~k:~,-,tn"-,t,,-r,,,-'s-u-p-p-o-rt-t-o-m-a-ir-lt-a-ir-l-c-o-re-'

-k;~~:~~;;;;'-F-,,-,t-ig-'l-,e->-D-,,-n-la-L-,e-./-F-a-ti-g-U-e-+r-L-,-,\\-A------+---N-/-A---+---N-,-;;-\\---1 k)~;lc;~I'r;d-~~~;;';IV~;I~)~~I:,'

and flow distribution 91.

~'ore support sbield assembly;

)utlet and vent valve nozzles Stnrctural Support to maintaiu core

'onfiguration and t10w distribution ILoss of Fracture Toughness/Neutron See Note I Ilrrad'iation Embrittlement, Void ISweiling MRp*189 Table 4*1.

Table 4*2 Category A 92,

'ore support shield assembly;

)utlet and veut valve nozzles Stnrctural Support to maintain core and t10w distribmion Loss of Material/Pitting and Crevice Water Chemistry N/A N/A 93,

'ore support shield assembly; Vent valve assemblv lockiug

~evice Stmctural Support to maintain core configuration and flow distribution

~'hanges in DimensionslVoid See Note I

~well' N/A Vent Valve locking devices arc being considered for the TMI-I RV Inspection plan, See Appendix D, response to LIAAI 112 94,

~ore Vent

~evice shield assembly; assembly lockiug Stmctural Support to maiutain core configuration and flow distribution

'racking/Stress Corrosion Cracking, See Note I Irradiation,Assisted Stress Corrosion

'racking N/A Vent Valve locking devtccs are being considered f()[ the TMI-I RV Inspection plan, See Appendix D, response to L1AAI

2, 95,

'ore support shield assembly; IVent valve assembly locking

!dcvice Structural Support to maintain core configuration and t10w distribution

<lb' Corrosion Cracking, Water Chemistry Ilrr'adiatiorl-,AlSsist,ed Stress Corrosion

'racking N/A N/A 96.

Core support shield assembly; Vent valve assembly locking device Structural Support to maintain core configuration and flow distribution

umulative Fatigue Damage/Fatigue rLAA I

N/A N/A Page E-10

Appendix AMR and MRP~227~A, Rev. 0 Comparison ANP-2952, Rev. 001 Component Type I,

Intended Function I

Aging Effect Rcquiring Managcment Aging Management Programs' Supporting Document and Location Disposition in Supporting Documcne 97.

shield assembly locking

""r'" Support to maintain core

!configuration and flow distributiou Loss of Fracture Toughuess. Neutron See Note I Irradiation EmbrittlemenL Void Swelling N/A Vent Valve locking devices are being considered fClr the TMI~ I RV Inspeetion plan.

See Appendix D, response to LlAAI

2

'ore support shield assembly; Vent valve assembly locking ievice Structural Support to maintain core

[configuration and t10w distribution Loss of Material/Pitting and Crevice

~orrOSlOn Water Chemistry N/A N/A N/A Category A Category A MRP-189 Table 4-1 MRP-189 Table 4-1

,Vater Chemistry See Note I

'See Note I Support to maintain core

'hanges in Dimensions/Void and flow distribution

~welling lore support shield assembly; Vent valve body Core support shield assembl:

Vent valve body 101.

99.

I<:,n"'j.,r'" S.upport to mai.ntai.~n core craCking/.Stress c..orroSi.on (.,raCking Icoufignration and t10w distribution Irradiation-Assisted Stress Corrosion

~---+-------------1~-----------

_n_lc_,k_i_n::-g

+

+-

--i Core support shield assembly; Structural Support to mamtam racking/Stress CorrosIOn Cracking, contiguration and flow distrib adiation-Assisted Stress Corrosion 102.

'ore support shield assembly; IVent valve body Stl1Jctnral Snpport to maintain core

'umnlative Fatigue Damage/Fatigue and flow distribution IrLAA N/A N/A j03.

(lore support shield assembly; Vent valve body

[Stl1Jt;tural Support to maintain core contignration and now distribution Loss of Fracture Toughness/Neutron Irradiation Embrittlemenl. Void Swelling ISee Note I MRP-189 Table 4-1 Category A 104 ore shield assemblY; Vent valve body Structural Support to maintain core and now distribution Loss of Fracture Tonghness/Thermal See Note I Aging Embrittlement MRP-189 Table 4-1 Category ;\\

107, lOS,

106,

'ore support shield assemblY' Structural Support to maintain core f

~alve b;dy configuration and tlow distribution support shidd assembly; IStructural Support to maintain core

~___

valve retaining ring configuration and t10w distribution

'ore support shield assembly; Structural Support to maintain core Vent valve retaining ring Icu>nllfiguration and flow distribution Loss of Material/Pitting and Crevice IWater Chemistry

'orrosion

'hanges in DimensionslVoid See Note I Swelling

rackiug/Stress Corrosion Craekiug, iSee Note I

'ation-Assisted Stress Corrosion N/A MRP-189 Table 4-1 MRP-189 Table 4-1 N!A Category A Category A Page E-11

Appendix E-AMR and MRP*227-A, Rev, 0 Comparison ANP-2952, Rev. 001 I

Line # I Component Type Intended Function Aging Effect Requiring Management Aging Management Programs l Supporting Document and Location Disposition in Supportin~

Doeument" lOX.

'ore Vent shield assemhly; retaining nng Stmetural Support to maintain core configuration and flow distribntion

'racking/Stress Corrosion Cracking, Vater Chemistry Irradiation-Assisted Stress Corrosion Cracking N/A II L N/A N/A Category A N/A N/A rvtRp*189 Table 4-1 oss of Material/Pitting and Crevice

,water Chemistry orrosion Fatigue Damage/Fatigne

[LAA Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement. Void Swelling tructural Snpport to mamtain core Jnfiguration and tlow distributIOn ructural Support to maintain core rrfiguration and flow distribntion

~ctural Support to malntam core Icou and tlow distribntion 109.

Core support shield assembly; Vent valve retaining ring

~e shield assembly; IV~nt retammg nng

112, Flow distributor assembly; Clmnninv ring Stmctural Support to maintain eore configuration and tlow distribution

'banges in Dimensions/Void See Note I welling rvtRP-189 Table 4-l.

Table 4-2 Category A

] 13, Flow distributor assembly;

'Iamping ring Structural Support to maintain core contiguration and tlow distribution

'racking/Stress Corrosion Crackiug, See Note I Irradiation-Assisted Stress Corrosiou iracking rvtRP-189 Tahle 4-1.

Tahle 4-2 Category A 114.

istributor assembly; ing ring

Stmctural Support to maintain core Icontiguratton and tlow distnbutron

'racking/Stress Corrosion Cracking, Water Chemistry Irradiation-/\\ssisted Stress Corrosion

'racking N/A N/A II Flow dist.ributor assembly; Clamping ring IStmctural Support to maintain core Icontiguration and tlow distnbution

\\ullulative Fatigue Damage/Fatigue trLAA N/A N/A 116<

Flow distributor assembly;

Iamping ring Stmetural Support to maintain core configuration and flow distribntion

< of Fracture Toughness/Neutron See Note 1 iation Embrittlement Void Uing MRP-IX9 Table4-1, Table 4-2 Category A

117, IIX<

Flow distribntor assembly; iStmetural Support to maintain core

.'Iamping ring onfiguration and tlow distribution Flow distributor assembly; FlowlStruetnral Support to maintain core distributor head and tlange configuration and tlow distribution ss of Material/Pitting and Crevice Vater Chemistry

'oITosion

<hanges in DimensionslVoid See Note I Swelling N/A MRP-189 Table 4-1, Table 4-2 N/A Category A

119, Flow distributor assembly; Flow Stmctural Support to maintain core flistributor head and flange configuration and tlow distribution

racking/Stress Corrosion Cracking, See Note I Irradiation-Assisted Stress Corrosion

'racking MRP-189 Table 4-1.

Table 4-2 Category A Page E-12

Appendix E-AMR and MRp*227*A, Rev. 0 Comparison ANP*2952, Rev. 001 N/A Disposition in Supporting Document" N/A Supporting Document and Location Aging Management I'rograms' Water Chemistry I

~I Aging Effect Requiring Line # I Component Type Intended Function Management

l------l------I------j 120.

IFiow distributor assembly; Flow Struc.tu.ral Support to maintain core g/Stress Corrosion Cracking, listributor head and confignration and flow distributIOn ion*AsSisted Stress Corrosion 121.

Flow distnbutor assembly; Support to maintain core

'umulative Fatigue Damage/Fatigue

~istributor head and flange and !low distribution

[rL\\A N/A 122.

IFlow distnbutor assembly; Flow'Structural Support to maintain core

~istributor head and flange contiguratlOn and !low distnbutlon ass of Fracture Toughness/Neutron rradiation Embrittlement, Void

,welling See Note I MRp*189 Table 4*1, Table 4*2 Category A 123.

IFlow distributor aSSemblY.'; Flow Struetural Support to maintain core

~istnbutor head and tlange and !low d'stribution ass of N1ateriallPitting and Crevice orrosion Water Chemistry N/A N/A 124.

Flow distributor assembly; Structural Support to maintain core lueore guide support plate configuration and flow distribution See Note I MRp*189 Table 4*1 Category A

125, Flow distributor assembly; Structural Support to maintain core Incore guide support plate confignration and flow distribution Corrosion Cracking, See Note I Iln*adiatiOll*P,ssist,ed Stress Corrosion

.'racking MRp*189 Table 4*1 Category A 126.

IFlow distributor assembly;

!Iucore guide support plate Structural Support to maintain core contiguration and flow distribotion "c'

Corrosion Cracking, Water Chemistry IIn*adiatior,*,;,ssisved Stress Corrosion

~'racking N/A N/A 127.

IFI distributor assembly;

[ncore guide support plat~

Structural Support to maintain core and flow distribution

~nmulative Fatigue Damage/Fatigue LAA N/A N/A 128.

Flow distributor assembly; Incore guide support plate Structural Support to maintain core configuration and flow distribution Loss of Fracture Toughness/Neutron See Note I Irradiation Fmbrittlement, Void Swelling MRP-189 Table 4*1 Category A 129.

130.

Flow distribntor assembly; Structural Support to maintain core Incore guide support plate and !low distribution IFI distributor assembly; Shell Structural Support to maintain core n

distributor bolts configuration and flow distribution Loss of Material/Pittiug and Crevice Vater Chemistry

'orrosion

~:hanges in DimensionslVoid See Note I

~welling NA MRp*189 Table 4*1 N/A Category A 131.

Flow distributor assembly; Shell Structural Support to maintain core fo,,' Of w~flow"",ib,"" bo'" 1'00"f~'""00 "'d now","ib0" 00

'racking/Stress Corrosion Cracking, See Note I Irradiation-Assisted Stress Corrosion

raeking tvIRp*23I Table 3*9 and Table 3*10 MRP-227*/\\

MRp*227*A changed these trom "Expansion" to "Primary" components.

Page E-13

Appendix E~ AMR and MRP-227-A, Rev. a Comparison ANP-2952, Rev. 001 Line #

Component Type Intended Function Aging Effect Requiring Management Aging Management Programs' Supporting Document and Location Disposition in Supporting Document!

132.

distributor assembly; ShellStructural Support to maintain core

, 0'0 distributor bolts configuration and flow distribution "0

Corrosion Cracking, Vater Chemistry

!lrr'adiatiorH'"siM,~dStress Corrosion

~'racking N/A N/A 133.

distributor assembly; Shell Structural Support to maintain core distributor bolts configuration and flow distribution

'umulative Fatigue Damage/Fatigue LAA N/A 134.

Flow distributor assembly; Shell Structural Support to maintaiu core forging-to,flow distributor bolts'onfiguration aud flow distribution loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement, Void Swelling MRP,189 Table 4-1 Category A N/A Category A N/A MRP,189 Table 4-1 Loss of Preload/Stress Relaxation See Note I None Short lived

'b Flow distributor assembly; Shell Structnral Support to maintain core distributor bolts confignration and flow distribution Fuel Assembly 136.

135.

137.

,,}"e>" grid assembly; Fuel

~lSsembly support pads Structural Support to maintain core and flow distribution

~'hanges in DimcnsionsiVoid See Note I ISwelling MRP-189 Table 4-1 Category A lower llrid Fuel lassemblv support Structural Support to maintain core configuration and flow distribution iracking/Strcss Corrosion Cracking, See Note I Irradiation-Assisted Stress Corrosion

'racking MRP,231 Table 3,8 and Table 3,10 Category A except Expansion Alloy X,750 dowel locking weld 139.

lower Fuel

[assembly support Structural Support to maintain core configuration and flow distribution

'racking/Stress Corrosion Cracking, Water Chemistry Irradiation-Assisted Stress Corrosion

'racking N/A N/A 140.

Lower assemblv; Fuel assembly support p,;ds Structural Support to maintain core and flow distribution

'umulative Fatigue Damage/Fatigue [rlAA N/A N/A 141.

Lower assemblv; Fuel assembly support p,;ds Structural Support to maintain core configuration and now distribution Loss of Fracture Toughness/Neutron See Note I Irradiation Ernbrittlement. Void Swelling MRP,231 Table 3-8 and Table 3,10 Expansion (IE) 142.

Lower grid assembly; Fuel Structural Support to maintain core assembly support pads configuration and flow distribution Loss of MaterialiPitting and Crevice Water Chemistry

~orrosion NlA N/A lower grid assembly; Guide Stmctural Support to maintain core

'hanges in Dimensions/Void See Note I

~.

RP,189

_-tb__o_c_k_s rc~o_n~fi,,;g~u~r~at~i(~m_a_n~d~f~lo~'~v~(~li_s_tr_ib~u~t_i(_ll1

_+S-w-c->I-II-*n,,;g~.

+-

I--_~..able 4-1

.'~,'~I:~ grid assembly; Guide Structural Support to maintain core r'racking/Stress Corrosion Cracking, See Note I UIULM configuration and flow distribution Ilrradiation,Assisted Stress Corrosion k~racking Category A Category A Page E-14

Appendix 1'- AMR and MRp*227*A, Rev. 0 Comparison ANP*2952, Rev. 001 Line #

Component Type Intended Funetion Aging Effect Requiring Management Aging Mauagement I>rograms l I Supporting

,Document and I

Location Disposition in Supportin~

Document*

145.

Lower grid assembly; Guide blocks IStru(;tural Support to maintain core Iconfiguration and flow distribution

~racking/Slress Corrosion Cracking, IrradiatiorH\\ssisted Stress Corrosion

'racking Water ChemIstry N/A N/A 146.

Lower grid assembly; Guide

[blocks 1147.

Lower grid assembly; Guide 1

Iblocks CorrosIon Cracking, Water Chemistry IIrradiatiorH',ssist,.:d Stress Corrosion N/A N/A MRp*189 Category A Table 4-1 N!A N!A MRP-189 Category A Table 4-1

VlRP*189 Category /\\

Table 4-1 N/,

N!A N/A N/A MRP-IH9 Category A Table4-1 MRP-189 Category A Table 4-1 MRP-IH9 Category A Table 4-1 MRP-IH9 Category A Table 4-1 N/A N/A LAA See Note I See Note I Iwater Chemistry

'Water ChemIstry iSee Note I See Note I See Note I See Note I

[rLAA

'umulative Fatigue Damage/Fatigue Loss of Material/Pitting and Crevice

'nIcking

~racking/Stress Corrosion Cracking; Irradiation-Assisted Stress Corrosion

'racking Loss of Preload/Stress Relaxation Loss of Fracture Toughness/Neutron Irradiation Emhrittlement, Void Swelling

~:hanges in Dimensions/VoId iSwelling

~

'raCkingIStress Corrosion Cracking, Irradiation-Assisted Stress Corrosion "racking ICraekirlg, Stress Corrosion Cracking, See Note I Ilrr'adiatiorh',ssist,.:d Stress Corrosion

~,'raeking LOSS of Fracture Toughuess/Neutron Irradiation Embrittlemcnt. Void Swelling f'umulative Fatigue Damage/Fatigue Structural Support to maintain core and flow distribution Snpport to maintain core and flow distribution Structural Support to maintain core and flow distribution Structural Support to maintain core and flow distribution Structural Support to maintain core conflguration and flow distribution Structural Support to maintain core configuration and flow distribution Structural Support to maintain core configuration and How distribution Structural Snpport to maintain core conflguration and flow distribution Structural Support to maintain core conflguration and flow distribution Structural Support to maintain core confignration and flow distribntion Structural Support to maintaiu core Iconllguration and flow distribution IStructnral Support to maintain core

!configuration and flow distribution ower grid assembly; Guidc locks bolts ower grid assembly; Guide locks bolts ower grid assembly; Guide ocks bolts ower grid assembly; lncore ide tube spider castings

.ower grid assembly; Guide locks Lower grid assembly; Incore guide tube spider castings Lower grid assembly; {ncore guide tube spider castings ILower grid assembly; Guide

[blocks Lower grid assembly; Guide

[blocks bolts ower grid assembly; Guide locks bolts

156, 152.

149.

150.

155.

154.

14H, 151 153.

Page 1'-15

Appendix E-AMR and MRP*227*A, Rev. 0 Comparison ANP*2952, Rev. 001 Line #

Component Type Lower grid assembly; lneon:

guide tube spider castings ILower grid Incore Iguidc tube spider 160.

Lower grid assembly; lucore guide tube spider castings Intended Function Structural Support to maintain core and flow distribution Structural Support malutam core configuration and flow distribution Structural Support to maintain core configuration and flow distribution Aging Effect Requiring Management

\\mmlalive Fatigue Damage/Fatigue Loss of Fracture Toughness/Neutron Ilrradiation Fmbrittlement, Void Swelling Loss of Fracture Toughness/Thermal Agi Embrittlement Aging Management Programs' rLAA See Note I See Note I I[ Supporting Document and Location N/A MRP*231 Table 3*8 and Table 3*9 MRp*231 Table 3-8 and Table 3*9 Disposition in Supporting Document1 N/A Primary (IE)

Primary 161 Lower grid assembly; lncore Imide tube spider castings Structural Snpport to maintain core aud flow distribntion Loss of Material/Pitting and Crevice W:

'orrosion Chemistry N/A N/A

'racking

'0 Corrosion Cracking, See Note I Iln'adiatiOll*/1lssisved Stress Corrosion 162.

163.

Lower grid assemblv; Lower Igrid and sbell forgit;gs Lower grid assembly; Lower Qrid and shell f'lrgings Stmctural Support to maintain core configuration and flow distribution Structural Support to maintain core configuration and flow distribution

'hanges in Dimensions/Void Swelling See Note I MRP-189 Table 4-1, Table 4-2 I

,mn 189 Table 4-1, Table 4-2 Category A Category A

164, 165.

166.

Lower grid assembly; Lower grid and shell forgings Lower grid assembly: Lower grid and shell f()rgings Lower grid assembly; Lower grid and shell forgings Structural Support to maintain core contignration and flow distrihntion Stmctural Support to maintain core and flow distribution jStructural Support to maintain core Icontign,rat'lon and flow distribution

'0 Corrosion Cracking.

Water Chemistry

!In'adiatiorH'lssist<ed Stress Corrosion

'racking

'mnulative Fatigue Damage/Fatigue TLAA ILoss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement. Void Swelling N/A N/A MRP-189 Table4-L Table 4-2 N/A N/A Category A 167 Lower grid assembly; Lower

!Structural Snpport to maintain core Loss of Material/Pitting and Crevice

\\Vater Chemistry grid and shell forgings and flow distribution "orrosion N/A N/A Category A 16~.

1()9.

Lower grid assembHY; Lower ructural Support to maintain core Changes in Dimensions/Void See Note 1 MRp*I~9 grid flO".' distributor plate mfig_ur_a_t_i(_ll__'U_ld_fl_O_W__d_is_,t_ri._b_u_ti_o_n_+s_we_IIlI_ing:::-

+-

-+__T_a_b_l_e_4_-_I_-+

~

Lower grid assembl ral Support to maintain core

'racking/Stress Corrosion Cracking, See Note 1 MRp*189 Category A 19rid flow distributor ration and flow distnbution Ilrr.adiation-Assisted Stress Corrosion Table 4-1

~'racklllg Lower grid assembly; Lower grid flow distributor plate IStructural Snpport to maintain core configuration and flow distribution Corrosion Cracking, Water Cbemistry IIn'adiatiOlH',ssist,ed Stress Corrosion Cracking N/A N/A Page E-16

Appendix AMR and MRP-227-A, Rev 0 Comparison ANP-2952, Rev, 001 Un<' I Component Type Intended Function Aging Effect Reqniring Management Aging Management Programs' I Snpporting IDocument and I

Location Disposition in Supportinlf Docnment 171 173.

174.

Lower grid assembly; Lower grid flow distributor plate Lower grid assembly:

grid 110w distnbutor plate Lower grid assemblv; Lower igrid 110;" distributo; plate

,Lower grid assembly; Lower rib section Structural Support to maintain core and flow distribntion Structural Support to mamtain core configuration and tlow distributiou Structural Support to maintain core

[Configuration and flow distribution IStructural Support to maintain core and flow distribution "umulative Fatigue Damage/Fatigue rLAA L,,,s of Fracture Toughness:Neutron See Note I Irradiation Embrittkment, Void Swelling Loss of Material/Pitting and Crevice Water Chemistry

'orrosion

'hauges iu DimensionsiVoid See Note I Swelling N/A MRp*189 Table 4*1 N/A MRP*189 Table 4*1 N/A Category A N:A Category A Lower grid assembly; Lower section

!Structural Support to maintain core Iconfign'ratiou and flow distribution

'rackiug/Stress Corrosion Cracking, See Note I Irradiation*Assisted Stress Corrosion

'racking MRP*189 Table 4*1 Category A 178.

179.

Lower assembly; Lower grid rib section Lower assembly; Lo\\ver grid rib~section

~

Lower grid assembly; Lower grid rib section ILower grid assembly; Lower jgrid rib section

,Structural Support to maintain core onfiguration and now distribution iStructural Support to maintain core and flow distnbutlon Structural Support to maintaiu core contignration and flow distribution I:>tru('~hmra, Support to ma111tain core aud l10w distribution

'racking/Stress Corrosion Crackiug, Water Chemistry Irradiation*Assisted Stress Corrosion

racking

'umulative Fatigue Damage/Fatigue TLAA oss of Fracture Tonghness/Nentron See Note I rradiation Embrittlement. Void Swelling Loss of Material/Pitting and Crevice Vater Chemistry

'orrosion N/A N/A MRP*189 Table 4*1 N/A N/A N/A Category A N/A I~O Lower grid assembly; Lower grid rib* to-shell fi:Jrging screws I:>truc,mra, Support to maintain core and 110w distribution Changes in Dimensions, Void Swelling See Note I MRP*189 Table 4*1 Category A 181.

Lower grid assembly~ Lower grid rib*to*shell forging screws

'Structural Support to maintain core confignration and flow distribution

~'racking:StressCortoston Crackmg, See Nelte I IIrradiation*Assisted Stress Corrosion k:racking MRp*I~9 Table 4*1 Category A IX2.

183.

ILower grid assembly; Lower Structural Snpport to maintain core Igrid rib~to*shell forging screws configuration and flow distribution ILower grid assembly; Lower Stmctnral Snpport to maintain core Wid rib*to*shell fc)rging screws configuration and !low distribution

'racking/Stress Corrosion Cracking, I'water Chemistry Irradiation~AssistedStress Corrosion I

'racking

'umnlative Fatigue Damage/Fatigue rLAA N/A N/A N/A N/A Page E-17

Appendix E~ AMR and MRP-227-A, Rev, 0 Comparison ANP-2952, Rev, 001 Line #

Component Type Intended Fnnction Aging Effect Rcquiring Management Aging Management Programs' I Supporting IDocument and I

Location Disposition in Supporting Document' 1M, Lower grid assembly; Lower grid rib,to'shell forging serews Stnlctura( Support to maintain core

'onfiguration and flow distribution Loss of Fracture Toughness! Neutron

'See Note I ilrrad,atlOn Embnttlement, Void Swelling MRP,189 Table 4,1 Category A

IH5, 186,

187, 188,

189, Lower grid assembly~ Lower grid rib,to'shell f'xging screws Lower grid assembly; Lower Intemals assembly-to~therrnal shield bolts Lower grid assembly; Lower

'intemal assembly,to,thermal shield bolts Lower grid assembly; Lower internals assembly-to-thermal shield bolts Lower grid assembly; Lower lintemal~ assembly-to-thermal shield bolts Struetural Support to maintain core and flow distribntion Structural Support to maiutain core configuration and flow distribution Structural Support to maintain core configuration and Ilow distribution Structural Support to maintain core configuration and flow distribntion Structnral Snpport to maintain core contiguration and flow distribution Loss of Preload/Stress Relaxation Changes in Dimensions/Void

,Swelling

'racking/Stress Corrosion Cracking.

Irradiation,Assisted Stress Corrosion

'racking

'racking/Stress Corrosion Cracking, Irradiation,Assisted Stress Corrosion

'racking

-'umulative Fatigue Damage/Fatigue See Note I See Note I See Note I Water Chemistry fLAA MRP,231 Table 3,8 MRP,189 Table 4,1 MRP,231 Table 3,8 and Table 3,10 N/A N/A No Additional Measures (ISR)

Category A Expansion (SCC)

N/A N/A 190.

191.

Lower assembly; Lower

'intemal assembly,t~,thermal shield bolts Lower grid assembly: Lower internals assembly-to-thermal shield bolts Structural Support to maintain core Loss of Fracture Toughness/Neutron See Note I

'onfiguration and flow distribution Irradiation Embrittlement. Void Swelling Structural Support to maintain core Loss of Preload/Stress Relaxation

~ee Note I confignration and flow distribution MRP,189 Table 4,1 lv1RP, I89 fable 4,1 Category A Category A

192, Lower grid assemblv: Orifice Iplugs

~

Structural Support to maintain core conflguratiou and flow distribution

'hanges in Dimensions/Void Swelling ISee Note I MRP,189 Table 4,1 Category A

193, 194, Lower grid assembly; Orifice iplugs ower grid assembly; Orifice plugs Stmctural Support to maintain core configuration and flow distribution I

iStructural Support to maintain core Iconfiguration and flow distribution Corrosion Cracking, See Note I Ilrradiat,orl-J'.ssist<~d Stress Corrosion

'racking/Stress Corrosion Cracking, Water Chemistry Irradiation-Assisted Stress Corrosion

'racking

!'v1RP-189 Table 4-1 N/A Category A N/A 195.

ILower grid assembly; Oritice IStructural Support to maintain core plugs

!configuration and flow distribution

~umulative Fatigue Damage/Fatigue

!TLAA N/A N/A Page E-18

Appendix AMR and MRP-227-A, Rev. 0 Comparison ANP-2952, Rev. 001 Line #

Component Type Intended Function Aging Effect Requiring Management Aging Management Prugrams l Supporting I Document and Location Disposition in Supporting Document2 ILowcr grid assembly; Onfice

~Iugs Structural Support to maintain core lconfigwtation and flow distribution iLoss of Fracture Tonghness/Neutron See Note I Ilrrad:iation Embrittlement, Void ISwelling MRp*189 Table 4-1 Category A 197.

Lower grid assernbly; Orifice

~)Iugs Structural Support to maintain core and flow distribution Loss of Material/Pitting aud Crevice Water Chemistry

~OIT(}SlOn N/A N/A 19X.

199.

Lower grid assemhly; Shock

!pads Lower grid assembly; Shock

!pads Structural Support to maintain core and flow distribution Structnral Support to maintain core Iconfigwtati.on and flow distribution

~'hangcs in Dimcnsions/Void See Note I

~welling

'racking/Stress Corrosion Cracking, See Notc I Irradiation*Assisted Stress Corrosion

'racking MRp*189 Table 4-1 MRP,189 Table 4-1 Category A Category A 20ll.

ILowcr grid assembly; Shock

~ads Structural Support to maintain core Iconfigul'ation and flow distribntion Corrosion Cracking, Water Chemistry IInmdliatioll-!\\.ss:isted Stress Corrosion N/A NA 20t.

Lower grid assembly; Shock

)ads to mamtam core configuration and flow distribution racking umulative Fatigue Damage/Fatigue LAA N/A N/A 202.

ILowcr grid assembly; Shock

[pads Structural Support to maintain core confIguration and flow distribution oss of Fracture Tonghness/Neutron radiation Embrittlement, Void welling See Note I MRP,189 Table 4-1 Category A N/A Category A (SCC) Expansion N/A Water Chemistry 203.

204.

205.

Lower grid assembly; Shock Structural Support to maintain core oss of Material/Pitting and Crevice pads and flow distribution

~roSion Lower grid assembly; Shock Structural Support to maintain core Iges in DimensionsiVoid See Note I MRP,189 1--~_~f-!p_,a_d_s_b_o_lt_s

-t_'o_I_'f_lg::.,u_r_a_t_io_n_,_m_d_fl_o_w_d_is_,t_ri_b_u_ti_o_n_

__in_g

~

_I--------_+--T-,-lb-l-e-4-,-I-+------__I ILower grid assembly; Shoek Stmetural Snpport to maintain core kin!!/Stress Corrosion Crackin!!,

See Note I MRP,231

~ads bolts

'onfiguration and How distribution

'ati~n,Assisted Stress Corrosi;;n Table 3,8 and racking Table 3-10 206.

Lower grid assembly; Shock pads bolts Structural Snpport to maintain core IconfigUlmtion and flow distribution

'racking/Stress Corrosion Cracking, Vater Chemistry N/A rradiation,Assisted Stress Corrosion

,'racking N/A 207.

Lower grid assembly; Shoek pads bolts Structnral Support to maintain core Icontiguration and flow distribution

'umnlative Fatigue Damage/Fatigue LAA N/A N/A 208.

Lower grid assembly; Sbock pads bolts Structural Support to maintain core conflgnration and flow distribution Loss of Fracture Toughness/Neutron Irradiation Embrittlemcnt, Void Swelling See Note I MRP,189 Table 4-1 Category A

Appendix E-AMR and MRP-227-A, Rev. 0 Comparison ANP-2952, Rev. 001 I

Line # I Component Type Intended Function Aging Effect Requiring Management Aging Management l>rograms I Supporting Document and Location Disposition in Supporting Documene cower grid assembly; Support tru~tural Support to maintain core post pipes and flow distnbution P.lcnUlTI c.'over and plenum IStrucmral Support to maintain core

'vlinder: Bottom t1ange,to, Iconfiguration and flow distribution Ipper gnd screws

'umulative Fatigue Damage/Fatigue LAA

.'umulative Fatigue Damage/Fatigue TLAA MRP,189 Category A Table 4-1 MRP-189 Category A Table 4-1 MRP-189 Category A Table 4-1 N/A N!A N!A N!A MRP-189 Category A Table 4-1 N!A N!A MRP-189 Category A Table 4-1 MRP-189 Category A Table 4-1 N!A N!A N!A

'N!A MRP-189 Category A Table 4-1 See Note I iSee Note I

~~hanues in Dimensions/Void Swelling Loss of Material/Pitting and Crevice Vater Chemistry "orrosion Changes m DimensiOl1S/Void Sce Note I Swelling lracking!Stress Corrosion Cracking, Water Chemistry rradiation-Assisted Stress Corrosion

'racking Corrosion Crackmg, See Note I II rradiatiorl-A,s:,tst"d Stress CorroSIon

'racking Loss of Fracture Toughness, Neutron See Note I Irradiauon Embritllement, Void Swelling

'Loss of Prdowl'Stress RdaxatlOn

~.':racking/Stress Corrosion Cracking, Vater Chemistry Ilrradiation-Assisted Stress Corrosion

('racking ILoss of Fracture Toughness!Neutron See Note 1 Ilrradiation Embrittlemeut, VOId jSwelling

'racking!Stress Corrosiou Crackinll, See Note 1 1rradiati~n-Assisted Stress Corrosi~n

~'racking Structnral Support to maintain core configuration and flow distribution Structural Support to maintain core configuration and flow distribution Structural Support to maiutain core cOIrtiguration and flow distribution iStructural Support to maintain core configuration and flow distribution IStructural Support to maintain core

'onfiguration and flow distributiou IStructural Support to maintain core Iconfiguratiou and flow distribution ructural Support to maintain core nfiguration and flow distribution

!Structural Support to maintain core IconfigUlrat,:on and flow distribution

!Structural Support to maintain core IconfigUll'llti,on and flow distribution

!Structural Support to maintaiu core IconfigUirati,on and flow distribution ower grid assembly; Support ost pIpes Lower grid assembly; Support

",,' hi "pc Lower grid assembly: Shock pads bolts Lower assembly; Support post pipes Lower grid assembly: Support post pipes ower grid assembly: Support post pipes Plenum cover and plenum

'vlinder; Bottom flange,to-upper grid screws C

IPlenum cover andph

~ylinder: Bottom kIpper grid screws IPlenmn cover and plenum

[cylinder: Bottom flange-to, tupper gnd screws

,Plenum cover and plenum k,ylinder: Bottom flangc-to-

~Jpper gnd screws 212.

20<).

211.

216.

218.

219.

21 213.

220.

214.

217.

1210.

Page E-20

Appendix E~ AMR and MRp*227*A, Rev, 0 Comparison ANP*2952, Rev. 001 Line #

Component Type 221.

Plenum cover and

'ylinder; Bottom upper grid screws 222.

Plenum cover and plenum

'vlinder: Plenum cover a~sembly Intended Function Stmctural Support to maintaiu core configuration and now distributiou Stnlctural Support to maintain core confIguration and now distributiou Aging Effect Requiring

\\lanagement Loss of l'reloarfStress Relaxation

~'halIges in DimensionsiVoid Swel Aging Management Programs l

See Note I See Note I Supporting Document and Location MRP-189 Table 4*1 MRP-189 Table 4-1, Table 4-2 Disposition in Supporting Document 2

Category A Category A 223.

Pleuum cover aud plenum

'vliuder' Pleuum cover a~sembl~

Stnlctural Support to maintain core couflguration and flow distribution Corrosion Cracking, See Note I Iln'adiatiOlH'\\ssisted Stress Corrosion Cracking MRP-189 Table 4-1, Table 4*2 Category A 224.

Plenum cover and plenum

'ylinder; Plenum cover

',ssembly 225.

Plenum cover and plenum

'ylinder: Plenum cover assembly 226.

Plenum cover and plenum

'ylinder; Plenum cover

'lssembly 227.

Plenum cover and plenum

'vlinder: Plenum cover a~sembly

~ c,'over and p,lenum Icytll1acr: Plenum cylinder 229.

Plenum cover and p.leuull1 cylinder; Plenum cylinder Stnlctural Support to maintain core Iconfiguration and flow distribution

!Structural Support to maintain core

'onfiguration and now distribution IStructural Support to maintain core

'onfIguration and now distribution Stmetural Support to maintaiu core configuration and flow distrIbution Stmetural Support to maintain core configuration and flow distribution Stmctural Snpport to maiutain core configuration and now distributiou

~racking/StressCorrosion Cracking, Irradiation*Assisted Stress Corrosion

'racking i'umnlative Fatigue Damage/Fatigue IIILoss of Fracture Toughness/Neutron Irradiation Embrittlement, Void ISwelling ILoss of Material/Pitting and Crevice

~'orroslOn 1

Changes 111 DimenSIOns, Void Swel

'rackin~'Stress Corrosion Crackiug, Ilrradiati~n-AssistedStress Corrosion t'racking Water Chemistry LAA See Note I Water Chemistry See Note I See Note I N/A N/A MRP-189 Table 4-1.

Table 4-2 N/A MRP-189 Table 4-1.

Table 4-2 MRP-189 Table 4-1.

Table 4-2 N/A N/A Category A N/A Category A I

Category A 230.

231.

Plenum cover and plenum cyliuder; Plenum cylinder Plenum cover and plenum cylinder; Plenum cylinder

,Structural Support to maintain core Iconfiguratlon and flow distribution Support to maintain core and flow distribution

~iracking/StressCorrosion Cracking, Water Chemistry Ilrradiation-Assisted Stress Corrosion

~~racking iumulative Fatigue Damage/Fatigue TLAA N/A N/A NfA N/A Page E-21

Appendix E~ AMR and MRP-227-A, Rev, 0 Comparison ANP-2952, Rev, 001 Line #

Component Type Intended Function Aging !£ffed Requiring Management Aging Management I>rograms' SUIJporting Document and Location Disposition in Supportiny Document

1232, Plenum cover and plenum

'ylinder: Plenum Loss of Material/Pitting and Crevice IWater Chemistry "orrosion 1:'33, Plenum cover and plenum

~ylinder; Plenum cylinder Struetnral Snpport to maintain core tion and flow distribution Structural Support to maintain core and flow distribntlon oss of Fracture fonghnessl1"eutron rradiation Embrlttlement, Void iSweiling iSee 1"ote I MRp*189 Table 4*1.

Table 4,2 N/A Category A N/A Plenum cover and cylinder; plates king/Stress Corrosion Cracking,

[See Note I iation*Assisted Stress Corrosion

king

1234, Ipi ryl cover and plennm Reinforcing plates tural Support to maintain core guration and flow distribution truetural Support to maintain core

'onfiguration and tlow distribution nges in Dimensions/Void See Note I MRP,189 Table 4*1 MRP,189 fable 4,1 Category A Category A

236, 240, 241.

242, 243.

244, Plenum cover and plenum

'ylinder; Reint()rcing plates Plenum cover and plenum

'ylinder; Reinf()rcing plates Plenum cover aud plcnum

'ylinder; ReinftJrcing plates IPlenum cover and plcnum

~ylinder; Reinforcing plates Plenum cover and plennm

'ylinder: Rib Pads Plenum cover and plenum

'ylinder; Rib Pads Plenum cover and plenum

'yhnder; Rib Pads Plenum cover and plenum

'ylinder; Rib Pads Plenum cover and plenum cylinder; Rib Pads Stmctural Snpport to maintain core configuration and flow distribution Structural Support to maintain core configuration and !low distribution Strnctural Support to maintain core confignration and !low distribntion Stmctural Snpport to maintain core and flow distribution Structural Support to maintain core configuration and !low distribution Structnral Support to maintain core

!co,"fiI5urati,.m and flow distribution

'Stmctural Support to maintain core configuration and tlow distribution Structural Support to maintain core configuration and !low distribution Stmctural Support to maintain core configuration and flow distribution king/Stress Corrosion Cracking, Water Chemistry

,Assisted Stress Corrosion umulative Fatigue Damage/Fatigue frLAA oss of Fracture Toughness/Neutron See Note I

[Irradiation Embrittlement, Void

[Swelling Loss of Material/Pitting and Crevice IWater Chemistry

'affoslOn

~hanges in Dimensions/Void See Note 1

~welling

'racking/Stress Corrosion Cracking, See Note I Irradiation-Assisted Stress Corrosion

'racking

'racking/Stress Corrosion Cracking, twatcr Chemistry Irradiation-Assisted Stress Corrosion

,'racking

'umulative Fatigue Damage/Fatigue frLAA Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement. Void Swelling N/A N/A MRp*189 Table 4-1 N/A MRP,189 Table 4,1 MRp*189 Table 4-1 N/A N/A MRp*189 Table 4*1 N/A N/A Category A N/A Category A Category A N/A N/A Category A Page E-22

Appendix E-AMR and MRP-227-A, Rev, 0 Comparison ANP-2952, Rev, 001 Line #

Component Type Intended Function Aging Effect Requiring Management Aging Management Programs l Supporting Document and Location Disposition in Supporting l)ocument1 Category A N/A MRP,189 Table 4,1 Water Chemistry See Note I

~han2es in Dimensions/Void jSwdl" lcoum cover and pleuum IStructural Support to maiutain core Loss of Material/Pitting aud Crevice ylinder; Rib Pads

~_d_I_'s_tr_ib_u_t_t_O_u_+-'_o_rr_o_s_i(_Jn

+_-------+-------1-------4 lenum cover and plenum Structural Support to maintaIn corc ylinder; Top tlange,to-cover conllguration and flow distribntion

[bolts

245, 247, Plenum cover and plenum

'ylinder; Top flange,to-cover bolts Structural Snpport to maintain core Iconflgnration and flow distribution IClackirtglStress (\\mosion Crackmg, See Note I Ilrr'adiatlOrH'lSsisted Stress CorrosIOn

~,'racking MRP,189 Table 4,1 Category A

248, Plenum cover and plenum Feyl Top flange,to,cover

~olts Structural Support to maintain core Icontlgul'ation and flow d"tnbutiou Corrosion Cracking1 Nater Chemistry Iln'adiatiOlH'lssisted Stress Corrosion

~'rackmg N/A N/A

249, Plenum cover and plenum

'ylinder; Top flange,to-cover

[bolts Structural Support to maintain core

~onfigurationand flow distribution

'umulative Fatigue Damage/Fatigue fLAA N/A

250, r

~~enum cover and pleuum

'hnder; Top flange-to*cover Its Stmctural Support to maiutain core configuration and flow distribution Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement, Void Swelling MRP,189 Table 4*1 Category /\\

251.

IPlenum cover and plenum

'ylinder; Top t1ange*to,cover bolts Structural Support to maintain core configuration and flow distribution Loss of Preload/Stress Relaxation See Note I MRp*189 Table 4*1 Category A

252, Reactor Vessellmernals; Incore Structural Support to maintain core Guide Tube Gussets eonfiguration and flow distribution ehauges In Dimensions!\\'Old See Note I ISwe11 MRP,189 Table4*I, Table 4*2 Category A

253, Reactor Vessel Internals; Incore Structural Support to maintain core

}uide Tube Gussets cont1guration and flow distribution

'racking/Stress Corrosion Cracking, Sce Note I Irmdiation-Assisted Stress Corrosion k,'racking MRp*189 Table4,1, Table 4*2 Category A

254, Reactor Vessel Interuals; Incore Structural Support to maintain core

',uide Tube Gussets configuration and flow distribution

.-0' Corrosion Cracking,

\\Vater Chemistry II rr'adiatiorl-f'lssist,ed Stress Corrosion N/A N/A

'raeking 255.

Reactor Vessel Internals; Incore Struetural Support to maintain core Guide Tube Gussets configuration and flow distribution

'nmulative Fatigue Damage/Fatigue LAA N/A N/A

256, Reactor Vessel Internals; Incore Structural Support to maintain core

,uide Tube Gussets configuration and flnw distribution Loss of Fracture Toughness/Neutron See Note I IIrradiation Fmbrittlement, Void

~welling MRP-189 Table 4*1, Table 4*2 Category A Page E*23

Appendix E-AMR and MRP*227*A, Rev. 0 Comparison ANP*2952, Rev. 001 Line #

Component Type Intended Funetion Aging Effed Requiring Management Aging Management Programs' I Supporting

'I\\)ocument and Loeation

\\)isposition in Supporting

\\)ocumene Reactor Vessel Internals; Incore Structural Support to maintain core Guide Tube Gussets and flow distribution Loss of Material/Pitting and Crevice

'affosion Water Chemistry N/A N/A Reactor Vcssellnternals; Incore Structural Support to maintaiu core

]uide Tube Nuts configuration and flow distributiou "hanges in Dimensions/Void Swelling ISee Note I MRP-189 Table 4-1 Category A 260.

,Reactor Vessel Internals; Incore Structural Support to rnaintain core

]uide Tube Nuts coufignration and flow distribution Reactor Vessel Imernals; Incore Structural Support to maintain core

]uide Tube Nnts confignration and flow distribution

'racking/Stress Corrosion Cracking, rradiation-Assisted Stress Corrosion

.'racking

-'racking/Stress Corrosion Cracking, rradiation-Assisted Stress Corrosion

~racking

!See Note I

~Vater Chcmistry MRP-189 Table 4*1 N/A Category A N/A N/A Category A MRP-189 Table 4-1 N/A

~ee Note 1

~LAA Fatigue oss of Fracture Toughness/Neutron rradiation Embrittlement, Void welling iuide Tube Nuts configuration and flow distribution Reactor Vessel Imernals; Ineore Structural Support to maintain core

!Reactor Vessel Internals; Incore Structural Support to maintain core

]uide Tube Nuts configuration and flow distribution

262, 261.

Reactor Vessel Internals; Ineore iStrt1(:tural Support to maintain core oss of Material/Pitting and Crevice

uicle Tube Nuts and flow distribution orrosion IWater Chemistry N/A N/A Category A Category A 264.

265.

Reactor Vessel Intemals; Ineore Support to maintain core t£~!ges in Dimensions/Void l:See Note I MRP-189

~~~~eiu[J,l,:d~e~T~u[J,b~e:>~S~p",id~e",'r:"s_~_~~~lll'ilS1l~'.()t~an1l.d~fl~o)\\w,:,*..'d~i,~st:n~b,,:u~tl~*

o,,:n:..._

_U_il_lg'---~~~~~ __~__---1

-+__T_a~b_le_4~-_I_+

--I Reactor Vessellntemals; Incore'Structural Support to maintain core

'king/Stress Corrosion Cracking, iSee Note 1 MRP-189

uide Tuhe Spiders configuration and flow distribntion iation*Assisted Stress Corrosion Table 4-1

'raekiug 266.

Reactor Vessellntemals; Incore Structural Support to maintain core Iiuide Tube Spiders configuration and flow distribution

'raeking/Stress Corrosion Cracking,

~ater Chemistry Irradiation-Assisted Stress Corrosion

~racking N/A N/A 267.

268.

Reactor Vessel Internals; Incore Struetural Support to maintain core

]uide Tube Spiders confignration aud flow distribution Reactor Vessel Intemals; Incore Structural Support to maintain core

uide Tube Spiders configuration and flow distrihution

'umulative Fatigue Damage/Fatigue fLAA Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement. Void Swelliug N/A MRP-23I I

Table 3-8 and I Table 3-9 N/A Primary 269.

Reactor Vessel Intemals; Incore I~tructu.ral Support to maintain core

]uide Tube Spiders leonfiguratiou and flow distributiou Loss of Fraeture Tonghness/Thermal See Note I IAging Embrittlement MRP-23 I Table 3-8 and Table 3-9 Primary Page E-24

Appendix AMR and MRP-227-A, Rev. 0 Comparison ANP-2952, Rev. 001 Line #

Component Type Intended Funetion Aging Effect Requiring Management Aging Management Programs t I Supporting

Document and Location Disposition in Supporting Document 2

270.

27.

272.

273.

27~.

275 276.

Reactor Vessel Internals; Ineole Strnctural Snpport to maintain core Gnide Tnbe Spiders conflgnration and flow distribution eactor Internals; Inwre Strnctural Support to maintain core nide Tubes and flow distribution eactor Vessel Internals; Incore Stmctural Support to maintain core

'tride Tubes conflgnration and flow distribution Reactor Vessel Internals; Incore Strnctural Support to maintain core Guide Tubes configuration and flow distribution Reactor Vessel Internals; Incore Stmctural Support to maintain core

.)uide Tubes and flow distribution Reactor Vessel Intemals; Incorc Strnetural Support to maintain core Guide Tubes configuration and flow distribution Reactor Vessel Internals; Incore Stntctural Support to maintain core

,uide Tnbes and flow distribution Loss of Material/Pitting and Crevice

'Water Chemistry

'"~",i,,n Chan:

in DimensionslVoid See Note I Swelling Corrosion Cracking,

'See Note I Iln'adiatiorl-P\\ssisted Stress Corrosion

~racking

/racking/Stress Corrosion Cracking,

'Water Chcmistry Irradiation,Assisted Stress Corrosion

/racking

/umulative Fatigue Damage/Fatigue TLAA Loss of Fracture Toughness:Neutmn See Note I IlTadiation Embrittlement, Void Swelling Loss of Material/Pitting and Crevice Water Chemistry

'orrosion N/A MRP-189 Table 4-1 MRP,189 Table 4-1 N/A N/A MRP-189 Table 4,1 N/A N/A Category A Category A N/A N/A Category A N/A 277 278.

279.

2XO.

281.

282.

Thermal Shield Thermal Shield

'Thermal Sliield Thermal Shield Thennal Shield Tbermal Shield ISliielding Shielding Shielding Shielding Sbielding Sliielding Cbanges in DimensionslVoid See Note I Swell

'racking/Stress Corrosion Cracking,

,ee Note I Irradiation,Assisted Stress Corrosion

'racking

~'raeking/StressCorrosion Cracking, Water Chemistry Irradiation,Assisted Stress COlTosion

~'raeking

'umulative Fatigue Damage/Fatigue LAA Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlemenl, Void

,welling I.oss of Material/Pitting and Crevice Water Chemistry

'orrosion MRP,189 Table 4,1, Table 4-2 MRP,189 Table 4,1, Table 4,2 N/A N/A MRP-189 Table4,l, Table 4-2 N/A Category A Category A N/A N/A Category A N/A.

Page E-25

Appendix E~ AMR and MRP-227-A, Rev, 0 Comparison ANp-2952, Rev, 001 Line #

Component Type Intended Funetion Aging Effect Reqniring Management Aging Management Programs l

Supporting Doeument and Location Disposition in Supportin~

Document" Upper grid assembly; Fuel asscmbh support pads IStructural Support to maiutain core Icoufiguration and flow distribution

~'hanges in Dimensions/Void Swelling See Note I MRP,189 Table 4,1 Category A N/A N/A Category A except for Expansion (Alloy X,750 dowel locking weld)

N/A N/A MRp*189 Table 4-1 and MRP-23I Table 3-8 and Table 3-10 See Note I

racking

'racking lumulative Fatigue Damage/Fatigue LAA ICr'ackirrgIStl'ess Corrosion Cracking, Ilrr'adiatiOlh'lSsist,cd Stress Corrosion ICI'ackirrg/Stl'ess Corrosion Cracking, IWater Chemistry Iln'adiatiol1'l',ssist,cd Stress Corrosion Structural Support to maintain core configuration and flow distribution Upper grid assembly; Fuel

'lSsembly support pads Upper grid assembl,

'lSsembly support pads Upper assembly; Fuel

~Is~~mbiy support pads

286, 285

~

tural S,'upport to mai,nta,in core guratlon and now distribution I----+---------~

tural Support to mamtain core guration and flow distnbution Upper grid assembly; Fuel

!assembly support pads Structural Support to maintain core configuration and flow distribution Loss of Fraetore Toughness,Neutron See Note I Irradiation Embrittlement. Void ISwelling MRP,189 Table 4-1 Category A IUpper Fuel lassembly support pads Structural Support to maintain core and flow distribution Loss of Material/Pitting and Crevice Vater Chemistry

~orrosion N/A N/A

[Upper grid assembly; Rib,to,

~ing screws Structural Support to maintain core and flow distribution

'hanges in DimensionsiVoid See Note I fSwelling MRP,189 Table 4,1 Category A 290 tUpper grid assembly; Rib,to, ing screws Structural Support to maintain core

'onfiguration and tlow distribntiou

'racking/Stress Corrosion Crackmg, Sce Note I Irradiation,AsslSted Stress Corrosion

~'racking MRP-189 Table 4,1 Category A 291.

tipper grid assembly; Rib-to-

,ring screws Struetural Support to maintain core configuration and flow distribntion

'racking/Stress Corrosion Cracking, Water Chemistry Irradiation,Assisted Stress Corrosion

~:racking N/A N/A

292, Upper grid assembly; Rib-to-ing screWS Sttuetural Snpport to maintain core configuration and flow distribution

~umulative Fatigue Damage/Fatigue LAA N/A N/A

293, Upper grid assembly; Rib-to, ing screws Structural Support to maintain core confignration and flow distribution Loss of Fracture Toughness/Neutron See Note I Irradiation Embrittlement Void fSwelling MRP-189 Table 4-1 Category A 294.

Upper grid assembly; Rib,to-ring screws Struetural Support to maintaiu core configuration and tlow distribution ILoss of Material/Pitting and Crevice Water Chemistry

~orrosion N/A N/A Page E-26

Appendix AMR and MRP-227~A, Rev. 0 Comparison ANP-2952, Rev. 001 Linc #

296.

Component Type Upper grid assembly; Upper grid n b section Upper grid assembly; Upper grid nb sectIon Intended Fnnction Support to maintain core and tlow Structural Support to maintain core contlguration and flow distribution Aging Effect Requil"ing Management k.*hanges in DimensionslVoid

'racking/Stress Corrosion Cracking, rradiation*Assisted Stress Corrosion C/racking Aging Management Programs' pice Note I See Note I Supporting Document and Location MRP,189 Table 4,1 MRP,189 Table 4,1 Disposition in Supporting Document1 Category A Category A 297.

Upper grid assembly; Upper grid section Structural Support to maintain core configuration aud flow distribution Corrosion Cracking, Water Chemistry Ilrradiatior"A,ssist"d Stress Corrosion

.'racking N/A N/A

'umulative Fatigue Damage/Fatigue 1LAA 298.

299.

300.

30!

302.

Upper grid assembly; Upper grid rib seetion Upper grid assembly; Upper g~id section Upper grid assembly; Upper grid rib section

!Upper grid assembly; Upper grid rmg forging Upper grid assembly; Upper arid ring forging Structural Support to maintain core configuration and flow distribution Stmetural Support to maintain core contlguration and tlow distribution Stmetural Support to maintain core configuration and tlow distribution

!Structural Support to maintain core configuration and tlow distribntion Structural Snpport to maintain core contlguration and flow distribution Loss of Fracture Toughness/Neutron Irradiatiou Embrittlement, Void

!Swelling Loss of Material/Pitting and Crevice

'orr08ion Changes in Dimensions/Void Swelling

~

/racking/StressCorrosion Cracking, Irradiation,Assisted Stress Corrosion I ~racking Sec Note I Water Chemistry See Note I See Note I N/A MRP,189 Table 4,1 N/A MRP,189 Table 4'1 MRP,189 Table 4,1 N/A Category A N/A Category A Category A 303.

304.

305.

I~S,~::~~~ld as.'semblY; Upper Igrid riug torging Iupper grid assembly; Upper grid ring forging Structural Support to maintain core configuration and flow distribution Structural Support to maintain core configuration and tlow distribution Structural Support to maintain core contiguration and flow distribution

'racking/Stress Cortosion Cracking, Water Chemistry Irradiation,Assisted Stress Corrosion

'racking

'umulative Fatigue Damage/Fatigue TLAA Loss of Fracture Touglmess/Neutrou

~ee Note I Irradiation Embrittlement, Void Swelling N/A N/A MRP,189 Table 4,1 N/A N/A Category A 306.

Upper grid assembly; Upper Structural Support to maintain core Loss of Material/Pitting and Crevice IVater Chemistry N/A N/A grid ring torging contlguration and flow distribution'orrosion I

Note 1 In the TMI-1 LRA said "A commitment Will be made In the UFSAR supplement to (1) partiCipate In the Industry programs tor Investlgattng and managing aging effects on reactor internals; (2) evaluate and implement the results of the industry programs as applicable to the reactor internals; and (3)

Upon completion ot these programs, but not less than 24 months betore entering the period of extended operation, submit an inspection plan for reactor internals to the NRC for review and approval."

Page E-27

Appendix AMR and MRP-227-A, Rev. 0 Comparison ANP-2952, Rev. 001 C"tAonrv A component items are those for which aging effects are below the screening criteria, so that age-related degradation significance is minimal.

only the ASME Boiler & Pressure Vessel Code Section XI Examination Category B-N-3 lSI visual examinations (VT-3) will be performed component to assess potential aging effects.

Page E-28 Summary of Commitments

Summary of Commitments Page 1 of 2 Summary of Commitments The following table identifies commitments made in this document. (Any other actions discussed in the submittal represent intended or planned actions. They are described to the NRC for the NRC's information and are not regulatory commitments.)

COMMITMENT COMMITTED COMMITMENT TYPE DATE OR ONE-TIME Programmatic "OUTAGE" ACTION (Yes/No)

(Yes/No)

Applicant/Licensee Action April 19, 2013 Yes No Item 2 (Table 0-2), based upon Section 4.2.2 of the Safety Evaluation Report (SER), requires a licensee to identify which Reactor Vessel Internals (RVI) components are within the scope of license renewal.

Exelon has identified that the RVI vent valve locking device should be further reviewed with MRP-227-A methodology to determine impact on the RVI Aging Management Program (AMP) and the enclosed plan. TMI is currently working with the Pressurized Water Reactor Owners Group (PWROG) to address the action item. Exelon will submit an update of the PWROG progress in evaluating this item and a schedule for showing when Exelon will submit the results of an evaluation by April 19, 2013.

Summary of Commitments Page 2 of 2 Applicant/Licensee Action April 19, 2013 Yes No Item 6 (Table D-2), based upon Section 4.2.6 of the SER, requires a licensee to justify the acceptability of certain inaccessible components that Table 4-4 of MRP-227-A identifies as expansion components.

Exelon will submit an update of the PWROG progress in evaluating these components and a schedule for showing when Exelon will submit an evaluation for continued service or a schedule for replacement by April 19, 2013.

Applicant/Licensee Action April 19, 2013 Yes No Item 7 (Table D-2), based upon Section 4.2.7 of the SER, requires that the licensee develop plant-specific analyses to demonstrate that there is not a loss of functionality of the Incore Monitoring Instrumentation (IMI) guide tube assembly spiders and Control Rod Guide Tube (CRGT) spacer castings due to loss of fracture toughness.

Exelon will submit an update of the PWROG progress in evaluating these components and a schedule for showing when Exelon will submit an evaluation for continued service or a schedule for replacement by April 19, 2013.

TMI-12-069, Submittal of Inspection Plan for Reactor Internals

Contents

Text

Subject:

Reference:

SUMMARY

7.0 REFERENCES

1.0 INTRODUCTION

2.0 BACKGROUND

SUMMARY

7.0 REFERENCES

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