NUREG-2221, Dfc, Supp 1, Technical Bases for Changes in the Subsequent License Renewal Guidance Documents, NUREG2191, Revision 1, Draft Report for Comment and NUREG2192, Revision 1, Draft Report for Comment Draft Report for Comment Tracked C

NUREG-2221 Dfc, Supp 1, Technical Bases for Changes in the Subsequent License Renewal Guidance Documents, NUREG2191, Revision 1, Draft Report for Comment and NUREG2192, Revision 1, Draft Report for Comment Draft Report for Comment Tracked C

ML23208A006
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Issue date:	07/31/2023
From:	Sayoc E NRC/NRR/DNRL/NLIB
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References
NUREG-2221 Suppl 1
Download: ML23208A006 (1)

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Draft Document: Tracked Changes Version NUREG-2221, Supplement 1, Technical Bases for Changes in the Subsequent License Renewal Guidance Documents, NUREG-2191, Revision 1, Dra Report for Comment and NUREG-2192, Revision 1, Dra Report for Comment, Dra Report for Comment, Track Changes Version This document is a companion to NUREG-2221, Supplement 1, Dra Report for Comment (Agencywide Documents Access and Management System Accession No. ML23180A208), the ocial document for public commenng purposes. This track changes version is published to facilitate ecient iden"caon of changes relave to NUREG-2221, and to assist parcipants in preparing to discuss changes in the subsequent license renewal guidance documents in upcoming public meengs.

Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version NUREG-2221 Supplement 1 Technical Bases for Changes in the Subsequent License Renewal Guidance Documents, NUREG-2191, Revision 1, Draft Report for Comment and NUREG-2192, Revision 1, Draft Report for Comment Draft Report for Comment Tracked Changes Version Office of Nuclear Reactor Regulation Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material Non-NRC Reference Material As of November 1999, you may electronically access Documents available from public and special technical NUREG-series publications and other NRC records at the libraries include all open literature items, such as books, NRCs Library at www.nrc.gov/reading-rm.html. Publicly journal articles, transactions, Federal Register notices, released records include, to name a few, NUREG-series Federal and State legislation, and congressional reports.

publications; Federal Register notices; applicant, licensee, Such documents as theses, dissertations, foreign reports and vendor documents and correspondence; NRC and translations, and non-NRC conference proceedings correspondence and internal memoranda; bulletins and may be purchased from their sponsoring organization.

information notices; inspection and investigative reports; licensee event reports; and Commission papers and their Copies of industry codes and standards used in a attachments. substantive manner in the NRC regulatory process are maintained at NRC publications in the NUREG series, NRC regulations, The NRC Technical Library and Title 10, Energy, in the Code of Federal Regulations Two White Flint North may also be purchased from one of these two sources: 11545 Rockville Pike Rockville, MD 20852-2738

1. The Superintendent of Documents U.S. Government Publishing Office These standards are available in the library for reference Washington, DC 20402-0001 use by the public. Codes and standards are usually Internet: https://bookstore.gpo.gov/ copyrighted and may be purchased from the originating Telephone: (202) 512-1800 organization or, if they are American National Standards, Fax: (202) 512-2104 from American National Standards Institute

2. The National Technical Information Service 11 West 42nd Street 5301 Shawnee Road New York, NY 10036-8002 Alexandria, VA 22312-0002 Internet: www.ansi.org Internet: https://www.ntis.gov/ (212) 642-4900 1-800-553-6847 or, locally, (703) 605-6000 Legally binding regulatory requirements are stated only in laws; NRC regulations; licenses, including technical A single copy of each NRC draft report for comment is specifications; or orders, not in NUREG-series publications.

available free, to the extent of supply, upon written The views expressed in contractor prepared publications in request as follows: this series are not necessarily those of the NRC.

The NUREG series comprises (1) technical and Address: U.S. Nuclear Regulatory Commission administrative reports and books prepared by the staff Office of Administration (NUREG-XXXX) or agency contractors (NUREG/CR-XXXX),

Digital Communications and Administrative (2) proceedings of conferences (NUREG/CP-XXXX),

Services Branch (3) reports resulting from international agreements Washington, DC 20555-0001 (NUREG/IA-XXXX),(4) brochures (NUREG/BR-XXXX), and (5) compilations of legal decisions and orders of the E-mail: Reproduction.Resource@nrc.gov Commission and the Atomic and Safety Licensing Boards and Facsimile: (301) 415-2289 of Directors decisions under Section 2.206 of the NRCs regulations (NUREG-0750), (6) Knowledge Management Some publications in the NUREG series that are posted prepared by NRC staff or agency contractors (NUREG/KM-at the NRCs Web site address www.nrc.gov/reading-rm/ XXXX).

doc-collections/nuregs are updated periodically and may DISCLAIMER: This report was prepared as an account of work differ from the last printed version. Although references to sponsored by an agency of the U.S. Government. Neither the material found on a Web site bear the date the material U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or implied, or assumes any was accessed, the material available on the date cited legal liability or responsibility for any third partys use, or the may subsequently be removed from the site. results of such use, of any information, apparatus, product, or process disclosed in this publication, or represents that its use by such third party would not infringe privately owned rights.

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Draft Document: Tracked Changes Version NUREG-2221 Supplement 1 Technical Bases for Changes in the Subsequent License Renewal Guidance Documents, NUREG-2191, Revision 1, Draft Report for Comment and NUREG-2192, Revision 1, Draft Report for Comment Draft Report for Comment Tracked Changes Version Manuscript Completed: July 2023 Date Published: July 2023 Office of Nuclear Reactor Regulation Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 COMMENTS ON DRAFT REPORT 2 Any interested party may submit comments on this report for consideration by the NRC staff.

3 Comments may be accompanied by additional relevant information or supporting data. Please 4 specify the report number NUREG-2221 in your comments and send them by the end of the 5 comment period specified in the Federal Register notice announcing the availability of this 6 report.

7 Addresses: You may submit comments by any one of the following methods. Please include 8 Docket ID NRC-2023-0096 in the subject line of your comments. Comments submitted in 9 writing or in electronic form will be posted on the NRC website and on the Federal rulemaking 10 website http://www.regulations.gov.

11 Federal Rulemaking Website: Go to http://www.regulations.gov and search for documents 12 filed under Docket ID NRC-2023-0096.

13 Mail comments to: Office of Administration, Mail Stop: TWFN-7-A60M, U.S. Nuclear 14 Regulatory Commission, Washington, DC 20555-0001, ATTN: Division of Resource 15 Management and Administration.

16 For any questions about the material in this report, please contact: Emmanuel Sayoc, Project 17 Manager, 301-415-4084, or by e-mail at Emmanuel.Sayoc@nrc.gov, and Carol Moyer, Sr.

18 Materials Engineer, 301-415-2153, or by e-mail at Carol.Moyer@nrc.gov. Please be aware that 19 any comments that you submit to the NRC will be considered a public record and entered in the 20 Agencywide Documents Access and Management System (ADAMS). Do not provide 21 information you would not want to be publicly available.

ii Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 ABSTRACT 2 This document, Draft NUREG-2221, Revision 0, Supplement 1, Technical Bases for Changes 3 in the Subsequent License Renewal Guidance Documents, Draft NUREG-2191, Revision 1, 4 and Draft NUREG-2192, Revision 1, Draft Report for Comment Supplement 1 (NUREG-2221, 5 Supplement 1) is a knowledge management and knowledge transfer document associated with 6 Draft NUREG-2191, Revision 1, Generic Aging Lessons Learned for Subsequent License 7 Renewal (GALL-SLR) Draft Report for Comment, (GALL-SLR Report, Revision. 1, GALL-SLR 8 Report, or simply GALL-SLR), and Draft NUREG-2192, Revision 1, Standard Review Plan for 9 Review of Subsequent License Renewal Applications for Nuclear Power Plants, Draft Report for 10 Comment (SRP-SLR Revision. 1, or simply SRP-SLR).

11 TThe initial iteration of NUREG-2221, (Agencywide Documents Access and Management 12 System (ADAMS) Accession No. ML17362A126) documenteded the technical changes and 13 bases that were made from the guidance contained in Revision 2 of NUREG-1801, Revision 2, 14 Generic Aging Lessons Learned (GALL) Report, (ADAMS Accession No. ML103490041), for 15 utilities applying for first license renewal, to the updated guidance for utilities wishing to apply for 16 subsequent license renewal (i.e., for operation from 60 to 80 years), published as NUREG-17 2191, Revision 0 (ADAMS Accession Nos. ML17187A031, and ML17187A204, for Volumes 1 18 and 2 respectively) (GALL-SLR, Revision. 0) in July 2017.

19 The iInitial iteration NUREG-2221, also documented the technical changes and bases for SLR 20 that were made from the guidance contained in, along with the technical basis for these 21 changes. Changes for the review of subsequent license renewal applications (SLRAs) from 22 Revision 2 of NUREG-1800, Standard Review Plan for Review of License Renewal 23 Applications for Nuclear Power Plants, (ADAMS Accession No. ML103490036) (SRP-LR) to 24 the updated guidance of NUREG-2192, Revision 0,Standard Review Plan for Review of 25 Subsequent License Renewal Applications for Nuclear Power Plants (ADAMS Accession No.

26 ML17188A158)(SRP-SLR, Rev.Revision 0)were also discussed in that document.

27 Consequently, that document (initial NUREG-2221) provided the underlying rationale that the 28 U.S. Nuclear Regulatory Commission (NRC) staff had used to develop the subsequent license 29 renewal guidance documents.

30 This publication is a draft supplement to the initial NUREG-2221, and it documents the technical 31 changes that were made in concurrent updates to the subsequent license renewal guidance 32 documents in 2023. This document provides the underlying rationale that the NRC staff used to 33 develop Draft NUREG-2191, Revision 1, of NUREG-2191 and Draft NUREG-2192, Revision 1 34 of NUREG-2192.

35 Paperwork Reduction Act Statement 36 This NUREG provides voluntary guidance for implementing the mandatory information 37 collections in 10 CFR Part 51 that are subject to the Paperwork Reduction Act of 1995 38 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of 39 Management and Budget (OMB) under control number 3150-0021. Send comments regarding 40 these information collections to the FOIA, Library, and Information Collections Branch 41 (T6A10M), U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001, or by email to 42 Infocollects.Resource@nrc.gov, and to the OMB reviewer at: OMB Office of Information and 43 Regulatory Affairs (3150-0021). Attn: Desk Officer for the Nuclear Regulatory Commission, 44 725 17th Street NW, Washington, DC 20503; email: oira_submission@omb.eop.gov.

45 Public Protection Notification iii Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 2 information or an information collection requirement unless the requesting document displays a 3 currently valid Office of Management and Budget control number.

4 iv Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 TABLE OF CONTENTS 2 ABSTRACT ................................................................................................................................. iii 3 TABLE OF CONTENTS ............................................................................................................... v 4 LIST OF TABLES ....................................................................................................................... vii 5 EXECUTIVE

SUMMARY

............................................................................................................. xi 6 LIST OF CONTRIBUTORS ....................................................................................................... xiii 7 ABBREVIATIONS AND ACRONYMS ....................................................................................... xv 8 1 INTRODUCTION .............................................................................................................. 1-1 9 1.1 Purpose and Organization of the Document............................................................. 1-1 10 2 CHANGES TO GENERIC AGING LESSONS LEARNED REPORT, SUBSEQUENT 11 LICENSE RENEWAL REVISION 0 AND THEIR TECHNICAL BASES .......................... 2-1 12 2.1 Overview of Changes to GALL-SLR Report Chapter I - Application of the 13 American Society of Mechanical Engineers Boiler and Pressure Vessel Code ....... 2-1 14 2.2 Overview of Changes to GALL-SLR Report, Chapters II, III, IV, V, VI, VII, and 15 VIII ............................................................................................................................ 2-1 16 2.3 Chapter IXUse of Terms General Changes .......................................................... 2-2 17 2.4 Chapter X Aging Management Programs That May Be Used to Demonstrate 18 Acceptability of Time-Limited Aging Analyses in Accordance with 10 CFR 19 54.21(c)(1)(iii) ........................................................................................................... 2-3 20 2.5 Chapter XI - Aging Management Programs ............................................................. 2-3 21 2.5.1 Mechanical Aging Management Programs (XI.M Series of AMPs) .............. 2-3 22 2.5.2 Structural Aging Management Programs (XI.S Series of AMPs) .................. 2-3 23 2.5.3 Electrical Aging Management Programs (XI.E Series of AMPs)................... 2-3 24 3 SUBSEQUENT LICENSE RENEWAL CHANGES TO STANDARD REVIEW PLAN 25 FOR REVIEW OF SUBSEQUENT LICENSE RENEWAL APPLICATIONS FOR 26 NUCLEAR POWER PLANTS, REVISION 2 AND THEIR TECHNICAL BASES ............ 3-1 27 3.1 SRP-SLR Chapter 1 - Administrative Information .................................................... 3-1 28 3.2 SRP-SLR Chapter 2 - Scoping and Screening ........................................................ 3-1 29 3.3 SRP-SLR Chapter 3 - Aging Management Review ................................................. 3-1 30 3.4 SRP-SLR Chapter 4 - Time-Limited Aging Analyses (TLAAs) ................................. 3-1 31 3.5 SRP-SLR Appendices A.1, A.2, A.3, and A.4 ........................................................... 3-2 32 4 CHANGES TO TECHNICAL BASES DOCUMENTED IN INITIAL NUREG-2221 .......... 4-1 33 5 REFERENCES ................................................................................................................. 5-1 34 v

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Draft Document: Tracked Changes Version 1 LIST OF TABLES 2 Table 1-1 Crosswalk Between NUREG-2191/NUREG-2192 and the Change 3 Summaries and Technical Bases Tables in NUREG-2221 ............................... 1-2 4 Table 2-1 Description of Table Columns for GALL-SLR Chapters II through VIII............... 2-2 5 Table 2-2 Description of Table Columns for GALL-SLR Chapter IX................................... 2-2 6 Table 2-3 Description of Table Columns for GALL-SLR Chapter X.................................... 2-3 7 Table 2-4 Description of Table Columns for GALL-SLR Chapter X.................................... 2-3 8 Table 2-5 New Aging Management Review Items Added in GALL-SLR Report 9 Revision 1, Chapter ll, Containment Structures.................................................. 2-4 10 Table 2-6 New Aging Management Review Items Added in GALL-SLR Report, 11 Revision 1, Chapter III, Structures and Component Supports............................ 2-4 12 Table 2-7 New Aging Management Review Items Added in GALL-SLR Report 13 Revision 1, Chapter IV, Reactor Vessel, Internals, and Reactor Coolant 14 System ............................................................................................................... 2-5 15 Table 2-8 Table New Aging Management Review Items Added in GALL-SLR Report 16 Revision 1, Chapter V, Engineered Safety Features ........................................ 2-15 17 Table 2-9 New Aging Management Review Items Added in GALL-SLR Report 18 Revision 1, Chapter Vl, Electrical Components ................................................ 2-16 19 Table 2-10 New Aging Management Review Items Added in GALL-SLR Report 20 Revision 1, Chapter VII, Auxiliary Systems ...................................................... 2-16 21 Table 2-11 New Aging Management Review Items Added in GALL-SLR Report 22 Revision 1, Chapter VIIl, Steam and Power Conversion System ..................... 2-21 23 Table 2-12 Deleted Aging Management Review Items From GALL-SLR Revision 0, 24 Chapter Il, Containment Structures .................................................................. 2-22 25 Table 2-13 Deleted Aging Management Review Items From GALL-SLR Revision 0, 26 Chapter IIl, Structures and Component Supports............................................. 2-22 27 Table 2-14 Deleted Aging Management Review Items From GALL-SLR Revision 0, 28 Chapter IV, Reactor Vessel, Internals, and Reactor Coolant System .............. 2-22 29 Table 2-15 Deleted Aging Management Review Items From GALL-SLR Revision 0, 30 Chapter V, Engineered Safety Features........................................................... 2-26 31 Table 2-16 Deleted Aging Management Review Items From GALL-SLR Revision 0, 32 Chapter VI, Electrical Components .................................................................. 2-26 33 Table 2-17 Deleted AMR Items, Chapter VII, Auxiliary Systems ........................................ 2-26 34 Table 2-18 Deleted Aging Management Review Items From GALL-SLR Revision 0, 35 Chapter VIII, Steam and Power Conversion System........................................ 2-27 36 Table 2-19 Changes to GALL-SLR Report, Revision 0, Chapter II Aging Management 37 Review Items and Technical Bases ................................................................. 2-27 38 Table 2-20 Changes to GALL-SLR Report, Revision 0, Chapter III Aging Management 39 Review Items and Technical Bases ................................................................. 2-27 40 Table 2-21 Changes to GALL-SLR Report, Revision 0, Chapter IV Aging Management 41 Review Items and Technical Bases ................................................................. 2-28 42 Table 2-22 Changes to GALL-SLR Report, Revision 0, Chapter V Aging Management 43 Review Items and Technical Bases ................................................................. 2-64 vii Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 Table 2-23 Changes to GALL-SLR Report, Revision 0, Chapter VI Aging Management 2 Review Items and Technical Bases ................................................................. 2-66 3 Table 2-24 Changes to GALL-SLR Report, Revision 0, Chapter VII Aging Management 4 Review Items Technical Bases ........................................................................ 2-66 5 Table 2-25 Changes to GALL-SLR Report, Revision 0, Chapter VIII Aging 6 Management Review Items and Technical Bases............................................ 2-67 7 Table 2-26 Chapter IX.B - Structures and Components, Differences From Chapter IX 8 GALL-SLR Report, Revision 0, and Their Technical Bases ............................. 2-68 9 Table 2-27 Chapter IX.C - Materials, Differences From Chapter IX GALL-SLR Report, 10 Revision 0, and Their Technical Bases ............................................................ 2-68 11 Table 2-28 Chapter IX.D - Environments, Differences From Chapter IX GALL-SLR, 12 Report, Revision 0, and Their Technical Bases ............................................... 2-69 13 Table 2-29 Chapter IX.E - Aging Effects, Differences From Chapter IX GALL-SLR 14 Report, Revision 0, and Their Technical Bases ............................................... 2-69 15 Table 2-30 Chapter IX.F - Aging Mechanisms, Differences from Chapter IX GALL-SLR 16 Report, Revision 0, and Their Technical Bases ............................................... 2-70 17 Table 2-31 Chapter IX.G - References, Differences From Chapter IX GALL-SLR 18 Report, Revision 0, and Their Technical Bases ............................................... 2-71 19 Table 2-32 GALL-SLR Report, Revision 1, Chapter X, Time-Limited Aging Analyses, 20 Differences From GALL-SLR Report, Revision 0, and Their Technical 21 Bases ............................................................................................................... 2-71 22 Table 2-33 GALL-SLR Report, Revision 1, Differences from Chapter XI, Mechanical 23 Aging Management Programs, Differences From GALL-SLR Report, 24 Revision 0, and Their Technical Bases ............................................................ 2-71 25 Table 2-34 GALL-SLR Report, Revision 1, Chapter XI, Structural Aging Management 26 Programs, Differences From GALL-SLR Report, Revision 0, and Their 27 Technical Bases ............................................................................................. 2-103 28 Table 2-35 GALL-SLR Report, Revision 1, Chapter XI, Electrical Aging Management 29 Programs, Differences From GALL-SLR Report, Revision 0, and Their 30 Technical Bases ............................................................................................. 2-106 31 Table 3-1 SRP-SLR, Revision 1, Chapter 1, Section 1.1, Administrative Information, 32 and Section 1.2, Integrated Plants Assessments and Aging Management 33 Reviews Differences from SRP-SLR, Revision 0, and Their Technical 34 Bases ................................................................................................................. 3-2 35 Table 3-2 SRP-SLR, Revision 1, Chapter 2, Scoping and Screening, Differences from 36 SRP-SLR, Revision 0, and Their Technical Bases............................................. 3-2 37 Table 3-3 SRP-SLR, Revision 1, Chapter 3.1, Reactor Vessels, Internals, Coolant 38 System, Differences from SRP-SLR, Revision 0, and Their Technical 39 Bases ................................................................................................................. 3-4 40 Table 3-4 SRP-SLR, Revision 1 Chapter 3.2, Engineered Safety Features, 41 Differences from SRP-SLR, Revision 0, and Their Technical Bases ............... 3-34 42 Table 3-5 SRP-SLR, Revision 1, Chapter 3.3, Auxiliary Systems, Differences from 43 SRP-SLR, Revision 0, and Their Technical Bases........................................... 3-42 44 Table 3-6 SRP-SLR, Revision 1, Chapter 3.4, Steam and Power Conversion 45 Systems, Differences from SRP-SLR, Revision 0, and Their Technical .......... 3-52 viii Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 Table 3-7 SRP-SLR, Revision 1, Chapter 3.5, Containments, Structures, and 2 Component Supports, Differences from SRP-SLR, Revision 0, and Their 3 Technical Bases ............................................................................................... 3-56 4 Table 3-8 SRP-SLR, Revision 1, Chapter 3.6, Electrical and Instrumentation Controls, 5 Differences from SRP-SLR, Revision 0 and Their Technical Bases ................ 3-64 6 Table 3-9 SRP-SLR, Revision 1, Chapter 4.1, Identification of Time-Limited Aging 7 Analysis, Differences from SRP-SLR, Revision 0, and Their Technical 8 Bases ............................................................................................................... 3-64 9 Table 3-10 SRP-SLR, Revision 1, Chapter 4.2 (Neutron Irradiation Embrittlement) 10 Differences from SRP-SLR, Revision 0, and Their Technical Bases ............... 3-64 11 Table 3-11 SRP-SLR, Revision 1, Chapter 4.3, Metal Fatigue, Differences from SRP-12 SLR, Revision 0, and Their Technical Bases ................................................... 3-68 13 Table 3-12 SRP-SLR, Revision 1, Chapter 4.4, Environmental Qualification of 14 Electrical Equipment, Differences from SRP-SLR, Revision 0, and Their 15 Technical Bases ............................................................................................... 3-68 16 Table 3-13 SRP-SLR, Revision 1, Chapter 4.5, Concrete Containment Unbonded 17 Tendon Prestress Analysis, Differences from SRP-SLR, Revision 0, and 18 Their Technical Bases ...................................................................................... 3-68 19 Table 3-14 SRP-SLR, Revision 1, Chapter 4.6, Containment Liner Plate, Metal 20 Containments, and Penetrations Fatigue Analysis, Differences from SRP-21 SLR, Revision 0, and Their Technical Bases ................................................... 3-69 22 Table 3-15 SRP-SLR, Revision 1, Chapter 4.7, Plant-Specific TLAA, Penetrations 23 Fatigue, Differences from SRP-SLR, Revision 0, and Their Technical 24 Bases ............................................................................................................... 3-69 25 Table 3-16 SRP-SLR, Revision 1, Chapter 5.0, Technical Specification Changes, 26 Differences from SRP-SLR, Revision 0, and Their Technical Bases ............... 3-70 27 Table 3-17 SRP-SLR, Revision 1, Appendices A.1, A.2, A.3, and A.4, Differences from 28 SRP-SLR, Revision 0, and Their Technical Bases........................................... 3-70 29 Table 4-1 Description of Table Columns for Technical Bases in Initial NUREG-2221 ....... 4-1 30 Table 4-2 Changes to Technical Bases in Initial NUREG-2221 ......................................... 4-1 31 32 ix Page Numbers May Not Align with Draft for Comment Version

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Draft Document: Tracked Changes Version 1 EXECUTIVE

SUMMARY

2 On July 14, 2017 (82 FR 32588), the U.S. Nuclear Regulatory Commission (NRC) announced 3 the issuance and availability of the following final subsequent license renewal guidance 4 documents:

5

• Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report 6 (NUREG-2191), and 7

• Standard Review Plan for Review of Subsequent License Renewal Applications forNuclear 8 Power Plants (SRP-SLR) (NUREG-2192) 9 Thoese subsequent license renewal (SLR) guidance documents describe methods acceptable 10 to the staff for implementing the license renewal rule, regulations in Title 10 of the Code of 11 Federal Regulations (10 CFR) Part 54, Requirements for Renewal of Operating Licenses for 12 Nuclear Power Plants, as well as techniques used by the staff in evaluating applications for 13 nuclear power plant (NPP) license renewals for operations from 60 to 80 years. Thosee 14 guidance documentsrevisions incorporated changes described in Interim Staff Guidance issued 15 since Revision 2 of NUREG-1801, Generic Aging Lessons Learned (GALL) Report, and 16 Revision 2 of NUREG-1800, Standard Review Plan for Review of License Renewal 17 Applications for Nuclear Power Plants, published in 2010, as well as findings from NRC staff 18 aging management program effectiveness audits, and comments from NRC staff and interested 19 stakeholders.

20 The initial NUREG-2221, Revision 0, NUREG-2221 provided a summary of changes and a 21 synopsis of the bases for thoese changes made as part of the development of Revision 0 of the 22 SRP-SLR, Rev.Revision 0, and Revision 0 of the GALL-SLR Report, Rev.Revision 0.

23 Supplement 1 to Draft NUREG-2221, Supplement 1, published herewith provides a summary of 24 changes and a synopsis of the bases for theose changes made as part of the development of 25 Revision 1 of the the Draft SRP-SLR, Rev, 1, published herewith and Revision 1 of the Draft 26 GALL-SLR Report, Rev.Revision 1. This draft supplement ese changes includes those changes 27 that were initiated by NRC staff as well as the changes made in by the staff in response to 28 public comments, as appropriate. This document provides the underlying rationale that the NRC 29 used in developing the revised SLR guidance.

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Draft Document: Tracked Changes Version 1 LIST OF CONTRIBUTORS Division of New and Renewed Licenses, Office of Nuclear Reactor Regulation Brian B. Smith Division Director Bernie B. Thompson Deputy Division Director L. Gibson Branch Chief S. Bloom Branch Chief A. Buford Branch Chief M. Mitchell Branch Chief A. Hiser Senior Technical Advisor J. Wise Senior Technical Advisor M. Sayoc Project Manager M. Yoo Senior Project Manager J. Hammock Project Manager B. Rogers Senior Project Manager B. Allik Mechanical Engineer I. Anchondo-Lopez Materials Engineer L. Alvarado Materials Engineer M. Benson Materials Engineer J. Collins Senior Materials Engineer D. Dijamco Materials Engineer C. Fairbanks Senior Materials Engineer B. Fu Materials Engineer T. Gardner Physical Scientist J. Gavula Mechanical Engineer E. Haywood Materials Engineer A. Johnson Senior Materials Engineer V. Kalikian Materials Engineer G. Makar Materials Engineer J. Medoff Senior Mechanical Engineer S. Min Materials Engineering C. Moyer Senior Materials Engineer E. Reichelt Senior Materials Engineer L. Terry Materials Engineer xiii Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version J. Tsao Senior Materials Engineer D. Widrevitz Materials Engineer M. Yoder Chemical Engineering O. Yee Materials Engineering Division of Engineering and External Hazards, Office of Nuclear Reactor Regulation J. Colaccino Branch Chief W. Morton Branch Chief J. Paige Branch Chief J. Cintron-Rivera Electrical Engineer B. Lehman Structural Engineer A. Istar Civil Engineer M. Marshall Senior Project Manager M. McConnell Senior Electrical Engineer A. Prinaris Civil Engineer L. Ramadan Electrical Engineer M. Sadollah Electrical Engineer G. Thomas Senior Civil Engineer G. Wang Civil Engineer 1

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Draft Document: Tracked Changes Version 1 ABBREVIATIONS AND ACRONYMS 2 °C degree(s) Celsius 3 °F degree(s) Fahrenheit 4 AAC all aluminum conductor 5 ACI American Concrete Institute 6 ADAMS Agencywide Documents Access and Management System 7 AERM aging effect requiring management 8 A/LAI applicant or / licensee action item 9 AMPs aging management programs 10 AMR aging management review 11 ANSI American National Standards Institute 12 ASM American Society for Metals 13 ASME American Society of Mechanical Engineers 14 ASME Code American Society of Mechanical Engineers Boiler and Pressure Vessel 15 Code 16 ASTM ASTM International (formerly American Society for Testing and Materials) 17 AUX auxiliary 18 19 B&PV boiler and pressure vessel 20 B&W Babcock & Wilcox 21 BAC boric acid concentrator 22 BMI bottom mounted instrumentation 23 BWR boiling water reactor 24 BWRVIP Boiling Water Reactor Vessel and Internals Project 25 26 CASS cast austenitic stainless steel 27 CE Combustion Engineering 28 CEA control element assembly 29 CFR Code of Federal Regulations 30 CFRP carbon fiber reinforced polymer 31 CRGT control rod guide tube 32 CLB current licensing basis 33 CRD control rod drive 34 CSB core support barrel 35 CSS core support shield 36 CUF cumulative usage factor 37 38 DNRL Division of New and Renewed Licenses 39 40 EFPY effective full-power year 41 EPRI Electric Power Research Institute 42 EQ environmental qualification xv Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 ESF emergency safety feature 2

3 FAC flow-accelerated corrosion 4 FD flow distributor 5 FE further evaluation 6 ft foot/feet 7 ft2 square-foot 8 FR Federal Register 9 FRN Federal Register Notice 10 FSAR Final Safety Analysis Report 11 FWST fire water storage tanks 12 13 GALL Generic Aging Lessons Learned 14 GALL-SLR Generic Aging Lessons Learned for Subsequent License Renewal 15 GL generic letter 16 GSI generic safety issue 17 18 HDPE high density polyethylene 19 HPSI high-pressure safety injection 20 HVAC heating, ventilation, and air conditioning 21 22 I&E inspection and evaluation 23 IASCC irradiation-assisted stress corrosion cracking 24 IE irradiation embrittlement 25 IGSCC intergranular stress corrosion cracking 26 ILRT integrated leak rate test 27 IMI incore monitoring instrument 28 IN Information Notice 29 in inch/inches 30 ISGs Interim staff Staff guidanceGuidance 31 ISI inservice inspection 32 ISP integrated surveillance program 33 ISR irradiation-enhance stress relaxation 34 35 ksi kilo pound(s) per square inch 36 37 LAW lower vertical (axial) weld 38 LBB leak-before-leak 39 LCB lower core barrel 40 LCOs limiting conditions for operation 41 LERs licensee event reports 42 LFW lower flange weld 43 LGWs lower girth welds 44 LOM loss of material xvi Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 LR license renewal 2 LRA license renewal application 3 LR-ISG license renewal Interim Staff Guidance 4 LRAs license renewal applications 5 LTOP low temperature overpressure protection 6

7 MAW middle vertical (axial) weld 8 M&TE measuring and test equipment 9 MEB metal enclosed bus 10 MEAP material, environment, aging effect program 11 MeV mega electron-volt(s) 12 Mg magnesium 13 MGWs middle girth welds 14 MIC microbiologically influenced corrosion 15 mm/yr millimeter per year 16 MRP Materials Reliability Program 17 MPa megapascal 18 mpy mils per year 19 mV millivolt (mV) 20 21 N/A not applicable 22 NACE National Association of Corrosion Engineers 23 NDE Non-destructive examination 24 NEI Nuclear Energy Institute 25 NFPA National Fire Protection Association 26 NPS nominal pipe size 27 NRC U.S. Nuclear Regulatory Commission 28 NSSS nuclear steam supply system 29 30 OCCW open-cycle cooling water 31 OD outside diameter 32 OE operating experience 33 ONWs outlet nozzle welds 34 35 P-T pressure-temperature 36 PDI performance demonstration initiative 37 PLL Predicted lower limit 38 PoF probability of failure 39 PTLRs pressure-temperature limit reports 40 PTS pressurized thermal shock 41 PVC polyvinyl chloride 42 PWR pressurized water reactor 43 PWRVI Pressurized Water Reactor Vessel and Internals 44 PWSCC primary water stress corrosion cracking 45 xvii Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 QA quality assurance 2

3 RAIs request for additional information 4 RCP reactor coolant pump 5 RCPB reactor coolant pressure boundary 6 RCS reactor coolant system 7 RIS Regulatory Issue Summary 8 RG Regulatory Guide 9 RIS Regulatory Issue Summary 10 RMI reflective metal insulation 11 RPV reactor pressure vessel 12 RV reactor vessel 13 RVI reactor vessel internalRWCU reactor water cleanup 14 RWT refueling water tank 15 16 17 SCs structures and components 18 SCC stress corrosion cracking 19 SG steam generator 20 Si silicon 21 SLC Standby Liquid Control 22 SLR subsequent license renewal 23 SLRAs subsequent license renewal applications 24 SPC steam and power conversion 25 SRP standard review plan 26 SRP-SLR Standard Review Plan for Review of Subsequent License Renewal 27 Applications for Nuclear Power Plants 28 SS stainless steel 29 SSHT surveillance specimen holder tube 30 SSCs systems, structures, and components 31 32 TE thermal embrittlement 33 TMI-1 Three Mile Island Unit 1 facility 34 TLAA time-limited aging analysis 35 TRs technical or topical reports 36 TS technical specifications 37 TSTF technical specification task force 38 39 UAW upper axial weld 40 UFSAR updated final safety analysis report 41 UGW upper girth weld 42 UHS ultimate heat sink 43 U.S.S United States 44 USACE U.S. Army Corps of Engineers 45 USAR updated safety analysis report 46 USAS United States of American Standards xviii Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 USE upper-shelf energy 2 UT ultrasonic testing 3 UTS upper thermal shield 4 UV ultraviolet 5

6 VS void swelling xix Page Numbers May Not Align with Draft for Comment Version

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Draft Document: Tracked Changes Version 1 1 INTRODUCTION 2 The iInitial NUREG-2221, Technical Bases for Changes in the Subsequent License Renewal 3 Guidance Documents NUREG-2191 and NUREG-2192, establishes describes the bases for 4 the changes that guideance inthe changes incorporated in constitute updates to NUREG-2191, 5 Revision 0,Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) 6 Report, (GALL-SLR RevRevision. 0) and NUREG-2192, Revision 0, Standard Review Plan for 7 Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR, 8 Revision. 0). Thoese two subsequent license renewal (SLR) guidance documents were 9 published on July 12, 2017. NUREG-2221 is a knowledge management transfer document.

10 This document draft NUREG-2221, Supplement 1, supplement provides a summary of notable 11 technical changes and the technical bases for the the changes made by the U.S. Nuclear 12 Regulatory Commission (NRC) staff in 2023, for toto generate the Revision 0 1 version of the 13 Draft GALL-SLR Report, and Revision 0 1 of the Draft SRP-SLR.

14 1.1 Many public comments resulted in changes to the GALL-SLR Report and 15 the SRP-SLR. NUREG-2222, Disposition of Public Comments on the Draft 16 Subsequent License Renewal Guidance Documents NUREG-2191 and 17 NUREG-2192, discusses the resolution of the public comments submitted 18 to the NRC staff.

19 1.21.1 Purpose and Organization of the Document 20 This document is organized into four sections followed by the references for each section.

21 Section 11 contains background and overview information. Section 22 summarizes the changes 22 to the GALL-SLR Report and the technical bases of these changes. Section 33 presents similar 23 information for changes to the SRP-SLR. Section 44 summarizes the changes to the Summary 24 of Significant Changes and Technical Bases for Changes information in the initial NUREG-25 2221.

26 Tables are used to summarize technical materials whenever possible. Generic changes are 27 discussed in the text content at the beginning of each subsection of Sections 22, 33, and 44, 28 followed by tables showingtables showing changes to the documents.

29 Table 1-1 helps the reader navigate between the tables that summarize the notable technical 30 changes and their technical bases.

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Draft Document: Tracked Changes Version 1 Table 1-1 Crosswalk Between NUREG-2191/NUREG-2192 and the Change Summaries 2 and Technical Bases Tables in NUREG-2221 Tables With Change Summaries andTechnical Source Document and Chapter Bases New aging management reviews (AMRs) - Table 2-5Table 2-1 and Table 2-6Table 2-2 Structural Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR Report), Revision 1Chapters II and III New AMRs - Mechanical GALL-SLR Report, Table 2-7Table 2-3, Table 2-8Table 2-4, Revision 1, Chapters IV, ViI, VII, and VIII Table 2-10Table 2-6, Table 2-11Table 2-7 New AMRs - Electrical GALL-SLR Report, Table 2-9Table 2-5 Revision 1, Chapter VI Deleted AMRs - Structural GALL-SLR Report, Table 2-12Table 2-8 and Table 2-13Table 2-9 Revision 1, Chapters II and III Deleted AMRs - Mechanical GALL-SLR Report, Table 2-14Table 2-10, Table 2-15Table 2-11, Revision 1, Chapters IV, V, VII, and VIII Table 2-17Table 2-13, Table 2-18Table 2-14 Deleted AMRs - Electrical GALL-SLR Report, Table 2-16Table 2-12 Revision 1, Chapter VI Revised AMRs - Mechanical GALL-SLR Report, Table 2-21Table 2-17, Table 2-22Table 2-18, Revision 1, Chapters IV, V, VII, and VIII Table 2-24Table 2-20, Table 2-25Table 2-21 Revised AMRs - Structural GALL-SLR Report, Table 2-19Table 2-15 and Table 2-20Table 2-16 Revision 1, Chapters II and III Revised AMRs - Electrical GALL-SLR Report, Table 2-23Table 2-19 Revision 1, Chapter VI GALL-SLR Report, Revision 1, Chapter IX - Use of Table 2-26Table 2-22 through Table 2-31Table 2-27 Terms GALL-SLR Report, Revision 1, Chapter X - Time- Table 2-32Table 2-28 Limited AgingAnalysis GALL-SLR Report, Revision 1, Chapter XI - Table 2-33Table 2-29 Mechanical GALL-SLR Report, Revision 1, Chapter XI - Table 2-34Table 2-30 Structural GALL-SLR Report, Revision 1, Chapter XI - Table 2-35Table 2-31 Electrical Standard Review Plan for Review of Subsequent Table 3-1Table 3-1 License Renewal Applications forNuclear Power Plants (SRP-SLR), Revision 1, Chapter 1 SRP-SLR, Revision 1, Chapter 2 Table 3-2Table 3-2 SRP-SLR, Revision 1, Chapter 3 Table 3-3Table 3-3 through Table 3-8Table 3-8 SRP-SLR, Revision 1, Chapter 4 Table 3-9Table 3-9 through Table 3-15Table 3-15 SRP-SLR, Revision 1, Chapter 5 Table 3-16Table 3-16 SRP-SLR, Revision 1, Appendices Table 3-17Table 3-17 1-2 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Tables With Change Summaries andTechnical Source Document and Chapter Bases Changes to Summary of Changes and Technical Table 4-2Table 4-1 Bases in NUREG-2221 1

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Draft Document: Tracked Changes Version 1 2 CHANGES TO GENERIC AGING LESSONS LEARNED REPORT, 2 SUBSEQUENT LICENSE RENEWAL REVISION 0 AND THEIR 3 TECHNICAL BASES 4 The technical changes to in the Generic Aging Lessons Learned for Subsequent License 5 Renewal (GALL-SLR) Report, Revision 1 were made to clarify or improve the guidance provided 6 in GALL-SLR Report Revision 0. The U.S. Nuclear Regulatory Commission (NRC) staff believes 7 that these changes make the GALL-SLR Report, Revision. 1 more useful to the applicants and 8 to NRC staff reviewing the safety aspects of applications for subsequent license renewal (SLR).

9 Additional changes have been made as a result of public comments received during the public 10 comment period. The final version of the Standard Review Plan for Review of Subsequent 11 License Renewal Applications for Nuclear Power Plants (SRP-SLR) NUREG-2192, Revision 1, 12 incorporates the revisions to the SRP-SLR, Revision 0, based on these types of technical 13 changes.

14 2.1 Overview of Changes to GALL-SLR Report Chapter I - Application of the 15 American Society of Mechanical Engineers Boiler and Pressure 16 VesselASME Code 17 Chapter ISection 1 provides a listing of American Society of Mechanical Engineers Boiler and 18 Pressure Vessel Code (ASME Code)Section XI Editions and Addenda that are acceptable for 19 use in aging management programs (AMPs). Since publication of the GALL-SLR Report, , in 20 July 2017, the NRC has promulgated several Title 10 of the Code of Federal Regulations (10 21 CFR) Part 50.55a rulemakings, incorporating them by reference in later editions of Section XI.

22 Further, the Commission issued SRM-SECY-21-0029, directing the staff to extend the inservice 23 inspection and inservice testing Code of record update interval specified in 10 CFR 50.55a.

24 Therefore, the NRC staff updated Table I-1 to include the latest Editions of Section XI 25 incorporated by reference in 10 CFR 50.55a. Also, the staff eliminated explicit reference to 26 10-year intervals, given the NRCs plans to update the Code of record update requirements.

27 2.2 Overview of Changes to GALL-SLR Report, Chapters II, III, IV, V, VI, VII, and 28 VIII 29 The aging management review (AMR) items in Chapters II, III, IV, V, VI, VII, and VIII of the 30 GALL-SLR Report, Revision 1, are divided into five categories:

31 1. The AMR items where the material/environmental/aging effect/program combination haves 32 not changed from an equivalent item in Revision 0 of the GALL-SLR Report and there is no 33 change in the recommendation regarding further evaluation (FE). These unchanged items 34 contain no entry (i,e, are blank) in the column that identifyies new (N), modified (M), edited 35 (E), or deleted (D) items in the tables in the GALL-SLR Report.

36 2. The AMR items that are new in Revision 1 of the GALL-SLR Report. For these items, there 37 is not a clear relationship with a similar item in the same chapter of Revision 0 of the GALL-38 SLR Report. These items are identified as new (N) in the column that identifies new (N),

39 modified (M), edited (E), or deleted (D) items in the tables of the GALL-SLR Report.

40 3. The AMR items where there is some change from Revision 0 of the GALL-SLR Report with 41 regard to the material, environment, aging effect, and AMP combination or the 42 recommendation regarding further evaluationFE. However, there is a clear relationship 2-1 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 between the AMR item in Revision 1 of the GALL-SLR Report and a related AMR item in 2 Revision 0 of the GALL-SLR Report. These items are identified as modified (M) in the 3 column that identifies new (N), modified (M), edited (E), or deleted (D) items in the tables of 4 the GALL-SLR Report.

5 4. The changes to some AMR items were minor and editorial in nature. These items are 6 identified as editorial (E) in the column that identifies new (N), modified (M), edited (E), or 7 deleted (D) items in the tables of the GALL-SLR Report.

8 5. The AMR items that were in Revision 0 but have been deleted in Revision 1 of the GALL-9 SLR Report are identified as deleted (D) in the column that identifies new (N), modified (M),

10 edited (E), or deleted (D) items in the tables of the GALL-SLR Report.

11 Table 2-1 through 2-21Table 2-25 present the changes to the AMR items that have been made 12 for the GALL-SLR Report, Revision 1. The following describes the information presented in 13 each column of these tables.

14 Table 2-1 Description of Table Columns for GALL-SLR Chapters II through VIII Column Heading Description Aging management reviews Identifies the item number in GALL-SLR Report Chapters II through VIII (AMR) Item No. presenting the detailed information summarized by this row. Using II.B1.2.CP-114 as an example: The first Roman numeral presents the GALL-SLR Chapter (II) which is followed by the subchapter (B1.2). The following letter identifies the discipline(s) that the precedent (P) is associated with (i.e., A for Auxiliary Systems, E for Engineered Safety Features Systems, L for Electrical Systems, R for Reactor Coolant Systems, T for Structures and Component Supports, S for Steam and Power Conversion Systems, and C for Containment Structures). The second letter P identifies that there is a precedent for the material-environment-aging effect-program combination. This nomenclature convention is found throughout NUREG-2191 and NUREG-2192.

Technical Bases for Provides background on the staffs technical position for making the Changes change.

15 2.3 Chapter IXUse of Terms General Changes 16 Changes are made to Chapter IX to include new structures and components, materials, 17 environments, and aging effects/mechanisms, and to help standardize expressions. Changes 18 are also made to clarify some of the use of terms that were included in GALL-SLR Report, 19 Revision 0. Specific changes to the use of terms for subchapters IX.B through IX.G are 20 summarized in Table 2-26Tables 2-22 through Table 2-312-27. The following describes the 21 information presented in each column of these tables.

22 Table 2-2 Description of Table Columns for GALL-SLR Chapter IX Column Heading Description Defined Term Identifies the term.

Summary of SignificantChanges Provides a summary of the change.

Technical Bases for Changes Provides background on the staffs technical position for making the change.

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Draft Document: Tracked Changes Version 1 2.4 Chapter X Aging Management Programs That May Be Used to 2 Demonstrate Acceptability of Time-Limited Aging Analyses in Accordance 3 with 10 CFR 54.21(c)(1)(iii) 4 The title of Chapter X was revised as this chapter provides a list of AMPs (and the program 5 element criteria for the AMPs) that are commonly used to demonstrate the acceptance of 6 generic or plant-specific time-limited aging analyses (TLAAs) in accordance with Title 10 of the 7 Code of Federal Regulations (10 CFR) 54.21(c)(1). Revisions to the TLAAs for mechanical, 8 structural, and electrical analyses are discussed in Table 2-32Table 2-28. The following 9 describes the information presented in each column of the table.

10 Table 2-3 Description of Table Columns for GALL-SLR Chapter X Column Heading Description Location of Change Identifies the AMP element that changed.

Summary of SignificantChanges Provides a summary of the change.

Technical Bases for Changes Provides background on the staffs technical position formaking the change.

11 2.5 Chapter XI - Aging Management Programs 12 Table 2-33Tables 2-29 through Table 2-352-31 present the changes to the AMPs that have 13 been made for the GALL-SLR Report, Revision 1. The following describes the information 14 presented in each column of these tables.

15 Table 2-4 Description of Table Columns for GALL-SLR Chapter X Column Heading Description Location of Change Identifies the AMP element that changed.

Summary of SignificantChanges Provides a summary of the change.

Technical Basis for Change Provides background on the staffs technical position for making the change.

16 2.5.1 Mechanical Aging Management ProgramsAMPs (XI.M Series of AMPs) 17 A summary of specific changes to the mechanical AMPs and their technical bases is provided in 18 Table 2-33Table 2-29.

19 2.5.2 Structural Aging Management ProgramsAMPs (XI.S Series of AMPs) 20 A summary of specific changes to the structural AMPs and their technical bases is provided in 21 Table 2-34Table 2-30.

22 2.5.3 Electrical Aging Management ProgramsAMPs (XI.E Series of AMPs) 23 A summary of specific changes to the electrical AMPs and their technical bases is provided in 24 Table 2-35Table 2-31.

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Draft Document: Tracked Changes Version 1 Table 2-5 Table 2-1 New Aging Management ReviewAMR Items Added in GALL-SLR 2 Report Revision 1, Chapter ll, Containment Structures New Aging Management ReviewAMR Item No. Technical Bases for Changes No new aging management review (AMR) items were added to Chapter II of the GALL-SLR Report, Revision 0.

3 Table 2-6 Table 2-2 New Aging Management ReviewAMR Items Added in GALL-SLR 4 Report, Revision 1, Chapter III, Structures and Component Supports New Aging Management Review (AMR) Item No. Technical Bases for Changes III.A4.T-36 This new item addresses irradiation aging effects for the reduction in fracture toughness and potential loss of intended function on reactor vessel (RV) steel structural supports and their assembled components (e.g., girder and columns, neutron shield tank, support skirt). Such aging effects could occur due to neutrons of energy spectrum E > 0.1 mega electron-volt (MeV) at potentially damaging radiation exposure levels that may be reached during the subsequent period of extended operation. To assess potential irradiation aging effects, applicants would perform a plant-specific further evaluation as recommended in the new Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR) Section 3.5.2.2.2.7 described in Table 3-7Table 3-7.

III.A4.T-37 This new item addresses monitoring through volumetric or surface examination of reactor vesselRV steel structural support assemblies for potential defect growth in areas of combined aging effects associated with radiation exposure and high tensile stresses (greater than> 6 kilo pounds per square inch [ksi]) due to any of current licensing basis (CLB) loading conditions. In areas where there is evidence of combined aging effects and high tensile stresses, recommended ongoing examination methods are to be consistent with American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, IWA-2220 and/or 2230, with qualification of personnel to be in accordance with IWA-2300.

III.A4.TP-36 This new item addresses irradiation aging effects for loss of intended function on non-metallic, nonferrous components other than concrete (e.g., lubricants and manganese bronze alloy) used in the RV structural support assembly. Such aging effects could occur with neutrons of energy spectrum E > 0.1 MeV at potentially damaging radiation exposure levels that may be reached during the subsequent period of extended operation. To assess potential irradiation aging effects, applicants perform a plant-specific further evaluation, as recommended in new SRP-SLR Section 3.5.2.2.2.7 described in Table 3-7Table 3-7.

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Draft Document: Tracked Changes Version 1 Table 2-7 Table 2-3 New Aging Management ReviewAMR Items Added in GALL-SLR 2 Report Revision 1, Chapter IV, Reactor Vessel, Internals, and Reactor 3 Coolant System New Aging Management Review (AMR) Item No. Technical Bases for Changes IV.A1.R-457 New aging management review (AMR) items to relate to new IV.A2.R-457 further evaluation (FE) Section 3.1.2.2.10.3.

IV.B2.RP-296a The new line item applies to aging management of cracking that may occur in Westinghouse-designed control rod guide tube (CRGT) assembly guide plates (guide cards). In Electric Power Research Institute (EPRI) Materials Reliability Program (MRP)

Comment #8, EPRI questioned the basis for including the new RP-296a item in the Interim Staff Guidance (ISG). The EPRI commented that the control rod guide tube (CRGT) guide cards did not screen in for any cracking mechanisms in MRP-227, Rev.Revision 1-A. The staff did not accept that comment as a basis omitting IV.B2.RP-296a as a new GALL-SLR item in Appendix B.1 of the ISG or the analogous comment in EPRI MPR Comment #1, which challenged the basis for referencing the new RP-296a item in the update of SRP-SLR Table 3.1-1 Item 053a, as given in Appendix A of the ISG.

The staff acknowledges that the EPRI MRP did not identify the CRGT guide plates as being susceptible to any cracking mechanisms in Item W1 of Table 4-3 in EPRI Report MRP-227, Rev.Revision 1-A. However, based on lessons learned from the staffs processing of the Surry subsequent license renewal application (SLRA), the staff confirmed that the guide cards screened in for SCC and fatigue cracking mechanisms, as referenced to the EPRI MRPs 80-year Expert Panel assessment of the components in MRP-2018-022 report. Therefore, the staff found it appropriate to include GALL-SLR Item IV.B2.RP-296a as the new AMR line item that addresses potential cracking in Westinghouse-design CRGT guide plates. The new RP-296a item appropriately cites SCC and fatigue as potential cracking mechanisms, as based on lessons learned from the processing of the Surry SLRA. Management of non-cracking effects in the guide plates is addressed by the staffs modification of GALL-SLR Item IV.B2.RP-296, with the technical bases for changes made to the RP-296 line item being addressed in Table 2-21Table 2-17 of this report.

The CRGT guide plates remain as EPRI MRP-defined Primary category components for Westinghouse-designed programs per Item W1 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.B2.RP-297a The new line items apply to aging management of cracking and IV.B2.RP-298a non-cracking effects that may occur in Westinghouse-designed control rod guide tube (CRGT) lower flange welds (LFWs).

In the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP divided its augmented inspection criteria for 2-5 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes Westinghouse-design CRGT LFWs into those that would be performed on the LFWs in the peripheral (outer) CRGT assemblies (which were defined as Primary category components per Item W2 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report), and those that would be performed on the LFWs in the remaining CRGT assemblies (i.e., the LFWs in the non-peripheral assemblies, which were defined as Expansion category components per Item W2.1 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report). Table IV.B2 in the GALL-SLR Rreport already includes Items IV.B2.RP-298 and IV.B2.RP-297 as applicable AMR line items for the managing cracking effects and non-cracking effects in peripheral CRGT assembly LFWs as the Primary designated components.

However, the previous version of the GALL-SLR Rreport did not include any AMR line items to address aging management of LFWs located in the non-peripheral CRGT assemblies. Thus, the staff developed the new IV.B2.RP-298a and IV.B2.RP-297a items to be the new AMR line items for cracking and non-cracking effect and mechanism combinations that apply to the LFWs in non-peripheral CRGT assemblies.

The cracking mechanisms cited for the non-peripheral CRGT LFWs in the new RP-298a item and the non-cracking mechanisms cited for the non-peripheral CRGT LFWs in the RP-297a item are based on lessons learned from the staffs processing and review of the RVI gap analysis in the Surry SLRA and the aging mechanisms that were cited for CRGT LFWs, as referenced to the EPRI MRPs 80-year Expert Panel assessment of the components in MRP-2018-022 report. This includes the cracking mechanisms of stress corrosion cracking (SCC),

irradiation-assisted stress corrosion cracking (IASCC) and fatigue, and the non-cracking mechanism of neutron irradiation embrittlement (IE), with thermal embrittlement (TE) being potentially applicable if the CRGT lower flanges are made from cast austenitic stainless steel (SS).

IV.B2.RP-345a The new line item applies to aging management of cracking that may occur in the core barrel flange of Westinghouse-designed reactor units. In EPRI MRP Comment #10, EPRI questioned the basis for including the new IV.B2.RP-345a item in the ISG.

Specifically, EPRI commented that: (1) cracking of the core barrel flange base metal did not need to be addressed by a new AMR line item and (2) the core barrel flange weld that is the flange location that is susceptible to potential cracking mechanisms. The staff did not accept that comment or EPRIs analogous comment in EPRI MRP Comment #1 for referencing the new RP-345a item Item in the update of SRP-SLR Table 3.1-1 Item 053c, as provided in Appendix A of the ISG.

The staff acknowledges that the EPRI MRP did not cite any cracking mechanisms for the core barrel flanges in Item W10 of Table 4-9 in the MRP-227, Revision. 1-A reportReport. However, the need for development of the new RP-345a items Items is 2-6 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes based on lessons learned from the staffs review and processing of the RVI gap analysis in the Surry SLRA. Specifically, in reactor vessel internal (RVI) gap analysis of the Surry SLRA, the staff confirmed that the applicant of the SLRA cited stress corrosion cracking (SCC) and fatigue as applicable cracking mechanisms for the core barrel flanges, as based on the results of the EPRI MRPs 80-year Expert Panel process performed for the components in MRP-2018-022. Thus, the staff developed the new IV.B2.RP-345a line item to address potential cracking in Westinghouse-design core barrel flanges, as based on the lessons learned from the staffs review and processing of the Surry SLRA.

The core barrel flange remains as an Existing Program component for Westinghouse-designed aging management programs (AMPsS) per Item W10 in Table 4-9 of the MRP-227-A or MRP-227, Revision. 1-A reportReport.

IV.B2.RP-280a The new line item applies to aging management of loss of fracture toughness or changes in dimension that may occur in Westinghouse-design core barrel assembly lower flange welds (LFWs). In EPRI MRP Comment #12, EPRI commented that the new line item is not appropriate or needed for the ISG because the core barrel LFWs are located near the bottom of the core barrel, where the excepted fluence exposures would not be high enough to induce irradiation- assisted or enhanced mechanisms in the welds (e.g., IE, void swelling [VS], or irradiated stress corrosion cracking [IASCC, as mentioned in another analogous comment, EPRI MRP Comment #11]). The staff did not accept that the rationale made in EPRI MRP Comment #12 was sufficient to exclude the GALL-SLR IV.B2.RP-280a iitem as a newly developed item for the ISG.

Specifically, the staff acknowledges that the EPRI MRP did not identify the core barrel LFWs as being RVI components that are susceptible to irradiation mechanisms (IE, VS, or IASCC) in Item W3.3 of Table 4-6 in the MRP-227, Revision. 1-A Rreport.

However, the MRP-227, Revision. 1-A report Report is based on a 60-year aging assessment, and the need for development of the new RP-280a item is based on lessons learned from the staffs review and processing of the RVI gap analysis in the Surry SLRA.

Specifically, in the 80-year RVI gap analysis of the Surry SLRA, the staff confirmed that the applicant for the SLRA indicated that the core barrel LFWs are located in fluence exposure zones high enough to screen the components in for IE and VS (and IASCC) mechanisms, as based on the results of the EPRI MRPs 80-year Expert Panel process performed for the components in MRP-2018-022. Thus, the staff developed the new IV.B2.RP-280a line item is based on the lessons learned and criteria docketed in the Surry SLRA and not on the 60-year assessment basis for the LFWs in Item W3.3 of Table 4-6 in the MRP-227, Revsion. 1-A Rreport.

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Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes The core barrel LFWs continue to be defined as EPRI MRP Expansion category components for Westinghouse-design RVI management programs.

IV.B3.RP-333a The new line item applies to aging management of loss of fracture toughness due to irradiation embrittlement (IE) in Combustion Engineering (CE)-design core support barrel (CSB) lower girth welds (LGWs), which may be referenced as lower flange welds

[LFWs]). In EPRI MRP Comment #21, EPRI commented that the new RP-333a Iitem is unnecessary and not needed for the objectives of the ISG because the LGWs/LFWs are located in portions of the core support barrel that do not receive sufficient fluence exposures to warrant initiation of irradiation mechanisms (e.g., IE or irradiation-assisted stress corrosion cracking [IASCC],

as mentioned in the analogous comment for the welds in EPRI MRP Comment #20). The staff did not accept that the rationale made in EPRI MRP Comment #21 was sufficient to exclude the GALL-SLR IV.B2.RP-333a iitem as a newly developed item for ISG Appendix B.1 or to exclude IE as an applicable irradiation mechanism for the CSB LGWs/LFWs.

Specifically, the staff developed the new RP-333a Iitem to be consistent with Item C5.1 in Table 4-5 of the MRP-227, Revision.

1-A Rreport, as modified by the EPRI MRPs response to RAI 26, Item a (ADAMS Accession ML17305A056), which was evaluated and approved in the staffs April 25, 2019 safety evaluation for the MRP-227, Revision. 1 Rreport (ADAMS Accession No. ML19081A001). In the request for additional information (RAI) response, the EPRI MPR downgraded the core support barrel (CSB) assembly lower girth welds (LGWs) as Expansion category components for Combustion Engineering (CE)-designed pressurized water reactors (PWRs), but screened the CSB LGWs/LFWs in for stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), fatigue, and neutron irradiation embrittlement (IE) aging mechanisms. The EPRI MRP-227, Revision. 1-A reportReport, as supplemented by the RAI response, is based on a 60-year assessment. Thus, any additional neutron fluence exposures to the CSB LGWs/LFWs over an 80-year licensed life would further support the screening of IASCC and IE as applicable irradiation mechanisms for the LGWs/LFWs. Thus, the staff developed the new IV.B2.RP-333a line item for the CSB LGWs/LFWs based on the aging mechanisms identified in the EPRI MPRs response to RAI 26, Item a (as referenced in ADAMS Accession ML17035A056), and not on the 60-year assessment basis for the LFWs in Item W3.3 of Table 4-6 in the MRP-227, Revision. 1-A Rreport.

The CSB LGWs/LFWs remain as EPRI-defined Expansion category components for CE PWR RVI management programs per Item C5.1 in Table 4-5 of the MRP-227, Revision. 1-A reportReport.

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Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes IV.B3.RP-338a The new line item applies to aging management of loss of fracture toughness due to neutron irradiation embrittlementIE in the fuel alignment plates of CE-designed PWRs whose core shrouds are assembled from welded full height shroud plates.

For Combustion Engineering-designed (CE-designed ) plants with this type of shroud design, the fuel alignment plate in the upper internals assembly remains as a Primary category component per Item C10 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Item C10, the EPRI MRP screened the fuel alignment plate as being potentially susceptible to the cracking mechanism of fatigue and the non-cracking mechanism of irradiation embrittlement (IE). Management of cracking in the fuel alignment plate is addressed by the existing AMR line item in GALL-SLR Item IV.B3.RP-338, as administratively edited in Table B.2 of the ISG. However, the previous version of the GALL-SLR Rreport did not include any AMR line item to address loss of fracture due to neutron irradiation embrittlementIE in the fuel alignment plates. . Thus, the staff developed the new IV.B2.RP-338a line item based on the EPRI MRPs criteria for the plates in Item C10 of Table 4-2 in the MRP-227, Revision. 1-A reportReport.

The fuel alignment plates remain as EPRI-defined Primary category components for CE PWR RVI management programs per Item C10 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.B3.RP-320a The new line item applies to aging management of stress corrosion cracking (SCC) in core stabilizing lugs and shims (as associated bolting) of Combustion Engineering (CE)-design PWRs. The staff did not receive any Nuclear Energy Institute (NEI) or EPRI MRP comments specific to the new IV.B2.RP-320a line item.

In Item C17 of Table 4-8 in the MRP-227, Revision 1-A reportRevision 1-A ReportReport, the EPRI MRP added the core stabilizing lugs and shims (and their associated bolts) in CE-designed PWRs as Existing Program components for CE-designed reactor units, with the applicable aging mechanism being identified and cited as stress corrosion cracking (SCC). However, the previous version of the GALL-SLR Rreport did not include any AMR line item to address cracking due to SCC in the core stabilizing lugs and shims. Thus, the staff developed the new IV.B2.RP-320a line item based on the EPRI MRPs criteria for the plates in Item C17 of Table 4-8 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.B4.RP.245c The new line item applies to aging management of loss of material due to wear and loss of preload due to thermal or irradiation-enhanced stress relaxation or creep in surveillance specimen holder tube (SSHT) bolts or studs. The line item only applies to the SSHT bolts or studs in the Babcock & Wilcox (B&W)-designed PWR at the Davis- Besse Nuclear Plant. The staff did not receive 2-9 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes any Nuclear Energy Institute (NEI) or EPRI MRP comments specific to the new IV.B2.RP-245c line item.

In Item B7.2 of Table 4-4 of the MRP-227, Revision 1-A ReportRevision 1-A Report, the EPRI MRP identified that the SSHT bolts or studs as Expansion category components for the EPRI MRP PWR internals program that applies to the Davis-Besse Nuclear Plant, with the applicable cracking mechanisms being identified and cited as stress corrosion cracking (SCC) and fatigue and the applicable non-cracking mechanism being identified and cited as wear and irradiation-enhance stress relaxation or creep (ISR/IC). However, the previous version of the GALL-SLR Rreport did not include any AMR line item to address loss of material due to wear or loss of preload due to ISR/IC in the SSHT bolts or studs. Thus, the staff developed the new IV.B2.RP-245c line item for the cited SSHT bolts based on the EPRI MRPs criteria for the SSHT bolts in Item B7.2 of Table 4-4 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The SSHT bolts remain as Expansion category components for the Davis- Besse RVI management program per Item B7.2 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.B4.RP-247c The new line item applies to aging management of loss of material due to wear and loss of preload due to thermal or irradiation-enhanced stress relaxation or creep in the lower core barrel (LCB) bolts of B&W-designed PWRs.

In Item B8 of Table 4-1 in the MRP-227, Revision 1-A Rreport, the EPRI MRP added the LCB bolts and their locking devices as Primary category components for B&W-designed PWRs, with the applicable aging mechanisms for the bolts being identified and cited as stress corrosion cracking (SCC), irradiation-assisted creep or stress relaxation (IC/ISR), fatigue, and wear. The GALL-SLR Rreport already included Item IV.B4.RP-247 to address cracking in the LCB bolts and Items IV.B4.RP-247a and IV.B4.RP-247b to address cracking and non-cracking effects in the LCB bolt locking devices. However, the previous version of the GALL-SLR Rreport did not include any AMR line item to address non-cracking effect and mechanism combinations (i.e., loss of material due to wear and loss of preload due to thermal and irradiation-enhanced stress relaxation or creep) in the LCB bolts.

Thus, the staff developed the new IV.B2.RP-247c line item for the referenced LCB bolts based on the EPRI MRPs criteria for the bolts in Item B8 of Table 4-1 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The LCB bolts (and the associated LCB bolt locking devices) remain as Primary category components for B&W-design RVI management programs per Item B8 in Table 4-1 of the MRP=227, Rev. 1-A reportRevision 1-A Report.

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Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes IV.B4.RP-252a In SLR-ISG-2021-01-PWRVI, the staff deleted the old version of GALL-SLR Item IV.B4.RP-252a (which pertained to management of cracking in vent valve top and bottom retaining rings and associated locking devices) and instead replaced it with a new version of the GALL-SLR Item IV.B4.RP-252a item that serves as the new AMR line item for managing loss of fracture toughness due to thermal aging embrittlement in the vent valve bodies that are included in B&W PWR internal designs.

In Item B2.1 of Table 4-4 in the MRP-227, Revision 1-A Rreport, the EPRI MRP included the vent valve bodies as designated Expansion category components for B&W-design PWRs, with the need for inspecting the components being tied to the results of Primary inspections that will be performed on the control rod guide tube (CRGT) spacer casting as the lead Primary cast austenitic stainless steel (CASS) components for the B&W programs. The EPRI identified that thermal embrittlementTE is an applicable loss of fracture toughness mechanism for the vent valve bodies because the valve bodies are made from CASS.

However, the previous version of the GALL-SLR Rreport did not include any AMR line items to address loss of fracture toughness in the vent valve bodies. Thus, the staff developed the IV.B4.RP-252a item for the vent valve bodies to be consistent with Item B2.1 in Table 4-4 of the MRP-227, Revision 1-A Rreport, and for practical purposes, the RP-252a item is being treated as a new item for the objectives of the ISG.

The vent valve bodes are Expansion category components for B&W-designed reactor units per Item B2.1 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.B4.RP-252b The new line item applies to aging management of non-cracking effect and mechanism combinations in specific types of original locking devices that are included the vent valve assemblies of B&W-design PWRs. The staff developed the new IV.B4.RP-252b item to be consistent with Items B4 and B5 in Table 4-1 of the MRP-227, Revision 1-A Rreport. The staff did not receive any NEI or EPRI MRP comments specific to the IV.B2.RP-252b item.

In Item B4 of Table 4-1, the EPRI MRP identified that the locking devices associated with the pressure plate, spring and spring retainer, and U-cover in the vent value assembly may be susceptible to loss of material that is induced by a wear mechanism in the components. Similarly, in Item B5 of Table 4-1, the EPRI MRP identified that the locking devices associated with the key ring and pin in the assembly may be susceptible to loss of fracture toughness that is induced by a thermal embrittlementTE mechanism. However, the previous version of the GALL-SLR Rreport did not include any AMR line item to address loss of material due to wear and loss of fracture toughness in specified types of original locking device components. Thus, the staff developed the new IV.B4.RP-252b item for the referenced original 2-11 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes locking devices to be consistent with Items B4 and B5 in Table 4-1 of the MRP-227, Revision 1-A reportRevision 1-A Report.

The referenced original locking devices are defined as Primary category components for B&W-designed reactor units per Items B4 and B5 in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.B4.RP-252c The new line item applies to aging management of cracking in specific types of original locking devices and modified locking devices that may be included the vent value assemblies of B&W-designed PWRs.

In Item B6 of Table 4-1, the EPRI MRP identified that the original locking devices (as associated with key ring and pin in B&W-designed plants that include the components) and modified locking devices (as associated with the bolt locking cup, jackscrew locking cup, and bolted block in B&W plants that include the components) may be susceptible to cracking that is induced by a stress corrosion crackingSCC mechanism (SCC). The EPRI MRP has designated these components as Primary category components for to the B&W-designed PWRs that have the types of modified locking devices in the plant designs. The previous version of the GALL-SLR Rreport did not include any AMR line item to address cracking in these types of components. Thus, the staff developed the new IV.B4.RP-252c Iitem to be consistent with Item B6 in Table 4-1 of the MRP-227, Revision 1-A reportRevision 1-A Report.

The referenced original locking devices and modified locking devices are defined as Primary category components per Items B5 and B6 in Table 4-1 of the MRP-227, Revision 1-A reportRevision 1-A Report. Note 4 in Table 4-1 of MRP-227, Rev.Revision 1-A defines which of the B&W-designed PWRs in the U.S.AUS include the referenced original locking devices and which of the B&W-designed PWRs in the U.S.AUS include the referenced modified locking devices.

IV.BR.RP-246c The new line items apply to aging management of Babcock and IV.B4.RP-246d Wilcox (B&W)-design core barrel assembly upper thermal shield IV.B4.RP-246e (UTS) bolts and their locking devises, as established in Item B7.1 of Table 4-4 in the MRP-227, Revision 1-A reportRevision 1-A Report. The staff did not receive any NEI or EPRI MRP comments specific to the development of the new IV.B4.RP-246c, IV.B4.RP-246d, and IV.B4.RP-246e line items.

In Item B7.1 of Table 4-4 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP designated that the UTS bolts and their locking devices as Expansion category components for B&W-designed PWRs, with the need for inspecting the UTS bolts and locking devices being dependent on the results of Primary inspections that will be performed on upper core barrel (UCB) bolts, lower core barrel (LCB) bolts, and flow distributor (FD) bolts that are included in the plant design and are 2-12 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes designated as Primary components for B&W-design programs. In Item B7.1, the EPRI MRP screened the UTS bolts in for the aging mechanism of stress corrosion cracking (SCC) and the UTS bolt locking devices in for the aging mechanisms of fatigue, distortion (the staff assumes distortion may be associated with void swelling), and wear.

The previous version of the GALL-SLR Rreport did not include any AMR line items to address aging in the UTS bolts and bolt locking devices. Thus, the staff developed the new line items to be consistent with the criteria in Item B7.1 of Table 4-4 in the MRP-227, Revision 1-A reportRevision 1-A Report, where: (1) the IV.B4.RP-246c Iitem has been developed to address management of cracking due to SCC in the UTS bolts, (2) the IV.B4.RP-246d Iitem has been developed to address management of cracking due to fatigue in the UTS bolt locking devices, and (3) the IV.B4.RP-246e Iitem has been developed to address management of loss of material due to wear and changes in dimension due to void swelling or distortion in the UTS bolt locking devices.

IV.B4.RP-386 The new line item applies to aging management of loss of material due to neutron irradiation embrittlementIE in the lower grid rib sections of B&W-design reactor units. The staff did not receive any NEI or EPRI MRP comments specific to the development of the new IV.B4.RP-386 Iitem.

In Item B10.3 in Table 4-4 of the MRP-227, Revision 1-A reportRevision 1-A Report, the EPRI MRP established that the lower grid rib section may be susceptible to loss of fracture toughness that may be induced by a neutron irradiation embrittlement (IE) mechanism. However, the previous version of the GALL-SLR Rreport did not include any AMR items to address loss of fracture toughness in the lower grid rib sections of B&W-design PWRs. Thus, the staff developed the new IV.B4.RP-386 item to be consistent with Item B10.3 in Table 4-4 of the MRP-227, Revision 1-A reportRevision 1-A Report.

The lower grid rib sections are defined as Expansion category components for B&W-designed reactor units per Item B10.3 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.C1.R-456 As addressed in NRC Bulletin 88-08, non-isolable branch lines IV.C2.R-456 connected to the reactor coolant system may be subject to unacceptable thermal stress that can cause thermal fatigue cracking and leakage failure. The NRC Bulletin 88-08 states that, when such piping is identified, actions should be taken to ensure that the piping will not be subject to unacceptable thermal stress.

Industry operating experience and evaluation indicate that, in some branch lines, thermal stratification or mixing cycles can occur due to the interaction between the hot swirl penetration from the reactor coolant system and the cold water in-leakage from a leaking valve. In other branch lines, thermal stratification or mixing 2-13 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes cycles can result from the interaction of the hot swirl penetration and the cold water in the normally cool, stagnant branch lines without a leaking valve. In addition, cold or hot fluid injections can cause thermal fatigue in the reactor coolant system as indicated in ASME Code Case N-716-1. Therefore, cracking due to thermal fatigue can occur due to cyclic stresses from the thermal stratification, mixing or injection cycles.

The industry guidance to manage the thermal fatigue in the PWR branch lines is described in EPRI MRP-146, Revision 2. The guidance provides methods for screening and evaluating the susceptibility of non-isolable branch lines to thermal fatigue. The MRP-146, Revision 2 also provides general guidance for monitoring valve in-leakage and thermal stress as needed and performing volumetric examinations on the susceptible locations (e.g., examination areas, volumes and frequencies). These guidelines continue to be enhanced based on the lessons learned from relevant operating experience and research activities. The Boiling Water Reactor Vessel and Internals Project (BWRVIP-155), Revision 1 also describes the evaluation of thermal fatigue susceptibility in the branch lines of BWR reactor coolant pressure boundary.

In comparison, the inservice inspection (ISI) requirements in Table IWB-2500-1 of ASME Code,Section XI do not include a specific examination item for thermal fatigue cracking in ASME Code Class 1 components (reactor coolant pressure boundary). However, alternative risk-informed inservice inspectionsISIs typically include an examination item for thermal fatigue cracking (e.g., as specified in ASME Code Case N-716-1 that has been approved in NRC Regulatory Guide (RG) 1.147, Revision 18). Therefore, the existing inservice inspectionISIs at plants may include the piping locations susceptible to thermal fatigue.

Currently, the SRP-SLR does not include a further evaluation section that addresses aging management for the piping locations susceptible to thermal fatigue. Therefore, new SRP-SLR Sections 3.1.2.2.16a and 3.1.3.2.16a are added to address the adequacy of a plant-specific aging management program (e.g., adequate selection of susceptible locations for inspections, timely detection of cracks and preventive action for valve in-leakage).

Changes are also made to the SRP-SLR section for references (Section 3.1.6). In addition, relevant changes are made to the aging management review (AMR) tables in the SRP-SLR and GALL-SLR Report.

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Draft Document: Tracked Changes Version 1 Table 2-8 Table 2-4 New Aging Management ReviewAMR Items Added in GALL-SLR 2 Report Revision 1, Chapter V, Engineered Safety Features New Aging Management Review (AMR) Item No. Technical Bases for Changes V.A.E-443b The staff added aluminum alloyAlloy 6063T6 to the list of materials V.A.E-443c that are not susceptible to stress corrosion cracking (SCC). This is V.A.E-443d a wrought material alloyed primary with Magnesium magnesium V.B.E-443b (Mg) and Silicon silicon (Si). It is a moderate strength precipitation V.B.E-443c hardened aluminum alloy in the peak-aged condition. The V.B.E-443d strengthening phase precipitated during the artificial aging of 6063 V.D1.E-443b is Mg2Si. Generally, 6xxx series alloys have satisfactory SCC V.D1.E-443c resistance and inservice performance. However, some 6xxx series V.D1.E-443d alloys are known to be susceptible to SCC when exposed to V.D2.E-443b certain atypical processing histories. The majority of 6xxx series V.D2.E-443c SCC testing and characterization has been performed on 6061T6, V.D2.E-443d which is known to be resistant to SCC. Much more limited SCC testing and characterization has been performed on 6063T6; although, results have been consistent with those of 6061T6. Alloy 6063 is a compositionally leaner version of 6061 which has been optimized for extrusion. The two alloys have the same strengthening mechanism and their nominal Mg/Si ratios are also similar. Therefore, it is expected that the SCC performance is comparable. Additionally, the known inservice performance of aluminum alloyAlloy 6063T6 has shown satisfactory SCC resistance across multiple industries. Based on the metallurgical characteristics, available laboratory testing, and known service history, the staff has determined that 6063T6 is not susceptible to SCC.

V.A.E-475 Subsequent to issuance of the GALL-SLR Report, the staff V.D1.E-475 recognized that to be consistent with other GALL-SLR Report V.D2.E-475 items associated with heat exchanger tubes, E-475 should have also cited reduction of heat transfer due to fouling. This is consistent with GALL Report Revision 2 item SP-41 where a material (i.e., stainless steel [SS]) that is not susceptible to loss of material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

Titanium components are subject to flow blockage due to fouling due to potential debris in the raw water environment.

V.B.R-457 New AMR items to relate to new further evaluationFE Section V.C.R-457 3.2.2.2.11.

V.D1.R-457 V.D2.R-457 V.E.R-457 V.A.E-478 This new line item was added to support the new AMP, GALL-SLR Report AMP XI.M43, High Density Polyethylene (HDPE)

Piping and Carbon Fiber Reinforced Polymer (CFRP) Repaired Piping. The technical basis for this new AMP can be found in Table 2-33Table 2-29, XI.M43 of this report.

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Draft Document: Tracked Changes Version 1 Table 2-9 Table 2-5 New Aging Management ReviewAMR Items Added in GALL-SLR 2 Report Revision 1, Chapter Vl, Electrical Components New Aging Management ReviewAMR Item No. Technical Bases for Changes No new aging management reviewAMR items were added to Chapter VI of the GALL-SLR Report, Revision 1.

3 Table 2-10 Table 2-6 New Aging Management ReviewAMR Items Added in GALL-SLR 4 Report Revision 1, Chapter VII, Auxiliary Systems New Aging Management Review (AMR) Item No. Technical Bases for Changes VII.I.AP-182 The staff added new AMR items to add carbon fiber reinforced VII.I.A-420 polymer (CFRP) repaired piping, crediting the new AMP XI.M43, VII.I.A-538 High Density Polyethylene (HDPE) Piping and Carbon Fiber VII.C1.A-792 Reinforced Polymer (CFRP) Repaired Piping. The new AMR items reflect the recent introduction and increasing use of CFRP repaired piping at reactor facilities. The unique aging issues and aging management approaches for CFRP repaired piping were considered to be most effectively addressed with a dedicated AMP.

VII.C1.A-400b Based on industry request, included AMR items for managing VII.C3.A-400b recurring internal corrosion of metallic components exposed to raw water that are not covered by Generic Letter (GL) 89-13 by the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program.

VII.D.A-414 Added new AMR items with air and condensation environments VII.D.A-416 since these environments were added to the scope of the Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks program with the issuance of SLR-ISG-2021-02-MECHANICAL, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance.

VII.E2.A-798 The staff modified the AMR Item V.A.E-434 and added AMR Item VII.E2.A-798, to note that the aging effects of loss of material, and long-term loss of material due to general corrosion on steel exposed to an environment of treated water and sodium pentaborate can be managed by the Water Chemistry and One-Time Inspection AMPs. No item was added to manage stress corrosion cracking of steel in this environment as the GALL-SLR already states that steel components typically are not susceptible to stress corrosion cracking and are mainly susceptible to loss of material.

The staff determined that this material, environment, aging effect program (MEAP) may be managed with the AMPs cited above because the Water Chemistry AMP can monitor and control the concentration of deleterious species in the water storage tanks that provide water to the Standby Liquid Control (SLC) system which contains the sodium pentaborate solution. Additionally, the One- Time Inspection AMP can verify the corrosion rate of the 2-16 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes steel components is low enough that loss of material is unlikely to cause a loss of intended function.

Several reports were reviewed by the staff to make this determination ([NUREG/CR-6001][EPRI Report 1010639][Metals Handbook Desk Edition, 2nd Edition][EPRI Report 1000975]).

These reports concluded that even though the pH of the SLC system varies with temperature, it is generally greater than 6.8 pH which is close to neutral [NUREG/CR-6001]. Additionally, these reports noted that the pH range in SLC systems tends to be between 6.8- - 8.5 [EPRI Report 1010639]. This would result in less corrosion of the steel as the corrosion rate of steel tends to decrease with an increasing (i.e. more basic) pH (i.e., more basic) and would need additional impurities (e.g., salts, oxygen) for appreciable corrosion to occur in this environment ([Metals Handbook Desk Edition, 2nd Edition)]. Additionally, one report found that corrosion rates of carbon and low-alloy steel, when exposed to varying concentrations of boric acid, were relatively low (0.05-1.1 millimeter per year [mm/yr] or 0.002-0.045 inches per year [in/yr]), when the temperature was below 60 degrees Celsius (°C) (140 degrees Fahrenheit [°F]). (0.05 - 1.1 mm/year (0.002 - 0.045 inches/year)), when the temperature was below 60°C (140°F)([EPRI Report 1000975)].

VII.G.A-805 A new AMR item for subliming compounds used as fireproofing/fire barriers is being added to NUREG-2191 because they are materials that are widely used throughout industry and are likely to be cited in future SLRAs. The aging effects and aging mechanisms for subliming compounds used as fireproofing/fire barriers exposed to air are based on the NRC staffs review and approval of applicants programs for aging management of fire protection materials listed in previous SLRAs. In addition, the aging effects and aging mechanisms are consistent with Section 6, Fire Barriers, of EPRI Report 3002013084, Long-Term Operations: Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), issued November 2018, and those cited by industry as part of SLRA lessons learned activities and public comments on the draft AMR item.

New AMR Iitem A-805 manages loss of material due to abrasion, flaking, and vibration; cracking/delamination due to chemical reaction and settlement; change in material properties due to gamma irradiation exposure; and separation for subliming compounds (Thermo-lag, Darmatt', 3M' Interam', and other similar materials) exposed to air.

The periodic inspections recommended by AMP XI.M26, Fire Protection, are capable of detecting these aging effects for these materials.

VII.G.A-806 A new AMR item for cementitious coatings used as fireproofing/fire barriers is being added to the GALL-SLR Report because they are materials that are widely used throughout 2-17 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes industry and are likely to be cited in future SLRAs. The aging effects and aging mechanisms for cementitious coatings used as fireproofing/fire barriers exposed to air are based on the NRC staffs review and approval of applicants programs for aging management of fire protection materials listed in previous SLRAs. In addition, the aging effects and aging mechanisms are consistent with Section 5, Structural Concrete Members, and Section 6, Fire Barriers, of EPRI 3002013084, and those cited by industry as part of SLRA lessons learned activities and public comments on the draft AMR item.

This item manages loss of material due to abrasion, exfoliation, elevated temperature, flaking, and spalling; cracking/delamination; change in material properties; and separation for cementitious coatings (Pyrocrete, BIO' K-10 Mortar, Cafecote, and other similar materials) exposed to air.

VII.G.A-807 A new AMR item for silicates used as fireproofing/fire barriers is being added to the GALL-SLR Report because they are materials that are widely used throughout industry and are likely to be cited in future SLRAs. The aging effects and aging mechanisms for silicates used as fireproofing/fire barriers exposed to air are based on the NRC staffs review and approval of applicants programs for aging management of fire protection materials listed in previous SLRAs. In addition, the aging effects and aging mechanisms are consistent with Section 6 of EPRI Report 3002013084, and those cited by industry as part of SLRA lessons learned activities and public comments on the draft AMR item.

New AMR Item A-807 manages loss of material due to abrasion and flaking; cracking/delamination due to settlement; change in material properties due to gamma irradiation exposure; and separation for silicates (Marinite, Kaowool', Cerafiber, Cera blanket, or other similar materials) exposed to air.

The periodic inspections recommended by AMP XI.M26 are capable of detecting these aging effects for these materials.

VII.H2.A-799 Two new AMR items on hHeat exchanger tubes are added to VII.H2.A-800 reflect that the Fuel Oil Chemistry program is capable of mitigating reduction of heat transfer for heat exchanger tubes by periodic sampling of fuel oil for contaminants that may cause the reduction of heat transfer due to fouling. The Fuel Oil Chemistry program can manage contaminants that would promote corrosion (e.g.,

water or microbial activity), particulate concentration, or other contaminants that tested for under ASTM D975 that could contribute to heat exchanger tube fouling. If operating experience, or plant specific configurations, indicate other fouling mechanisms for a fuel oil environment may be present or the Fuel Oil Chemistry program alone is not sufficient to manage aging, the staff may need to evaluate whether the Fuel Oil Chemistry program is appropriate to manage these aging effects and if a One-Time Inspection is needed for a given plant.

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Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes VII.H2.A-801 The staff noted that the GALL-SLR Report recommends the use of VII.H2.A-802 the Fuel Oil Chemistry and One- Time Inspection AMPs to manage loss of material of several different materials that are exposed to a fuel oil environment. These new AMR items credit the Fuel Oil Chemistry program to minimize contaminants which could lead to loss of material, and the One- Time Inspection program to verify the effectiveness of the Fuel Oil Chemistry program. The use of the Fuel Oil Chemistry program can minimize contaminants regardless of the material of the affected component. Therefore, the staff has reasonable assurance that it will be effective in managing loss of material for nickel alloy strainer elements exposed to fuel oil.

VII.C3.A-482a The staff added aluminum alloyAlloy 6063T6 to the list of materials VII.C3.A-482b that are not susceptible to systems, structures, and VII.C3.A-482c componentsSCC. This is a wrought material alloyed primary with VII.E5.A-482a mMagnesium (Mg) and sSilicon (Si). It is a moderate strength VII.E5.A-482b precipitation hardened aluminum alloy in the peak-aged condition.

VII.E5.A-482c The strengthening phase precipitated during the artificial aging of VII.H1.A-482a 6063 is Mg2Si. Generally, 6xxx series alloys have satisfactory SCC VII.H1.A-482b resistance and inservice performance. However, some 6xxx series VII.H1.A-482c alloys are known to be susceptible to SCC when exposed to certain atypical processing histories. The majority of 6xxx series SCC testing and characterization has been performed on 6061T6, which is known to be resistant to SCC. Much more limited SCC testing and characterization has been performed on 6063T6; although, results have been consistent with those of 6061T6. Alloy 6063 is a compositionally leaner version of 6061 which has been optimized for extrusion. The two alloys have the same strengthening mechanism and their nominal Mg/Si ratios are also similar. Therefore, it is expected that the SCC performance is comparable. Additionally, the known inservice performance of aluminum alloyAlloy 6063T6 has shown satisfactory SCC resistance across multiple industries. Based on the metallurgical characteristics, available laboratory testing, and known service history, the staff has determined that 6063T6 is not susceptible to SCC.

VII.C1.A-795a Subsequent to issuance of the GALL-SLR Report, the staff VII.C2.A-795b recognized that to be consistent with item A-767, A-795a and A-VII.C3.A-795a 795b should have also cited reduction of heat transfer due to VII.E4.A-795a fouling. This is consistent with GALL Report Revision 2, item SP-VII.H2.A-795a 41 where a material (i.e., stainless steel [SS]) that is not susceptible to loss of material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

Titanium components are subject to flow blockage due to fouling due to potential debris in the raw water environment. Based on the staffs review of industry operating experience, it is possible that flow blockage due to fouling can occur in the closed cycle cooling water environment.

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Draft Document: Tracked Changes Version New Aging Management Review (AMR) Item No. Technical Bases for Changes VII.G.A-789 The sStructure and/or cComponent was changed from fire damper assemblies to fire damper housing because the housing is the passive component of the fire damper assembly that is subject to aging management. The applicable material was revised to metallic because fire damper housings are typically constructed of steel or stainless steelSS. The applicable aging effects were revised to account for loss of material due to general, pitting, and crevice corrosion, and cracking due to SCC because the elastomer aging effects of hardening, loss of strength, and shrinkage do not apply to metallic components. The fire damper housing is potentially subject to the cited aging effects. For example, steel materials would not be subject to SCC; however, stainless steelSS materials would be. The periodic inspections recommended by GALL-SLR Aging Management ProgramAMP Report XI.M26 are capable of detecting these aging effects.

VII.E2.A798 The staff modified the AMR Item V.A.E-434 and added AMR Item VII.E2.A798, to note that the aging effects of loss of material, and long-term loss of material due to general corrosion on steel exposed to an environment of treated water and sodium pentaborate can be managed by the Water Chemistry and One-Time Inspection AMPs. No item was added to manage stress corrosion cracking of steel in this environmental as the GALL-SLR already states that steel components typically are not susceptible to stress corrosion cracking and are mainly susceptible to loss of material.

The staff determined that this MEAP may be managed with the AMPs cited above because the Water Chemistry AMP can monitor and control the concentration of deleterious species in the water storage tanks that provide water to the Standby Liquid Control (SLC) system which contains the sodium pentaborate solution. Additionally, the One-Time Inspection AMP can verify the corrosion rate of the steel components is low enough that loss of material is unlikely to cause a loss of intended function.

Several reports were reviewed by the staff to make this determination. These reports concluded that even though the pH of the SLC system varies with temperature, it is generally greater than 6.8 pH which is close to neutral. Additionally, these reports noted that the pH range in SLC systems tends to be between 6.8-

- 8.5. This would result in less corrosion of the steel as the corrosion rate of steel tends to decrease with an increasing (i.e.,

more basic) pH (i.e., more basic) and would need additional impurities (e.g., salts, oxygen) for appreciable corrosion to occur in this environment (Metals Handbook Desk Edition, 2nd Edition).

Additionally, one report found that corrosion rates of carbon and low -alloy steel, when exposed to varying concentrations of boric acid, were relatively low (0.05- - 1.1 mm/year [(0.002- - 0.045 inches/year])), when the temperature was below 60 °C (140 °F)

(EPRI Report 1000975).

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Draft Document: Tracked Changes Version 1 Table 2-11 Table 2-7 New Aging Management ReviewAMR Items Added in GALL-SLR 2 Report Revision 1, Chapter VIIl, Steam and Power Conversion System New Aging Management Review (AMR) Item No. Technical Bases for Changes VIII.E.S-450a The staff added aluminum alloyAlloy 6063T6 to the list of materials VIII.E.S-450b that are not susceptible to SCC. This is a wrought material alloyed VIII.E.S-450c primary with mMagnesium (Mg) and sSilicon (Si). It is a moderate VIII.G.S-450a strength precipitation hardened aluminum alloy in the peak-aged VIII.G.S-450b condition. The strengthening phase precipitated during the artificial VIII.G.S-450c aging of 6063 is Mg2Si. Generally, 6xxx series alloys have satisfactory stress corrosion cracking (SCC) resistance and inservice performance. However, some 6xxx series alloys are known to be susceptible to SCC when exposed to certain atypical processing histories. The majority of 6xxx series SCC testing and characterization has been performed on 6061T6, which is known to be resistant to SCC. Much more limited SCC testing and characterization has been performed on 6063T6; although, results have been consistent with those of 6061T6. Alloy 6063 is a compositionally leaner version of 6061 which has been optimized for extrusion. The two alloys have the same strengthening mechanism and their nominal Mg/Si ratios are also similar.

Therefore, it is expected that the SCC performance is comparable.

Additionally, the known inservice performance of aluminum alloyAlloy 6063T6 has shown satisfactory SCC resistance across multiple industries. Based on the metallurgical characteristics, available laboratory testing, and known service history, the staff has determined that 6063T6 is not susceptible to SCC.

VIII.D1.S-482 Subsequent to issuance of the GALL-SLR Report, the staff VIII.D2.S-482 recognized that to be consistent with other GALL-SLR Report VIII.E.S-482 items associated with heat exchanger tubes, S-482 should have VIII.F.S-482 also cited reduction of heat transfer due to fouling. This is consistent with GALL Report, Revision 2, Iitem SP-41 where a material (i.e., stainless steelSS) that is not susceptible to loss of material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

Titanium components are subject to flow blockage due to fouling due to potential debris in the raw water environment.

VIII.H.S-484 The staff added a new AMR item to add carbon fiber reinforced polymer (CFRP) repaired piping, crediting the new AMP XI.M43, High Density Polyethylene (HDPE) Piping and Carbon Fiber Reinforced Polymer (CFRP) Repaired Piping. The new AMR item reflects the recent introduction and increasing use of CFRP repaired piping at reactor facilities. The unique aging issues and aging management approaches for CFRP repaired piping were considered to be most effectively addressed with a dedicated AMP.

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Draft Document: Tracked Changes Version 1 Table 2-12 Table 2-8 Deleted Aging Management ReviewAMR Items From GALL-SLR 2 Revision 0, Chapter Il, Containment Structures Aging Management ReviewAMR Item No. Technical Bases for Changes No aging management reviewAMR items were deleted from Chapter ll GALL-SLR Report, Revision 0.

3 Table 2-13 Table 2-9 Deleted Aging Management ReviewAMR Items From GALL-SLR 4 Revision 0, Chapter IIl, Structures and Component Supports Aging Management Review (AMR) Item No. Technical Bases for Changes III.A6.TP-25 This item was deleted from NUREG-2191, Volume 1, since it is a duplicate to GALL-SLR Iitem no. III.A6.T-34. In addition, this GALL-SLR item is associated with Group 6 structures and the associated SRP-SLR AMR item (i.e., 3.5-1, 054) is only intended to address all groups of structures, except Group 6.

5 Table 2-14 Table 2-10 Deleted Aging Management ReviewAMR Items From GALL-SLR 6 Revision 0, Chapter IV, Reactor Vessel, Internals, and Reactor Coolant 7 System Aging Management Review (AMR) Item No. Technical Bases for Changes IV.B2.RP-356 The staff deleted GALL-SLR Item IV.B2.RP-356 in SLR-ISG-2021-01-PWRVI. The previous version of the RP-356 item in Table IV.B2 of GALL-SLR Rreport was included to address loss of material due to wear in stainless steel or nickel alloy Westinghouse-design control rod guide tube (CRGT) support pins (split pins) that are exposed to a reactor coolant and neutron flux environment.

For the Interim Staff Guidance (ISG) update, the staff modified the corresponding cracking item for the split pins in GALL-SLR Item IV.B2.RP-355 to limit the scope of line item only to CRGT spilt pins that are made from nickel alloy (X-750) materials and to include loss of material due wear as an additional aging effect and mechanism combination for the RP-355 item (i.e., in addition to cracking due to SCC or fatigue), where the Aging Management Program (AMP) XI.M16A aging management basis for the pins would be based on a component-specific evaluation per the MRP-227, Rev.Revision 1-A guidelines. For spilt pins made from Type 316 or Type 316L stainless steel (SS) materials, the MRP-227, Rev.Revision 1-A guidelines placed the components in the No Additional Measures category.

Since loss of material for CRGT spilt pins made from X-750 nickel alloy materials is now addressed by the modification of GALL-SLR Item IV.B2.RP-355 in the ISG and since CRGT spilt pins made from Type 316 or Type 316L SS materials are now within the scope of the AMR for No Additional Measures components in 2-22 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes GALL-SLR Item IV.B2.RP-265, GALL-SLR Item IV.2.RP-356 is no longer needed and has been deleted in SLR-ISG-2021-01-PWRVI.

IV.B2.RP-278 The staff deleted GALL-SLR Items IV.B2.RP-278 and IV.B2.RP-IV.B2.RP-278a 278a in SLR-ISG-2021-01-PWRVI. The previous versions of the RP-278 and RP-278a items in Table IV.B2 of the GALL-SLR Rreport were included to address management of cracking due to SCC or fatigue and loss of fracture toughness due to neutron irradiation embrittlement in Westinghouse-design core barrel outlet nozzle welds (ONWs).

In EPRIs MRP-227-A (Rev.Revision 0) report, the EPRI MRP designated that the core barrel ONWs were Expansion components for Westinghouse-design reactor vessel internal (RVI) management programs, where the need for inspecting the ONWs would be dependent on the results of primary inspections performed on the upper flange weld (UFW) in the core barrel assembly, as designated for inspection per Item W3 of Table 4-3 in MRP-227, Rev.Revision 1-A. However, in Items W3.1, W3.2, W3.3 and W3.4 of Table 4-6 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP replaced the ONWs with the core barrel assembly upper girth weld (UGW),

upper axial welds (UAWs), lower flange weld (LFW), and lower support forging or casting as the new Expansion components linked to the Primary core barrel UFW inspections.

Since the core barrel assembly ONWs are now within the scope of the staffs AMR line item for Westinghouse-design No Additional Measures category components (See GALL-SLR Item IV.B2.RP-265), the GALL-SLR Items IV.B2.RP-278 and IV.B2.RP-278a are no longer necessary and have been deleted in SLR-ISG-2021 PWRVI.

IV.B2.RP-382 The staff deleted GALL-SLR Items IV.B2.RP-382, IV.B3.RP-382 IV.B3.RP-382 and IV.B4.RP-382 in SLR-ISG-2021-01-PWRVI.

IV.B4.RP-382 The previous RP-382 items in the GALL-SLR Rreport were the AMR line items that could be used for RVI component aging management if the applicants GALL-SLR AMP XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD program was credited for aging management of the component(s).

However, the staff determined that the RP-382 AMR line items are redundant with staffs modification of GALL-SLR Item IV.E.R-444 in Appendix B.4 of the ISG.

Similarly, Item 032 in NUREG-2192 (SRP-SLR Report) Table 3.1-1 has also been deleted in accordance with these line item changes (Refer to the line item entry for Item 032 in Table 3-3a in this report). The modified version of SRP-SLR Table 3.1-1, Item 114 in Appendix A of the ISG is the SRP-SLR item that references the staffs modified version GALL-SLR Item IV.E.R-444 for pressurized water reactor (PWR) reactor internal components (RVI) that are defined as ASME Section XI Class 1 interior attachments to the reactor vessel (RV) or as ASME Section XI 2-23 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Class 1 core support structure components. Therefore, GALL-SLR Items IV.B2.RP-382, IV.B3.RP-382, and IV.B4.RP-382 have been deleted in SLR-ISG-2021-01-PWRVI.

IV.B3.RP-326a The staff deleted GALL-SLR Item IV.B3.RP-326a in SLR-ISG-2021-01-PWRVI. The previous version of the RP-326a item in Table IV.B3 of the GALL-SLR Rreport addressed cracking (due to SCC or fatigue) in the core shrouds of CE-designed PWRs. The RP-326a item applied to those CE PWR designs where the core shroud in the plant design is fabricated from two welded vertical shroud sections.

Based on the staffs partial acceptance of NEI Comment #3 on draft SLR-ISG-PWRVI-2020-XX, and the generic request in the comment, the staff confirmed that EPRI did not screen in any cracking mechanisms (i.e., SCC, irradiation-assisted stress corrosion cracking [IASCC], fatigue or overload) for these types of welded CE core shroud assemblies in Tables 3-2, 4-2 or 5-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. The assemblies did screen in for IE and void swelling (VS) in Item C4a of Table 4-2 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, which are covered by the RP-326 item on the previous page. So consistent with the staffs basis for partially accepting NEI Comment #3, the staff confirmed that GALL-SLR Item IV.B2.RP-326a item is no longer needed for the final version of the ISG and has been deleted in Appendix B.2 of SLR-ISG-2021-01-PWRVI.

IV.B3.RP-400 The staff deleted GALL-SLR Item IV.B3.RP-400 in SLR-ISG-2021-01-PWRVI. The prior version of the IV.B3.RP-400 item in Table IV.B3 of GALL-SLR Rreport was included to address cracking and loss of material due to wear in CE-design thermal shield positioning pins.

In the staffs review of EPRI Report MRP-227, Rev.Revision 1, the staff agreed that CE-design thermal shield positioning pins could be placed in the No Additional Measures category for CE-design RVI management programs. These components are now No Additional Measures category components per MRP-227, Rev.Revision 1-A criteria and are now covered by the line item for CE-design No Additional Measures components, as given in GALL-SLR Item IV.B3.RP-306. Therefore, GALL-SLR Item IV.B3.RP-400 is no longer necessary and has been deleted in Appendix B.2 of SLR-ISG-2021-01-PWRVI.

IV.B3.RP-334a The staff deleted GALL-SLR Item IV.B3.RP-334a in SLR-ISG-2021-01-PWRVI based on the staffs decision to fold the CE plant design applicability statement of the RP-334a item into staffs modification of GALL-SLR Item IV.B3.RP-336 in Appendix B.2 of the ISG. The previous version of the IV.B3.RP-334a item applied to management of loss of material due to wear, loss of fracture toughness and loss of preload in fuel alignment pins of CE-designed PWRs that have welded shrouds fabricated from two vertical shroud sections. The existing GALL-SLR Item IV.B4.RP-2-24 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes 336 item is the corresponding item for the fuel alignment pins in welded core shrouds that use full height shroud plates.

The staffs modification of the IV.B3/RP-336 item in SLR-ISG-2021-01-PWRVI allows the RP-336 item to be applied for the management of non-cracking effects in the fuel alignment pins of CE-design plants with welded core shrouds assembled from either two vertical sections or from full height shroud plates. Since the GALL-SLR Item IV.B4.RP-334a is no longer necessary, the staff deleted the GALL-SLR Item IV.B4.RP-334a item in Appendix B.2 of SLR-ISG-2021-01-PWRVI.

IV.B4.RP-400 The staff deleted GALL-SLR Items IV.B4.RP-400 and IV.B4.RP-IV.B4.RP-401 401 in SLR-ISG-2021-01-PWRVI. The previous versions of GALL-SLR Items IV.B4.RP-400 and IV.B4.RP-401 were included in Table IV.B4 of the GALL-SLR Rreport to address on aging management of cracking and loss of fracture toughness in the upper (top) flange welds that are located in the core shield assemblies of Babcock and Wilcox (B&W)-designed reactors.

The staff confirmed that the core support shield top flange welds are no longer designated as B&W Primary category or Expansion category components in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. Thus, the staff deleted the GALL-SLR IV.B4.RP-400 and IV.B4.RP-401 line items in order to be consistent with the revised program in the MRP-227, Revision 1-A reportRevision 1-A Report and with Section 3.6.4 (Page 45) of the staffs safety evaluation SE for the MRP-227, Revision 1-A reportRevision 1-A Report. Instead management of loss of material due to wear and loss of preload in the core support shield (CSS) top flange connections is covered by the existing item in GALL-SLR Iitem IV.B4.RP-251, and by Item B1.d in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, which identifies that the CSS top flanges are Primary category components for B&W-design RVI management programs.

IV.B4.RP-254 The staff deleted GALL-SLR Items IV.B4.RP-254, IV.B4.RP-254a, IV.B4.RP-254a and IV.B4.RP-254b in SLR-ISG-2021-01-PWRVI. The previous IV.B4.RP-254b versions of these line items in Table IV.B4 of the GALL-SLR Rreport only applied to specific types of internals components in the lower grid assembly of the Three Mile Island Unit 1 (TMI-1) facility. However, the licensee for TMI- Unit 1 has made an owner decision to decommission the plant. Since the GALL-SLR IV.B4.RP-254, IV.B4.RP-254a, and IV.B4.RP-254b line items are no longer necessary, the staff deleted them in Appendix B.3 of the ISG.

IV.B4.RP-249a The staff deleted GALL-SLR Items IV.B4.RP-249a, IV.B4.RP-IV.B4.RP-244a 244a, IV.B4.RP-250a, IV.B4.RP-252a, IV.B4.RP-258a and IV.B4.RP-250a IV.B4.RP-259a in SLR-ISG-2021-01-PWRVI.

IV.B4.RP-252a IV.B4.RP-258a The prior versions of these GALL-SLR Items applied to cracking in IV.B4.RP-259a B&W-design baffle plates, baffle-to-former bolt/core barrel core barrel-former bolt locking devices, core barrel assemblies and 2-25 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes welds, vent valve top and bottom retaining rings and locking devices, incore monitoring instrument (IMI) guide tube spiders, and IMI guide tube spiders-to-lower grid rib sections welds. In MRP-227, Rev.Revision 1-A, In MRP-227, Rev.Revision 1-A, the Electric Power Research Institute (EPRI) MRP identified that these components did not screen in for any cracking mechanisms (e.g.,

SCC, IASCC, fatigue or component overload). Therefore, consistent comments received by the EPRI MRP or Framatome on this matter, the staff deleted these AMR items in Appendix B.3 of the ISG.

1 Table 2-15 Table 2-11 Deleted Aging Management ReviewAMR Items From GALL-SLR 2 Revision 0, Chapter V, Engineered Safety Features Aging Management ReviewAMR Item No. Technical Bases for Changes No aging management reviewAMR items were deleted from Chapter V GALL-SLR Report, Revision 0.

3 Table 2-16 Table 2-12 Deleted Aging Management ReviewAMR Items From GALL-SLR 4 Revision 0, Chapter VI, Electrical Components Aging Management ReviewAMR Item No. Technical Bases for Changes No aging management reviewAMR items were deleted from Chapter VI From GALL-SLR Report, Revision 0.

5 Table 2-17 Table 2-13 Deleted AMR Items, Chapter VII, Auxiliary Systems Aging Management ReviewAMR Item No. Technical Bases for Changes No aging management reviewAMR items were deleted from Chapter VII of GALL-SLR Report, Revision 0.

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Draft Document: Tracked Changes Version 1 Table 2-18 Table 2-14 Deleted Aging Management ReviewAMR Items From GALL-SLR 2 Revision 0, Chapter VIII, Steam and Power Conversion System Aging Management ReviewAMR Item No. Technical Bases for Changes No aging management reviewAMR items were deleted from Chapter VIII of GALL-SLR Report, Revision 0 3 Table 2-19 Table 2-15 Changes to GALL-SLR Report, Revision 0, Chapter II Aging 4 Management ReviewAMR Items and Technical Bases Aging Management ReviewAMR Item No. Technical Bases for Changes II.A1.CP-147 Modifications to the GALL-SLR Report aging management II.A1.CP-67 review (AMR) items and Standard Review Plan for Review of II.A1.CP-102 Subsequent License Renewal Applications for Nuclear Power II.A1.CP-34 Plants (SRP-SLR) Table 3.5-1 line items with associated further II.A2.CP-70 evaluations provide the option to use plant-specific II.A2.CP-104 II.A2.CP-53 enhancements to GALL-SLR Report AMP XI.S2, ASME Section II.A3.CP-37 XI, Subsection IWL, and/or GALL-SLR Report AMP XI.S6, II.B1.1.CP-49 Structures Monitoring, in lieu of a plant-specific AMP. The II.B1.2.CP-99 option to use plant-specific enhancements increases the II.B1.2.CP-110 efficiency of subsequent license renewal application reviews by II.B1.2.CP-57 limiting the use of AMR Note E designations for plant-specific II.B2.1.CP-107 aging management activities when aging effects are managed II.B2.1.CP-142 through a plant-specific AMP.

II.B2.2.CP-99 II.B2.2.CP-110 II.B2.2.CP-57 II.B2.2.CP-64 II.B3.1.CP-53 II.B3.1.CP-83 II.B3.1.CP-65 II.B3.2.CP-135 II.B3.2.CP-121 II.B3.2.CP-122 II.B3.2.CP-108 II.B4.CP-37 5 Table 2-20 Table 2-16 Changes to GALL-SLR Report, Revision 0, Chapter III Aging 6 Management ReviewAMR Items and Technical Bases Aging Management ReviewAMR Item No. Technical Bases for Changes III.A1.TP-204 Modifications to the GALL-SLR Report aging management review III.A1.TP-67 (AMR) items and Standard Review Plan for Review of Subsequent III.A1.TP-108 License Renewal Applications for Nuclear Power Plants (SRP-III.A1.TP-114 III.A2.TP-204 SLR) Modifications to GALL-SLR Report AMR items and SRP-SLR III.A2.TP-67 Table 3.5-1 line items with associated further evaluations provide III.A2.TP-108 the option to use, in lieu of a plant-specific AMP, plant-specific III.A2.TP-114 enhancements to GALL-SLR Report AMP XI.S6, Structures III.A3.TP-204 Monitoring, or other selected AMPs. The option to use plant-2-27 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version III.A3.TP-67 specific enhancements increases the efficiency of subsequent III.A3.TP-108 license renewal applications reviews by limiting the use of AMR III.A3.TP-114 Note E to plant-specific aging management activities when aging III.A4.TP-204 III.A4.TP-305 effects are managed through a plant-specific AMP.

III.A4.TP-114 III.A4.T-35 III.A5.TP-204 III.A5.TP-67 III.A5.TP-108 III.A5.TP-114 III.A6.TP-220 III.A6.TP-110 III.A6.TP-109 III.A7.TP-204 III.A7.TP-67 III.A7.TP-108 III.A8.TP-204 III.A8.TP-67 III.A8.TP-108 III.A9.TP-204 III.A9.TP-67 III.A9.TP-108 No AMR items were changed from Chapter III of GALL-SLR Report, Revision 0.

1 Table 2-21 Table 2-17 Changes to GALL-SLR Report, Revision 0, Chapter IV Aging 2 Management ReviewAMR Items and Technical Bases Aging Management Review (AMR) Item No. Technical Bases for Changes IV.B1.R-95 To reflect the deletion of Standard Review Plan for Review of IV.B1.R-94 Subsequent License Renewal Applications for Nuclear Power IV.B1.R-92 Plants (SRP-SLR) Section 3.1.2.2.12.

IV.B1.R-96 IV.B1.R-93 IV.B1.R-97 IV.B1.R-99 IV.B1.R-105 IV.B1.R-100 IV.B1.R-422 IV.B1.R-98 IV.B1.RP-182 To reflect the deletion of SRP-SLR Section 3.1.2.2.13.

IV.B1.RP-200 IV.B1.RP-219 IV.B1.RP-220 IV.B1.R-416 IV.B1.R-417 IV.B1.R-419 IV.B2.RP-301 The upper core plate alignment pins that are the subject of GALL-SLR Item IV.B2.RP-301 remain as Existing Program components for Westinghouse-design reactor vessel internal (RVI) management programs per Item W15 in Table 4-9 of the MRP-2-28 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes 227, Rev. 1-A reportMRP-227, Revision 1-A Report. The RP-301 item applies to management of cracking in the alignment pins.

In Item W15 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the Electric Power Research Institute (EPRI) MRP only screened the upper core plate alignment pins in for stress corrosion cracking (SCC) as the applicable cracking mechanism.

However, based on lessons learned from the staffs processing of the Surry Subsequent License Renewal Application (SLRA) RVI gap analysis results, the staff confirmed that the applicant included fatigue as additional cracking mechanism for the core plate alignment pins (i.e., in addition to SCC) by referencing the assessment for the core plate alignment pins in Electric Power Research Institute's (EPRIs) MRP-2018-022 report. The administrative edits of the IV.B2.RP-301 item and the cracking mechanisms cited for the RP-301 item are consistent with the cracking mechanisms cited for the fuel alignment pins in the Surry SLRA.

IV.B2.RP-271 The baffle-to-former bolts (which are the topic of the RP-271 item IV.B2.RP-272 for cracking effect and mechanism combinations and the RP-272 item for non-cracking effect and mechanism combinations) remain as leading Primary category components for Westinghouse-design RVI management programs per Item W6 in Table 4-3 of the MRP-227, Rev.Revision 1-A report.

Based on lessons learned from the staffs review of the RVI gap analysis in the Surry SLRA, the staff confirmed that the applicant screened the baffle-to-former bolts in for irradiation stress corrosion cracking (irradiation-assisted stress corrosion cracking

[SCC] or irradiation-assisted stress corrosion cracking [IASCC]),

fatigue, wear, neutron irradiation embrittlement (IE), void swelling (VS), and irradiation-enhanced stress relaxation or creep (ISR/IC) aging mechanisms, as cited consistent with the EPRI MRPs 80-year Expert Panel assessment for the baffle-to-former bolts in EPRIs MRP-2018-022 report. Loss of material due to wear was added to the RP-272 item based on the information in the Surry SLRA. The edited version of the RP-271 item includes cracking due to IASCC and fatigue. The edited version of the RP-272 item appropriately includes the IE, VS/distortion, ISR/IC and wear aging mechanisms.

The MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report includes and adequately addresses accessibility criteria for specified Primary, Expansion or Existing Program components. Therefore, from a generic perspective, there is no need for the staff to reference or address accessibility criteria in any of the AMR line items for pressurized water reactor (PWR) reactor vessel internal (RVI) components in the SRP-SLR or GALL-SLR Rreports, including the staffs updates of GALL-SLR Items IV.B2.RP-271 and IV.B2.RP-272 in SLR-ISG-2021 PWRVI.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The change to include the parenthetical clause (includes corner bolts) in the updates of the Structure and/or Component column entries of the GALL-SLR Items IV.B2.RP-271 and IV.B2.RP-272 line items is based on the staffs receipt and acceptance of EPRI MPR Comment #4 on draft SLR-ISG-PWRVI-2020-XX, in which EPRI had clarified that the corner bolts are a subset of the baffle-to-former bolts. The change in the RP-271 item to cite irradiation-assisted SCC as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B2.RP-270 The baffle and former plates (which are the topic of the RP-270a IV.B2.RP-270a item for cracking effect and mechanism combinations and the RP-270 item for non-cracking effect and mechanism combinations) remain as leading Primary category components for Westinghouse-design RVI management programs per Item W7 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The staff acknowledges that in Item W7 of Table 4-3 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP only screened the baffle and former plates in for irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC) and distortion (void swelling [VS]) as applicable mechanisms for the plates. However, based on lessons learned from the staffs review of the RVI gap analysis in the Surry SLRA, the staff confirmed that the applicant screened the baffle and former plates in for IASCC, fatigue, neutron irradiation embrittlementIE and VS aging mechanisms, as assessed with the EPRI MRPs 80-year Expert Panel assessment of the plates in EPRIs MRP-2018-022 report. Thus, the staff revised GALL-SLR Item IV.B2.RP-270 in the Interim Staff Guidance (ISG) to reference the applicable non-cracking effect and mechanism combinations cited for the plates in the Surry SLRA; similarly, the staff revised GALL-SLR Item IV.B2.RP-270a in the ISG to reference the cracking mechanisms cited for the plates in the Surry SLRA.

The change in the RP-270a item to cite irradiation-assisted SCC as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B2.RP-275 The baffle -edge bolts (which are the topic of modified versions of IV.B2.RP-354 the RP-275 item for cracking effect and mechanism combinations and the RP-354 item for non-cracking effect and mechanism combinations) remain as leading Primary category components for Westinghouse-design RVI management programs per Item W7 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. The staff acknowledges that, in EPRI MRP Comment

1. 4 for these line items, EPRI commented that: The modified text deleted all plants with baffleedge bolts and replaced it with corner bolts. This is not correct. Corner bolts are a subset of baffleformer bolts, not baffleedge bolts. Note that bBracket bolts are a subset of baffleedge bolts.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The staff acknowledges that in Item W7 of Table 4-3 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP only screened the baffle edge bolts in for irradiation-assisted stress corrosion cracking (IASCC) and distortion (void swelling

[VS]) as applicable mechanisms for the bolts. However, based on lessons learned from the staffs review of the RVI gap analysis in the Surry SLRA, the staff confirmed that the applicant screened the baffle edge bolts in for IASCC, fatigue, wear, neutron irradiation embrittlement (IE), VS, and irradiated-enhanced stress relaxation or creep (ISR/IC) mechanisms, as referenced to the EPRI MRPs 80-year Expert Panel assessment of the baffle edge bolts in EPRIs MRP-2018-022 report. Thus, the staff revised GALL-SLR Item IV.B2.RP-354 in the ISG to reference the applicable non-cracking effect and mechanism combinations cited for the baffle edge bolts in the Surry SLRA; similarly, the staff revised GALL-SLR Item IV.B2.RP-275 in the ISG to reference the cracking mechanisms cited for the baffle edge bolts in the Surry SLRA.

The staff also partially accepted EPRIs perspective in Comment

1. 4 and agreed that the component descriptions for the RP-275 and RP-354 line items in the draft SLR-ISG-PWRVI-2020-XX should not have been adjusted to include reference of corner bolts, as previously designated by a change of the component-specific parenthetical explanation in the line items (corner bolts).

To resolve the EPRI comment, the staff administratively edited the GALL-SLR Items IV.B2.RP-275 and IV.B2.RP-354 line items by removing the parenthetical clause (all plants with baffle edge bolts) from the scope of the line items. Therefore, for the final ISG, the component descriptions in the RP-275 and RP-354 items now state: Baffle-to-former assembly: baffle edge bolts..

IV.B2.RP-273 The barrel-to-former bolts (which are the topic of the RP-273 item IV.B2.RP-274 for cracking effect and mechanism combinations and the RP-274 item for non-cracking effect and mechanism combinations) remain as Expansion category components for Westinghouse-design RVI management programs per Item W6.1 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The staff acknowledges that, in Item W6.1 of Table 4-6 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP only screened the barrel-to-former bolts in for irradiation-assisted stress corrosion cracking (IASCC), fatigue, neutron irradiation embrittlement (IE), void swelling (VS), and irradiation-enhanced stress relaxation or creep (ISR/IC) aging mechanisms. However, based on lessons learned from the staffs review of the RVI gap analysis in the Surry SLRA, the staff confirmed that the applicant screened the barrel-to-former bolts in for wear as an additional aging mechanism for the bolts (i.e., in addition to IASCC, fatigue, IE, VS, and ISR/IC), as referenced to the EPRI MRPs 80-year Expert Panel assessment of the barrel-to-former bolts in EPRIs MRP-2018-022 report. Thus, the staff 2-31 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes revised GALL-SLR Item IV.B2.RP-274 in the ISG to reference the applicable non-cracking effect and mechanism combinations cited (including loss of material due to wear) for the baffle edge bolts in the Surry SLRA; similarly, the staff revised GALL-SLR Item IV.B2.RP-273 in the ISG to reference the cracking mechanisms (i.e., IASCC and fatigue) cited for the barrel-to-former bolts in the Surry SLRA.

The change in the RP-273 item to cite irradiation-assisted SCC as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B2.RP-292 The bottom mounted instrumentation (BMI) column bodies (which IV.B2.RP-293 are is the topic of the RP-293 Iitem for cracking effect and mechanism combinations and the RP-292 Iitem for non-cracking effect and mechanism combinations) remains as the Expansion category components for Westinghouse-design RVI management programs per Item W2.2 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The staff acknowledges that, in Item W2.2 of the MRP-227, Rev.

1-A reportMRP-227, Revision 1-A Report, EPRI screened the BMI column bodies only in for fatigue and neutron irradiation embrittlement (IE) aging mechanisms. However, based on lessons learned from the staffs review of the RVI gap analysis in the Surry SLRA, the staff confirmed that the applicant screened wear in as an additional aging mechanism for the BMI column bodies, as referenced to the EPRI MRPs 80-year Expert Panel assessment of the BMI column bodies in EPRIs MRP-2018-022 report. Thus, for the final ISG, the staff edited the GALL-SLR IV.B2.RP-293 to cite SCC and fatigue as the applicable cracking mechanisms for the BMI column bodies and modified the IV.B2.RP-292 to cite loss or material due to wear as an additional non-cracking effect and mechanism combination for the BMI column bodies (i.e., in addition to loss of fracture toughness due to neutron irradiation embrittlement).

Based on the staffs response to and acceptance of EPRI MRP Comments #5 and #6 in Appendix H of SLR-ISG-2021-01-PWRVI, the staff confirmed that IASCC did not need to be included as a cited cracking mechanism for GALL-SLR Item IV.B2.RP-293 and that changes in dimension due to void swelling or distortion did not need to be included as a cited non-cracking effect and mechanism combination for GALL-SLR Item IV.B2.RP-292.

IV.B2.RP-296 The guide plates (guide cards) in the control rod guide tube (CRGT) assemblies of Westinghouse-designed PWRs (which are the topic of the amended version of Item IV.B2.RP-296 for cited non-cracking effect and mechanism combinations) remain as leading Primary category components for Westinghouse-design RVI management programs per Item W1 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The staff acknowledges that, in Item W1 of Table 4-3 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the CRGT guide cards in only for the aging mechanism of wear, as based on relevant operating experience (OE) with wear occurring in the components. However, based on lessons learned from the staffs review of the RVI gap analysis in the Surry SLRA, the staff confirmed that the applicant identified that the CRGT guide cards could also be susceptible to the aging mechanism of thermal embrittlement (TE) if the guide cards were fabricated from cast austenitic stainless steel materials (e.g., CF8 cast austenitic stainless steel [CASS] materials) materials. Thus, for the final ISG, the revised the GALL-SLR IV.B2.RP-296 items were made to account for the lessons learned taken from the staffs past processing of the Surry SLRA, as evaluated in the staffs final safety evaluation report for the application (ADAMS Accession No. ML20052F523, dated March 9, 2020).

IV.B2.RP-297 The lower flange welds (LFWs) in the peripheral CRGT IV.B2.RP-298 assemblies of Westinghouse-designed PWRs (which are the topic of the amended RP-298 item for cited non-cracking effect and mechanism combinations and the RP-297 item for cited non-cracking effect and mechanism combinations) remain as leading Primary category components for Westinghouse-design RVI management programs per Item W2 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. The LFWs in the remaining (non-peripheral) CRGT assemblies (which are the topics of the new GALL-SLR IV.B2.RP-297a and IV.B2.RP-298a Items) were identified as Expansion category components per Item W2.1 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

Based on the staffs review of the RVI gap analysis for RVI components in the Surry SLRA, the staff confirmed that the applicant screened the CRGT LFWs in for SCC, IASCC, fatigue, and irradiation embrittle mechanisms, and additionally for thermal aging embrittlement if the components were made from cast austenitic stainless steel materials. Therefore, based on the EPRI MRP criteria for peripheral and non-peripheral CRGT LFWs in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report and the lessons learned from the staffs review of the Surry SLRA, the staff modified the GALL-SLR IV.B2.RP-297 and IV.B2.RP-298 items in the ISG to limit the scope of the line items only to those CRGT LFWs in the peripheral (outer) CRGT assemblies (as assigned as the appropriate Primary category components), with the updated of the RP-298 item citing the applicable SCC, IASCC, and fatigue cracking mechanisms, and the updated of the RP-297 appropriately citing the applicable loss of fracture toughness irradiation embrittlementIE mechanism, with thermal embrittlementTE also being applicable if the CRGT LFWs are made from CASS materials.

Additionally, the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report adequately addresses accessibility of PWR RVI 2-33 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes components. Therefore, from a generic perspective, there is no need for the staff to reference or address accessibility criteria in any of the AMR line items for PWR RVI components in the SRP-SLR or GALL-SLR Rreports, including GALL-SLR Items IV.B2.RP-297 and IV.B2.RP-298.

IV.B2.RP-355 The prior version of the RP-355 item in the GALL-SLR Rreport addressed cracking in the CRGT spilt pins independent of whether the spilt pins were made from stainless steel or nickel alloy materials; similarly, the prior version of the RP-356 item addressed loss of material in Westinghouse-design CRGT split pins independent of whether the pins were made from stainless steel or nickel alloy materials.

In the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP adjusted its aging management criteria for CRGT spilt pins to require aging management as Existing Program category components only if the pins were made from nickel alloy (X-750) materials, where the pins had not be replaced with pins made from either Type 316 or 316L austenitic stainless steelSS materials and where aging management of the nickel alloy pins would need to be based on a component-specific evaluation of the pins. For replacement pins made from stainless steel 316 or 316L materials, EPRI placed the pins in the No Additional Measures category of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report (Refer to GALL-SLR Item IV.B2.RP-265, as referenced in this SLR-ISG).

The staff has amended SRP-SLR Table 3.1-1, Item 028 and the GALL-SLR RP-355 and RP-356 items to be consistent with the updated basis in MRP-227, Rev.Revision 1-A for the components.

This required: (1) restricting the scope of the RP-355 item only to CRGT split pins made from of nickel alloy (X-750) materials, (2) adding loss of material due to wear to the RP-355 item, and (3) clarifying that the AMP XI.16A, PWR Vessel Internals basis would be based on a plant-specific evaluation of the pins. Since CRGT spilt pins made from Type 316 or Type 316L SS materials are now designated as EPRI MRP No Additional Measures components, the RP-356 item is no longer necessary for the scope of GALL-SLR AMR line items specified in SLR-ISG-2021-01-PWRVI. Instead, license renewal (LR) or subsequent license renewal (SLR) applicants of Westinghouse-designed PWRs may now use GALL-SLR Item IV.B2.RP-265 (which is the generic GALL-SLR AMR item for Westinghouse-design No Additional Measures components if the CRGT spilt pins are made from either Type 316 or 316L stainless steel material).

IV.B2.RP-345 The core barrel flanges in the core barrel assemblies of Westinghouse-designed PWRs (which are the topic of the amended version of Item IV.B2.RP-345 for managing loss of material due to wear in the flanges) remain as Existing Program category components for Westinghouse-design RVI management programs per Item W10 in Table 4-9 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Water Cchemistry programs are not designed to monitor for loss of material that may be induced by a physical degradation mechanism, such as wear, erosion, or abrasion. Thus, the staff deleted reference of GALL-SLR AMP XI.M2, Water Chemistry, from the Aging Management Program (AMP)/TLAA column entry in Item IV.B2.RP-345.

IV.B2.RP-280 The upper girth welds (UGWs), upper axial welds (UAWs), and lower flange welds (LFWs) in the core barrel assemblies of Westinghouse-designed PWRs (which are the topic of Item IV.B2.RP-280) are designated as Expansion category components for Westinghouse-design RVI management programs per Items W3.1, W3.2, and W3.3 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. Item IV.B2.RP-280 now addresses cracking in the core barrel assembly UGWs, UAWs, and LFWs. The inspections of these core barrel Expansion category welds are linked to the Primary inspections of the core barrel upper flange weld (UFW) per Item W3 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The staff modified the RP-280 Iitem in order to: (1) keep all of the core barrel Expansion category welds linked to EPRI MRP Item W3 in a singular AMR line item that addresses cracking of the components, including the core barrel UGW, UAWs, and LFW, and (2) reflect the change in the inspection category for the components from Primary category (as previously indicated by the SRP Table 3.1-1, Item 053a reference in the line item) to Expansion category (as now indicated by the SRP-SLR Table 3.1-1, 053b reference in the line item).

The staff also used lessons learned from the staffs processing of the Surry SLRA RVI gap analysis results for the cracking mechanisms cited in the RP-280 line item. Specifically, in the RVI gap analysis of the Surry SLRA, the applicant cited EPRIs MRP-2018-022 Eexpert Ppanel process for these core barrel Expansion category welds and screened the core barrel assembly LFWs for stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), and fatigue cracking mechanisms and the core barrel assembly UAWs and UGWs in for SCC and fatigue cracking mechanisms. Thus, the staff also adjusted the GALL-SLR IV.B2.RP-280 item to cite SCC, IASCC, and fatigue as the applicable cracking mechanisms for the final version of the line item in the ISG. Based on the past review of the Surry SLRA RVI gap analysis results, the staff did not rely on EPRIs Comment #11 rationale that IASCC should not be cited as a referenced cracking mechanism for the core barrel LFWs.

IV.B2.RP-387 The core barrel assembly lower girth welds (LGWs) (which are the IV.B2.RP-388 topic of the RP-387 item for cracking effects and the RP-388 item for non-cracking effects) remain as Primary category components for Westinghouse-design RVI management programs 2-35 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes per Item W4 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. Item W4 in Table 4-3 does not include core barrel UGWs as components for the line item. Therefore, the staff removed the core barrel UGWs from the scope of the IV.B2.RP-387 and IV.B2.RP-388 line items.

The staff also used lessons learned from the staffs processing of the Surry SLRA RVI gap analysis results for the cracking mechanisms cited in the RP-387 line iItem and the non-cracking effect and mechanisms cited in the RP-388 Iitem. Specifically, in the Surry SLRA, the applicant screened the core barrel assembly LGWs in for stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC),

fatigue, neutron irradiation embrittlement (IE), and void swelling aging mechanisms, by referencing the EPRI MRPs 80-year Expert Panel results for the components in MRP-2018-022. However, the previous version of GALL-SLR Item IV.B2.RP-388 did not address void swelling in the LGWs. Thus, the staff added changes in dimension due to void swelling or distortion as an additional non-cracking effect and mechanism combination in the revised version of GALL-SLR Item IV.B2.RP-388, as updated in Appendix B.1 of the final ISG.

Cracking of the core barrel assembly UGWs has been incorporated into and is now addressed by the staffs revision of GALL-SLR Item IV.B2.RP-280 in Appendix B.1 of the final ISG (rRefer to the staffs technical basis statement for the RP-280 item on the previous page of this report table). Management of loss of fracture toughness due to neutron irradiation embrittlementIE and changes in dimension due to void swelling or distortion in the UGWs is addressed by the staffs development of the new GALL-SLR IV.B2.RP-280a line item in Appendix B.1 of the ISG (rRefer to staffs technical basis statement for the new RP-280a Iitem as given provided in the Table 2-7Table 2-3 Ssupplement of this report).

IV.B2.RP-387a The core barrel assembly middle vertical (axial) welds (MAWs)

IV.B2.RP-388a MAWs and lower vertical (axial) welds (LAWs)LAWs (which are the topic of the RP-387a Iitem for cracking effects and the RP-388a Iitem for non-cracking effects) are identified as Expansion category components for Westinghouse-design RVI management programs per Items W4.2 and W4.3 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. Therefore, to be consistent with the EPRI MRPs criteria in the W4.2 and W4.3 items, the staff modified GALL-SLR Items IV.B2.RP-387a and IV.B2.RP-388a to include both the core barrel MAWs and LAWs as the referenced core barrel assembly components cited in the line items.

Based on lessons learned from the staffs review of the RVI gap analysis in the SLRA for Surry Nuclear Plant, Units 1 and 2, the past applicant screened the core barrel assembly MAWs and LAWs screened in for the aging mechanisms of stress corrosion 2-36 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes cracking (SCC), irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), fatigue, neutron irradiation embrittlement (IE), and void swellingVS, as referenced to in the EPRI MRPs 80-year Eexpert Ppanel assessment of the components in MRP-2018-022. However, the previous version of the Item IV.B2.RP-388a item did not address void swelling in the MAWs and LAWs. Therefore, the staff added changes in dimension due to void swelling or distortion as an additional non-cracking effect and mechanism combination for the revised version of the RP-388a Iitem in Appendix B.1 of the final ISG.

IV.B2.RP-276 The core barrel assembly upper flange welds (UFWs) in Westinghouse-designed PWRs (which are the topic of the RP-276 Iitem for cracking effects and mechanisms) remain as leading Primary category components for Westinghouse-design RVI management programs per Item W3 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

Based on lessons learned from the staffs review of the RVI gap analysis in the SLRA for Surry Nuclear Plant, Units 1 and 2, the staff confirmed that the past applicant screened the core barrel assembly UFWs in for the aging mechanisms of stress corrosion cracking (SCC) and fatigue, as referenced to the EPRI MRPs 80-year Eexpert Ppanel assessment of the components in MRP-2018-022. Therefore, the staff edited GALL-SLR Item IV.B2.RP-276 in the ISG to reference both of these cracking mechanisms.

IV.B2.RP-285 The staffs update of the RP-399 Iitem addresses cracking in the IV.B2.RP-399 clevis insert components and the staffs update of the RP-285 item addresses non-cracking effect and mechanisms in the clevis insert components.

Specifically, the clevis insert assemblies and their components are treated by the EPRI MRP as applicable Existing Program components, as reflected in Item W14 of Table 4-9 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report; however, dowels were not included in the scope of Item W14. However, based on lessons learned from the staffs review of the RVI gap analysis in the SLRA for Surry Nuclear Plant, Units 1 and 2, the past applicant included three types of clevis insert assembly components would be inspected under the program, as linked to the EPRI MRPs 80-year Eexpert Ppanel assessment of the clevis insert assemblies in MRP-2018-022: (1) clevis insert bolts or screws, (2) clevis insert dowels, and (3) clevis insert surfaces. The Surry SLRA gap analysis identified that the clevis insert components are either susceptible to SCC or fatigue (but not both) as an applicable cracking mechanisms. For purposes of this reviewIn order to keep the update of GALL-SLR Item IV.B2.RP-399 simple, the staff is assuming that the clevis insert assembly components can be susceptible to either of the SCC or fatigue cracking mechanisms.

The MRP-227, Rev.Revision 1-A guidelines list the components as being made from of nickel alloy materials, but the staff has 2-37 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes conservatively included stainless steel in the RP-285 and RP-399 Iitems just in case a plant has specific clevis insert assembly components made from stainless steel. The design of a plants clevis inserts may have had the nickel alloy base metal modified with an outer layer of stellite as a precaution for protecting the components against wear, so stellite has been added to the RP-285 Iitem as a potential material for the clevis insert surfaces.

The staffs inclusion of changes in dimension due to void swelling or distortion in the final version of the GALL-SLR Item IV.B2.RP-285 item is based on recently reported operating experienceOE with distortion of a clevis insert assembly at the Ginna Nuclear Plant facility. Inclusion of changes in dimension in GALL-SLR Item IV.B2.RP-285 does not alter the EPRI MRPs basis for inspecting the clevis insert assemblies or there components in Item W14 of Table 4-9 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

IV.B2.RP-288 The lower core plates (including XL types, which are the topic of IV.B2.RP-289 the RP-289 Iitem for cracking effects and the RP-288 Iitem for non-cracking effects) remain as Existing Program category components for Westinghouse-design plants per Items W12a and W12b in Table 4-9 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

In the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP explained that the lower internals assemblies in Westinghouse-designed PWRs include either a normal sized lower core plate or an XL lower core plate for plants with 1.4-foot (ft) cores. The prior versions of the line items could have been interpreted that the lower internals assemblies of the plants included both types of core plates. The line items have been administratively edited to correct this.

Based on lessons learned from the staffs review of the RVI gap analysis in the SLRA for Surry Nuclear Plant, Units 1 and 2, the staff confirmed that the applicant screened the lower core plate in for the aging mechanisms of irradiation-assisted stress corrosion cracking (IASCC), fatigue, wear, irradiation embrittlement (IE) and void swelling (VS), as linked to the EPRI MRPs 80-year Eexpert Ppanel assessment of the components in MRP-2018-022. The revised version of GALL-SLR Item IV.B2.RP-288 in the ISG now includes both the IASCC and fatigue cracking mechanisms. The revised version of GALL-SLR Item IV.B2.RP-289 in the ISG now includes the aging effects associated with the wear, IE, and VS mechanisms.

IV.B2.RP-290a The lower support forging or casting in Westinghouse-designed IV.B2.RP-291a PWRs (which are the topic of the RP-291a Iitem for cracking effects and the RP-290a Iitem for non-cracking effects) remain as an Expansion category component for Westinghouse-design RVI management programs per Item W3.4 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The previous versions of the RP-290a and RP-291a Iitems in the GALL-SLR Rreport reported that the lower support casting or forging was located in the lower support structure. However, the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report identifies that the lower support casting or forging is located in the lower internals assembly of the plants. The line items have been edited to reference the assembly cited for the components referenced in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

In item W3.4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, EPRI screened the lower support forging or casting in for stress corrosion cracking (SCC), and additionally for thermal embrittlementTE if the lower support was a casting made from CASS. In the gap analysis of the Surry SLRA, the staff confirmed that the applicant screened lower support forgings for the units in for both SCC and fatigue cracking mechanisms. The scope of GALL-SLR Item IV.B2.RP-291a in the ISG includes both the SCC and fatigue mechanisms. The scope of GALL-SLR Item IV.B2.RP-290a in the ISG is now limiting only to lower support castings made from CASS, with the applicable aging effect and mechanism combination being cited as loss of fracture toughness due to thermal embrittlement. Thus, the revised versions of the RP-290a and RP-291a Iitems are now consistent with Item W3.4 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, as adjusted in RP-291a for lessons learned obtained from the staffs review of Surry SLRA.

IV.B2.RP-291 The lower support column bodies (both cast on non-cast types)

IV.B2.RP-294 remain as Expansion category components for Westinghouse-IV.B2.RP-290 design RVI management programs per Item W4.4 in Table 4-6 of IV.B2.RP-295 the MRP227, Rev. 1-A reportRevision 1-A Report.

The staff acknowledges that in Item W4.4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, EPRI only screened the lower support column bodies in for a stress corrosion cracking (SCC) mechanism, with thermal embrittlementTE being applicable if the column bodies were made from cast austenitic stainless steel (CASS). However, based on lessons learned obtained from the staffs review of the past Surry SLRA, the staff confirmed that the past Surry applicant screened the lower support column bodies in for the aging mechanisms of irradiation assisted stress corrosion crackingIASCC (irradiation-assisted SCC or IASCC), fatigue, wear, neutron irradiation embrittlement (IE), and void swelling (VS), and additionally for thermal aging embrittlement (TE) if the components are fabricated from CASS.

Thus, the staff used lessons learned from the past SLRA review as the main basis for the adjustments of GALL-SLR Items IV.B2.RP-291 and IV.B2.RP-294 and cited cracking due to IASCC or fatigue as the listed cracking effect and mechanisms for the AMR line items.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The staff also used these lessons learned as the main basis for revising GALL-SLR Item IV.B2.RP-290 and cited the applicable non-cracking effect and mechanism combinations as loss of fracture toughness due to thermal aging embrittlement and neutron irradiation embrittlement and changes in dimension due to void swelling or distortion for the cast column body types. In a similar fashion, the staff used these lessons learned as the main basis for revising GALL-SLR Item IV.B2.RP-295 and cited the applicable non-cracking effect and mechanism combinations as loss of fracture toughness due to neutron irradiation embrittlement and changes in dimension due to void swelling or distortion for the forged column body types.

IV.B2.RP-286 The lower support column bolts located in the lower support IV.B2.RP-287 assemblies of Westinghouse-designed PWRs (which are the topic of the RP-286 Iitem for cracking effects and the RP-287 Iitem for non-cracking effects) have been identified as Expansion category components for Westinghouse-design RVI management programs per Item W6.2 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The existing version of the Item IV.B2.RP-286 item in the GALL-SLR Rreport already addresses cracking due to irradiation-assisted IASCC or fatigue of the lower support column bolts and the staffs edit of the item to cite the stress corrosion mechanism as IASCC is strictly an administrative change to make the item consistent with other AMR line items that cite IASCC as an applicable mechanism.

In regard to the staffs modification of GALL-SLR Item IV.B2.RP-287, the staff acknowledges that Item W6.2 of the MRP-227, Rev.

1-A reportMRP-227, Revision 1-A Report only screened the lower support column bolts in for irradiation-assisted stress corrosion cracking (IASCC), fatigue, irradiation embrittlement (IE), and irradiation stress relaxation/irradiation-assisted creep (ISR/IC) mechanisms. However, based on lessons learned obtained from the staffs review of the past Surry SLRA, the staff confirmed that the past applicant screened the lower support column bolts in for wear and void swelling (VS) as additional aging mechanisms for the lower support column bolts.

Thus, the staff used lessons learned from the past SLRA review as the main basis for the modification of GALL-SLR Item IV.B2.RP-287 and, although the line item appropriately addressed IE and ISR/IC of the bolts, it did not address changes in dimension that could be induced by distortion or a VS mechanism or loss of material due to wear in the bolts. Therefore, the staff modified GALL-SLR Item IV.B2.RP-287 to include changes in dimension due to void swelling or distortion and loss of material due to wear as additional non-cracking effect and mechanism combinations for the line item.

IV.B2.RP-302 The thermal shield flexures located in the thermal shield assemblies of Westinghouse-designed reactors (which are the 2-40 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes topic of the RP-302 Iitem for cracking effects and the topic of the RP-302a Iitem for non-cracking effects) remain as Primary category components for Westinghouse-design RVI management programs per Item W9 in Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

Based on lessons learned from the staffs review of the Surry SLRA gap analysis, the staff confirmed that the past applicant did not screen the thermal shield flexures in the Surry units for void swelling (VS), irradiation embrittlement (IE), or irradiation-enhanced stress relaxation or irradiation-enhanced creep (ISR/IC),

as assessed per EPRIs 80-yYear Expert Panel analysis of the components in MRP-2018-022. Specifically, the gap analysis indicated that the projected 80-year fluence exposures of the thermal shield flexures are in a fluence zone lower than the threshold for screening the thermal shield flexures in for the Surrys referenced irradiation mechanisms. The staff confirmed that the gap analysis did screen the thermal shield flexures in for the aging mechanisms of stress corrosion cracking (SCC), fatigue, and wear. Thus, the staff accepted EPRIs comment implications that the RP-302 and RP-302a line Iitems for the thermal shield flexures should not include citation of irradiation-induced aging mechanisms.

Based on these confirmations, the staff edited GALL-SLR Item IV.B2.RP-302 item to cite both SCC and fatigue as the applicable cracking mechanisms for the line item. Based on the staffs acceptance of the EPRI MRP Comment #18, the staff confirmed that GALL-SLR Item IV.B2.RP-302a did not need to be edited or modified in the ISG, as the existing version of the line item in GALL-SLR already addresses loss of material due to wear in the thermal shield flexures.

IV.B2.RP-290b The upper core plate in the upper internals assembly of IV.B2.RP-291b Westinghouse-designed PWRs (which are the topic of the RP-291b item for cracking effects and the topic of the RP-290b item for non-cracking effects) remains as an Expansion category component for Westinghouse-design RVI management programs per Item W4.1 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

In Item W4.1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened Westinghouse-design upper core plates in for fatigue, wear, and irradiation embrittlement (IE) aging mechanisms. However, based on lessons learned obtained from the staffs review of the past Surry SLRA, the staff confirmed that the past applicant screened the upper core plates in the Surry units in for IASCC as an additional aging mechanism for the plates (i.e., in addition to fatigue, wear, and IE). Thus, the staff used MRP-227, Rev.Revision 1-A, as supplemented by lessons learned from the staffs past review of the Surry SLRA review, as the basis for adjusting GALL-SLR Items IV.B2.RP-290b and IV.B2.RP-291b in the final ISG.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Therefore, for the staffs update of GALL-SLR Item IV.B2.RP-290b item, the staff added loss of fracture toughness due to neutron irradiation embrittlement as an additional non-cracking effect and mechanism combination for the line item (i.e., in addition to citation of loss of material due to wear. Similarly, for the staffs update of GALL-SLR Item IV.B2.RP-291b, the staff added IASCC as an additional cracking mechanism for the line item (i.e., in addition to the reference of fatigue as an applicable cracking mechanism).

IV.B3.RP-312 The instrument guide tubes in the peripheral (outer) control IV.B3.RP-313 element assembly (CEA) shroud assemblies of Combustion Engineering (CE)-designed PWRs have been designated as Primary category components for CE-design RVI management programs per Item C11 in Table 4-2 of the report. The linked expansion components are the instrument guide tubes in the remaining (non-peripheral) CEA shrouds assemblies per Item C11.1 in Table 4-6 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

Therefore, staff edited GALL-SLR Item IV.B3.RP-312 to clearly identify that the scope of the line item applies to management of cracking in the instrument guide tubes of the peripheral CEA shroud assemblies, as identified as EPRI Primary category in MRP-227, Rev.Revision 1-A. Similarly, the staff edited GALL-SLR Item IV.B3.RP-313 to clearly indicate that the scope of the line item applies to the management of cracking in the guide tubes of the remaining CEA shroud assemblies, as identified as Expansion category components in MRP-227, Rev.Revision 1-A.

The RP-312 and RP-313 line items items appropriately identifying the applicable effect and mechanism combination as cracking due to SCC or fatigue.

IV.B3.RP-319 The guide lugs in the core shroud assemblies and guide lug IV.B3.RP-320 inserts and bolts in the upper internals assemblies of Combustion Engineering (CE)-designed PWRs (which are the topic of the RP-320 item for cracking effects and the topic of the RP-319 Iitem for non-cracking effects) remains as Existing Program category components for CE-design RVI management programs per Items C13 and C14 in Table 4-8 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

The core shroud/upper internals assembly lugs and lug inserts and bolts were screened in for fatigue, wear and irradiation-enhanced stress relaxation or creep (ISR/IC) in Table 3-2 of the MRP-227, Revision 1- A Rreport. The aging effect and mechanism combinations in amended versions of the RP-319 and RP-320 Iitems are consistent with those in the MRP-227, Rev.Revision 1-A basis, with the RP-319 Iitem citing the non-cracking basis as loss of material due to wear; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep, and the RP-320 Iitem citing the cracking basis as cracking due to fatigue.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes IV.B3.RP-358 The GALL- SLR Items IV.B3.RP-358 and IV.B3.RP-318 items IV.B3.RP-318 apply to specific referenced components (including the core side surfaces, core shroud plates and plate joints, and core shroud bolts and locking devices) that are located in the core shroud assemblies of CE-designed PWRs that have bolted core shroud designs. The information in Item C4 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report has designated the specified components as being Primary category components for CE-design RVI management programs.

These core shroud assembly components were screened in for irradiation-assisted stress corrosion crackingIASCC (irradiation-assisted SCC or IASCC), void swelling (VS), and neutron irradiation embrittlement (IE) aging mechanisms per Table 3-2 of the MRP-227, Revision 1- A reportReport. To be consistent with the this screening basis, the staff amended GALL-SLR Item RP-358, which addresses cracking due to irradiation-assistedIA SCC in the specified components. and tThe staff also amended version of the GALL-SLR Item IV.B3.RP-318, which addresses loss of fracture toughness due to neutron irradiation embrittlementIE and changes in dimension due to void swelling or distortion in the specified components.

The change in the RP-358 Iitem to cite irradiation-assisted SCC as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-316 Other than the minor adjustment of the RP-316 Iitem, the existing versions of these line items are consistent with the EPRI MRPs criteria for barrel-shroud bolts in bolted CE core shroud assembly designs, as designated in Item C1.2 of Table 4-5 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

In Item C1.2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, EPRI MRP designate that the barrel-shroud bolts are Expansion category components for CE-design PWRs that have these types of bolted core shroud designs. The EPRI MRP screened the bolts in for irradiation-assisted stress corrosion cracking (IASCC), fatigue, neutron irradiation embrittlement (IE),

and irradiation-enhanced stress relaxation or creep (ISR/IC) aging mechanisms. The staffs edit of GALL-SLR Item IV.B3.RP-316 in the ISG addresses cracking due to IASCC or fatigue in the barrel-shroud bolts and the existing version of the GALL-SLR Item IV.B3.RP-317 in NUREG-2191, Volume 1 addresses management of loss of fracture toughness due to neutron irradiation embrittlementIE and loss of preload due to thermal or irradiation-enhanced stress relaxationISR or creep in the barrel-shroud bolts.

IV.B3.RP-314 The referenced core shroud bolts are the topic of the RP-314 IV.B3.RP-315 item for cracking effect and mechanism combinations and the RP-315 Iitem for non-cracking effect and mechanism combinations.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The referenced core shroud bolts remain as Primary category components for CE -plants that have bolted core shroud designs, as designated in Item C1 of Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 3-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, EPRI screened the core shroud bolts in for irradiation-assisted stress corrosion cracking (IASCC), fatigue, neutron irradiation embrittlement (IE), void swelling (VS), and irradiation-enhanced stress relation or creep (ISR/IC) aging mechanisms. Therefore, the staff amended GALL-SLR Item IV.B3.RP-314 to cite the aging effect and mechanism combination in the line item as Cracking due to IASCC or fatigue and GALL-SLR Item IV.B3.RP-315 to cite the aging effect and mechanism combinations in the line item as loss of preload due to thermal and irradiation-enhanced stress relaxation or creep; loss of fracture toughness due to neutron irradiation embrittlementIE; changes in dimension due to void swelling or distortion.

The change in the RP-315 Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-326 The core shroud assemblies referenced in GALL-SLR Item IV.B3.RP-326 apply to Combustion Engineering (CE)-designed plants whose core shrouds are welded in two vertical shroud sections. These assemblies are the topic of the RP-326 Iitem for referenced non-cracking effect and mechanism combinations.

For CE-designed plants that are designed with this type of shroud assembly, the shroud assemblies (including the weld seams between the shroud segments) remain as Primary components for CE-design RVI management programs per Item C4a in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. The existing version of the Item IV.B3.RP-326 line item in Table IV.B3 of NUREG-2191, Volume 1 (i.e., the GALL- SLR Rreport) included a parenthetical phrase in the Structure and/or Component column entry of the line that relates to the component accessibility and coverage criteria that apply to the components.

However, these criteria are already adequately established and addressed in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. As a result, there is no need to include such descriptions in the line items. Instead the staff edited parenthetical phrase in Structure and/or Component column entry of the line item to clarify used as the replacement phrases relate to the actual core shroud components that will be inspected in accordance with Item C4a of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

In Item C4a of Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP only cited neutron irradiation embrittlement (IE) and void swelling as listed aging mechanisms for these types of shroud assemblies. The RP-326 2-44 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Iitem continues to cite loss of fracture toughness due to neutron irradiation embrittlementIE; changes in dimension due to void swelling or distortion as the applicable non-cracking effect and mechanism combinations for the shroud assembly components, which is consistent with non-cracking mechanisms cited for the components in the MRP-227. Rev. 1-A reportRevision 1-A Report.

IV.B3.RP-322 The core shroud plate-to-former plate welds referenced in GALL-IV.B3.RP-359 SLR Items IV.B3.RP-322 and IV.B3.RP-359 apply to Combustion Engineering (CE)-designed plants whose core shrouds are welded in two vertical shroud sections. These welds are the topic of the RP-322 Iitem for cracking effect and mechanism combinations and the RP-359 Iitem for non-cracking effect and mechanism combinations.

For CE-designed PWRs whose core shrouds are designed and assembled with two vertical sections, the core shroud-plate-to-former plate welds remain as Primary components for the RVI management programs per Item C2 in Table 4-2 of the MRP-227, Rev.Revision 1-1 report. In Table 3-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the shroud plates and former plates (and their plate-to-plate welds) in for irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), neutron irradiation embrittlement (IE),

and void swelling (VS) aging mechanisms. The staff revised GALL-SLR Items IV.B3.RP-322 and IV.B3.RP-359a to be consistent with the aging mechanisms cited for the core shroud plate-to-former plate welds in Item C2 of MRP-227, Rev.Revision 1-A, with the RP-322 Iitem citing those associated with cracking of the welds (i.e., IASCC) and the RP-359 Iitem citing those associated with the non-cracking effects (i.e., IE and VS/distortion) that are attributed to the welds.

The change in the RP-322 Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-323 The remaining core shroud axial welds referenced in the RP-IV.B3.RP-359a 323 and RP-359a Iitems apply to Combustion Engineering (CE)-

designed plants whose core shrouds are welded in two vertical shroud sections. These welds are the topic of the RP-323 Iitem for cracking effect and mechanism combinations and the RP-359a Iitem for non-cracking effect and mechanism combinations.

For CE-designed PWRs whose core shrouds are designed and assembled with two vertical sections, the referenced axial welds are identified as Expansion category components for the RVI management programs per Item C2.1 in Table 4-5 of the MRP-227, Rev.Revision 1-1 A Rreport. In Item C2.1 the EPRI MRP screened the remaining core shroud axial welds in for irradiation-assisted stress corrosion cracking (IASCC) and neutron irradiation embrittlement (IE) aging mechanisms.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The staff revised GALL-SLR Items IV.B3.RP-323 and IV.B3.RP-359a to be consistent with the aging mechanisms cited for the remaining axial welds in Item C2.1 of MRP-227, Rev.Revision 1-A, with the RP-322 citing those associated with cracking of the welds (i.e., IASCC) and the RP-359 citing those associated with the non-cracking effects that are attributed to the welds (i.e., IE, but not VS or distortion).

The change in the RP-323 Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-325 The remaining core shroud axial welds and the core shroud ribs IV.B3.RP-361 and rings referenced in GALL-SLR Items IV.B3.RP-325 and IV.B3.RP-361 apply to Combustion Engineering (CE)-designed plants whose core shrouds are fabricated from welded full -height shroud plates. These components are the topic of the RP-325 Iitem for cracking effect and mechanism combinations and the RP-361 Iitem for non-cracking effect and mechanism combinations.

For CE-designed PWRs with these types of core shroud designs, the referenced axial welds, ribs and rings are identified as Expansion category components for the RVI management programs per Item C3.1 or C3.2 in Table 4-5 of the MRP-227, Rev.Revision 1-1A Rreport. In Items C3.1 and C3.2, the EPRI MRP screened the components in for irradiation-assisted stress corrosion cracking (IASCC) and neutron irradiation embrittlement (IE) aging mechanisms. The staff revised GALL-SLR Items IV.B3.RP-325 and IV.B3.RP-361 to be consistent with the collective set of components and aging mechanisms cited in Items C3.1 and C3.2 of MRP-227, Rev.Revision 1-A, with the RP-325 citing those associated with cracking of the components (i.e.,

IASCC) and the RP-361 citing those associated with the non-cracking effects that are attributed to the welds (i.e., IE).

The change in the RP-325 Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-324 The core shroud plates referenced in GALL-SLR Items IV.B3.RP-IV.B3.RP-360 324 and IV.B3.RP-360 apply to Combustion Engineering (CE)-

designed plants whose core shrouds are fabricated from welded full -height shroud plates. These components are the topic of the RP-324 Iitem for cracking effect and mechanism combinations and the RP-360 Iitem for non-cracking effect and mechanism combinations.

The referenced shroud plates are identified as Primary category components for the RVI management programs per Item C3 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 3-2 of the MRP-227, Rev. 1-A reportMRP-227, 2-46 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Revision 1-A Report, the EPRI MRP screened the shroud plates in for irradiation-assisted stress corrosion cracking (IASCC), neutron irradiation embrittlement (IE), and void swelling (VS) aging mechanisms. The edited or modified versions of the RP-324 and RP-360 Iitems are consistent with the cited aging mechanisms, with RP-324 citing those associated with cracking of the components (i.e., IASCC) and RP-360 citing those associated with the non-cracking effects attributed to the components (i.e., IE and VS). This required the staffs addition of changes in dimension due to void swelling or distortion as an additional non-cracking effect and mechanism combination for the RP-360 Iitem.

Additionally, for the RP-324 Item, the previous inclusion of the phrase at the core mid plane (+3 feet in height) as visible from the core side of the shroud in the component description of the line item related to specific location and coverage criteria for the axial weld seams that were defined in the earlier MRP-227 Revision 1-A Report -A report. These criteria are no longer included in the updated guidelines in MRP-227, Rev.Revision 1-A Report.

As a result of this change in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the staff deleted this phrase from the scope of component description in the revised GALL-SLR IV.B3.RP-324 Iitem.

The change in the RP-324 Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-328 The core support barrel (CSB) assembly flexure weld cited in the modified version of GALL-SLR Item IV.B3.RP-328 applies to all Combustion Engineering (CE)-designed plants with welded core shroud assembly designs. The flexure welds are the topic of the RP-328 item for cracking effect and mechanism combinations.

In EPRI MRP Letter No. MRP 2020-012 (dated May 4, 2020), the EPRI MRP clarified that the CSB flexure weld in Combustion Engineering (CE)-designed plants was one of two circumferential welds in the lower flange of the CSB assembly, with the flexure weld being identified as a Primary category component per Item C7 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. The EPRI MRP explained that the weld joins the CSB lower flange to a flexure in the lower support structure. In line item C7, the EPRI MRP cites fatigue and stress corrosion cracking (SCC) as the applicable cracking mechanisms for the CSB flexure weld.

The EPRI MRP also explained that the other circumferential weld adjoins the lower flange to the core support barrel and has been renamed and referenced as the CSB lower girth weld (LGW) in Item C5.1 of Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, 2-47 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Revision 1-A Report (in order to avoid confusion with the flexure weld). The CSB LGW was downgraded-graded to the Expansion category for CE-designed plants and is now addressed by the staffs revision of GALL-SLR Item IV.B3.RP-333 for cracking mechanisms. Irradiation embrittlementThe IE in the CSB LGW is now addressed by the new GALL-SLR Item IV.B2.RP-333a (rRefer to the technical basis statement for the RP-333a item in Table 2-7 Table 2-3 of this report)..

Since the RP-328 Iitem is the GALL-SLR item that correlates to the applicable Primary category LFW component, the component for the RP-328 Iitem has been changed to reflect the CSB flexure weld as the applicable Primary category component for the line item. The revised version of GALL-SLR Item IV.B2.RP-328 continues to reference the applicable cracking mechanisms as SCC and fatigue.

IV.B3.RP-362 The core support barrel (CSB) middle girth welds (MGWs) cited in IV.B3.RP-362a the modified versions of GALL-SLR Items IV.B3.RP-362 and IV.B3.RP-362a apply to all Combustion Engineering (CE)-

designed plants. The MGWs are the topic of the RP-362a Iitem for cracking effect and mechanism combinations and the topic of the RP-362 Iitem for non-cracking effect and mechanism combinations.

The CSB MGWs have been identified as Primary category components per Item C6 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Item C6, the EPRI MRP screened the CSB MGWs in for stress corrosion cracking (SCC),

irradiation-assisted stress corrosion cracking (IASCC), and neutron irradiation embrittlement (IE) aging mechanisms. The staff modified GALL-SLR Items IV.B3RP-362 and IV.B3.RP-362a to cite the CSB MGWs as the applicable Primary category components for the line items and to reference the aging mechanisms cited in Item C6 of MRP-227, Rev.Revision 1-A, with RP-362a citing those associated with cracking of the MGWs (i.e.,

SCC and IASCSC) and RP-362 citing those associated with the non-cracking effects attributed to the components (IE).

The change in the RP-362a Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-362b The core support barrel (CSB) middle vertical (axial) welds IV.B3.RP-362c (MAWs) and lower vertical (axial) welds (LAWs) cited in the modified versions of the RP-362 and RP-362a Iitems apply to all Combustion Engineering (CE)-designed plants. The CSB MAWs and LAWs are the topic of the RP-362c Iitem for cracking effect and mechanism combinations and the topic of the RP-362b Iitem for non-cracking effect and mechanism combinations.

The CSB MAWs and LAWs have been identified as Expansion category components per Items C6.1 and C6.2 in Table 4-5 of the 2-48 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Items C6.1 and C6.2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the CSB MAWs and LAWs in for stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), and neutron irradiation embrittlement (IE) aging mechanisms.

However, the prior versions of the RP-362b and RP-362c Iitems did not include the MAWs. Therefore, the staff modified GALL-SLR Items IV.B3.RP-362b and IV.B3.RP-362c to cite the CSB MAWs and LAWs as the applicable Primary category components for the line items and to reference the aging mechanisms cited in Items C6.1 and C6.2 of MRP-227, Rev.Revision 1-A, with RP-362c citing those associated with cracking of the components (i.e., SCC and IASCC) and RP-362b citing those associated with the non-cracking effects attributed to the components (i.e., IE).

The change in the RP-362c Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-333 The core support barrel (CSB) lower girth weld (LGW) cited in the RP-333 Iitem applies to all Combustion Engineering (CE)-

designed plants. The LGW is the topic of the RP-333 Iitem for cited cracking effect and mechanisms combinations.

In EPRIs MRP-2020-012 letter (dated May 4, 2020), EPRI clarified that this girth weld is the CE-design circumferential weld that adjoins the CSB lower flange to the CSB. The EPRI MRP confirmed that the weld is an Expansion component per Item C5.1 in Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Item C5.1, EPRI MRP cited that the CSB LGW is susceptible to stress corrosion cracking (SCC) and fatigue aging mechanisms. However, in EPRIs response to request for additional information (RAI) 26 on MRP-227, Rev.Revision 1 dated October 16, 2017 (ADAMS Accession No. ML17305A056), the EPRI MRP clarified that the referenced CSB LGW is also susceptible to IASCC and neutron irradiation embrittlement (IE) as additional aging mechanisms for the weld (i.e., in addition to SCC and fatigue).

Thus the staff modified the scope of GALL-SLR Item IV.B3.RP-333 to cite the CSB LGW (LFW) that is referenced in Item C5.1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report and to reference the cracking mechanisms (i.e., SCC, IASCC and fatigue) referenced for CSB LGWs in EPRI Letter of October 16, 2017.

Management of the remaining aging effect and mechanism combination (i.e., loss of fracture toughness due to neutron irradiation embrittlement) in the CSB LGW is being address by the development of a new line item, GALL-SLR Item IV.B3.RP-333a 2-49 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes (Refer to the technical basis for the new RP-333a line in the NUREG-2221, Table 2-14Table 2-10 Supplement of this report).

IV.B3.RP-332 The core support barrel (CSB) assembly upper flange cited in the edited version of GALL-SLR Item IV.B3.RP-332 applies to all Combustion Engineering (CE)-designed plants. The CSB upper flange is the topic of the RP-332 Iitem for cited non-cracking effect and mechanism combinations.

The CSB upper flange remains as an Existing Program category component for CE-design RVI management programs per Item C16 in Table 4-8 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 4-8 of the report, the EPRI MRP screened the CSB upper flange in for wear as the applicable aging mechanism of concern. The staffs change to GALL-SLR Item IV.B3.RP-332 is a simple administrative change to make the component description consistent with Item C16 in Table 4-8 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, and the line item remains consistent with aging mechanism basis in Item C16 by citing the applicable aging effect and mechanism combination as loss of material due to wear.

IV.B3.RP-327 The core support barrel (CSB) assembly upper flange weld (UFW) cited in the edited version of GALL-SLR Item IV.B3.RP-327 applies to all Combustion Engineering (CE)-designed plants. The CSB UFW is the topic of the RP-332 Iitem for cited cracking effect and mechanism combinations.

The CSB UFW is identified as a Primary category component for CE-design RVI management programs per Item C5 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 3-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the UFW in for stress corrosion cracking (SCC) and wear aging mechanisms. The staffs change to GALL-SLR Item IV.B3.RP-327 is a simple administrative change to make the component description consistent with Item C5 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, and the line item remains consistent with aging mechanism basis in Item C5 by citing the applicable aging effect and mechanism combination as cracking due to SCC.

Management of loss of material due to wear in the CSB UFW is adequately addressed in GALL-SLR Item IV.B3.RP-332 and by implementation of the EPRI MRPs Existing Program protocols for inspecting the upper flange containing the welds for evidence of wear, as defined in Item C16 in Table 4-8 of the MRP-227, Rev.

1-A reportMRP-227, Revision 1-A Report. Refer to the technical basis statement for edits to GALL-SLR Item IV.B3.RP-332 on the previous page ofdescribed above in this technical basis statement table.

IV.B3.RP-329 The core support barrel (CSB) assembly upper circumferential IV.B3.RP-455 (girth) weld (UGW) and upper vertical (axial) welds (UAWs) cited in GALL-SLR Items IV.B3.RP-329 and IV.B3.RP-455 apply to all 2-50 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Combustion Engineering (CE)-designed plants. The CSB UGW and UAWs are the topic of the RP-329 Iitem for cited cracking effect and mechanism combinations and the R-455 Iitem for cited non-cracking effect and mechanism combinations.

The CSB UGW and UAWs are identified as Expansion category components for CE-design RVI management programs per Items C5.2 and C5.3 in Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 4-5 of the MRP-227, Rev. 1-A reportRevision 1-A Report, the EPRI MRP screened the CSB UGWs and UAWs in for stress corrosion cracking (SCC) and neutron irradiation embrittlement (IE) aging mechanisms. The staff modified the component descriptions in GALL-SLR Items IV.B3.RP-329 and IV.B3.RP-455 to be consistent with the component descriptions in Items C5.2 and C5.3 of Table 4-5 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. The modified version of GALL-SLR Item IV.B3.RP-329 continues to cite the applicable aging effect and mechanism as cracking due to SCC and the modified version of GALL-SLR Item IV.B3.RP-455 continues to cite the applicable non-cracking effect and mechanism as loss of fracture toughness due to neutron irradiation embrittlement.

IV.B3.RP-357 The incore instrumentation (ICI) thimble tubes cited in GALL-SLR Item IV.B3.RP-357 apply to all Combustion Engineering (CE)-

designed plants. The ICI thimble tubes are the topic of the RP-357 Iitem for loss of material due wear.

For the SLR-ISG-2021-01-PWRVI updates, the staff decided to break the SRP-SLR line item reference for Combustion Engineering (CE)-designed ICI thimble tubes out of SRP-SLR Table 3.1-1, Item 028 in order to decouple the mixing Westinghouse-designated components (i.e., CRGT support pins

[split pins]) and CE-designated components in the same SRP-SLR Item. The ICI thimble tubes (which were previously referenced in the SRP-SLR Table 3.1-1 028 Iitem) are referenced in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. Specifically, in Table 3-2 (page 3-26) of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP designates that the ICI thimble tubes (lower) are Existing Program components for CE-design RVI management programs.

Therefore, the staff realigned reference of the RP-357 Iitem out of SRP-SLR Table 3.1-1, Item 028, and into SRP-SLR Table 3.1-1, Item 056c, is consistent with this type of basis. This also resulted in the need for the staff to edit and re-align the SRP-SLR Table 3.1-1 item reference in GALL-SLR Item IV.B3.RP-357 as 3.1-1, 056c in order to ensure appropriate cross linking with SRP-SLR Table 3.1-1, Item 056c.

IV.B3.RP-363 The core support columns cited in GALL-SLR Items IV.B3.RP-363 IV.B3.RP-364 and IV.B3.RP-364 apply to all Combustion Engineering (CE)-

designed plants that are designed with full height bolted core shroud assemblies or half height welded core shroud assemblies 2-51 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes (i.e., shroud assemblies assembled from two vertical sections).

The core support columns are the topic of the RP-363 Iitem for cracking effect and mechanism combinations and the RP-364 Iitem for non-cracking effect and mechanism combinations.

The core support columns in these CE plant designs are identified as Expansion category components for the RVI management programs per Item C6.3 in Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 3-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the core support columns in for stress corrosion cracking (SCC), irradiation assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), fatigue, and neutron irradiation embrittlement (IE) aging mechanisms, and additionally for thermal aging embrittlement (TE) if the components were fabricated from cast austenitic stainless steel materials (CASS).

The staffs modified version of GALL-SLR Item IV.B3.RP-363 is consistent with the cited cracking mechanism basis by citing the applicable aging effect and mechanism combination as cracking due to SCC, IASCC, or fatigue. The staffs modified version of GALL-SLR Item IV.B3.RP-364 is consistent with the cited non-cracking mechanism basis by citing the applicable aging effect and mechanism combination as Loss of fracture toughness due to neutron irradiation and thermal embrittlement (TE for CASS materials only).

The change in the RP-363 Iitem to cite irradiation-assisted SCC mechanism as IASCC is strictly administrative edit to make the line item consistent with other GALL-SLR AMR line items that cite IASCC as a referenced mechanism.

IV.B3.RP-334 The fuel alignment pins referenced in GALL-SLR Item IV.B3.RP-IV.B3.RP-336 334 apply to Combustion Engineering (CE)-designed plants with welded core shroud assemblies that are assembled from two vertical sections and address cracking in the pins.

The fuel alignment pins referenced in the modified version of GALL-SLR Item IV.B3.RP-336 apply to all CE-designed plants with welded core shrouds that are fabricated from two vertical sections or with core shrouds that are fabricated from full -height shroud plates, and address non-cracking effect and mechanism combinations in the pins.

The fuel alignment pins in these plant designs are identified as Existing Program category components for CE-design RVI management programs per Item C15a or C15b in Table 4-8 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 4-8, the EPRI MRP screened the fuel alignment pins in for the following aging mechanisms: (1) for those in CE plants with core shrouds assembled from full -height plates, stress corrosion cracking (SCC), irradiation assisted stress corrosion cracking (IASCC), fatigue,. irradiation embrittlement (IE), and irradiation-2-52 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes enhanced stress relaxation or creep (ISR/IC); and (2) for those in CE plants with welded core shrouds assembled from two vertical sections, wear, IE, and ISR/IC.

The staffs edited version of GALL-SLR Item-334 is consistent with the cited cracking mechanism basis by citing the applicable aging effect and mechanism combination as cracking due to SCC, IASCC, or fatigue. The change in the RP-334 Iitem to cite the irradiation SCC mechanisms as IASCC is strictly an administrative change. The staffs modified version of GALL-SLR Item-336 is consistent with the cited non-cracking mechanism basis by citing the applicable aging effect and mechanism combinations as loss of material due to wear; loss of fracture toughness due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep. The scope of the staffs modification of GALL-SLR Item IV.B3.RP-336 incorporates the fuel alignment pins previously in the scope of GALL-SLR Item IV.B3.RP334a, which was deleted in the ISG.

IV.B3.RP-335 The lower core support beams cited in the modified version of the RP-335 item apply to all Combustion Engineering (CE)-designed PWRs, except for those with core shroud assembles assembled from full height shroud plates. The lower core support beams are the topic of the RP-335 Iitem for cracking effect and mechanism combinations.

The lower core support beams in these plant designs have been identified as Expansion category components for CE-design RVI management programs per Item C5.4 in Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the lower core support beams in for stress corrosion cracking (SCC) and, fatigue. The edited version of the RP-335 Iitem is consistent with the cited cracking mechanism basis by citing the applicable aging effect and mechanism combination as cracking due to SCC or fatigue.

IV.B3.RP-343 The core support plates cited in the edited versions of the RP-IV.B3.RP-365 343 and RP-365 Iitems apply to all Combustion Engineering (CE)-designed PWRs whose plant designs include core support plates in the lower support structure of the reactor. The core support plates are the topic of the RP-343 Iitem for cracking effect and mechanism combinations and the RP-365 Iitem for non-cracking effect and mechanism combinations.

The core support plates in these plant designs have been identified as Primary category components for CE-design RVI management programs per Item C9 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the core support plates in for fatigue and neutron irradiation embrittlement (IE) aging mechanisms. The edited version of the RP-343 Iitem is consistent with the cited 2-53 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes cracking mechanism basis and the edited version of the RP-365 item is consistent with the IE mechanism basis.

IV.B3.RP-342 The deep beams cited in the edited versions of the RP-342 and IV.B3.RP-366 RP-366 Iitems apply to those Combustion Engineering (CE)-

designed PWRs that have welded core shrouds made from full height shroud plates. The deep beams are the topic of the RP-342 Iitem for cracking effect and mechanism combinations and the RP-366 Iitem for non-cracking effect and mechanism combinations.

The deep beams in these plant designs have been identified as Primary category components for CE-design RVI management programs per Item C12 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 3-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the deep beams in for stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), fatigue and neutron irradiation embrittlement (IE) aging mechanisms. The edited version of the RP-342 Iitem is consistent with the cited cracking mechanism basis and the edited version of the RP-366 Iitem is consistent with the IE mechanism basis.

IV.B3.RP-330 As clarified in EPRI MRP Letter No. MRP 2020-012 (dated May 4, IV.B3.RP-331 2020), the core support column bolts cited in the RP-330 and RP-331 Iitems only apply to the reactor internals design at the Palisades Nuclear Power Plant.. The core support column bolts are the topic of the RP-330 Iitem for cracking effect and mechanism combinations and the RP-331 Iitem for non-cracking effect and mechanism combinations.

The core support column bolts at Palisades have been identified as Expansion category components for CE-design RVI management programs per Item C1.1 in Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the core support column bolts in for irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), fatigue and neutron irradiation embrittlement (IE) aging mechanisms. The edited version of the RP-330 Iitem is consistent with the cited cracking mechanism basis and the edited version of the RP-331 Iitem is consistent with the IE mechanism basis.

IV.B3.RP-338 The fuel alignment plates cited in the RP-338 Iitem apply to all Combustion Engineering (CE)-designed PWRs with welded core shrouds made from full height shroud plates. The fuel alignment plates are the topic of the RP-338 Iitem for cracking effect and mechanism combinations (a new item, GALL-SLR Item IV.B3.RP-338a, has been developed to address non-cracking effect and mechanism combinations that apply to the fuel alignment plates).

2-54 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes The fuel alignment plates in these plant designs have been identified as Primary category components for CE-design RVI management programs per Item C10 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In Item C10 of Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP screened the fuel alignment plates in for fatigue and neutron irradiation embrittlement (IE) aging mechanisms. The edited version of the RP-338 Iitem is consistent with the cited cracking mechanism basis and the new RP-338a Iitem is consistent with the IE mechanism basis.

IV.B4.RP-245 The surveillance specimen holder tube (SSHT) bolts/studs cited in the RP-245 Iitem and the associated bolt locking devices cited in the RP-245a and RP-245b Iitems only apply to the Davis-Besse Nuclear Plantnuclear plant. The SSHT bolts/studs are the subject of the RP-245 Iitem for cracking effect and mechanism combinations. The SSHT bolt locking devices are the subject of the RP-245a Iitem for cracking effect and mechanism combinations and the RP-245b Iitem for non-cracking effect and mechanism combinations.

The SSHT bolts/studs, and locking devices, remain as Expansion category components for the RVI program of the Davis- Besse plant per Item B7.2 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. The line item does not list nuts as a component subject to aging management. In the B7.2 item, the EPRI MRP screened the SSHT bolts/studs in for stress corrosion cracking (SCC), fatigue, wear, and irradiation-enhanced stress relaxation or creep (ISR/IC) aging mechanisms, and the associated locking devices in for fatigue and wear mechanisms.

The modified version of the RP-245 Iitem is consistent cited cracking mechanisms of SCC and fatigue. The existing versions of the RP-245a and RP-245b Iitems are consistent with the identification of fatigue and wear as applicable mechanisms for the locking devices.

Loss of material due to wear and loss of preload in the SSHT bolts/studs are addressed by the new GALL-SLR Item IV.B4.RP-245c (rRefer to the Appendix B.9 entry for the RP-245c Iitem).

IV.B4.RP-247 The lower core barrel (LCB) bolts cited in the RP-247 Iitem and the associated bolt locking devices cited in the RP-247a and RP-247b Iitems apply to all Babcock and Wilcox (B&W)-designed PWRs. The LCB bolts are the subject of the RP-247 Iitem for cracking effect and mechanism combinations. The LCB bolt locking devices are the subject of the RP-247a Iitem for cracking effect and mechanism combinations and the RP-247b Iitem for non-cracking effect and mechanism combinations.

The LCB bolts and LCB bolt locking devices remain as Primary category components for the B&W-design RVI management programs per Item B8 in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B8 Iitem, the EPRI MRP screened the LCB bolts in for stress corrosion cracking 2-55 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes (SCC), fatigue, wear, and irradiation-enhanced stress relaxation or creep (ISR/IC) aging mechanisms, and the associated locking devices in for fatigue and wear mechanisms. The modified version of the RP-247 Iitem is consistent cited cracking mechanisms of SCC and fatigue. The staff deleted LCB bolt locking devices from the scope of the RP-247 because they are already within the scope of the AMRs in the RP-247a and RP-247b Iitems. The existing versions of the RP-247a and RP-247b Iitems for the LCB bolt locking devices are consistent with the identification of fatigue and wear as applicable mechanisms for the components.

Loss of material due to wear and loss of preload in the LCB bolts are addressed by the new GALL-SLR Item IV.B4.RP-247c (rRefer to the SLR-ISG-2021-01-PWRVI Appendix B.9 technical basis entry for the new RP-247c Iitem).

IV.B4.RP-240 The baffle-to-former bolts cited in the modified RP-240 and RP-IV.B4.RP-241 241 Iitems apply to all Babcock and Wilcox (B&W)-designed PWRs. The baffle-to-former bolts are the subject of the RP-241 Iitem for cracking effect and mechanism combinations and the RP-240 Iitem for non-cracking effect and mechanism combinations.

The baffle-to-former bolts remain as Primary category components for the B&W-design RVI management programs per Item B9 in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B9 Iitem, the EPRI MRP screened the baffle-to-former bolts in for irradiation-assisted stress corrosion cracking (Irradiation-assisted SCC or IASCC), fatigue, overload, wear, neutron irradiation embrittlement (IE), and irradiation-assisted stress relaxation or creep (ISR/IC) aging mechanisms.

The modified RP-241 Iitem is consistent with the citing of the applicable cracking mechanisms of IASCC, overload, and fatigue and the modified RP-240 Iitem is consistent with the citing of the applicable non-cracking mechanisms of wear, IE, and ISR/IC.

Screws do not need to be referenced in the modified versions of the line items because the terminology is considered to be somewhat redundant with the referencing of bolts in the line items.

IV.B4.RP-243 The baffle-to-baffle bolts and core barrel-to-former bolts cited in IV.B4.RP-244 the edited versions of the RP-243 and RP-244 Iitems apply to all Babcock and Wilcox (B&W)-designed PWRs. The baffle-to-baffle bolts and the core barrel-to-former bolts are the subject of the RP-244 Iitem for cracking effect and mechanism combinations and the RP-243 Iitem for non-cracking effect and mechanism combinations.

The baffle-to-baffle bolts and core barrel-to-former bolts remain as Expansion category components for the B&W-design RVI management programs per Items B9.1 and B9.2 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B9.1 and B9.2 Iitems, the EPRI MRP screened the bolts for irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC), neutron irradiation embrittlement (IE), fatigue, 2-56 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes overload, wear, and irradiation-enhanced stress relaxation or creep (ISR/IC) aging mechanisms. The edited version of the RP-244 Iitem is consistent with the cited cracking mechanisms of IASCC, fatigue, and overload. The edited version of the RP-243 Iitem is consistent with the cited non-cracking mechanisms of IE, wear, and ISR/IC.

Item B9.1 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report already adequately defines and differentiates between Expansion-based aging management criteria for external baffle-to-baffle bolts from those established for internal baffle-to-baffle bolts. Therefore, there is no need to differentiate external baffle-to-baffle bolt types from internal baffle-to-baffle bolt types in the RP-243 and RP-244 line Iitems.

IV.B4.RP-240a The locking devices (including locking welds) for baffle-to-former IV.B4.RP-241a bolts and internal baffle-to-baffle bolts cited in the edited or modified versions of the RP-240a and RP-241a Iitems apply to all Babcock and Wilcox (B&W)-designed PWRs. The locking devices are the subject of the RP-241a Iitem for cracking effect and mechanism combinations and the RP-240a Iitem for non-cracking effect and mechanism combinations.

The cited locking devices are identified as Primary category components for the B&W-design RVI management programs per Item B11 in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B11 item, the EPRI MRP screened the locking devices in for irradiation-assisted stress corrosion cracking (irradiation-assisted SCC or IASCC) and neutron irradiation embrittlement (IE) aging mechanisms. The edited version of the RP-241a Iitem is consistent with the cited IASCC mechanism and the edited version of the RP-240a Iitem is consistent with the cited IE mechanism.

IV.B4.RP-243a The locking devices (including locking welds) for external baffle-to-baffle bolts and core barrel-to-former bolts cited in the edited/modified versions of the RP-243a Iitem apply to all Babcock and Wilcox (B&W)-designed PWRs. The locking devices are the subject of the RP-243a Iitem for non-cracking effect and mechanism combinations.

The cited locking devices are identified as Expansion category components for the B&W-design RVI management programs per Item B11.1 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B11.1 Iitem, the EPRI MRP screened the locking devices in for the neutron irradiation embrittlement (IE) aging mechanism. The modifications of the RP-243a line make the GALL-SLR item consistent with Item B11.1 in Table 4-4 of MRP-227, Rev.Revision 1-A.

IV.B4.RP-248 The upper core barrel (UCB) bolts cited in the RP-248 Iitem and the associated bolt locking devices cited in the RP-248a and RP-248b Iitems apply to all Babcock and Wilcox (B&W)-designed PWRs. The UCB bolts are the subject of the RP-248 Iitem for 2-57 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes cracking effect and mechanism combinations. The UCB bolt locking devices are the subject of the RP-248a Iitem for cracking effect and mechanism combinations and the RP-248b Iitem for non-cracking effect and mechanism combinations.

The UCB bolts and UCB bolt locking devices remain as Primary category components for the B&W-design RVI management programs per Item B7 in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B7 Iitem, the EPRI MRP screened the UCB bolts in for stress corrosion cracking (SCC) and additionally for fatigue if the bolts had yet to be replaced at the plants. The EPRI MRP screened the corresponding UCB bolt locking devices in for fatigue and wear mechanisms. For simplicity, the modified version of the RP-248 Iitem assumes that SCC and fatigue apply generically, regardless of whether the UCB bolts have been replaced at the facility. The staff deleted UCB bolt locking devices from the scope of the RP-248 because they are already within the scope of the AMRs in the RP-248a and RP-248b Iitems. The existing versions of the RP-248a and RP-248b Iitems for the UCB bolt locking devices are consistent with the identification of fatigue and wear as applicable mechanisms for the components.

IV.B4.RP-252 The vent valve assembly (VVA) top and bottom retaining rings cited in the RP-252 Iitem applies to all Babcock and Wilcox (B&W)-designed PWRs.

The vent valve assemblyVVA top and bottom retaining rings are identified as Primary category components for B&W-design RVI management programs per Items B3.a and B3.b in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B3.a and B3.b Iitems, the EPRI MRP screened the retaining rings in for thermal aging embrittlement (TE) as the applicable aging mechanism. The existing version of the RP-252 Iitem is consistent with the component nomenclature and citing of TE as an applicable non-cracking mechanism in Items B3.a and B3.b of Table 4-1, MRP-227, Rev.Revision 1-A.

IV.B4.RP-256 The flow distributor (FD) bolts cited in the RP-256 Iitem and the IV.B4.RP-256a associated bolt locking devices cited in the RP-256a and RP-IV.B4.RP-256b 256b Iitems apply to all Babcock and Wilcox (B&W)-designed PWRs. The FD bolts are the subject of the RP-256 Iitem for cracking effect and mechanism combinations. The FD bolt locking devices are the subject of the RP-256a Iitem for cracking effect and mechanism combinations and the RP-256b Iitem for non-cracking effect and mechanism combinations.

The FD bolts and FD bolt locking devices remain as Primary category components for the B&W-design RVI management programs per Item B12 in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B12 Iitem, the EPRI MRP screened the FD bolts in for stress corrosion cracking (SCC) and fatigue cracking mechanisms and the corresponding FD bolt locking devices in for fatigue and wear mechanisms. The staff 2-58 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes edits of the RP-256 Iitem make it consistent with the SCC and fatigue mechanisms attributed to the FD bolts. The existing versions of the RP-256a and RP-256b Iitems for the FD bolt locking devices are consistent with the identification of fatigue and wear as applicable mechanisms for the components.

IV.B4.RP-259 The incore monitoring (IMI) guide tube spider-to-lower grid rib section welds cited in the RP-259 Iitem apply to all Babcock and Wilcox (B&W)-designed PWRs. These welds are the subject of the RP-259 Iitem for non-cracking effect and mechanism combinations.

The IMI guide tube spider-to-lower grid rib section welds were identified as Primary category components for B&W-design RVI management programs per Item B15 in Table 4-1 of the MRP-227, Rev.Revision 1-A. In the B15 Iitem, the EPRI MPR screened the welds in for neutron irradiation embrittlement (IE) aging mechanisms.

However, in Table 3-1 of MRP-227, Rev.Revision 1-A, the EPRI MRP identifies that the applicable spider-to-lower grid rib sections welds are made from Type 308L stainless steel weld filler metals.

Therefore, the staff deleted nickel alloy welds as a listed weld filler metal type for the IMI guide tube spider-to-lower grid rib section welds. The staff also deleted thermal aging embrittlement (TE) as a listed embrittlement mechanism in order to make the aging mechanisms consistent with those listed in Item B15 of Table 4-1 in MRP-227, Rev.Revision 1-A.

IV.B4.RP-262 As cited in Item B13.2 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the dowel-to-lower grid fuel assembly support pad welds cited in the RP-262 Iitem applies to all Babcock and Wilcox (B&W)-designed PWRs, except for the PWR at the Davis- Besse Nuclear Plantnuclear plant. The support pad welds are the subject of the RP-262 Iitem for cracking effect and mechanism combinations.

The dowel-to-lower grid fuel assembly support pad welds were identified as Expansion category components for the B&W-design RVI management programs at these plants per Item B13.2 in Table 4-4 of the MRP-227, Rev.Revision 1-A. In the B13.2 Iitem, the EPRI MPR screened the support pad welds in for the stress corrosion cracking aging mechanism. The edited version of the RP-262 Iitem is consistent with the cited aging mechanism basis.

If the owner of the Davis- Besse Nuclear Plantnuclear plant opts to submit a subsequent license renewal applicationSLRA for its facility, the owner may use GALL-SLR Item IV.B4.R-423 as the alternate AMR line item for aligning to the alternate, unit-specific dowel-to-lower grid fuel assembly support pad configuration basis called out by MRP-227, Rev.Revision 1-A, Table 4-4, Note 2.

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Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes IV.B4.RP-261 The dowel-to-guide block welds cited in the modified version of the RP-261 item applies to all Babcock and Wilcox (B&W)-designed PWRs, except for the PWR at the Davis- Besse Nuclear Plantnuclear plant. The support pad welds are the subject of the RP-261 Iitem for cracking effect and mechanism combinations.

The dowel-to-guide block welds were identified as Primary category components for the B&W-design RVI management programs at these plants per Item B13 in Table 4-1 of the MRP-227, Rev.Revision 1-A. In the B13 item, the EPRI MPR screened the support pad welds in for the stress corrosion cracking aging mechanism. The modified version of the RP-261 Iitem is consistent with the cited aging mechanism basis.

If the owner of the Davis- Besse Nuclear Plantnuclear plant opts to submit a subsequent license renewal applicationSLRA for its facility, the owner may use GALL-SLR Item IV.B4.R-423 as the alternate AMR line item for aligning to the alternate, unit-specific dowel-to-guide block configuration basis called out by MRP-227, Rev.Revision 1-A, Table 4-1, Note 9.

IV.B4.RP-246 In the MRP-227, Rev.Revision 1-! -A Report, the EPRI MRP IV.B4.RP-246a established the upper thermal shield (UTS) bolts are located in the IV.B4.RP-246b core barrel assemblies of Babcock and Wilcox (B&W)-designed PWRs and that the lower thermal shield (LTS) bolts are located in the lower grid assembly of B&W-designed. This required the staff to delete the UTS bolts from the scope of the RP-246 Iitem and to delete the associated UTS bolt locking devices from the scope of the RP-246a and RP-246b Iitems.

The UTS bolts are now covered by the new GALL-SLR IV.B4.RP-246c Iitem, and the associated UTS bolt locking devices are now covered by the new GALL-SLR IV.B4.RP-246d and IV.B4.RP-246e Iitems.

This leaves the LTS bolts as the cited components in the modified RP-246 Iitem and the associated LTS bolt locking devices as the cited components in the modified RP-246a and RP-246b Iitems.

The LTS bolts are the subject of the RP-246 Iitem for cracking effect and mechanism combinations. The LTS bolt locking devices are the subject of the RP-246a Iitem for cracking effect and mechanism combinations and the RP-246b Iitem for non-cracking effect and mechanism combinations. For simplicity of the modified version of RP-246, the staff considers the term studs/nuts to be synonymous with the term bolts, so studs/nuts are not referenced in the RP-246 Iitem.

The LTS bolts and LTS bolt locking devices remain as Expansion category components for the B&W-design RVI management programs per Item B8.1 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B8.1 Iitem, the EPRI MRP screened the LTS bolts in for the stress corrosion cracking (SCC) aging mechanism, and the associated locking devices in for 2-60 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes fatigue and wear mechanisms. The modified version of the RP-246 Iitem is consistent cited cracking mechanism of SCC. The existing versions of the RP-246a and RP-246b Iitems for the LTS bolt locking devices are consistent with the identification of fatigue and wear as applicable mechanisms for the components.

IV.B4.RP-260 The lower grid assembly pads, pad-to-rib section welds, dowels, IV.B4.RP-260a cap screws and associated locking devices cited in the RP-260 and RP-260a Iitems apply to all Babcock and Wilcox (B&W)-

designed PWRs. The RP-260a Iitems addresses cracking effect mechanism combinations in the components and the RP-260 Iitem addresses non-cracking effect and mechanism combinations in the components.

The lower grid assembly pads, pad-to-rib section welds, dowels, cap screws and associated locking devices are identified as Expansion category components for B&W-design RVI management programs per Item B15.1 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In the B15.1 Iitem, the EPRI MRP screened the components in for cracking and for neutron irradiation embrittlement (IE). The edited version of the RP-260 Iitem is consistent with the designation of IE as a listed non-cracking mechanism. The B15.1 Iitem in the MRP-227, Rev.

1-A reportMRP-227, Revision 1-A Report did not specify any specific mechanisms for inducing cracking of the components. The staff assumes that stress corrosion cracking (SCC) and fatigue are the mechanisms that may induce potential cracking of the components, as referenced in the RP-260a Iitem.

The material columns in the RP-260 and RP-260a Iitems already acknowledge that the components could be fabricated usingfrom nickel-based alloy materials, so there is no reason to reference X-750 as a potential nickel alloy material in the component descriptions of the line items.

IV.B4.RP-251a The plenum cover weldment rib pads, plenum cover support flanges, and plenum cover support rings cited in the RP-251a Iitem apply to all Babcock and Wilcox (B&W)-designed PWRs. The RP-251a iitems addresses non-cracking effect and mechanism combinations in the components.

The plenum cover weldment rib pads, plenum cover support flanges, and plenum cover support rings are identified as Primary category components for B&W-design RVI management programs per Items B1.a, B1.b, and B1.c in Table 4-1 of the MRP-227, Rev.

1-A reportMRP-227, Revision 1-A Report. In the B1.a. B1.b, and B1.c Iitems, the EPRI MRP screened the components in for loss or material due to wear and loss of preload due to wear as the applicable non-cracking aging effect and mechanism combinations for the components. The modified version of the RP-251a is consistent with the bases in the B1.a. B1.b, and B1.c Iitems.

IV.B4.RP-352 The dowel-to-upper grid fuel assembly support pad welds cited in the edited version of the RP-352 Iitem applies to all Babcock and 2-61 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes Wilcox (B&W)-designed PWRs, except for the PWR at the Davis-Besse Nuclear Plantnuclear plant. The support pad welds are the subject of the RP-352 Iitem for cracking effect and mechanism combinations.

The dowel-to-upper grid fuel assembly support pad welds are identified as Expansion category components for the B&W-design RVI management programs at these plants per Item B13.1 in Table 4-4 of the MRP-227, Rev.Revision 1-A. In the B13.1 item, the EPRI MPR screened the support pad welds in for the stress corrosion cracking aging (SCC) mechanism. The edited version of the RP-352 Iitem is consistent with the cited aging mechanism basis.

If the owner of the Davis- Besse Nuclear Plantnuclear plant opts to submit a subsequent license renewal application for its facility, the owner may use GALL-SLR Item IV.B4.R-423 as the alternate AMR line item for aligning to the alternate, unit-specific dowel-to-upper grid fuel assembly support pad configuration basis called out by MRP-227, Rev.Revision 1-A, Table 4-4, Note 2.

IV.B2.R-423 The previous versions of the R-423 and R-424 Iitems were IV.B2.R-424 extremely restrictive in that they could only be applied for cases IV.B3.R-423 where a SLR applicant was applying a plant-specific aging IV.B3.R-424 management program for its PWR RVI components. The IV.B4.R-423 modifications of the R-423 and R-424 Iitems now allow the line IV.B4.R-424 items to be applied for additional cases, and specifically for cases where the applicant is using its GALL-SLR XI.M16A-based AMP as its program, but where the referenced MRP-227, Rev.Revision 1-A protocols for a specified component are being adjusted based on site-specific or component-specific considerations. This will broaden the scope of the R-423 and R-424 Iitems so that they can be more readily applied and used in applicable subsequent license renewal applications.

For more information, refer to the technical basis statement in Appendix B.3 of SLR-ISG-2021-01-PWRVI for the staffs analogous changes proposed to SRP-SLR Table 3.1-1, Items 118 and 119.

IV.B4.RP-258 The incore monitoring instrument (IMI) guide tube spiders cited in the RP-258 Iitem apply to all Babcock and Wilcox (B&W)-

designed PWRs. The IMI guide tube spiders are the subject of the RP-258 Iitem for non-cracking effect and mechanism combinations.

The IMI guide tube spiders were identified as Primary category components for B&W-design RVI management programs per Item B15 in Table 4-1 of the MRP-227, Rev.Revision 1-A. In the B15 iItem, the EPRI MPR screened the IMI guide tube spiders in for cracking and for neutron irradiation embrittlement (IE) and thermal aging embrittlement (TE) aging mechanisms. The existing version of the RP-258 Iitem is consistent with the referencing of IE and 2-62 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes TE as the applicable non-cracking mechanisms for the IMI guide tube spider components.

IV.A2.RP-154 Change for stainless steel (SS) bottom-mounted instrument BMI guide tubes exposed to reactor coolant, IV.A2.RP154, was to remove the plant- specific aging management programAMP. This line item is just the SS portion of the bottom mounted instrumentBMI guide tubes external to the bottom head. For other SS materials in the primary circuit with the concern for cracking due to primary water stress corrosion cracking (SCC), the application of AMP XI.M1 ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD, along with XI.M2 Water Chemistry has been shown to be adequate to address this aging mechanism, primary water SCC, for this material, SS. Therefore, the update to reference these AMPs is now recommended.

IV.C2.R-05 The staff re-evaluated the guidance provided in Section 3.1.2.2.6, Item 2 which states Further evaluation is recommended of a plant-specific program for these components to ensure that this aging effect is adequately managed and Section 3.1.3.2.6, Item 2 which states that A plant-specific AMP should be evaluated to manage cracking due to SCC in CASS PWR Class 1 reactor coolant system piping and piping components exposed to reactor coolant that do not meet the carbon and ferrite content guidelines of NUREG-0313. The guidance in NUREG-0313, Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary, Revision 2, was published on January 1988. As the title suggests it was intended to provide guidance concerning intergranular stress corrosion cracking susceptibility of BWR piping and included guidelines on CASS components. Specifically, it highlighted the potential of SCC for certain CASS components if they did not meet the recommended ferrite and carbon content. While the recommendations in NUREG-0313 are still very relevant to BWRs, current operating experience of CASS components in PWRs does not merit to elevate this AMR item to a Further Evaluation. There is no current operating experience that indicates that this is a problem for CASS components in PWRs that requires further evaluation.

Section 3.1.2.2.6, Item 2 is deleted as referenced NUREG-0313 is applicable to BWRs.

IV.D1.RP-367 The staff added a discussion of plant-specific steam generator IV.D1.RP-358 (SG) design parameters that should be evaluated against the IV.D2.RP-185 industry analyses (EPRI 3002002850) to determine whetherif a given plant is bounded by the industry analyses for SG divider plate cracking. This includes potential use of the checklist in EPRI letter SGMPIL1602 to demonstrate that plant-specific parameters are bound by the industry analyses. This is meant to provide clarity to determine whether the industry analyses are applicable and bounding. Additionally, the reference to a plant-specific AMP was replaced with the One-Time Inspection AMP because the GALL-SLR Rreport states that a plant-specific AMP, may include a Oone-Ttime Iinspection that is capable of detecting cracking to verify the effectiveness of the water chemistry and steam 2-63 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management Review (AMR) Item No. Technical Bases for Changes generator programs and the absence of primary water stress corrosion cracking (PWSCC) in the divider plate assemblies. The One-Time Inspection AMP fulfills this recommendation from the GALL-SLR and eliminates the need for a plant-specific AMP to be evaluated.

IV.E.R-444 Added reactor interior attachments to list of examples to be consistent with PWR reactor internal components that are defined as ASMES Section XI Class 1 interior attachments to the reactor vessel.

1 Table 2-22 Table 2-18 Changes to GALL-SLR Report, Revision 0, Chapter V AMRAging 2 Management Review Items and Technical Bases Aging Management ReviewAMR Item No. Technical Bases for Changes V.A.E-20 The staff noted that for other material and environment V.D1.E-20 combinations in the GALL- SLR Report, reduction of heat transfer V.D2.EP-74 due to fouling is the only aging effect associated with an intended function of heat transfer. The Water Chemistry Aging Management Program (AMP) can be used to minimize the potential for deposits that can lead to fouling through the control of primary side water chemistry. Additionally, the One- Time Inspection AMP will help to verify the effectiveness of the Water Chemistry AMP. The GALL- SLR recommends the use of the Water Chemistry and Steam Generator AMPs (Aging Management Review [AMR] Table 1 Iitem 3.1-1, 111) to manage the reduction of heat transfer due to fouling in nickel alloy tubes.

The use of the Water Chemistry and One- Time Inspection AMPs provide an analogous approach (i.e., water chemistry control and an inspection to verify effectiveness) to managing the reduction of heat transfer on primary side nickel alloy heat exchanger tubes.

The staffs review of the Turkey Point subsequent license renewal application (SLRA) demonstrates that stainless steel (SS) and nickel alloy have similar aging effects when exposed to treated borated water. The GALL- SLR recommends the use of the Water Chemistry and One- Time Inspection AMPs to manage the reduction of heat transfer in stainless steel heat exchanger tubes.

Because stainless steelSS and nickel alloy experience similar aging effects it is reasonable to use the same AMPs to manage the aging effects in nickel alloy materials.

V.A.E-475 Subsequent to issuance of the GALL-SLR Report, the staff V.D1.E-475 recognized that to be consistent with other GALL-SLR Report V.D2.E-475 items associated with heat exchanger tubes, E-475 should have also cited reduction of heat transfer due to fouling. This is consistent with GALL Report Revision 2 item SP-41 where a material (i.e., stainless steel) that is not susceptible to loss of material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

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Draft Document: Tracked Changes Version Aging Management ReviewAMR Item No. Technical Bases for Changes Titanium components are subject to flow blockage due to fouling due to potential debris in the raw water environment.

V.A.E-415 During its review of recent SLRA plant-specific operating V.B.EP-54 experience, in response to the staffs observation regarding dark V.B.E-415 corrosion product layers indicative of graphitic corrosion on the V.C.E-415 internal surfaces of malleable iron fittings exposed to a closed-V.D1.EP-52 cycle cooling water environment (ADAMS Accession No.

V.D1.EP-54 ML22010A129), the staff has revised the guidance documents V.D1.E-415 (i.e., GALL-SLR Report and SRP-SLR) to include malleable iron V.D2.EP-54 as a material susceptible to selective leaching.

V.D2.E-415 V.A.E-401 The staff has accepted opportunistic inspections, in lieu of periodic V.B.E-401 inspections, as an acceptable alternative for buried internally V.D1.E-401 coated/lined fire water system piping provided: (a) flow tests and V.D1.E-414 internal piping inspections will occur at intervals specified in V.D2.E-401 National Fire Protection Association (NFPA 25), or as modified by AMP XI.M27, Table XI.M27-1; and (b) through-wall flaws in the V.D2.E-414 piping can be detected through continuous system pressure monitoring. Examples of the staffs acceptance of this alternative approach are documented in the Safety Evaluation Report Related to the License Renewal of Fermi 2 Nuclear Power Plant (ADAMS Accession No. ML16190A241) and the Safety Evaluation Report Related to the Subsequent License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3 (ADAMS Accession No. ML20044D902). Based on recent OE involving ruptures ofin the buried fire water system piping system due to age-related degradation (ADAMS Accession No. ML19294A044), the staff added a third condition for using this alternative approach related to plant- specific operating experience (OE). The staff notes that the subject OE involved degradation of the external surfaces of the piping; however, degradation of internal coatings/linings could also result in significant degradation of buried fire water system piping.

The GALL-SLR Report discusses the reason for citing specific AMPs to manage recurring internal corrosion rather than a plant-specific AMP in the section titled Explanation of the Use of Multiple Aging Management Programs in Aging Management Review Items. For the associated AMR item in the SRP-SLR (Iitem 3.3-1-127), the listed environments still include closed-cycle cooling water even though NUREG-2221, Technical Bases for Changes in the Subsequent License Renewal Guidance Documents NUREG-2191 and NUREG-2192, Table 2-13, notes that the associated item in Table C2, Closed-Cycle Cooling Water System, was deleted because recurring internal corrosion is not anticipated in this system. These changes corrects this error in conjunction with the adjustments above for the use of multiple AMPs.

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Draft Document: Tracked Changes Version 1 Table 2-23 Table 2-19 Changes to GALL-SLR Report, Revision 0, Chapter VI Aging 2 Management ReviewAMR Items and Technical Bases Aging Management Review (AMR)

Item No. Technical Bases for Changes VI.A.LP-28 The AMR item is modified to incorporate industry operating experience to enhance aging management of high voltage insulators to include additional susceptible materials (toughened glass; polymers silicone rubber; fiberglass, aluminum alloy) and aging effects (peeling of silicone rubber sleeves for polymer insulators; or glazing degradation for porcelain insulators).

VI.A.LP-32 The AMR item is modified to incorporate industry operating experience to enhance aging management of high voltage insulators to include additional susceptible materials (toughened glass; polymers silicone rubber; fiberglass, aluminum alloy).

No AMR items were changed from Chapter VI of GALL-SLR Report, Revision 0.

3 Table 2-24 Table 2-20 Changes to GALL-SLR Report, Revision 0, Chapter VII Aging 4 Management ReviewAMR Items Technical Bases Aging Management ReviewAMR Item No. Technical Bases for Changes Item No. VII.G.A-789 The sStructure and/or cComponent was changed from fire damper assemblies to fire damper housing because the housing is the passive component of the fire damper assembly that is subject to aging management. The applicable material was revised to metallic because fire damper housings are typically constructed of steel or stainless steel (SS). The applicable aging effects were revised to loss of material due to general, pitting, and crevice corrosion, and cracking due to stress corrosion cracking (SCC) because the elastomer aging effects of hardening, loss of strength, and shrinkage do not apply to metallic components. The fire damper housing is potentially subject to the cited aging effects. For example, steel materials would not be subject to SCC; however, stainless steel SS materials would be subject to SCC. The periodic inspections recommended by GALL-SLR Aging Management Program (AMP) Report XI.M26 are capable of detecting these aging effects.

VII.C1.A-787a Based on a review of current subsequent license renewal VII.C1.A-787c applications (SLRAs), the staff noted that an applicant had cited VII.G.A-787b a polyvinyl chloride (PVC) piping component (chemical addition VII.E5.A-787d tank) in its essential service water system. The tank is internally exposed to treated water. In Revision 0 of the GALL- SLR Report, Iitem A 787b is the only AMR item citing PVC piping components exposed to treated water. This item cites AMP XI.M27, Fire Water System, to manage loss of material due to wear and flow blockage due to fouling (raw water only). Given that the applicants tank was not located in the fire water system, AMP XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components, was cited as the applicable AMP in lieu of AMP XI.M27. To eliminate the potential for future 2-66 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Aging Management ReviewAMR Item No. Technical Bases for Changes generic note E items (different AMP than that cited in the GALL-SLR Report), Iitems A 787a and A 787c were revised to include treated water as an applicable environment. The staff concluded that the addition of treated water to these two AMR items is acceptable because the treated water environment is less aggressive than the existing raw water environment. As a result, the cited AMPs will be adequate to manage loss of material.

Items A 787a and A 787c were further revised to state that flow blockage due to fouling is only applicable to the raw water environment as stated in SRP- SLR Iitem 3.3.1 253.

In the process of incorporating this change, the staff noted an error in SRP- SLR Iitem 3.3.1 253. Based on the potential for accumulation of potential particles in the wastewater water flow stream, PVC piping and piping components are susceptible to flow blockage due to fouling in the wastewater water environment. This is consistent with Table 2- 20, Changes to Existing GALL Report Revision 2 Chapter VII AMR Items Technical Bases, Iitem AP 269.

The staff further noted that the bases for AP 269 states that based on The PVC Pipe - Design and Installation - Manual of Water Supply Practices, M23, American Water Works Association, 2nd Edition, 2002, PVC pipe is well suited to applications where abrasive conditions are anticipated. The staff concluded that it is reasonable to conclude that loss of material due to wear would not occur due to abrasive particle impingement or flow perturbations in low flow applications. The aging effects requiring management for A 787a, A 787b, A 787c, and A 787d were changed accordingly.

1 Table 2-25 Table 2-21 Changes to GALL-SLR Report, Revision 0, Chapter VIII Aging 2 Management ReviewAMR Items and Technical Bases Aging Management ReviewAMR Item No. Technical Bases for Changes VIII.A.SP-28 During its review of recent subsequent license renewal VIII.A.SP-27 application (SLRA) plant-specific operating experience (OE), in VIII.E.SP-26 response to the staffs observation regarding dark corrosion VIII.E.SP-27 product layers indicative of graphitic corrosion on the internal VIII.E.S-415 surfaces of malleable iron fittings exposed to a closed-cycle VIII.F.SP-27 cooling water environment (ADAMS Accession No.

VIII.F.S-415 ML22010A129), the staff has revised guidance documents (i.e.,

VIII.G.SP-28 GALL-SLR and Standard Review Plan for Review of Subsequent VIII.G.SP-26 License Renewal Applications for Nuclear Power Plants [SRP-VIII.G.SP-27 SLR] Report) to include malleable iron as a material susceptible VIII.G.S-415 to selective leaching.

VIII.D1.S-482 Subsequent to issuance of the GALL-SLR Report, the staff VIII.D2.S-482 recognized that to be consistent with other GALL-SLR Report VIII.E.S-482 items associated with heat exchanger tubes, E-475 should have VIII.F.S-482 also cited reduction of heat transfer due to fouling. This is consistent with GALL Report Revision 2 Iitem SP-41 where a material (i.e., stainless steel [SS]) that is not susceptible to loss of 2-67 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

Titanium components are subject to flow blockage due to fouling due to potential debris in the raw water environment.

No aging management reviewAMR items were changed from Chapter VIII of GALL-SLR Revision 0.

1 Table 2-26 Table 2-22 Chapter IX.B - Structures and Components, Differences From 2 Chapter IX GALL-SLR Report, Revision 0, and Their Technical Bases Defined Term Summary of Significant Changes Technical Basis for Change Eelectrical Added examples of organic polymers for Added examples of organic polymers:

insulation clarity. ethylene propylene rubber, silicone rubber, ethylene propylene diene monomer (EPDM), and cross-linked polyethylene.

Eexisting Updated the referenced EPRI Report Updated the referenced EPRI report to programs from EPRI Report No. 1022863 (MRP- reflect the most recent revision that has components 227-A) to EPRI Report No. 3002017168 been approved by NRC staff.

(MRP-227, Revision 1-A).

Eexpansion components No Additional Measures components Primary components 3 Table 2-27 Table 2-23 Chapter IX.C - Materials, Differences From Chapter IX GALL-4 SLR Report, Revision 0, and Their Technical Bases Defined Term Summary of Significant Changes Technical Basis for Change Carbon fiber Added the new term. The new term was added to support the reinforced polymer new AMP, GALL-SLR Report AMP (CFRP) XI.M43, High Density Polyethylene (HDPE) Piping and Carbon Fiber Reinforced Polymer (CFRP) Repaired Piping. The technical basis for this new AMP can be found in Table 2-29Table 2-29, XI.M43 of this report.

Malleable Iron Revised term: (a) to clarify the material The material properties description for properties of malleable iron; and (b) this material was expanded to be similar based on its inclusion to GALL- SLR to the existing definition of ductile iron.

Report AMP XI.M33, Selective Leaching. The staffs basis for the inclusion of malleable iron as a material susceptible to selective leaching is documented in Table 2-29, GALL-SLR Differences from Chapter XI, Mechanical Aging 2-68 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Defined Term Summary of Significant Changes Technical Basis for Change Management Programs, GALL Report Revision 2 and Their Technical Bases.

Stellite Added the new Mmaterial terminology In the subsequent license renewal to GALL-SLR Table IX.C in SLR-ISG- application (SLRA) for Surry Nuclear 2021-01-PWRVI due to reference of Station, Units 1 and 2, the applicant stellite in the aging management identified that certain reactor vessel reviewAMR line items for specific types (RV) components made from stainless of pressurized water reactor (PWR) steel or nickel alloy materials (e.g.,

reactor vessel internal (RVI) clevis inserts or fuel alignment pins) components. were fabricated with stellite surface layers to make the components more resistant to wear. Reference of stellite was not previously included in Table IX.C of the GALL-SLR Rreport, but the material is referenced as a cited material in MRP-227, Revision 1-A reportRevision 1-A Report. The addition of stellite makes the contents of GALL-SLR Table IX.C up- to- date with materials referenced for PWR reactor vessel internalRVI components in the MRP-227, Revision 1-A reportRevision 1-A Report.

The staff has adopted a definition for stellite by ASTM International.

Vvarious Included carbon fiber reinforced polymer The change was added to support the polymeric in the list of examples of polymers used new AMP, GALL-SLR Report AMP materials in mechanical applications that are XI.M43, High Density Polyethylene addressed as specific material types. (HDPE) Piping and Carbon Fiber Reinforced Polymer (CFRP) Repaired Piping. The technical basis for this new AMP can be found in Table 2-29Table 2-29, XI.M43 of this report.

1 Table 2-28 Table 2-24 Chapter IX.D - Environments, Differences From Chapter IX 2 GALL-SLR, Report, Revision 0, and Their Technical Bases Defined Term Summary of Significant Changes Technical Basis for Change No differences from Chapter IX of GALL-SLR Report, Revision 0.

3 Table 2-29 Table 2-25 Chapter IX.E - Aging Effects, Differences From Chapter IX 4 GALL-SLR Report, Revision 0, and Their Technical Bases Defined Term Summary of Significant Changes Technical Basis for Change Crack growth Added that increase in crack size can Clarification that crack size can also be attributed to static loading. increase to static loading in addition to cyclic loading.

Cracking Added additional context to the use of These changes were added to support the term as it relates to polymeric the new AMP, GALL-SLR Report AMP materials and carbon fiber reinforced XI.M43, High Density Polyethylene polymer (CFRP) piping. (HDPE) Piping and Carbon Fiber 2-69 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Defined Term Summary of Significant Changes Technical Basis for Change Reinforced Polymer (CFRP) Repaired Piping. The technical basis for this new AMP can be found in Table 2-29Table 2-29, XI.M43 of this report.

Flow Blockage Added clarification that accumulations These changes were added to support of particulate fouling, biofouling, or the new AMP, GALL-SLR Report AMP macro fouling also includes XI.M43, High Density Polyethylene delamination/disbanding of CFRP (HDPE) Piping and Carbon Fiber repaired piping. Reinforced Polymer (CFRP) Repaired Piping. The technical basis for this new AMP can be found in Table 2-29Table 2-29, XI.M43 of this report.

Loss of Material Added additional context to the use of These changes were added to support the term as it relates to polymeric the new AMP, GALL-SLR Report AMP materials. XI.M43, High Density Polyethylene (HDPE) Piping and Carbon Fiber Reinforced Polymer (CFRP) Repaired Piping. The technical basis for this new AMP can be found in Table 2-29Table 2-29, XI.M43 of this report.

1 Table 2-30 Table 2-26 Chapter IX.F - Aging Mechanisms, Differences from Chapter IX 2 GALL-SLR Report, Revision 0, and Their Technical Bases Defined Term Summary of Significant Changes Technical Basis for Change Crevice Corrosion Clarified that crevice corrosion can result The previous discussion did not from situations beyond those with sufficiently explain the variety of dissimilar materials or designed circumstances where crevice corrosion crevices. could occur. Clarification includes reference to a new term differential aeration corrosion.

Differential Added this new term as an overarching The staffs reviews of previous operating Aeration corrosion mechanism that applies to experience identified this mechanism as Corrosion crevice corrosion and configurations being associated with corrosion in where varying oxygen concentrations various air-to-water and soil-to-air across a component can lead to interfaces.

accelerated corrosion.

Wear Added additional context to the use of These changes were added to support the term as it relates to carbon fiber the new AMP, GALL-SLR Report AMP reinforced polymer (CFRP). XI.M43, High Density Polyethylene (HDPE) Piping and Carbon Fiber Reinforced Polymer (CFRP) Repaired Piping. The technical basis for this new AMP can be found in Table 2-29Table 2-29, XI.M43 of this report.

No differences from Chapter IX of GALL-SLR Report, Revision 0.

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Draft Document: Tracked Changes Version 1 Table 2-31 Table 2-27 Chapter IX.G - References, Differences From Chapter IX GALL-2 SLR Report, Revision 0, and Their Technical Bases Defined Term Summary of Significant Changes Technical Basis for Change N/A Added Reference 43: EPRI Technical Reference added to support changes to Report No. 3002017168, Materials terms: Eexisting programs Reliability Program: Pressurized Water components; Eexpansion components; Reactor Internals Inspection and No Additional Measures components; Evaluation Guidelines (MRP-227, Primary components in Table 2-22 of Revision 1-A). this document.

No new references for Chapter IX of GALL-SLR Report, Revision 0.

3 Table 2-32 Table 2-28 GALL-SLR Report, Revision 1, Chapter X, Time-Limited Aging 4 AnalysesTLAAs, Differences From GALL-SLR Report, Revision 0, and Their 5 Technical Bases Location of Change Summary of Significant Changes Technical Basis for Change X.M2: Neutron Fluence Monitoring Monitoring and Aging management program (AMP) The added references to this AMP Trending X.M2 is revised to reference provide examples of acceptable approaches that have been found to approaches from recent reviews.

Acceptance Criteria be acceptable in recent staff reviews These examples provided acceptable of extended beltline and reactor justification to apply the methods vessel internals fluence calculations, used for fluence calculations in the as RG 1.190 is not applicable, and traditional reactor vessel beltline, to the U.S. Nuclear Regulatory the extended beltline and to reactor Commission (NRC) staff continues to vessel internal components.

develop regulatory guidance for such calculations.

X.E1: Environmental Qualification of Electric Components AMP X.E1 The AMP program description and The staff determined that the AMP Environmental scope should include mechanical program description and scope should Qualification of components associated with electrical be clarified to include mechanical Electrical equipment, such as gaskets, seals, O components associated with electrical EquipmentProgram rings, etc. equipment to avoid further questions Description and clarify the intent of the program.

Scope of Program 6 Table 2-33 Table 2-29 GALL-SLR Report, Revision 1, Differences from Chapter XI, 7 Mechanical Aging Management Programs, Differences From GALL-SLR 8 Report, Revision 0, and Their Technical Bases Location of Change Summary of Significant Changes Technical Bases for Changes XI-01: Final Safety Analysis ReportFSAR Supplement Summaries Table XI-01, For aging management programs Although the implementation schedules Implementation (AMPsS) XI.M27, XI.M29, XI.M30, for the associated programs stated that Schedule Column XI.M32, XI.M33, XI.M41, and XI.M42, the activities begin X years prior to the clarify that the inspections only begin subsequent period of extended operation, within the specified time period (e.g., 10- the staff had only intended that the years) prior to the subsequent period of activities would be conducted within the 2-71 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes extended operation and not at the specified time period and not necessarily specified time. begin at that time.

XI.M2: Water Chemistry Program AMP XI.M2, Water Chemistry, was EPRI issued 3002010645, Pressurized Description revised to include the latest revision of Water Reactor Secondary Water Electric Power Research Institute (EPRI) Chemistry Guidelines, Revision 8, in References guidelines for boiling water reactor (BWR) 2017 from the previous version and pressurized water reactor (PWR). (1016555). According to EPRI, a committee of industry experts collaborated in reviewing data and generating water-chemistry guidelines, which should be used at all nuclear plants, that has been endorsed by the utility chemistry community. Approved precedent for use of the more recent version of the above guideline is documented in the NRC staffs safety evaluation reportSER for subsequent license renewal of Surry Units 1 and 2 (Agencywide Documents Access Management System [(ADAMS])

Accession No. ML20052F523)

EPRI has issued BWRVIP-190, BWR Water Chemistry Guidelines -

Mandatory, Needed, and Good Practice Guidance. Revision 1. Consistent with the staffs evaluation of an exception documented in NUREG-2205, Safety Evaluation Report Related to the License Renewal of LaSalle County Station, Units 1 and 2, September 2016, Section 3.0.3.2.1, Water Chemistry, the staff finds the use of BWRVIP-190, Revision 1, BWR Vessel and Internals Project, Volume 1, BWR Water Chemistry Guidelines - Mandatory, Needed, and Good Practice Guidance, EPRI 3002002623, dated April 24, 2014, acceptable to cite.

XI.M3: Reactor Head Closure Stud Bolting Preventiveative Item (d) of the Ppreventive Aactions Item (d) of the Ppreventive Aactions Actions program element of GALL-SLR AMP program element describes the material XI.M3 is changed to allow either yield- strength criteria to prevent the Corrective Actions strength criterion (<150 ksi) or ultimate- susceptibility of reactor head closure stud tensile-strength criterion (170 ksi) for materials to SCC (including intergranular Table XI-01 use of low alloy steels resistant to stress stress corrosion crackingIGSCC). These corrosion cracking (SCC). Either of the criteria are defined in terms of material material strength criteria may be used strength thresholds, below which the low regardless of whether existing reactor alloy steels are resistant to SCC.

head closure studs or newly installed Specifically, the program element uses studs are addressed in the preventive the 170-ksi ultimate-tensile-strength 2-72 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes actions. Corresponding changes were criterion (170 ksi) for existing studs and made to the Ccorrective Aactions the 150 -ksi yield-strength criterion (<150 program element and FSAR Ssupplement ksi) for newly installed studs to provide summarizedy in Table XI-01. SCC resistance.

These slightly different criteria are described in the guidance in Regulatory Guide (RG) 1.65, Materials and Inspections for Reactor Vessel Closure Studs, Revision 0 and 1, respectively.

The initial version of the Rregulatory Gguide (Revision 0) used the ultimate-tensile-strength criterion. This criterion was typically used to select the materials of the original studs installed at the start of the plant operation. In recent years, yield-strength-based criteria have been widely used to select alloy steels or other materials resistant to SCC. The approach using a yield-strength criterion for closure studs is described in the more recent revision of RG 1.65 (Revision 1).

As discussed above, Iitem (d) of the Ppreventive Aactions program element reflects the evolution in guidance development over time (i.e., the ultimate-tensile-strength criterion for the existing studs and yield-strength criterion for newly installed studs). The U.S. Nuclear Regulatory Commission (NRC) staff notes that either the yield-strength criterion or ultimate-tensile-strength criterion is sufficient to select SSC-resistant low alloy steels for the aging management. Therefore, changes are made to Iitem (d) to allow either of the criteria for the selection of SCC-resistant stud materials. Accordingly, the Corrective Actions program element and Final Safety Analysis Report (FSAR)

Ssupplement (GALL-SLR Table XI-01) are revised. The NRC staff expect these changes will reduce the issuance of unnecessary request for additional information (RAIs).

XI.M12: Thermal Aging Embrittlement of Cast Austenitic Stainless Steel (CASS)

Program The program description and the Scope The NUREG/CR-4513, Revision 2 was Description oOf Program and Acceptance Criteria published in 2016 to provide the updated program elements in GALL-SLR AMP screening criteria and fracture toughness Scope of Program XI.M12 are changed to reference the (FT) estimation methods for cast-screening criteria and fracture toughness austenitic stainless steel (CASS) 2-73 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes Acceptance Criteria estimation methods in NUREG/CR-4513, materials that are susceptible to thermal Revision 2 with errata (March 2021). The aging embrittlement [1]. Subsequently, references section and FSAR typographical errors in NUREG/CR-4513, Ssupplement (GALL-SLR XI-01) are were Revision 2 were corrected in the errata also updated. dated March 15, 2021 [2]. The screening criteria and FT estimation methods in NUREG/CR-4513, Revision 2 with errata are consistent with those in NUREG/CR-4513, Revision 2 published in 2016.

The updated criteria and methods are based on the evaluation of additional CASS materials with a ferrite content up to 40 percent. The maximum ferrite content of 40 percent evaluated in NUREG/CR-4513, Revision 2 is an extension from the maximum ferrite content of 25 percent evaluated in Revision 1 of the NUREG report.

Therefore, changes are made to GALL-SLR AMP XI.M12 to reference and use the updated screening criteria and FT estimation methods for CASS materials.

XI.M9: BWR Vessel Internals Scope of Program Aging management programThe AMP The AMP XI.M9 describes an aging XI.M9 was updated to standardize cited management programAMP acceptable to Monitoring & Boiling Water Reactor Vessel and the staff for boiling water reactor (BWR)

Trending Internals Project (BWRVIP) references reactor vessel internals (RVIs). BWR and reflect recent developments in the RVIs are currently age managed through Operating BWRVIP guidance. The AMP XI.M9 was a series of BWR Vessels and Internals Experience updated to include Code Case N-889 for Project (BWRVIP) guidance documents, calculating irradiation- assisted crack many of which have been reviewed and References growth rates. approved by the NRC. This update to AMP XI.M9 corrects, updates, and standardizes BWRVIP document references in the text. This update also reflects two recent developments in BWRVIP guidance. First, the fracture toughness and flaw evaluation guidance in BWRVIP-100, Revision 1-A is currently being updated due to recent data collected in material harvesting programs. Therefore, the NRC staff added a description of the potential changes to BWRVIP guidance and how subsequent license renewal applicants should respond. Second, the BWRVIP-315 topical report was in an advanced stage of NRC review at the time of this proposed GALL-SLR update. The NRC staff added a reference to BWRVIP-315 under Scope of the Program and 2-74 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes provided instructions to subsequent license renewal applicantsSLRAs concerning limitations and applicant action items. The staff removed reference to low alloy steels in Section 5 of XI.M9, since there are no low alloy steel BWR RVI components. Also in Section 5, the NRC staff added a reference to Code Case N-889 and provided guidance to subsequent license renewal applicantsSLRAs for its use.

Finally, the staff added a brief description of recent top guide cracking operating experience (OE) in Section 10 of XI.M9.

XI.M16A : PWR Vessel Internals Program The staff modified the program The change in the program description is Description description to indicate that PWR vessel consistent with the staffs assumption internals programs will be based on the that, by the time a PWR- designed updated inspection and evaluation (I&E) nuclear plant will have entered into the guidelines in EPRI Report No. subsequent period of extended operation 3002017168 (MRP-227, Revision 1-A). for plant, the licensee will have converted its PWR vessel internals program over to Because MRP-227, Revision 1-A, the updated program defined in the MRP-represents aging management for 60 227, Revision 1-A reportRevision 1-A years of plant operation, the staff clarified Report.

that a gap analysis of the reactor internals is needed if the AMP is based on MRP- In Section 7 of the MRP-227, Revision 1-227, Revision 1-A as a starting point for A reportRevision 1-A Report, the EPRI the AMP that will be applied during the MRP calls for the AMPs to be converted subsequent period of extended operation. over to the I&E guidelines in MRP-227, Revision 1-A by January 1, 2022. Thus, The staff amended the program for those licensees that decide to submit description to clarify that programs for SLRAs of their PWRs, the licensees will Westinghouse and Combustion have converted their PWR vessel Engineering (CE)- designed PWRs group internals programs over to MRP-227, the RVI components group the Revision 1-A before the plants enter into components into either Primary, the subsequent period of extended Expansion, Existing Program, and No operation.

Additional Measures inspection categories and that the associated Like the preceding guidelines in EPRI programs for B&WBabcock and Wilcox- Report No. 1022863 (i.e., MRP-227-A),

designed designed PWRs only group the the EPRI MRPs I&E guidelines in MRP-RVI components into Primary, 227, Revision 1-A only assessed the Expansion or No Additional Measures PWR RVI components for Westinghouse, categories. The staff also amended the CE and B&W- designed plants for program description to include a sentence operating cycles and neutron fluence that describes the types of analyses that exposures over a cumulative 60-year are used for the EPRI MRPs integrated service life. The methods in MRP-227, sample-based selection process. Revision 1-A do not account for the potential impact that additional cycles and fluence imparted during a subsequent 2-75 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes The staff defined that, in the updated period of extended operation would have version of AMP XI.M16A, MRP-227 (as on the I&E protocols defined for the supplemented) refers to MRP-227, components in the MRP-227, Revision 1-Revision 1-A criteria as supplemented by A reportRevision 1-A Report. Although a gap analysis. This definition applies reports like MRP-2018-022 may be used through the revised AMP. to assess those impacts, the reports have yet to be docketed with the NRC or endorsed for use by the staff. Thus, gap analyses will still be needed if the applicants program is based on MRP-227, Revision 1-A as a starting point for the AMP.

PWR vessel internals AMPsS for B&W-designed PWRs do not include Existing Program inspection categories.

Scope of Program The staff updated the Sscope of The first significant change to the Scope Pprogram program element to clarify that of Programscope of program element is the program is based on MRP-227 (as consistent with changes made to the supplemented), or on a staff- approved program description that specify a gap generic report that assesses aging over analysis will be needed if MRP-227, Revision 1-A is used as the starting point an 80-year service life.

for the version of the AMP that will be applied during the subsequent period of The staff updated the program element to extended operation. The technical basis include applicable supplemental guidance for coordinating the MRP-227, Revision or reports (e.g., WCAPs, B&W report, 1-A methods with a gap analysis has alert letters) as being within the scope of been given in the previous technical basis the program. statement for the program description of the AMP. As previously clarified in the GALL-SLR version of the AMP, the Scope of Programscope of program element covers the possibility that the industry may develop a generic 80-year report for PWR RVI components that is endorsed by the staff and that it would be perfectly acceptable for the applicant to adopt that report as the basis for the AMP without the need for performing a gap analysis of the components.

The second significant change is designed to eliminate unnecessary enhancements. Although these PWR vessel internals programs are based on the latest staff-endorsed version of MRP-227 I&E guidelines (currently Revision 1-A), the programs may include supplemental guidelines or reports that are endorsed by the NRC. The staff cannot preclude an applicant from including these types of supplemental methods in the scope of its AMP. The 2-76 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes staff would anticipate that the applicant would define any supplemental guidelines (i.e., beyond those in MRP-227, Rev.Revision 1-A) in the Scope of Programscope of program element discussion in the technical basis document for the AMP and provide file copies of the documents containing the methods in its in-house audit portal site directory for the PWR vessel internals AMP. The current SLRA review process would allow the staff to review the supplemental methods as part of the staffs in-office audit review of the AMP.

Parameters The staff made a minor edit of the last The program description for the AMP was Monitored or paragraph in the Pparameters amended to clarify that if MRP-227, Inspected Mmonitored or Iinspected program Rev.Revision 1-A is used as the starting element to indicate that parameters point for the AMP that will be applied monitored or inspected by the program during the subsequent period of extended are based on those defined and operation, the gap analysis being applied established in MRP-227 (as to the components may include methods supplemented). in supplemental guidance or reports. The Sscope of Pprogram was also amended to allow use of supplemental guidelines.

Thus, the change to the Parameters Monitored or Inspectedparameters monitored or inspected element will allow the parameters monitored or inspected by the AMP to be based those in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, or as established in supplemental guidelines, including those referenced in the gap analysis section of the SLRA. For the current status of PWRVI vessel internals programs proposed in PWR SLRAs, those defined in the staff-endorsed MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report should be sufficient, unless the results of the gap analysis demonstrate a need for adjusting the EPRI MRPs I&E criteria for a specified RVI component evaluated in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

Detection of Aging The staff amended the first paragraph of The changes to the specific criteria in the Effects the Ddetection of Aaging Eeffects Ddetection of Aaging Eeffects program program element to indicate that RVI element update the RVI component-component-specific inspections are as specific bases to be consistent with those established in Section 4 of MRP-227 (as defined MRP-227, Rev. 1-A reportMRP-supplemented). 227, Revision 1-A Report or as modified by the results of the applicants gap analysis for a specified RVI component.

2-77 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes The staff amended the second to last paragraph in the program element to indicate that component-specific inspection coverages are established in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, or as amended in the gap analysis for the components.

The staff amended the last paragraph of the program element to indicate that justifications of the appropriateness of component-specific inspection methods should be based on those in the MRP-22, Rev. 1-A reportRevision 1-A Report (and not those that previously defined in the MRP-227-A reportReport).

Otherwise, the criteria in the Ddetection of Aaging Eeffects element remain as previously written and issued in the corresponding program element in AMP XI.M16A of the GALL-SLR reportReport.

Monitoring and The staff amended the first paragraph of The changes to the specific criteria in the Trending the monitoring Monitoring and Ttrending Mmonitoring and Ttrending program program element to specify the methods element update the RVI component-for monitoring, recording, evaluating and specific bases to be consistent with those trending data resulting from the programs defined MRP-227, Rev. 1-A reportMRP-inspections are given in MRP-227 (as 227, Revision 1-A Report or as modified supplemented) and that the inspection by the results of the applicants gap frequencies are established in Section 4 analysis for a specified RVI component.

of MRP-227 (as supplemented).

Otherwise, the criteria in the detection Detection of Aaging effects element remain as previously written and issued in the corresponding program element in AMP XI.M16A of the GALL-SLR reportReport.

Acceptance Criteria The staff amended the Aacceptance The changes account for the fact that the Ccriteria program element in GALL-SLR current component-specific I&E criteria in AMP XI.M16A to indicate that the MRP-227, Rev.Revision 1-A are based component-specific acceptance criteria on an assessment of aging over a are in Table 5-1, 5-2, or 5-3 of Section 5 cumulative 60-year licensed service life of MRP-227, Rev.Revision 1-A or else in and that the acceptance criteria MRP-227 (as supplemented). established in Table 5-1, 5-2, or 5-3 of MRP-227, Rev.Revision 1-A may be The staff also amended the Aacceptance superseded (on a component-specific Ccriteria program element to establish basis) by the results of an 80-year RVI that the acceptance criteria for some gap analysis or by the corresponding Expansion category components may be tables in a version of MRP-227 that established through performance of a covers an 80-year cumulative licensed component-specific analysis, particularly service life.

if the component type is inaccessible to inspection or the industry has yet to 2-78 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes establish an adequate inspection for the Additionally, the staff has already component. approved disposition by component-specific analysis in the staffs April 25, 2019 safety evaluation for MRP-227, Rev.Revision 1-A; however, the prior version of the Aacceptance Ccriterion program element did not account for this possibility. For example, in Items B10.1 and B10.2 of Table 4-4 in MRP-227, Rev.Revision 1-A, the EPRI MRP established the Expansion category B&W core barrel welds and former plates (as linked to the Primary category baffle plates) would be dispositioned by component-specific analysis if the results of primary inspections performed on the baffle plates triggered sample-expansion to the core barrel welds and former plates. Similarly, for Westinghouse baffle-former bolts, the EPRI MRP identifies in Item W6 of Table 5-1, MRP-227, Rev.Revision 1-A, that the acceptance criteria for the ultrasonic bolt inspections is to be established by performance of licensee technical justification for the bolt type. Thus, the staffs amendment of the Aacceptance Ccriteria program accounts for acceptance criteria that are defined and established in a component specific component-specific analysis or component-specific technical justification.

Corrective Actions The staff amended the Ccorrective Similar to analogous changes to other Aactions program element to reference program elements in the AMP, the MRP-227 (as supplemented). Otherwise corrective actions will be established by the Ccorrective Aactions program either those defined in the MRP-227, element remains as previously Revision 1-A reportRevision 1-A Report established and written in the analogous or as modified by the applicants gap program element of AMP XI.16A in the analysis, which may include and establish GALL-SLR Report. supplemental methods and alternative corrective action bases for the components.

Confirmation The staff amended the Cconfirmation The EPRI MRPs latest criteria for Process Pprocess program element to indicate implementing these types of programs in that the implementation criteria for these accordance with the guidance in NEI 03-programs are established of in Section 7 08 are given in Section 7 of the MRP-227, in MRP-227 Rev.Revision 1 or else as Rev. 1-A reportMRP-227, Revision 1-A Report. But since MRP-227, defined in Nuclear Energy Institute (NEI)

Rev.Revision 1-A was based on an 03-08 or other guidance documents, assessment of aging over a 60-year reports, or guidelines that are referenced service life, the minor adjustment of the for the AMP. Otherwise the Cconfirmation Pprocess program Cconformation Pprocess program elements now allows additional implementation and confirmation activity 2-79 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes element remains as previously criteria to be used as part of the established and written in the analogous confirmation and implementation activities program element of AMP XI.16A in the of the program, particularly if they are GALL-SLR Report. defined and called as part of the gap analysis methodology for the AMP (or even a version of MRP-227 that covers an 80-year licensed service life). As an example of this, the use of the additional criteria in EPRIs MRP-2018-022 guidance may provide additional criteria for adjusting the MRP-227, Revision 1-A criteria for CE or Westinghouse components over a cumulative 80-year licensed service life.

Administrative The staff amended the second paragraph The staff confirmed that the EPRI MRPs Controls of the Aadministrative Ccontrols bases for implementing MRP-based program element to clarify that Section 7 PWRVI Vessel Internals pPrograms in of the MRP-227, Revision 1-A accordance with NEI-03-08 is given in reportRevision 1-A Report establishes the Section 7 of the MRP-227, Revision 1-A basis for implementing the PWRVI Vessel reportRevision 1-A Report. The MRP-Internal pProgram in accordance with the industry guidelines set in NEI-03-08. The 227, Revision 1-A reportRevision 1-A staff also added a sentence to clarify that Report no longer includes any criteria to administrative needs for updating the report deviations from the MRP-AMP implementation procedures consist established I&E criteria to the NRC within with updated industry guidelines within 45 days of their discovery.

the scope of the AMP fall within the Aadministrative Ccontrols program The additional clarification on program element of the AMP and do not need to enhancements that solely involve needs be subject to individual programmatic for procedural updates is included to enhancements. reduce unnecessary-necessary burden.

Both the industry and staff agree that The staff deleted the previous sentence these types of PWRVI Vessel Internals that established a 45 day window for pPrograms (i.e., MRP-based programs) reporting programmatic MRP-227-based are living programs that are periodically deviations to the NRC. updated as new guidelines develop for the inspection or evaluation of PWR RVI components. Thus, the staff would expect licensee to keep their programs and related procedures up- to- date as new guidance develops and is issued by the industry relative to aging management needs of PWR RVI components.

Applicants have already demonstrated to the staff that they are already performing the appropriate updates of the programs and related procedures on an as needed basis. Since these types of activities fully fall within the scope of the Aadministrative Ccontrols program element of the AMP, there is no need for applicants to include additional AMP programmatic enhancements for 2-80 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes procedural update needs that would already fall within the scope of the Aadministrative Ccontrols program element of GALL-SLR AMP XI.M16A.

Operating The staff updated the Ooperating The changes are administrative and are Experience Eexperience program element to change analogous to those made to the any reference of the applicable report referencing of MRP-227, Rev.Revision 1-from MRP-227-A to MRP-227, A in the other program elements of the Rev.Revision 1-A. AMP.

References Staff added references for the MRP-227, The staffs changes in the references in Revision 1-A reportRevision 1-A Report the AMP are considered to be (EPRI Topical ReporttR No. 3002017168) administrative edits of the AMP. The and the staffs April 25, 2019 safety updated reference list is intended to keep evaluation (SE) for the MRP-227, the references in the AMP up- to- date Revision 1-A. The staff also added a with changes made to GALL AMP reference for the staffs correspondence XI.M16, PWR Vessel Internals, in SLR-letter to the EPRI MRP that endorsed ISG-2021-01-PWRVI.

MRP-227.,

GALL-SLR Table The staff amended the FSAR Supplement The changes to the FSAR Supplement XI-01 summary description example for GALL- summary description example are SLR AMP XI.M16A, PWR Vessel considered to be administrative and are Internals, to reference MRP-227, analogous to those made for the Revision 1-A as the applicable report for referencing of MRP-227, Rev.Revision 1-the AMP. A in the staffs update of GALL-SLR AMP XI.M16A, PWR Vessel Internals, in SLR-ISG-2021-01-PWRVI.

XI.M17: Flow-Accelerated Corrosion Program Clarify that commitments made in Previous staff audits noted that some Description response to NRC Generic Letter (GL) 89- commitments for a long-term FAC 08 were for an ongoing flow-accelerated monitoring program delineated in GL 89-corrosion (FAC) monitoring program. 08 had been considered one-time commitments and not ongoing commitments. For license renewal, the staff views the commitments in response to GL 89-08 to be ongoing commitments, remaining in effect, and part of the current licensing basis.

Program Add information that software quality The NSAC-202L notes that the Description assurance (QA) activities should continue CHECWORKS' code was developed in even though these activities are not accordance with QA policies requiring a required by the FAC program software formal software plan, detailed program QA classification. documentation, and a list of program bugs. However, the staff has found that, in most cases, the software QA classification for FAC software does not require any of the QA activities currently being performed on the FAC software.

The staff has determined that the currently performed QA activities should continue in order to provide reasonable 2-81 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes assurance that the effects of aging will be adequately managed.

Program Add that the guidance in EPRI During its review of the Surry SLRA Description 3002005530 can be used to manage loss (ML20052F523), the staff determined that of material due to erosion mechanisms. the guidance in EPRI 3002005530, Recommendations for an Effective Program Against Erosive Attack, can be used as the basis for a program to manage erosion mechanisms.

Detection of Aging Clarify by adding specific The staff added the specificity regarding Effects recommendations from EPRI guidance scope expansions based on previous NSAC202L for scope expansion due to SLRAs which did not appear to document unexpected or inconsistent inspection plant requirements for expanding results. inspection results, and because of previous operating experience noting the importance of inspection scope expansions (Licensee Event Report

[(LER]) 286/2018-003 and IN 2019-08).

As noted in LER 286/2018-003, a contributing cause to the FAC event was inadequate procedural guidance for inspection scope expansions. The GALL-SLR guidance was updated to highlight the importance of not only inspecting two diameters downstream from the affected component, but also to inspect the next two most susceptible components in the line as predicted by CHECWORKS' and to inspect corresponding components on other trains with a similar configuration to the one displaying wear.

Detection of Aging Include clarification that erosion The EPRI TR-112657, Revised Risk-Effects susceptibility screening provided in EPRI Informed Inservice Inspection Evaluation 3002005530 can be used to augment Procedure, includes specific guidance erosion location identification, except that regarding exclusion of erosion-cavitation system exclusion should be based on 100 consideration if flow occurs less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year instead of the 2 percent of hours per year. This is in contrast to the 2 operating time. percent operational time (approximately 275 hours0.00318 days <br />0.0764 hours <br />4.546958e-4 weeks <br />1.046375e-4 months <br />) specified in EPRI 3002005530. In addition, the severity of cavitation and operating experienceOE should be used to validate screening results based on previous reviews where severe cavitation was only occurring during opening and closing of valves caused very high wear rates.

Acceptance Add a safety factor of 2.0 from EPRI The EPRI 3002005530, Section 6.10 Criteria 3002005530 for erosion mechanism re- discusses Safety Factor determination inspection interval determinations. and states that the minimum should never be less than 2.0. Cases where a safety factor greater than 2.0 are also discussed in that section.

2-82 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes Administrative Add information that software quality The staff has found that, in most cases, Controls assurance (QA) activities should continue the software QA classification for FAC even though these activities are not software does not require any of the QA required by the FAC program software activities that are currently being QA classification. performed on the FAC software.

The staff has determined that the currently performed QA activities should continue in order to provide reasonable assurance that the effects of aging will be adequately managed.

Operating Add recently issued Information Notice Issues identified at Indian Point and Experience (IN) 2019-08 and associated Licensee Davis-Besse were discussed in IN 2019-Event ReportsLERs. 08 where legacy issues from initial modeling resulted in loss of pressure boundary integrity.

References Add EPRI 3002005530, During its review of the Surry SLRA Recommendations for an Effective (ML20052F523), the staff determined that Program Against Erosive Attack. In the guidance in EPRI 3002005530, can addition, add Information NoticeIN 2019- be used as the basis for a program to 08, Flow-Accelerated Corrosion Events, manage erosion mechanisms and the associated Licensee Event Reports (LER) from Indian Point, Information NoticeIN 2019-08, along with (286/2018-003) and Davis-Besse the associated LERs from Indian Point (346/2015-002). and Davis-Besse, were added to highlight legacy issues from initial FAC model development that resulted in several events.

XI.M19: Steam Generators Program Update the references to NUREG-1430, In September 2021, the NRC published Description Standard Technical Specifications - Revision 5 of NUREG-1430, Standard Babcock and Wilcox Plants, NUREG- Technical Specifications - Babcock and References 1431, Standard Technical Specifications Wilcox Plants (ML21272A363 [(Volume

- Westinghouse Plants, and NUREG- 1])); NUREG-1431, Standard Technical 1432, Standard Technical Specifications Specifications - Westinghouse Plants

- Combustion Engineering Plants, to (ML21259A155 [(Volume 1])); and Revision 5. NUREG-1432, Standard Technical Specifications - Combustion Engineering Plants (ML21258A421 [(Volume 1])).

Therefore, the Program Description and the References were updated to reference Revision 5 of the Sstandard Ttechnical Sspecifications.

Preventive Actions Add clarification regarding the type of Other than stress corrosion crackingSCC, corrosion that SG tube plugs may the staff is unaware of U.S. operating experience and add clarification that experienceOE of SG tube plugs extensive deposit buildup on the experiencing other types of corrosion in secondary side of SGs could affect tube the US. Therefore, the clarification was integrity. made to reflect the type of corrosion SG tube plugs may experience, which is stress corrosion cracking.

2-83 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes The clarification regarding extensive buildup of deposits on the secondary side of SGs was made to clearly state where extensive deposit buildup is expected in the SG.

Parameters Add clarification that the Steam Based on a review of current subsequent Monitored or Generators and Water Chemistry license renewal applicationsSLRAs, the Inspected programs are used to manage cracking staff noted that applicants may omit due to primary water stress corrosion aging management reviewAMR items for cracking (PWSCC) of divider plate managing cracking due to PWSCC if, assemblies and tube-to-tubesheet welds, after further evaluation, a plant-specific even if it is determined that use of the AMP is determined not to be required for One-Time Inspection program is not the divider plate assemblies or the tube-needed to confirm the effectiveness of the to-tubesheet welds. However, because Steam Generators and Water Chemistry the divider plate assemblies and tube-to-programs at mitigating PWSCC. tubesheet welds, are susceptible to PWSCC, the intent is that cracking due to In addition, references to Sections PWSCC for the divider plate assemblies 3.1.2.2.11 and 3.1.3.2.11 in NUREG- and the tube-to-tubesheet welds, be 2192, Standard Review Plan for Review managed by the Steam Generators and of Subsequent License Renewal Water Chemistry programs. Use of the Applications for Nuclear Power Plants, One-Time Inspection AMP (beyond the dated July 2017 (ADAMS Accession No. Steam Generators and Water Chemistry ML17188A158), were added, which programs) to confirm the effectiveness of contain the review procedures for the Steam Generators and Water determining whetherif a plant-specific Chemistry programs at mitigating AMP is required. cracking due to PWSCC, may be necessary depending, in part, on the materials of construction of the divider plate assemblies and the tube-to-tubesheet welds. Reference to Sections 3.1.2.2.11 and 3.1.3.2.11 in NUREG-2192 were added because they provide the review procedures for determining whetherif use of the One-Time Inspection AMP is necessary.

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Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes Parameters Changes were made to the inspection NRC-approved Technical Specification Monitored or frequency of divider plate assemblies, Specifications Task Force (TSTF), TSTF-Inspected tube-to-tubesheet welds, heads (channel 577, Revision 1, Revised Frequencies or lower/upper heads), and tubesheets to for Steam Generator Tube Inspections be consistent with the maximum (ADAMS Package Accession No.

inspection interval in the standard ML21099A086), increased the maximum technical specifications. inspection interval for thermally treated Alloy 690 (Alloy 690TT) SG tubing.

Specifically, the maximum inspection interval for Alloy 690TT is 96 Effective Full Power Months (EFPM), which is longer than the current 72 EFPM inspection frequency of the divider plate assemblies, tube-to-tubesheet welds, heads (channel or lower/upper heads),

and tubesheets in the GALL. Therefore, the inspection frequency of divider plate assemblies, tube-to-tubesheet welds, heads (channel or lower/upper heads),

and tubesheets was updated to reflect the maximum inspection interval for units with Alloy 690TT SG tubing.

Parameters Update references to address recent In December 2020, EPRI published EPRI Monitored or EPRI guidelines for SGs. 3002018267, PWR Primary-to-Inspected Secondary Leak Guidelines, Revision 5.

Therefore, EPRI 3002018267 replaces EPRI 1022832.

In December 2021, EPRI published EPRI 3002020909, Steam Generator Integrity Assessment Guidelines, Revision 5.

Therefore, EPRI 3002020909 replaces EPRI 3002007571.

Acceptance Criteria Update references to address recent In November 2016, EPRI published EPRI EPRI guidelines for SGs. 3002007856, Steam Generator In Situ Pressure Test Guidelines, Revision 5.

Therefore, EPRI 3002007856 replaces EPRI 1025132.

References Several references were updated to cite Since the last publication of Volume 2 of the latest revision and to correct titles and NUREG-2191, Generic Aging Lessons report numbers. In addition, TSTF-577, Learned for Subsequent License Revision 1, and NUREG-2192, Standard Renewal (GALL-SLR) Report (ADAMS Review Plant for Review of Subsequent Accession No. ML17187A204), new License Renewal Applications for Nuclear revisions of previously cited references Power Plants, were added as have been issued, which, in some references. instances, resulted in a new report number. The NRC staff also identified minor errors in the titles of previously cited references. Therefore, the references were updated to cite the latest revisions and new report numbers, and 2-85 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes those references with errors in the title were corrected.

A reference to NRC-approved TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (ADAMS Package Accession No. ML21099A086), was added. Revision 5 of the standard technical specifications incorporate TSTF-577.

A reference to NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, dated July 2017 (ADAMS Accession No. ML17188A158), was added because it is referenced in the Parameters Monitored or Inspection section of AMP XI.M19.

XI.M26: Fire Protection Program Change fire damper assembly to fire Based on a review of current subsequent Description damper housing to clarify that the fire license renewal applicationsSLRAs, the damper housing is the passive staff noted that AMR items have been Scope of Program component of a fire damper assembly included for fire damper assemblies or that is subject to aging management. fire damper housings. Fire damper Parameters assembly suggests the entire component Monitored or (e.g., housing, damper) is subject to Inspected aging management while fire damper housing suggests only a portion of the Detection of Aging component is subject to aging Effects management. Therefore, clarification is needed regarding which components of a Monitoring and fire damper assembly are passive Trending components and are subject to aging management.

Acceptance Criteria NUREG-2192 defines passive structures and components as those that perform their intended functions without moving parts or change in configuration or properties in accordance with 10 CFR 54.21(a)(1)(i). The fire damper housing does not perform its intended function with moving parts; however, the other fire damper assembly components, including the damper, do perform their intended function with moving parts.

Treating the fire damper itself as an active component not subject to aging management is consistent with the treatment of other dampers. Specifically, 2-86 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes 10 CFR 54.21(a)(1)(i) states that ventilation dampers are excluded from aging management, and SRP-SLR Table 2.1-6 states that only the housings of dampers, louvers, and gravity dampers associated with valves are subject to aging management.

GALL-SLR Item VII.G.A-789, SRP Item 3.3-1, 255 in Volume 1 of NUREG-2191 is revised to address cracking and loss of material for metallic fire damper housings exposed to air by the Fire Protection program.

SRP-SLR Table 2.1-6 is revised to state that the fire damper housing is subject to aging management.

Program Add statement that the GALL-SLR Report Based on a review of current subsequent Description AMP XI.M26 is complemented by GALL- license renewal applicationsSLRAs, the SLR Report AMP XI.S5. In addition, add staff noted instances where only the a statement that the Structures Masonry Walls program or the Fire Monitoring and Fire Protection program Protection program is cited to monitor would together manage applicable aging applicable aging effects for masonry walls effects for structural fire barriers, and that that are considered fire barriers.

the Masonry Walls and Fire Protection However, GALL-SLR Report AMP XI.S5 programs would together manage states, The aging effects on masonry applicable aging effects for masonry walls walls that are considered fire barriers are that are considered fire barriers. also managed by the GALL-SLR Report AMP XI.M26, Fire Protection, as well as being managed by this program. This statement is consistent with GALL-SLR AMR Iitem VII.G.A-626, SRP item 3.3-1, 179 in Volume 1 of NUREG-2191, which cites both programs for managing applicable aging effects for masonry walls that are considered fire barriers.

Therefore, the statement that the Fire Protection program is complemented by the Masonry Walls program is added to be consistent with GALL-SLR Report AMP XI.S5 and GALL-SLR AMR Iitem VII.G.A-626, SRP item 3.3-1, 179. This addition is consistent with the existing statement that the Fire Protection program is complemented by the Structures Monitoring program.

Based on a review of current subsequent license renewal applicationsSLRAs, the staff also noted instances where only the Structures Monitoring program or the Fire Protection program is cited to monitor 2-87 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes applicable aging effects for structural fire barriers. However, GALL-SLR AMR Iitem VII.G.A-90, SRP Iitem 3.3-1, 060 cites both the Structures Monitoring and Fire Protection programs for managing applicable aging effects for structural fire barriers (walls, ceilings, and floors).

Changes are not required to GALL-SLR AMP Report XI.M26 because it already states that the Fire Protection pProgram is complemented by the Structures Monitoring program.

Scope of Program Add that materials used to secure fire Based on a review of current subsequent wraps are subject to aging management license renewal applicationsSLRAs, the by the Fire Protection program. staff noted that either it was unclear whether AMR items were included, or no AMR items were included for materials used to secure fire wraps.

The clarification is being made because EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Effects for Structures and Structural Components (Structural Tools), dated November 2018, states that materials used to secure the fire wrap areis considered part of the fire wrap. Therefore, since the fire wrap is subject to aging management so is the material that is used to secure fire wrap.

Monitoring and Add clarification that the results of The clarification is needed because fire Trending inspections for all aging effects, not just barrier penetration seals and materials cracking and loss of material, are trended used as fireproofing/fire barriers may to provide for timely detection of aging have aging effects, other than cracking effects. In addition, add clarification that and loss of material, therefore, inspection fire barriers include walls, ceilings, floors, results of all aging effects, not just and other fire barrier materials and that cracking and loss of material, are to be the results of inspections of fire barrier trended to provide for timely detection of walls, ceilings, and floors and other fire aging effects.

barrier materials are trended to provide for timely detection of aging effects. Clarification is needed to indicate that fire barriers include walls, ceilings, and floors and other fire barrier materials and that the results of inspections of that fire barriers include walls, ceilings, and floors and other fire barrier materials are trended to provide for timely detection of aging effects.

Detection of Aging Add clarification that separation of seals Clarification is needed because fire Effects can also be from ceilings and floors, not barrier penetration seals can be used with just from walls and components. ceilings and floors, not just with walls and Acceptance Criteria components. Therefore, separation of fire 2-88 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes barrier penetration seals can also be from ceilings and floors.

References Several references to cite the latest Since the last publication of Volume 2 of revision and to make formatting changes NUREG-2191, Generic Aging Lessons for consistency. In addition, EPRI Learned for Subsequent License 3002013084, Long-Term Operations: Renewal (GALL-SLR) Report (ADAMS Subsequent License Renewal Aging Accession No. ML17187A204), new Effects for Structures and Structural revisions of several of the previously cited Components (Structural Tools), and references have been issued. The staff NRC IN-89-52, Potential Fire Damper also made minor formatting changes for Operational Problems, were added as consistency.

references.

A reference to EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Effects for Structures and Structural Components (Structural Tools), was added because it is now referenced in the Scope of Program section of GALL-SLR AMP Report XI.M26.

A reference to NRC IN-89-52, Potential Fire Damper Operational Problems, was added because it provides relevant operating history for fire dampers.

XI.M27: Fire Water System Program Added replacing or testing using Information regarding replacing or testing Description guidance in National Fire Protection dry sprinklers and fast response Association (NFPA) 25 for dry sprinklers sprinklers because these type sprinklers and fast response sprinklers. may exist in nuclear power plants, and it is consistent with NFPA 25. The AMP XI.M27 currently only includes replacing or testing of sprinklers which areis at a different frequency from the frequency in NFPA 25 for dry sprinklers and fast response sprinklers.

Detection of Aging Deleted reference to fire hydrant hose The tests and inspections were deleted Effects hydrostatic tests and gasket inspections. because fire hydrant hoses and gaskets are typically excluded from aging management review based on SRP-SLR Table 2.1-3, Specific Staff Guidance on Screening, Consumables, Iitems (1) and (4).

Detection of Aging The recommended extent of standpipe The staff has revised the Effects and hose system flow tests is reduced if recommendations for this testing for two the tests conducted no earlier than 5 reasons. The purpose of this testing is to years prior to the subsequent period of detect potential flow blockage due to extended operation meet pressure and fouling and loss of material. The fire water flow criteria. New footnote nNo. 11 to system for plants entering the subsequent Table XI.M27-1. period of extended operation will have been in service for at least 60 years; with the program change allowing testing 2-89 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes results to be monitored as early as 5 years prior to the subsequent period of extended operation. Based on its review of many fire water system aging management programsAMPs, the staff has concluded that the internal environment for the fire water system (i.e., water supply) has either been the same or, at some plants, water supplies have been modified to take its source from a less adverse environment (e.g.,

modifying the system to take suction from city water or well water in lieu of a river).

Given that the environment has been the same or less adverse and the acceptance criteria has been met, the design pressure at the required flow, it is reasonable to assume that a reduced sample size would be adequate to provide continued confirmation that the fire water system will meet its intended function in relation to these tests.

Detection of Aging The recommended drain down level for The staff has concluded that it is Effects hydrant barrels was revised based on the reasonable to assume that: (a) water that plant-specific frost line and operating is in a hydrant barrel below the frost line experience. New footnote nNo. 12 was will not freeze because of heat provided added to Table XI.M27-1. by the Earth earth below the frost line, which is supported by national standards for the installation of fire service mains and their appurtenances including fire hydrants; and (b) national consensus standards for fire piping, such as Section 10.4.2.1 of NFPA 24 and Section 3.3.9.1 of NFPA 25, only require that the hydrant isolation valve be installed below the frost line.

The staff reviewed the following:

• NFPA 24, Standard for the Installation of Private Fire Service Mains and Their Appurtenances, Section 10.4.2.1, Protection for Piping, states that the top of the pipe shall be buried not less than 12 inches (in) below the frost line for the locality.

• Section 3.3.9.1 of NFPA 25 and Section 3.4.1.1 of NFPA 24 state that the control valve for a dry barrel hydrant is located below the frost line.

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Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes

• The Manual of Water Supply Practices, M17, Fire Hydrants:

Installation, Field Testing, and Maintenance, section titled Types of Dry Barrel Hydrants states that the main valve is located below the normal frost line to protect the hydrant from freezing.

Detection of Aging The recommended periodicity of fire Like Similar to the basis for the inclusion Effects pump suction screen and strainer of the new footnote No. 11 to Table inspections was revised based on the XI.M27 1, it is reasonable to assume that suction source of the fire pump and plant- the amount of internally generated debris specific operating experience. New (i.e., loss of material) from the piping system would not vary after 55 years of footnote nNo. 13 was added operating the fire water system with Table XI.M27-1. exposure to the same environment.

Accordingly, the examination results for the screens and strainers should predict future performance. However, this is not the case for fire water pumps which take suction directly from sources of makeup with the potential for bulk debris. For example, degraded cooling tower fill and storm generated debris into an ocean or river, which could eventually accumulate on the suction screens or strainers later in life. Fire pump suction strainers was added because depending on the installation, there may be fire pump suction screens and strainers (see NFPA 25 Figure A.8.2.2).

Detection of Aging The recommended sample size and Like the basis for the inclusion of the new Effects periodicity of conducting main drain tests footnote nNo. 11 to Table XI.M27 1, it is was revised based on test results and reasonable to assume that the amount of plant-specific operating experience. New internally generated debris (i.e., loss of footnote nNo. was added 14 to Table material) from the piping system would not vary after 55 years of operating the XI.M27-1.

fire water system with exposure to the same environment. Accordingly, the test results and plant- specific operating experienceOE can provide effective enough insights such that the extent of testing and periodicity changes can still provide reasonable assurance that the system will meet its intended function. In addition, the reduced sampling size is consistent with the number of recommended tests or inspections (i.e.,

20% percent) in several sampling based AMPs (e.g., XI.M38, Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components).

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Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes Detection of Aging The recommendation for inspecting the Although NFPA 25 states that exterior Effects exterior surfaces of insulated fire water visual examinations should be conducted storage tanks was revised to allow on an annual basis, conducting insulation removal and inspection inspections consistent with GALL- SLR consistent with AMP XI.M29, Outdoor Report AMP XI.M29, provides reasonable assurance that loss of material and and Large Atmospheric and Metallic cracking (e.g., aluminum tanks) will be Storage Tanks. In addition, the reference adequately managed for these tanks.

to footnote nNo. 10 was added to Table Refueling outage interval inspections of XI.M271 was deleted and the inspections the external surfaces of the tank are are recommended as occurring on a consistent with GALL- SLR Report AMP refueling outage interval in lieu of XI.M29.

annually. New footnote nNo. 15 was added to Table XI.M27-1.

Detection of Aging The NFPA 25 Section column in Table Specificity was removed while still Effects XI.M271 was revised to Periodicity. referencing the appropriate portions of Footnotes were revised accordingly. NFPA 25, so that the title of the tests or examinations remains while providing applicants flexibility in meeting the appropriate portions of NFPA 25.

Detection of Aging The recommendation for performing main Although Section 13.2.5 of NFPA 25 Effects drain tests was revised to include that full states, When there is a 10 percent flow pressures should not be compared reduction in full flow pressure when only to the immediately prior test result. compared to the original acceptance test New footnote nNo. 16 was added to or previously performed tests, the cause Table XI.M27-1. of the reduction shall be identified and corrected if necessary., The staff notes that if the test-to-test pressure monitoring only uses the immediately prior test result, significant degradation of the fire water system supply over several years may not be identified while still being less than a 10% percent reduction from the previous test.

Detection of Aging A recommendation to visually inspect and Mainline strainers may become blocked Effects clean the mainline strainers of the private or corrode over time, therefore, consistent fire service main was added. New with NFPA 25 Section 7.2.2.3, a footnote nNo. 17 was added to Table recommendation was added to visually XI.M27-1. inspect and clean the mainline strainers annually and after each significant flow.

NFPA 25 Sections 7.2.2.3 and A.7.2.2.3 provide additional information on significant flow.

XI.M29: Outdoor and Large Atmospheric Metallic Storage Tanks Detection of Aging Clarify that the 1 square-footft2 sections of During its the Staffs review of a Effects insulation should be taken from multiple recentNorth Anna SLRA, an the applicant locations to ensure ascertain that the proposed taking all 1 -foot piping samples samples are representative of the entire from a single excavation location without exterior of the tank. any technical justification for this position.

SRP-SLR Section A.1.2.3.4 notes that when sampling is used to represent a 2-92 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes larger population of [components],

applicants provide the basis for the inspection population and sample size.

The staff added additional guidance to alleviate future questions.

XI.M32: One-Time Inspection Parameters As revised by Revision 1 to the SLR AMP The staff deleted footnote no. 3 because Monitored or XI.M32, footnote no. 3, which stated, it has concluded that the more rigorous Inspected Visual inspections conducted to detect examination techniques cited in AMP potential loss of material or cracking of XI.M32 should be conducted when SS and aluminum alloy support members; periodic inspections will not be conducted welds; bolted connections; support during the subsequent period of extended anchorage to building structure exposed operation. These techniques can detect to air or condensation (see SRP-SLR minor indications of loss of material and Section 3.5.2.2.2.4) may be conducted cracking. If the one One-Ttime consistent with those for the GALL-SLR Iinspections and plant- specific operation Report AMP XI.S6, Structures experienceOE do not reveal loss of Monitoring. was deleted. material or cracking, periodic inspections will not be conducted during the subsequent period of extended operation.

As a result, it is important to demonstrate that the environment conditions will not promote loss of material or cracking by more rigorous examination techniques during the Oone- Ttime Iinspection.

For a support, minor loss of material or cracking that might not be detectable during a one- time walkdown inspection will likely not impact the intended function of the support; however, the staff has concluded that growth of loss of material or cracking will become more evident during periodic inspections of supports.

Program Incorporation of an incubation period into Since the One--Time Inspection program Description AMP XI.M32 for repairs or replacements is based on a 50 to 60--year incubation that are used to correct a condition period, inspections on recently installed Detection of Aging adverse to quality that is related to repairs or replacements do not provide Effects plant-specific operating experience. objective evidence that adverse aging effects are not occurring at a rate that would cause a loss of intended function during the subsequent period of extended operation. This scenario occurred during the review of a carbon fiber wrap for the Indian Point LRA, when it was found that a Oone-Ttime Iinspection was proposed for a repair that was only in service for about five 5 years. The review resulted in several management discussions and RAIs.

XI.M33: Selective Leaching 2-93 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes Scope of Program Eliminate the conditional exclusion of LR-ISG-2011-03, Changes to the and Detection of buried components with external coatings Generic Aging Lessons Learned (GALL)

Aging Effects from inspection. Report, Revision 2 AMP XI.M41, Buried and Underground Piping and Tanks, introduced the recommendation regarding reducing the number of selective leaching inspections for buried piping based on the presence of external coatings. The basis provided in LR-ISG-2011-03 for this reduction is coatings can prevent or mitigate selective leaching in buried components. These recommendations were subsequently moved from AMP XI.M41 to AMP XI.M33 with the issuance of LR-ISG-201501, Changes to Buried and Underground Piping and Tank Recommendations. These recommendations, which were later incorporated into GALL-SLR Report AMP XI.M33, provided conditions where externally coated buried piping may be excluded from the scope of the Selective Leaching program and provided conditions where the inspection sample size may be reduced by 50% percent.

During the staffs review of the Surry a subsequent license renewal applicationSLRA, two ruptures occurred in buried gray cast iron piping associated with the fire protection system due to selective leaching (ADAMS Accession No. ML19310E716). Prior to the ruptures occurring, the applicant had responded to a staffs request to clarify if all buried fire protection piping is externally coated by stating specifications require buried cast iron fire protection piping to be coated with bituminous coating (ADAMS Accession No. ML19183A386). The staff notes that in this instance, external coatings were ineffective in preventing or mitigating selective leaching of the buried fire protection piping.

Although external coatings were ineffective in this operating experienceOE example, the staff recognizes that external coatings can be effective in preventing or mitigating selective leaching of buried piping based on site-specific 2-94 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes parameters such as coating types utilized, results of inspections of these coatings, soil corrosivity as determined by soil testing in the vicinity of buried piping susceptible to selective leaching, etc.

Based on the need for site-specific information, the staff is removing the current generic recommendations in the AMP with respect to reducing or eliminating selective leaching inspections of buried piping based on the presence of external coatings. An applicant still has the option to take an exception with technical justification to inspect a reduced sample size of buried components.

Scope of Program, Inclusion of malleable iron as a material During its review of recent SLRA plant-Detection of Aging susceptible to selective leaching. specific operating experienceOE, in Effects, and response to the staffs observation Acceptance regarding dark corrosion product layers Criteria indicative of graphitic corrosion on the internal surfaces of malleable iron fittings exposed to a closed -cycle cooling water environment, an applicant revised its SLRA to reflect that malleable iron components exposed to environments where selective leaching could occur will be managed for loss of material due to selective leaching. Based on this new operating experienceOE, the staff has revised guidance documents (i.e., GALL-SLR and SRP-SLR Report) to include malleable iron as a material susceptible to selective leaching. The staffs revisions to guidance documents are similar to those incorporated when ductile iron was added as a susceptible material in 2016.

In addition, due to similarities in microstructure between malleable iron and ductile iron, the staff revised GALL-SLR Report AMP XI.M33 to reflect that these two materials may be grouped together in sample populations. The staff notes ductile iron and malleable iron consist of spherical graphite nodules and irregularly shaped graphite nodules, respectively, embedded in iron (whereas gray cast iron has a semicontinuous network of graphite flakes embedded in iron).

Detection of Aging Clarify that a technical justification for NUREG-2222, Disposition of Public Effects using the extent of inspections in the Comments on the Draft Subsequent License Renewal Guidance Documents 2-95 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes AMP should be provided for gray cast NUREG-2191 and NUREG-2192, iron piping exposed to soil. provides the basis for reducing the extent of inspections for selective leaching during the subsequent period of extended operation (i.e., 3% percent with a maximum of 10 components per GALL-SLR guidance) when compared to the extent of inspections for selective leaching during the initial period of extended operation (i.e., 20% percent with a maximum of 25 components per GALL Report, Revision 2 guidance). Part of the basis for reducing the extent of inspections is that industry operating experienceOE had not identified instances of loss of material due to selective leaching which had resulted in a loss of intended function for the component.

The NRC issued Information Notice (IN) 202004, Operating Experience Regarding Failure of Buried Fire Protection Main Yard Piping, to inform the industry of operating experienceOE involving the loss of function of buried gray cast iron fire water main yard piping due to multiple factors, including graphitic corrosion (i.e., selective leaching),

overpressurization, low cycle fatigue, and surface loads. As noted in the IN, a contributing cause to the failures of buried gray cast iron piping at Surry Power Station (SPS) was the external reduction in wall thickness at several locations due to graphitic corrosion. Based on recent industry operating experienceOE, the staff revised GALL-SLR Report AMP XI.M33 to reflect that a technical justification is provided in the SLRA when using the sample size recommend in the AMP for gray cast iron piping exposed to soil. Alternatively, an applicant may elect to use the sample size recommended in Revision 2 of the GALL Report (i.e., 20 percent with a maximum of 25 components) for this population.

Detection of Aging Clarify that 1-foot pipe samples should be During the sStaffsits review of the a Effects taken from multiple locations to ensure recentNorth Anna SLRA, an the applicant proposed taking all 1-foot piping samples 2-96 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes that the sample is representative of the from a single excavation location without entire population. any technical justification for this position.

SRP-SLR Section A.1.2.3.4 notes that when sampling is used to represent a larger population of [components],

applicants provide the basis for the inspection population and sample size.

The staff added additional guidance to alleviate future questions.

Detection of Aging Include soil parameter consistency when The current factors to consider when Effects providing a basis for reducing the total reducing the number of inspections at number of inspections at multiunit sites multiunit sites focus on aqueous for buried components. environments; however, components susceptible to selective leaching are also commonly exposed to a soil environment (e.g., buried cast iron fire water system piping).

Acceptance Revise the pointer related to not crediting Due to an editorial error, the discussion Criteria material properties of dealloyed portions related to not crediting the material from criterion (c) to criterion (d). properties of the dealloyed portion of a component in any evaluations inadvertently referred to criterion (c) instead of criterion (d). Criterion (c) refers to a superficial dealloyed layer which would not involve an evaluation, whereas criterion (d) would involve an evaluation to show that system design requirements would be met.

Operating Include recent operating experienceOE. The cited operating experienceOE Experience contributed to the program changes associated with: (a) inspection reductions that credit external coatings and common soil environments; and (b) the addition of malleable iron as a material susceptible to selective leaching.

XI.M21A. Closed Water Treated System Detection of Aging Clarify that 1-foot pipe samples should be During its the sStaffs review of the a Effects taken from multiple locations to ensure recentNorth Anna SLRA, an the applicant that the samples are representative of the proposed taking all 1-foot piping samples entire population. from a single excavation location without any technical justification for this position.

SRP-SLR Section A.1.2.3.4 notes that when sampling is used to represent a larger population of [components],

applicants provide the basis for the inspection population and sample size.

The staff added additional guidance to alleviate future questions.

XI.M29: Outdoor and Large Atmospheric Metallic Storage Tanks 2-97 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes 2

Detection of Aging Clarify that the 1 square-foot (ft ) sections During its review of a recent SLRA, an Effects of insulation should be taken from applicant proposed taking all 1-foot piping multiple locations to ensure that the samples from a single excavation location samples are representative of the entire without any technical justification for this exterior of the tank. position. SRP-SLR Section A.1.2.3.4 notes that when sampling is used to represent a larger population of

[components], applicants provide the basis for the inspection population and sample size. The staff added additional guidance to alleviate future questions.

XI.M36: External Surfaces Monitoring of Mechanical Components Detection of Aging Clarify that 1-foot ft pipe samples should During its the Sstaffs review of the a Effects be taken from multiple locations to ensure recentNorth Anna SLRA, an the applicant that the samples are representative of the proposed taking all 1-foot piping samples entire population. from a single excavation location without any technical justification for this position.

SRP-SLR Section A.1.2.3.4 notes that when sampling is used to represent a larger population of [components],

applicants provide the basis for the inspection population and sample size.

The staff added additional guidance to alleviate future questions.

XI.M38: Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components Detection of Aging Clarify that 1-foot pipe samples should be During its the sStaffs review of the a Effects taken from multiple locations to ensure recentNorth Anna SLRA, an the applicant that the samples are representative of the proposed taking all 1-foot piping samples entire population. from a single excavation location without any technical justification for this position.

The SRP-SLR Section A.1.2.3.4 notes that when sampling is used to represent a larger population of [components],

applicants provide the basis for the inspection population and sample size.

The staff added additional guidance to alleviate future questions.

XI.M41: Buried and Underground Piping and Tanks Acceptance Revise AMP XI.M41, Buried and The LR-ISG-2011-03, Changes to the Criteria Underground Piping and Tanks, to clarify Generic Aging Lessons Learned (GALL) that when the 100 millivolt (mV) criterion Report Revision 2 Aging Management is utilized to protect copper alloy or Program (AMP) XI.M41, Buried and aluminum alloy components, applicants Underground Piping and Tanks, must explain in the application why the provides the following recommendation:

effects of mixed potentials are minimal and why the most anodic metal in the [w]hen the 100 mV criterion is utilized in system is adequately protected. lieu of the -850 mV CSE [copper/copper sulfate reference electrode] criterion for steel piping, or where copper or 2-98 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes aluminum components are protected, applicants must explain in the application why the effects of mixed potentials are minimal and why the most anodic metal in the system is adequately protected.

This recommendation was removed when LR-ISG-2015-01, Changes to Buried and Underground Piping and Tank Recommendations, was issued. LR-ISG-2015-01 addressed the issue of mixed potentials for steel components by introducing the concept of confirmatory testing (i.e., verifying external loss of material rate through the use installed electrical resistance [(ER]) corrosion rate probes). However, LR-ISG-2015-01 did not address the issue of mixed potentials for copper alloy and aluminum alloy components, which was an oversight by the staff.

The staff considered including confirmatory testing for aluminum alloy components; however, ER probes are intended to indicate metal loss by general corrosion (i.e., not suited for aluminum where pitting and crevice corrosion are the aging mechanisms in a soil environment). Specifically, the staff reviewed the following:

Corrosion Tests and Standards:

Application and Interpretation states [i]n the same way as mass loss on corrosion test specimens, resistance measurements on electrical resistance probes indicate metal loss by general corrosion. Pitting is generally not noticeable until near the end of probe life, where the effect of pitting becomes runaway on the resistance measurement.

Corrosion Rate Probes for Soil Environments, American Society for Metals (ASM) Handbook Volume 13C, states [t]he ER technique does not function well in pitting environments because corrosion pits could be interpreted as thinning of the sensor cross-sectional area and thus as a uniform corrosion rate.

2-99 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes Regarding copper alloy components, ER probes for buried environments predominately contain corrosion rate elements constructed from carbon steel; therefore, it is unclear if proposing the use of ER probes with copper corrosion rate elements is practical. In addition, based on its review of the first three subsequent license renewal applications, which did not cite this component, material, and environment combination (i.e., copper alloy piping exposed to soil),

the staff did not identify a need to provide a specific recommendation in the AMP to address the issue of mixed potentials for copper alloy components.

To address the issue of mixed potentials for copper alloy and aluminum alloy components, the staff reinserted the recommendation from LR-ISG-201103 regarding the use of the 100 mV criterion in a mixed metal environment.

Recommendations regarding validating the use of the 100 mV criterion (through the use of ER probes) for steel piping were not revised.

Detection of Aging Revise GALL-SLR AMP XI.M41 to clarify An applicant inadequately addressed an Effects that evaluation of plant- specific operating RAI regarding several through-wall leaks experience (OE) includes out of scope in buried piping by stating that the Operating buried components if they are components were not within the scope of Experience representative of in scope buried license renewal. During a clarification call components (e.g., similar material for the RAI, the staff stated that in-composition, degradation mechanisms, scopescope buried components might coatings, soil conditions, history of have the same material composition, cathodic protection). degradation mechanisms, coatings, soil conditions, and history of cathodic protection. Therefore, it was unclear to the staff why aging experienced on the out of scope components would not be equally applicable to in-scopescope buried components. The staff requested that the applicant provide a basis for why the out of scope components (where selfrevealingself-revealing issues had been identified) were not representative of in-scopescope components or provide a basis for why additional inspections, beyond those recommended in AMP XI.M41, were not appropriate.

Detection of Aging Revise GALL-SLR AMP XI.M41 GALL-SLR AMP XI.M41 Preventive Effects Preventive Action Category F inspection Action Category F states, [i]nspection recommendations to clarify that this criteria provided for Category F piping is 2-100 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes inspection category is based on a used for those portions of in-scope buried cathodic protection system being piping which cannot be classified as installed; however, it is not meeting Category C, D, or E. It was not the staffs performance criteria. intent that Preventive Action Category F would be used where cathodic protection was not installed. In this case, the applicant would develop plant- specific inspection quantities.

Preventive Actions Revise the Ppreventive Aactions Recent operating experienceOE at a program element of GALL-SLR AMP station with a renewed license revealed a XI.M41 to recommend external coatings significant failure of prestressed concrete for underground cementitious piping. cylindrical piping (PCCP) in an underground environment. This piping was not in scope; however, it was exposed to the same environment as in-scopescope piping. Physical deterioration of the cement and corrosion of the internal prestressed wire reduced the pipes strength and led to a local rupture.

External coatings could have helped prevent this failure.

In addition, Concrete Pressure Pipe -

Manual of Water Supply Practices recommends barrier coatings for atmospheric exposure of concrete pressure pipe where the exposed line may be subjected to large temperature fluctuations, wetting and drying cycles, freezing and thawing cycles, and atmospheric carbonation.

Detection of Aging Revise GALL-SLR AMP XI.M41 to include Preventive Action Category E of GALL-Effects EPRI Report 3002005294, Soil Sampling SLR Table XI.M41, Inspection of Buried and Testing Methods to Evaluate the and Underground Piping and Tanks, Corrosivity of the Environment for Buried currently references American Water Piping and Tanks at Nuclear Power Works Association (AWWA) C105, Plants, Table 9-4, Soil Corrosivity Index Polyethylene Encasement for Ductile-from BPWORKS, as an additional approach to determine soil corrosivity. Iron Pipe Systems, and Table A.1, Soil Test Evaluation, to determine soil corrosivity. Nine points or less indicates noncorrosive soil using AWWA C105, Table A.1.

As an alternative to using this AWWA standard, the staff finds that a threshold of ten points or less using the carbon steel column in Table 9-4 of EPRI Report 3002005294 to be acceptable for determining noncorrosive soil for carbon steel. The staffs basis is documented in the Safety Evaluation Report Related to 2-101 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes the Subsequent License Renewal of Surry Power Station, Units 1 and 2 (ADAMS Accession No. ML ML20052F520), Section 3.0.3.2.20, Buried and Underground Piping and Tanks.

The staff notes that Preventive Action Category E of GALL-SLR Table XI.M412 also applies to aluminum and copper alloys (i.e., not just carbon steel). Based on its review of the columns in Table 9-4 of EPRI Report 3002005294 associated with aluminum and copper alloys, the staff finds a threshold of ten points or less to also be acceptable for these two material types.

XI.M42: Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks Detection of Aging Clarify that 1-foot pipe samples should be During the sStaffsits review of the a Effects taken from multiple locations to ensure recentNorth Anna SLRA, an the applicant that the samples are representative of the proposed taking all 1-foot piping samples entire population. from a single excavation location without any technical justification for this position.

The SRP-SLR Section A.1.2.3.4 notes that when sampling is used to represent a larger population of [components],

applicants provide the basis for the inspection population and sample size.

The staff added additional guidance to alleviate future questions.

XI.M43: High Density Polyethylene (HDPE) Piping and Carbon Fiber Reinforced Polymer (CFRP)

Repaired Piping GALL-SLRP The AMP XI.43 is a new AMP. This AMP The technical basis for the new AMP is to Report, Chapter XI manages the aging of buried high density manage the effects of age-related polyethylene (HDPE) piping and carbon degradation mechanisms that are fiber reinforced polymer (CFRP) repaired applicable to HDPE piping and CFRP piping. This program manages aging repaired piping. This new AMP reflects through preventive, mitigative, inspection the recent introduction and increasing use and in some cases, performance of CFRP repaired piping at reactor monitoring activities. This AMP manages facilities. The unique aging issues and aging effects such as loss of material, aging management approaches for CFRP cracking, disbondment, damage and repaired piping and HDPE piping were leaking. Preventive actions and considered to be most effectively inspection intervals are defined, addressed with a dedicated AMP.

depending on the environment and the Depending on the material, the AMP type of material. addresses the preventive and mitigative techniques and may include external coatings, cathodic protection, and the quality of backfill. In addition, depending on the material, inspection activities may 2-102 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Bases for Changes include electrochemical verification of the effectiveness of cathodic protection, nondestructive evaluation of pipe wall thicknesses, pressure testing of the pipe, volumetric inspections, and visual inspections of the pipe from the exterior and/or interior.

1 Table 2-34 Table 2-30 GALL-SLR Report, Revision 1, Chapter XI, Structural Aging 2 Management Programs, Differences From GALL-SLR Report, Revision 0, 3 and Their Technical Bases Location of Change Summary of Significant Changes Technical Basis for Change XI.S1: ASME Section XI, Subsection IWE Preventive Action Added wording to add ASTM ASTM F3125/F3125M-21 specification International (ASTM) F3125 bolts to the consolidates and replaces six ASTM list of bolts that require preventive standards that include: A325, A325M, actions. A490, A490M, F1852, and F2280.

XI.S3: ASME Section XI, Subsection IWF Preventive Action Added wording to add ASTM F3125 bolts ASTM F3125/F3125M-21 specification to the list of bolts that require preventive consolidates and replaces six ASTM actions. standards that include: A325, A325M, A490, A490M, F1852 and F2280.

XI.S6 : Structures Monitoring Scope of Program Added reference to AMP XI.M26, Fire Based on a review of current subsequent Protection, to clarify that together the license renewal applications (SLRAs),

Detection of Aging Structures Monitoring program and the the staff noted instances where only the Effects Fire Protection program manage the Structures Monitoring program or the applicable aging effects for reinforced Fire Protection program is cited to concrete structural fire barriers (walls, manage applicable aging effects for ceilings, and floors). reinforced concrete structural fire barriers. However, GALL-SLR AMR Iitem VII.G.A-90, SRP item 3.3-1, 060 cites both the Structures Monitoring program and the Fire Protection program for managing applicable aging effects for reinforced concrete structural fire barriers (walls, ceilings, and floors).

Adding reference to AMP XI.M26 to AMP XI.S6 is consistent with GALL-SLR AMR Iitem VII.G.A-90, SRP item 3.3-1, 060 in Volume 1 of NUREG-2191, which cites both programs for managing applicable aging effects for reinforced concrete structural fire barriers (walls, ceilings, and floors), and clarifies in the AMP 2-103 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Basis for Change XI.S6 description that both programs manage the applicable aging effects.

Preventive Action Added wording to add ASTM F3125 bolts ASTM F3125/F3125M-21 specification to the list of bolts that require preventive consolidates and replaces six ASTM actions. standards that include: A325, A325M, A490, A490M, F1852 and F2280.

Detection of Aging Clarify ambiguities when applying the Effects term groundwater/soil in the GALL-SLR AMP.

XI.S7: Inspection of Water-Control Structures Associated with Nuclear Power Plants Preventive Action Added wording to add ASTM F3125 bolts ASTM F3125/F3125M-21 specification to the list of bolts that require preventive consolidates and replaces six ASTM actions. standards that include: A325, A325M, A490, A490M, F1852 and F2280.

Detection of Aging Clarify ambiguities when applying the The forward slash from the Effects term groundwater/soil in the GALL-SLR groundwater/soil term is replaced and AMP. spelled-out to clearly communicate its usage within the XI.S6, Structures Monitoring, and the XI.S7, Inspection of Water Control Structures Associated with Nuclear Power Plants programs. For example, when the guidance document refers to an aggressive groundwater/soil it is intended to be used as aggressive groundwater or aggressive soil; and when the guidance document refers to a nonaggressive groundwater/soil it is intended to be used as nonaggressive groundwater and nonaggressive soil.

XI.S8: Protective Coating Monitoring and Maintenance Program Revisionses made to the frequency of ASTM International (formerly American Description inservice coating inspection monitoring to Society for Testing and Materials) allow the inspection of coatings meeting Specification D5163-08, Standard Guide Detection of Aging GALL-SLR AMP XI.S8 Element 6, for Establishing a Program for Condition Effects Acceptance Criteria, to be performed Assessment of Coating Service Level I on a frequency not to exceed 6 years, Coating Systems in Nuclear Power Monitoring and based on trending of the total amount of Plants. West Conshohocken, Trending permitted degraded coatings. Pennsylvania. ASTM International, 2008, paragraph 6, notes that the licensee shall Operating Updates GALL-SLR Report AMP XI.S8 determine the frequency of inservice Experience to reference Regulatory Guide (RG) coating inspections. ASTM D5163-08, 1.54, Service Level I, II, III, and In- paragraph 6, also notes that it is a good References Scope License Renewal Protective practice to perform inspections during Program Coatings Applied to Nuclear Power each refueling outage at . Aan interval of Description Plants, Revision 3, issued April 2017, as not to exceed< 6 years based on station it is the most current revision at the time operating experienceOE may be justified Detection of Aging of this change. if coatings meet the acceptance criteria Effects (AMP XI.S8 Element 6) and trending 2-104 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Basis for Change activities for the total amount of Monitoring and degraded coatings in containment Trending indicate that a margin will be maintained (AMP XI.S8 Element 5). A qualified Operating nuclear coating specialist performs a Experience coating condition assessment report to determine the priority of repairs to be References conducted during the current outage and repairs that can be postponed to a future date (ASTM D5163-08, paragraph 11.1.2). Trending of the total amount of degraded coatings allowed in containment iswill also performed.

RG 1.54, Revision 3, contains the most up-to-date NRC guidance on the selection, application, qualification, inspection, and maintenance of protective coatings applicable under GALL-SLR Report AMP XI.S8.

For an applicant to extend the inspection interval stated in the GALL-SLR Report (each refueling outage), an applicant must demonstrate that margin to the ECCS suction strainer operability limits for coating debris will be maintained during the subsequent period of extended operation based on operating experience and trending of degraded/unqualified coatings. If plant-specific operating experience OE identifies coating degradation mechanisms that indicate the potential to exceed the ECCS suction strainer debris margin, an applicant may not be able to extend the inspection intervals beyond each refueling outage. Applicants that extend the inspection interval to longer than each refueling outage may need to provide trending of degraded and unqualified coatings and review operating experience for more than the previous two coating monitoring reports.

This is because an extension of the inspection interval may result in periods of time without inspections that are longer than the time period covered by the previous two refueling outages.

Additionally, an applicant may need to consider covering a time period greater than the proposed interval to provide margin for trending of coatings and to 2-105 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of Significant Changes Technical Basis for Change account for variations in degraded coatings recorded during a typical inspection.

RG 1.54, Revision 3, contains the most up-to-date NRC guidance on the selection, application, qualification, inspection, and maintenance of protective coatings applicable under GALL-SLR Report AMP XI.S8. For an applicant to demonstrate that an inspection interval of longer than every refueling outage is appropriate, it is necessary to identify aging effects such as blistering, cracking, flaking, peeling, rusting, and physical damage and to demonstrate acceptable historical coating performance. This is because coating degradation mechanisms can cause large amounts of coatings to become degraded/unqualified in time periods of less than 6 years (the maximum interval). The applicant will need to account for aging effects such as blistering, cracking, flaking, peeling, rusting, and physical damage for the containment coatings to demonstrate that the coating will be able to perform its safety function during all inspection intervals through the subsequent period of extended operation.

1 Table 2-35 Table 2-31 GALL-SLR Report, Revision 1, Chapter XI, Electrical Aging 2 Management Programs, Differences From GALL-SLR Report, Revision 0, 3 and Their Technical Bases Changed Program Elements Summary of Significant Changes Technical Basis for Change XI.E1: Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program Aging management program (AMP) is AMP is based on a visual inspection of Description clarified with an improved definition of an accessible cables and connections. Past adverse localized environment. An reviews and operating experience (OE)

Detection of Aging adverse localized environment is an indicated the management of cables Effects environment that exceeds the most specifically those that are coated with fire limiting environment (e.g., temperature, retardant material need to be visually radiation, or moisture) for the electrical inspected. The staff concluded that this insulation of cables that are coated with change should clarify the intent of the fire retardant material and connectors. program and provide additional guidance on the cable and electrical insulation 2-106 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Changed Program Elements Summary of Significant Changes Technical Basis for Change The Detection of Aging Effects program material coated with fire retardant adds, Cable and connection electrical material type subject to aging.

insulation are inspected to identify cable and connection insulation coated with fire retardant material installed in an adverse localized environment.

XI.E3A: Electrical Insulation for Inaccessible Medium-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program The proposed revisions add inspection of The staff finds that there is no need to Description manholes with water level monitoring perform annual inspections for manholes and alarms that result in consistent, that have an installed water level Scope of Program subsequent pump out of accumulated monitoring and alarm system if there are water prior to wetting or submergence of provisions for a timely response to level Preventative cable at least once every five years. alarms. Manholes with water level Actions Also, the proposed revisions add monitoring and alarms, and timely pump inspection of manholes following event- out, prevent water accumulation from Parameters driven occurrences such as heavy rain, wetting or submerging cables. There is Monitored or rapid thawing of ice and snow, or no adverse industry OEoperating Inspected flooding, only when water level experience for the level monitoring monitoring indicates water is equipment. Therefore, the staff finds that Acceptance accumulating. Based on the review of a inspecting manholes with installed water Criteria previous SLRA, manholes with water level monitoring and alarms every five level monitoring and alarms are self- years is acceptable. Additionally, Table XI-01 monitoring, and therefore do not require because of the level transmitters annual inspection for water continuous monitoring and alarms, there accumulation. is no need for event-driven inspections if there is no water accumulation.

Therefore, the staff finds acceptable a practice of inspecting manholes with water level monitoring and alarms following event-driven occurrences, only when the water level monitoring indicates water is accumulating. These water level monitoring systems are widely used in the industry, are very reliable, and can cope with a variety of operating conditions encountered in manholes at nuclear power plants. The water level monitoring system is self-monitoring. If it fails, indication will be shown in the control room. This proposed change provides continuous monitoring of water level in manholes rather than annual inspection of water level in manholes.

XI.E3B: , Electrical Insulation for Inaccessible Instrument and Control Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program The proposed revisions add inspection of The staff finds that there is no need to Description manholes with water level monitoring perform annual inspections for manholes 2-107 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Changed Program Elements Summary of Significant Changes Technical Basis for Change and alarms that result in consistent, that have an installed water level Scope of Program subsequent pump out of accumulated monitoring and alarm system if there are water prior to wetting or submergence of provisions for a timely response to level Preventative cable at least once every five years. alarms. Manholes with water level Actions Also, the proposed revisions add monitoring and alarms, and timely pump inspection of manholes following event- out, prevent water accumulation from Parameters driven occurrences such as heavy rain, wetting or submerging cables. There is Monitored or rapid thawing of ice and snow, or no adverse industry operating Inspected flooding only when water level experienceOE for the water level monitoring indicates water is monitoring equipment. Therefore, the Acceptance accumulating. Based on the review of a staff finds that inspecting manholes with Criteria previous SLRA, manholes with water installed water level monitoring and level monitoring and alarms are self- alarms every five years is acceptable.

Table XI-01 monitoring, and therefore do not require Additionally, because of the level annual inspection for water transmitters continuous monitoring and accumulation. alarms, there is no need for event-driven inspections if there is no water accumulation. Therefore, the staff finds acceptable a practice of inspecting manholes with water level monitoring and alarms following event-driven occurrences, only when the water level monitoring indicates water is accumulating. These water level monitoring systems are widely used in the industry, are very reliable and can cope with a variety of operating conditions encountered in nuclear power plant manholes. The water level monitoring system is self-monitoring. If it fails, indication will be shown in the control room. This proposed change provides continuous monitoring of water level in manholes rather than annual inspection of water level in manholes.

XI.E3C: , Electrical Insulation for Inaccessible Low-Voltage Power Cables Not Subject to 10 CFR 50.49 Environmental Qualification Requirements Program The proposed revisions add inspection of The staff finds that there is no need to Description manholes with water level monitoring perform annual inspections for manholes and alarms that result in consistent, that have an installed water level Scope of Program subsequent pump out of accumulated monitoring and alarm system if there are water prior to wetting or submergence of provisions for a timely response to level Preventative cable at least once every five years. alarms. Manholes with water level Actions Also, the proposed revisions add monitoring and alarms, and timely pump inspection of manholes following event- out, prevent water accumulation from Parameters driven occurrences such as heavy rain, wetting or submerging cables. There is Monitored or rapid thawing of ice and snow, or no adverse industry operating Inspected flooding, only when water level experienceOE for the level monitoring monitoring indicates water is equipment. Therefore, the staff finds that accumulating. Based on the review of a inspecting manholes with installed water 2-108 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Changed Program Elements Summary of Significant Changes Technical Basis for Change Detection of Aging previous SLRA, manholes with water level monitoring and alarms every 5five Effects level monitoring and alarms are self- years is acceptable.

monitoring, and therefore do not require Acceptance annual inspection for water Additionally, because of the level Criteria accumulation. transmitters continuous monitoring and alarms, there is no need for event-driven Corrective Actions inspections if there is no water accumulation. Therefore, the staff finds Table XI-01 an acceptable level of inspecting manholes with water level monitoring and alarms following event driven occurrences, only when the water level monitoring indicates water is accumulating. These water level monitoring systems are widely used in the industry, are very reliable, and can cope with a variety of operating conditions encountered in nuclear power plant manholes. The water level monitoring system is self-monitoring. If it fails, indication will be shown in the control room. This proposed change provides continuous monitoring of water level in manholes rather than annual inspection of water level in manholes.

XI.E7: , High-Voltage Insulators Program The proposed revisions add polymer The staff added polymer and toughened Description high-voltage (HV) insulators to the scope glass HV insulators to the scope and and program elements of GALL-SLR program elements of GALL-SLR AMP Scope of Program AMP XI.E7. The current AMP addresses XI.E7. Polymer and toughened glass HV porcelain insulators, however, polymer insulators are being used in some Parameters insulators have been utilized in some nuclear plant sites and are not currently Monitored or nuclear plant sites and should be discussed in GALL-SLR. Polymer HV Inspected addressed accordingly. Polymer HVhigh insulators include different voltage (HV) insulators include different material/environment and aging effects Detection of Aging material/environment and aging effects not previously considered in GALL-SLR Effects not previously considered in GALL-SLR and SRP-SLR. Adding polymer insulators and GALL-SRP. to this AMP enables use of this AMPits Acceptance use to manage aging of porcelain as well Criteria This also clarifies the scope of the as polymer HV insulators. Polymer HV insulators included under this program. insulators are typically composed of Table XI-01 Although the term high-voltage is used material such as fiberglass, silicone throughout AMP XI.E7, this program rubber (SIR), ethylene propylene rubber includes all insulators used in power (EPR), epoxy, silicone gel, sealants, systems operating at nominal system ductile iron, aluminum, aluminum alloys, voltages greater than 1 kV and equal to steel, steel alloys, malleable iron, and or less than 765 kV, and installed on in- galvanized metals. Exposure to air-scope portions of switchyards, outdoor can cause degradation and transmission lines, and power systems. aging effects that can result in reduced insulation resistance due to deposits and 2-109 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Changed Program Elements Summary of Significant Changes Technical Basis for Change surface contamination, reduced insulation resistance due to polymer degradation as well as loss of material caused by wind blowing on transmission conductors, all of which may require aging management. This component material/environment combination has not previously been evaluated in GALL-SLR and is considered a site-specific condition to be evaluated by the applicant.

Polymer HV insulators have been shown to have unique failure modes with little advance indications. Surface buildup of contamination can be worse for SIR (compared to porcelain insulators) due to absorption by silicone oil, especially in late stages of service life.

Typical aging degradation and mechanisms for polymer HV insulators include (but are not limited to) the following:

• Deposits and buildup of surface contamination causing reduced insulation resistance, arcing and flashover

• Polymer degradation caused by thermal degradation of organic material, radiolysis and photolysis of ultra-violet (UV )sensitive material, oxidation, and moisture intrusion

• Stress corrosion cracking of glass fibers due to sheath degradation

• Swelling of SIR layer due to chemical contamination

• Sheath wetting caused by chemicals absorbed by oil from SIR compound

• Brittle fracture of rods resulting from discharge activity, flash under, and flashover

• Chalking and crazing of insulator surfaces resulting in contamination, arcing, and flashover

• Water penetration through the sheath followed by electrical failure

• Bonding failure at rod and sheathing interface 2-110 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Changed Program Elements Summary of Significant Changes Technical Basis for Change

• Water ingress through end fittings causing flash under, corrosion, and fracture of glass fibers Additionally, aggressive environment due to presence of excrement from birds and rodents containing chemicals such as uric acid, phosphates, and ammonia can accelerate degradation.

This focus of this program is on certain classes of insulators commonly used in nuclear power plant applications, not on a particular voltage range definition. The term high-voltage insulator is recognized in the industry to apply to types of power conductor insulators used across a wide range of conductor voltages. Given that there are multiple standards that define voltage ranges (low, medium, high, extra high) differently, this AMP does not use any one definition but instead clarifies the specific voltage rating range that within the scope of this program.

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Draft Document: Tracked Changes Version Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 3 SUBSEQUENT LICENSE RENEWAL CHANGES TO STANDARD 2 REVIEW PLAN FOR REVIEW OF SUBSEQUENT LICENSE 3 RENEWAL APPLICATIONS FOR NUCLEAR POWER PLANTSSRP-4 LR, REVISION 2 AND THEIR TECHNICAL BASES 5 Many changes have been made to NUREG-2192, the Standard Review Plan for Review of 6 Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), Revision 0, 7 document. Some changes are the result of lessons learned and experience from the staffs 8 reviews of subsequent license renewal applications (SLRAs), including those from NRC 9 approved Interim Staff Guidance (ISG) documents. Additional changes are the result of public 10 comments that were received during the public comment period. Revision 1 of NUREG-2192 11 has consolidated these changes. This section provides a summary of notable technical changes 12 that were made in Revision 1 to the SRP-SLR and provides the technical basis for each change.

13 There are many changes that have been made to the Standard Review Plan for Review of 14 Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), Revision 0, 15 document. Some changes are the result of lessons learned and experience from the staffs 16 reviews of subsequent license renewal applications. Additional changes are the result of public 17 comments that were received during the public comment period. Revision 1 of NUREG-2192 18 has consolidated these changes. This section provides a summary of notable technical changes 19 that were made in Revision 1 to the SRP-SLR and provides the technical basis for each change.

20 The specific changes to each SRP-SLR chapter are discussed in Sections 3.13.1 through 21 3.53.5 of this document. A summary of the changes to each chapter and their technical bases 22 are presented in Tables Table 3-13-1 through Table 3-173-17.

23 3.1 SRP-SLR Chapter 1 - Administrative Information 24 There are no major technical changes in Chapter 1 of the SRP-SLR, Revision 0.

25 3.2 SRP-SLR Chapter 2 - Scoping and Screening 26 There were are no major technical changes to SRP-LR Chapter 2 of SRP-SLR, Revision 0, with 27 the exception of clarifications for complex assemblies and to the requirements of the Station 28 Blackout Rule. The changes and technical bases for this these changes areis shown in 29 Table 3-2Table 3-2.

30 3.3 SRP-SLR Chapter 3 - Aging Management Review 31 There are six subchapters to the SRP-SLR Chapter 3 on aging management review (AMR).

32 Subchapter 3.1 discusses aging management of reactor vessel (RV), internal, and reactor 33 coolant system. Subchapter 3.2 deals with aging management of engineered safety features; 34 Subchapter 3.3 covers auxiliary systems; Subchapter 3.4 discuses steam and power conversion 35 system; sSubchapter 3.5 discusses containments, structures, and component supports; and 36 subchapter Subchapter 3.6 discusses electrical and instrumentation and controls. The changes 37 and technical bases for these changes are shown in Table 3-3Tables 3-3 through Table 3-83-8, 38 respectively.

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Draft Document: Tracked Changes Version 1 3.4 SRP-SLR Chapter 4 - Time-Limited Aging Analyses (TLAAs) 2 There are seven subchapters to the SRP-SLR Chapter 4 on generic and plant-specific time-3 limited aging analyses (TLAAs). Subchapter 4.1 discusses how to recognize when a TLAA may 4 be appropriate, and changes to that subchapter are summarized in Table 3-9Table 3-9, along 5 with the technical bases for these changes. Subchapter 4.2 deals with reactor vesselRV neutron 6 embrittlement; subchapter Subchapter 4.3 covers metal fatigue; Ssubchapter 4.4 discusses the 7 environmental qualification of electrical equipment; Ssubchapter 4.5 presents a discussion of 8 concrete containment tendon prestress; Ssubchapter 4.6 discusses inservice local metal 9 containment corrosion analyses; and Ssubchapter 4.7 discusses other plant-specific safety 10 analyses that may involve other time-limited assumptions. The changes and technical bases for 11 these changes are shown in Table 3-9Tables 3-9 through Table 3-153-15, respectively.

12 3.5 SRP-SLR Appendices A.1, A.2, A.3, and A.4 13 There were no changes to the appendices in the SRP-SLR, Revision 0.

14 Table 3-1 SRP-SLR, Revision 1, Chapter 1, Section 1.1, Administrative Information, 15 and Section 1.2, Integrated Plants Assessments and Aging Management 16 Reviews Differences from SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of Significant Changes Technical Bases for Changes Section 1.1 Administrative Information No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

Section 1.2 Integrated Plant Assessments and Aging Management Reviews No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

17 Table 3-2 SRP-SLR, Revision 1, Chapter 2, Scoping and Screening, Differences from 18 SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change Table 2.1-2, Issue The guidance for the Complex The Statements of Consideration for Title 10 Complex Assemblies, issue in SRP-SLR, Code of Federal Regulations (10 CFR) Part Assemblies Table 2.1-2, Specific Staff Guidance 54 published in the Federal Register Notice on Scoping, was modified to clarify 60 FR 22461,of May 8, 1995, Nuclear the evaluation of complex Power Plant License Renewal; Revisions, assemblies, performed to identify the 60 FR 22461, states, in part, Passive parts structures and components (SCs) within the scope of subsequent of structures and components that only license renewal and subject to aging perform active functions do not require an management review (AMR). aging management review. The Structures and componentsSCs that perform both passive and active functions require an aging management reviewAMR for their intended passive function[s] only.

Accordingly, Table 2.2-2 was changed to indicate that if complex assemblies 3-2 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version performed passive intended functions (which meet the criteria of 10 CFR Part 54.4(a) to be included within the scope of license renewal), either solely or in addition to active intended functions, the passive intended functions should be evaluated to identity any SCs subject to AMR.

Table 2.1-6, Item New item in SRP-SLR Table 2.1-6 to Based on a review of current subsequent 128 state that the fire damper housing is license renewal applications, the staff noted subject to aging management. that AMR items have been included for fire damper assemblies or fire damper housings.

Fire damper assembly suggests the entire component (e.g., housing, damper) is subject to aging management while fire damper housing suggests only a portion of the component is subject to aging management. Therefore, clarification is needed regarding which components of a fire damper assembly are passive components and are subject to aging management.

NUREG-2192 defines passive structures and components as those that perform their intended functions without moving parts or change in configuration or properties in accordance with 10 CFR 54.21(a)(1)(i). The fire damper housing does not perform its intended function with moving parts; however, the other fire damper assembly components, including the damper, do perform their intended function with moving parts.

Treating the fire damper itself as an active component not subject to aging management is consistent with the treatment of other dampers. Specifically, 10 CFR 54.21(a)(1)(i) states that ventilation dampers are excluded from aging management, and SRP-SLR Table 2.1-6 states that only the housings of dampers, louvers, and gravity dampers associated with valves are subject to aging management.

Section 2.5.2.1.1 Added clarifying language to the Updated to conform with the requirements of requirements of the Station Blackout the SBO rule.

(SBO) Rule. Specifically, to address components within the scope of the SBO.

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Draft Document: Tracked Changes Version 1 Table 3-3 SRP-SLR, Revision 1, Chapter 3.1, Reactor Vessels, Internals, Coolant 2 System, Differences from SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change Section 3.1.3.5 Add reference to the FSAR This information was inadvertently omitted Final Safety Supplement information contained in when the FSAR Supplement information was Analysis Report GALL-SLR Table X-01 and Table XI- relocated from the SRP-SLR tables to the (FSAR) 01. The scope of this section was GALL-SLR tables.

Supplement expanded to include other types of cyclical loading analyses that may qualify as time-limited aging analysis (TLAAs) for these components, as defined in SRP-SLR Section 4.3. In addition, the further evaluation (FE)

Aacceptance criteriaCriteria and Rreview Pprocedure guidelines in SRP-SLR were amended to indicate that monitoring of cumulative usage factor (CUF) analyses for Class 1 components may be based on stress-based monitoring methods.

Section 3.1.2.2.10 The Standard Review Plan for Recent industry operating experience (OE)

Section 3.1.3.2.10 Subsequent License Renewal (SRP- indicates that significant wear can occur on Section 3.1.6 SLR, NUREG-2192, July 2017) the OD of piping due to system vibrations Table 3.1-1, Item Section 3.1.2.2.10, Loss of Material and interactions with certain types of 141 Due to Wear currently has Items 1 reflective metal insulation (RMI). Specifically, and 2, which address loss of material piping that uses RMI with an end cap of thin due to wear for pressurized water sheet metal has the potential for wear up to reactor (PWR) control rod drive 360 degrees around its circumference.

(CRD) head penetration nozzles and Repeated movement of the RMI end cap in stainless steel (SS) thermal sleeves contact with a pipe OD can cause loss of of PWR CRD head penetration material on the OD of the subject pipe.

nozzles, respectively. An additional Multiple instances of such wear are known to item, Item 3, is needed to address have occurred. An occurrence of this type of new industry operating experience material loss due to RMI end cap wear was related to loss of material due to reported in 2006, and was the subject of IN wear on the outside diameter (OD) of 2007-21, issued by the NRC on June 11, ASME Code Class 1 and Class 2 2007.

small-bore piping. Specifically, a reference is made to the NRC During outage activities in fFall of 2006, the Information Notice (IN) 2007-21, licensee at Catawba, Unit 1 removed RMI on Pipe Wear Due to Interaction of small-bore ASME Code Class 2 piping for a Flow-Induced Vibrations and planned valve replacement. The licensee Reflective Metal Insulation, identified multiple wear marks on the OD of Supplement 1, issued on December stainless steel piping. It was determined that 11, 2020. Its purpose was to alert the wear marks were the result of licensees of nuclear power reactors interactions between the stainless steelSS on recent operating experience piping and the stainless steelSS RMI end related to wear of nuclear power caps, caused by vibration. The licensee plant piping caused by flow-induced initially identified three locations with metal vibration and interaction of certain loss. During the extent of condition review, type of insulation. additional 81 discrete wear marks were identified over a 150-foot length of piping. All 3-4 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change of the wear marks were located inside containment and at insulation end caps.

In December of 2020 IN 2007-21 was revised due to more recent occurrences of metal loss on the OD of ASME Code Class 1 small-bore piping at two other nuclear power plants. The most recent known occurrence is summarized below.

During an outage in the spring of 2020, workers at Arkansas Nuclear One, Unit 2 identified multiple wear marks on American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME) Code Class 1 pressurizer spray piping. The wear marks were identified as a result of piping inspections in response to vibration related failures of snubber connections to the pressurizer spray piping. The wear marks ranged from surface scratches to deeper groves that were nearly 360 degrees around the OD circumference of the pipe.

The licensee determined that the wear was caused by vibration-induced interactions of the RMI end caps and the OD surfaces of the subject piping.

In the overview of the original IN 2007-21 and its subsequent supplement of 2020, it is apparent that the observed OD pipe wear for both the ASME Code Class 1 and 2 small-bore piping was discovered as a result of unrelated inspections. There are currently no specific ASME Code requirements to remove insulation from piping and inspect the piping for degradation due to RMI wear.

This type of wear, if present and undetected, could have a significant impact on the integrity of ASME Code Class 1 and 2 small-bore piping because: (1) small-bore piping has wall thickness values that are significantly less than those for large bore piping, (2) small-bore piping systems are more susceptible to vibration, and (3) there are no specific ASME Code requirements to inspect piping for RMI wear.

Based on the observed degradation, the licensees referenced in the updated IN 2007-21 have performed engineering evaluations as well as completed extent of 3-5 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change conditions and installed a modified RMI to eliminate the pipe wear. The modified insulation has an end cap as a piece of flat sheet metal that looks like a cuff, band or strip. The modified end cap touches the pipe as a flat piece of metal parallel to the pipe, not as a sharp edge; thereby eliminating the potential for excessive wear. Temporary modifications have also included installation of cuffs on the OD of the pipe where the end caps are located.

More recently, an applicant in the process of applying for a subsequent license renewal (SLR) reviewed the operating experience OE in the updated IN 2007-21, determined that the operating experience could be applicable at its units, and updated its aging management program to check for the RMI end cap wear on its ASME Code Class 1 small-bore piping (Ref. 40).

Relevant SRP-SLR sections are being updated to add a further evaluationFE to determine whetherif a plant seeking license renewal has evaluated the use of RMI in its integrated plant assessment and the potential of OD wear on its population of ASME Code Class 1 and 2 small-bore piping. Related references are also added in the reference section (SRP-SLR Section 3.1.6).

Table 3.1-1, Item The staff made the following changes The Electric Power Research Institute 028 to SRP-SLR Table 3.1-1, Item 028 in (EPRI) MRP revised its aging management SLR-ISG-2021-01-PWRVI: criteria for Westinghouse-design control rod guide tube (CRGT) split pins and CE-design

1) 1) Deleted reference of thermal shield positioning pins and incore Combustion Engineering (CE)-design monitoring instrument (IMI) thimble tube in components (i.e., thermal shield the MRP-227, Rev. 1-A reportMRP-227, repositioning pins and incore Revision 1-A Report. Therefore, the staff instrumentation [(ICI]) thimble tubes) modified its AMR criteria for Westinghouse-from the scope of the Item 028a and design CRGT split pins in AMR Item 028 and realigned the aging management deleted reference of CE-design reactor review (AMR) line items for the vessel internal (RVI) thermal shield components as described in the positioning pins and incore instrument (ICI)

Technical Basis for Changes thimble tubes components from the scope of column entry for this line item. AMR Item 028.

2) Edited the component For the AMR criteria that apply to the description to clarify it only applies to Westinghouse-design CRGT support pins Westinghouse-design control rod (split pins), the staff modified Item 028 to allow application and use of GALL-SLR Item 3-6 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change guide tube (CRGT) support pins (split IV.B2.RP-355 if the CRGT split pins in the pins). plant design have yet to be replaced and are made from X-750 nickel alloy materials.

3) Added parenthetical statements Otherwise, the modified version of SRP-SLR to the GALL-SLR Item references in Table 3.1-1, Item 028 allows use of GALL-the GALL-SLR column entry of Item SLR Item IV.B2.RP-265 (i.e., No Additional 028 that clarify when the GALL-SLR Measures) if the CRGT split pins have been items may be applied for use in an replaced and are made from austenitic incoming SLRA. stainless steelSS materials, and can be placed in the No Additional Measures

4) Administratively edited the item category of components. The modified to cite the irradiation-assisted stress version Item 028 also allows use of GALL-corrosion cracking mechanism as SLR Item IV.E.R-444 if the CRGT split pins IASCC. are defined in the CLB as ASME Section XI Code -Class components and the ISI

5) Deleted GALL-SLR Items program is credited for aging management IV.B2.RP-356, IV.B3.RP-357, and of the pins. The changes to Item 28 for IV.B3.RP-400 as referenced GALL- Westinghouse-design CRGT spilt pins SLR Items in Item 028. should make the AMR criteria consistent with those in MRP-227, Rev.Revision 1-A.

The EPRI MRP changed the inspection categories for the referenced CE-design thermal shield positioning pins and ICI thimble tubes in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

Specifically, the EPRI MRP downgraded the CE-design thermal shield repositioning pins to No Additional Measures components in the updated report. Therefore, based on that change, the staff realigned its AMR criteria for the thermal shield positioning pins to the AMR line items for CE-design No Additional Measures components, which are given in SRP-SLR Table 3.1-1, Item 055b and in GALL-SLR Item IV.B2.RP-306. Additionally, EPRI MRP has identified that CE-design ICI thimble tubes are Existing Program category components per the line item entry for these components in Table 3-2 (page 3-

26) of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. In this table, EPRI MRP identifies that the ICI thimble tubes are susceptible to the aging mechanism of wear.

The existing item in GALL-SLR IV.B3.RP-357 is consistent with this basis, with the exception that it is now appropriate to the reference of the RP-357 item as being aligned to SRP-SLR Item 056c (the item for CE-design Existing Program components that are subject to non-cracking aging effect and mechanism combinations), and not Item 028.

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Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change IASCC was included in Item 028 based on lessons learned from the Surry SLRA gap analysis, which referenced inclusion of IASCC as an applicable mechanism for the CRGT split pins per criteria in the MRP-2018-022 reportReport.

Table 3.1-1, Item The staff deleted SRP-SLR Table In the SRP-SLR and GALL-SLR rReports, 032 3.1-1, Item 032 in SLR-ISG-2021 SRP-SLR Table 3.1-1, Item 032, and the PWRVI. linked items in GALL-SLR Items IV.B2.RP-382, IV.B3.RP-382 and IV.B4.RP-382 provided the SRP-SLR and GALL-SLR AMR line items for Westinghouse-designed, Combustion Engineering (CE)-designed, and Babcock and Wilcox (B&W)-designed reactor internals that are categorized as ASME Section XI Code Class components.

During the staff ISG review, the staff determined that Item 032 is bounded by and redundant with the scope of the AMR in SRP-SLR Table 3.1-1, Item 114, as updated in SLR-ISG-2021-01-PWRVI. Similarly, the staff determined that the scope of GALL-SLR Items IV.B2.RP-382, IV.B3.RP-382 and IV.B4.RP-382 are bounded by and redundant with the scope of the AMR in GALL-SLR Item IV.E.R-444, as updated in SLR-ISG-2021-01-PWRVI. Therefore, SRP-SLR Table 3.1-1, Item 032 and the corresponding GALL-SLR RP-382 items were deleted in SLR-ISG-2021-01-PWRVI.

For more details, refer to the comment and technical basis for changes being made to SRP-SLR Table 3.1-1, Item 114 in this table.

Table 3.1-1, Item The staff made the following changes SRP-SLR Table 3.1-1, Item 051a remains as 051a to SRP-SLR Table 3.1-1, Item 051a AMR line item for Babcock and Wilcox in SLR-ISG-2021-01-PWRVI: (B&W) Primary category reactor vessel internal (RVI) components subject to

1) 1) Added GALL-SLR Item cracking.

IV.B4.RP-252c as a new referenced GALL-SLR item in Item 051a. 1) In the staff-approved basis in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A

2) Deleted GALL-SLR Items Report, the EPRI MRP designated specific IV.B4.RP-249a, IV.B4.RP-252a, B&W-design vent valve original locking IV.B4.RP-258a, IV.B4.RP-259a, and devices and modified locking devices as new IV.B4.RP-400 as referenced GALL- Primary category component per Items B5 SLR items in Item 051a. and B6 in Table 4-1 of the report. Since the staff has developed new GALL-SLR Item

3) Administratively edited Item 051a IV.B4.RP-252c to address cracking of these to cite the irradiation-assisted stress Primary category locking devices, this required addition of the new GALL-SLR Item 3-8 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change corrosion cracking mechanism as as an additional referenced item for SRP-IASCC. SLR Table 3.1-1, Item 051a.

2) Based on acceptance of EPRI MRP Ccomments #29 and #32, the staff confirmed that the B&W- design RVI components that were previously within the scope of GALL-SLR Items IV.B4.RP-249a, IV.B4.RP-252a, IV.B4.RP-258a, and IV.B4.RP-259a did not screen in for any cracking mechanisms in MRP-227, Rev.Revision 1-A. Therefore those RP line items were deleted or completely revised as non-cracking items in Appendix B.3 of the ISG and the referencing of these GALL-SLR item was either deleted from Item 051a or deleted from 051a and moved as a referenced item to one of non-cracking SRP-SLR items for B&W- design components (i.e., either SRP-SLR Table 3.1-1 Item 058a or 058b) in Appendix A of the ISG. For example in the ISG, the staff modified the existing IV.B4.RP-252a item to address loss of fracture toughness due to thermal aging embrittlement in B&W- design vent valve bodies instead of cracking in B&W- design core support shield (CSS) vent valve top and bottom retaining rings, which was the topic of the previous version of the RP-252a item. However, the vent valve bodies are B&W- design Expansion category components per item B2.1 in Table 4-4 of MRP-227, Rev.Revision 1-A. Therefore, the staff deleted the IV.B4.RP-252a item as a referenced item in SRP-SLR Table 3.1-1, Item 051a and instead added it as a new reference item for SRP-SLR Table 3.1-1, Item 058b, as updated in Appendix A of the ISG (i.e., Item 058b is the proper item in SRP-SLR Table 3.1-1 for B&W-design Expansion category components in MRP-227, Rev.Revision 1-A that are subject to non-cracking effect and mechanism combinations, including the vent valve bodies).

The previous inclusion of GALL-SLR Item IV.B4.RP-400 in the GALL-SLR Rreport only applied to cracking of specific B&W-designed RVI components that were included in the design of the Three Mile Island Unit 1 (TMI-1) reactor. The owner of TMI-1 has made the decision to decommission the reactor. Therefore, GALL-3-9 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change SLR Item IV.B4.RP-400 is no longer needed for the objectives of the updates in SLR-ISG-2021-01-PWRVI and has been deleted. The staff adjusted AMR Item 051a accordingly to delete Item IV.B4.RP-400 as a referenced GALL-SLR item for the line item.

3) The change to cite the irradiation-assisted stress corrosion cracking mechanism as IASCC is an administrative edit for consistency with other AMR line items that cite IASCC as an aging mechanism.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Item 051b 051b to SRP-SLR Table 3.1-1, Item 051b remains as the AM line item for B&W in SLR-ISG-2021-01-PWRVI: Expansion category RVI components subject to cracking.

1) Added GALL-SLR Items IV.B4.RP-246c and IV.B4.RP-246d 1) 1) In Item B7.1 in Table 4-4 of the as new referenced GALL-SLR items MRP-227, Rev. 1-A reportMRP-227, in Item 051b. Revision 1-A Report, the EPRI MRP identifies that the upper thermal shield (UTS)

Per the staffs acceptance of Nuclear bolts and their locking devices are Energy Institute (NEI) Comment #3, Expansion category components that are and similar generic comment bases located in the core barrel assemblies of in EPRI MRP Comments #29 and B&W-designed PWRs. The prior GALL-SLR

1. 32, and the staffs bases for IV.IV.B4-RP-246 and IV.B4.RP-246a items resolving these comments, the Final covering cracking in both the lower thermal version of the ISG no longer includes shield (LTS) bolt and bolt locking devices a new GALL-SLR IV.B4.RP-375 item and UTS bolt and bolt locking devices on the subject of cracking in B&W- indicated that all of the components are design lower grid rib sections. located in the lower grid assembly of the plants. As a result of these MRP-227,

2) Deleted GALL-SLR Items Rev.Revision 1-A changes, the staff deleted IV.B4.RP-244a, IV.B4.RP-250a, reference of the UTS bolt and bolt locking IV.B4.RP-254 and IV.B4.RP-254a as devices from the scope of existing Item referenced GALL-SLR items in Item IV.B4.RP-246 and IV.B4.RP-246a items, and 051b. instead, developed new GALL-SLR Items IV.B4.RP-246c and IV.B4.RP-246d to

3) Administratively edit the line item address cracking of UTS bolts and UTS bolt to cite the irradiation-assisted stress locking devices, as located in the proper corrosion cracking mechanism as core barrel assembly containing the IASCC. components. The staff adjusted SRP-SLR Table 3.1-1, Item 051b accordingly to include Items IV.B4.RP-246c and IV.B4.RP-246d as new GALL-SLR Item references for the line item.

Based on the staffs partial acceptance of NEI #3, and similar generic comments made in EPRI MRP Comments #29 and #32, the staff confirmed that the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report did not 3-10 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change screen B&W-design lower grid rib sections in for any cracking mechanisms. Thus, the IV.B4.RP-375 Iitem, as previously proposed in the draft ISG for cracking in the lower grid rib sections, is not being included as a new item for the final version of the ISG and is not referenced as a GALL-SLR item reference for the revision of SRP-SLR Table 3.1-1, Item 051b in the ISG.

2) Based on the staffs partial acceptance of NEI #3, and similar generic comments made in EPRI MRP Comments #29 and #32, the staff confirmed that the MRP-227, Rev.

1-A reportMRP-227, Revision 1-A Report did not screen B&W-design external baffle-to-baffle bolts, core barrel-to-former bolts, core barrel cylinders, or former plates in for any cracking mechanisms. Therefore, the staff deleted GALL-SLR Items IV.B4.RP-244a and IV.B4.RP-250a in Appendix B.3 of SLR-ISG-2021-01-PWRVI and as referenced GALL-SLR items for the update of AMR Item 051b in the ISG.

The previous inclusion of GALL-SLR Items IV.B4.RP-254 and IV.B4.RP-254a in the GALL-SLR Rreport only applied to cracking of specific B&W-designed lower grid assembly bolts and bolt locking devices that were included in the design of the Three Mile Island Unit 1 (TMI-1) reactor. The owner of TMI-1 has made the decision to decommission the reactor. Therefore, the staff deleted GALL-SLR Items IV.B4.RP-254 and IV.B4.RP-254a in Appendix B.3 of SLR-ISG-2021-01-PWRVI and as referenced GALL-SLR items for the update of AMR Item 051b in in the ISG.

3) The change to cite the irradiation-assisted stress corrosion cracking mechanism as IASCC is an administrative edit for consistency with other AMR line items that cite IASCC as an aging mechanism.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Items 052a and 052a to SRP-SLR Table 3.1-1, Items 052a 052b remain as the AMR line items for Table 3.1-1, Item and 052b in SLR-ISG-2021 Combustion Engineering (CE) Primary and 052b PWRVI: Expansion category RVI components that may be subject to cracking.

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Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change

1) Moved the reference of GALL- 1) In the staff-approved basis in the MRP-SLR Item IV.B3.RP-363 from SRP- 227, Rev.Revision 1 report, the EPRI MRP SLR Table 3.1-1, Item 052a to SRP- amended the inspection category for CE-SLR Table 3.1-1, Item 052b. design lower support structure core support columns from Primary category

2) Deleted Item IV.B3.RP-326a as a components (as designated in Table 4-2 of referenced GALL-SLR item for Item the MRP-227-A report) to Expansion 052a. category components, as indicated in Item C6.3 of Table 4-5 in the MRP-227, Rev. 1-A

3) Administratively edited the 052a reportMRP-227, Revision 1-A Report. As a and 052b line items to cite the result of this component category irradiation-assisted stress corrosion designation change, the staff amended cracking mechanism as IASCC. GALL-SLR Item IV.B3.RP-363 to link its SRP-SLR item reference to that in SRP-SLR Table 3.1-1, Item 052b (and not to Item 052a) and adjusted the references of GALL-SLR Item IV.B3.RP-363 in SRP-SLR Table 3.1-1, Items 052a and 052b accordingly by moving the reference of the RP-363 item from Item 052a to Item 052b.

2) Based on the staffs partial acceptance of NEI Comment #3, the staff confirmed that the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report did not screen CE-design core shroud assembly components (for CE plant designs with core shrouds that are assembled in two vertical sections) in for any cracking mechanisms. Thus, IV.B3.RP-326a Iitem, as previously as previously included in the GALL-SLR Rreport for cracking in these shroud components, is being deleted in Appendix B.2 of the final version of the ISG.

3) The change to cite the irradiation-assisted stress corrosion cracking mechanism as IASCC is an administrative edit for consistency with other AMR line items that cite IASCC as an aging mechanism.

Table 3.1-1, Item The staff made the following changes SRP-SLR Table 3.1-1, Item 052c remains as 052c to SRP-SLR Table 3.1-1, Item 052c the AMR line item for CE Existing Program in SLR-ISG-2021-01-PWRVI: category RVI components subject to cracking.

1) Added GALL-SLR Item IV.B3.RP-320a as a new GALL-SLR In Item C17 of Table 4-8 in the MRP-227, item reference in SRP-SLR Table Revision 1-A reportRevision 1-A Report, the 3.1-1, Item 052c EPRI MRP added the core stabilizing lugs and shims (and their associated bolts) as

2) Administratively edited the line Existing Program components for plants item to cite the irradiation-assisted with CE-designed reactor internals, with the 3-12 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change stress corrosion cracking applicable aging effect being cited as mechanism as IASCC. cracking due to stress corrosion cracking (SCC). The staff developed new GALL-SLR Item IV.B3.RP-320a to be consistent with this basis. The staff adjusted SRP-SLR Table 3.1-1, Item 052c to include Item IV.B3.RP-320a as a new GALL-SLR item reference for the line item.

Other designated CE-design Existing Program components (e.g., the fuel alignment pins per Item C15a in MRP-227, Rev.Revision 1-A, Table 4-8 and GALL-SLR Item IV.B3.RP-334) have been identified as being susceptible to the cracking mechanisms of stress corrosion cracking (SCC), IASCC, or fatigue. The 052c line item cites all of these cracking mechanisms, which is appropriate for the generic basis of the line item.

The change to cite the irradiation-assisted stress corrosion cracking mechanism as IASCC is an administrative edit for consistency with other AMR line items that cite IASCC as an aging mechanism.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Item 053a 053a to SRP-SLR Table 3.1-1, Item 053a remains as the AMR line item for in SLR-ISG-2021-01-PWRVI: Westinghouse Primary category RVI components subject to cracking.

1) Moved the reference of GALL-SLR Item IV.B2.RP-280 from SRP- 1) Per the criteria in Items W3.1, W3.2, and SLR Table 3.1-1, out of Item 053a W3.3 of Table 4-6 in MRP-227, and into Item 053b. Rev.Revision 1-A, the EPRI MRP designated Westinghouse-design core barrel

2) Added GALL-SLR Iitem lower flanges welds (LFWs), upper IV.B2.RP-296a as a new GALL-SLR circumferential (girth) welds (UGWs), and item reference in SRP-SLR Table upper vertical (axial) welds (UAWs) as 3.1-1, Item 053a. Expansion category components for the programs. As a result, the staff amended

3) Administratively edited the line GALL-SLR Item IV.B2.RP-280 to be item to cite the irradiation-assisted consistent with the EPRI MRPs Expansion stress corrosion cracking category criteria for the LFWs, UGWs and mechanism as IASCC. UAWs in MRP-227, Rev.Revision p 1-A and realigned the GALL-SLR item from SRP-SLR Table 3.1-1, Item 053a to Iitem 053b.

The staff adjusted the reference of GALL-SLR Item IV.B2.RP-280 in SRP-SLR Table 3.1-1 from Item 053a to Item 053b accordingly.

2) The staff developed new GALL-SLR Item IV.B2.RP-296a to address potential 3-13 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change cracking that may occur in control rod guide tube (CRGT) assembly guide plates (guide cards). Although the guide cards were not identified as being susceptible to cracking in Table 4-3 of MRP-227, Rev.Revision 1-A, the components were screened in for fatigue as part of the Surry SLRA , as referenced to EPRIs 80-year Expert Panel assessment for the components in MRP-2018-022 report.

As a result, the staff adjusted SRP-SLR Table 3.1-1, Item 053a accordingly to include Item IV.B2.RP-296a as a new GALL-SLR item reference for the line item. Based on the lessons learned from the Surry SLRA, the staff did not accept EPRI MRP Comments #1 and #8 that fatigue should not be screened in as a cracking mechanism for CRGT guide cards, as cited in the new RP-296a item for the guide cards.

3) The change to cite the irradiation-assisted stress corrosion cracking mechanism as IASCC is an administrative edit for consistency with other AMR line items that cite IASCC as an aging mechanism.

Table 3.1-1, Item The staff made the following changes SRP-SLR Table 3.1-1, Item 053b remains as 053b to SRP-SLR Table 3.1-1, Item 053b the AMR line item for Westinghouse in SLR-ISG-2021-01-PWRVI: Expansion category RVI components subject to cracking.

1) Moved reference of GALL-SLR Item IV.B2.RP-280 into Item 053b 1) In Items W3.1, W3.2, and W3.3 of Table from its previous referenced location 4-6 in the MRP-227, Rev. 1-A reportMRP-in Item 053a. 227, Revision 1-A Report, the EPRI MRP designated Westinghouse-design core barrel

2) Added GALL-SLR Iitem lower flanges welds (LFWs), upper IV.B2.RP-298a as a new GALL-SLR circumferential (girth) welds (UGWs), and item reference in SRP-SLR Table upper vertical (axial) welds (UAWs) as 3.1-1, Item 053b. Expansion category components for the programs. Based on these changes, the

3) Deleted GALL-SLR Item staff amended GALL-SLR Item IV.B2.RP-IV.B2.RP-278 as a referenced GALL- 280 to be consistent with the EPRI MRPs SLR item in SRP-SLR Table 3.1-1, Expansion category criteria for the LFWs, Item 053b. UGWs and UAWs in MRP-227, Rev.Revision 1-A and adjusted the

4) Administratively edited the line reference of GALL-SLR Item IV.B2.RP-280 item to cite the irradiation-assisted in SRP-SLR Table 3.1-1 from Item 053a to stress corrosion cracking Item 053b accordingly.

mechanism as IASCC.

2) In Item W2 of Table 4-3 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP only assigned the LFWs in peripheral (outer) CRGT 3-14 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change assemblies of Westinghouse-designed PWRs as Primary category components, with the inspections expanding to the LFWs in the non-peripheral (remaining) assemblies (as the designated Expansion components) per Item W2.1 in Table 4-6 of MRP-227, Rev.Revision 1-A) if unacceptable degradation was detected in the peripheral CRGT assembly LFWs. The previous version of the GALL-SLR Rreport did not include a line item for cracking of the non-peripheral CRGT LFWs. Therefore, the staff developed new GALL-SLR Iitem IV.B2.RP-298a to address cracking in the non-peripheral CRGT assembly LFWs and adjusted SRP-SLR Table 3.1-1, Item 053b accordingly to cite new GALL-SLR Item IV.B2.RP-298a item as a new GALL-SLR item reference for Item 053b.

3) In MRP-227-A, the core barrel outlet nozzle welds (ONWs) covered by GALL-SLR Item IV.B2.RP-278 were designated as the Expansion components for Primary inspections performed on the core barrel upper flange weld (UFW). However, in MRP-227, Rev.Revision 1-A, the EPRI MRP deleted the ONWs as Expansion components and replaced them with the core barrel assembly upper girth weld (UGW), lower flange weld (LFW), upper axial welds (UAWs), and lower support forging or casting as the applicable new Expansion components for Primary UFW inspections.

Therefore, the staff deleted GALL-SLR Item IV.B2.RP-278 in SLR-ISG-2021-01-PWRVI and the reference of the GALL-SLR Item IV.B2.RP-278 from Item 053b.

4) The change to cite the irradiation-assisted stress corrosion cracking mechanism as IASCC is an administrative edit for consistency with other AMR line items that cite IASCC as an aging mechanism.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Items 053c and 053c to SRP-SLR Table 3.1-1, Items 053c 059c remain as the AMR line items for Table 3.1-1, Item and 059c in SLR-ISG-2021 Westinghouse Existing Program category 059c PWRVI: RVI components subject to cracking or non-cracking effects.

1) Added stellite as an additional potential material of fabrication for 3-15 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Westinghouse-design Existing 1) Stellite has been added as an additional Program components in SRP-SLR material for SRP-SLR Table 3.1-1, Item 059c Table 3.1-1, Item 059c. in order to account for the possibility that the materials used in the design of some reactor

2) Added GALL-SLR Item internals may have an outside, wear -

IV.B2.RP-345a as a new GALL-SLR resistant surface layer made from stellite.

reference for SRP-SLR Table 3.1-1, For example, in Dominion Energy Item 053c. Companys subsequent license renewal application for Surry Nuclear Plant, Units 1

3) Deleted Item IV.B2.RP-355 as a and 2, the licensee identified that clevis GALL-SLR item reference in SRP- inserts and fuel alignment pins in the units SLR Table 3.1-1, Item 053c. included an outside, wear -resistant stellite layer on the nickel alloy or stainless steelSS

4) Administratively edited the 053c materials used to fabricate the components.

line item to cite the irradiation- As a result, the staff adjusted SRP-SLR assisted stress corrosion cracking Table 3.1-1, Items 053c and 059c to include mechanism as IASCC. stellite as a potential material of fabrication for the line item.

2) The staff developed new GALL-SLR Item IV.B2.RP.345a as part of the ISG update efforts to address potential cracking in the Westinghouse-design core barrel flanges similar to the manner that GALL-SLR Item IV.B2.RP-345 is used to address loss of material due to wear on the flanges. In the Appendix C gap analysis results for the Surry subsequent license renewal application (SLRA), Dominion identified that stress corrosion cracking (SCC) and fatigue were applicable cracking mechanisms for the flanges based on the Expert Panel basis in the MRP-2018-022 report. As a result of these criteria, the staff determined it was prudent to develop the new GALL-SLR Item IV.B2.RP-345a item to address cracking in the core barrel flange components. The staff adjusted SRP-SLR Table 3.1-1, Item 053c accordingly to include Item IV.B2.RP-345a as a new GALL-SLR item reference for the line item. Based on the Surry SLRA lessons learned, the staff did not accept EPRIs basis in EPRI MRP Comments #2 and #10 that the new RP-345a item on cracking of Westinghouse core bare flanges is not appropriate for the ISG.

3) The staff deleted GALL-SLR Item IV.B2.RP-355 as a referenced item for SRP-SLR Table 3.1-1, Item 053c because it was redundant with the existing referencing of GALL-SLR Item IV.B2.RP-355 in SRP-SLR Table 3.1-1, Item 028. As modified in the 3-16 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change ISG, SRP-SLR Table 3.1-1, Item 028 remains as the applicable SRP-SLR line item for addressing cracking and loss of material due to wear in Westinghouse-designed control rod guide tube support (CRGT) pins (split pins), which in turn links to use of the GALL-SLR RP-355 item if the split pins made from X-750 nickel alloy materials. Refer to the technical basis statement in this table for Item 028 for additional information.

4) The change to cite the irradiation-assisted stress corrosion cracking mechanism as IASCC in Item 053c is an administrative edit for consistency with other AMR line items that cite IASCC as aging mechanism.

Table 3.1-1, Item The staff only made a minor The staffs change is strictly administrative.

054 administrative edit of SRP-SLR Table The technical bases for managing loss of 3.1-1, Item 054 in SLR-ISG-2021 material due to wear in Westinghouse-PWRVI to clarify that the line item design BMI flux thimble tubes, as defined in and GALL-SLR Item IV.B2.RP-284 GALL-SLR Table 3.1-1, Item 054 and GALL-are only applicable to the bottom SLR Item IV.B2.RP-284, remain the same as mounted instrumentation (BMI) flux defined in the NUREG-2191 and NUREG-thimble tubes in Westinghouse- 2192 reports and explained in NUREG-designed PWRs. 2221.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Items 056a and 056a to SRP-SLR Table 3.1-1, Items 056a 056b remain as the AMR line items for CE Table 3.1-1, Item and 056b in SLR-ISG-2021 Primary and Expansion category RVI 056b PWRVI: components that are subject to non-cracking effect and mechanism combinations.

1) Moved the reference of GALL-SLR Item IV.B3.RP-364 from SRP- 1) In the staff-approved basis in the MRP-SLR Table 3.1-1, Item 056a to Item 227, Rev.Revision 1 report, the EPRI MRP 056b. amended the inspection category for CE-design lower support structure core support

2) Added GALL-SLR Iitem columns from Primary category (as IV.B3.RP-338a as a new referenced designated in Table 4-2 of the MRP-227-A GALL-SLR item for SRP-SLR Table report) to Expansion category, as indicated 3.1-1, Item 056a. in Item C6.3 of Table 4-5 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A

3) Added GALL-SLR Item Report. As a result of this component IV.B3.RP-333a as a new referenced category designation change, the staff has GALL-SLR items for SRP-SLR Table amended GALL-SLR Item IV.B3.RP-364 to 3.1-1, Item 056b. link its SRP-SLR item reference to that in SRP-SLR Table 3.1-1, Item 056b. Similarly, staff adjusted the references of GALL-SLR Item IV.B3.RP-364 in SRP-SLR Table 3.1-1 accordingly by moving the reference of the RP-364 item from Item 056a to Item 056b.

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Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change

2) The staff developed new GALL-SLR Item IV.B3.RP-338a as part of the ISG update efforts. For CE plants with welded core shroud designs that utilize full height shroud plates, the Primary category fuel alignment plate in the upper internals assembly screened in for cracking (fatigue) and loss of fracture toughness (irradiation embrittlement [IE]) per Item C10 in Table 4-2 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. Management of cracking in the fuel alignment plate is addressed by the existing AMR in GALL-SLR Item IV.B3.RP-338. The staff added GALL-SLR Item IV.B3.RP-338a in ISG Appendix B.2 to address management of loss of fracture toughness due to neutron irradiation embrittlementIE in the fuel alignment plate. The staff adjusted SRP-SLR Table 3.1-1, Item 056a accordingly to include Item IV.B.RP-338a as a new GALL-SLR item reference for the plates.

3) The staff developed new GALL-SLR Item IV.B3.RP-333a as part of the ISG update efforts. For CE core support barrel lower girth welds (LGWs), the Expansion category LGWs screened in for cracking (SCC, IASCC, and fatigue) and loss of fracture toughness (irradiation embrittlement

[IE]) per Item C5.1 in Table 4-5 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report. Management of cracking in the core support barrel LGWs is addressed by the modified AMR in GALL-SLR Item IV.B3.RP-333. The staff added GALL-SLR Item IV.B3.RP-338a in ISG Appendix B.2 to address management of loss of fracture toughness due to neutron irradiation embrittlementIE in the LGWs. The staff adjusted SRP-SLR Table 3.1-1, Item 056b accordingly to include Item IV.B.RP-333a as a new GALL-SLR item reference for the LGWs.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Item 056c 056c to SRP-SLR Table 3.1-1, Item 056c remains as the AMR line item for CE in SLR-ISG-2021-01-PWRVI: Existing Program category RVI components that are subject to non-cracking

1) Moved the reference of GALL- effect and mechanism combinations.

SLR Item IV.B3.RP-357 from SRP-SLR Table 3.1-1, Item 028 to Item 1) The staffs technical bases for moving 056c. the referencing of GALL-SLR Item IV.B3.RP-3-18 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change 357 from SRP-SLR Table 3.1-1 Item 028 to

2) Deleted reference of GALL-SLR Item 056c have been adequately addressed Item IV.B3.RP-334a from SRP-SLR and discussed in the technical basis entry for Table 3.1-1, Item 056c. changes made to SRP-SLR Table 3.1-1, Item 028 in this table. The criteria relate to the updated I&Einspection and evaluation criteria for CE-design incore instrument (ICI) thimble tubes (lower) in the MRP-227, Rev.

1-A reportMRP-227, Revision 1-A Report, which screened in for the mechanism of wear in Table 3-2 of the report.

2) In SLR-ISG-2021-01-PWRVI, the staff deleted GALL-SLR Item IV.B3.RP-334a based on the staffs decision to fold the CE plant design applicability statement in the RP-334a item (i.e., the prior line item applied to CE plants with welded full height shroud designs) into GALL-SLR Item IV.B3.RP-336 (which applied to CE plants with welded shrouds fabricated from two vertical shroud sections). The modification to the component description and the applicable shroud types in the RP-336 item now accounts for the component description and applicable shroud types cited in the prior GALL-SLR Item IV.B3.RP-334a. These changes to the RP-336 item permitted the staff to delete the RP-334a item due to its redundancy of the modified RP-336 item.

The staff adjusted SRP-SLR Table 3.1-1, Item 056c accordingly to delete reference of the RP-334a item.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Item 058a 058a to SRP-SLR Table 3.1-1, Item 058a remains as the AMR line item for B&W in SLR-ISG-2021-01-PWRVI: Primary category RVI components that are subject to non-cracking aging effect and

1) Added GALL-SLR Item mechanism combinations.

IV.B4.RP-247c as a new referenced GALL-SLR item for Item 058a. 1) The staff developed new GALL-SLR Item IV.B4.RP-247c as part of the ISG

2) Added GALL-SLR Item update efforts in order to be consistent with IV.B4.RP-252b as a new referenced Item B8 in Table 4-1 of the MRP-227, GALL-SLR item for Item 058a. Revision 1-A reportRevision 1-A Report for B&W-design, Primary category lower core

3) Deleted GALL-SLR Item barrel (LCB) bolts. In Item B8, EPRI IV.B4.RP-401 as a referenced GALL- screened the LCB bolts in for irradiation-SLR item for Item 058a enhanced creep/stress relaxation (ISR/IC),

wear, stress corrosion cracking (SCC), and fatigue aging mechanisms. Cracking of the LCB bolts is being addressed in the ISG by the staffs modification of the GALL-SLR 3-19 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Item IV.B4.RP-247 item in the ISG. The staff developed the new GALL-SLR Item IV.B4.RP-247c to address loss of material due to wear and loss of preload due to ISR/IC in the LCB bolts. SRP-SLR Table 3.1-1, Item 058a was administratively edited to reference the new GALL-SLR Item IV.B4.RP-247c item.

2) The staff developed new GALL-SLR Item IV.B4.RP-252b as part of the ISG update efforts in order to be consistent with Items B4 and B5 in Table 4-1 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report for various types of original locking devices (Primary components) located in the vent valve assemblies of B&W-designed reactors. For the original pressure plate, spring retainer, spring, and U-cover components, EPRI screened the components in for wear; for the original key, ring, and pin components, EPRI screened the components in for thermal embrittlement (TE). Thus, the new GALL-SLR Item IV.B4.RP-252b item addresses loss of material due to wear in the original pressure plate, spring retainer, spring, and U-cover components and loss of fracture toughness due to TE in the original key, ring, and pin components. The staff adjusted SRP-SLR Table 3.1-1, Item 058a to reference the GALL-SLR Item IV.B4.RP-252b item as a new referenced GALL-SLR item for Item 058a.

3) The staffs previous inclusion of GALL-SLR Item IV.B4.RP-401 in the GALL-SLR Rreport only applied to the management of non-cracking effects in specific types of B&W-designed RVI components that were included in the design of the Three Mile Island Unit 1 (TMI-1) reactor and that were designated by the EPRI MRP as Primary category components for TMI-1. The owner of TMI-1 has made the decision to decommission the reactor. Therefore, the staff no longer needed GALL-SLR Item IV.B4.RP-401 item for the purposes of the updates in SLR-ISG-2021-01-PWRVI and the staff has deleted the IV.B4.RP-401 item in Appendix B.3 of the ISG. The staff adjusted the SRP-SLR Table 3.1-1 058a item accordingly by deleting the reference of the GALL-SLR Item IV.B4.RP-401 from Item 058a.

3-20 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Item 058b 058b to SRP-SLR Table 3.1-1, Item 058b remains as the AMR line item for B&W in SLR-ISG-2021-01-PWRVI: Expansion category components that are subject to non-cracking effect and

1) Added GALL-SLR Item mechanism combinations.

IV.B4.RP-245c as a new referenced GALL-SLR item for Item 058b. 1) The staff developed the new GALL-SLR Item IV.B4.RP-245c item in the ISG in order

2) Added GALL-SLR Item to be consistent with Item B2.1 in Table 4-4 IV.B4.RP-246e as a new referenced of the MRP-227, Rev. 1-A reportMRP-227, GALL-SLR item for Item 058b. Revision 1-A Report for B&W Expansion

3) Added GALL-SLR Item category surveillance specimen holder tube IV.B4.RP-252a as a new referenced bolts (surveillance specimen holder tube GALL-SLR item for Item 058b. [SSHT] bolts, Davis- Besse only). In Item B2.1, EPRI screened the SSHT bolts in for

4) Added GALL-SLR Item the aging mechanisms of wear and IV.B4.RP-386 as a new referenced irradiation-enhanced stress relaxation or GALL-SLR item for Item 058b. creep (ISR/IC). Thus, the new RP-245c item accounts for loss of material due to

5) Deleted GALL-SLR Item wear and loss of preload due to thermal or IV.B4.RP-254b as a referenced irradiation- enhanced stress relaxation or GALL-SLR item for Item 058b. creep in the SSHT bolts. The staff adjusted SRP-SLR Table 3.1-1, Item 058b accordingly to reference the GALL-SLR Item IV.B4.RP-245c item.

2) The prior GALL-SLR line item covering loss of material due to wear and changes in dimension in B&W-design Expansion category lower thermal shield (LTS) bolt locking devices and upper thermal shield (UTS) bolt locking devices was GALL-SLR Item IV.B4.RP-246b, which identified that the locking devices were located in the lower grid assembly. However, in Item B7.1 of Table 4-4 in MRP-227, Rev.Revision 1-A, the EPRI MRP identified that the UTS bolt locking devices are located in the core barrel assembly and screened them in for the mechanisms of wear and distortion.

Therefore, the staff needed to delete the UTS bolt locking devices from the RP-346b item and developed the new IV.B4.RP-246e item to address wear and distortion in the UTS bolt locking devices. The staff adjusted SRP-SLR Table 3.1-1, 058b item to reference GALL-SLR Item IV.B4.RP-246e.

3) The staff developed the new GALL-SLR Item IV.B4.RP-252a item in the ISG in order to be consistent with Item B2.1 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report for B&W Expansion category vent valve bodies, which screened 3-21 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change in for the aging mechanism of thermal embrittle (TE). Thus, the new RP-252a item accounts for loss of fracture toughness due to TE in the valve bodies. The staff adjusted SRP-SLR Table 3.1-1, Item 058b accordingly to reference GALL-SLR Item IV.B4.RP-252a.

4) The staff developed the new GALL-SLR Item IV.B4.RP-386 item in the ISG in order to be consistent with Item B2.1 in Table 4-4 of the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report for B&W Expansion category lower grid rib sections, which screened in for the aging mechanism of irradiation embrittlement (IE). Thus, the new RP-386 item accounts for loss fracture toughness due to IE in the rib sections. The staff adjusted the SRP-SLR Table 3.1-1, Item 058b accordingly to reference the GALL-SLR Item IV.B4.RP-386 item.

5) The previous inclusion of GALL-SLR Item IV.B4.RP-254b in the GALL-SLR report Report only applied to cracking of specific lower grid assembly bolt locking devices included in the design of the Three Mile Island Unit 1 (TMI-1) reactor. The owner of TMI-1 has made the decision to decommission the reactor. Therefore, GALL-SLR Item IV.B4.RP-254b is no longer needed for the purposes of the updates in the ISG and has been deleted. The staff adjusted Item 058b accordingly to delete Item IV.B4.RP-254b as a GALL-SLR item reference in the line item.

Table 3.1-1, Item The staff made the following changes The SRP-SLR Table 3.1-1, Item 059b 059b to SRP-SLR Table 3.1-1, Item 059b remains as the AMR line item for in SLR-ISG-2021-01-PWRVI: Westinghouse Expansion category RVI components that are subject to non-cracking

1) Added GALL-SLR Item aging effect and mechanism combinations.

IV.B2.RP-280a as a new referenced GALL-SLR item for Item 059b. 1) The staff developed Item IV.B2.RP-280a as a new GALL-SLR item reference for in

2) Added GALL-SLR Item the ISG in order to be consistent with Item IV.B2.RP-297a as a new referenced W3.3 in Table 4-6 of the MRP-227, Rev. 1-A GALL-SLR item for Item 059b. reportMRP-227, Revision 1-A Report, as modified by the additional lessons learned

3) Deleted GALL-SLR Item bases for Westinghouse-design core barrel IV.B2.RP-278a as a referenced assembly lower flange welds (LFWs) from GALL-SLR item in Item 059b. the staffs past processing of the Surry SLRA. In Item W3.3, the EPRI MRP designated that these LFWs are Expansion 3-22 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change category components for Westinghouse RVI management programs. In EPRI MRP Comment #12 on the ISG, EPRI MRP commented that GALL-SLR Item IV.B2.RP.280a should be omitted from the scope of the ISG due to the fact the core barrel LFW is not subject to irradiation embrittlement (IE) and is located far from the reactor core region. However, in the gap analysis of the staff-approved Surry SLRA, the past applicant cited IE and void swelling (VS) as applicable aging mechanisms for the Surry LFWs based on the information in contained in MRP-2018-022. Thus, the docketed SLRA information formed a sufficient basis for the establishment of the new GALL-SLR RP-280a item based on the reporting of IE and VS as being applicable to the Surry core barrel LFWs.

The staff also adjusted SRP-SLR Table 3.1-1, Item 059b to include reference of the new GALL-SLR Item IV.B2.RP-280a, which remains valid for the objective of the ISG.

2) In Item W2 of Table 4-3 in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report, the EPRI MRP assigned only the LFWs in peripheral (outer) CRGT assemblies as Primary category components for Westinghouse-design RVI programs, with inspections expanding to the LFWs in the non-peripheral (remaining)

CRGT assemblies (per Item W2.1 in Table 4-6 of the report) if an unacceptable level of degradation was detected in the peripheral CRGT LFWs. The EPRI MRP designated that these LFWs are susceptible to cracking (SCC, fatigue) and irradiation embrittlementIE and TEthermal embrittlement aging (IE and TE) mechanisms. The GALL-SLR Rreport did not include an AMR line item to address loss of fracture toughness due to IE or TE in the non-peripheral CRGT LFWs, which required the staffs development of the new GALL-SLR Item IV.B2.RP-297a to address this gap. The staff adjusted SRP-SLR Table 3.1-1, Item 059b accordingly to reference Item IV.B4.RP-297a as a new GALL-SLR item reference in the line item.

3) In MRP-227-A, the core barrel outlet nozzle welds (ONWs) were designated as 3-23 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change the Expansion components for Primary inspections performed on the core barrel upper flange weld (UFW). However, in MRP-227, Rev.Revision 1-A, the EPRI MRP deleted the ONWs as Expansion components and replaced them with the core barrel assembly upper girth weld (UGW), lower flange weld (LFW), upper axial welds (UAWs), and lower support forging or casting as the new applicable Expansion category components for the Primary category core barrel UFW inspections.

Therefore, the staff deleted GALL-SLR Item IV.B2.RP-278a as part of the ISG update efforts and adjusted SRP-SLR Table 3.1-1, Item 059b accordingly to delete reference of GALL-SLR Item IV.B2.RP-278a. The core barrel ONWs are now covered by the staffs Items for B&W-design No Additional Measures components (e.g., SRP-SLR Table 3.1-1, Item 055a and GALL-SLR AMR Item IV.B4.RP-236).

Table 3.1-1, Item The staff made the following changes The inclusion of stellite in Item 059c, and in 059c to SRP-SLR Table 3.1-1, Item 059c SRP-SLR Table 3.1-1, Item 119 as well, was in SLR-ISG-2021-01-PWRVI: requested by members of the U.S. nuclear power as part of their initial set of

1) Modified the Component recommendations for development of the column entry of the item to include ISG. Based on lessons learned from the RVI stellite as an additional (penitential) gap analysis previously provided in the Surry material of fabrication. SLRA, the staff confirmed that the previous applicant for the Surry SLRA identified that some of the RVI components (e.g., clevis inserts or fuel alignment pins) were designed with an outer, wear resistant stellite surface layer. Thus, the staff also found it would be appropriate to add stellite as a potentially applicable component material in the SRP-SLR Table 3.1-1 059c line item, as updated in Appendix A of the ISG.

Table 3.1-1, Item The staff made the following changes The staff modified SRP-SLR Table 3.1-1, 114 to SRP-SLR Table 3.1-1, Item 114 in Item 114 to generically cover any reactor SLR-ISG-2021-01-PWRVI: coolant system components that may be defined in the current licensing basis as

1) Added ASME Code Class 1 ASME Code Class components and the reactor interior attachments to the list aging effects and mechanisms that may of components in the Component apply to these types of components. This column entry for the line item. includes administrative adjustments of Item 114 to include ASME Code Class 1 reactor

2) Added primary water stress interior attachments and to incorporate the corrosion cracking (PWSCC), previous criteria for ASME Code Class-irradiation-assisted stress corrosion defined PWR reactor internals that were cracking (IASCC), and fatigue as previously within the scope of SRP-SLR 3-24 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change additional listed mechanisms for all Table 3.1-1 Item 032, which is being deleted or some of the components listed in in SLR-ISG-2021-01-PWRVI.

the line item.

Upon the staffs review of Items 032 and 114 in SRP-SLR Table 3.1-1, the staff found that the scope of components in Item 114 would bound those components that were previously include in Item 032. Thus, the AMR objective in SRP-SLR Table 3.1-1, Item 032 was determined to be redundant with that in SRP-SLR Table 3.1-1, Item 114, and therefore SRP-SLR Table 3.1-1, Item 032 was deleted from the scope of SRP-SLR Table 3.1-1, as updated in ISG Appendix A.

The aging mechanism of IASCC was added to Item 114 in order to account for the possibility that a specified PWR RVI component may be a nickel alloy or stainless steel SS reactor internal core support structure component located in the vicinity of the reactor core. For completeness, the mechanism of primary water stress corrosion cracking (PWSCC) was added to and included in Item 114 to account for cases where the component is defined as an ASME Code Class 1 steam generator component and PWSCC is a plausible cracking mechanism for the component. The mechanism of fatigue was added administratively to account for the possibility that a Code Class component may be susceptible to fatigue or cyclic loading mechanisms.

Table 3.1-1, Item The staff made the following changes The staff modified SRP-SLR Table 3.1-1, 118 to SRP-SLR Table 3.1-1, Items 118 Items 118 and 119 to be consistent with Table 3.1-1, Item and 119 in SLR-ISG-2021 changes to the referenced GALL-SLR items 119 PWRVI: linked to the SRP-SLR items (i.e., GALL-SLR Items IV.B2.R-423, IV.B3.R-423 and

1) Based on receipt and IV.B4.R-423, as linked to SRP-SLR Table acceptance of EPRI MRP Comments 3.1-1, Item 118 for cracking effect and

1. 15 and #17, modified the Structure mechanism combinations, and GALL-SLR and/or Component column entries of Items IV.B2.R-424, IV.B3.R-424 and the SRP-SLR Table 3.1-1 #118 and IV.B4.R-424, as linked to SRP-SLR Table

1. 119 items to include the words or 3.1-1, Item 119 for non-cracking effect and LRA/SLRA-specified reactor vessel mechanism combinations).

internal component.

The staff changes to the SRP-SLR-Table

2) Administratively edited Item 118 3.1-1 AMR Items 118 and 119 allow more to cite irradiation-assisted stress flexibility on when the #118 and #119 items, corrosion cracking as IASCC. and the associated GALL-SLR R-423 and R-424 type items, can be adopted and 3-25 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change

3) Based on receipt and used for development of AMR line items in acceptance of EPRI MRP Comments an incoming PWR SLRA. Use of SRP-SLR

1. 15, the staff modified the Aging Table 3.1-1, Items 118 and 119 (and the Management Program (AMP)/TLAA applicable referenced GALL-SLR items) may column entry of the SRP-SLR Table now be used for GALL-SLR XI.M16A-based 3.1-1 #118 item to include the words programs where the referenced MRP-227, or AMP XI.M16A, "PWR Vessel Rev.Revision 1-A protocols for a specified Internals," and AMP XI.M2, "Water PWR RVI component are adjusted based on Chemistry" (SCC and IASCC only), site-specific or component-specific with an adjusted site-specific or considerations. This will broaden the scope component-specific aging of SRP-SLR Table 3.1-1, Items 118 and 119 management basis for a specified (and the associated GALL-SLR R-423 reactor vessel internal component. items referenced by Item 118 and GALL-SLR R-424 items referenced by Item 119)

4) Based on receipt and so that they can be more readily applied and acceptance of EPRI MRP Comment used in applicable subsequent license

1. 17, the staff modified the Aging renewal applications.

Management Program (AMP)/TLAA column entry of the SRP-SLR Table Based on the staff acceptance of EPRI MRP 3.1-1 #119 item to include the words Comments #15 and #17 on draft SLR-ISG-or AMP XI.M16A, "PWR Vessel PWRVI-2020-XX, the staff implemented the Internals," with an adjusted site- changes recommended by EPRI for the specific or component-specific aging Structure and/or Component and Aging management basis for a specified Management Program (AMP)/TLAA column reactor vessel internal component. entries of Items #118 and #119, as specified in the previous column entry of this line item.

5) Added stellite as a new nickel-based alloy material for SRP-SLR In MRP 2018-022, the EPRI MRP added Table 3.1-1, AMR Item 119. stellite as a type of wear -resistant nickel-based alloy for specified stainless steel RVI components that were designed with an outer stellite surface layer for wear-resistance objective. The staff amended AMR Item 119 consistent with this change in MRP-2018-022.

Table 3.1-1, Item The staff administratively edited the The SRP-SLR Table 3.1-1 029, 041, and 029 SRP-SLR Table 3.1-1 029, 041, and 103 items involve specified types of BWR Table 3.1-1, Item 103 to cite the mechanism of RVI components. Although the line items do 041 irradiation-assisted stress corrosion not involve PWR RVI components, the Table 3.1-1, Item cracking (or irradiation-assisted changes involve administrative edits in order 103 SCC) as IASCC. to make the referencing of IASCC in the 029, 041, and 103 items consistent with the manner it is referenced in the corresponding SRP-SLR line items on cracking that were updated in Appendix A of SLR-ISG-2021 PWRVI and apply to PWR RVI components in SRP-SLR Table 3.1-1.

Section 3.1.2.2.6 Item 1 iswas deleted. The In the current GALL-SLR and SRP-SLR, the Section 3.1.3.2.6 recommendation iswas to remove item was edited or modified from the Table 3.1-1, Item this item from the FE guidance. previous documents. Previously in the 019 Removal of this item from FE would September 2005 GALL report, it was listed with the reactor vessel closure head flange 3-26 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change allow for more efficient and leak detection line, and it was recommended consistent review of an SLRA. to have a plant- specific program. When split up and transferred to the latest versions the nickel Ni-alloy version and SS version were separated for bottom-mounted instrument guide tubes. The recommendation for a plant-specific aging management program migrated as well.

However, this line item coveresis just the SS portion of the bottom mounted instrument guide tubes external to the bottom head. For other SS materials in the primary circuit with the concern for cracking due to primary water stress corrosion cracking (SCC), the application of AMP XI.M1 ASME Section XI Inservice Inspection, Subsections IWB, IWC and IWD, along with XI.M2 Water Chemistry has been shown to be adequate to address this aging mechanism, primary water SCC, for this material, SS, as noted in Item 033 (shown below.)

Section 3.1.2.2.6 Item 2 is deleted. The The staff re-evaluated the guidance provided Section 3.1.3.2.6 recommendation is to remove this in Section 3.1.2.2.6, Item 2 which states Table 3.1-1, Item item from the FE guidance. Removal Further evaluation is recommended of a 020 of this item from FE would allow for plant-specific program for these components more efficient and consistent review to ensure that this aging effect is adequately of an SLRA. managed and Section 3.1.3.2.6, Item 2 which states that A plant-specific AMP should be evaluated to manage cracking due to SCC in cast austenitic stainless steel (CASS) PWR Class 1 reactor coolant system piping and piping components exposed to reactor coolant that do not meet the carbon and ferrite content guidelines of NUREG-0313. The guidance in NUREG-0313, Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary, Revision 2, was published on January 1988. As the title suggests it was intended to provide guidance concerning intergranular stress corrosion cracking susceptibility of BWR piping and included guidelines on CASS components. Specifically, it highlighted the potential of SCC for certain CASS components if they did not meet the recommended ferrite and carbon content.

While the recommendations in NUREG-0313 are still very relevant to BWRs, current operating experience of CASS components in PWRs does not merit to elevate this AMR 3-27 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change item to a Further Evaluation. There is no current operating experienceOE that indicates that this is a problem for CASS components in PWRs that requires further evaluationFE.

Section 3.1.2.2.9 The staff updated the AMR further 1) The updated acceptance criteria are acceptance criteria to base them on based on staffs assumption that, if the the updated EPRI MRP EPRI-defined living AMP is based on I&Einspection and evaluation GALL-SLR AMP XI.M16A, the program guidelines in EPRI Technical Report being applied to the subsequent period of 3002017168 (MRP-227, extended operation will be based on the Rev.Revision 1-A): updated I&E guidelines in MRP-227, Rev.Revision 1-A. This is based on the

1) Clarified if GALL AMP XI.M16A EPRI-defined Needed Requirement in and EPRI MRP-based program is Section 7.3 of the MRP-227, Rev. 1-A used for aging management, the reportMRP-227, Revision 1-A Report, that program for the period of extended establishes the program will need to covert operation will be based on MRP-227, over and implement the updated guidelines Rev. 1-A reportMRP-227, Revision 1- in the MRP-227, Rev.Revision 1-A by A Report. January 1, 2022. Thus, for applicants that decide to submit SLRAs for their PWRs, the

2) Clarified that, if MRP-227, PWR vessel internals programs will have Rev.Revision 1-A is used for the converted over to the MRP-227, program, the assessments of the RVI Rev.Revision 1-A guidelines by the time the components will still need to be licensees will have entered into the subject to a gap analysis. subsequent period of extended operation for their PWR units.

3) Eliminated the discussion related to SLRA responses to the 2) If the AMP is based on MRP-227, applicant/licensee action items Rev.Revision 1-A as a starting point, a gap (A/LAIs) on the previous analysis will still be necessary, as the methodology in MRP-227-A. Rev.Revision 1-A version of the report is still based on an assessment of aging in the RVI

4) Clarified that a SLR applicant of components over a 60-year service life. The a PWR may address the A/LAI on the staffs previous criteria in SPR-SLR Section MRP-227, Rev. 1-A reportMRP-227, 3.1.2.2.9 and in GALL-SLR AMP XI.M16A, Revision 1-A Report as part of the PWR Vessel Internals, for requesting Ooperating Eexperience program performance of a gap analysis still remain element of its PWR vessel internals valid even if the program is updated to be program and in the technical basis based on the updated I&E guidelines in the document for the AMP. MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report.

3) Responses to the A/LAIs on MRP-227-A are no longer necessary PWR SLRAs because they were adequately resolved by the EPRI MRP in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report and closed out by the staff in the April 25, 2019 safety evaluation for the report.

4) The sole A/LAI on the guidelines in the MRP-227, Rev. 1-A reportMRP-227, 3-28 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Revision 1-A Report was issued by the staff in relation to resolving operating experience associated with the cracking of baffle-to-former bolts or core shroud bolts. Thus, an applicant of a PWR unit may address the A/LAI on MRP-227, Rev.Revision 1-A as part of its Ooperating Eexperience program element discussion provided in its AMP or in the technical basis document for the AMP. A separate response to the A/LAI does not need to be included in Appendix C of the SLRA.

Section 3.1.2.2.12 Section 3.1.2.2.12 was deleted. Section 3.1.2.2.12 instructed subsequent Table 3.1-1, Item license renewal applicantsSLRAs to perform 029 further evaluationFE on aging management Table 3.1-1, Item programsAMPs for IASCCirradiation 041 assisted stress corrosion cracking. However, Table 3.1-1, Item since publication of the SRP-SLR, the 103 Electric Power Research InstituteEPRI submitted the BWRVIP-315 topical report for NRC review. This topical report provided the industrys evaluation of aging management programsAMPs for irradiation assisted stress corrosion crackingIASCC for operations beyond 60 years. At the time of this revision, the BWRVIP-315 topical report was in an advanced stage of NRC review. The NRC determined that further evaluationFE for irradiation assisted stress corrosion crackingIASCC was no longer necessary, given the BWRVIP-315 topical report.

Section 3.1.2.2.13 Section 3.1.2.2.13 was deleted. Section 3.1.2.2.13 instructed subsequent Table 3.1-1, Item license renewal applicantsSLRAs to perform 099 further evaluationFE on aging management programsAMPs for loss of fracture toughness. However, since publication of the SRP-SLR, the Electric Power Research InstituteEPRI submitted the BWRVIP-315 topical report for NRC review. This topical report provided the industrys evaluation of aging management programsAMPs for loss of fracture toughness for operations beyond 60 years. At the time of this revision, the BWRVIP-315 topical report was in an advanced stage of NRC review. The NRC determined that further evaluationFE for loss of fracture toughness was no longer necessary, given the BWRVIP-315 topical report.

Section 3.1.3.2.6 Item 2 is deleted. The The staff re-evaluated the guidance provided recommendation is to remove this in Section 3.1.2.2.6, Item 2 which states 3-29 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change item from the FE guidance. Removal Further evaluation is recommended of a of this item from FE would allow for plant-specific program for these components more efficient and consistent review to ensure that this aging effect is adequately of an SLRA. managed and Section 3.1.3.2.6, Item 2 which states that A plant-specific AMP should be evaluated to manage cracking due to SCC in CASS PWR Class 1 reactor coolant system piping and piping components exposed to reactor coolant that do not meet the carbon and ferrite content guidelines of NUREG-0313. The guidance in NUREG-0313, Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary, Revision 2, was published on January 1988. As the title suggests it was intended to provide guidance concerning intergranular stress corrosion cracking susceptibility of BWR piping and included guidelines on CASS components. Specifically, it highlighted the potential of SCC for certain CASS components if they did not meet the recommended ferrite and carbon content. While the recommendations in NUREG-0313 are still very relevant to BWRs, current operating experience of CASS components in PWRs does not merit to elevate this AMR item to a Further Evaluation. There is no current operating experience that indicates that this is a problem for CASS components in PWRs that requires further evaluation. Section 3.1.3.2.6, Item 2 is deleted as referenced NUREG-0313 I applicable to BWRs.

Section 3.1.3.2.9 Changes are analogous to those The technical bases for changes being made to AMR acceptance criteria in made to AMR acceptance criteria in SRP-SRP-SLR Section 3.1.2.2.9. SLR Section 3.1.2.2.9 also apply to the changes being made to the AMR review Additionally, the staff clarified the procedures of SRP-SLR Section 3.1.3.2.9.

AMR items in GALL-SLR Items IV.B2.R-423, IV.B3.R-423, or Since a gap analysis will be needed if MRP-IV.B4.R-423 for cracking effects or 227, Rev.Revision 1-A is used as the mechanisms, or GALL-SLR Items starting point AMP, the staff expanded the IV.B4.R-424, IV.B3.R-424, or scope of the GALL-SLR R-423 items and IV.B4.R-424 for non-cracking effects R-424 items to allow use of the generic or mechanisms may be used if the AMR line items, even if the program was MRP I&E protocols for a specified being based on the program defined in component in the MRP-227, GALL-SLR AMP XI.M16A, and the EPRI-Rev.Revision 1-A are being adjusted defined protocols for a specified component 3-30 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change on a site-specific or component- in MRP-227, Rev.Revision 1-A were being specific basis. adjusted on a site-specific or component-specific basis. The changes will provide a broader scope of the R-423 and R-424 items and will allow the amended GALL-SLR AMR items to be used even if the EPRI MRP I&E protocols for a given component in the MRP-227, Rev. 1-A reportMRP-227, Revision 1-A Report are being adjusted as a result of the gap analysis or operating experience considerations. The AMRs in the R-423 and R-424 type line items are no longer limited only to plant-specific RVI aging management programsAMPs.

Section The AMR further evaluationFE For the SRP-SLR Section 3.1.2.2.11, 3.1.2.2.11,, subsections were revised, to clarify Subsection Item 1 guidelines that apply to Subsections 1 and the plant-specific parameters to be PWR SG divider plates, the staff added 2 evaluated against industry analyses additional guidance on the plant-specific Table 3.1-1, Item to determine whetherif a given plant parameters that should be compared to 025 is bounded by industry analyses for industry analyses that show the analyses are steam generator (SG) divider plate cracking. Additionally, reference to a applicable and bounding for a given plant.

plant-specific AMP for plants that are Additionally, the reference to a plant-specific not bounded is replaced with a AMP was replaced with a reference to the reference to the One-Time Inspection One-Time Inspection AMP for applicants that AMP. would need to manage cracking due to PWSCC in their SG divider plates.

Section 3.1.2.2.16 Revised the last paragraph to state The statement in Section 3.1.2.2.16 that loss that the applicant may mitigate or of material does not require management prevent loss of material using a contradicts the statement in SLR-SRP barrier coating rather than saying the Section A.1.2.1, Item 5, which states that loss of material does not require even with a prevention or mitigation management if a barrier coating is program, including a coating, an aging effect used. Added a statement that the should be identified as applicable for SLR, applicant should identify loss of and the AMR should consider the adequacy material as applicable for SLR and of the AMP referencing the prevention or identify the AMP that will be used to mitigation program (e.g., coating). Section manage the integrity of the coating. A.1.2.1, Item 5, correctly states the need for an AMP to manage coating integrity.

Section 3.1.3.2.16 Revised the wording in the last The statement in Section 3.1.3.2.16 that paragraph to delete the statement barrier coatings make aging management that loss of material (LOM) does not unnecessary is inconsistent with Section require aging management if a A.1.2.1, Item 5, which correctly states that barrier coating is used. Deleted the an AMP is needed to manage the integrity of statement that the reviewer verifies the coating that prevents the aging effect.

the barrier coating is impermeable. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

3-31 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Section The SRP-SLR Sections 3.1.2.2.16a As addressed in NRC Bulletin 88-08, non-3.1.2.2.176a and 3.1.3.2.16a are added to isolable branch lines connected to the Section address a further evaluationFE for reactor coolant system may be subject to 3.1.3.2.176a aging management of thermal fatigue unacceptable thermal stress that can cause Table 3.1-1 Item in the reactor coolant system. In the thermal fatigue cracking and leakage failure.

140 further evaluationFE, the applicant The NRC Bulletin 88-08 states that, when evaluates the adequacy of a plant- such piping is identified, actions should be specific program for the aging taken to ensure that the piping will not be management (e.g., adequate subject to unacceptable thermal stress.

selection of susceptible locations for inspections, timely detection of Industry operating experienceOE and cracks and preventive action for evaluation indicate that, in some branch valve in-leakage as needed). lines, thermal stratification or mixing cycles can occur due to the interaction between the hot swirl penetration from the reactor coolant system and the cold water in-leakage from a leaking valve. In other branch lines, thermal stratification or mixing cycles can result from the interaction of the hot swirl penetration and the cold water in the normally cool, stagnant branch lines without a leaking valve. In addition, cold or hot fluid injections can cause thermal fatigue in the reactor coolant system as indicated in ASME Code Case N-716-1. Therefore, cracking due to thermal fatigue can occur due to cyclic stresses from the thermal stratification, mixing or injection cycles.

The industry guidance to manage the thermal fatigue in the PWR branch lines is described in EPRI MRP-146, Revision 2.

The guidance provides methods for screening and evaluating the susceptibility of non-isolable branch lines to thermal fatigue.

MRP-146, Revision 2 also provides general guidance for monitoring valve in-leakage and thermal stress as needed and performing volumetric examinations on the susceptible locations (e.g., examination areas, volumes and frequencies). These guidelines continue to be enhanced based on the lessons learned from relevant operating experience and research activities.

BWRVIP-155, Revision 1 also describes the evaluation of thermal fatigue susceptibility in the branch lines of BWR reactor coolant pressure boundary.

In comparison, the inservice inspection (ISI) requirements in Table IWB-2500-1 of ASME Code,Section XI do not include a specific 3-32 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change examination item for thermal fatigue cracking in ASME Code Class 1 components (reactor coolant pressure boundary). However, alternative risk-informed ISIsinservice inspections typically include an examination item for thermal fatigue cracking (e.g., as specified in ASME Code Case N-716-1 that has been approved in NRC Regulatory Guide [RG] 1.147, Revision 18). Therefore, the existing inservice inspectionISIs at plants may include the piping locations susceptible to thermal fatigue.

Currently, the SRP-SLR does not include a further evaluationFE section that addresses aging management for the piping locations susceptible to thermal fatigue. Therefore, new SRP-SLR Sections 3.1.2.2.16a and 3.1.3.2.16a are added to address the adequacy of a plant-specific aging management programAMP (e.g., adequate selection of susceptible locations for inspections, timely detection of cracks and preventive action for valve in-leakage).

Changes are also made to the SRP-SLR section for references (Section 3.1.6). In addition, relevant changes are made to the aging management review (AMR) tables in the SRP-SLR and GALL-SLR Report.

Section The SRP-SLR Sections 3.1.2.2.10.2 , The SRP-SLR Sections 3.1.2.2.10,. Item 2 3.1.2.2.10.2 Item .2 and 3.1.3.2.10, Item .2 are and 3.1.3.2.10., Item 2 describe a further Section changed to add additional examples evaluationFE to manage loss of material due 3.1.3.2.10.2 for the wear locations in control rod to wear in CRD thermal sleeves. As an drive (CRD) thermal sleeves. The example of the wear locations, the SRP-SLR added examples address the wear sections refer to the location where the degradation near the bottom of the thermal sleeve exists from the CRD head thermal sleeve and at the thermal penetration nozzle inside the reactor vessel sleeve upper flange location. (RV). The wear at this location results from the interactions between the thermal sleeve outer surface and the head penetration nozzle. This type of wear is called thermal sleeve outer-diameter (OD) wear.

In addition, industry operating experienceOE indicates that wear can occur in the following locations of the thermal sleeves: (1) near the bottom of the thermal sleeve (wear at this location is called thermal sleeve inner-diameter [(ID]) wear); and (2) thermal sleeve upper flange location (thermal sleeve flange 3-33 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change wear). The thermal sleeve ID wear is due to the interactions between the thermal sleeve inner surface and the drive rod passing through the thermal sleeve near the bottom of the thermal sleeve. The thermal sleeve flange wear is caused by the interactions between the bottom side of the flange and the CRD penetration housing near the top of the thermal sleeve.

Changes are made to the SRP-SLR sections to add ID and flange wear locations. Related references are also added in the reference section (SRP-SLR Section 3.1.6).

1 Table 3-4 Table 3-4 SRP-SLR, Revision 1 Chapter 3.2, Engineered Safety Features, 2 Differences from SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change Section 3.2.3.5 Add reference to the FSAR This information was inadvertently omitted Final Safety Supplement information contained in when the FSAR Supplement information Analysis Report GALL-SLR Table X-01 and Table XI- was relocated from the SRP-SLR tables (FSAR) 01. to the GALL-SLR tables.

Supplement Section 3.2.2.2.11 New further evaluation (FE) sections Recent industry operating experienceOE Section 3.2.3.2.11 in SRP-SLR Section 3.2 is needed indicates that significant wear can occur on Section 3.2.6 to address new industry operating the OD of piping due to system vibrations Table 3.2-1, Item experience related to loss of material and interactions with certain types of 135 due to wear on the outside diameter reflective metal insulation (RMI).

(OD) of American Society of Specifically, piping that uses RMI with an Mechanical Engineers Boiler and end cap of thin sheet metal has the Pressure Vessel Code (ASME potential for wear up to 360 degrees Code) Class 1 and Class 2 small- around its circumference. Repeated bore piping. Specifically, a reference movement of the RMI end cap in contact is made to the NRC Information with a pipe OD can cause loss of material Notice (IN) 2007-21, Pipe Wear on the OD of the subject pipe. Multiple Due to Interaction of Flow-Induced instances of such wear are known to have Vibrations and Reflective Metal occurred. An occurrence of this type of Insulation, Supplement 1, issued on material loss due to RMI end cap wear was December 11, 2020. Its purpose was reported in 2006, and was the subject of IN to alert licensees of nuclear power 2007-21, issued by the NRC on June 11, reactors on recent operating 2007.

experience (OE) related to wear of nuclear power plant piping caused During outage activities in fFall of 2006, the by flow-induced vibration and licensee at Catawba, Unit 1 removed RMI interaction of certain type of on small-bore ASME Code Class 2 piping insulation. for a planned valve replacement. The licensee identified multiple wear marks on the OD of stainless steel (SS) piping. It was determined that the wear marks were the 3-34 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change result of interactions between the SSstainless steel piping and the stainless steelSS RMI end caps, caused by vVibration. The licensee initially identified three locations with metal loss. During the extent of condition review, additional 81 discrete wear marks were identified over a 150-foot length of piping. All of the wear marks were located inside containment and at insulation end caps.

In December of 2020, IN 2007-21 was revised due to more recent occurrences of metal loss on the OD of ASME Code Class 1 small-bore piping at two other nuclear power plants. The most recent known occurrence is summarized below.

During an outage in the spring of 2020, workers at Arkansas Nuclear One, Unit 2 identified multiple wear marks on ASME Code Class 1 pressurizer spray piping. The wear marks were identified as a result of piping inspections in response to vibration related failures of snubber connections to the pressurizer spray piping. The wear marks ranged from surface scratches to deeper groves that were nearly 360 degrees around the OD circumference of the pipe. The licensee determined that the wear was caused by vibration-induced interactions of the RMI end caps and the OD surfaces of the subject piping.

In the overview of the original IN 2007-21 and its subsequent supplement of 2020, it is apparent that the observed OD pipe wear for both the ASME Code Class 1 and 2 small-bore piping was discovered as a result of unrelated inspections. There are currently no specific ASME Code requirements to remove insulation from piping and inspect the piping for degradation due to RMI wear.

This type of wear, if present and undetected, could have a significant impact on the integrity of ASME Code Class 1 and 2 small-bore piping because: (1) small-bore piping has wall thickness values that are significantly less than those for large bore piping, (2) small-bore piping systems are more susceptible to vibration, and (3) there 3-35 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change are no specific ASME Code requirements to inspect piping for RMI wear.

Based on the observed degradation, the licensees referenced in the updated IN 2007-21 have performed engineering evaluations as well as completed extent of conditions and installed a modified RMI to eliminate the pipe wear. The modified insulation has an end cap as a piece of flat sheet metal that looks like a cuff, band or strip. The modified end cap touches the pipe as a flat piece of metal parallel to the pipe, not as a sharp edge; thereby eliminating the potential for excessive wear. Temporary modifications have also included installation of cuffs on the OD of the pipe where the end caps are located.

More recently, an applicant in the process of applying for a subsequent license renewal (SLR) reviewed the operating experienceOE in the updated IN 2007-21, determined that the operating experienceOE could be applicable at its units, and updated its aging management program (AMP) to check for the RMI end cap wear on its ASME Code Class 1 small-bore piping (Ref. 40).

Relevant SRP-SLR sections are being updated to add a further evaluationFE to determine whetherif a plant seeking license renewal has evaluated the use of RMI in its integrated plant assessment and the potential of OD wear on its population of ASME Code Class 1 and 2 small-bore piping. Related references are also added in the reference section (SRP-SLR Section 3.2.6).

Section 3.2.2.2.2 Revised the last paragraph to state The statement in Section 3.2.2.2.2 that loss that the applicant may mitigate or of material does not require management prevent loss of material using a contradicts the statement in SLR-SRP barrier coating rather than saying the Section A.1.2.1, Item 5, which states that loss of material does not require even with a prevention or mitigation management if a barrier coating is program, including a coating, an aging used. Added a statement that the effect should be identified as applicable for applicant should identify loss of SLR, and the AMR should consider the material as applicable for SLR and adequacy of the AMP referencing the identify the AMP that will be used to prevention or mitigation program (e.g.,

manage the integrity of the coating. coating). Section A.1.2.1, Item 5, correctly 3-36 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change states the need for an AMP to manage coating integrity.

Section 3.2.3.2.2 Revised the wording in the last The statement in Section 3.2.3.2.2 that paragraph to delete the statement barrier coatings make aging management that LOM does not require aging unnecessary is inconsistent with Section management if a barrier coating is A.1.2.1, Item 5, which correctly states that used. Deleted the statement that the an AMP is needed to manage the integrity reviewer verifies the barrier coating of the coating that prevents the aging is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.2.2.2.4 Revised the last paragraph to state The statement in Section 3.2.2.2.4 that that the applicant may mitigate or cracking due to SCC does not require prevent cracking due to stress management contradicts the statement in corrosion cracking (SCC) using a SLR-SRP Section A.1.2.1, Item 5, which barrier coating rather than saying states that even with a prevention or cracking due to SCC does not mitigation program, including a coating, an require management if a barrier aging effect should be identified as coating is used. Added a statement applicable for SLR, and the AMR should that the applicant should identify consider the adequacy of the AMP cracking due to SCC as applicable referencing the prevention or mitigation for SLR and identify the AMP that program (e.g., coating). Section A.1.2.1, will be used to manage the integrity Item 5, correctly states the need for an of the coating. AMP to manage coating integrity.

Section 3.2.3.2.4 Revised the wording in the last The statement in Section 3.2.3.2.4 that paragraph to delete the statement barrier coatings make aging management that cracking due to SCC does not unnecessary is inconsistent with Section require aging management if a A.1.2.1, Item 5, which correctly states that barrier coating is used. Deleted the an AMP is needed to manage the integrity statement that the reviewer verifies of the coating that prevents the aging the barrier coating is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.2.2.2.8 Revised the wording in the last The purpose of this change is to make the paragraph to clarify that the wording about barrier coatings consistent applicant may mitigate or prevent with that of other FE sections that required cracking due to SCC using a barrier changes to the coatings discussion.

coating. Added a statement that the applicant should identify cracking due to SCC as applicable for SLR and identify the AMP that will be used to manage the integrity of the coating.

Section 3.2.3.2.8 Revised the wording in the fourth The statement in Section 3.2.3.2.8 that paragraph to delete the statement barrier coatings make aging management that cracking due to SCC does not unnecessary is inconsistent with Section require aging management if a A.1.2.1, Item 5, which correctly states that barrier coating is used. Deleted the an AMP is needed to manage the integrity 3-37 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change statement that the reviewer verifies of the coating that prevents the aging the barrier coating is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.2.2.2.10 Revised the wording in the last The purpose of this change is to make the paragraph to clarify that the wording about barrier coatings consistent applicant may mitigate or prevent with that of other FE sections that required LOM using a barrier coating. Added changes to the coatings discussion.

a statement that the applicant should identify LOM as applicable for SLR and identify the AMP that will be used to manage the integrity of the coating.

Section 3.2.2.2.10 Revised the wording to indicate that During the first revision or interim staff LOM due to pitting and crevice guidance /(ISG) to the SRP--SLR, the staff corrosion need not be managed if added a provision to the further the type of aluminum is not evaluationFE sections associated with loss susceptible to cracking and of material due to pitting and crevice plant--specific operating corrosion of aluminum components experienceOE does not reveal any exposed to air or condensation. This issues related to loss of material due change allowed an alternative to conducting to pitting or crevice corrosion. a Oone--Ttime Iinspection to detect loss of material. Loss of material need not be managed if: (a) the aluminum material is not susceptible to cracking; and (b) plant--specific operating experienceOE does not reveal any issues related to loss of material due to pitting or crevice corrosion.

The staff included this alternative because:

(a) it is unlikely that pitting or crevice corrosion in aluminum components would lead to a loss of intended function; and (b) if loss of material has not been identified as an issue after 40 years (the earliest point at which a subsequent license renewal application [SLRA] can be submitted, it is unlikely that it will lead to a loss of intended function during the subsequent period of extended operation. This alternative was not allowed for aluminum materials that are susceptible to stress corrosion crackingSCC because pitting or crevice corrosion might, but not necessarily, be a precursor to cracking.

Section 3.2.3.2.10 Revised the wording in the last The statement in Section 3.2.3.2.10 that paragraph to delete the statement barrier coatings make aging management that LOM does not require aging unnecessary is inconsistent with Section management if a barrier coating is A.1.2.1, Item 5, which correctly states that used. Deleted the statement that the an AMP is needed to manage the integrity of the coating that prevents the aging 3-38 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change reviewer verifies the barrier coating effect. Because an AMP will be used to is impermeable. manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.2.3.2.10 Revised the wording to indicate that During the first revision or /ISG to the LOM due to pitting and crevice SRP-SLR, the staff added a provision to the corrosion need not be managed if further evaluationFE sections associated the type of aluminum is not with loss of material due to pitting and susceptible to cracking and crevice corrosion of aluminum components plant--specific operating exposed to air or condensation. This experienceOE does not reveal any change allowed an alternative to conducting issues related to loss of material due a Oone--Ttime Iinspection to detect loss of to pitting or crevice corrosion. material. Loss of material need not be managed if: (a) the aluminum material is not susceptible to cracking; and (b) plant--specific operating experienceOE does not reveal any issues related to loss of material due to pitting or crevice corrosion.

The staff included this alternative because:

(a) it is unlikely that pitting or crevice corrosion in aluminum components would lead to a loss of intended function; and (b) if loss of material has not been identified as an issue after 40 years (the earliest point at which a SLRA can be submitted, it is unlikely that it will lead to a loss of intended function during the subsequent period of extended operation. This alternative was not allowed for aluminum materials that are susceptible to stress corrosion crackingSCC because pitting or crevice corrosion might, but not necessarily, be a precursor to cracking.

Section 3.4.2.2.9 Revised the wording to indicate that During the first revision or /ISG to the LOM due to pitting and crevice SRP--SLR, the staff added a provision to corrosion need not be managed if the further evaluationFE sections the type of aluminum is not associated with loss of material due to susceptible to cracking and pitting and crevice corrosion of aluminum plant-specific operating components exposed to air or experienceOE does not reveal any condensation. This change allowed an issues related to loss of material due alternative to conducting a Oone--Ttime to pitting or crevice corrosion. Iinspection to detect loss of material. Loss of material need not be managed if: (a) the aluminum material is not susceptible to cracking; and (b) plant--specific operating experienceOE does not reveal any issues related to loss of material due to pitting or crevice corrosion. The staff included this alternative because: (a) it is unlikely that pitting or crevice corrosion in aluminum components would lead to a loss of intended function; and (b) if loss of material 3-39 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change has not been identified as an issue after 40 years (the earliest point at which a SLRA can be submitted, it is unlikely that it will lead to a loss of intended function during the subsequent period of extended operation. This alternative was not allowed for aluminum materials that are susceptible to stress corrosion crackingSCC because pitting or crevice corrosion might, but not necessarily, be a precursor to cracking.

Table 3.2-1 Item Included reduction of heat transfer The staff noted that for other material and 19 for nickel alloy internally exposed to environment combinations in the GALL-treated borated water/ SLR Report, reduction of heat transfer due to fouling is the only aging effect associated with an intended function of heat transfer.

The Water Chemistry AMP can be used to minimize the potential for deposits that can lead to fouling through the control of primary side water chemistry. Additionally, the One- Time Inspection AMP will help to verify the effectiveness of the Water Chemistry AMP. The GALL- SLR recommends the use of the Water Chemistry and Steam Generator AMPs (AMR Table 1 Iitem 3.1-1, 111) to manage the reduction of heat transfer due to fouling in nickel alloy tubes. The use of the Water Chemistry and One- Time Inspection AMPs provide an analogous approach (i.e.,

Wwater Cchemistry control and an inspection to verify effectiveness) to managing the reduction of heat transfer on primary side nickel alloy heat exchanger tubes.

The staffs review of the Turkey Point subsequent license renewal applicationSLRA demonstrates that stainless steelSS and nickel alloy have similar aging effects when exposed to treated borated water. The GALL- SLR recommends the use of the Water Chemistry and One- Time Inspection AMPs to manage the reduction of heat transfer in stainless steelSS heat exchanger tubes.

Because stainless steelSS and nickel alloy experience similar aging effects it is reasonable to use the same AMPs to manage the aging effects in nickel alloy materials.

3-40 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Table 3.2-1, Item Added malleable iron as an During its review of recent SLRA plant-036 applicable material. specific operating experienceOE, in Table 3.2-1, Item response to the staffs observation 037 regarding dark corrosion product layers Table 3.2-1, Item indicative of graphitic corrosion on the internal surfaces of malleable iron fittings 074 exposed to a closed-cycle cooling water environment (ADAMS Accession No. ML22010A129), the staff has revised guidance documents (i.e., GALL-SLR Report and SRP-SLR) to include malleable iron as a material susceptible to selective leaching.

Table 3.2-1, Item Added lubricating oil and The staff has accepted opportunistic 072 condensation as applicable inspections, in lieu of periodic inspections, Table 3.2-1, Item environments. as an acceptable alternative for buried 073 internally coated or/ lined fire water system piping provided: (a) flow tests and internal piping inspections will occur at intervals specified in NFPA 25, or as modified by AMP XI.M27, Table XI.M27-1; and (b) through-wall flaws in the piping can be detected through continuous system pressure monitoring. Examples of the staffs acceptance of this alternative approach are documented in the Safety Evaluation Report Related to the License Renewal of Fermi 2 Nuclear Power Plant (ADAMS Accession No. ML16190A241) and the Safety Evaluation Report Related to the Subsequent License Renewal of Peach Bottom Atomic Power Station, Units 2 and 3 (ADAMS Accession No. ML20044D902). Based on recent OE involving ruptures of buried fire water system piping due to age-related degradation (ADAMS Accession No. ML19294A044), the staff added a third condition for using this alternative approach related to plant- specific OEoperating experience. The staff notes that the subject OE involved degradation of the external surfaces of the piping; however, degradation of internal coatings or /linings could also result in significant degradation of buried fire water system piping.

The GALL-SLR Report discusses the reason for citing specific AMPs to manage recurring internal corrosion rather than a plant-specific AMP in the section titled 3-41 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Explanation of the Use of Multiple Aging Management Programs in Aging Management Review Items. For the associated AMR item in the SRP-SLR (item 3.3-1-127), the listed environments still include closed-cycle cooling water even though NUREG-2221, Technical Bases for Changes in the Subsequent License Renewal Guidance Documents NUREG-2191 and NUREG-2192, Table 2-13, notes that the associated item in Table C2, Closed-Cycle Cooling Water System, was deleted because recurring internal corrosion is not anticipated in this system. These changes corrects this error in conjunction with the adjustments above for the use of multiple AMPs.

Table 3.2-1, Item Added Fflow blockage due to Subsequent to issuance of the GALL-SLR 132 fouling as an applicable aging effect Report, the staff recognized that to be or /mechanism. consistent with other GALL-SLR Report items associated with heat exchanger tubes, E-475 should have also cited reduction of heat transfer due to fouling.

This is consistent with GALL Report Revision 2 Iitem SP-41 where a material (i.e., stainless steelSS) that is not susceptible to loss of material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

Titanium components are subject to flow blockage due to fouling due to potential debris in the raw water environment.

Table 3.2-1, Item Added (Inspection of Internal Flow blockage due to fouling is not an 134 Surfaces only) to flow blockage due applicable aging effect requiring to fouling. management for the external environment of polymeric components.

1 Table 3-5 SRP-SLR, Revision 1, Chapter 3.3, Auxiliary Systems, Differences from 2 SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change Section 3.3.3.5 Add reference to the FSAR This information was inadvertently omitted Final Safety Supplement information contained in when the FSAR Supplement information Analysis Report GALL-SLR Table X-01 and Table was relocated from the SRP-SLR tables to (FSAR) XI-01. the GALL-SLR tables.

Supplement Section 3.3.2.2.3 Revised the last paragraph to state The statement in Section 3.3.2.2.3 that that the applicant may mitigate or cracking due to SCC does not require 3-42 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change prevent cracking due to stress management contradicts the statement in corrosion cracking (SCC) using a SLR-SRP Section A.1.2.1, Item 5, which barrier coating rather than saying states that even with a prevention or cracking due to SCC does not mitigation program, including a coating, an require management if a barrier aging effect should be identified as coating is used. Added a statement applicable for SLR, and the AMR should that the applicant should identify consider the adequacy of the AMP cracking due to SCC as applicable referencing the prevention or mitigation for subsequent license renewal program (e.g., coating). Section A.1.2.1, (SLR) and identify the aging Item 5, correctly states the need for an management program (AMP) that AMP to manage coating integrity.

will be used to manage the integrity of the coating.

Section 3.3.3.2.3 Revised the wording in the last The statement in Section 3.3.3.2.3 that paragraph to delete the statement barrier coatings make aging management that cracking due to SCC does not unnecessary is inconsistent with Section require aging management if a A.1.2.1, Item 5, which correctly states that barrier coating is used. Deleted the an AMP is needed to manage the integrity statement that the reviewer verifies of the coating that prevents the aging the barrier coating is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluation (FE) review.

Section 3.3.2.2.4 Revised the last paragraph to state The statement in Section 3.3.2.2.4 that that the applicant may mitigate or loss of material does not require prevent loss of material using a management contradicts the statement in barrier coating rather than saying SLR-SRP Section A.1.2.1, Item 5, which the loss of material does not require states that even with a prevention or management if a barrier coating is mitigation program, including a coating, an used. Added a statement that the aging effect should be identified as applicant should identify loss of applicable for SLR, and the AMR should material as applicable for SLR and consider the adequacy of the AMP identify the AMP that will be used to referencing the prevention or mitigation manage the integrity of the coating. program (e.g., coating). Section A.1.2.1, Item 5, correctly states the need for an AMP to manage coating integrity.

Section 3.3.3.2.4 Revised the wording in the last The statement in Section 3.3.3.2.4 that paragraph to delete the statement barrier coatings make aging management that LOM does not require aging unnecessary is inconsistent with Section management if a barrier coating is A.1.2.1, Item 5, which correctly states that used. Deleted the statement that the an AMP is needed to manage the integrity reviewer verifies the barrier coating of the coating that prevents the aging is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.3.2.2.8 Revised the wording in the last The purpose of this change is to make the paragraph to clarify that the wording about barrier coatings consistent applicant may mitigate or prevent with that of other FE sections that required cracking due to SCC using a barrier changes to the coatings discussion.

coating. Added a statement that the 3-43 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change applicant should identify cracking due to SCC as applicable for SLR and identify the AMP that will be used to manage the integrity of the coating.

Section 3.3.3.2.8 Revised the wording in the fourth The statement in Section 3.3.3.2.8 that paragraph to delete the statement barrier coatings make aging management that cracking due to SCC does not unnecessary is inconsistent with Section require aging management if a A.1.2.1, Item 5, which correctly states that barrier coating is used. Deleted the an AMP is needed to manage the integrity statement that the reviewer verifies of the coating that prevents the aging the barrier coating is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.3.2.2.10 Revised the wording to indicate that During the first revision/ISG to the SRP-SLR, loss of material due to pitting and the staff added a provision to the further crevice corrosion need not be evaluation sections associated with loss of managed if the type of aluminum material due to pitting and crevice corrosion used is not susceptible to cracking of aluminum components exposed to air or and plant-specific operating condensation. This change allowed an experience does not reveal any alternative to conducting a one-time issues related to loss of material due inspection to detect loss of material. Loss of to pitting or crevice corrosion. material need not be managed if: (a) the aluminum material used is not susceptible to cracking; and (b) plant-specific operating experience does not reveal any issues related to loss of material due to pitting or crevice corrosion. The staff included this alternative because: (a) it is unlikely that pitting or crevice corrosion in aluminum components would lead to a loss of intended function; and (b) if loss of material has not been identified as an issue after 40 years (the earliest point at which a SLRA can be submitted), it is unlikely that it will lead to a loss of intended function during the subsequent period of extended operation.

This alternative was not allowed for aluminum materials that are susceptible to stress corrosion cracking because pitting or crevice corrosion might, but not necessarily, be a precursor to cracking.

Section 3.3.2.2.10 Revised the wording in the last The purpose of this change is to make the paragraph to clarify that the wording about barrier coatings consistent applicant may mitigate or prevent with that of other FE sections that required LOM using a barrier coating. Added changes to the coatings discussion.

a statement that the applicant should identify LOM as applicable for SLR and identify the AMP that 3-44 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change will be used to manage the integrity of the coating.

Section 3.3.3.2.10 Revised the wording in the last The statement in Section 3.3.3.2.10 that paragraph to delete the statement barrier coatings make aging management that LOM does not require aging unnecessary is inconsistent with Section management if a barrier coating is A.1.2.1, Item 5, which correctly states that used. Deleted the statement that the an AMP is needed to manage the integrity reviewer verifies the barrier coating of the coating that prevents the aging is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Table 3.3-1, Item Revised the items to include High- These items are revised to include 30a density polyethylene (HDPE) and/or components, aging effects/mechanism, Table 3.3-1, Item carbon fiber reinforced polymer and/or GALL-SLR item(s) for crediting new 104 (CFRP) repaired piping as components, as applicable, clarify AMP XI.M43, High Density Polyethylene Table 3.3-1, Item the associated aging mechanisms (HDPE) Piping and Carbon Fiber Reinforced 133 and effects, and refer to the new Polymer (CFRP) Repaired Piping, to Table 3.3-1, Item AMP XI.M43, High Density manage the effects of age-related 194 Polyethylene (HDPE) Piping and Table 3.3-1, Item degradation mechanisms that are applicable Carbon Fiber Reinforced Polymer 196 (CFRP) Repaired Piping.. to HDPE piping and CFRP repaired piping.

Table 3.3-1, Item This new AMP reflects the recent 210 Added GALL-SLR Item VII.I.AP-182 introduction and increasing use of CFRP Table 3.3-1, Item to item 104. repaired piping at reactor facilities. The 253 Added GALL-SLR Item VII.I.A-420 unique aging issues and aging management to item 133.

approaches for CFRP repaired piping and Added GALL-SLR Item VII.I.A-538 to item 194. HDPE piping (previously managed by AMP Added GALL-SLR Item VII.C1.A-792 XI.M41) were considered to be most to item 253. effectively addressed with a dedicated AMP.

Table 3.3-1, Item Added nickel -alloy as an applicable The staff noted that the GALL-SLR Report 071 material and updated applicable recommends the use of the Fuel Oil GALL-SLR Report items. Chemistry and One- Time Inspection AMPs to manage loss of material of several different materials that are exposed to a fuel oil environment. These new AMR items credit the Fuel Oil Chemistry program to minimize contaminants which could lead to loss of material, and the One- Time Inspection program to verify the effectiveness of the Fuel Oil Chemistry program. The use of the Fuel Oil Chemistry program can minimize contaminants regardless of the material of the affected component. Therefore, the staff has reasonable assurance that it will be effective in managing loss of material for nickel alloy strainer elements exposed to fuel oil.

3-45 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Table 3.3-1, Item Added malleable iron as an During its review of recent SLRA plant-072 applicable material. specific operating experience (OE), in Table 3.3-1, Item response to the staffs observation regarding 140 dark corrosion product layers indicative of graphitic corrosion on the internal surfaces of malleable iron fittings exposed to a closed-cycle cooling water environment (ADAMS Accession No. ML22010A129), the staff has revised guidance documents (i.e., GALL-SLR Report and SRP-SLR) to include malleable iron as a material susceptible to selective leaching.

Table 3.3-1, Item Added heat exchanger components GALL-SLR Table VII.J only addresses 114 and tanks as applicable components. components with material/environment combinations that do not have aging effects expected to degrade their intended function.

Because copper alloy heat exchanger tubes have aging effects requiring management, the component description for this item needs to include other than tubes, after heat exchanger components. The historical response to public comment 045-062 (ML17362A143), which stated that a change to this item will be made, never occurred.

The comment response had stated that the basis for AP-144 was equally applicable to tanks and heat exchangers components in addition to piping and piping components.

However, the response also noted that reduction of heat transfer due to fouling of copper alloy heat exchanger tubes is addressed separately in item A-565.

Because copper alloy heat exchanger tubes exposed to air and condensation have an aging effect that could degrade their ability to perform their intended function, they are excluded from GALL-SLR Report Table VII.J for components with no aging effects requiring management. GALL-SLR Report Iitems A-419, A-565, and A-716 address reduction of heat transfer due to fouling of copper alloy heat exchanger tubes exposed to air and condensation environments.

Table 3.3-1, Item Removed close cycle cooling water For the associated AMR item in SRP-SLR 127 as a susceptible environment. (Iitem 3.3.1-127), the listed environments erroneously include closed-cycle cooling water even though NUREG-2221, Table 2-3-46 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change 13 notes that the associated item in Table C2, Closed-Cycle Cooling Water System, was deleted because recurring internal corrosion is not anticipated in this system.

This oversight is being corrected in conjunction with the adjustments above for use of multiple AMPs.

Table 3.3-1, Item Added air-dry, air, and condensation Air and condensation environments were 138 as applicable environments and added to the scope of the Internal Table 3.3-1, Item updated applicable GALL-SLR Coatings/Linings for In-Scope Piping, Piping 139 Report items. Components, Heat Exchangers, and Tanks program with the issuance of SLR-ISG-2021-02-MECHANICAL, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance.

Table 3.3-1, Item Removed ultraviolet light and ozone Modified to incorporate industry operating 175 as applicable environments. experienceOE to update aging management of piping, piping components, and tanks made of fiberglass exposed to water and soil environments, by removing exposure to ultraviolet light and ozone as a cause of cracking, blistering, and loss of material.

Table 3.3-1, Item Removed ultraviolet light and ozone Modified to incorporate industry operating 175 as applicable environments. experience to update aging management of piping, piping components, and tanks made of fiberglass exposed to water and soil environments, by removing exposure to ultraviolet light and ozone as a cause of cracking, blistering, and loss of material.

Table 3.3-1, Item Added steel as an applicable Modified to note that the aging effects of 203 material and updated applicable loss of material, and long-term loss of GALL-SLR Report items. material due to general corrosion on steel exposed to an environment of treated water and sodium pentaborate can be managed by the Water Chemistry and One-Time Inspection AMPs. No item was added to manage stress corrosion cracking of steel in this environment as the GALL-SLR already states that steel components typically are not susceptible to stress corrosion cracking and are mainly susceptible to loss of material.

The staff determined that this material, environment, aging effect program (MEAP) may be managed with the AMPs cited above because the Water Chemistry AMP can monitor and control the concentration of deleterious species in the water storage tanks that provide water to the Standby 3-47 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Liquid Control (SLC) system which contains the sodium pentaborate solution.

Additionally, the One- Time Inspection AMP can verify the corrosion rate of the steel components is low enough that loss of material is unlikely to cause a loss of intended function.

Several reports were reviewed by the staff to make this determination [NUREG/CR-6001][EPRI Report 1010639][Metals Handbook Desk Edition, 2nd Edition][EPRI Report 1000975]. These reports concluded that even though the pH of the SLC system varies with temperature, it is generally greater than 6.8 pH which is close to neutral

[NUREG/CR-6001]. Additionally, these reports noted that the pH range in SLC systems tends to be between 6.8- - 8.5

[EPRI Report 1010639]. This would result in less corrosion of the steel as the corrosion rate of steel tends to decrease with an increasing (i.e. more basic) pH (i.e., more basic) and would need additional impurities (e.g., salts, oxygen) for appreciable corrosion to occur in this environment

([Metals Handbook Desk Edition, 2nd Edition)]. Additionally, one report found that corrosion rates of carbon and low-alloy steel, when exposed to varying concentrations of boric acid, were relatively low (0.05- - 1.1 mm/year [(0.002- - 0.045 inches/year])), when the temperature was below 60 °C (140 °F)([EPRI Report 1000975)].

Table 3.3-1, Item Clarified applicable materials a The Structure and/or Component was 255 metallic, replaced the term, changed from fire damper assemblies to assemblies, with housings, and fire damper housing because the housing removed hardening, loss of strength, is the passive component of the fire damper and shrinkage due to elastomer assembly that is subject to aging degradation as applicable aging management. The applicable material was effects/mechanisms. revised to metallic because fire damper housings are typically constructed of steel or stainless steel. The applicable aging effects were revised to loss of material due to general, pitting, and crevice corrosion, and cracking due to SCC because the elastomer aging effects of hardening, loss of strength, and shrinkage do not apply to metallic components. The fire damper housing is potentially subject to the cited aging effects.

For example, steel materials would not be 3-48 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change subject to SCC; however, stainless steel materials would be. The periodic inspections recommended by GALL-SLR AMP Report XI.M26 are capable of detecting these aging effects.

Table 3.3-1, Item Added flow blockage due to fouling Subsequent to issuance of the GALL-SLR 261 as an applicable aging Report, the staff recognized that to be effect/mechanism. consistent with other GALL-SLR Report items associated with heat exchanger tubes, E-475 should have also cited reduction of heat transfer due to fouling. This is consistent with GALL Report Revision 2 Iitem SP-41 where a material (i.e., stainless steel) that is not susceptible to loss of material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

Table 3.3-1, Item Added (Inspection of Internal Flow blockage due to fouling is not an 263 Surfaces only) to flow blockage due applicable aging effect requiring to fouling. management for the external environment of polymeric components.

Table 3.3-1, Item New Table 3.3-1 Iitem. New item to note that the aging effects of 264 loss of material, and long-term loss of material due to general corrosion on steel exposed to an environment of treated water and sodium pentaborate can be managed by the Water Chemistry and One-Time Inspection AMPs. No item was added to manage stress corrosion cracking of steel in this environment as the GALL-SLR already states that steel components typically are not susceptible to stress corrosion cracking and are mainly susceptible to loss of material.

The staff determined that this MEAP may be managed with the AMPs cited above because the Water Chemistry AMP can monitor and control the concentration of deleterious species in the water storage tanks that provide water to the Standby Liquid Control (SLC) system which contains the sodium pentaborate solution.

Additionally, the One- Time Inspection AMP can verify the corrosion rate of the steel components is low enough that loss of material is unlikely to cause a loss of intended function.

Several reports were reviewed by the staff to make this determination [NUREG/CR-3-49 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change 6001][EPRI Report 1010639][Metals Handbook Desk Edition, 2nd Edition][EPRI Report 1000975]. These reports concluded that even though the pH of the SLC system varies with temperature, it is generally greater than 6.8 pH which is close to neutral

[NUREG/CR-6001]. Additionally, these reports noted that the pH range in SLC systems tends to be between 6.8- - 8.5

[EPRI Report 1010639]. This would result in less corrosion of the steel as the corrosion rate of steel tends to decrease with an increasing (i.e. more basic) pH (i.e., more basic) and would need additional impurities (e.g., salts, oxygen) for appreciable corrosion to occur in this environment

([Metals Handbook Desk Edition, 2nd Edition)]. Additionally, one report found that corrosion rates of carbon and low-alloy steel, when exposed to varying concentrations of boric acid, were relatively low (0.05- - 1.1 mm/year [(0.002- - 0.045 inches/year])), when the temperature was below 60 °C (140 °F)([EPRI Report 1000975)].

Table 3.3-1, Item New Table 3.3-1 items. Two new items on hHeat exchanger tubes 265 are added to reflect that the Fuel Oil Table 3.3-1, Item Chemistry program is capable of mitigating 266 reduction of heat transfer for heat exchanger tubes by periodic sampling of fuel oil for contaminants that may cause the reduction of heat transfer due to fouling. The Fuel Oil Chemistry program can manage contaminants that would promote corrosion (e.g., water or microbial activity), particulate concentration, or other contaminants that tested for under ASTM D975 that could contribute to heat exchanger tube fouling. If operating experienceOE, or plant specific configurations, indicate other fouling mechanisms for a fuel oil environment may be present or the Fuel Oil Chemistry program alone is not sufficient to manage aging, the staff may need to evaluate whether the Fuel Oil Chemistry program is appropriate to manage these aging effects and if a One-Time Inspection is needed for a given plant.

Table 3.3-1, Item New Table 3.3-1 item. A new item for subliming compounds used 267 as fireproofing/fire barriers is being added because they are materials that are widely used throughout industry and are likely to be 3-50 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change cited in future SLRAs. The aging effects and aging mechanisms for subliming compounds used as fireproofing/fire barriers exposed to air are based on the NRC staffs review and approval of applicants programs for aging management of fire protection materials listed in previous SLRAs. In addition, the aging effects and aging mechanisms are consistent with Section 6, Fire Barriers, of EPRI Report 3002013084, Long-Term Operations: Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), issued November 2018, and those cited by industry as part of SLRA lessons learned activities and public comments on the draft AMR item.

The new item manages loss of material due to abrasion, flaking, and vibration; cracking/delamination due to chemical reaction and settlement; change in material properties due to gamma irradiation exposure; and separation for subliming compounds (Thermo-lag, Darmatt', 3M' Interam', and other similar materials) exposed to air.

The periodic inspections recommended by AMP XI.M26, Fire Protection, are capable of detecting these aging effects for these materials.

Table 3.3-1, Item New Table 3.3-1 item. A new item for cementitious coatings used 268 as fireproofing/fire barriers is being added because they are materials that are widely used throughout industry and are likely to be cited in future SLRAs. The aging effects and aging mechanisms for cementitious coatings used as fireproofing/fire barriers exposed to air are based on the NRC staffs review and approval of applicants programs for aging management of fire protection materials listed in previous SLRAs. In addition, the aging effects and aging mechanisms are consistent with Section 5, Structural Concrete Members, and Section 6, Fire Barriers, of EPRI 3002013084, and those cited by industry as part of SLRA lessons learned activities and public comments on the draft AMR item.

This item manages loss of material due to abrasion, exfoliation, elevated temperature, 3-51 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change flaking, and spalling; cracking/delamination; change in material properties; and separation for cementitious coatings (Pyrocrete, BIO' K-10 Mortar, Cafecote, and other similar materials) exposed to air.

Table 3.3-1, Item New Table 3.3-1 item. A new item for silicates used as 269 fireproofing/fire barriers is being added because they are materials that are widely used throughout industry and are likely to be cited in future SLRAs. The aging effects and aging mechanisms for silicates used as fireproofing/fire barriers exposed to air are based on the NRC staffs review and approval of applicants programs for aging management of fire protection materials listed in previous SLRAs. In addition, the aging effects and aging mechanisms are consistent with Section 6 of EPRI Report 3002013084, and those cited by industry as part of SLRA lessons learned activities and public comments on the draft AMR item.

The new item manages loss of material due to abrasion and flaking; cracking/delamination due to settlement; change in material properties due to gamma irradiation exposure; and separation for silicates (Marinite, Kaowool', Cerafiber, Cera blanket, or other similar materials) exposed to air.

The periodic inspections recommended by AMP XI.M26 are capable of detecting these aging effects for these materials.

1 Table 3-6 SRP-SLR, Revision 1, Chapter 3.4, Steam and Power Conversion Systems, 2 Differences from SRP-SLR, Revision 0, and Their Technical Location of Change Summary of the Change Technical Basis for Change Section 3.4.3.5 Add reference to the FSAR This information was inadvertently omitted Final Safety Supplement information contained in when the FSAR Supplement information Analysis Report GALL-SLR Table X-01 and Table was relocated from the SRP-SLR tables to (FSAR) XI-01. the GALL-SLR tables.

Supplement Section 3.4.2.2.2 Revised the last paragraph to state The statement in Section 3.4.2.2.2 that that the applicant may mitigate or cracking due to SCC does not require prevent cracking due to stress management contradicts the statement in corrosion cracking (SCC) using a SLR-SRP Section A.1.2.1, Item 5, which barrier coating rather than saying states that even with a prevention or 3-52 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change cracking due to SCC does not mitigation program, including a coating, an require management if a barrier aging effect should be identified as coating is used. Added a statement applicable for SLR, and the aging that the applicant should identify management review (AMR) should cracking due to SCC as applicable consider the adequacy of the AMP for subsequent license renewal referencing the prevention or mitigation (SLR) and identify the aging program (e.g., coating). Section A.1.2.1, management program (AMP) that Item 5, correctly states the need for an will be used to manage the integrity AMP to manage coating integrity.

of the coating.

Section 3.4.3.2.2 Revised the wording in the last The statement in Section 3.4.3.2.2 that paragraph to delete the statement barrier coatings make aging management that LOM does not require aging unnecessary is inconsistent with Section management if a barrier coating is A.1.2.1, Item 5, which correctly states that used. Deleted the statement that the an AMP is needed to manage the integrity reviewer verifies the barrier coating of the coating that prevents the aging is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluation (FE) review.

Section 3.4.2.2.3 Revised the last paragraph to state The statement in Section 3.4.2.2.3 that that the applicant may mitigate or loss of material does not require prevent loss of material using a management contradicts the statement in barrier coating rather than saying SLR-SRP Section A.1.2.1, Item 5, which the loss of material does not require states that even with a prevention or management if a barrier coating is mitigation program, including a coating, an used. Added a statement that the aging effect should be identified as applicant should identify loss of applicable for SLR, and the AMR should material as applicable for SLR and consider the adequacy of the AMP identify the AMP that will be used to referencing the prevention or mitigation manage the integrity of the coating. program (e.g., coating). Section A.1.2.1, Item 5, correctly states the need for an AMP to manage coating integrity.

Section 3.4.3.2.3 Revised the wording in the last The statement in Section 3.4.3.2.3 that paragraph to delete the statement barrier coatings make aging management that LOM does not require aging unnecessary is inconsistent with Section management if a barrier coating is A.1.2.1, Item 5, which correctly states that used. Deleted the statement that the an AMP is needed to manage the integrity reviewer verifies the barrier coating of the coating that prevents the aging is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.4.2.2.7 Revised the wording in the last The purpose of this change is to make the paragraph to clarify that the wording about barrier coatings consistent applicant may mitigate or prevent with that of other FE sections that required cracking due to SCC using a barrier changes to the coatings discussion.

coating. Added a statement that the applicant should identify cracking due to SCC as applicable for SLR and identify the AMP that will be 3-53 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change used to manage the integrity of the coating.

Section 3.4.3.2.7 Revised the wording in the fourth The statement in Section 3.4.3.2.7 that paragraph to delete the statement barrier coatings make aging management that cracking due to SCC does not unnecessary is inconsistent with Section require aging management if a A.1.2.1, Item 5, which correctly states that barrier coating is used. Deleted the an AMP is needed to manage the integrity statement that the reviewer verifies of the coating that prevents the aging the barrier coating is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.4.2.2.9 Revised the wording in the last The purpose of this change is to make the paragraph to clarify that the wording about barrier coatings consistent applicant may mitigate or prevent with that of other FE sections that required LOM using a barrier coating. Added changes to the coatings discussion.

a statement that the applicant should identify LOM as applicable for SLR and identify the AMP that will be used to manage the integrity of the coating.

Section 3.4.3.2.9 Revised the wording in the last The statement in Section 3.4.3.2.9 that paragraph to delete the statement barrier coatings make aging management that LOM does not require aging unnecessary is inconsistent with Section management if a barrier coating is A.1.2.1, Item 5, which correctly states that used. Deleted the statement that the an AMP is needed to manage the integrity reviewer verifies the barrier coating of the coating that prevents the aging is impermeable. effect. Because an AMP will be used to manage coating integrity, an evaluation of coating permeability is not necessary for this further evaluationFE review.

Section 3.4.2.2.9 Revised the wording to indicate that During the first revision/ISG to the SRP-SLR, loss of material due to pitting and the staff added a provision to the further crevice corrosion need not be evaluation sections associated with loss of managed if the type of aluminum is material due to pitting and crevice corrosion not susceptible to cracking and of aluminum components exposed to air or plant-specific operating experience condensation. This change allowed an does not reveal any issues related to loss of material due to pitting or alternative to conducting a one-time crevice corrosion. inspection to detect loss of material. Loss of material need not be managed if: (a) the aluminum material is not susceptible to cracking; and (b) plant-specific operating experience does not reveal any issues related to loss of material due to pitting or crevice corrosion. The staff included this alternative because: (a) it is unlikely that pitting or crevice corrosion in aluminum components would lead to a loss of intended function; and (b) if loss of material has not 3-54 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change been identified as an issue after 40 years (the earliest point at which a SLRA can be submitted), it is unlikely that it will lead to a loss of intended function during the subsequent period of extended operation.

This alternative was not allowed for aluminum materials that are susceptible to stress corrosion cracking because pitting or crevice corrosion might, but not necessarily, be a precursor to cracking.

Table 3.4-1, Item Added malleable iron as an During its review of recent SLRA plant-032 applicable material. specific operating experienceOE, in Table 3.4-1, Item response to the staffs observation regarding 033 dark corrosion product layers indicative of Table 3.4-1, Item graphitic corrosion on the internal surfaces of 068 malleable iron fittings exposed to a closed-cycle cooling water environment (ADAMS Accession No. ML22010A129), the staff has revised guidance documents (i.e., GALL-SLR Report and SRP-SLR) to include malleable iron as a material susceptible to selective leaching.

Table 3.4-1, Item Added flow blockage due to fouling Subsequent to issuance of the GALL-SLR 134 as an applicable aging Report, the staff recognized that to be effect/mechanism. consistent with other GALL-SLR Report items associated with heat exchanger tubes, E-475 should have also cited reduction of heat transfer due to fouling. This is consistent with GALL Report Revision 2 item SP-41 where a material (i.e., stainless steel) that is not susceptible to loss of material (a potential source of fouling products), is susceptible to reduction of heat transfer due to fouling.

Table 3.4-1, Item Added (Inspection of Internal Flow blockage due to fouling is not an 135 Surfaces only) to flow blockage due applicable aging effect requiring to fouling. management for the external environment of polymeric components.

Table 3.4-1, Item Revised the item to include carbon The item is revised to include CFRP repaired 125 fiber reinforced polymer (CFRP) piping components and credit the new AMP repaired piping as applicable XI.M43, High Density Polyethylene (HDPE) components and to refer to the new AMP XI.M43, High Density Piping and Carbon Fiber Reinforced Polymer Polyethylene (HDPE) Piping and (CFRP) Repaired Piping, to manage the Carbon Fiber Reinforced Polymer effects of age-related degradation (CFRP) Repaired Piping.. mechanisms that are applicable to HDPE piping and CFRP repaired piping. This new Added GALL-SLR Item VIII.H.S-484 to item 125. AMP reflects the recent introduction and increasing use of CFRP repaired piping at 3-55 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change reactor facilities. The unique aging issues and aging management approaches for CFRP repaired piping and HDPE piping (previously managed by AMP XI.M41) were considered to be most effectively addressed with a dedicated AMP.

1 Table 3-7 SRP-SLR, Revision 1, Chapter 3.5, Containments, Structures, and 2 Component Supports, Differences from SRP-SLR, Revision 0, and Their 3 Technical Bases Location of Change Summary of the Change Technical Basis for Change Section 3.5.2.2.1.2 Modifies SRP-SLR Section 3.5 Modifications to SRP-SLR Further Section 3.5.3.2.1.2 Further Evaluation sections to Evaluation sections provide the option to Section 3.5.2.2.1.7 manage the effects of aging in use plant-specific enhancements to GALL-Section 3.5.3.2.1.7 concrete for the following: SLR Report AMP XI.S2, ASME Section XI Section 3.5.2.2.1.8 Subsection IWL, AMP XI.S6, Structures Section 3.5.3.2.1.8

• Rreduction of strength and modulus Monitoring, or other AMPs in lieu of a plant-Section 3.5.2.2.1.9 of elasticity due to elevated specific AMP. The option to use plant-Section 3.5.3.2.1.9 temperature (greater than> 66 specific enhancements to GALL-SLR Section 3.5.2.2.2.1 degrees Celsius [ C](150 degrees Report AMPs increases the efficiency of Section 3.5.3.2.2.1 Fahrenheit [ F]) general: greater than subsequent license renewal application Section 3.5.2.2.2.2 93 Cdegrees Celsius (200 F (SLRA) reviews by limiting the use of AMR Section 3.5.3.2.2.2 degrees Fahrenheit) local). Note E designations for plant-specific Section 3.5.2.2.2.3 aging management activities when aging Section 3.5.3.2.2.3 *L loss of material (spalling, scaling) effects are managed by a plant-specific Section 3.5.2.2.2.6 and cracking due to freeze-thaw. AMP.

Section 3.5.3.2.2.6

• C cracking due to expansion from reaction with aggregates.

• I increase in porosity and permeability; loss of strength due to leaching of calcium hydroxide and carbonation

• R reduction of strength; loss of mechanical properties due to irradiation (i.e., radiation interactions with material and radiation-induced heating) for boiling-water reactor (BWR) and pressurized-water reactor (PWR) components in question, including those located in inaccessible areas.

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Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Section 3.5.2.2.1.5 Modifies SRP-SLR Further ASME Code,Section III, Division 1, includes Section 3.5.3.2.1.5 Evaluation sections to provide the provisions to analytically address option to perform a further evaluation cumulative fatigue damage (cracking due to based on ASME Code,Section III, cyclic loading) through detailed fatigue Division 1, Subsection NE, fatigue analysis or fatigue waiver analysis. If the waiver analysis for containment code criteria for a fatigue waiver are metallic pressure-retaining boundary satisfied, then a detailed fatigue analysis is components that are subject to cyclic not required. SRP-SLR Section 4.6.1, loading but have no current licensing Areas of Review, states that ASME Code basis (CLB) fatigue analysis. fatigue analyses and fatigue waiver analyses that are in the CLB may be time-limited aging analyses (TLAAs).

The fatigue waiver analysis described in this change is a TLAA, except that it will not be in the CLB at the time of a subsequent license renewal application (SLRA) submittal. It therefore does not meet the sixth criterion of 10 CFR 54.3, Definitions, for TLAA, which states that the analysis [is]

contained or incorporated by reference in the CLB. Nevertheless, as indicated above, performing a fatigue waiver analysis in accordance with the ASME Code is a technically acceptable approach to analytically address the effects of cyclic loading (fatigue aging effects) for containment metallic pressure-retaining boundary components. Therefore, satisfying the six conditions for fatigue waiver analysis in the ASME Code for applicable component materials provides an acceptable technical basis to demonstrate that a detailed fatigue analysis is not required, and cracking due to cyclic loading is not an aging effect requiring management. Therefore, the revised further evaluation section and modified AMR line items in this change provide one acceptable approach to address the aging effect of cracking due to cyclic loading for specific containment metallic pressure-retaining boundary components in lieu of supplemental surface examinations or performing or crediting an appropriate leak-rate test pursuant to Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, to 10 CFR 3-57 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Part 50, Domestic licensing of production and utilization facilities, as discussed in GALL-SLR Report AMP XI.S1, ASME Section XI, Subsection IWE, for which no CLB fatigue analysis exists at the time of SLRA submittal.

Section 3.5.2.2.2.4 The staff added aluminum alloyAlloy This is a wrought material alloyed primary 6063T6 to the list of materials that with magnesium (Mg) and silicon (Si). It is a are not susceptible to SCC. moderate strength precipitation hardened aluminum alloy in the peak-aged condition.

The strengthening phase precipitated during the artificial aging of Alloy 6063 is Mg2Si.

Generally, 6xxx series alloys have satisfactory SCC resistance and inservice performance. However, some 6xxx series alloys are known to be susceptible to SCC when exposed to certain atypical processing histories. The majority of 6xxx series SCC testing and characterization has been performed on Alloy 6061T6, which is known to be resistant to SCC. Much more limited SCC testing and characterization has been performed on Alloy 6063T6, although results have been consistent with those of Alloy 6061T6. Alloy 6063 is a compositionally leaner version of 6061 that has been optimized for extrusion. The two alloys have the same strengthening mechanism and their nominal Mg/Si ratios are also similar. Therefore, it is expected that the SCC performance is comparable.

Additionally, the known inservice performance of aluminum alloy 6063T6 has shown satisfactory SCC resistance across multiple industries. Based on the metallurgical characteristics, available laboratory testing, and known service history, the staff has determined that Alloy 6063T6 is not susceptible to SCC.

Sections Added new further evaluation (FE) To understand the serviceability of an aged 3.5.2.2.2.87 and acceptance criteria Section structure, it is important to understand the Section 3.5.2.2.2.7 (and corresponding applicable aging mechanisms, and more 3.5.3.2.2.87 review procedure Section 3.5.3.2.2.7) importantly, their effects on the ability of that Table 3.5-1, Item to address combined effects of aging structure to safely operate during the associated with radiation exposure of subsequent period of extended operation 102 the reactor vessel (RV) structural (SPEO). Reduction in fracture toughness Table 3.5-1, Item support assembly (e.g., reduction in due to irradiation embrittlement (IE) from 103 fracture toughness of RV steel accumulated neutron exposure through the 3-58 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change structural support components and SPEO could occur in BWR and PWR potential loss of intended function for reactor vessel (RV) steel structural support the support assembly as a whole, components (e.g., RV steel girder and including nonconcrete, non-metallic, column supports, neutron shield tank, nonferrous components and/or support skirt). Reduction in fracture materials). toughness (and other potentially combined aging effects associated with irradiation New aging management review such as loss of preload and distortion) could (AMR) line items 101 and 102 compromise the structural integrity of the (notwithstanding those considered above steel structural components and additionally by an applicant as plant- could also result in loss of intended function specific AMRs) associated with the of the RV structural support assembly, new further evaluationFE and including related nonconcrete, non-metallic, corresponding references are also components or /materials (e.g., Lubrite in added to SRP-SLR Table 3.5-1. sliding surfaces of the assembly) and nonconcrete, nonferrous components or

/materials (e.g., manganese bronze alloy).

To address the above concerns, a new further evaluationFE acceptance criteria, Section 3.5.2.2.2.7, and corresponding review procedures Section 3.5.3.2.2.7, are added to determine whetherif a plant-specific aging management program (AMP) or plant-specific enhancements to selected GALL-SLR AMPs are needed to manage the effects of aging due to combined mechanisms that could lead to loss of intended function, for example those attributed to irradiation, corrosive media (boric acid), large temperature variations, cyclic loading, and stress, in steel and related nonconcrete, non-metallic, nonferrous components of the RV structural support assembly for the SPEO. New AMR line Iitems 101 and 102 (not withstanding those added by applicants as plant-specific AMRs) associated with the further evaluationFE are also added to SRP-SLR Table 3.5-1.

The criteria and technical evaluation procedures (with the exception of the structural consequence analysis in Section 4.5) in NUREG-1509 Radiation Effects on Reactor Pressure Vessel Supports, May 1996 provide one acceptable methodology for performing a further evaluationFE for irradiation embrittlementIE of the RV steel structural support components.

Alternatively, applicant -proposed methodologies are acceptable on the basis 3-59 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change that comprehensive inspections and examinations of the RV structural support assembly noted above precede and are included on an ongoing basis in the recommended further evaluationFE analysis subject to GALL-SLR guidance and the ASME Code requirements.

Section 3.5.2.2.2.7 Added new Further Evaluation The staff added wooden poles as a Section 3.5.3.2.2.7 Ssection to address loss of material, new structural component requiring an Table 3.5-1, Item and/or changes in material properties aging management reviewAMR since 101 of wooden poles due to weathering, current guidance does not provide chemical degradation, insect recommendations to adequately infestation, repeated wetting and manage aging effects for this drying, or fungal decay. component. During recent SLRAs, licensees identified standing wooden poles as a structural componentstructural component within the scope of subsequent license renewalSLR that required aging management reviewAMR. However, the lack of a clear guidance has resulted in inefficiencies during the review process, in part, because industrys recommendations and guidelines for the inspections of wooden poles are different from those normally recommended by the GALL-SLR Rreport for other structural components.

Wooden poles are generally used on site for power distribution and function as structural supports for utility line distribution, or for support of other essential electrical components (e.g.,

cables, power conductors, pole transformers) with a safety-related function or a function related to NRC regulations such as station blackout.

These wooden poles, typically, are treated with wood preservatives that protect them from deterioration.

Although, these preservatives typically have a limited life expectancy, it is possible to significantly increase the service life of wooden poles through inspections, remediations, and management of prevalent aging effects. For plants entering the subsequent period of extended operation, it is expected that these wooden poles will remain in service 3-60 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change past the service life of the original preservative and therefore will be susceptible to deterioration. Thus, a plant-specific AMP or plant-specific enhancements to an existing aging management programAMP is recommended to adequately manage the aging effects in wooden poles for the loss of material, and/or changes in material properties due to weathering, chemical degradation, insect infestation, repeated wetting and drying, or fungal decay during the subsequent period of extended operation.

Decay of wooden poles is usually a gradual deterioration caused by fungi and other low forms of plant life. In most cases, the decay of wooden poles will be just below the groundline where the conditions of moisture, temperature and air are most favorable for the loss of material, and/or changes in material properties due to growth of fungi and other deteriorations due to the site-specific environmental conditions.

Factors affecting the service life of wooden poles are the species of wood, type and thoroughness of treatment, geographical location, and soil conditions. Since these factors are considered site-specific, it is recommended to develop further evaluationFE criteria to adequately address the site-specific conditions and establish criteria for a plant-specific AMP or plant-specific enhancements to an existing aging management programAMP required to manage the effects of aging for wooden poles during the subsequent period of extended operation. It is expected that the type and frequency of periodic inspections will vary by region and be determined based on site- specific conditions.

Although visual inspection might be considered a good first step, visual inspection alone will not detect majority of defective wooden poles since most decay tends to occur underground 3-61 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change and/or internally. Thus, the use of existing aging management programAMP, such as the Structures Monitoring program, needs to be enhanced with additional inspection methods, frequency, and acceptance criteria to adequately detect and manage the effects of aging for wooden poles before there is a loss of intended function.

Section 3.5.3.5 Add reference to the FSAR This information was inadvertently omitted Final Safety Supplement information contained in when the FSAR Supplement information Analysis Report GALL-SLR Table X-01 and Table XI- was relocated from the SRP-SLR tables to (FSAR) 01. the GALL-SLR tables.

Supplement Section 3.5.6 Added the following references. New references are added for the new further evaluationFE Sections 3.5.2.2.2.7

21. NUREG-1509, Radiation Effects and 3.5.3.2.2.7 in SRP-SLR to address on Reactor Pressure Vessel combined aging effects associated with Supports, U.S. Nuclear irradiation of RV steel structural support Regulatory Commission, assembly components.

Washington DC, May 1996.

22. ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix A, Analytical Evaluation of Flaws, New York, New York:

The American Society of Mechanical Engineers, 1989 Edition up to Edition incorporated by reference in 10 CFR 50.55a.

23 ASTM E693-17, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacement Per Atom (DPA), ASTM International.

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Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Table 3.5-1, Item Modified to provide the option to The option to use plant-specific 003 use plant-specific enhancements to enhancements increases the efficiency of Table 3.5-1, Item GALL-SLR Report AMP XI.S2, subsequent license renewal application 011 ASME Section XI, Subsection reviews by limiting the use of AMR Note Table 3.5-1, Item IWL, and/or GALL-SLR Report E designations for plant-specific aging 012 AMP XI.S6, Structures Monitoring, management activities when aging effects Table 3.5-1, Item in lieu of a plant-specific AMP. are managed through a plant-specific 014 AMP.

Table 3.5-1, Item 042 Table 3.5-1, Item 043 Table 3.5-1, Item 047 Table 3.5-1, Item 048 Table 3.5-1, Item 049 Table 3.5-1, Item 050 Table 3.5-1, Item 051 Table 3.5-1, Item 097 Table 3.5-1, Item Added a recommended Further Modification reflects the updated SRP-SLR 027 Evaluation section. Section 3.5.2.2.1.5 Further Evaluation Table 3.5-1, Item section.

040 Table 3.5-1, Item Updated the applicable GALL-SLR Item III.A6.TP-25 was deleted since it is a 054 Report items. duplicate to GALL-SLR Iitem no. III.A6.T-34.

In addition, this GALL-SLR item is associated with Group 6 structures and the associated SRP-SLR AMR item (i.e., Table 3.5-1, 054) is only intended to address all groups of structures, except Group 6.

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Draft Document: Tracked Changes Version 1 Table 3-8 SRP-SLR, Revision 1, Chapter 3.6, Electrical and Instrumentation Controls, 2 Differences from SRP-SLR, Revision 0 and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change Section 3.6.3.5 Add reference to the FSAR This information was inadvertently Final Safety Supplement information contained in omitted when the FSAR Supplement Analysis Report GALL-SLR Table X-01 and Table XI- information was relocated from the SRP-(FSAR) 01. SLR tables to the GALL-SLR tables.

Supplement Table 3.6-1, Item Added toughened glass, polymers, Modified to incorporate industry operating 002 silicone rubber, fiber glass, and experience to enhance aging management aluminum alloy to the list of of high voltage insulators to include applicable materials and clarified that additional susceptible materials (toughened loss of material is applicable to glass; polymers silicone rubber; fiberglass, metallic connectors. aluminum alloy).

Table 3.6-1, Item Added toughened glass, polymers, Modified to incorporate industry operating 003 silicone rubber, fiber glass, and experience to enhance aging management aluminum alloy as applicable of high voltage insulators to include materials and added peeling of additional susceptible materials (toughened silicone rubber sleeves for polymer glass; polymers silicone rubber; fiberglass, insulators or degradation of glazing aluminum alloy) and aging effects (peeling on porcelain insulators as applicable of silicone rubber sleeves for polymer aginge effects or /mechanisms. insulators; or glazing degradation for porcelain insulators).

3 Table 3-9 SRP-SLR, Revision 1, Chapter 4.1, Identification of Time-Limited Aging 4 AnalysisTLAA, Differences from SRP-SLR, Revision 0, and Their Technical 5 Bases Location of Change Summary of the Change Technical Basis for Change No changes from SRP-SLR, Revision 0, to SRP-SLR Revision 1.

6 Table 3-10 SRP-SLR, Revision 1, Chapter 4.2 (Neutron Irradiation Embrittlement) 7 Differences from SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change SRP-SLR Section 4.2: Reactor Pressure Vessel Neutron Embrittlement Analysis 4.2.1 Areas of Review Section 4.2.1 was revised to Revision to Section 4.2.1 to provide provide details from Regulatory context from Regulatory Issue Issue Summary (RIS) 2014-11 Summary (RIS) 2014-11 in terms of and to ensure coordination the traditional geometric beltline as between the U.S. Nuclear defined in 10 CFR Part 50, Appendix Regulatory Commission (NRC) G, and all other reactor vessel (RV) staff review of time-limited aging ferritic materials with projected neutron analysis (TLAAs) related to fluence values greater than 1 xx 1017 reactor pressure vessel (RPV) Newton per square centimeter embrittlement. (Nn/cm2) (E > 1 MeV).

3-64 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change Additionally, Section 4.2.1 was revised to acknowledge the common use of material properties, copper and nickel content values for reactor vesselRV materials in several RPV embrittlement TLAAs and provide guidance to ensure coordination between the NRC staff reviews.

4.2.3.1.2.1 10 CFR Section 4.2.3.1.2.1 was revised Revision to Section 4.2.3.1.2.1 and 54.21(c)(1)(i) to provide additional guidance to Section 4.2.3.1.2.2 is due to the NRC the NRC staff to ensure the staff review experience from 4.2.3.1.2.2 10 CFR appropriate disposition of upper- subsequent license renewal 54.21(c)(1)(ii) shelf energyUSE TLAA based applications (SLRAs).

on the different circumstances.

Section 4.2.3.1.2.1 is revised to clarify Section 4.2.3.1.2.2 was revised that the assessment of additional to provide additional context reactor vessel RV materials that were and guidance for the NRC staff not previously addressed in the current review of USE TLAAs. licensing basis (CLB) due to the projected neutron fluence exposure at the end of the subsequent period of extended operation. The assessment of these reactor vesselRV materials constitutes a revision to CLB analysis for USE; thus, a disposition in accordance with 10 CFR 54.21(c)(1)(i).for the USE TLAA of these additional reactor vesselRV materials is not appropriate.

Section 4.2.3.1.2.2 was revised to provide additional context and guidance for the NRC staff review of USE TLAAs that involve: (1) revision to CLB material property information as part of the application and (2) the inclusion of reactor vesselRV materials not previously addressed in the CLB but are now necessary due to the projected neutron fluence exposure at the end of the subsequent period of extended operation. Additionally, Section 4.2.3.1.2.2 was revised to provide guidance consistent with NRC Regulatory Guide (RG) 1.99, Revision

2. for the NRC staff review of the applicants use of surveillance data in the USE TLAAs.

4.2.3.1.3.1 10 CFR Section 4.2.3.1.3.1 was revised Revision to Section 4.2.3.1.3.1 and 54.21(c)(1)(i) to provide additional guidance to Section 4.2.3.1.3.2 is due to the NRC the NRC staff to ensure the staff review experience from appropriate disposition of 3-65 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change 4.2.3.1.3.2 10 CFR pressurized thermal shock (PTS) subsequent license renewal 54.21(c)(1)(ii) TLAA based on the different applicationsSLRAs.

circumstances.

Section 4.2.3.1.3.1 is revised to clarify Section 4.2.3.1.3.2 was revised that the assessment of additional to provide additional context reactor vesselRV materials that were and guidance for the NRC staff not previously addressed in the current review of PTS TLAAs. licensing basis (CLB) due to the projected neutron fluence exposure at the end of the subsequent period of extended operation. The assessment of these reactor vesselRV materials constitutes a revision to CLB analysis for PTS; thus, a disposition in accordance with 10 CFR 54.21(c)(1)(i).for the PTS TLAA of these additional reactor vesselRV materials is not appropriate.

Section 4.2.3.1.3.2 was revised to provide additional context and guidance for the NRC staff review of PTS TLAAs that involve: (1) revision to CLB material property information as part of the application and (2) the inclusion of reactor vesselRV materials not previously addressed in the CLB but are now necessary due to the projected neutron fluence exposure at the end of the subsequent period of extended operation. Additionally, Section 4.2.3.1.3.2 was revised to provide guidance consistent with 10 CFR 50.61 and NRC Regulatory Guide 1.99, Revision 2. for the NRC staff review of the applicants use of surveillance data in the PTS TLAAs.

4.2.3.1.4 Section 4.2.3.1.4 was revised to Revision to Section 4.2.3.1.4 is due to Pressure- provide additional details and the NRC staff review experience from Temperature guidance related to adjusted subsequent license renewal Limits reverence temperature, which applicationsSLRAs.

are used for determining Pressure-Temperature Limits. Based on past experience, some applicants have provided a TLAA for Aadjusted Rreference Ttemperature.

However, there is not specific acceptance criteria or limits for adjusted reference temperature; rather it is a key input for Pressure-Temperature Limits. As such, Section 4.2.3.1.4 is revised to include additional background and guidance in 3-66 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change the event Adjusted Reference Temperature is identified as a TLAA in the subsequent license renewal applicationSLRAs.

This additional guidance is consistent with the guidance established in SRP-SLR Sections 4.2.3.1.2.2 and 4.2.3.1.3.2 for USE and PTS TLAAs, respectively, and NRC Regulatory Guide 1.99, Revision 2, due to the common use of material properties, copper and nickel content values for reactor vesselRV materials in these reactor pressure vessel embrittlement analyses.

Section 4.2.3.1.5, Elimination Sections 4.2.2.1.5 and 4.2.3.1.5 BWRVIP-329-A is an NRC-approved of Boiling Water Reactor were revised to incorporate the topical report. The purpose of Circumferential Weld guidance from Boiling Water BWRVIP-329-A is to use NRC safety Inspections Reactor Vessel and Internals goals and probabilistic fracture (Acceptance Criteria) Project (BWRVIP)-329-A, BWR mechanics analysis procedures that Vessel and Internals Program, have been developed since the Section 4.2.3.1.5, Elimination Updated Probabilistic Fracture publication of BWRVIP-05 to update of Boiling Water Reactor Mechanics Analyses for BWR the evaluation procedure and Circumferential Weld RPV Welds to Address acceptance criteria specified in Inspections Extended Operations as it BWRVIP-74-A for providing relief from (Review Procedures) relates to boiling water reactor examination of circumferential welds.

(BWR) circumferential weld The results of this report identify the inspections. combinations of beltline material conditions for the BWR fleet that will ensure regulatory safety goals are satisfied for the postulated transient.

The results from this report can be used to demonstrate that reactor pressure vesselsRPV in the BWR fleet have margins against failure that satisfy regulatory criteria through at least an 80-year operating interval for the postulated, low temperature isothermal pressure transient.

Section 4.2.2.1.6 Boiling Sections 4.2.2.1.6 and 4.2.3.1.6 The BWRVIP-329-A is an NRC-Water Reactor Axial Welds were revised to incorporate the approved topical report. The purpose (Acceptance Criteria) guidance from BWRVIP-329-A, of BWRVIP-329-A is to use NRC BWR Vessel and Internals safety goals and probabilistic fracture Section 4.2.3.1.6 Boiling Program, Updated Probabilistic mechanics analysis procedures that Water Reactor Axial Welds Fracture Mechanics Analyses have been developed since the (Review Procedures) for BWR RPV Welds to Address publication of BWRVIP-05 to update Extended Operations as it the evaluation procedure and relates to boiling water reactor acceptance criteria specified in axial welds. BWRVIP-74-A for assessing axial weld integrity. The results of this report identify the combinations of beltline material conditions for the BWR fleet 3-67 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change that will ensure regulatory safety goals are satisfied for the postulated transient. The results from this report can be used to demonstrate that reactor pressure vesselsRPV in the BWR fleet have margins against failure that satisfy regulatory criteria through at least an 80-year operating interval for the postulated, low temperature isothermal pressure transient.

Section 4.2.2.1.4.3 of 10 Section 4.2.2.1.4.3 was revised Revision to Sections 4.2.2.1.4.3 and CFR 54.21(c)(1)(iii) to clarify the circumstances in 4.2.3.1.4.3 are due to the NRC staff which pressure-temperature P-T review experience from subsequent 4.2.3.1.4.3 10 CFR limits would be updated to license renewal applicationsSLRAs.

54.21(c)(1)(iii) account for the subsequent period of extended operation. Current guidance in these sections of the SRP-SLR could be interpreted that Section 4.2.3.1.4.3 was revised updated P-T limits must be established to clarify the circumstances in under the appropriate regulatory which P-T limits would be process prior to the plants entry into updated to account for the the subsequent period of extended subsequent period of extended operation even if the current terms of operation. applicability for the P-T limits have not been exceeded.

1 Table 3-11 SRP-SLR, Revision 1, Chapter 4.3, Metal Fatigue, Differences from SRP-2 SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

3 Table 3-12 SRP-SLR, Revision 1, Chapter 4.4, Environmental Qualification of Electrical 4 Equipment, Differences from SRP-SLR, Revision 0, and Their Technical 5 Bases Location of Change Summary of the Change Technical Basis for Change No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

6 Table 3-13 SRP-SLR, Revision 1, Chapter 4.5, Concrete Containment Unbonded 7 Tendon Prestress Analysis, Differences from SRP-SLR, Revision 0, and 8 Their Technical Bases Location of Change Summary of the Change Technical Basis for Change No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

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Draft Document: Tracked Changes Version 1 Table 3-14 SRP-SLR, Revision 1, Chapter 4.6, Containment Liner Plate, Metal 2 Containments, and Penetrations Fatigue Analysis, Differences from SRP-3 SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

4 Table 3-15 SRP-SLR, Revision 1, Chapter 4.7, Plant-Specific TLAA, Penetrations 5 Fatigue, Differences from SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change Table 4.7-1 The staff added EPRI MRP In the license renewal application Cycle-based and Fluence- (LRA) for the Waterford Nuclear Plant, Based Analyses in Support of the licensee identified the supporting MRP-227 as an additional MRP analyses in the MRP-191, example of a potential, plant- Revision 1 report as a plant-specific specific time-limited aging TLAA for the LRA. The licensee used analysis (TLAA) for PWR-design its PWR Vessel Internals Program and nuclear plants. the RVI component-specific inspection and evaluation I&E criteria called out in the MRP-227-A report to disposition the TLAA in accordance with the TLAA acceptance criterion in 10 CFR 54.21(c)(1)(iii).

The addition of this type of TLAA to the list of potential, plant-specific TLAAs for PWR-designed facilities accounts for the possibility that some LRA or SLRA applicants may identify these types of analysis as plant-specific TLAAs for their LRAs or SLRAs.

Section 4.7.4 SRP-SLR Sections 4.7.4, The SRP-SLR Section 4.7 describes Section 4.7.5 Evaluation Findings and 4.7.5, the SLR review plan for the other Section 4.7.6 References for the other plant- plant-specific TLAAs (i.e., TLAAs Table 4.7-1 specific time-limited aging other than those addressed in SRP-analyses (TLAAs) are SLR Sections 4.1 through 4.6). The renumbered to Sections 4.7.5 SRP-SLR Sections 4.7.1, 4.7.2 and and 4.7.6, respectively. 4.7.3 provide general guidance for the Relevant references are also areas of review, acceptance criteria, added in the renumbered and review procedures for the TLAA Section 4.7.6, References. reviews, respectively. However, these The new SRP-SLR Section sections do not address guidance that 4.7.4 is added to provide pertains to specific TLAAs within the additional guidance for the scope of SRP-SLR Section 4.7.

TLAA reviews within the scope of SRP-SLR Section 4.7. Recently, the U.S. Nuclear Regulatory Specifically, Sections 4.7.4.1 Commission (NRC) staff reviewed and 4.7.4.2 are added to subsequent license renewal (SLR) provide specific guidance for the applications for the first time and review of a leak-before-leak noted that the applications included (LBB) TLAA and a pump casing LBB TLAAs and cast austenitic 3-69 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Location of Change Summary of the Change Technical Basis for Change flaw tolerance TLAA, stainless steel (CASS) pump casing respectively. flaw tolerance TLAAs. Currently, the SRP-SLR does not include specific guidance for the review of these TLAAs. Therefore, new SRP-SLR sections are added to provide specific guidance for the review of an LBB TLAA and a CASS pump casing flaw tolerance TLAA.

The guidance for the LBB TLAA is based on SRP (NUREG-0800)

Section 3.6.3. The guidance for the pump casing flaw tolerance TLAA is based on the provisions in ASME Code Case N-481 with a reference to the NRC staff-approved PWROG-17033-NP-A Report, Revision 1.

1 Table 3-16 SRP-SLR, Revision 1, Chapter 5.0, Technical Specification Changes, 2 Differences from SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

3 Table 3-17 SRP-SLR, Revision 1, Appendices A.1, A.2, A.3, and A.4, Differences from 4 SRP-SLR, Revision 0, and Their Technical Bases Location of Change Summary of the Change Technical Basis for Change No changes from SRP-SLR, Revision 0, to SRP-SLR, Revision 1.

5 3-70 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 4 CHANGES TO TECHNICAL BASES DOCUMENTED IN INITIAL 2 NUREG-2221 3 After the issuance of the initial NUREG-2221, U.S. Nuclear Regulatory Commission (NRC) staff 4 made revisions to either the Summary of Significant Changes or Technical Bases for 5 Changes information that were documented within NUREG-2221 for certain Generic Aging 6 Lessons Learned for Subsequent License Renewal (GALL-SLR) Report and Standard Review 7 Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-8 SLR) changes. These changes do not affect the GALL-SLR Report or SRP-SLR items within the 9 tables of NUREG-2221, but provide a revised summary of significant changes or technical basis 10 to support the change from the GALL Report and SRP-LR to Revisions 0 of the GALL-SLR 11 Report and SRP-SLR. These revised summary of significant changes or technical bases have 12 been made as a result of lessons learned from the staffs review of subsequent license renewal 13 applications (SLRAs) as all as public comments received during the public comment period.

14 A summary of specific changes to the Summary of Significant Changes or Technical Bases 15 for Changes in the initial NUREG-2221 is provided in Table 4-2 Table 4-1. The technical bases 16 documented in this Supplement to NUREG-2221 is are intended to supersede the technical 17 bases documented in the initial NUREG-2221 for that particular table item. The following 18 describes the information presented in each column of these tables.

19 Table 4-1 Description of Table Columns for Technical Bases in Initial NUREG-2221 Column Heading Description Location of Change Identifies the location in the initial NUREG-2221 of the applicable change to either or both the Summary of Significant Changes and Technical Bases for Change.

Revised Summary of Provides the revised summary of the change to supersede the current Changes entry in the initial NUREG-2221. N/A means that there a summary of the change is not included for this particular change to the GALL-SLR Report or SRP-SLR. No change from initial NUREG-2221 entry means that there is no revision to the current summary in the initial NUREG-2221 table entry.

Revised Technical Bases Provides the revised technical bases of the change to supersede the for Change current entry in the initial NUREG-2221.

20 Table 4-2 Changes to Technical Bases in Initial NUREG-2221 Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change Initial NUREG- Not applicable (N/A) The SRP-SLR Section 3.5.2.2.2.4, Cracking Due 2221, Table 2-2 to Stress Corrosion Cracking, and Loss of Material Due to Pitting and Crevice Corrosion, III.B1.1.T-36a was revised and new line items, T-36 a- - c and III.B1.2.T-36a T-37 a- - c were added to address aluminum and III.B1.3.T-36a stainless steel (SS) support members; welds; III.B1.1.T-36b bolted connections; and support anchorage to III.B1.2.T-36b building structure exposed to air or condensation.

III.B1.3.T-36b The basis for the potential for aluminum and III.B1.1.T-36c stainless-steelSS components to experience loss III.B1.2.T-36c of material and cracking is established in the 4-1 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change III.B1.3.T-36c GALL-SLR Report and SRP-SLR Supplemental III.B2.T-37a Staff Guidance Document Supplement issued on III.B3.T-37a March 29, 2016, Agencywide Documents Access III.B4.T-37a and Management System (ADAMS) Accession III.B5.T-37a No. ML16041A090. An overview of this basis is III.B2.T-37b as follows.

III.B3.T-37b III.B4.T-37b The staff has concluded that air and III.B5.T-37b condensation environments can be aggressive if III.B2.T-37c halides are present. Halides can be present due III.B3.T-37c to leakage from flanged connections or valve III.B4.T-37c packing through insulation and raw water leakage III.B5.T- from flanged connections or valve packing. The 37c staff accepts that pressure boundary leakage would be considered as event driven and not as a potential source to transport halides to the surface of the aluminum component. However, SRP-SLR, Section A.1.2.1 states that, leakage from bolted connections should not be considered as abnormal events. Although bolted connections are not supposed to leak, experience shows that leaks do occur, and the leakage could cause corrosion. The outdoor air environment can contain halides due to nearby salted roads, ocean mist, cooling tower fallout if treatment chemicals contain halides, and nearby industrial facilities.

Loss of material due to pitting or crevice corrosion, and cracking due to stress corrosion cracking (SCC) of stainless steelSS components were addressed in GALL Report Revision 2 for in Chapters V, VII, and VIII. During the development of the GALL-SLR Report, the staff recognized that stainless steelSS support members should be addressed in addition to piping system components. As a result, SRP-SLR, Section 3.5.2.2.2.4 was revised and new AMR items were added.

Loss of material of aluminum components was addressed in GALL Report Revision 2 for in Chapters V, VII, and VIII by Pperiodic Iinspection programs. During the development of the GALL-SLR Report and SRP-SLR, the staff concluded that cracking of aluminum components should be addressed through a further Further Eevaluation (FE) Ssection. During the development of the GALL-SLR Report and SRP-SLR, the staff concluded that it may not be necessary to conduct periodic inspections of aluminum components in order to manage aging 4-2 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change effects associated with aluminum components.

The staff noted that Oone-Ttime Iinspections, as described by aging management program (AMP)

XI.M32, One-Time Inspection, for the subsequent license renewal (SLR) period would occur after no less than 50 years of operation.

The staff concluded that a Oone-Ttime Iinspection of aluminum components prior to entry in the subsequent period of extended operation coupled with a search of plant-specific operating experience (OE) related to loss of material of aluminum components would provide sufficient input to determine whether periodic inspections should be conducted. As a result, SRP-SLR, Section 3.5.2.2.2.4 was revised to address loss of material and cracking, and new AMR items were added.

If the OE search or one-time inspection results in conducting a periodic inspection of piping, piping components, and tanks, GALL-SLR Report, AMP XI.M36 recommends either surface examinations, American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section XI VT-1 inspections, or visual inspections where it has been analytically demonstrated that surface cracks can be detected by leakage prior to a crack challenging the structural integrity or intended function of the component. The staff did not include this specificity for the periodic inspection of aluminum and stainless steelSS support members; welds; bolted connections; and support anchorage to building structure in AMP XI.S3 and AMP XI.S6 because piping, piping components and tanks are less flaw tolerant than supports in that minor through-wall loss of material or cracking will result in leakage. The leakage, in and of itself, may not result in a loss of intended function; however, it could impact components in the vicinity of the flaw. In contrast, for a support, minor loss of material or cracking that might not be detectable during a walkdown inspection will likely not impact the intended function of the support and the staff has concluded that additional loss of material or crack growth will likely become more evident during periodic inspections of supports.

In contrast, the SRP- SLR recommends that one-time inspections for loss of material and cracking of aluminum and stainless steelSS supports be conducted in accordance with AMP XI.M32. The more rigorous examination techniques cited in 4-3 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change AMP XI.M32 can detect minor indications of loss of material and cracking. As a result, in the absence of adverse indications, it is reasonable to conclude that the environmental conditions will not cause loss of material or cracking in the future. If the one- time inspections and plant-specific operation experience (OE) do not reveal loss of material or cracking, no further inspections will be conducted during the subsequent period of extended operation.

A less rigorous approach to periodic inspections of supports, as compared to piping, is demonstrated as follows. The ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Table IWF-2500-1, Examination Categories, for Class 1, Class 2, and Class 3 Piping Supports, and Supports Other Than Piping Supports, requires that an owner conduct VT-3 inspections. The purpose of a VT-3, as stated in IWA-2213 is:

VT3 examination is conducted to determine the general mechanical and structural condition of components and their supports by verifying parameters such as clearances, settings, and physical displacements; and to detect discontinuities and imperfections, such as loss of integrity at bolted or welded connections, lose or missing parts, debris, corrosion, wear, or erosion. VT-3 includes examination for conditions that could affect operability or functional adequacy of constant load and springtype supports.

In contrast, the purpose of VT-1 examinations, as stated in IWA-2211, which are used to inspect pressure retaining components (e.g., nuts; bolts; flange surfaces; internal core support structures; welded attachments to Class 3 vessels, piping, pumps, and valves) is: VT-1 examination is conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion. This demonstrates the utilization of a more rigorous inspection methodology for pressure retaining components versus supports.

Initial NUREG- N/A The previous line items stated that the AMP in 2221, Table 2-17 GALL AMP XI.M2, Water Chemistry, may be used to manage any cracking that may occur in 4-4 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change IV.D1.RP-367 the components as a result of these types of IV.D1.RP-385 aging mechanisms, as assessed in conjunction IV.D2.RP-185 with the further evaluation (FE) of acceptance Acceptance Ccriteria and Rreview Pprocedure guidelines in Sections 3.1.2.2.11 and 3.1.3.2.1.11 of the SRP-LR Revision 2 report and their subsections. The further evaluationFE criteria basically recommended that the applicants perform an evaluation of these steam generator (SG) components to determine whether additional aging management activities or a plant-specific AMP would need to be implemented (i.e.,

in addition to implementation of the Water Chemistry program) in order to ensure adequate detection and management of cracking that may occur in the divider plates and tube-to-tubesheet welds during the period of extended operation.

The aging management review (AMR) Iitem No.

025 in Table 3.1-1 of the SRP-SL Revision 2 report referenced the following AMR items in NUREG-1801, Revision 2 for aging management:

(a) AMR Iitem IV.D1.RP-367 for primary side divider plates in recirculating steam generators that are made from either nickel alloy materials or steel with nickel alloy cladding and are exposed to a reactor coolant environments, (b) AMR Iitem IV.D1.RP-385 for the tube-to-tubesheet welds in recirculating steam generators that are made from nickel alloy materials and are exposed to a reactor coolant environment, and (c) AMR Iitem IV.D2.RP-185 for the tube-to-tubesheet welds in once-through steam generatorsSGs that are made from nickel alloy materials and are exposed to a reactor coolant environment.

An update of the staffs aging management guidelines for these components were issued in NRC License Renewal Interim Staff Guidance (LR-ISG) No. 2016-01, Changes to Aging Management Guidance for Various Steam Generator Components, dated December 7November 30, 2016 (ADAMS Accession No. ML16237A383).

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Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change Therefore, the staff determined that the previous versions of AMR Iitems IV.D1.RP-367, IV.D1.RP-385, and IV.D2.RP-185 in the GALL Revision 2 report were acceptable for retention in NUREG-2191, but with the need for certain modifications of the line item. Specifically, the staff updated the further evaluationFE acceptance Acceptance criteriaCriteria guidelines in SRP-SLR Section 3.1.2.2.11 and review procedure guidelines in SRP-SLR 3.1.3.2.11 to be consistent with the changes made to these sections in LR-ISG-2016-01, Changes to Aging Management Guidance for Various Steam Generator Components,. The AMR Iitems IV.D1.RP-367, IV.D1.RP-385, and IV.D2.RP-185 were then updated to indicate that the AMPs in GALL-SLR AMP XI.M2, Water Chemistry, and AMP XI.M19, Steam Generators, may be used to manage cracking in the components, when coupled to a the staffs further evaluationFE guidelines in SRP-SLR Section 3.1.2.2.11.1 for the assessment of divider plates in recirculating SGs (i.e., the subject of the AMR in Iitem IV.D1.RP-367) and in SRP-SLR Section 3.1.2.2.11.2 for the assessment of SG tube-to-tubesheet welds in recirculating and once-through SG designs (i.e., the subject of the AMRs in Iitems IV.D1.RP-385 and IV.D2.RP-185). AMR Iitem No. 025 in Table 3.1-1 in Table 3.1-1 of NUREG-2192 was modified accordingly.

Under the updated guidelines in LR-ISG-2016-01, implementation of the AMPs in GALL AMP XI.M19, Steam Generators, and GALL AMP XI.M2, Water Chemistry, are acceptable bases for managing any cracking that may occur in these components, when subject to and evaluated in accordance with the staffs updated further evaluationFE acceptance criteria for these components in SRP-SLR Sections 3.1.2.2.11.1 and 3.1.2.2.11.2. The corresponding review procedures for performing these reviews of these AMR items are given in SRP-SLR Sections 3.1.3.2.11.1 and 3.1.3.2.11.2.

The staff added a discussion of plant- specific SG design parameters that should be evaluated against the industry analyses (EPRI 3002002850) to determine whetherif a given plant is bounded by the industry analyses for SG divider plate cracking. This includes potential use of the checklist in EPRI letter SGMP IL 16 02 to demonstrate that plant- specific parameters are bound by the industry analyses. This is meant to 4-6 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change provide clarity to determine whether the industry analyses are applicable and bounding.

Additionally, the reference to a plant- specific AMP was replaced with the One- Time Inspection AMP because the GALL- SLR Rreport states that a plant- specific AMP may include a Oone- Ttime Iinspection that is capable of detecting cracking to verify the effectiveness of the Wwater Cchemistry and Ssteam Ggenerator programs and the absence of primary water stress corrosion cracking (PWSCC) in the divider plate assemblies. The One- Time Inspection AMP fulfills this recommendation from the GALL- SLR and eliminates the need for a plant- specific AMP to be evaluated.

Initial NUREG- No change from initial Changes were made to this program in a 2221, NUREG-2221 entry. consistent manner consistent with the updated Table 2-29 aging management guidance for steam generator AMP XI.M19 SG components described in License Renewal Steam Interim Staff Guidance (LR-ISG) 2016-01, Generators, Changes to Aging Management Guidance for Program Various Steam Generator Components. The Description technical bases of these changes are described in Scope of Program LR-ISG-2016-01 in detail (ADAMS Accession No.

Parameters ML16237A383). The associated Federal Register Monitored or Notice is 81 FR 88276 (December 7, 2016).

Inspected Detection of The LR-ISG-2016-01 also contains the staffs Aging Effects dispositions of public comments for the draft LR-Monitoring and ISG-2016-01 and indicates that the staff intends Trending to incorporate corresponding changes to the SLR Acceptance guidance. The technical bases as well as related Criteria background information are summarized below.

References The SRP-LR, Revision 2, Sections 3.1.2.2.11 and 3.1.3.2.11, Cracking due to Primary Water Stress Corrosion Cracking describe further evaluationFE regarding primary water stress corrosion cracking (PWSCC) in steam generator SG nickel alloy divider plate assemblies and tube-to-tubesheet welds. The main concern discussed in these further evaluationFE sections is that, when these components are fabricated with PWSCC-susceptible nickel alloy materials (e.g., Alloy 600/82/182), PWSCC could occur and such cracking could propagate into adjacent reactor coolant pressure boundary components (e.g., steam generator SG heads and tubesheets).

4-7 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change The further evaluationFE guidance in the SRP-LR is, in part, based on foreign operating experience that PWSCC occurred in steam generator SG divider plate assemblies. In previous license renewal applications (LRAs), applicants typically committed to inspection or analysis approaches that will confirm that PWSCC is not occurring in these components or any potential PWSCC does not affect the integrity or design functions of steam generator (SG) components.

Since the development of the further evaluationFE guidance in the SRP-SLR, the industry performed additional evaluations, tests and analyses regarding operating experience (including inspection results), characterization of material compositions in terms of susceptibility to PWSCC, potential significance of PWSCC to the integrity and design functions of SG components, and inspection activities credible to manage the aging effect. Based on these activities and findings, EPRI submitted to the NRC staff, EPRI 3002002850, Steam Generator Management Program: Investigation of Crack Initiation and Propagation in the Steam Generator Channel Head Assembly.

In its review of the Electric Power Research Institute (EPRI) report and related information, the staff found a need to update the guidance in the SRP-SLR further evaluationFE sections and GALL AMP XI.M19, Steam Generators, as further summarized below.

The susceptibility of a material to stress corrosion crackingSCC depends on three main factors:

susceptible material, conducive environment, and sufficiently high tensile stress. Therefore, these factors need to be considered in the evaluation of material susceptibility to PWSCC.

The cracks due to PWSCC in divider plate assemblies (foreign operating experience) tend to be very shallow (approximately 0.08 inches) and have not grown in depth since detection. These cracks are located in divider plates that were provided primarily by one manufacturer.

In addition, the cracks discussed above are believed to have initiated as a result of significant cold work introduced through surface grinding and stub runner distortion primarily attributed to hydrostatic testing of the steam generators. All but one of these instances of PWSCC have been 4-8 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change observed in the divider plate assemblies that are approximately 1.3 inches thick. Analyses by the industry in the foreign country further indicated that distortion of the stub runner is only expected to occur in thinner divider plates (i.e., 1.3 inches thick or less).

The foreign operating experience (OE) also indicates that fabrication issues (e.g., a misalignment between the stub runner plate and the divider plate after welding and subsequent realignment) may cause additional residual stresses and strains.

The U.S. industry has performed analyses assuming a fully degraded divider plate assembly. These analyses conclude that the potential degradation does not affect the design functions or safety-related analyses of steam generator components.

Based on crack growth and fracture mechanics analyses, cracks due to PWSCC in the divider plate assemblies are highly unlikely to affect the integrity of other pressure boundary components (such as the channel head and tube-to-tubesheet welds).

The inservice inspections performed in accordance with Section XI of ASME Code include periodic volumetric inspections of steam generatorSG head welds and tubesheet-to-head welds. The examination can confirm the structural integrity of the steam generatorSG head welds and tubesheet-to-head welds.

With respect to the tube-to-tubesheet welds, the weld chromium content for Alloy 690 tubes and Alloy 82 tubesheet cladding can range from approximately 24 to 26 percent% chromium and the weld chromium content for Alloy 690 tubes and Alloy 182 tubesheet cladding can range from approximately 21 to 23 percent%. In addition, the steam generatorSG tubesheet is in compression.

The staff has not identified any instances where cracks have been reported in the tubesheet cladding. Although it is unlikely that any inspections looking specifically for cracking have been performed, if cracking were prevalent, it would have most likely been detected during the performance of steam generator tube inspections.

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Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change Foreign and domestic operating experienceOE indicates that loss of material due to boric acid corrosion can occur in the steel base material of the steam generator channel head and tubesheet. This operating experienceOE is discussed in NRC Information Notice 2013-20, Steam Generator Channel Head and Tubesheet Degradation. One means to effectively manage this aging effect is to control the reactor water chemistry to mitigate the loss of material due to boric acid corrosion for the base material ifn the event that the cladding is compromised and to perform periodic visual inspections of the cladded surfaces within the steam generatorSG to detect anomalous conditions (e.g., rust stains).

Based on the discussion above, general visual inspections of steam generator (SG) head interior surfaces (including the divider plates and tubesheets) are necessary as part of the steam generator program. These inspections are intended to identify signs that cracking or loss of material may be occurring (e.g., through identification of rust stains or other abnormal conditions such as distortion of divider plate assembly).

As further details are described in LR-ISG-2016-01, the staff finds that, if the industry analyses (EPRI 3002002850) are bounding and applicable to the applicants steam generators, use of the One-Time Inspection program may not be necessary to manage cracking of PWSCC for divider plate assemblies and tube-to-tubesheet welds in accordance with the revised GALL-SLR Report, AMP XI.M19 and SRP-LR further evaluationFE sections (along with AMP XI.M2, Water Chemistry). However, if the industry analyses are not bounded and applicable to the applicants steam generatorsSGs, use of the One-Time Inspection program may be necessary.

The GALL-SLR, AMP XI.M19 and XI.M2 are also used to manage loss of material due to boric acid corrosion for steam generator heads and tubesheets. These bases and changes are consistently applied to the corresponding GALL-SLR Report and SRP-SLR guidance.

In addition, the inspection frequency of the general visual inspections added to AMP XI.M19 is also consistent with the maximum inspection interval allowed by the steam generatorSG tube 4-10 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change inspection requirements in Technical Specifications.

Initial NUREG- As revised by Revision 1 to The staff deleted footnote no. 3 because it has 2221, the SLR AMP XI.M32, concluded that the more rigorous examination Table 2-29 footnote no. 3, which stated, techniques cited in AMP XI.M32 should be AMP XI.M32 Visual inspections conducted when periodic inspections will not be One-Time conducted to detect conducted during the subsequent period of Inspection potential loss of material or extended operation. These techniques can detect cracking of SS and minor indications of loss of material and cracking.

Parameters aluminum alloy support If the Oone- Ttime Iinspections and plant- specific Monitored or members; welds; bolted operation experienceOE do not reveal loss of Inspected (Third connections; support material or cracking, periodic inspections will not Entry) anchorage to building be conducted during the subsequent period of structure exposed to air or extended operation. As a result, it is important to condensation (see SRP- demonstrate that the environment conditions will SLR Section 3.5.2.2.2.4) not promote loss of material or cracking by more may be conducted rigorous examination techniques.

consistent with those for the GALL-SLR Report AMP For a support, minor loss of material or cracking XI.S6, Structures that might not be detectable during a one- time Monitoring. was deleted. walkdown inspection will likely not impact the intended function of the support; however, the staff has concluded that growth of loss of material or cracking will become more evident during periodic inspections of supports.

Initial NUREG- Table 3.0-1 in NUREG- The staff updated the FSAR Supplements in 2221, Table 3-1 1800 Revision 2, which Table 3.0-1 for consistency with the AMP updates provided examples of Final in Chapter XI of the GALL-SLR Report, or with the Table 3.0-1, as Safety Analysis Report FSAR Ssupplement summary descriptions for previously given (FSAR) Ssupplements for these types of AMPs provided in past industry-in Chapter 3.0, of aging management submitted license renewal applications (LRAs).

NUREG-1800, programsAMPs, was Revision 2 updated, relocated to the The corresponding table (NUREG-2191, Table XI-GALL-SLR Report Section 01, FSAR Supplement Summaries for GALL-SLR XI, and renumbered as Report Chapter XI Aging Management Table XI 01. Programs,) was not included in the SRP-SLR because the information was not considered to be limited to guidance to the NRC staff reviewers, but related to broader considerations by applicants during their license renewal application development.

Initial NUREG- These AMR further For the SRP-SLR Section 3.1.2.2.11, Subsection 2221, Table 3-3 evaluationFE subsections 1 guidelines that apply to PWR SG divider plate provide the staffs assemblies, the staff added additional paragraph Section acceptance criteria and guidance and criteria that clarified when a 3.1.2.2.11, review procedures for prospective SLR applicantSLRA for a PWR-Subsections 1 managing primary water design with recirculating SGs would need to and 2 stress corrosion cracking propose use of the One- Time Inspection AMP, in (PWSCC) in pressurized addition to the Steam Generators and Water water reactor (PWR) steam Chemistry AMPs, for managing cracking due to generator (SG) divider plate PWSCC in their SG divider plate assemblies.

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Draft Document: Tracked Changes Version Revised Summary of Revised Technical Bases for Location of Change Significant Changes Change assemblies and tube-to-tubesheet welds made from For the SRP- SLR Section 3.1.2.2.11, Subsection nickel alloy materials. Some 1 guidelines that apply to PWR SG divider plate changes were made to the assemblies, the staff added additional guidance previous versions of these on the plant- specific parameters that should be further evaluationFE compared to industry analyses that show the guidelines for these analyses are applicable and bounding for a components in Section given plant. Additionally, the reference to a plant-3.1.2.2.11, Subsections 1 specific AMP was replaced with a reference to and 2 of NUREG-1800, the One- Time Inspection AMP for applicants that Revision 2. would need to manage cracking due to PWSCC in their SG divider plate assemblies.

The AMR further evaluationFE subsections For the SRP-SLR Section 3.1.2.2.11, Subsection were revised, to clarify the 1 guidelines that apply to PWR SG tube to-plant- specific parameters to tubesheet welds, the staff made some minor be evaluated against technical adjustments of the previous guidelines industry analyses to for managing PWSCC in these weld in Section determine whetherif a given 3.1.2.2.11, Subsection 2, of the NUREG-1800, plant is bounded by industry Revision 2 report. However, these changes do analyses for SG divider not alter the general approach for managing plate assembly cracking. PWSCC in PWR tube-to-tubesheet welds.

Additionally, reference to a plant- specific AMP for plants that are not bounded is replaced with a reference to the One- Time Inspection AMP.

1 2

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14 Fontana, M.G. Corrosion Engineering. McGraw Hill. pp. 86-90. 1986.

15 FPL. Mano K. Nazar. FPL Letter (January 30) to NRC, L-2018-004, Turkey Point Units 3 and 4 16 Subsequent License Renewal Application, ADAMS Accession No. 18037A812. Juno Beach, 17 FL: Florida Power & Light Company. 2018.

18 ISO. ISO 15589-1, Petroleum and Natural Gas Industries-Cathodic Protection of Pipeline 19 Transportation Systems-Part 1: On Land Pipelines, Vernier, Geneva, Switzerland: International 20 Organization for Standardization, November 2003.

21 Lee, S., P.T. Kuo, K. Wichman, and O. Chopra. Flaw Evaluation of Thermally Aged Cast 22 Stainless Steel in Light-Water Reactor Applications. International Journal of Pressure Vessels 23 and Piping. pp. 37-44. 1997.

24 Licensee Event Report 237/2007-003, Dresden Unit 2, High Pressure Coolant Injection System 25 Declared Inoperable. ADAMS Accession No. ML072750663.

26 https://lersearch.inl.gov/LERSearchCriteria.aspx. September 2007.

27 Licensee Event Report 254/2009-004, Quad Cities Unit 1, Pinhole Leak in Core Spray Piping 28 Results in Loss of Containment Integrity and Plant Shutdown for Repairs. ADAMS Accession 29 No. ML093170206. https://lersearch.inl.gov/LERSearchCriteria.aspx. November 2009.

30 Licensee Event Report 277/2006-003, Peach Bottom Unit 2, Elbow Leak on Piping Attached to 31 Suppression Pool Results in Loss of Containment Integrity. ADAMS Accession No.

32 ML063420059. https://lersearch.inl.gov/LERSearchCriteria.aspx. December 2006 33 Licensee Event Report 286/2018-003, Indian Point Unit 3, Manual Reactor Trip Due to a Steam 34 Leak on a High Pressure Feedwater Heater. ADAMS Accession No. ML18341A122.

35 https://lersearch.inl.gov/LERSearchCriteria.aspx. November 2018.

36 Licensee Event Report 346/2015-002, Davis-Besse, Improper Flow Accelerated Corrosion 37 Model Results in 4-Inch Steam Line Failure and Manual Reactor Trip. ADAMS Accession No.

38 ML15194A013. https://lersearch.inl.gov/LERSearchCriteria.aspx. July 2015.

5-7 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 Licensee Event Report 374/2013-001, LaSalle Unit 2, Pin Hole Leaks Identified in High 2 Pressure Core Spray Piping. ADAMS Accession No. ML13168A576.

3 https://lersearch.inl.gov/LERSearchCriteria.aspx. June 2013.

4 Licensee Event Report 374/2015-001, LaSalle Unit 2, High Pressure Core Spray Inoperable 5 Due to Division 3 Diesel Generator Cooling Water Pump Casing Leak. ADAMS Accession 6 No. ML15058A462. https://lersearch.inl.gov/LERSearchCriteria.aspx. February 2015.

7 Licensee Event Report 483/1999-003, Callaway, Manual Reactor Trip due to Heater Drain 8 System Pipe Rupture Caused by Flow Accelerated Corrosion. ADAMS Accession 9 No. ML003712775. https://lersearch.inl.gov/LERSearchCriteria.aspx. May 2000.

10 Licensee Event Report 499/2005-004, South Texas Project Unit 2, Inoperability of Essential 11 Cooling Water 2A and 2B Trains. ADAMS Accession No. ML053410155.

12 https://lersearch.inl.gov/LERSearchCriteria.aspx. November 2005.

13 Licensee Event Report 369/2014-002, Degraded Condition due to Rejectable Flaws on 1B and 14 1C Safety Injection Lines. https://lersearch.inl.gov/LERSearchCriteria.aspx. September 27, 15 2014.

16 Metals Handbook Desk Edition, Second Edition, Corrosion Characteristics of Carbon and Alloy 17 Steels, 1998 (Non-Public).

18 NACE. D.O. Sprowls and R.H. Brown, Stress Corrosion Mechanisms for Aluminum Alloys, 19 Fundamental Aspects of Stress Corrosion Cracking, NACE, 1969, p 466-506.

20 NRL. B.F. Brown, Stress-Corrosion Cracking in High Strength Steels and in Titanium and 21 Aluminum Alloys, Naval Research Laboratory, ARPA No. 878, 1972.

22 NBS. B.F. Brown, Stress Corrosion Cracking Control Measures, National Bureau of Standards, 23 NBS Monogr 156, June 1977.

24 NEI. NEI 97-06, Steam Generator Program Guidelines. Revision 3. Washington, D.C.: Nuclear 25 Energy Institute. January 2011.

26 NFPA. NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor 27 Electric Generating Plants, 2001 Edition. Quincy, Massachusetts: National Fire Protection 28 Association. 2001.

29 NRC. NUREG-1800, (Ch. 3.3 - End) Standard Review Plan for Review of License Renewal 30 Applications for Nuclear Power Plants. ADAMS Accession No. ML012070409.

31 Washington, DC: U.S. Nuclear Regulatory Commission. July 2001.

32 NRC. NUREG-1801, Generic Aging Lessons Learned (GALL) Report. Revision 2. ADAMS 33 Accession No. ML103490041. Washington, D.C.: U.S. Nuclear Regulatory Commission.

34 December 2010.

35 NRC. NUREG-1950, Disposition of Public Comments and Technical Bases for Changes in the 36 License Renewal Guidance Documents NUREG-1801 and NUREG-1800. Washington, DC:

37 ADAMS Accession No. ML11116A062. U.S. Nuclear Regulatory Commission. April 2011.

5-8 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 NRC. NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal 2 (GALL-SLR) Report. Volume 1. ADAMS Accession No. ML17187A031. Washington, D.C.: U.S.

3 Nuclear Regulatory Commission. July 2017.

4 NRC. NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal 5 (GALL-SLR) Report. Volume 2. ADAMS Accession No. ML17187A204. Washington, D.C.: U.S.

6 Nuclear Regulatory Commission. July 2017.

7 NRC. NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal 8 Applications for Nuclear Power Plants (SRP-SLR). ADAMS Accession No. ML17188A158.

9 Washington, DC: U.S. Nuclear Regulatory Commission. July 2017.

10 NRC. NUREG-2221, Technical Bases for Changes in the Subsequent License Renewal 11 Guidance Documents NUREG-2191 and NUREG-2192. ADAMS Accession No.

12 ML17362A126. Washington, DC: U.S. Nuclear Regulatory Commission. December 2017.

13 NRC. NUREG-2222, Disposition of Public Comments on the Draft Subsequent License 14 Renewal Guidance Documents NUREG-2191 and NUREG-2192. ADAMS Accession No.

15 ML17362A143. Washington, DC: U.S. Nuclear Regulatory Commission. December 2017.

16 NRC. Staff Requirements Memoranda, SRM-SECY-21-0029, Rulemaking Plan on Relaxation 17 of Inservice Testing and Inservice Inspection Program Update Frequencies Required in 10 CFR 18 50.55a, ADAMS Accession No. ML21312A490. Washington, DC: U.S. Nuclear Regulatory 19 Commission. November 2011.

20 NRC. Bulletin 88-02, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes.

21 Washington, DC: U.S. Nuclear Regulatory Commission. February 1988.

22 NRC. Information Notice 94-05, Potential Failure of Steam Generator Tubes Sleeved with 23 Kinetically Welded Sleeves. Washington, DC: U.S. Nuclear Regulatory Commission. January 24 1994.

25 NRC. NUREG-1430, Standard Technical Specifications - for Babcock and Wilcox Pressurized 26 Water Reactors Plants. Volume 1, Revision 54. ADAMS Accession No. Volume 1, 27 ML21272A363; Volume 2, ML21272A370. Washington DC: U.S. Nuclear Regulatory 28 Commission. September 2021.

29 NRC. NUREG-1431, Standard Technical Specifications - for Westinghouse Pressurized Water 30 Reactors Plants. Volume 1, Revision 54. ADAMS Accession No. Volume 1, ML21259A155; 31 Volume 2, ML21259A159. Washington DC: U.S. Nuclear Regulatory Commission. September 32 2021.

33 NRC. NUREG-1432, Standard Technical Specifications - for Combustion Engineering 34 Pressurized Water Reactors Plants. Volume 1, Revision 54. ADAMS Accession No. Volume 1, 35 ML21258A421; Volume 2, ML21258A424. Washington DC: U.S. Nuclear Regulatory 36 Commission. September 2021.

37 NRC. NUREG/CR-6001, Aging Assessment of BWR Standby Liquid Control Systems, dated 38 August 17, 1992. ADAMS Accession No. ML040340671.

5-9 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 NRC. GALL-SLR and SRP-SLR Supplemental Staff Guidance. ADAMS Accession No.

2 ML16041A090. Washington, DC: U.S. Nuclear Regulatory Commission. March 2016.

3 NRC. TSTF-577, Revised Frequencies for Steam Generator Tube Inspections. Revision 1.

4 ADAMS Accession No. ML21060B434. Rockville MD: Technical Specifications Task Force.

5 March 2021.

6 NRC. Information Notice 2007-21, Pipe Wear Due to Interaction of Flow-Induced Vibrations, 7 Supplement 1, ADAMS Accession No. ML20225A204. Washington, DC: U.S. Nuclear 8 Regulatory Commission. December 11, 2020.

9 NRC. Information Notice 2002-21, Pipe Wear Due to Interaction of Flow Induces Vibrations, 10 ADAMS No. ML071150051. Washington, DC: U.S. Nuclear Regulatory Commission. June 11, 11 2007.

12 NRC. Supplement 1 to PBN SLRA, ADAMS Accession No. ML21111A155. April 21, 2021.

13 NRC. Information Notice 2007-21, Pipe Wear Due to Interaction of Flow-Induced Vibrations, 14 Supplement 1, ADAMS Accession No. ML20225A204. Washington, DC: U.S. Nuclear 15 Regulatory Commission. December 11, 2020 16 NRC. NUREG/CR-4513, Estimation of Fracture Toughness of Cast Stainless Steels During 17 Thermal Aging in LWR Systems. Revision 2. ADAMS Accession No. ML16145A082.

18 Washington, DC: U.S. Nuclear Regulatory Commission. May 2016 19 NRC. NUREG/CR-4513, Estimation of Fracture Toughness of Cast Stainless Steels During 20 Thermal Aging in LWR Systems. Revision 2 with errata. ADAMS Accession No. ML16145A082.

21 Washington, DC: U.S. Nuclear Regulatory Commission. March 2021 22 NRC. NUREG/CR-4513, Estimation of Fracture Toughness of Cast Stainless Steels During 23 Thermal Aging in LWR Systems. Revision 1. ADAMS Accession No ML052360554.

24 Washington, DC: U.S. Nuclear Regulatory Commission. August 1994 25 NRC. Generic Letter 89-13, Service Water Systems Problems Affecting Safety Related 26 Equipment.

27 NRC. Generic Letter 89-13, Service Water Systems Problems Affecting Safety Related 28 Equipment, Supplement 1.

29 NRC. Final Safety Evaluation of the BWRVIP-234: Thermal Aging and Neutron Embrittlement 30 Evaluation of Cast Austenitic Stainless Steel for BWR Internals. ADAMS Accession No.

31 ML16096A002. Washington, DC: U.S. Nuclear Regulatory Commission. June 22, 2016.

32 NRC. Surry Power Station, Units 1 and 2 - Final Safety Evaluation Report for the Subsequent 33 License Renewal Application Review, ADAMS Accession No. ML20052F520. Washington, DC:

34 U.S. Nuclear Regulatory Commission. March 9, 2020.

35 NRC. Safety Evaluation Report Related to the Subsequent License Renewal of Turkey Point 36 Generating Units 3 and 4, ADAMS Accession No. ML19191A057. Washington, DC: U.S.

37 Nuclear Regulatory Commission. July 2019.

5-10 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 NRC. Generic Letter 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling 2 Water Reactors. Washington, DC: U.S. Nuclear Regulatory Commission. July 1994.

3 NRC. IE Bulletin 80-07, BWR Jet Pump Assembly Failure. Washington, DC: U.S. Nuclear 4 Regulatory Commission. April 1980.

5 NRC. IE Bulletin 80-07, Supplement 1, BWR Jet Pump Assembly Failure. Washington, DC:

6 U.S. Nuclear Regulatory Commission. May 1980.

7 NRC. IE Bulletin 80-13, Cracking in Core Spray Spargers. Washington, DC: U.S. Nuclear 8 Regulatory Commission. May 1980.

9 NRC. Information Notice 88-03, Cracks in Shroud Support Access Hole Cover Welds.

10 Washington, DC: U.S. Nuclear Regulatory Commission. February 1988.

11 NRC. Information Notice 92-57, Radial Cracking of Shroud Support Access Hole Cover Welds.

12 Washington, DC: U.S. Nuclear Regulatory Commission. August 1992.

13 NRC. Information Notice 93-101, Jet Pump Hold-Down Beam Failure. Washington, DC: U.S.

14 Nuclear Regulatory Commission. December 1993.

15 NRC. Information Notice 94-42, Cracking in the Lower Region of the Core Shroud in Boiling 16 Water Reactors. Washington, DC: U.S. Nuclear Regulatory Commission. June 1994.

17 NRC. Information Notice 95-17, Reactor Vessel Top Guide and Core Plate Cracking.

18 Washington, DC: U.S. Nuclear Regulatory Commission. March 1995.

19 NRC. Information Notice 97-02, Cracks Found in Jet Pump Riser Assembly Elbows at Boiling 20 Water Reactors. Washington, DC: U.S. Nuclear Regulatory Commission. February 1997.

21 NRC. Information Notice 97-17, Cracking of Vertical Welds in the Core Shroud and Degraded 22 Repair. Washington, DC: U.S. Nuclear Regulatory Commission. April 1997.

23 NRC. Information Notice 2007-02, Failure of Control Rod Drive Mechanism Lead Screw Male 24 Coupling at Babcock and Wilcox-Designed Facility. Washington, DC: U.S. Nuclear Regulatory 25 Commission. March 2007.

26 NRC. Letter from Christopher I. Grimes, U.S. Nuclear Regulatory Commission, License 27 Renewal and Standardization Branch, to Douglas J. Walters, Nuclear Energy Institute, License 28 Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Stainless Steel 29 Components. ADAMS Accession No. ML003717179. May 19, 2000.

30 NRC. NUREG-1544, Status Report: Intergranular Stress Corrosion Cracking of BWR Core 31 Shrouds and Other Internal Components. Washington, DC: U.S. Nuclear Regulatory 32 Commission. March 1996.

33 NRC. NUREG-0313, Technical Report on Material Selection and Process Guidelines for BWR 34 Coolant Pressure Boundary, Revision 2, January 1988. (ADAMS Accession No.

35 ML031470422).

5-11 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 NRC. NUREG/CR-4513, Estimation of NRC. NRC. Fracture Toughness of Cast Stainless 2 Steels during Thermal Aging in LWR Systems. Revision 1. Washington, DC: U.S. Nuclear 3 Regulatory Commission. August 1994.

4 NRC. NUREG/CR-6923, Expert Panel Report on Proactive Materials Degradation 5 Assessment. Washington, DC: U.S. Nuclear Regulatory Commission. March 2007.

6 NRC. Memorandum from Joseph J. Holonich, U.S. Nuclear Regulatory Commission, Licensing 7 Processes Branch, to Dennis C. Morey, U.S. Nuclear Regulatory Commission, Licensing 8 Processes Branch, Summary of the May 27, 2021, Meeting between the U.S. Nuclear 9 Regulatory Commission Staff and the Electric Power Research Institute to Discuss 10 Nonconservatism in BWRVIP-100, Revision 1-A. ADAMS Accession No. ML21153A003. June 11 8, 2021.

12 NRC. Regulatory Guide 1.65, Revision 0, Materials and Inspections for Reactor Vessel Closure 13 Studs, October 1973.

14 NRC. Regulatory Guide 1.65, Revision 1, Materials and Inspections for Reactor Vessel Closure 15 Studs, April 2010.

16 NRC. NRC Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant 17 Systems. ADAMS Accession No. ML031220144. Washington, DC: U.S. Nuclear Regulatory 18 Commission. June 22, 1988.

19 NRC. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for 20 Nuclear Power Plants, Revision 1, Section 3.6.3, Leak-before-break Evaluation Procedures.

21 ADAMS Accession No. ML063600396. Washington, DC: U.S. Nuclear Regulatory Commission.

22 March 2007.

23 NRC. Safety Evaluation Report Related to the Subsequent License Renewal of Turkey Point 24 Generating Units 3 and 4, dated December 2019 (ADAMS Accession No. ML19191A057).

25 NRC. 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear 26 Power Plants. Washington, DC: U.S. Nuclear Regulatory Commission. 2021.

27 NRC. Information Notice 2018-10, Thermal Sleeve Flange Wear Leads to Stuck Control Rod at 28 Foreign Nuclear Plant. ADAMS Accession No. ML18214A710. Washington, DC: U.S. Nuclear 29 Regulatory Commission. August 29, 2018.

30 31 NRC. 10 CFR Part 50, Appendix R, Fire Protection Program for Nuclear Power Facilities 32 Operating Prior to January 1, 1979. Washington, DC: U.S. Nuclear Regulatory Commission.

33 2021.

34 NRC. 10 CFR 50.48, Fire protection. Washington, DC: U.S. Nuclear Regulatory Commission.

35 2021.

36 NRC. Generic Letter 92-08, Thermo-Lag 330-1 Fire Barrier. ADAMS Accession No.

37 ML031130425. Washington, DC: U.S. Nuclear Regulatory Commission. December 17, 1992.

5-12 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 NRC. Information Notice 88-56, Potential Problems with Silicone Foam Fire Barrier Penetration 2 Seals. ADAMS Accession No. ML031150042. Washington, DC: U.S. Nuclear Regulatory 3 Commission. August 4, 1988.

4 NRC. Information Notice 91-47, Failure of Thermo-Lag Fire Barrier Material to Pass Fire 5 Endurance Test. ADAMS Accession No. ML031190452. Washington, DC: U.S. Nuclear 6 Regulatory Commission. August 6, 1991.

7 NRC. Information Notice 94-28, Potential Problems with Fire-Barrier Penetration Seals.

8 ML031060475. Washington, DC: U.S. Nuclear Regulatory Commission. April 5, 1994.

9 NRC. Information Notice 97-70, Potential Problems with Fire Barrier Penetration Seals.

10 ADAMS Accession No. ML031050108. Washington, DC: U.S. Nuclear Regulatory Commission.

11 September 19, 1997.

12 NRC. Regulatory Guide 1.189, Fire Protection for Nuclear Power Plants. Revision 4. ADAMS 13 Accession No. ML21048A441. Washington, DC: U.S. Nuclear Regulatory Commission. May 14 2021.

15 NRC. Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing 16 Light-Water Nuclear Power Plants. Revision 2. ADAMS Accession No. ML21048A448.

17 Washington, DC: U.S. Nuclear Regulatory Commission. May 2021.

18 NRC. Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section 19 Xi, Division 1, Revision 18. ADAMS Accession No. ML16321A336. Washington, DC: U.S.

20 Nuclear Regulatory Commission. March 2017.

21 NRC. Information Notice 89-52, Potential Fire Damper Operational Problems. ADAMS 22 Accession No. ML031180663. Washington, DC: U.S. Nuclear Regulatory Commission. June 23 1989.

24 NRC. Bulletin 87-01, Thinning of Pipe Walls in Nuclear Power Plants. ADAMS Accession 25 No. ML031210862. Washington, DC: U.S. Nuclear Regulatory Commission. July 1987.

26 NRC. Generic Letter 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning. ADAMS 27 Accession No. ML031200731. Washington, DC: U.S. Nuclear Regulatory Commission.

28 May 1989.

29 NRC. Information Notice 89-53, Rupture of Extraction Steam Line on High Pressure Turbine.

30 ADAMS Accession No. ML031180660. Washington, DC: U.S. Nuclear Regulatory Commission.

31 June 1989.

32 NRC. Information Notice 91-18, High-Energy Piping Failures Caused by Wall Thinning.

33 ADAMS Accession No. ML031190529. Washington, DC: U.S. Nuclear Regulatory Commission.

34 March 1991.

35 NRC. Information Notice 91-18, High-Energy Piping Failures Caused by Wall Thinning.

36 Supplement 1. ADAMS Accession No. ML082840749. Washington, DC: U.S. Nuclear 37 Regulatory Commission. December 1991.

5-13 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 NRC. Information Notice 92-35, Higher than Predicted Erosion/Corrosion in Unisolable Reactor 2 Coolant Pressure Boundary Piping inside Containment at a Boiling Water Reactor. ADAMS 3 Accession No. ML031200365. Washington, DC: U.S. Nuclear Regulatory Commission.

4 May 1992.

5 NRC. Information Notice 93-21, Summary of NRC Staff Observations Compiled During 6 Engineering Audits or Inspections of Licensee Erosion/Corrosion Programs. ADAMS Accession 7 No. ML031080042. Washington, DC: U.S. Nuclear Regulatory Commission. March 1993.

8 NRC. Information Notice 95-11, Failure of Condensate Piping Because of Erosion/Corrosion at 9 a Flow Straightening Device. ADAMS Accession No. ML031060332. Washington, DC:

10 U.S. Nuclear Regulatory Commission. February 1995.

11 NRC. Information Notice 97-84, Rupture in Extraction Steam Piping as a Result of Flow-12 Accelerated Corrosion. ADAMS Accession No. ML031050037. Washington, DC: U.S. Nuclear 13 Regulatory Commission. December 1997.

14 NRC. Information Notice 99-19, Rupture of the Shell Side of a Feedwater Heater at the Point 15 Beach Nuclear Plant. ADAMS Accession No. ML031040409. Washington, DC: U.S. Nuclear 16 Regulatory Commission. June 1999.

17 NRC. Information Notice 2006-08, Secondary Piping Rupture at the Mihama Power Station in 18 Japan. ADAMS Accession No. ML052910008. Washington, DC: U.S. Nuclear Regulatory 19 Commission. March 2006.

20 NRC. Information Notice 2019-08, Flow-Accelerated Corrosion Events. ADAMS Accession No.

21 ML19065A123. Washington, DC: U.S. Nuclear Regulatory Commission. October 2019.

22 NRC. License Renewal Interim Staff Guidance LR-ISG-2012-01, Wall Thinning Due to Erosion 23 Mechanisms. ADAMS Accession No. ML12352A057. Washington, DC: U.S. Nuclear 24 Regulatory Commission. April 2013.

25 NRC. License Renewal Interim Staff Guidance LR-ISG-2021-01, Updated Aging Management 26 Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors of Subsequent 27 License Renewal Guidance. ADAMS Accession No. ML20217L203. Washington, DC: U.S.

28 Nuclear Regulatory Commission. January 2021.

29 NRC. NUREG-1344, Erosion/Corrosion-Induced Pipe Wall Thinning in U.S. Nuclear Power 30 Plants. ADAMS Accession No. ML20247A046. Washington, DC: U.S. Nuclear Regulatory 31 Commission. April 1989.

32 NRC. NUREG/CR-6031, Cavitation Guide for Control Valves. ADAMS Accession No.

33 ML17187A204. Washington DC: U.S. Nuclear Regulatory Commission. April 1993.

34 NSAC. NSAC-202L-R2, Recommendations for an Effective Flow-Accelerated Corrosion 35 Program. Palo Alto, California: Electric Power Research Institute, Nuclear Safety Analysis 36 Center (NSAC). April 1999.

37 NSAC. NSAC-202L-R3, Recommendations for an Effective Flow-Accelerated Corrosion 38 Program (1011838). Palo Alto, California: Electric Power Research Institute, Nuclear Safety 39 Analysis Center (NSAC). May 2006.

5-14 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version 1 NSAC. NSAC-202L-R4, Recommendations for an Effective Flow-Accelerated Corrosion 2 Program (3002000563). Palo Alto, California: Electric Power Research Institute, Nuclear Safety 3 Analysis Center (NSAC). November 2013.

4 OECD-NEA. McDevitt, M., Childress T., Hoehn M and McGill R. Analysis and Impact of Recent 5 Thermal Fatigue Operating Experience in the USA. Fourth International Conference on Fatigue 6 of Nuclear Reactor Components. Sevilla, Spain. Organisation for Economic Co-operation and 7 Development (OECD)/ Nuclear energy Agency (NEA). 2015. Document No.

8 NEA/CSNI/R(2017)2/ADD1.

9 Oregon State University. Morrell, J. Jeffrey, Estimated Service Life of Wood Poles, Technical 10 Bulletin No. 16-U-101, North American Wood Pole Council, Oregon State University, February 11 2016.

12 PWROG. Presentation, Industry Plans to Address Thermal Sleeve Operating Experience.

13 ADAMS Accession No. ML18254A400. Cranberry Township, Pennsylvania: PWR Owners 14 Group. September 12, 2018.

15 USDA. U.S. Department of Agriculture, Wood Pole Inspection and Maintenance, Rural Utility 16 Service (RUS) Bulletin 1730B-121, August 13, 2013.

17 USDA. U.S. Department of Agriculture Report, Durability of Preservative-Treated Wood Utility 18 Poles in Guam, May 1986.

19 Westinghouse. James A. Gresham. Westinghouse Letter (July 17) to NRC, LTR-NRC-18-53, 20 NSAL-18-1 Revision 0, Thermal Sleeve Flange Wear Leads to Stuck Control Rod. ADAMS 21 Accession No. ML18198A275. Pittsburgh, Pennsylvania: Westinghouse Electric Company.

22 2018.

23 Westinghouse. PWR Owners Group Report Nos. PWROG-17033-NP-A and PWROG-17033-P-24 A (non-proprietary and proprietary versions), Revision 1, Update for Subsequent License 25 Renewal: WCAP-13045, Compliance to ASME Code Case N-481 of the Primary Loop Pump 26 Casings of Westinghouse Type Nuclear Steam Supply Systems. ADAMS Accession Nos.

27 ML19319A188 and ML19319A195 (non-proprietary and proprietary versions). Pittsburgh, 28 Pennsylvania: Westinghouse Electric Company. November 2019.

29 5-15 Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (12-2010) (Assigned by NRC, Add Vol., Supp., Rev.,

NRCMD 3.7 and Addendum Numbers, if any.)

BIBLIOGRAPHIC DATA SHEET NUREG-2221, (See instructions on the reverse) Supplement 1 Draft

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED Technical Bases for Changes in the Subsequent License Renewal Guidance Documents, MONTH YEAR NUREG-2191, Revision 1, Draft Report for Comment and NUREG-2192, Revision 1, Draft July 2023 Report for Comment

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT Technical

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U. S. Nuclear Regulatory Commission, and mailing address; if contractor, provide name and mailing address.)

Division of New and Renewed Licenses Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above", if contractor, provide NRC Division, Office or Region, U. S. Nuclear Regulatory Commission, and mailing address.)

Same as above

10. SUPPLEMENTARY NOTES When finalized, this report will supplement NUREG-2221 (Rev. 0)

11. ABSTRACT (200 words or less)

This document is a knowledge management and knowledge transfer document associated with Draft NUREG-2191, Revision 1, Generic Aging Lessons Learned for Subsequent License Renewal Draft Report for Comment, (GALL-SLR Report, Rev. 1,), and Draft NUREG-2192, Revision 1, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, Draft Report for Comment (SRP-SLR Rev. 1).

The initial iteration of NUREG-2221, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17362A126) documented the technical changes and bases that were made from the guidance contained in NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report, (ML103490041), for utilities applying for first license renewal, to the updated guidance for utilities wishing to apply for subsequent license renewal (i.e., for operation from 60 to 80 years), published as NUREG-2191, Revision 0 (ML17187A031 and ML17187A204, for Volumes 1 and 2 respectively).

This publication is a draft supplement to the initial NUREG-2221, and it documents the technical changes that were made in concurrent updates to the subsequent license renewal guidance documents in 2023. This document provides the underlying rationale that the NRC staff used to develop Draft NUREG-2191, Revision 1, and Draft NUREG-2192, Revision 1.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT License Renewal Further Evaluations unlimited Long-term Operations Technical Bases 14. SECURITY CLASSIFICATION Aging (This Page)

Nuclear Safety unclassified Aging Mechanisms (This Report)

Aging Effects unclassified Aging Management Programs 15. NUMBER OF PAGES Subsequent License Renewal Second License Renewal 16. PRICE NRC FORM 335 (12-2010)

Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Page Numbers May Not Align with Draft for Comment Version

Draft Document: Tracked Changes Version Page Numbers May Not Align with Draft for Comment Version

Page Numbers May Not Align with Draft for Comment Version Draft Document: Tracked Changes Version NUREG-2221 Technical Bases for Changes in the Subsequent License Renewal July 2023 Supplement 1, Draft Guidance Documents, NUREG-2191, Revision 1, Draft Report for Comment and NUREG-2192, Revision 1, Draft Report for Comment