ML20156A343

SLR-ISG-PWRVI-2020-XX Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors Draft Interim Staff Guidance

ML20156A343
Person / Time
Issue date:	07/17/2020
From:	William Burton NRC/NRR/DNRL
To:
Burton W
References
Download: ML20156A343 (92)

Text

SLR-ISG-PWRVI-2020-XX Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors Draft Interim Staff Guidance July 2020

ML20156A343 CAC: TM3021 OFFICE Author:DNLR:NVIB BC:DNLR:NVIB PM:DNLR:NLRP PM:DNRL:NLRP NAME JMedoff HGonzalez WBurton DDrucker DATE 6/8/2020 6/9/2020 6/9/2020 6/10/2020 OFFICE BC:DNLR:NLRP QTE OGC LA:DRO:IRSB NAME LGibson JDougherty STurk BCurran w/comments DATE 6/15/2020 6/19/2020 7/10/2020 7/14/2020 OFFICE PM:DRO:IRSB D:NRR:DRO D:DNRL NAME TGovan CMiller ABradford DATE 7/17/2020 7/17/2020 7/17/2020 DRAFT INTERIM STAFF GUIDANCE UPDATED AGING MANAGEMENT CRITERIA FOR REACTOR VESSEL INTERNAL COMPONENTS FOR PRESSURIZED-WATER REACTORS SUBSEQUENT LICENSE RENEWAL GUIDANCE SLR-ISG-PWRVI-2020-XX PURPOSE The U.S. Nuclear Regulatory Commission (NRC) staff is issuing this draft subsequent license renewal (SLR) interim staff guidance (ISG) to provide clarifying guidance to facilitate staff and industry understanding of the aging management of systems, structures, and components required by Title 10 of the Code of Federal Regulations (10 CFR) Part 54, Requirements for renewal of operating licenses for nuclear power plants (Ref. 1).

This draft SLR-ISG identifies proposed revisions to the guidance for pressurized-water reactor (PWR) vessel internal components in NUREG 2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), issued July 2017 (Ref. 2), and in NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL SLR) Report, issued July 2017 (Ref. 3).

The guidance in this SLR-ISG supersedes in total the previous guidance in License Renewal (LR)-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors, dated June 3, 2013 (Ref. 4), which is related to NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report, issued December 2010 (Ref. 5), and NUREG 1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants (SRP-LR), issued December 2010 (Ref. 6).

BACKGROUND The NRC staff has reviewed three applications to extend plant operations to 80 years (i.e., for SLR) for Turkey Point Nuclear Generating Units 3 and 4 (Turkey Point); Peach Bottom Atomic Power Station, Units 2 and 3 (Peach Bottom); and Surry Power Station, Units 1 and 2 (Surry).

During these reviews, both the staff and applicants have identified ways to make the preparation and review of future subsequent license renewal applications (SLRAs) more effective and efficient.

RATIONALE Public meetings took place on March 28, 2019; December 12, 2019; February 20, 2020; March 25, 2020; April 3, 2020; and April 7, 2020, between the staff and industry representatives to discuss staff and industry experience in the preparation and review of the initial license renewal application (LRA) for River Bend Station, Unit 1, which piloted the optimized 18-month review process for SLRAs, as well as the reviews of the first three SLRAs for Turkey Point, Peach Bottom, and Surry.

The guidance document changes issued in this SLR-ISG are based on the updated inspection and evaluation (I&E) guidelines in Electric Power Research Institute (EPRI) Materials Reliability

SLR-ISG-PWRVI-2020-XX Page 2 of 9 Program (MRP) Topical Report No. 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), issued December 2019 (Ref. 7), which the NRC staff found acceptable for referencing in licensing applications in its safety evaluation dated April 25, 2019 (Ref. 8), and approved for use in the staffs letters to the EPRI MRP dated February 19, 2020 (Ref. 9), and July 7, 2020 (Ref. 10).

The NRC is issuing this draft SLR-ISG to accomplish the following five objectives:

(1) GALL-SLR Report and SRP-SLR Guidance Changes: Update the staffs guidance for PWR reactor vessel internal (RVI) components in the GALL-SLR Report and SRP-SLR to account for changes in I&E criteria for PWR RVI components made in MRP-227, Revision 1-A, and in other relevant industry documents (e.g., EPRI MRP expert panel reports for 80-year RVI component assessments or in relevant industry interim guidance documents or alert letters).

(2) Clarification on the Use of MRP-227, Revision 1-A: Clarify whether incorporation and adoption of MRP-227, Revision 1-A, may be used as the starting basis for the PWR Vessel Internals Aging Management Program (AMP) and whether reference to the criteria in MRP-227, Revision 1-A, in a PWR applicants SLRA will need to be subject to the performance of an RVI component-specific gap analysis.

(3) Reduction of Unnecessary Burden for PWR SLRAs: Provide additional clarifications on PWR Vessel Internals AMP programmatic change bases that are considered to be administrative and that will no longer need to be within the scope of AMP-identified exceptions or enhancements.

(4) Resolution of Applicant/Licensee Action Items (A/LAIs): Resolve whether the staffs A/LAIs in its safety evaluation for the I&E guidelines in EPRI TR No. 1022863, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), dated December 16, 2011 (Ref. 11), and A/LAI No. 1 in the staffs safety evaluation for the I&E guidelines in MRP-227, Revision 1-A, dated April 25, 2019, need to be addressed in an initial LRA or an SLRA.

(5) Closure of Regulatory Information Summary (RIS) 2011-07: Provide the staffs basis for closing previous guidance matters raised in RIS 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management, dated July 21, 2011 (Ref. 12).

CURRENT REGULATORY FRAMEWORK The NRC defines and establishes the staffs rules for submitting and receiving Commission approval of LRAs or SLRAs in 10 CFR Part 54. Pursuant to the requirements specified in 10 CFR 54.21(a)(1), a license renewal applicant is required to perform an integrated plant assessment of its facility to determine those systems, structures, or components (SSCs) that are within the scope of an aging management review (AMR). In 10 CFR 54.21(a)(1), the NRC defines SSCs subject to an AMR as those SSCs that perform an intended function in accordance with the requirements defined in 10 CFR 54.4, Scope, without moving parts or a change in configuration, and that are not subject to replacement based on a qualified life or specified time period (sometimes referred to as passive, long-lived components). For those SSCs that are within the scope of an AMR, 10 CFR 54.21(a)(3) requires the applicant to demonstrate that the effects of aging on the SSCs will be adequately managed so that the

SLR-ISG-PWRVI-2020-XX Page 3 of 9 intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation.

The requirements in 10 CFR 54.21(a)(1) and 10 CFR 54.21(a)(3) apply to subsequent periods of extended operation that may be proposed in an SLRA for a U.S light-water reactor facility.

The PWR RVI components that are within the scope of this SLR-ISG are those that are required to be the subject of an AMR pursuant to the integrated plant assessment requirements in 10 CFR 54.21(a)(1).

The guidance in this SLR-ISG provides a process that may be used to determine whether a specified PWR RVI component will need to be managed for specified aging effects in accordance with the requirements defined in 10 CFR 54.21(a)(3).

DISCUSSION AMP XI.M16A, PWR Vessel Internals, of the GALL Report, Revision 2, and the associated AMR line items in both the GALL Report, Revision 2, and SRP-LR, Revision 2, provide aging management guidance for PWR vessel internals based on the initial submitted version of MRP-227, Revision 0, dated December 2008 (Ref. 13). LR-ISG-2011-04 updated GALL Report Revision 2 AMP XI.M16A to be consistent with MRP-227-A (Ref. 14), which the NRC staff approved in a safety evaluation dated December 16, 2011 (Ref. 11). The staff also updated the AMR line items for PWR RVI components in both the GALL Report, Revision 2, and SRP-LR, Revision 2, to make them consistent with MRP-227-A.

The NRC issued the GALL-SLR Report and SRP-SLR in 2017 to address plant operation for a period up to 80 years. The AMR line items were based on those provided in LR-ISG-2011-04, as adjusted for relevant operating experience or industry recommendations that were developed after the issuance of MRP-227-A. However, these AMR line items did not represent a complete analysis for 80 years of operations.

GALL-SLR Report AMP XI.M16A and SRP-SLR Section 3.1.2.2.9 were based on MRP-227-A, which is an analysis for 60 years of plant operation. These GALL-SLR Report and SRP-SLR sections used the term MRP-227-A (as supplemented) to describe either the use of MRP-227-A as supplemented by a gap analysis to enhance the program for an 80-year operating period, or the use of acceptable generic guidance such as an approved revision of MRP-227 that considers an operating period of 80 years. For example, in SRP-SLR Section 3.1.2.2.9, the staff clarified that if a gap analysis is needed for the programmatic basis, the analysis should consider the extension of time-dependent cyclical loads and neutron irradiation exposures through the end of an 80-year cumulative licensing period to identify changes to inspections of PWR RVI components from those defined for the specified components in MRP-227-A. The staff also explained that an SLRA does not need to include a gap analysis of the RVI components if the AMP is based on a site-specific or staff-approved generic industry program whose evaluation of aging in the RVI components is based on an 80-year assessment.

The revisions in this SLR-ISG to the information for PWR RVI components in the GALL-SLR Report and SRP-SLR reflect the revised I&E guidelines in MRP-227, Revision 1-A. While Revision 1-A is an update of the guidance in MRP-227-A that reflects the operating experience since the issuance of MRP-227-A, Revision 1-A only assesses PWR RVI components through the end of a 60-year licensing term. Thus, even if an applicant revises its PWR vessel internals program (or analogous AMP for the RVI components) based on MRP-227, Revision 1-A, the

SLR-ISG-PWRVI-2020-XX Page 4 of 9 program in the SLRA will need a gap analysis to identify enhancements to the program that are necessary to address an 80-year operating period. As described in SRP-SLR Section 3.1.2.2.9 (as updated in this SLR-ISG), the SLRA should include and discuss the gap analysis methods and results. As a result of these considerations, the staff considers that it is appropriate to issue this SLR-ISG that covers updated aging management criteria and bases for PWR RVI components.

APPLICABILITY All holders of operating licenses for nuclear power reactors under 10 CFR Part 50, Domestic licensing of production and utilization facilities (Ref. 15), except those that have permanently ceased operations and have certified that fuel has been permanently removed from the reactor vessel.

GUIDANCE The NRC provides requirements for the submission and review of applications to extend plant operations beyond the initial 40-year operating period in 10 CFR Part 54.

The GALL-SLR Report and SRP-SLR provide guidance to licensees that wish to extend their plant operating licenses from 60 years to 80 years, and to the NRC staff who will review the SLRAs.

The staff and nuclear industry have identified a number of areas for which future SLRAs and staff reviews can be completed more effectively and efficiently. A series of SLR-ISGs captures these areas, known as lessons learned.

The NRC staff considers that the information in this ISG provides an acceptable approach for managing aging in PWR vessel internal components within the scope of 10 CFR Part 54 and will improve the quality, uniformity, effectiveness, and efficiency of NRC staff reviews of future SLRAs.

IMPLEMENTATION The NRC staff will use the information discussed in this draft SLR-ISG to determine whether, pursuant to 10 CFR 54.21(a)(3), an SLRA demonstrates that the effects of aging on structures and components subject to an AMR are adequately managed so their intended functions will be maintained consistent with the current licensing basis for the subsequent period of extended operation. This draft ISG contains an update in redline/strikeout of the GALL-SLR Report and SRP-SLR sections related to the aging management of pressurized-water RVIs. An applicant may reference this SLR-ISG in an SLRA to demonstrate that the AMPs at the applicants facility correspond to those described in the GALL-SLR Report. If an applicant credits an AMP as updated by this ISG, it is incumbent on the applicant to ensure that the conditions and operating experience at the plant are bounded by the conditions and operating experience for which this draft ISG was evaluated. If these bounding conditions are not met, it is incumbent on the applicant to address any additional aging effects and augment its AMPs.

For AMPs that are based on this ISG, the NRC staff will review and verify whether the applicants AMPs are consistent with those described in this ISG, including applicable plant conditions and operating experience.

SLR-ISG-PWRVI-2020-XX Page 5 of 9 ACTIONS SLR-ISG Objectives 1 and 2GALL-SLR Report and SRP-SLR Guidance Changes and Clarifications on the Use of MRP-227, Revision 1-A This SLR-ISG updates the following sections or tables in the GALL-SLR Report or SRP-SLR to ensure consistency with guidance in MRP-227, Revision 1-A:

• commodity group-based AMR line items for PWR RVI components in Table 3.1-1 of the SRP-SLR

• AMR line items for these components in Table IV.B2 of the GALL-SLR Report

• AMR line items for these components in Table IV.B3 of the GALL-SLR Report

• AMR line items for these components in Table IV.B4 of the GALL-SLR Report

• generic AMR line items applying to PWR RVI components in Section IV.E and Table IV.E of the GALL-SLR Report

• AMR further evaluation acceptance criteria for PWR RVI components in SRP-SLR Section 3.1.2.2.9 and AMR further evaluation review procedures for PWR RVI components in SRP-SLR Section 3.1.3.2.9

• the program description, program elements, and program references in GALL-SLR Report AMP XI.M16A

• the final safety analysis report (FSAR) supplement example for a PWR vessel internals program specified in Table 3.0-1 of the SRP-SLR

• material definitions in GALL-SLR Report Table IX.C to add a new definition for stellite materials, which may apply to the design of specific types of PWR RVI components

• SRP-SLR Table 4.7-1 to include MRP-based fluence and cycle analyses for PWR RVI components as potential plant-specific time-limited aging analyses (TLAAs) for PWR SLRAs The appendices included in this SLR-ISG provide the updated versions of these sections, line items, or tables.

MRP-227, Revision 1-A, is based (in part) on an assessment of Westinghouse, Combustion Engineering (CE), and Babcock & Wilcox (B&W)-designed reactor internals over a 60-year cumulative licensed service life for the reactors. Thus, PWR SLR applicants who transition their programs to use MRP-227, Revision 1-A as the starting basis for their AMPs, will need to perform and include a gap analysis for their PWR RVI components in their SLRAs.

These actions satisfy Objectives 1 and 2, as stated in the Rationale section of this SLR-ISG.

SLR-ISG-PWRVI-2020-XX Page 6 of 9 SLR-ISG Objective 3Reduction of Unnecessary Burden for PWR SLRAs The PWR Vessel Internals program discussed in this SLR-ISG is based on MRP-227, Revision 1-A. The programs in SLRAs may also include implementation of additional inspection guidance developed by the EPRI MRP, industry vendors, or owners organizations (e.g., Westinghouse, CE, B&W, or the PWR Owners Group). The NRC has updated the Scope of Program element in GALL-SLR Report AMP XI.M16A to clarify that the scope of PWR vessel internals programs may include all industry guidelines that apply to the RVI components. The Administrative Controls and Confirmation Process elements in GALL-SLR Report AMP XI.M16A identify that the program implements these guidelines in accordance with an applicants industry initiative processes in accordance with Nuclear Energy Institute 03-08, Guideline for the Management of Materials Issues, Revision 3, dated February 2017 (Ref. 16).

The staff acknowledges that, as the industry generates supplemental guidance, the plant procedures for these programs may not be up to date with the new methods recommended for the components. Activities to update and maintain the procedures are explicitly identified in the Confirmatory Processes and Administrative Controls elements of the AMP.

These clarifications satisfy Objective 3, as stated in the Rationale section of this SLR-ISG.

SLR-ISG Objective 4Resolution of A/LAIs The safety evaluation for MRP-227-A identified a number of A/LAIs to be addressed by those applicants or licensees using that topical report to satisfy the aging management requirements of 10 CFR 54.21(a)(3).

The staffs approval basis in the April 25, 2019, safety evaluation for MRP-227, Revision 1-A, was sufficient to close all A/LAIs previously issued by the staff on MRP-227-A. Therefore, responses to the A/LAIs on MRP-227-A do not need to be included in a PWR SLRA or in a PWR LRA where the PWR vessel internals program for the SLRA or LRA is based on the I&E guidelines in MRP-227, Revision 1-A.

The safety evaluation for MRP-227, Revision 1-A, did identify one A/LAI, which pertains to an applicants basis for resolving generic operating experience with the occurrence of cracking in Westinghouse-designed baffle-former bolts or CE-designed core shroud bolts. Since A/LAI No. 1 on MRP-227, Revision 1-A, is applicable to an SLR applicants basis for addressing relevant operating experience, it is acceptable for the applicant to address its resolution of A/LAI No. 1 as part of its bases for addressing relevant operating experience for the baffle-former bolts or core shroud bolts in the Operating Experience program element of the applicants PWR Vessel Internals AMP, or in the applicants technical basis document or procedure for the AMP. A separate SLRA section addressing the A/LAI would not be necessary. The clarifications made in this Actions section satisfy Objective 4, as referenced in the Rationale section of this SLR-ISG.

SLR-ISG Objective 5Closure of RIS 2011-07 The staffs guidance in RIS 2011-07 addresses differences in aging management criteria for a plants PWR RVI components based on the timing of the initial LRA submittal and the applicability and specified guidance criteria in the GALL Report version referenced in the LRA.

The guidance in RIS 2011-07 no longer applies to future license renewal or SLR applicants because LRAs will be submitted in accordance with the criteria in either the GALL-SLR Report or the GALL Report, Revision 2, and SLRAs will be submitted in accordance with the GALL-SLR

SLR-ISG-PWRVI-2020-XX Page 7 of 9 Report. As such, the staff is formally closing the guidance of RIS 2011-07 in SLR-ISG-PWRVI-2020-XX.

The clarification made in this Actions section satisfies Objective 5, as referenced in the Rationale section of this SLR-ISG.

NEWLY IDENTIFIED SYSTEMS, STRUCTURES, AND COMPONENTS UNDER 10 CFR 54.37(b)

Any structures and components identified in this SLR-ISG as requiring aging management that were not previously identified in earlier versions of the SRP-SLR or GALL-SLR Report are considered to be newly identified structures and components under 10 CFR 54.37(b).

Specifically, the staffs update of AMR items and GALL-SLR Report AMP XI.M16A in this SLR-ISG is based (in part) on the EPRI MRPs analysis of PWR RVI components in MRP-227, Revision 1-A. Any new components identified for aging management in this SLR-ISG are based on the EPRI MRPs analysis and decision to place new PWR RVI components in the Primary, Expansion, or Existing Program categories of MRP-227, Revision 1-A, in addition to those that these categories previously included in MRP-227-A.

BACKFITTING AND ISSUE FINALITY DISCUSSION Discussion to be provided in the final ISG.

CONGRESSIONAL REVIEW ACT Discussion to be provided in the final ISG.

FINAL RESOLUTION By July 1, 2027, the staff will transition this information into NUREG-2191 (GALL-SLR Report) and NUREG-2192 (SRP-SLR). Following the transition of this guidance to NUREG-2191 and NUREG-2192, this ISG will be closed.

APPENDICES A. Proposed Revisions to SRP-SLR Table 3.1-1 B.1 Proposed Revisions to GALL-SLR Report Table IV.B2, Reactor Vessel Internals (PWR)Westinghouse B.2 Proposed Revisions to GALL-SLR Report Table IV.B3, Reactor Vessel Internals (PWR)Combustion Engineering B.3 Proposed Revisions to GALL-SLR Report Table IV.B4, Reactor Vessel Internals (PWR)Babcock & Wilcox B.4 Proposed Revisions to GALL-SLR Report Table IV.E, Common Miscellaneous Material/Environment Combinations C. Proposed Revisions to SRP-SLR Section 3.1.2.2.9, (AMR Further Evaluation Acceptance Criteria) and SRP-SLR Section 3.1.3.2.9 (AMR Further Evaluation Review Procedures)

SLR-ISG-PWRVI-2020-XX Page 8 of 9 D. Proposed Revisions to GALL-SLR Report AMP XI.M16A, PWR Vessel Internals, and Related FSAR Supplement Example in GALL-SLR Report Table XI-01 E. Proposed Revision to GALL-SLR Report Table IX.C, Use of Terms for Materials F. Proposed Revisions to SRP-SLR Table 4.7-1, Examples of Potential Plant-Specific TLAA Topics G. List of Abbreviations Commonly Used in SLR-ISG-PWRVI-2020-XX REFERENCES

1. U.S. Code of Federal Regulations, Requirements for renewal of operating licenses for nuclear power plants, Part 54, Chapter 1, Title 10, Energy.

2. NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, July 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17188A158).

3. NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, July 2017 (ADAMS Accession Nos. ML17187A031 and ML17187A204).

4. NRC Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors, June 3, 2013 (ADAMS Accession No. ML12270A436).

5. NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report, December 2010 (ADAMS Accession No. ML103490041).

6. NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, December 2010 (ADAMS Accession No. ML103490036).

7. EPRI Technical Report No. 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), December 2019 (ADAMS Accession No. ML19339G350).

8. NRC Safety Evaluation, Final Safety Evaluation for Electric Power Research Institute Topical Report MRP-227, Revision 1, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline, April 25, 2019 (ADAMS Accession No. ML19081A001).

9. Letter from J. Holonich (NRC) to Brian Burgos (EPRI), U.S. Nuclear Regulatory Commission Verification Letter for Electric Power Research Institute Topical Report MRP 227, Revision 1, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline, February 19, 2020 (ADAMS Accession No. ML20006D152).

SLR-ISG-PWRVI-2020-XX Page 9 of 9

10. Email from J. Holonich (NRC) to K. Amberge (EPRI), Transmittal of MRP-227, Rev 1-A Supplemental Information -A Verification, July 7, 2020 (ADAMS Accession No. ML20175A149).

11. NRC, Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227),

Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines, December 16, 2011 (ADAMS Accession No. ML11308A770).

12. NRC Regulatory Information Summary 2011-07, License Renewal Submittal Information for Pressurized Water Reactor Internals Aging Management, July 21, 2011 (ADAMS Accession No. ML111990086).

13. EPRI Technical Report No. 1022863, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), December 2011 (ADAMS Accession Nos. ML12017A194, ML12017A196, ML12017A197, ML12017A191, ML12017A192, ML12017A195, and ML12017A199).

14. EPRI Technical Report No. 1016596, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227 Revision 0),

December 2008 (ADAMS Accession Nos. ML090160204 (Cover letter from EPRI MRP) and ML090160206 (Final Report).

15. U.S. Code of Federal Regulations, Domestic licensing of production and utilization facilities, Part 50, Chapter 1, Title 10, Energy.

16. NEI 03-08, Revision 3, Guideline for the Management of Materials Issues February 2017 (ADAMS Accession No. ML19079A253).

APPENDIX A PROPOSED REVISIONS TO SRP-SLR TABLE 3.1-1, Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report Proposed revisions to NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), Table 3.1-1, Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL-SLR Report, are provided in redline format. The revised items below supersede the respective items in SRP-SLR, Revision 0, Table 3.1-1.

SLR-ISG-PWRVI-2020-XX: Appendix A Page 2 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item M 028 PWR Westinghouse-specific Loss of material due to AMP XI.M16A, "PWR Yes (SRP-SLR IV.B2.RP-356 "Existing Programs" wear; cracking due to Vessel Internals," and Section 3.1.2.2.9) IV.B3.RP-357 components: Stainless SCC, irradiation-assisted AMP XI.M2, "Water IV.B3.RP-400 steel, nickel alloy SCCIASCC, fatigue Chemistry" (for SCC IV.B2.RP-355 (if Westinghouse , and X-750 mechanisms only) AMP XI.M16A is control rod guide tube credited for support pins (split pins), aging and Combustion management)

Engineering thermal shield positioning pins; Zircaloy-4 IV.E.R-444 (if Combustion Engineering components are incore instrumentation defined as thimble tubes exposed to ASME Section reactor coolant and XI category neutron flux components and the XI.M1 ISI AMP is credited for aging management)

IV.B2.RP-265 (if components can be placed in the No Additional Measures category)

M 029 BWR Nickel alloy core shroud Cracking due to SCC, AMP XI.M9, "BWR Yes (SRP-SLR IV.B1.R-94 and core plate access hole IGSCC, irradiation- Vessel Internals," and Section cover (welded covers) assisted SCCIASCC AMP XI.M2, "Water 3.1.2.2.12) exposed to reactor coolant Chemistry"

SLR-ISG-PWRVI-2020-XX: Appendix A Page 3 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item MD 032 PWR Stainless steel, nickel Cracking, loss of material AMP XI.M1, "ASME No IV.B2.RP-382 alloy, or CASS reactor due to wear Section XI Inservice IV.B3.RP-382 vessel internals, core Inspection, IV.B4.RP-382 support structure (not Subsections IWB, already referenced as IWC, and IWD" ASME Section XI Examination Category B-N-3 core support structure components in MRP-227-A), exposed to reactor coolant and neutron flux M 041 BWR Nickel alloy core shroud Cracking due to SCC, AMP XI.M9, "BWR Yes (SRP-SLR IV.B1.R-95 and core plate access hole IGSCC, irradiation- Vessel Internals," and Section cover (mechanical covers) assisted SCCIASCC AMP XI.M2, 3.1.2.2.12) exposed to reactor coolant "Water Chemistry" M 051a PWR Stainless steel, nickel alloy Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B4.RP-241 Babcock & Wilcox reactor irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B4.RP-241a internal "Primary" SCCIASCC, fatigue Internals," and IV.B4.RP-242a components exposed to AMP XI.M2, "Water IV.B4.RP-247 reactor coolant, neutron Chemistry" (for SCC IV.B4.RP-247a flux mechanisms only) IV.B4.RP-248 IV.B4.RP-248a IV.B4.RP-249a IV.B4.RP-252a IV.B4.RP-252e IV.B4.RP-256 IV.B4.RP-256a IV.B4.RP-258a IV.B4.RP-259a IV.B4.RP-261 IV.B4.RP-400

SLR-ISG-PWRVI-2020-XX: Appendix A Page 4 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item M 051b PWR Stainless steel, nickel alloy Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B4.RP-244 Babcock & Wilcox reactor irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B4.RP-244a internal "Expansion" SCCIASCC, fatigue, Internals," and IV.B4.RP-245 components exposed to overload AMP XI.M2, "Water IV.B4.RP-245a reactor coolant, neutron Chemistry" (for SCC IV.B4.RP-246 flux mechanisms only) IV.B4.RP-246a IV.B4.RP-246c IV.B4.RP-246d IV.B4.RP-252b IV.B4.RP-254 IV.B4.RP-254a IV.B4.RP-260a IV.B4.RP-262 IV.B4.RP-352 IV.B4.RP-250a IV.B4.RP-386 M 052a PWR Stainless steel, nickel alloy Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B3.RP-312 Combustion Engineering irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B3.RP-314 reactor internal "Primary" SCCIASCC, fatigue Internals," and IV.B3.RP-322 components exposed to AMP XI.M2, "Water IV.B3.RP-324 reactor coolant, neutron Chemistry" (for SCC IV.B3.RP-326a flux mechanisms only) IV.B3.RP-327 IV.B3.RP-328 IV.B3.RP-342 IV.B3.RP-358 IV.B3.RP-362a IV.B3.RP-363 IV.B3.RP-338 IV.B3.RP-343

SLR-ISG-PWRVI-2020-XX: Appendix A Page 5 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item M 052b PWR Stainless steel, nickel alloy Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B3.RP-313 Combustion Engineering irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B3.RP-316 reactor internal SCCIASCC, fatigue Internals," and IV.B3.RP-323 "Expansion" components AMP XI.M2, IV.B3.RP-325 exposed to reactor coolant, "Water Chemistry" (for IV.B3.RP-329 neutron flux SCC mechanisms IV.B3.RP-330 only) IV.B3.RP-333 IV.B3.RP-335 IV.B3.RP-362c IV.B3.RP-363 M 052c PWR Stainless steel, nickel alloy Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B3.RP-320 Combustion Engineering irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B3.RP-320a reactor internal "Existing SCCIASCC, fatigue Internals," and IV.B3.RP-334 Programs" components AMP XI.M2, exposed to reactor coolant, "Water Chemistry" (for neutron flux SCC mechanisms only)

M 053a PWR Stainless steel, nickel alloy Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B2.RP-270a Westinghouse reactor irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B2.RP-271 internal "Primary" SCCIASCC, fatigue Internals," and IV.B2.RP-275 components exposed to AMP XI.M2, IV.B2.RP-276 reactor coolant, neutron "Water Chemistry" (for IV.B2.RP-280 flux SCC mechanisms IV.B2.RP-296a only) IV.B2.RP-298 IV.B2.RP-302 IV.B2.RP-387 M 053b PWR Stainless steel Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B2.RP-273 Westinghouse reactor irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B2.RP-278 internal "Expansion" SCCIASCC, fatigue Internals," and IV.B2.RP-280 components exposed to AMP XI.M2, IV.B2.RP-286 reactor coolant and "Water Chemistry" (for IV.B2.RP-291 neutron flux SCC mechanisms IV.B2.RP-291a only) IV.B2.RP-291b IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-298a IV.B2.RP-387a

SLR-ISG-PWRVI-2020-XX: Appendix A Page 6 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item M 053c PWR Stainless steel, nickel Cracking due to SCC, AMP XI.M16A, Yes (SRP-SLR IV.B2.RP-289 alloy, or stellite irradiation-assisted "PWR Vessel Section 3.1.2.2.9) IV.B2.RP-301 Westinghouse reactor SCCIASCC, fatigue Internals," and IV.B2.RP-345a internal "Existing AMP XI.M2, IV.B2.RP-346 Programs" components "Water Chemistry" (for IV.B2.RP-399 exposed to reactor coolant, SCC mechanisms IV.B2.RP-355 neutron flux only)

M 054 PWR Stainless steel Loss of material due to AMP XI.M37, No IV.B2.RP-284 Westinghouse-design wear "Flux Thimble Tube bottom mounted Inspection" instrument system flux thimble tubes (with or without chrome plating) exposed to reactor coolant and neutron flux M 056a PWR Stainless steel (SS, Loss of fracture toughness AMP XI.M16A, Yes (SRP-SLR IV.B3.RP-315 including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B3.RP-318 martensitic SS) or nickel embrittlement and for Internals" IV.B3.RP-359 alloy Combustion CASS, martensitic SS, IV.B3.RP-360 Engineering reactor and PH SS due to thermal IV.B3.RP-362 internal "Primary" aging embrittlement; IV.B3.RP-364 components exposed to changes in dimensions IV.B3.RP-366 reactor coolant and due to void swelling, IV.B3.RP-365 neutron flux distortion; loss of preload IV.B3.RP-326 due to thermal and IV.B3.RP-338a irradiation-enhanced stress relaxation, creep; loss of material due to wear

SLR-ISG-PWRVI-2020-XX: Appendix A Page 7 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item M 056b PWR Stainless steel (SS, Loss of fracture toughness AMP XI.M16A, Yes (SRP-SLR IV.B3.RP-317 including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B3.RP-331 martensitic SS) embrittlement and for Internals" IV.B3.RP-333a Combustion Engineering CASS, martensitic SS, IV.B3.RP-359a "Expansion" reactor and PH SS due to thermal IV.B3.RP-361 internal components aging embrittlement; IV.B3.RP-362b exposed to reactor coolant changes in dimensions IV.B3.R-455 and neutron flux due to void swelling, IV.B3.RP-364 distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation, creep; loss of material due to wear M 056c PWR Stainless steel (SS, Loss of fracture toughness AMP XI.M16A, Yes (SRP-SLR IV.B3.RP-319 including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B3.RP-332 martensitic SS) or nickel embrittlement and for Internals" IV.B3.RP-334a alloy Combustion CASS, martensitic SS, IV.B3.RP-336 Engineering reactor and PH SS due to thermal IV.B3.RP-357 internal "Existing aging embrittlement; Programs" components changes in dimensions exposed to reactor coolant due to void swelling, and neutron flux distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation, creep; loss of material due to wear

SLR-ISG-PWRVI-2020-XX: Appendix A Page 8 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item M 058a PWR Stainless steel (SS, Loss of fracture toughness AMP XI.M16A, Yes (SRP-SLR IV.B4.RP-240 including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B4.RP-240a martensitic SS), nickel embrittlement and for Internals" IV.B4.RP-242 alloy Babcock & Wilcox CASS, martensitic SS, IV.B4.RP-247b reactor internal "Primary" and PH SS due to thermal IV.B4.RP-247c components exposed to aging embrittlement; or IV.B4.RP-248b reactor coolant and changes in dimensions IV.B4.RP-249 neutron flux due to void swelling or IV.B4.RP-251 distortion; or loss of IV.B4.RP-251a preload due to wear; or IV.B4.RP-252 loss of material due to IV.B4.RP-252d wear IV.B4.RP-256b IV.B4.RP-258 IV.B4.RP-259 IV.B4.RP-401 M 058b PWR Stainless steel (SS, Loss of fracture toughness AMP XI.M16A, Yes (SRP-SLR IV.B4.RP-245b including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B4.RP-245c martensitic SS), nickel embrittlement and for Internals" IV.B4.RP-246b alloy Babcock & Wilcox CASS, martensitic SS, IV.B4.RP-246e reactor internal and PH SS due to thermal IV.B4.RP-254b "Expansion" components aging embrittlement; or IV.B4.RP-260 exposed to reactor coolant changes in dimensions IV.B4.RP-243 and neutron flux due to void swelling, or IV.B4.RP-243a distortion; or loss of IV.B4.RP-250 preload due to thermal IV.B4.RP-252c and irradiation-enhanced IV.B4.RP-386a stress relaxation, or creep; or loss of material due to wear

SLR-ISG-PWRVI-2020-XX: Appendix A Page 9 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item M 059b PWR Stainless steel (SS, Loss of fracture toughness AMP XI.M16A, Yes (SRP-SLR IV.B2.RP-274 including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B2.RP-278a martensitic SS) embrittlement and for Internals" IV.B2.RP-280a Westinghouse reactor CASS, martensitic SS, IV.B2.RP-287 internal "Expansion" and PH SS due to thermal IV.B2.RP-290 components exposed to aging embrittlement; IV.B2.RP-290a reactor coolant and changes in dimensions IV.B2.RP-290b neutron flux due to void swelling, IV.B2.RP-292 distortion; loss of preload IV.B2.RP-295 due to thermal and IV.B2.RP-297a irradiation-enhanced IV.B2.RP-388a stress relaxation, creep; loss of material due to wear M 059c PWR Stainless steel (SS, Loss of fracture toughness AMP XI.M16A, Yes (SRP-SLR IV.B2.RP-285 including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B2.RP-288 martensitic SS), or nickel embrittlement and for Internals" IV.B2.RP-299 alloy, or stellite CASS, martensitic SS, IV.B2.RP-345 Westinghouse reactor and PH SS due to thermal internal "Existing aging embrittlement; Programs" components changes in dimensions exposed to reactor coolant due to void swelling, and neutron flux distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation, creep; loss of material due to wear M 103 BWR Stainless steel, nickel alloy Cracking due to SCC, AMP XI.M9, "BWR Yes (SRP-SLR IV.B1.R-422 reactor internal IGSCC, irradiation- Vessel Internals," and Section IV.B1.R-100 components exposed to assisted SCCIASCC AMP XI.M2, 3.1.2.2.12) IV.B1.R-105 reactor coolant and "Water Chemistry" IV.B1.R-92 neutron flux IV.B1.R-93 IV.B1.R-96 IV.B1.R-97 IV.B1.R-98 IV.B1.R-99

SLR-ISG-PWRVI-2020-XX: Appendix A Page 10 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item N 114 BWR/PWR Reactor coolant system Cracking due to SCC, AMP XI.M1, "ASME No IV.E.R-444 components defined as IGSCC, PWSCC, IASCC Section XI Inservice ASME Section XI Code (SCC mechanisms for Inspection, Class components (ASME stainless steel, nickel alloy Subsections IWB, Code Class 1 reactor components only), fatigue, IWC, and IWD," and coolant pressure boundary or cyclic loading; loss of AMP XI.M2, components, reactor material due to general "Water Chemistry" vessel interior corrosion (steel only), (water chemistry-attachments, or core pitting corrosion, crevice related or corrosion-support structure corrosion, or wear related aging effect components, ; or ASME mechanisms only)

Class 2 or 3 components -

including ASME defined appurtenances, component supports, and associated pressure boundary welds, or components subject to plant-specific equivalent classifications for these ASME code classes)

N 118 PWR Stainless steel, nickel alloy Cracking due to SCC, Plant-specific aging Yes (SRP-SLR IV.B2.R-423 PWR reactor vessel irradiation-assisted management program Section 3.1.2.2.9) IV.B3.R-423 internal components SCCIASCC, cyclic or AMP XI.M16A, IV.B4.R-423 exposed to reactor coolant, loading, fatigue "PWR Vessel neutron flux Internals," and AMP XI.M2, "Water Chemistry" (SCC and IASCC only), with adjusted site-specific or component-specific aging management basis for a given component

SLR-ISG-PWRVI-2020-XX: Appendix A Page 11 of 11 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL SLR Report New, Modified, Deleted, Further Edited Aging Management Evaluation Item ID Type Component Aging Effect/Mechanism Program (AMP)/TLAA Recommended GALL-SLR Item N 119 PWR Stainless steel, nickel alloy Loss of fracture toughness Plant-specific aging Yes (SRP-SLR IV.B2.R-424 PWR reactor vessel due to neutron irradiation management program Section 3.1.2.2.9) IV.B3.R-424 internal components embrittlement or thermal or AMP XI.M16A, IV.B4.R-424 exposed to reactor coolant, aging embrittlement; "PWR Vessel neutron flux changes in dimensions Internals," with due to void swelling or adjusted site-specific distortion; loss of preload or component-specific due to thermal and aging management irradiation-enhanced basis for a given stress relaxation or creep; component loss of material due to wear

APPENDIX B PROPOSED REVISIONS TO GALL-SLR REPORT TABLES IV.B2, IV.B3, AND IV.B4

APPENDIX B.1 PROPOSED REVISIONS TO GALL-SLR REPORT TABLE IV.B2, REACTOR VESSEL INTERNALS (PWR)WESTINGHOUSE NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Table IV.B2, Reactor Vessel Internals (PWR)Westinghouse, addresses the Westinghouse pressurized-water reactor (PWR) vessel internals, which consist of components in the upper internals assembly, the control rod guide tube (CRGT) assembly, the core barrel assembly, the baffle/former assembly, the lower internals assembly, lower support assembly, thermal shield assembly, bottom-mounted instrumentation system, and alignment and interfacing components.

Proposed revisions to Table IV.B2 of the GALL-SLR Report are provided in redline format.

These AMR items supersede the respective items in GALL-SLR Report, Revision 0, Table IV.B2.

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 2 of 16 GALL-SLR Report Table IV.B2 Proposed Revisions IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-301 3.1-1, 053c Alignment and Stainless steel Reactor Cracking due to AMP XI.M16A, Yes interfacing coolant and SCCIASCC or "PWR Vessel components: upper neutron flux fatigue Internals," and core plate AMP XI.M2, alignment pins (fuel "Water alignment pins) Chemistry" (for SCC mechanisms only)

M IV.B2.RP-299 3.1-1, 059c Alignment and Stainless steel Reactor Loss of material due AMP XI.M16A, Yes interfacing coolant and to wear; loss of "PWR Vessel components: upper neutron flux fracture toughness Internals" core plate due to neutron alignment pins (fuel irradiation alignment pins) embrittlement M IV.B2.RP-271 3.1-1, 053a Baffle-to-former Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: coolant and irradiation-assisted "PWR Vessel accessible baffle- neutron flux SCCIASCC or Internals," and to-former bolts fatigue AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 3 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-272 3.1-1, 059a Baffle-to-former Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: coolant and toughness due to "PWR Vessel accessible baffle- neutron flux neutron irradiation Internals" to-former bolts embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep; loss of material due to wear M IV.B2.RP-270 3.1-1, 059a Baffle-to-former Stainless steel Reactor Changes in AMP XI.M16A, Yes assembly: baffle coolant and dimensions due to "PWR Vessel and former plates neutron flux void swelling or Internals" distortion; loss of fracture toughness due to neutron irradiation embrittlement M IV.B2.RP-270a 3.1-1, 053a Baffle-to-former Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: baffle coolant and irradiation-assisted "PWR Vessel and former plates neutron flux SCCIASCC or Internals," and fatigue AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 4 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-275 3.1-1, 053a Baffle-to-former Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: baffle- coolant and irradiation-assisted "PWR Vessel edge bolts (all neutron flux SCCIASCC or Internals," and plants with baffle- fatigue AMP XI.M2, edge, corner bolts) "Water Chemistry" (for SCC mechanisms only)

M IV.B2.RP-354 3.1-1, 059a Baffle-to-former Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: baffle- coolant and toughness due to "PWR Vessel edge bolts (all neutron flux neutron irradiation Internals" plants with baffle- embrittlement; edge, corner bolts) changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep; loss of material due to wear M IV.B2.RP-273 3.1-1, 053b Baffle-to-former Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: barrel- coolant and irradiation-assisted "PWR Vessel to-former bolts neutron flux SCCIASCC or Internals," and fatigue AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 5 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-274 3.1-1, 059b Baffle-to-former Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: barrel- coolant and toughness due to "PWR Vessel to-former bolts neutron flux neutron irradiation Internals" embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep; loss of material due to wear M IV.B2.RP-293 3.1-1, 053b Bottom-mounted Stainless steel Reactor Cracking due to AMP XI.M16A, Yes instrumentation coolant and SCC, IASCC, or "PWR Vessel system: bottom- neutron flux fatigue Internals," and mounted AMP XI.M2, instrumentation "Water (BMI) column Chemistry" (for bodies SCC mechanisms only)

M IV.B2.RP-292 3.1-1, 059b Bottom-mounted Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes instrumentation coolant and toughness due to "PWR Vessel system: bottom- neutron flux neutron irradiation Internals" mounted embrittlement; instrumentation changes in (BMI) column dimension due to bodies void swelling or distortion; loss of material due to wear

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 6 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-296 3.1-1, 059a Control rod guide Stainless steel Reactor Loss of material due AMP XI.M16A, Yes tube (CRGT) (including coolant and to wear; loss of "PWR Vessel assemblies: CRGT CASS) neutron flux fracture due to Internals" guide plates neutron irradiation (cards) embrittlement, and for CASS, thermal aging embrittlement)

N IV.B2.RP-296a 3.1-1, 053a Control rod guide Stainless steel Reactor Cracking due to AMP XI.M16A, Yes tube (CRGT) (including coolant and SCC or fatigue "PWR Vessel assemblies: CRGT CASS) neutron flux Internals," and guide plates AMP XI.M2, (cards) "Water Chemistry" (for SCC mechanisms only)

M IV.B2.RP-298 3.1-1, 053a Control rod guide Stainless steel Reactor Cracking due to AMP XI.M16A, Yes tube (CRGT) coolant and SCC, IASCC, or "PWR Vessel assemblies: CRGT neutron flux fatigue Internals," and lower flange welds AMP XI.M2, (accessible)in outer "Water (peripheral) CRGT Chemistry" (for assemblies SCC mechanisms only)

N IV.B2.RP-298a 3.1-1, 053b Control rod guide Stainless steel Reactor Cracking due to AMP XI.M16A, Yes tube (CRGT) coolant and SCC, IASCC, or "PWR Vessel assemblies: lower neutron flux fatigue Internals," and flange welds in AMP XI.M2, remaining (non- "Water peripheral) CRGT Chemistry" (for assemblies SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 7 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-297 3.1-1, 059a Control rod guide Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes tube (CRGT) (including coolant and toughness due to "PWR Vessel assemblies: CRGT CASS) neutron flux thermal aging and Internals" lower flange welds neutron irradiation (accessible)in outer embrittlement and (peripheral) CRGT for CASS, due to assemblies thermal aging embrittlement N IV.B2.RP-297a 3.1-1, 059b Control rod guide Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes tube (CRGT) (including coolant and toughness due to "PWR Vessel assemblies: lower CASS) neutron flux neutron irradiation Internals" flange welds in the embrittlement, and remaining (non- for CASS, due to peripheral) CRGT thermal aging assemblies embrittlement M IV.B2.RP-355 3.1-1, Control rod guide Stainless steel, Reactor Cracking due to AMP XI.M16A, Yes 053c028 tube (CRGT) nickelNickel coolant and SCC or fatigue; loss "PWR Vessel assemblies: guide alloy (X-750) neutron flux of material due to Internals," and tube support pins wear AMP XI.M2, (split pins) "Water Chemistry" (for SCC mechanisms only) - using component-specific evaluation per MRP guidelines MD IV.B2.RP-356 3.1-1, 028 Control rod guide Stainless steel, Reactor Loss of material due AMP XI.M16A, Yes tube (CRGT) nickel alloy coolant and to wear "PWR Vessel assemblies: guide neutron flux Internals" tube support pins (split pins)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 8 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-345 3.1-1, 059c Core barrel Stainless steel Reactor Loss of material due AMP XI.M16A, Yes assembly: core coolant and to wear "PWR Vessel barrel flange neutron flux Internals," and AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)"

N IV.B2.RP-345a 3.1-1, 053c Core barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: core coolant and SCC or fatigue "PWR Vessel barrel flange neutron flux Internals," and AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

MD IV.B2.RP-278 3.1-1, 053b Core barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: core coolant and SCC or fatigue "PWR Vessel barrel outlet nozzle neutron flux Internals," and welds AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

MD IV.B2.RP-278a 3.1-1, 059b Core barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: core coolant and toughness due to "PWR Vessel barrel outlet nozzle neutron flux neutron irradiation Internals" welds embrittlement

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 9 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-280 3.1-1, Core barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes 053a053b assembly: lower coolant and SCC, irradiation- "PWR Vessel flange weld (core neutron flux assisted Internals," and barrel flange weld- SCC,IASCC (lower AMP XI.M2, to-support plate flange weld only), or "Water weld), upper fatigue Chemistry" (for circumferential SCC (girth) weld, and mechanisms upper vertical only)

(axial) welds N IV.B2.RP-280a 3.1-1, 059b Core barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly; lower coolant and toughness due to "PWR Vessel flange weld (core neutron flux neutron irradiation Internals" barrel-to-support embrittlement plate weld)

M IV.B2.RP-387 3.1-1, 053a Core barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: upper coolant and SCC, irradiation- "PWR Vessel core barrel and neutron flux assisted Internals," and lower core barrel SCCIASCC, or AMP XI.M2, circumferential fatigue "Water (girth) welds Chemistry" (for SCC mechanisms only)

M IV.B2.RP-388 3.1-1, 059a Core barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: upper coolant and toughness due to "PWR Vessel core barrel and neutron flux neutron irradiation Internals" lower core barrel embrittlement, circumferential changes in (girth) welds dimension due to void swelling or distortion

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 10 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-387a 3.1-1, 053b Core barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: upper coolant and SCC, irradiation- "PWR Vessel core barrel and neutron flux assisted Internals," and lower core SCCIASCC, or AMP XI.M2, barrelmiddle fatigue "Water vertical (axial) Chemistry" (for welds and lower SCC vertical (axial) mechanisms welds only)

M IV.B2.RP-388a 3.1-1, 059b Core barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: upper coolant and toughness due to "PWR Vessel core barrel and neutron flux neutron irradiation Internals" lower core embrittlement; barrelmiddle changes in vertical (axial) dimension due to welds and lower void swelling or vertical (axial) distortion welds M IV.B2.RP-276 3.1-1, 053a Core barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: upper coolant and irradiation-assisted "PWR Vessel core barrel flange neutron flux SCC or fatigue Internals," and weld AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

M IV.B2.RP-285 3.1-1, 059c Lower internals Nickel Reactor Loss of material due AMP XI.M16A, Yes assemblyAlignment alloyStainless coolant and to wear; loss of "PWR Vessel and interfacing steel, nickel neutron flux preload due to Internals" components: clevis alloy (including thermal andor insert inserts alloy 600, irradiation-enhanced (including bolts or X-750), stellite stress relaxation or screws, and clevis (for insert creep (bolts and insert surfaces) surfaces only) screws only);

changes in dimension due to distortion

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 11 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-399 3.1-1, 053c Lower internals Stainless steel, Reactor Cracking due to AMP XI.M16A, Yes assemblyAlignment nickel alloy coolant and primary water SCC, "PWR Vessel and interfacing (including neutron flux irradiation-assisted Internals," and components: clevis Alloy 600, SCC, or fatigue AMP XI.M2, insert inserts X-750), stellite "Water (including bolts or (for insert Chemistry" (for screws , dowels, surfaces only) SCC and clevis insert mechanisms surfaces) only)

M IV.B2.RP-289 3.1-1, 053c Lower internals Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: lower coolant and irradiation-assisted "PWR Vessel core plate andor neutron flux SCCIASCC or Internals," and extra-long (XL) fatigue AMP XI.M2, lower core plate "Water Chemistry" (for SCC mechanisms only)

M IV.B2.RP-288 3.1-1, 059c Lower internals Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: lower coolant and toughness due to "PWR Vessel core plate andor neutron flux neutron irradiation Internals" extra-long (XL) embrittlement; loss lower core plate of material due to wear M IV.B2.RP-291a 3.1-1, 053b Lower Stainless steel Reactor Cracking due to AMP XI.M16A, Yes supportinternals coolant and SCC or fatigue "PWR Vessel assembly: lower neutron flux Internals," and support forging or AMP XI.M2, casting "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 12 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-290a 3.1-1, 059b Lower Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes supportinternals coolant and toughness due to "PWR Vessel assembly: lower neutron flux neutron irradiation Internals" support forging or embrittlement (and casting thermal aging embrittlement for CASS, PH SS, and martensitic SS)

M IV.B2.RP-291 3.1-1, 053b Lower support Cast austenitic Reactor Cracking due to AMP XI.M16A, Yes assembly: lower stainless steel coolant and irradiation-assisted "PWR Vessel support column neutron flux SCCIASCC or Internals," and bodies (cast) fatigue AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

M IV.B2.RP-290 3.1-1, 059b Lower support Cast austenitic Reactor Loss of fracture AMP XI.M16A, Yes assembly: lower stainless steel coolant and toughness due to "PWR Vessel support column neutron flux thermal aging and Internals" bodies (cast) neutron irradiation embrittlement; changes in dimension due to void swelling or distortion M IV.B2.RP-294 3.1-1, 053b Lower support Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: lower coolant and irradiation-assisted "PWR Vessel support column neutron flux SCCIASCC or Internals," and bodies (non-cast) fatigue AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 13 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-295 3.1-1, 059b Lower support Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: lower coolant and toughness due to "PWR Vessel support column neutron flux neutron irradiation Internals" bodies (non-cast) embrittlement; changes in dimension due to void swelling or distortion M IV.B2.RP-286 3.1-1, 053b Lower support Stainless steel, Reactor Cracking due to AMP XI.M16A, Yes assembly: lower nickel alloy coolant and irradiation-assisted "PWR Vessel support column neutron flux SCCIASCC or Internals," and bolts fatigue AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

M IV.B2.RP-287 3.1-1, 059b Lower support Stainless steel, Reactor Loss of fracture AMP XI.M16A, Yes assembly: lower nickel alloy coolant and toughness due to "PWR Vessel support column neutron flux neutron irradiation Internals" bolts embrittlement; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep; changes in dimension due to void swelling or distortion; loss of material due to wear

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 14 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B2.R-423 3.1-1, 118 Reactor vessel Stainless steel, Reactor Cracking due to Plant-specific Yes internal nickel alloy coolant, SCC, irradiation- aging components, or neutron flux assisted management specified reactor SCCIASCC, cyclic program, or vessel internal loading, fatigue AMP XI.M16A, component with a "PWR Vessel site-specific or Internals," and component-specific AMP XI.M2, aging management "Water basis Chemistry" (SCC and IASCC only),

for cases where a specified component is subject to a site-specific or component-specific aging management basis N IV.B2.R-424 3.1-1, 119 Reactor vessel Stainless steel, Reactor Loss of fracture Plant-specific Yes internal nickel alloy coolant, toughness due to aging components, or neutron flux neutron irradiation management specified reactor embrittlement or program, or vessel internal thermal aging AMP XI.M16A, component with a embrittlement; PWR Vessel site-specific or changes in Internals, for component-specific dimensions due to cases where a aging management void swelling or specified basis distortion; loss of component is preload due to subject to a thermal and site-specific or irradiation-enhanced component-stress relaxation or specific aging creep; loss of management material due to wear basis

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 15 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation MD IV.B2.RP-382 3.1-1, 032 Reactor vessel Stainless steel, Reactor Cracking due to AMP XI.M1, No internals: ASME nickel alloy, coolant and fatigue, SCC, or "ASME Section Section XI, cast austenitic neutron flux irradiation-assisted XI Inservice Examination stainless steel SCC; loss of Inspection, Category B-N-3 material due to wear Subsections core support IWB, IWC, and structure IWD" components (not already identified as "Existing Programs" components in MRP-227-A)

M IV.B2.RP-302a 3.1-1, 059a Thermal shield Stainless steel Reactor Loss of material due AMP XI.M16A, Yes assembly: thermal coolant and to wear; loss of "PWR Vessel shield flexures neutron flux fracture toughness Internals" due to neutron irradiation embrittlement; loss of preload due to irradiation-enhanced stress relaxation or creep M IV.B2.RP-302 3.1-1, 053a Thermal shield Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: thermal coolant and SCC, IASCC, or "PWR Vessel shield flexures neutron flux fatigue Internals," and AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.1 Page 16 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B2 Reactor Vessel Internals (PWR)Westinghouse New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B2.RP-291b 3.1-1, 053b Upper internals Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly; upper coolant and IASCC or fatigue "PWR Vessel core plate neutron flux Internals," and AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

M IV.B2.RP-290b 3.1-1, 059b Upper internals Stainless steel Reactor Loss of material due AMP XI.M16A, Yes assembly; upper coolant and to wear: loss of "PWR Vessel core plate neutron flux fracture toughness Internals" due to neutron irradiation embrittlement

APPENDIX B.2 PROPOSED REVISIONS TO GALL-SLR REPORT TABLE IV.B3, REACTOR VESSEL INTERNALS (PWR)COMBUSTION ENGINEERING NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Table IV.B3, Reactor Vessel Internals (PWR)Combustion Engineering, addresses the Combustion Engineering (CE) pressurized-water reactor (PWR) vessel internals, which consist of components in the upper internals assembly, the control element assembly (CEA), the core support barrel assembly, the core shroud assembly, and the lower support structure assembly, and incore instrumentation components.

Proposed revisions to Table IV.B3 of the GALL-SLR Report are provided in redline format.

These AMR items superseded the respective items in GALL-SLR Report, Revision 0, Table IV.B3.

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 2 of 16 GALL-SLR Report Table IV.B3 Proposed Revisions IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-313 3.1-1, 052b Control element Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (CEA): coolant and SCC or fatigue "PWR Vessel shroud assemblies:) - neutron flux Internals," and Shroud Assemblies: AMP XI.M2, remaining instrument "Water guide tubes (i.e., Chemistry" (for guide tubes in non- SCC peripheral mechanisms CEAcontrol element only) shroud assemblies)

M IV.B3.RP-312 3.1-1, 052a Control element Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (CEA): coolant and SCC or fatigue "PWR Vessel shroud assemblies) - neutron flux Internals," and Shroud Assemblies: AMP XI.M2, instrument guide "Water tubes in peripheral Chemistry" (for CEA shroud SCC assemblies mechanisms only)

M IV.B3.RP-320 3.1-1, 052c Core shroud and Stainless steel Reactor Cracking due to AMP XI.M16A, Yes upper internals coolant and fatigue "PWR Vessel assemblies (all neutron flux Internals" plants):: guide lugs; insertguide lug inserts and bolts M IV.B3.RP-319 3.1-1, 056c Core shroud and Stainless steel Reactor Loss of material AMP XI.M16A, Yes upper internals coolant and due to wear; Loss "PWR Vessel assemblies (all neutron flux of preload due to Internals" plants):: guide lugs; thermal and insertguide lug inserts irradiation-and bolts enhanced stress relaxation or creep

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 3 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-358 3.1-1, 052a Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assemblies (for coolant and irradiation-assisted "PWR Vessel bolted core shroud neutron flux SCCIASCC Internals," and assemblies): AMP XI.M2, assembly "Water components, Chemistry" (for including core side SCC surfaces, shroud mechanisms plates and former only) platesplate joints, and bolts and bolt locking devices M IV.B3.RP-318 3.1-1, 056a Core shroud Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assemblies (for coolant and toughness due to "PWR Vessel bolted core shroud neutron flux neutron irradiation Internals" assemblies): embrittlement; assembly changes in components, dimensions due to including core side void swelling or surfaces, shroud distortion plates and former platesplate joints, and bolts and bolt locking devices M IV.B3.RP-316 3.1-1, 052b Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assemblies (for coolant and irradiation-assisted "PWR Vessel bolted core shroud neutron flux SCCIASCC or Internals," and assemblies): barrel- fatigue AMP XI.M2, shroud bolts "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 4 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-314 3.1-1, 052a Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assemblies (for coolant and irradiation-assisted "PWR Vessel bolted core shroud neutron flux SCCIASCC or Internals," and assemblies): core fatigue AMP XI.M2, shroud bolts "Water Chemistry" (for SCC mechanisms only)

M IV.B3.RP-315 3.1-1, 056a Core shroud Stainless steel Reactor Loss of preload AMP XI.M16A, Yes assemblies (for coolant and due to thermal and "PWR Vessel bolted core shroud neutron flux irradiation- Internals" assemblies): core enhanced stress shroud bolts relaxation or creep; loss of fracture toughness due to neutron irradiation embrittlement; changes in dimension due to void swelling or distortion M IV.B3.RP-326 3.1-1, 056a Core shroud Stainless steel Reactor Changes in AMP XI.M16A, Yes assembly (for welded coolant and dimensions due to "PWR Vessel shroud designs neutron flux void swelling or Internals" assembled in two distortion; loss of vertical sections): fracture toughness assembly due to neutron components, irradiation (including monitoring embrittlement of the gap opening at the core shroud re-entrant cornersthe horizontal seam between the upper and lower shroud segments)

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 5 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-326a 3.1-1, 052a Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (for welded coolant and SCCIASCC or "PWR Vessel shroud designs neutron flux fatigue Internals," and assembled in two AMP XI.M2, vertical sections): "Water assembly Chemistry" (for components, SCC (including monitoring mechanisms of the gap opening at only) the core shroud re-entrant cornersthe horizontal seam between the upper and lower shroud segments)

M IV.B3.RP-322 3.1-1, 052a Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (for welded coolant and irradiation-assisted "PWR Vessel core shroud designs neutron flux SCCIASCC Internals," and assembled in two AMP XI.M2, vertical sections): "Water core shroud plate-to- Chemistry" (for former plate welds SCC mechanisms only)

M IV.B3.RP-359 3.1-1, 056a Core shroud Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly (for welded coolant and toughness due to "PWR Vessel core shroud designs neutron flux neutron irradiation Internals" assembled in two embrittlement; vertical sections): changes in core shroud plate-to- dimensions due to former plate welds void swelling or distortion

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 6 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-323 3.1-1, 052b Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (for welded coolant and irradiation-assisted "PWR Vessel core shroud designs neutron flux SCCIASCC Internals," and assembled in two AMP XI.M2, vertical sections): "Water remaining axial welds Chemistry" (for SCC mechanisms only)

M IV.B3.RP-359a 3.1-1, 056b Core shroud Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly (for welded coolant and toughness due to "PWR Vessel core shroud designs neutron flux neutron irradiation Internals" assembled in two embrittlement; vertical sections): changes in remaining axial welds dimensions due to void swelling or distortion M IV.B3.RP-325 3.1-1, 052b Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (for core coolant and irradiation-assisted "PWR Vessel shroud designs neutron flux SCCIASCC Internals," and assembled with full- AMP XI.M2, height shroud plates): "Water remaining axial Chemistry" (for welds, ribs, and rings SCC mechanisms only)

M IV.B3.RP-361 3.1-1, 056b Core shroud Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly (for core coolant and toughness due to "PWR Vessel shroud designs neutron flux neutron irradiation Internals" assembled with full- embrittlement height shroud plates):

remaining axial welds, ribs, and rings

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 7 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-360 3.1-1, 056a Core shroud Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly (for core coolant and toughness due to "PWR Vessel shroud designs neutron flux neutron irradiation Internals" assembled with full- embrittlement; height shroud plates): changes in shroud plateplates dimension due to (including visible axial void swelling or weld seams at the distortion core shroud re-entrant corners and at the core midplane)

M IV.B3.RP-324 3.1-1, 052a Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (core coolant and irradiation-assisted "PWR Vessel shroud designs neutron flux SCCIASCC Internals," and assembled with full- AMP XI.M2, height shroud plates): "Water shroud plates, Chemistry" (for (including visible axial SCC weld seams at the mechanisms core shroud re- only) entrant corners, and at the core mid-plane

(+3 feet in height) as visible from the core side of the shroud)

M IV.B3.RP-328 3.1-1, 052a Core support barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: lower core coolant and SCC or fatigue "PWR Vessel barrel flangeflexure neutron flux Internals," and weld AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 8 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-362 3.1-1, 056a Core support barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: lower coolant and toughness due to "PWR Vessel cylindermiddle neutron flux neutron irradiation Internals" circumferential (girth) embrittlement weldsweld M IV.B3.RP-362a 3.1-1, 052a Core support barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: lower coolant and SCC or irradiation- "PWR Vessel cylindermiddle neutron flux assisted Internals," and circumferential (girth) SCCIASCC AMP XI.M2, weldsweld "Water Chemistry" (for SCC mechanisms only)

M IV.B3.RP-362c 3.1-1, 052b Core support barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: lower coolant and SCC or irradiation- "PWR Vessel cylindermiddle neutron flux assisted Internals," and vertical (axial) welds SCCIASCC AMP XI.M2, and lower vertical "Water (axial) welds Chemistry" (for SCC mechanisms only)

M IV.B3.RP-362b 3.1-1, 056b Core support barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: lower coolant and toughness due to "PWR Vessel cylindermiddle neutron flux neutron irradiation Internals" vertical (axial) welds embrittlement and lower vertical (axial) welds M IV.B3.RP-333 3.1-1, 052b Core support barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: lower girth coolant and SCC, IASCC, or "PWR Vessel weld (lower flange neutron flux fatigue Internals," and weld) AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 9 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B3.RP-333a 3.1-1, 056b Core support barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: lower girth coolant and toughness due to "PWR Vessel weld (lower flange neutron flux neutron irradiation Internals" weld) embrittlement MD IV.B3.RP-400 3.1-1, 028 Core support barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: thermal coolant and SCC, irradiation- "PWR Vessel shield positioning neutron flux assisted SCC or Internals," and pins fatigue; loss of AMP XI.M2, material due to "Water wear Chemistry" (for SCC mechanisms only)

M IV.B3.RP-332 3.1-1, 056c Core support barrel Stainless steel Reactor Loss of material AMP XI.M16A, Yes assembly: upper core coolant and due to wear "PWR Vessel barrel flange neutron flux Internals" M IV.B3.RP-327 3.1-1, 052a Core support barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: upper core coolant and SCC "PWR Vessel support barrel flange neutron flux Internals," and weld AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

N IV.B3.R-455 3.1-1, 056b Core support barrel Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly: upper coolant and toughness due to "PWR Vessel cylinder (base neutron flux neutron irradiation Internals," and metalcircumferential embrittlement AMP XI.M2, (girth) weld and upper "Water vertical (axial) welds) Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 10 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-329 3.1-1, 052b Core support barrel Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: upper coolant and SCC "PWR Vessel cylinder (base metal neutron flux Internals," and and AMP XI.M2, welds)circumferential "Water (girth) weld and upper Chemistry" (for core barrel flange SCC (flange base mechanisms metal)vertical (axial) only) welds M IV.B3.RP-357 3.1-1, Incore instruments Zircaloy-4 Reactor Loss of material AMP XI.M16A, Yes 028056c (ICI): ICI thimble coolant and due to wear "PWR Vessel tubes - lower neutron flux Internals" M IV.B3.RP-363 3.1-1, Lower support Stainless steel Reactor Cracking due to AMP XI.M16A, Yes 052a052b structure (all plants coolant and SCC, irradiation- "PWR Vessel with either full height neutron flux assisted Internals," and bolted or half height SCCIASCC, or AMP XI.M2, welded shroud fatigue "Water plates): core support Chemistry" (for column SCC weldscolumns mechanisms only)

M IV.B3.RP-364 3.1-1, Lower support Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes 056a056b structure (all plants (including coolant and toughness due to "PWR Vessel with either full height CASS) neutron flux neutron irradiation Internals" bolted or half height and thermal welded shroud embrittlement (TE plates): core support for CASS materials column only) weldscolumns

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 11 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-334 3.1-1, 052c Lower support Stainless steel Reactor Cracking due to AMP XI.M16A, Yes structure (for CE coolant and SCC, irradiation- "PWR Vessel plants with core neutron flux assisted Internals," and shroud designs SCCIASCC, or AMP XI.M2, assembled in two fatigue "Water vertical sections or Chemistry" (for withfrom full-height SCC shroud plates): fuel mechanisms alignment pins only)

M IV.B3.RP-336 3.1-1, 056c Lower support Stainless steel Reactor Loss of material AMP XI.M16A, Yes structure (for CE coolant and due to wear; loss "PWR Vessel plants with core neutron flux of fracture Internals" shroud designs toughness due to assembled in two neutron irradiation vertical sections or embrittlement; loss from full height of preload due to shroud plates): fuel thermal and alignment pins irradiation-enhanced stress relaxation or creep MD IV.B3.RP-334a 3.1-1, 056c Lower support Stainless steel Reactor Loss of material AMP XI.M16A, Yes structure (designs coolant and due to wear; loss "PWR Vessel assembled with full- neutron flux of fracture Internals" height shroud plates): toughness due to fuel alignment pins neutron irradiation embrittlement; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 12 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-335 3.1-1, 052b Lower support Stainless steel Reactor Cracking due to AMP XI.M16A, Yes structure (all CE coolant and SCC, irradiation- "PWR Vessel plants except those neutron flux assisted SCC, or Internals," and with welded core fatigue AMP XI.M2, shroud designs "Water assembled withfrom Chemistry" (for full-height shroud SCC plates): lower core mechanisms support beams only)

M IV.B3.RP-343 3.1-1, 052a Lower support Stainless steel Reactor Cracking due to AMP XI.M16A, Yes structure (for CE coolant and fatigue "PWR Vessel plant designs with a neutron flux Internals," and core support plate): AMP XI.M2, core support plate "Water Chemistry" (for SCC mechanisms only)

M IV.B3.RP-365 3.1-1, 056a Lower support Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes structure (for CE coolant and toughness due to "PWR Vessel plant designs with a neutron flux neutron irradiation Internals" core support plate): embrittlement core support plate M IV.B3.RP-342 3.1-1, 052a Lower support Stainless steel Reactor Cracking due to AMP XI.M16A, Yes structure (designs for coolant and SCC, irradiation- "PWR Vessel CE plants with neutron flux assisted Internals," and welded core shrouds SCCIASCC, or AMP XI.M2, shroud designs fatigue "Water assembled withfrom Chemistry" (for full height shroud SCC plates): deep beams mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 13 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B3.RP-366 3.1-1, 056a Lower support Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes structure (for CE coolant and toughness due to "PWR Vessel plants with welded neutron flux neutron irradiation Internals" core shroud designs embrittlement with assembled from full height shroud plates): deep beams M IV.B3.RP-330 3.1-1, 052b Lower support Stainless steel Reactor Cracking due to AMP XI.M16A, Yes structure: (for CE coolant and irradiation-assisted "PWR Vessel plants with bolted neutron flux SCCIASCC or Internals," and designs): core fatigue AMP XI.M2, support column bolts "Water Chemistry" (for SCC mechanisms only)

M IV.B3.RP-331 3.1-1, 056b Lower support Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes structure: (for CE coolant and toughness due to "PWR Vessel plants with bolted neutron flux neutron irradiation Internals" designs): core embrittlement support column bolts

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 14 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B3.R-423 3.1-1, 118 Reactor vessel Stainless Reactor Cracking due to Plant-specific Yes internal components steel, nickel coolant, SCC, irradiation- aging or reactor vessel alloy neutron flux assisted management internal component- SCCIASCC, cyclic program, or specific basis for a loading, fatigue AMP XI.M16A, specified RVI "PWR Vessel component Internals," and AMP XI.M2, "Water Chemistry" (SCC and IASCC only),

for cases where a specified component is subject to a site-specific or component-specific aging management basis

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 15 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B3.R-424 3.1-1, 119 Reactor vessel Stainless Reactor Loss of fracture Plant-specific Yes internal components steel, nickel coolant, toughness due to aging or reactor vessel alloy neutron flux neutron irradiation management internal component- embrittlement or program, or specific basis for a thermal aging AMP XI.M16A, specified RVI embrittlement; PWR Vessel component changes in Internals, for dimensions due to cases where a void swelling or specified distortion; loss of component is preload due to subject to a thermal and site-specific or irradiation- component-enhanced stress specific aging relaxation or management creep; loss of basis material due to wear MD IV.B3.RP-382 3.1-1, 032 Reactor vessel Stainless Reactor Cracking due to AMP XI.M1, No internals: ASME steel, nickel coolant and fatigue, SCC, or "ASME Section Section XI, alloy, cast neutron flux irradiation-assisted XI Inservice Examination austenitic SCC; loss of Inspection, Category B-N-3 core stainless steel material due to Subsections support structure wear IWB, IWC, and components (not IWD" already identified as "Existing Programs" components in MRP-227-A)

M IV.B3.RP-338 3.1-1, 052a Upper internals Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (designs for coolant and fatigue "PWR Vessel CE plants with core neutron flux Internals" shrouds shroud designs assembled withfrom full height shroud plates): fuel alignment plate

SLR-ISG-PWRVI-2020-XX: Appendix B.2 Page 16 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B3 Reactor Vessel Internals (PWR)Combustion Engineering New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B3.RP-338a 3.1-1, 056a Upper internals Stainless steel Reactor Loss of fracture AMP XI.M16A, Yes assembly (for CE coolant and toughness due to "PWR Vessel plants with core neutron flux neutron irradiation Internals" shroud designs embrittlement assembled from full height shroud plates):

fuel alignment plate N IV.B3.RP-320a 3.1-1, 052c Alignment and Stainless, Reactor Cracking due to AMP XI.M16A, Yes Interfacing steel, nickel coolant and SCC "PWR Vessel Components: core alloy neutron flux Internals" stabilizing lugs, shims and bolts

APPENDIX B.3 PROPOSED REVISIONS TO GALL-SLR REPORT TABLE IV.B4, REACTOR VESSEL INTERNALS (PWR)BABCOCK & WILCOX NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Table IV.B4, Reactor Vessel Internals (PWR)Babcock &

Wilcox, addresses the Babcock & Wilcox (B&W) pressurized-water reactor (PWR) vessel internals, which consist of components in the plenum cover assembly, the upper grid assembly, the control rod guide tube (CRGT) assembly, the core support shield assembly, the core barrel assembly, the lower grid assembly, the incore monitoring instrument (IMI) guide tube assembly, and the flow distributor assembly.

Proposed revisions to Table IV.B4 of the GALL-SLR Report are provided in redline format.

These AMR items supersede the respective items in GALL-SLR Report, Revision 0, Table IV.B4.

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 2 of 16 GALL-SLR Report Table IV.B4 Proposed Revisions IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B4.RP-245 3.1-1, 051b Core barrel assembly Stainless Reactor Cracking due to AMP XI.M16A, Yes (applicable to Davis steel, nickel coolant and SCC "PWR Vessel Besse only): alloy neutron flux Internals," and surveillance AMP XI.M2, specimen holder tube "Water (SSHT) studs/nuts or Chemistry" bolts N IV.B4.RP-245c 3.1-1, 058b Core barrel assembly Stainless Reactor Loss of material AMP XI.M16A, Yes (applicable to Davis steel, nickel coolant and due to wear; loss "PWR Vessel Besse only): alloy neutron flux of preload due to Internals surveillance thermal or specimen holder tube irradiation-(SSHT) studs or bolts enhanced stress relaxation or creep M IV.B4.RP-247 3.1-1, 051a Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: accessible steel, nickel coolant and SCC or fatigue "PWR Vessel lower core barrel alloy neutron flux Internals," and (LCB) bolts and AMP XI.M2, locking devices "Water Chemistry" N IV.B4.RP-247c 3.1-1, 058a Core barrel Stainless Reactor Loss of material AMP XI.M16A, Yes assembly: lower core steel, nickel coolant and due to wear; loss "PWR Vessel barrel (LCB) bolts alloy neutron flux of preload due to Internals thermal and irradiation-enhanced stress relaxation or creep M IV.B4.RP-249a 3.1-1, 051a Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: baffle steel coolant and SCC, irradiation- "PWR Vessel plates neutron flux assisted SCC, Internals," and cyclic AMP XI.M2, loading,IASCC or "Water fatigue Chemistry"

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 3 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B4.RP-241 3.1-1, 051a Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: steel coolant and SCC, irradiation- "PWR Vessel baffle/former neutron flux assisted Internals," and assembly: baffle-to- SCCIASCC, AMP XI.M2, former bolts and fatigue, or "Water screws overload Chemistry" (for SCC mechanisms only)

M IV.B4.RP-240 3.1-1, 058a Core barrel Stainless Reactor Loss of fracture AMP XI.M16A, Yes assembly: baffle-to- steel coolant and toughness due to "PWR Vessel former bolts and neutron flux neutron irradiation Internals" screws embrittlement; loss of preload due to thermal and irradiation-enhanced stress relaxation; loss of material due to wear M IV.B4.RP-250a 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: core barrel steel coolant and irradiation-assisted "PWR Vessel cylinder (including neutron flux SCCIASCC or Internals," and vertical and fatigue AMP XI.M2, circumferential seam "Water welds); former plates Chemistry" (irradiation-assisted SCCIASCC only)

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 4 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B4.RP-244 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: external steel coolant and irradiation-assisted "PWR Vessel and internal baffle-to- neutron flux SCCIASCC, Internals," and baffle bolts and core fatigue, or AMP XI.M2, barrel-to-former bolts overload "Water Chemistry" (irradiation-assisted SCCIASCC only)

M IV.B4.RP-243 3.1-1, 058b Core barrel Stainless Reactor Loss of fracture AMP XI.M16A, Yes assembly: external steel coolant and toughness due to "PWR Vessel and internal baffle-to- neutron flux neutron irradiation Internals" baffle bolts and core embrittlement; loss barrel-to-former bolts of preload due to thermal and irradiation-enhanced stress relaxation or creep; loss of material due to wear M IV.B4.RP-240a 3.1-1, 058a Core barrel Stainless Reactor Loss of fracture AMP XI.M16A, Yes assembly: locking steel coolant and toughness due to "PWR Vessel devices (including neutron flux neutron irradiation Internals" locking welds) of embrittlement; loss baffle-to-former bolts of material due to and internal baffle-to- wear baffle bolts M IV.B4.RP-241a 3.1-1, 051a Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: locking steel coolant and SCC, irradiation- "PWR Vessel devices (including neutron flux assisted Internals," and locking welds) of SCCIASCC, AMP XI.M2, baffle-to-former bolts fatigue, or "Water and internal baffle-to- overload Chemistry" (for baffle bolts SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 5 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B4.RP-244a 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: locking steel coolant and irradiation-assisted "PWR Vessel devices (including neutron flux SCCIASCC or Internals," and locking welds) of fatigue AMP XI.M2, external baffle-to- "Water baffle bolts and core Chemistry" barrel-to-former bolts (irradiation-assisted SCCIASCC only)

M IV.B4.RP-243a 3.1-1, 058b Core barrel Stainless Reactor Loss of fracture AMP XI.M16A, Yes assembly: locking steel coolant and toughness due to "PWR Vessel devices (including neutron flux neutron irradiation Internals" locking welds) of embrittlement; loss external baffle-to- of material due to baffle bolts and core wear barrel-to-former bolts M IV.B4.RP-248 3.1-1, 051a Core support shield Stainless Reactor Cracking due to AMP XI.M16A, Yes (CSS) assembly: steel, nickel coolant and SCC or fatigue "PWR Vessel accessible upper core alloy neutron flux Internals," and barrel (UCB) bolts AMP XI.M2, and locking devices "Water Chemistry" (SCC only)

M IV.B4.RP-252 3.1-1, 058a Core support shield Stainless Reactor Loss of fracture AMP XI.M16A, Yes (CSS) Vent valve steel, coolant and toughness due to "PWR Vessel assembly: CSS vent including neutron flux thermal aging Internals" valve top and bottom CASS and embrittlement retaining rings (valve PH steels body components)

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 6 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B4.RP-252a 3.1-1, 051a Core support shield Stainless Reactor Cracking due to AMP XI.M16A, Yes (CSS) Vent valve steel coolant and SCC or fatigue "PWR Vessel assembly: CSS vent neutron flux Internals," and valve top and bottom AMP XI.M2, retaining rings; vent "Water valve locking devices Chemistry" (for (valve body SCC components) mechanisms only)

N IV.B4.RP-252b 3.1-1, 051b Vent valve assembly: CASS Reactor Cracking due to AMP XI.M16A, Yes vent valve bodies coolant and SCC or fatigue "PWR Vessel neutron flux Internals," and AMP XI.M2, "Water Chemistry" (SCC only)

N IV.B4.RP-252c 3.1-1, 058b Vent valve assembly: CASS Reactor Loss of fracture AMP XI.M16A, Yes vent valve bodies coolant and toughness due to "PWR Vessel neutron flux thermal aging Internals" embrittlement

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 7 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B4.RP-252d 3.1-1, 058a Vent valve assembly: Stainless Reactor Loss of material AMP XI.M16A, Yes original locking steel coolant and due to wear (for "PWR Vessel devices (associated neutron flux locking devices Internals" with the pressure associated with the plate, spring retainer, pressure plate, spring, U-cover, key spring and spring ring, and pin in the retainer, and U assembly) cover in the assembly);

loss of fracture toughness due to thermal aging embrittlement (for locking devices associated with the key ring and pin in the assembly)

N IV.B4.RP-252e 3.1-1, 051a Vent valve assembly: Stainless Reactor Cracking due to AMP XI.M16A, Yes original locking steel coolant and SCC or fatigue "PWR Vessel devices (associated neutron flux (fatigue only for Internals," and with the key ring, pin listed original AMP XI.M2, in the assembly); locking devices) "Water modified locking Chemistry" (for devices (associated SCC with lock cup, mechanisms jackscrew locking cup only) and bolted block in the assembly -

Oconee 1, 2, and 3 and ANO-1 only)

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 8 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation MD IV.B4.RP-400 3.1-1, 051a Core support shield Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: upper (top) steel coolant and SCC "PWR Vessel flange weld neutron flux Internals," and AMP XI.M2, "Water Chemistry" MD IV.B4.RP-401 3.1-1, 058a Core support shield Stainless Reactor Loss of fracture AMP XI.M16A, Yes assembly: upper (top) steel coolant and toughness due to "PWR Vessel flange weld neutron flux neutron irradiation Internals" embrittlement M IV.B4.RP-256a 3.1-1, 051a Flow distributor Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: flow steel, nickel coolant and fatigue "PWR Vessel distributor (FD) bolt alloy neutron flux Internals" locking devices M IV.B4.RP-256b 3.1-1, 058a Flow distributor Stainless Reactor Loss of material AMP XI.M16A, Yes assembly: flow steel, nickel coolant and due to wear; "PWR Vessel distributor (FD) bolt alloy neutron flux changes in Internals" locking devices dimensions due to distortion or void swelling or distortion M IV.B4.RP-256 3.1-1, 051a Flow distributor Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: flow steel, nickel coolant and SCC or fatigue "PWR Vessel distributor (FD) bolts alloy neutron flux Internals," and AMP XI.M2, "Water Chemistry" M IV.B4.RP-258a 3.1-1, 051a Incore Monitoring StainlessCast Reactor Cracking due to AMP XI.M16A, Yes Instrument (IMI) austenitic coolant and SCC, irradiation- "PWR Vessel guide tube assembly: stainless neutron flux assisted Internals," and IMI guide tube steel SCCIASCC, or AMP XI.M2, spiders (castings) fatigue "Water Chemistry" (SCC and irradiation-assisted SCCIASCC only)

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 9 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation M IV.B4.RP-259a 3.1-1, 051a Incore Monitoring Stainless Reactor Cracking due to AMP XI.M16A, Yes Instrument (IMI) steel coolant and SCC, irradiation- "PWR Vessel guide tube assembly: neutron flux assisted Internals," and IMI guide tube spider- SCCIASCC, or AMP XI.M2, to-lower grid rib fatigue "Water sectionssection welds Chemistry" (SCC and irradiation-assisted SCCIASCC only)

M IV.B4.RP-259 3.1-1, 058a Incore Monitoring Stainless Reactor Loss of fracture AMP XI.M16A, Yes Instrument (IMI) steel, nickel coolant and toughness due to "PWR Vessel guide tube assembly: alloy neutron flux thermal aging, Internals" IMI guide tube spider- neutron irradiation to-lower grid rib embrittlement sectionssection welds M IV.B4.RP-262 3.1-1, 051b Lower grid assembly: Nickel alloy Reactor Cracking due to AMP XI.M16A, Yes accessible alloy X- coolant and SCC "PWR Vessel 750 dowel-to-lower neutron flux Internals," and grid fuel assembly AMP XI.M2, support pad locking "Water welds (all plants Chemistry" (for except Davis Besse) SCC mechanisms only)

M IV.B4.RP-261 3.1-1, 051a Lower grid assembly: Nickel alloy Reactor Cracking due to AMP XI.M16A, Yes alloy X-750 dowel-to- coolant and SCC "PWR Vessel guide block welds (all neutron flux Internals," and plants except Davis AMP XI.M2, Besse) "Water Chemistry" MD IV.B4.RP-254b 3.1-1, 058b Lower grid assembly: Nickel Alloy Reactor Loss of material AMP XI.M16A, Yes alloy X-750 lower grid coolant and due to wear; "PWR Vessel shock pad bolt neutron flux changes in Internals" locking devices dimensions due to (Three Mile Island void swelling or Unit 1 only) distortion

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 10 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation MD IV.B4.RP-254a 3.1-1, 051b Lower grid assembly: Nickel alloy Reactor Cracking due to AMP XI.M16A, Yes alloy X-750 lower grid coolant and fatigue "PWR Vessel shock pad bolt neutron flux Internals" locking devices (Three Mile Island Unit 1 only)

MD IV.B4.RP-254 3.1-1, 051b Lower grid assembly: Nickel alloy Reactor Cracking due to AMP XI.M16A, Yes alloy X-750 lower grid coolant and SCC "PWR Vessel shock pad bolts neutron flux Internals," and (Three Mile Island AMP XI.M2, Unit 1 only) "Water Chemistry" M IV.B4.RP-246a 3.1-1, 051b Lower grid assembly: Stainless Reactor Cracking due to AMP XI.M16A, Yes upper thermal shield steel, nickel coolant and fatigue "PWR Vessel (UTS) bolt locking alloy neutron flux Internals" devices and lower thermal shield (LTS) bolt locking devices M IV.B4.RP-246b 3.1-1, 058b Lower grid assembly: Stainless Reactor Loss of material AMP XI.M16A, Yes upper thermal shield steel, nickel coolant and due to wear; "PWR Vessel (UTS) bolt locking alloy neutron flux changes in Internals" devices and lower dimensions due to thermal shield (LTS) void swelling or bolt locking devices distortion M IV.B4.RP-246 3.1-1, 051b Lower grid assembly: Stainless Reactor Cracking due to AMP XI.M16A, Yes upper thermal shield steel, nickel coolant and SCC "PWR Vessel (UTS) bolts and lower alloy neutron flux Internals," and thermal shield (LTS) AMP XI.M2, bolts "Water Chemistry" N IV.B4.RP-246c 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: steel; nickel coolant and SCC "PWR Vessel upper thermal shield alloy neutron flux Internals," and (UTS) bolts AMP XI.M2, Water Chemistry

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 11 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B4.RP-246d 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: steel; nickel coolant and fatigue "PWR Vessel upper thermal shield alloy neutron flux Internals" (UTS) bolt locking devices N IV.B4.RP-246e 3.1-1, 058b Core barrel Stainless Reactor Loss of material AMP XI.M16A, Yes assembly: steel; nickel coolant and due to wear; "PWR Vessel upper thermal shield alloy neutron flux changes in Internals" (UTS) bolt locking dimension due to devices void swelling or distortion M IV.B4.RP-260 3.1-1, 058b Lower grid fuel Stainless Reactor Loss of fracture AMP XI.M16A, Yes assembly: (a) steel, nickel coolant and toughness due to "PWR Vessel accessible pads; (b) alloy neutron flux neutron irradiation Internals" accessible, pad-to-rib embrittlement section welds; (c) accessible alloy X-750, dowels, cap screws and their locking devices M IV.B4.RP-260a 3.1-1, 051b Lower grid fuel Stainless Reactor Cracking due to AMP XI.M16A, Yes assembly: (a) pads; steel, nickel coolant and SCC or fatigue "PWR Vessel (b), pad-to-rib section alloy neutron flux Internals," and welds; (c) alloy AMP XI.M2, X-750, dowels, cap "Water screws and their Chemistry" (for locking devices SCC mechanisms only)

M IV.B4.RP-251a 3.1-1, 058a Plenum cover Stainless Reactor Loss of material AMP XI.M16A, Yes assembly: plenum steel coolant and due to wear; loss "PWR Vessel cover weldment rib neutron flux of preload (wear) Internals" pads and, plenum cover support flange, plenum cover support ring

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 12 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B4.R-423 3.1-1, 118 Reactor vessel Stainless Reactor Cracking due to Plant-specific Yes internal components, steel, nickel coolant, SCC, irradiation- aging or reactor vessel alloy neutron flux assisted management internal component- SCCIASCC, cyclic programPlant-specific basis for a loading, fatigue specific aging specified RVI management component program, or AMP XI.M16A, "PWR Vessel Internals," and AMP XI.M2, "Water Chemistry" (SCC and IASCC only),

for cases where a specified component is subject to a site-specific or component-specific aging management basis

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 13 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B4.R-424 3.1-1, 119 Reactor vessel Stainless Reactor Loss of fracture Plant-specific Yes internal components, steel, nickel coolant, toughness due to aging or reactor vessel alloy neutron flux neutron irradiation management internal component- embrittlement or program, or specific basis for a thermal aging AMP XI.M16A, specified RVI embrittlement; PWR Vessel component changes in Internals, for dimensions due to cases where a void swelling or specified distortion; loss of component is preload due to subject to a thermal and site-specific or irradiation- component-enhanced stress specific aging relaxation or management creep; loss of basis material due to wear MD IV.B4.RP-382 3.1-1, 032 Reactor vessel Stainless Reactor Cracking due to AMP XI.M1, No internals: ASME steel, nickel coolant and fatigue, SCC, or "ASME Section Section XI, alloy, cast neutron flux irradiation-assisted XI Inservice Examination austenitic SCC; loss of Inspection, Category B-N-3 core stainless material due to Subsections support structure steel wear IWB, IWC, and components (not IWD" already identified as "Existing Programs" components in MRP-227-A)

M IV.B4.RP-352 3.1-1, 051b Upper grid assembly: Nickel alloy Reactor Cracking due to AMP XI.M16A, Yes alloy X-750 dowel-to- coolant and SCC "PWR Vessel upper grid fuel neutron flux Internals," and assembly support AMP XI.M2, pad welds (all plants "Water except Davis-Besse) Chemistry" (for SCC mechanisms only)

SLR-ISG-PWRVI-2020-XX: Appendix B.3 Page 14 of 16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table B4 Reactor Vessel Internals (PWR)Babcock & Wilcox New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.B4.RP-386 3.1-1, 051b Lower Grid Stainless Reactor Cracking due to AMP XI.M16A, Yes Assembly: lower grid steel coolant and SCC or fatigue "PWR Vessel rib section neutron flux Internals," and AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

N IV.B4.RP-386a 3.1-1, 058b Lower Grid Stainless Reactor Loss of fracture AMP XI.M16A, Yes Assembly: lower grid steel coolant and toughness due to "PWR Vessel rib section neutron flux neutron irradiation Internals," and embrittlement AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

APPENDIX B.4 PROPOSED REVISIONS TO GALL-SLR REPORT TABLE IV.E, COMMON MISCELLANEOUS MATERIAL/ENVIRONMENT COMBINATIONS NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Table IV.E, Common Miscellaneous Material/Environment Combinations, addresses miscellaneous material/environment combinations that may be found throughout the reactor vessel, internals, and reactor coolant systems, structures, and components.

Proposed revisions to Table IV.E of the GALL-SLR Report are provided in redline format. This AMR item supersedes the respective item in GALL-SLR Report, Revision 0, Table IV.E.

SLR-ISG-PWRVI-2020-XX: Appendix B.4 Page 2 of 2 GALL-SLR Report Table IV.E Proposed Revisions IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Table E Common Miscellaneous Material/Environment Combinations New, Aging Modified, Structure Management Deleted, SRP Item and/or Aging Program Further Edited Item Item (Table, ID) Component Material Environment Effect/Mechanism (AMP)/TLAA Evaluation N IV.E.R-444 3.1-1, 114 Reactor coolant system Any Applicable Cracking due to AMP XI.M1, No components: Components internal or SCC, IGSCC "ASME defined as ASME Section XI external (stainless steel or Section XI components (e.g., ASME environment nickel alloy Inservice Code Class 1 reactor components only), Inspection, coolant pressure boundary cyclic loading; loss Subsections components, reactor interior of material due to IWB, IWC, attachments, or core general corrosion and IWD," and support structure (steel only), pitting AMP XI.M2, components, ASME Class 2 corrosion, crevice "Water or 3 components, including corrosion, wear Chemistry" associated pressure- (water retaining welds) not chemistry-managed by other AMR line related or items in GALL-SLR corrosion-Chapter IV related aging effect mechanisms only)

APPENDIX C PROPOSED REVISIONS TO SRP-SLR SECTION 3.1.2.2.9 (AMR FURTHER EVALUATION ACCEPTANCE CRITERIA) AND SRP-SLR SECTION 3.1.3.2.9 (AMR FURTHER EVALUATION REVIEW PROCEDURES)

NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), Sections 3.1.2.2.9 and 3.1.3.2.9, provide staff guidance for the acceptance criteria and review procedures, respectively, for the further evaluation item related to aging management of pressurized-water reactor vessel internals. These sections are reproduced below in their entirety with revisions provided in redline format, and supersede SRP-SLR, Revision 0, Sections 3.1.2.2.9 and 3.1.3.2.9.

SRP-SLR Further Evaluation Proposed Revisions 3.1.2.2.9 Aging Management of Pressurized Water Reactor Vessel Internals (Applicable to Subsequent License Renewal Periods Only)

Electric Power Research Institute (EPRI) Topical Report (TR)-1022863, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML12017A191 through ML12017A197 and ML12017A199), provides provided the industrys current aging managementinitial set of aging management inspection and evaluation (I&E) recommendations for the reactor vessel internal (RVI) components that are included in the design of a PWR facility. Since the issuance of MRP-227-A on January 9, 2012, EPRI updated its I&E guidelines for the PWR RVI components in Topical Report No. 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A) (ADAMS Accession No. ML19339G350). MRP-227, Revision 1-A, incorporated the industrys bases for resolving operating experience and industry lessons learned resulting from component-specific inspections performed since the issuance of MRP-227-A in January 2012. The staff found the guidelines in MRP-227, Revision 1-A, acceptable, as documented in a staff-issued safety evaluation dated April 25, 2019 (ADAMS Accession No. ML19081A001) and approved the topical report for use as documented in the staffs letters to the EPRI Materials Reliability Program (MRP) dated February 19, 2020 and July 7, 2020 (ADAMS Accession Nos. ML20006D152 and ML20175A149).

In this report MRP-227, Revision 1-A, the EPRI Materials Reliability Program (MRP) identified that the following aging mechanisms may be applicable to the design of the RVI components in these types of facilities: (a) stress corrosion cracking (SCC), (b) irradiation-assisted stress corrosion cracking (IASCC), (c) fatigue, (d) wear, (e) neutron irradiation embrittlement, (f) thermal aging embrittlement, (g) void swelling and irradiation growth or component distortion, or (h) thermal or irradiation-enhanced stress relaxation or irradiation enhanced creep. The methodology in MRP-227-A was approved by the NRC in a safety evaluation dated December 16, 2011 (ADAMS Accession No. ML11308A770), which includes those plant-specific applicant/licensee action items that a licensee or applicant applying the MRP-227-A report would need to address and resolve and apply to its licensing basis.

The EPRI MRPs functionality analysis and failure modes, effects, and criticality analysis bases for grouping Westinghouse-designed, B&W-designed and Combustion Engineering (CE)-designed RVI components into these the applicable inspection categories (as evaluated in MRP-227, Revision 1-A) was were based on an assessment of aging effects and relevant

SLR-ISG-PWRVI-2020-XX: Appendix C Page 2 of 4 time-dependent aging parameters through a cumulative 60-year licensing period (i.e., 40 years for the initial operating license period plus an additional 20 years during the initial period of extended operation). The EPRI MRPs has not assessedassessment in MRP-227, Revision 1-A, did not evaluate whether operation of Westinghouse-designed, B&W-designed and CE-designed reactors during an SLR operating period (60 to 80 years) would have any impact on the existing susceptibility rankings and inspection categorizations for the RVI components in these designs, as defined in MRP-227, Revision 1-A or its the applicable MRP background documents (e.g., MRP-191, Revision 1, for Westinghouse-designed or CE-designed RVI components or MRP-189, Revision 2, for B&W-designed components).

As described in GALL-SLR Report AMP XI.M16A, the applicant may use the MRP-227, Revision 1-A based AMP as an initial reference basis for developing and defining the AMP that will be applied to the RVI components for the subsequent period of extended operation.

However, to use this alternative basis, GALL-SLR Report AMP XI.M16A recommends that the MRP-227, Revision 1-A based AMP be enhanced to include a gap analysis of the components that are within the scope of the AMP. The gap analysis is a basis for identifying and justifying any potential changes to the MRP-227, Revision 1-A based program that may beare necessary to provide reasonable assurance that the effects of age-related degradation will be managed during the subsequent period of extended operation. The criteria for the gap analysis are described in GALL-SLR Report AMP XI.M16A. If a gap analysis is needed to establish the appropriate aging management criteria for the RVI components, the applicant has the option of including the gap analysis in the SLRA for its reactor unit(s) or making the gap analysis and any supporting gap analysis documents available in the in-office audit portal for the SLRA review.

Subsequent license renewal (SLR) applicants for units of a PWR design will no longer need to include separate SLRA Appendix C section responses in resolution of the A/LAIs previously issued on MRP-227-A because the A/LAIs were resolved and closed by the staff in the April 25, 2019, safety evaluation for MRP-227, Revision 1-A. The sole A/LAI issued by the staff in the safety evaluation dated April 25, 2019, relates to an applicants methods and timing of inspections that will be applied to the baffle-to-former bolts or core shroud bolts in the plant design. Since an applicants resolution of this A/LAI can be appropriately addressed in the Operating Experience program element discussion for the AMP and in the applicants basis document for the AMP, a separate SLRA Appendix C response for the A/LAI is unnecessary.

Alternatively, the PWR SLRA may define a plant-specific AMP for the RVI components to demonstrate that the RVI components will be managed in accordance with the requirements of 10 CFR 54.21(a)(3) during the proposed subsequent period of extended operation.

Components to be inspected, parameters monitored, monitoring methods, inspection sample size, frequencies, expansion criteria, and acceptance criteria are justified in the SLRA. The If the AMP is a plant-specific program, the NRC staff will assess the adequacy of the plant-specific AMP against the criteria for the 10 AMP program elements that are defined in Section A.1.2.3 of SRP-SLR Appendix A.1.

3.1.3.2.9 Aging Management of Pressurized Water Reactor Vessel Internals (Applicable to Subsequent License Renewal Periods Only)

EPRI TR-1022863Topical Report No. 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A)

(ADAMS Accession Nos. ML19339G350ML12017A191 through ML12017A197 and ML12017A199), provides the industrys current updated aging management recommendations for the RVI components that are included in the design of a PWR facility, based on an analysis

SLR-ISG-PWRVI-2020-XX: Appendix C Page 3 of 4 of plant operation for 60 years. The review procedures in this section are based on the staffs assumption that a PWR SLR applicants PWR vessel internals AMP will be based on the I&E guidelines in MRP-227, Revision 1-A for the AMP that will be applied and implemented during the subsequent period of extended operation. The rationale for this assumption is based on the MRP-defined Needed Requirement in Section 7.3 of MRP-227, Revision 1-A, which states that the update of MRP-based program shall be implemented by January 1, 2022.

In this reportMRP-227, Revision 1-A, the EPRI MRP identified that the following aging mechanisms may be applicable to the design of the RVI components in these types of facilities:

(a) stress corrosion cracking (SCC), (b) irradiation-assisted stress corrosion cracking (IASCC),

(c) fatigue, (d) wear, (e) neutron irradiation embrittlement, (f) thermal aging embrittlement, (g) void swelling and irradiation growth or distortion, or (h) thermal or irradiation-enhanced stress relaxation or irradiation enhanced creep. The methodology in The staff approved MRP-227, Revision 1-A was approved by the NRC in a safety evaluation dated December 16, 2011April 25, 2019 (ADAMS Accession No. ML11308A770ML19081A001)., which includes In that safety evaluation, the staff resolved and closed all those plant-specific applicant/licensee action items (A/LAIs) that were previously issued on the previous version of the I&E guidelines (i.e., a licensee or applicant applyingthose in the MRP-227-A report). The assessments of RVI components in the MRP-227, Revision 1-A, report and the MRP-defined background reports for MRP-227, Revision 1-A have not been updated based on an assessment of aging effects over an 80-year operating period.

If a plant-specific AMP is proposed for the RVI components, the reviewer evaluates the adequacy of the applicants AMP on a case-by-case basis against the criteria for plant-specific AMP program elements defined in Sections A.1.2.3.1 through A.1.2.3.10 of SRP-SLR Appendix A.1. The reviewer verifies that the applicant has defined both the type of performance monitoring, condition monitoring, preventative monitoring, or mitigative monitoring AMP activities that will be used for aging management of the RVI components and the specific program element criteria for the AMP that will be used to manage age-related effects in the RVI components during the subsequent period of extended operation.

If a PWR applicant for SLR proposes to use GALL-SLR Report AMP XI.M16A, PWR Vessel Internals, as the basis for aging management, the staff reviews the program elements of the AMP against the program element criteria defined in AMP XI.M16A. The staff verifies that the applicant has addressed the relevancy of the A/LAI for MRP-227, Revision 1-A in the Operating Experience program element of the AMP, or in the applicants technical basis document or procedure for the AMP. The staff also verifies that the proposed program includes a gap analysis that provides the identification and justification of:

• Components that screen in for additional aging effects or mechanisms when assessed for aging through the end of the subsequent period of extended operation

• Components that previously screened in for an aging effect or mechanism and the severity of that aging effect or mechanism could significantly increase during the subsequent period of extended operation

• Changes to the existing MRP-227, Revision 1-A program characteristics or criteria, including, but not limited to, changes in inspection categories, inspection criteria, or primary-to-expansion component criteria and relationships

SLR-ISG-PWRVI-2020-XX: Appendix C Page 4 of 4 The If a gap analysis is needed to establish the appropriate aging management criteria for the RVI components, the staff evaluates the adequacy and justification of the gap analysis in the safety evaluation report for the SLRA. , sSpecifically, the staffs review should focus on the following aspects of the gap analysis:

• The gap analysis methodology

• The components that screened in for additional aging effects or mechanisms when assessed for aging through the end of the subsequent period of extended operation

• The components for which a previously screened in aging effect or mechanism has been identified as potentially more severe during the subsequent period of extended operation

• Components whose AMP inspection categories have changed from those previously identified for the components in MRP-227, Revision 1-A

• Proposed changes to the aging management program characteristics or criteria identified in the SLRA For those RVI components that screened in for additional aging effects or mechanisms, or that are subject to site-specific or component-specific changes in the EPRI MRPs I&E protocols for the components, the staff also confirms that the applicant has included and justified appropriate AMR line items for the components. The applicant may use the updated version of GALL-SLR Report Item IV.B2.R-423, IV.B3.R-423, or IV.B4.R-423 to address any RVI component for which the EPRI MRP I&E protocols for managing cracking or specific cracking mechanisms in the component are being updated or adjusted on a site-specific or component-specific basis. The applicant may use the updated version of GALL-SLR Report Items IV.B2.R-424, IV.B3.R-424, or IV.B4.R-424 to address any RVI component for which the EPRI MRP I&E protocols for managing non-cracking effects or mechanisms in the component are being updated or adjusted on a site-specific or component-specific basis.

Otherwise an applicant may use an NRC-approved generic methodologyreport such as an approved revision of MRP-227 that considers an operating period of 80 years. In this case, the staff reviews any responses to action items on the aging management methods that may be identified in the NRC approval of the generic methodologyreport.

APPENDIX D PROPOSED REVISIONS TO GALL-SLR REPORT AMP XI.M16A, PWR VESSEL INTERNALS, AND RELATED FSAR SUPPLEMENT EXAMPLE IN GALL-SLR REPORT TABLE XI-01 NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Aging Management Program (AMP) XI.M16A, PWR Vessel Internals, describes one acceptable way to manage aging effects related to pressurized-water reactor (PWR) vessel internals for subsequent license renewal. This AMP is reproduced below in its entirety, with revisions provided in redline format. It supersedes GALL-SLR Report, Revision 0, AMP XI.M16A.

This appendix also provides a redline version of the AMP XI.M16A final safety analysis report (FSAR) supplement summary in GALL-SLR Report Table XI-01, FSAR Supplement Summaries for GALL-SLR Report Chapter XI Aging Management Programs. This entry modifies GALL-SLR Report, Revision 0, Table XI-01.

GALL-SLR Report Aging Management Program XI.M16A Proposed Revisions XI.M16A PWR VESSEL INTERNALS Program Description This program is used to manage the effects of age-related degradation mechanisms that are applicable to the pressurized water reactor (PWR) reactor vessel internal (RVI) components.

These aging effects include: (a) cracking, including stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and cracking due to fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to thermal aging and neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

In the absence of an acceptable generic methodology report such as an approved revision of Materials Reliability Program (MRP)-227 that considers an operating period of 80 years, this program may be based on an existing plant program that is consistent with Electric Power Research Institute (EPRI) Technical Topical Report No. 30020171681022863, Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines, (MRP-227, Revision 1-A), which is implemented in accordance with Nuclear Energy Institute (NEI) 03-08, Guideline for the Management of Materials Issues. The staff approved found the augmented updated inspection and evaluation (I&E) guidelines and criteria for PWR RVI components acceptable, as documented in the staffs safety evaluation of April 25, 2019 (ADAMS Accession No. ML19081A001), and approved the use of MRP-227, Revision 1-A, for PWR-specific design bases in the staffs letters to the EPRI MRP dated February 19, 2020 and July 7, 2020 (ADAMS Accession Nos. ML20006D152 and ML20175A149)NRC Safety Evaluation (SE), Revision 1, on MRP-227 by letter dated December 16, 2011.

Because the guidelines of MRP-227, Revision 1-A, are based on an analysis of the RVI that considers the operating conditions up to a 60-year operating period, these guidelines are supplemented through a gap analysis that identifies enhancements to the program that are

SLR-ISG-PWRVI-2020-XX: Appendix D Page 2 of 10 needed to address an 80-year operating period. In this program, the term MRP-227-A (as supplemented) is used to describe either MRP-227, Revision 1-A, as supplemented by this gap analysis, or an acceptable generic methodology report such as an approved revision of MRP-227 that considers an operating period of 80 years.

The program applies the guidance in MRP-227-A (as supplemented) for inspecting, evaluating, and, if applicable, dispositioning non-conforming RVI components at the facility. These examinations provide reasonable assurance that the effects of age-related degradation mechanisms will be managed during the period of extended operation. The program includes expanding periodic examinations and other inspections, if the extent of the degradation identified exceeds the expected levels.

The methodology used in the development of MRP-227, Revision 1-A, guidance for selecting RVI components for inclusion in the inspection sample is based on a four-step ranking process.

Through this process, the RVIs for all threeWestinghouse and Combustion Engineering (CE)

PWR designs were assigned to one of the following four groupsinspection categories:

Primary, Expansion, Existing Programs, and or No Additional Measures. Through this process, the RVIs for Babcock & Wilcox (B&W) PWR designs were assigned to one of the following three inspection categories: Primary, Expansion, or No Additional Measures.

Definitions of each group category are provided in MRP-227, Revision 1-A.

In the absence of an acceptable generic methodology such as an approved revision of MRP-227 that considers an operating period of 80 years, the gap analysis described below is used to provide reasonable assurance that the aging management for the RVI components identified in the four groups is appropriate for 80 years of operation.

The result of this four-step sample selection process is a set of Primary internals component locations for each of the three plant designs that are inspected because they are expected to show the leading indications of the degradation effects., with another set The category of Expansion internals component locations that are is specified to expand the sample should the indications from the Primary components be more severe than anticipated.

The degradation effects in a third set of internals locations (which apply only to the RVI components in Westinghouse- or CE-designed PWRs) are deemed to be adequately managed by Existing Programs, such as American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI,1 Examination Category B-N-3, examinations of core support structures. A fourth set of internals locations are deemed to require No Additional Measures.

In the absence of an acceptable generic report such as an approved revision of MRP-227 that considers an operating period of 80 years, the gap analysis described below is used to provide reasonable assurance that the aging management activities designated for the RVI components identified in the four groups is appropriate for 80 years of operation. The gap analysis may include and incorporate supplemental guidelines developed and recommended for the RVI components.

If the program is based on MRP-227, Revision 1-A, with a gap analysis, the inspection categories, inspection criteria, and other program characteristics required byestablished in 1

GALL-SLR Report Chapter I, Table 1, identifies the ASME Code Section XI editions and addenda that are acceptable to use for this AMP.

SLR-ISG-PWRVI-2020-XX: Appendix D Page 3 of 10 MRP-227, Revision 1-A, are identified and justified for each component in the applicable program elements. The justification should focus on the aging management of the any additional aging considerations (i.e., new aging effect/mechanism) during the subsequent period of extended operation. The acceptance criteria in the Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR),

Section 3.1.2.2.9 and the review procedures in Section 3.1.3.2.9 provide additional information.

Evaluation and Technical Basis

1. Scope of Program: The scope of the program includes all RVI components based on the plants applicable nuclear steam supply system design. The scope of the program applies the methodology and guidanceguidelines in MRP-227-A (as supplemented),

which provides an augmented inspection and flaw evaluation methodology guidelines for assuring the functional integrity of safety-related internals in commercial operating U.S.

PWR nuclear power plants designed by Babcock & Wilcox (B&W), Combustion Engineering (CE), and Westinghouse. Since these types of AMPs are considered to be living programs by the licensees implementing the programs, the scope of program may also include additional reports, documents or guidelines recommended for implementation by the EPRI MRP, PWR Owners Group, or industry vendors. This may include (but is not limited to) applicable WCAP or BAW technical/topical reports issued by Westinghouse or B&W, or supplemental guidelines or industry alert letters issued by the EPRI MRP, PWR Owners Group, or industry vendors.

The scope of components includes core support structures, those RVI components that serve an intended license renewal safety function pursuant to criteria in Title 10 of the Code of Federal Regulations (10 CFR) 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). In addition, ASME Code,Section XI includes inspection requirements for PWR removable core support structures in Table IWB-2500-1, Examination Category B-N-3, which are in addition to any inspections that are implemented in accordance with MRP-227-A (as supplemented).

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation. The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are managed in accordance with an applicants AMP that corresponds to GALL-SLR Report AMP XI.M1, ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.

This program element specifies if whether the program is based on an existing program that is consistent with MRP-227, Revision 1-A, with a gap analysis, or if itthe program is based on an acceptable generic methodology report that covers an 80-year service life for the RVI components, such as an approved revision of MRP-227 that considers an operating period of 80 years. If based on MRP-227, Revision 1-A, with a gap analysis, the scope of the program focuses on identification and justification of the following:

a. Components that screen in for additional aging effects or mechanisms when assessed for the 60-80 year operating period.

SLR-ISG-PWRVI-2020-XX: Appendix D Page 4 of 10

b. Components that previously screened in for an aging effect or mechanism and the severity of that aging effect or mechanism could significantly increase for the 60-80 year operating period.

c. Changes to the existing MRP-227, Revision 1-A, program characteristics or criteria, including but not limited to changes in inspection categories, inspection criteria, or primary-to-expansion component criteria and relationships.

2. Preventive Actions: The program relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms [e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms (SCC, PWSCC, or IASCC)]. Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program, as described in GALL-SLR Report AMP XI.M2, Water Chemistry.

3. Parameters Monitored or Inspected: The program manages the following age-related degradation effects and mechanisms that are applicable in general to RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by thermal aging and neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

For the management of cracking, the program monitors for evidence of surface-breaking linear discontinuities if a visual inspection technique is used as the non-destructive examination (NDE) method or for relevant flaw presentation signals if a volumetric ultrasonic testing (UT) method is used as the NDE method. For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components. For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement. Instead, the impact of loss of fracture toughness on component integrity is indirectly managed by: (1) using visual or volumetric examination techniques to monitor for cracking in the components, and (2) applying applicable reduced fracture toughness properties in the flaw evaluations, in cases where cracking is detected in the components and is extensive enough to necessitate a supplemental flaw growth or flaw tolerance evaluation. The program uses physical measurements to monitor for any dimensional changes due to void swelling or distortion.

Specifically, the program implements the parameters monitored/inspected criteria consistent with the applicable tables in Section 4, Aging Management Requirements, in MRP-227-A (as supplemented).

4. Detection of Aging Effects: The inspection methods are defined and established in Section 4 of MRP-227-A (as supplemented). Standards for implementing the inspection methods are defined and established in MRP-228. In all cases, well-established inspection methods are selected. These methods include volumetric UT examination methods for detecting flaws in bolting and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an

SLR-ISG-PWRVI-2020-XX: Appendix D Page 5 of 10 alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). VT-3 visual methods may be applied for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component, as evaluated for reduced fracture toughness properties, is known and the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. VT-3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g., redundant bolts or pins used to secure a fastened RVI assembly).

In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss of preload caused by thermal and irradiation- enhanced stress relaxation and creep.

The program adopts the guidance in MRP-227-A (as supplemented) for defining the Expansion Criteria that need to be applied to the inspection findings of Primary components and for expanding the examinations to include additional Expansion components. RVI component inspections are performed consistent with the inspection frequency and sampling bases for Primary components, Existing Programs components, and Expansion components in MRP-227-A (as supplemented).

In some cases (as defined in MRP-227, Revision 1-A), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimensions due to void swelling or distortion.

Inspection coverages for Primary and Expansion RVI components are implemented consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, onthose established in MRP-227, Revision 1-A, or as modified by a gap analysis.

This program element should justify the appropriateness of the inspection methods, sample size criteria, and inspection frequency criteria for managing the effects of degradation during the subsequent period of extended operation, including any changes to these criteria from their prior assessment in MRP-227, Revision 1-A.

5. Monitoring and Trending: The methods for monitoring, recording, evaluating, and trending the data that result from the programs inspections are given in Section 6 of MRP-227-A (as supplemented) and its subsections. Component reinspection frequencies for Primary and Expansion category components are defined in specific tables in Section 4 of the MRP-227-A report (as supplemented). The examination and re-examinations that are implemented in accordance with MRP-227-A (as supplemented), together with the criteria specified in MRP-228 for inspection methodologiesstandards, inspection procedures, and inspection personnel, provide for timely detection, reporting, and implementation of corrective actions for the aging effects and mechanisms managed by the program.

The program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in

SLR-ISG-PWRVI-2020-XX: Appendix D Page 6 of 10 cases where cracking is detected in an RVI component and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program includes criteria to evaluate the aging effects in the inaccessible portions of the components and the resulting impact on the intended function(s) of the components. For redundant components (such as redundant bolts, screws, pins, keys, or fasteners, some of which are accessible to inspection and some of which are not accessible to inspection), the program includes criteria to evaluate the aging effects in the population of components that are inaccessible by the applicable inspection technique and the resulting impact on the intended function(s) of the assembly containing the components.

Flaw evaluation methods, including recommendations for flaw depth sizing and for crack growth determinations as well as for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications, are defined in MRP-227-A (as supplemented).

6. Acceptance Criteria: Section 5 of MRP-227-A (as supplemented), which includes Table 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-designed RVIs, provides the specific examination and flaw evaluation acceptance criteria for the Primary and Expansion RVI component examination methods. Consistent with the criteria in MRP-227, Revision 1-A, the acceptance criteria for some Expansion category components may be established through performance of a component-specific analysis or component replacements, particularly if the components are inaccessible for inspection or the industry has yet to develop adequate inspection methods for the components. For RVI components addressed by examinations performed in accordance with the ASME Code,Section XI, the acceptance criteria in IWB-3500 are applicable. For RVI components covered by other Existing Programs, the acceptance criteria are described within the applicable reference document. As applicable, the program establishes acceptance criteria for any physical measurement monitoring methods that are credited for aging management of particular RVI components.

This program element should justify the appropriateness of the acceptance criteria for managing the effects of degradation during the subsequent period of extended operation, including any changes to acceptance criteria based on the gap analysis.

7. Corrective Actions: Results that do not meet the acceptance criteria are addressed in the applicants corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, Corrective Action, of 10 CFR Part 50, Appendix B. Appendix A of the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the corrective actions element of this AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events.

SLR-ISG-PWRVI-2020-XX: Appendix D Page 7 of 10 The implementation of the guidance in MRP-227-A (as supplemented), plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Other alternative corrective actions bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Alternative corrective actions not approved or endorsed by the NRC will be submitted for NRC approval prior to their implementation.

8. Confirmation Process: The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, Corrective Action, of 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the recommendations of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable. The implementation of the guidance in Section 7 of MRP-227-A (as supplemented), in conjunction with NEI 03-08 and other guidance documents, reports, or methodologies guidelines referenced in this AMP, provides an acceptable level of quality and an acceptable basis for confirming the quality of inspections, flaw evaluations, and corrective actions.

9. Administrative Controls: Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR Part 50, Appendix B, associated with managing the effects of aging. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

The administrative controls for these types of programs, including their implementing procedures and review and approval processes, are implemented in accordance with the recommended industry guidelines and criteria in NEI 03-08, and are under existing site 10 CFR 50 Appendix B, Quality Assurance Programs, or their equivalent, as applicable.

The basis defined in Section 7 of MRP-227, Revision 1-A, found acceptable as documented in the staffs safety evaluation dated April 25, 2019, provides the basis for implementing the program in accordance with NEI 03-08. Administrative activities for keeping the program implementation procedures up to date with the various industry reports within the scope of the AMP (e.g.,MRP-227, Revision 1-A) fall within the scope of this Administrative Controls program element. The evaluation in Section 3.5 of the NRCs SE, Revision 1, on MRP-227-A provides the basis for endorsing NEI 03-08. This includes endorsement of the criteria in NEI 03-08 for notifying the NRC of any deviation from the I&E methodology in MRP-227-A and justifying the deviation no later than 45 days after its approval by a licensee executive.

10. Operating Experience: The review and assessment of relevant operating experience (OE) for its impacts on the program, including implementing procedures, are governed by NEI 03-08 and Appendix A of MRP-227, Revision 1-A. Consistent with MRP-227,

SLR-ISG-PWRVI-2020-XX: Appendix D Page 8 of 10 Revision 1-A, the reporting of inspection results and OE is treated as a Needed category item under the implementation of NEI 03-08.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry OE including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

References 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants. Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR Part 50.55a, Codes and Standards. Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code,Section V, Nondestructive Examination. 2004 Edition 2. New York, New York: American Society of Mechanical Engineers.

_____. ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. New York, New York: American Society of Mechanical Engineers. 2008.

EPRI. EPRI Topical Report No. 1016596, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 0). Palo Alto, California: Electric Power Research Institute. 2008.

_____. EPRI Technical Topical Report No. 1022863, Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A).

Agencywide Documents Access and Management System (ADAMS) Accession No. ML12017A193 (Transmittal letter from the EPRI-MRP) and ADAMS Accession Nos.

ML12017A194, ML12017A196, ML12017A197, ML12017A191, ML12017A192, ML12017A195 and ML12017A199, (Final Report). Palo Alto, California: Electric Power Research Institute. December 2011.

_____. EPRI 1016609, Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228). (Non-publicly available ADAMS Accession No. ML092120574). The non-proprietary version of the report may be accessed by members of the public at ADAMS Accession No. ML092750569. Palo Alto, California: Electric Power Research Institute.

July 2009.

_____. EPRI Topical Report 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A). ADAMS Accession No. ML19339G350. Palo Alto, California: Electric Power Research Institute.

December 2019.

2 GALL-SLR Report Chapter I, Table 1, identifies the ASME Code Section XI editions and addenda that are acceptable to use for this AMP.

SLR-ISG-PWRVI-2020-XX: Appendix D Page 9 of 10 NEI. NEI 03-08, Revision 23, Guideline for the Management of Materials Issues. ADAMS Accession No. ML19079A253ML101050337. Washington, DC: Nuclear Energy Institute.

January 2010February 2017.

NRC. License Renewal Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors. ADAMS Accession No. ML12270A436. Washington, DC: U.S. Nuclear Regulatory Commission.

June 3, 2013.

_____. License Renewal Interim Staff Guidance LR-ISG-2011-05, Ongoing Review of Operating Experience. ADAMS Accession No. ML12044A215. Washington, DC: U.S. Nuclear Regulatory Commission. March 16, 2012.

_____. Safety Evaluation from Robert A. Nelson (NRC) to Neil Wilmshurst (EPRI), Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. ADAMS Accession No. ML11308A770.

Washington, DC: U.S. Nuclear Regulatory Commission. December 16, 2011.

_____. Final Safety Evaluation for Electric Power Research Institute Topical Report MRP-227, Revision 1, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guideline. ADAMS Accession No. ML19081A001. Washington, D.C: U.S. Nuclear Regulatory Commission. April 25, 2019.

_____. Letter from Joe Holonich (NRC) to Brian Burgos (EPRI), U.S. Nuclear Regulatory Commission Verification Letter for Electric Power Research Institute Topical Report MRP-227, Revision 1, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline. ADAMS Accession No. ML20006D152. Washington, D.C: U.S.

Nuclear Regulatory Commission. February 19, 2020.

_____. Email from Joe Holonich (NRC) to Kyle Amberge (EPRI), Transmittal Email MRP 227, Rev 1-A Supplemental Information -A Verification. ADAMS Accession No. ML20175A149.

Washington, D.C: U.S. Nuclear Regulatory Commission. July 7, 2020.

SLR-ISG-PWRVI-2020-XX: Appendix D Page 10 of 10 GALL-SLR Report Table XI-01 Proposed Revisions Table XI-01. FSAR Supplement Summaries for GALL-SLR Report Chapter XI Aging Management Programs AMP GALL-SLR Description of Program Implementation Program Schedule XI.M16A PWR Vessel The program relies on implementation of Program, accounting Internals the inspection and evaluation guidelines for the impacts of a gap in EPRI Technical Report No. 1022863 analysis, is 3002017168 (MRP-227, Revision 1-A) implemented 6 months and EPRI Technical Report No. 1016609 prior to the subsequent (MRP-228) to manage the aging effects period of extended on the reactor vessel internal operation, or components, as supplemented by a gap alternatively, a plant-analysis that identifies enhancements to specific program may the program that are needed to address be implemented an 80-year operating period. 6 months prior to the subsequent period to Alternatively, the program relies on extended operation.

implementation of an acceptable generic report such as an approved revision of MRP-227 that considers an operating period of 80 years This program is used to manage (a) cracking, including stress corrosion cracking, primary water stress corrosion cracking, irradiation-assisted stress corrosion cracking, and cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging, neutron irradiation embrittlement, or void swelling; (d) dimensional changes due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation enhanced stress relaxation or creep.

[The applicant is to provide additional details to describe the gap analysis associated with the AMP.]

APPENDIX E PROPOSED REVISION TO GALL-SLR REPORT TABLE IX.C, USE OF TERMS FOR MATERIALS NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, Table IX.C, Use of Terms for Materials, defines many generalized materials used in the aging management review tables in Chapters II through VIII of the GALL-SLR Report. The table below adds the term stellite and its usage to Table IX.C.

GALL-SLR Report Table IX.C Revisions IX.C Use of Terms for Materials Term Usage in this document Stellite ASTM International provides a technical definition of stellite in ASTM MNL46, Metallographic and Materialographic Specimen Preparation, Light Microscopy, Image Analysis and Hardness Testing:

Stellite is a special cobalt-based alloy with 46-65 % Co, 25-25 % Cr, and 5-20 % W. The material is very wear resistant

APPENDIX F PROPOSED REVISIONS TO SRP-SLR TABLE 4.7-1, EXAMPLES OF POTENTIAL PLANT-SPECIFIC TLAA TOPICS NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants (SRP-SLR), Table 4.7-1, Examples of Potential Plant-Specific TLAA Topics, provides examples of potential plant-specific time-limited aging analyses (TLAAs) that license renewal applicants have identified. This table is reproduced below in its entirety, with changes provided in redline format. This table supersedes SRP-SLR, Revision 0, Table 4.7-1.

SRP-SLR Table 4.7-1 Proposed Revisions Table 4.7-1 Examples of Potential Plant-Specific TLAA Topics BWRs Re-flood thermal shock of the reactor pressure vessel Re-flood thermal shock of the core shroud and other reactor vessel internals Loss of preload for core plate rim hold-down bolts Erosion of the main steam line flow restrictors Susceptibility to irradiation-assisted stress corrosion cracking PWRs Reactor pressure vessel underclad cracking Leak-before-break Reactor coolant pump flywheel fatigue crack growth Response to NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification Response to NRC Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Cooling Systems EPRI MRP cycle-based and fluence-based analyses in support of MRP-227 BWRs and PWRs Fatigue of cranes (crane cycle limits)

Fatigue of the spent fuel pool liner Corrosion allowance calculations Flaw growth due to stress corrosion cracking Predicted lower limit

APPENDIX G LIST OF ABBREVIATIONS USED IN SLR-ISG-PWRVI-2020-XX ADAMS Agencywide Document Access Management System A/LAI applicant/licensee action item AMR aging management review AMP aging management program ANO-1 Arkansas Nuclear One ASME American Society of Mechanical Engineers BMI bottom-mounted instrumentation B&W Babcock & Wilcox Company (currently part of the AREVA corporate complex of private companies)

CASS cast austenitic stainless steel CE Combustion Engineering Company (currently owned by Westinghouse Electric Company)

CEA control element assembly CFR Code of Federal Regulations CRGT control rod guide tube CSS core support shield GALL NUREG-1801, Generic Aging Lessons Learned (GALL) Report GALL-SLR NUREG-2191, Volumes 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal Applications (GALL-SLR)

Report FD flow distributor FSAR final safety analysis report EPRI Electric Power Research Institute IASCC irradiation-assisted stress corrosion cracking I&E inspection and evaluation

SLR-ISG-PWRVI-2020-XX: Appendix G Page 2 of 3 IMI incore monitoring instrument or incore monitoring instrumentation ISG interim staff guidance LCB lower core barrel LR license renewal LRA license renewal application LTS lower thermal shield MRP Materials Reliability Program NRC U.S. Nuclear Regulatory Commission OE operating experience PH precipitation hardened PWR pressurized-water reactor PWSCC primary water stress corrosion cracking RIS regulatory information summary RVI reactor vessel internal SCC stress corrosion cracking SSC structure, system, and component SLR subsequent license renewal SLRA subsequent license renewal application SRP-LR NUREG 1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants SRP-SLR NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants SS stainless steel SSHT surveillance specimen holder tube TLAA time-limited aging analysis TR topical report

SLR-ISG-PWRVI-2020-XX: Appendix G Page 3 of 3 UAW upper axial weld (upper vertical weld)

UCB upper core barrel UTS upper thermal shield XL extra-long X-750 generic reference to a type of nickel-based alloy metal that may be trademarked by industry manufacturers of the material