ML20217L203

Final SLR-ISG-2021-01-PWRVI Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors

ML20217L203
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Issue date:	01/08/2021
From:	James Medoff NRC/NRR/DNRL/NVIB
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MITCHELL J
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ML20352A217	List: ML20217L203
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SLR-ISG-2021-01-PWRVI
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SLR-ISG-2021-01-PWRVI Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors Interim Staff Guidance January 2021

ML20217L203 CAC: TM3021 OFFICE Author:DNLR:NVIB PM:DNRL:NLRP BC:DNLR:NVIB BC:DNLR:NLRP NAME JMedoff JMitchell HGonzalez LGibson DATE 11/17/20 12/3/20 12/04/20 12/4/20 OFFICE SL:DNLR PM:DRO:IRSB QTE OGC NAME AHiser TGovan JDougherty STurk DATE 12/11/20 12/9/20 12/10/20 12/16/20 OFFICE D:DNRL D:NRR:DRO NAME ABradford GSuber for CMiller DATE 12/21/20 01/06/21 INTERIM STAFF GUIDANCE UPDATED AGING MANAGEMENT CRITERIA FOR REACTOR VESSEL INTERNAL COMPONENTS FOR PRESSURIZED-WATER REACTORS SUBSEQUENT LICENSE RENEWAL GUIDANCE SLR-ISG-2021-01-PWRVI PURPOSE The U.S. Nuclear Regulatory Commission (NRC) staff is issuing this 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 SLR-ISG identifies 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-2021-01-PWRVI Page 2 of 10 Program (MRP) Topical Report No. 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), issued June 2020 (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 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-2021-01-PWRVI Page 3 of 10 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-2021-01-PWRVI Page 4 of 10 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 provides guidance to licensees that wish to extend their plant operating licenses from 60 years to 80 years, and SRP-SLR provides guidance 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 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 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 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-2021-01-PWRVI Page 5 of 10 ACTIONS SLR-ISG Objectives 1 and 2GALL-SLR Report and SRP-SLR Guidance Changes and Clarification 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 to address anticipated aging effects associated with the requested 80-year operating period.

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

SLR-ISG-2021-01-PWRVI Page 6 of 10 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 is not 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-2021-01-PWRVI Page 7 of 10 Report. As such, the staff is formally closing the guidance of RIS 2011-07 in SLR-ISG-2021-01-PWRVI.

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 Issuance of this ISG does not constitute a backfit as defined in 10 CFR 50.109(a)(1) and is not otherwise inconsistent with the issue finality provisions in 10 CFR Part 52, Licenses, certifications, and approvals for nuclear power plants. Thus, the NRC staff did not prepare a backfit analysis for the issuance of this ISG.

The NRC staffs position is based upon the following considerations:

• The ISG positions do not constitute backfitting, inasmuch as the ISG is guidance directed to the NRC staff with respect to its regulatory responsibilities. The ISG provides interim guidance to the staff on how to review certain requests. Changes in guidance intended for use by only the staff are not matters that constitute backfitting as that term is defined in 10 CFR 50.109, Backfitting, or that involve the issue finality provisions of 10 CFR Part 52.

• Backfitting and issue finalitywith certain exceptions discussed in this sectiondo not apply to current or future applicants. Applicants and potential applicants are not, with certain exceptions, the subject of either the Backfit Rule or any issue finality provisions under 10 CFR Part 52. This is because neither the Backfit Rule nor the issue finality provisions of 10 CFR Part 52 were intended to apply to every NRC action that substantially changes the expectations of current and future applicants. The exceptions to the general principle are applicable whenever a 10 CFR Part 50 operating license applicant references a construction permit or a 10 CFR Part 52 combined license applicant references a license (e.g., an early site permit) or an NRC regulatory approval (e.g., a design certification rule) (or both) for which specified issue finality provisions apply. The NRC staff does not currently intend to impose the positions represented in this ISG in a manner that constitutes backfitting or is inconsistent with any issue finality provision of 10 CFR Part 52. If in the future the NRC staff seeks to impose positions stated in this ISG in a manner that would constitute backfitting or be inconsistent with these issue finality provisions, the NRC staff must make the requisite showing as set

SLR-ISG-2021-01-PWRVI Page 8 of 10 forth in the Backfit Rule or address the regulatory criteria set forth in the applicable issue finality provision, as applicable, that would allow the staff to impose the position.

• The NRC staff has no intention to impose the ISG positions on existing nuclear power plant licensees either now or in the future (absent a voluntary request for a change from the licensee). The staff does not intend to impose or apply the positions described in the ISG to existing (i.e., already issued) licenses (e.g., operating licenses and combined licenses). Hence, the issuance of this ISGeven if considered guidance subject to the Backfit Rule or the issue finality provisions in 10 CFR Part 52 would not need to be evaluated as if it were a backfit or as being inconsistent with issue finality provisions. If, in the future, the NRC staff seeks to impose a position in the ISG on holders of already issued licenses in a manner that would constitute backfitting or does not provide issue finality as described in the applicable issue finality provision, then the staff must make a showing as set forth in the Backfit Rule or address the criteria set forth in the applicable issue finality provision that would allow the staff to impose the position.

CONGRESSIONAL REVIEW ACT This ISG is a rule as defined in the Congressional Review Act (5 U.S.C. 801-808). However, the Office of Management and Budget has not found it to be a major rule as defined in the Congressional Review Act.

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. Revisions to SRP-SLR Table 3.1-1 B.1 Revisions to GALL-SLR Report Table IV.B2, Reactor Vessel Internals (PWR)

Westinghouse B.2 Revisions to GALL-SLR Report Table IV.B3, Reactor Vessel Internals (PWR)

Combustion Engineering B.3 Revisions to GALL-SLR Report Table IV.B4, Reactor Vessel Internals (PWR)

Babcock & Wilcox B.4 Revisions to GALL-SLR Report Table IV.E, Common Miscellaneous Material/Environment Combinations C. 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)

D. Revisions to GALL-SLR Report AMP XI.M16A, PWR Vessel Internals, and Related FSAR Supplement Example in GALL-SLR Report Table XI-01

SLR-ISG-2021-01-PWRVI Page 9 of 10 E. Revision to GALL-SLR Report Table IX.C, Use of Terms for Materials F. Revisions to SRP-SLR Table 4.7-1, Examples of Potential Plant-Specific TLAA Topics G. List of Abbreviations Commonly Used in SLR-ISG-2021-01-PWRVI H. Disposition of Public Comments 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 Topical Report No. 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), June 2020 (ADAMS Accession No. ML20175A112).

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).

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).

SLR-ISG-2021-01-PWRVI Page 10 of 10

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 Topical 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 Topical 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 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 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.

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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" Page 2 of 11

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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-252c IV.B4.RP-256 IV.B4.RP-256a IV.B4.RP-258a IV.B4.RP-259a IV.B4.RP-261 IV.B4.RP-400 Page 3 of 11

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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 IASCC, fatigue, overload "PWR Vessel Section 3.1.2.2.9) IV.B4.RP-244a internal "Expansion" Internals," and IV.B4.RP-245 components exposed to 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-250a IV.B4.RP-254 IV.B4.RP-254a IV.B4.RP-260a IV.B4.RP-262 IV.B4.RP-352 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-338 IV.B3.RP-342 IV.B3.RP-343 IV.B3.RP-358 IV.B3.RP-362a IV.B3.RP-363 Page 4 of 11

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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 Page 5 of 11 only) IV.B2.RP-291b IV.B2.RP-293 IV.B2.RP-294 IV.B2.RP-298a IV.B2.RP-387a

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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 IASCC, fatigue "PWR Vessel Section 3.1.2.2.9) IV.B2.RP-301 Westinghouse reactor 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.B2.RP-326 alloy Combustion CASS, martensitic SS, IV.B3.RP-338a Engineering reactor and PH SS due to thermal IV.B3.RP-359 internal "Primary" aging embrittlement; IV.B3.RP-360 components exposed to changes in dimensions IV.B3.RP-362 reactor coolant and due to void swelling, IV.B3.RP-364 neutron flux distortion; loss of preload IV.B3.RP-365 due to thermal and IV.B3.RP-366 irradiation-enhanced stress relaxation, creep; loss of material due to wear Page 6 of 11

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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.RP-364 and neutron flux due to void swelling, IV.B3.R-455 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 Page 7 of 11

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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-252b 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-243 including CASS, PH SS or due to neutron irradiation "PWR Vessel Section 3.1.2.2.9) IV.B4.RP-243a martensitic SS), nickel embrittlement and for Internals" IV.B4.RP-245b alloy Babcock & Wilcox CASS, martensitic SS, IV.B4.RP-245c reactor internal and PH SS due to thermal IV.B4.RP-246b "Expansion" components aging embrittlement; or IV.B4.RP-246e exposed to reactor coolant changes in dimensions IV.B4.RP-250 and neutron flux due to void swelling, or IV.B4.RP-252a distortion; or loss of IV.B4.RP-254b preload due to thermal IV.B4.RP-260 and irradiation-enhanced IV.B4.RP-386 stress relaxation, or creep; or loss of material due to wear Page 8 of 11

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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.B4.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 Page 9 of 11 neutron flux IV.B1.R-93 IV.B1.R-96 IV.B1.R-97 IV.B1.R-98 IV.B1.R-99

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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 or SCCIASCC, cyclic or AMP XI.M16A, IV.B4.R-423 LRA/SLRA-specified loading, fatigue "PWR Vessel reactor vessel internal Internals," and AMP component exposed to XI.M2, "Water reactor coolant, neutron Chemistry" (SCC and flux IASCC only), with an adjusted site-specific or component-specific Page 10 of 11 aging management basis for a specified reactor vessel internal component

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System SLR-ISG-2021-01-PWRVI: Appendix A 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 Loss of fracture toughness Plant-specific aging Yes (SRP-SLR IV.B2.R-424 alloy, stellite PWR reactor due to neutron irradiation management program Section 3.1.2.2.9) IV.B3.R-424 vessel internal embrittlement or thermal or AMP XI.M16A, IV.B4.R-424 components or LRA/SLRA- aging embrittlement; "PWR Vessel specified reactor vessel changes in dimensions Internals," with an internal component due to void swelling or adjusted site-specific exposed to reactor coolant, distortion; loss of preload or component-specific neutron flux due to thermal and aging management irradiation-enhanced basis for a specified stress relaxation or creep; reactor vessel internal loss of material due to component wear Page 11 of 11

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

APPENDIX B.1 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.

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.

GALL-SLR Report Table IV.B2 Revisions SLR-ISG-2021-01-PWRVI: Appendix B.1 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 SCC or fatigue "PWR Vessel components: upper neutron flux Internals," and core plate AMP XI.M2, alignment pins "Water Chemistry" (for SCC mechanisms only)

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, (includes corner "Water bolts) Chemistry" (for SCC mechanisms only)

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; (includes corner changes in bolts) dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation or creep; loss of Page 2 of 15 material due to wear

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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)

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 bolts) "Water Chemistry" (for SCC mechanisms only)

Page 3 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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 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)

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 Page 4 of 15 irradiation-enhanced stress relaxation or creep; loss of material due to wear

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-293 3.1-1, 053b Bottom-mounted Stainless steel Reactor Cracking due to AMP XI.M16A, Yes instrumentation coolant and SCC or fatigue "PWR Vessel system: bottom- neutron flux 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; loss instrumentation of material due to (BMI) column wear bodies 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 toughness Internals" guide plates due to thermal aging (cards) embrittlement (CASS only)

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)

Page 5 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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)

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 Page 6 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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 plant-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)

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 Page 7 of 15 mechanisms only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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 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) changes in dimension due to void swelling or distortion Page 8 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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 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 Page 9 of 15 vertical (axial) distortion welds

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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 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) "Water (including bolts or Chemistry" (for screws , dowels, 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 Page 10 of 15 extra-long (XL) fatigue AMP XI.M2, lower core plate "Water Chemistry" (for SCC mechanisms only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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; changes in dimension due to void swelling or distortion 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)

M IV.B2.RP-290a 3.1-1, 059b Lower Stainless Cast Reactor Loss of fracture AMP XI.M16A, Yes supportinternals austenitic coolant and toughness due to "PWR Vessel assembly: lower stainless steel 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 Page 11 of 15 SCC mechanisms only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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)

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 Page 12 of 15 SCC mechanisms only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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 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 LRA/SLRA- SCCIASCC, cyclic program, or specified reactor loading, fatigue AMP XI.M16A, vessel internal "PWR Vessel component Internals," and AMP XI.M2, "Water Chemistry" (SCC and IASCC only),

with an adjusted site-specific or component-specific aging management basis for a specified reactor vessel Page 13 of 15 internal component

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-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 stellite (as a neutron flux neutron irradiation management LRA/SLRA- wear-resistant embrittlement or program, or specified reactor surface) thermal aging AMP XI.M16A, vessel internal embrittlement; PWR Vessel component changes in Internals, with dimensions due to an adjusted void swelling or site-specific or distortion; loss of component-preload due to specific aging thermal and management irradiation-enhanced basis for a stress relaxation or specified creep; loss of reactor vessel material due to wear internal component 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)

Page 14 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.1 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-302 3.1-1, 053a Thermal shield Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly: thermal coolant and SCC or fatigue "PWR Vessel shield flexures neutron flux Internals," and AMP XI.M2, "Water Chemistry" (for SCC mechanisms only)

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 Page 15 of 15

APPENDIX B.2 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.

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.

GALL-SLR Report Table IV.B3 Revisions SLR-ISG-2021-01-PWRVI: Appendix B.2 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 Page 2 of 15 relaxation or creep

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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)

Page 3 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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-Page 4 of 15 entrant cornersthe horizontal seam between the upper and lower shroud segments)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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 MD IV.B3.RP-326a 3.1-1, 052a Core shroud Stainless steel Reactor Cracking due to AMP XI.M16A, Yes assembly (designs coolant and SCC or fatigue "PWR Vessel assembled in two neutron flux Internals," and vertical sections): AMP XI.M2, assembly "Water components, Chemistry" (for including monitoring SCC of the gap opening at mechanisms the core shroud re- only) entrant corners 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 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-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 Page 5 of 15 SCC mechanisms only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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-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 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 plates dimension due to (including visible axial void swelling or weld seams at the distortion core shroud re-Page 6 of 15 entrant corners and at the core midplane)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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-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)

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 Page 7 of 15 Chemistry" (for SCC mechanisms only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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-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)

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 Page 8 of 15 wear Chemistry" (for SCC mechanisms only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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-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)

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 Page 9 of 15 tubes - lower neutron flux Internals"

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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-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 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 Page 10 of 15 from full height of preload due to shroud plates): fuel thermal and alignment pins irradiation-enhanced stress relaxation or creep

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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 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 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 Page 11 of 15 plant designs with a neutron flux neutron irradiation Internals" core support plate): embrittlement core support plate

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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-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)

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 Page 12 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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 LRA/SLRA- alloy neutron flux assisted management specified reactor SCCIASCC, cyclic program, or vessel internal loading, fatigue AMP XI.M16A, component "PWR Vessel Internals," and AMP XI.M2, "Water Chemistry" (SCC and IASCC only),

with an adjusted site-specific or component-specific aging management basis for a specified reactor vessel internal component Page 13 of 15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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 LRA/SLRA- alloy, stellite neutron flux neutron irradiation management specified reactor (as a wear- embrittlement or program, or vessel internal resistant thermal aging AMP XI.M16A, component surface) embrittlement; PWR Vessel changes in Internals, with dimensions due to an adjusted void swelling or site-specific or distortion; loss of component-preload due to specific aging thermal and management irradiation- basis for a enhanced stress specified relaxation or reactor vessel creep; loss of internal material due to component 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" Page 14 of 15 shrouds shroud designs assembled withfrom full height shroud plates): fuel alignment plate

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.2 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 Page 15 of 15

APPENDIX B.3 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.

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-2021-01-PWRVI: Appendix B.3 GALL-SLR Report Table IV.B4 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 Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes assembly (applicable steel, nickel coolant and or fatigue "PWR Vessel to Davis Besse only): alloy neutron flux Internals," and surveillance AMP XI.M2, specimen holder "Water tube (SSHT) Chemistry" studs/nuts or bolts N IV.B4.RP- 3.1-1, 058b Core barrel Stainless Reactor Loss of material due AMP XI.M16A, Yes 245c assembly (applicable steel, nickel coolant and to wear; loss of "PWR Vessel to Davis Besse only): alloy neutron flux preload due to Internals surveillance thermal or irradiation-specimen holder enhanced stress tube (SSHT) studs or relaxation or creep bolts M IV.B4.RP-247 3.1-1, 051a Core barrel Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes assembly: accessible steel, nickel coolant and 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- 3.1-1, 058a Core barrel Stainless Reactor Loss of material due AMP XI.M16A, Yes 247c assembly: lower core steel, nickel coolant and to wear; loss of "PWR Vessel barrel (LCB) bolts alloy neutron flux preload due to Internals thermal and irradiation-enhanced stress relaxation or creep M IV.B4.RP- 3.1-1, 051a Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes 249a assembly: baffle steel coolant and SCC, irradiation- "PWR Vessel Page 2 of 14 plates neutron flux assisted SCC, cyclic Internals," and loading, fatigue AMP XI.M2, "Water Chemistry"

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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 SCCIASCC, Internals," and assembly: baffle-to- fatigue, or overload AMP XI.M2, former bolts and "Water screws 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- 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes 250a assembly: core steel coolant and irradiation-assisted "PWR Vessel barrel cylinder neutron flux SCC or fatigue Internals," and (including vertical AMP XI.M2, and circumferential "Water seam welds); former Chemistry" plates (irradiation-assisted SCC only)

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- neutron flux SCCIASCC, fatigue, Internals," and to-baffle bolts and or overload AMP XI.M2, core barrel-to-former "Water Page 3 of 14 bolts Chemistry" (irradiation-assisted SCCIASCC only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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-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- neutron flux neutron irradiation Internals" to-baffle bolts and embrittlement; loss of core barrel-to-former preload due to bolts thermal and irradiation-enhanced stress relaxation or creep; loss of material due to wear M IV.B4.RP- 3.1-1, 058a Core barrel Stainless Reactor Loss of fracture AMP XI.M16A, Yes 240a assembly: locking steel coolant and toughness due to "PWR Vessel devices (including neutron flux neutron irradiation Internals" locking welds) of embrittlement; loss of baffle-to-former bolts material due to wear and internal baffle-to-baffle bolts M IV.B4.RP- 3.1-1, 051a Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes 241a assembly: locking steel coolant and SCC, irradiation- "PWR Vessel devices (including neutron flux assisted SCCIASCC, Internals," and locking welds) of fatigue, or overload AMP XI.M2, baffle-to-former bolts "Water and internal baffle- Chemistry" (for to-baffle bolts SCC mechanisms only)

MD IV.B4.RP- 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes 244a assembly: locking steel coolant and irradiation-assisted "PWR Vessel devices (including neutron flux SCCor fatigue Internals," and welds) of external AMP XI.M2, baffle-to-baffle bolts "Water and core barrel-to- Chemistry" former bolts (irradiation-Page 4 of 14 assisted SCC only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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- 3.1-1, 058b Core barrel Stainless Reactor Loss of fracture AMP XI.M16A, Yes 243a assembly: locking steel coolant and toughness due to "PWR Vessel devices (including neutron flux neutron irradiation Internals" locking welds) of embrittlement; loss of external baffle-to- material due to wear baffle bolts and core barrel-to-former bolts M IV.B4.RP-248 3.1-1, 051a Core support shield Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes (CSS) assembly: steel, nickel coolant and or fatigue "PWR Vessel accessible upper alloy neutron flux Internals," and core barrel (UCB) AMP XI.M2, bolts and locking "Water devices 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 or embrittlement retaining rings (valve precipitation body components) hardened (PH) stainless steels MD IV.B4.RP- 3.1-1, 051a Core support shield Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes 252a (CSS) assembly: steel coolant and or fatigue "PWR Vessel CSS vent valve top neutron flux Internals," and and bottom retaining AMP XI.M2, rings; vent valve "Water locking devices Chemistry" (for (valve body SCC components) mechanisms only)

Page 5 of 14

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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- 3.1-1, 058b Vent valve assembly: CASS Reactor Loss of fracture AMP XI.M16A, Yes 252a vent valve bodies coolant and toughness due to "PWR Vessel neutron flux thermal aging Internals," and embrittlement AMP XI.M2, "Water Chemistry" (SCC only)

N IV.B4.RP- 3.1-1, 058a Vent valve assembly: Stainless Reactor Loss of material due AMP XI.M16A, Yes 252b original locking steel coolant and to wear (for locking "PWR Vessel devices (associated neutron flux devices associated Internals" with the pressure with the pressure plate, spring retainer, plate, spring and spring, U-cover, key spring retainer, and U ring, and pin in the cover in the assembly) 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- 3.1-1, 051a Vent valve assembly: Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes 252c original locking steel coolant and or fatigue (fatigue "PWR Vessel devices (associated neutron flux only for listed original Internals," and with the key ring, pin locking devices) AMP XI.M2, in the assembly); "Water modified locking Chemistry" (for devices (associated SCC with lock cup, mechanisms Page 6 of 14 jackscrew locking only) cup and bolted block in the assembly -

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

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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 SCC AMP XI.M16A, Yes assembly: upper steel coolant and "PWR Vessel (top) 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 steel coolant and toughness due to "PWR Vessel (top) flange weld neutron flux neutron irradiation Internals" embrittlement M IV.B4.RP- 3.1-1, 051a Flow distributor Stainless Reactor Cracking due to AMP XI.M16A, Yes 256a assembly: flow steel, nickel coolant and fatigue "PWR Vessel distributor (FD) bolt alloy neutron flux Internals" locking devices M IV.B4.RP- 3.1-1, 058a Flow distributor Stainless Reactor Loss of material due AMP XI.M16A, Yes 256b assembly: flow steel, nickel coolant and to wear; changes in "PWR Vessel distributor (FD) bolt alloy neutron flux dimensions due to Internals" locking devices distortion or void swelling or distortion M IV.B4.RP-256 3.1-1, 051a Flow distributor Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes assembly: flow steel, nickel coolant and or fatigue "PWR Vessel distributor (FD) bolts alloy neutron flux Internals," and AMP XI.M2, "Water Chemistry" MD IV.B4.RP- 3.1-1, 051a Incore Monitoring Stainless Reactor Cracking due to AMP XI.M16A, Yes 258a Instrument (IMI) steel coolant and SCC, irradiation- "PWR Vessel guide tube neutron flux assisted SCC, or Internals," and assembly: IMI guide fatigue AMP XI.M2, tube spiders "Water Chemistry" (SCC and irradiation-Page 7 of 14 assisted SCC only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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-258 3.1-1, 058a Incore Monitoring Stainless Reactor Loss of fracture AMP XI.M16A, Yes Instrument (IMI) steel, coolant and toughness due to "PWR Vessel guide tube including neutron flux thermal aging and Internals" assembly: IMI CASS neutron irradiation Incore guide tube embrittlement or spiders (castings) thermal aging embrittlement (for spiders made from CASS)

MD IV.B4.RP- 3.1-1, 051a Incore Monitoring Stainless Reactor Cracking due to AMP XI.M16A, Yes 259a Instrument (IMI) steel coolant and SCC, irradiation- "PWR Vessel guide tube neutron flux assisted SCC, or Internals," and assembly: IMI guide fatigue AMP XI.M2, tube spider-to-lower "Water grid rib sections Chemistry" welds (SCC and irradiation-assisted SCC 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 alloy neutron flux thermal aging, Internals" assembly: IMI guide neutron irradiation tube spider-to-lower embrittlement grid rib sections welds M IV.B4.RP-262 3.1-1, 051b Lower grid assembly: Nickel alloy Reactor Cracking due to SCC AMP XI.M16A, Yes accessible alloy X- coolant and "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 including alternate SCC Page 8 of 14 weld configuration at mechanisms Davis Besse) only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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-261 3.1-1, 051a Lower grid assembly: Nickel alloy Reactor Cracking due to SCC AMP XI.M16A, Yes alloy X-750 dowel-to- coolant and "PWR Vessel guide block welds neutron flux Internals," and (all plants except AMP XI.M2, Davis Besse) "Water Chemistry" MD IV.B4.RP- 3.1-1, 058b Lower grid assembly: Nickel Alloy Reactor Loss of material due AMP XI.M16A, Yes 254b alloy X-750 lower coolant and to wear; changes in "PWR Vessel grid shock pad bolt neutron flux dimensions due to Internals" locking devices void swelling or (Three Mile Island distortion Unit 1 only)

MD IV.B4.RP- 3.1-1, 051b Lower grid assembly: Nickel alloy Reactor Cracking due to AMP XI.M16A, Yes 254a alloy X-750 lower coolant and fatigue "PWR Vessel grid 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 SCC AMP XI.M16A, Yes alloy X-750 lower coolant and "PWR Vessel grid shock pad bolts neutron flux Internals," and (Three Mile Island AMP XI.M2, Unit 1 only) "Water Chemistry" M IV.B4.RP- 3.1-1, 051b Lower grid assembly: Stainless Reactor Cracking due to AMP XI.M16A, Yes 246a upper thermal shield steel, nickel coolant and fatigue "PWR Vessel (UTS) bolt locking alloy neutron flux Internals" devices and lower thermal shield (LTS) bolt/stud locking devices M IV.B4.RP- 3.1-1, 058b Lower grid assembly: Stainless Reactor Loss of material due AMP XI.M16A, Yes 246b upper thermal shield steel, nickel coolant and to wear; changes in "PWR Vessel Page 9 of 14 (UTS) bolt locking alloy neutron flux dimensions due to Internals" devices and lower void swelling or thermal shield (LTS) distortion bolt/stud locking devices

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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-246 3.1-1, 051b Lower grid assembly: Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes upper thermal shield steel, nickel coolant and "PWR Vessel (UTS) bolts and alloy neutron flux Internals," and lower thermal shield AMP XI.M2, (LTS) bolts or "Water studs/nuts Chemistry" N IV.B4.RP- 3.1-1, 051b Core barrel Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes 246c assembly: steel; nickel coolant and "PWR Vessel upper thermal shield alloy neutron flux Internals," and (UTS) bolts AMP XI.M2, Water Chemistry N IV.B4.RP- 3.1-1, 051b Core barrel Stainless Reactor Cracking due to AMP XI.M16A, Yes 246d assembly: steel; nickel coolant and fatigue "PWR Vessel upper thermal shield alloy neutron flux Internals" (UTS) bolt locking devices N IV.B4.RP- 3.1-1, 058b Core barrel Stainless Reactor Loss of material due AMP XI.M16A, Yes 246e assembly: steel; nickel coolant and to wear; changes in "PWR Vessel upper thermal shield alloy neutron flux dimension due to Internals" (UTS) bolt locking void swelling or devices distortion M IV.B4.RP-260 3.1-1, 058b Lower grid fuel Stainless Reactor Loss of fracture AMP XI.M16A, Yes assembly: (a) pads, steel, nickel coolant and toughness due to "PWR Vessel (b) pad-to-rib section alloy neutron flux neutron irradiation Internals" welds, (c) alloy X- embrittlement 750, dowels, cap screws and their locking devices Page 10 of 14

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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- 3.1-1, 051b Lower grid fuel Stainless Reactor Cracking due to SCC AMP XI.M16A, Yes 260a assembly: (a) pads; steel, nickel coolant and or fatigue "PWR Vessel (b), pad-to-rib alloy neutron flux Internals," and section welds; (c) AMP XI.M2, alloy X-750, dowels, "Water cap screws and their Chemistry" (for locking devices SCC mechanisms only)

M IV.B4.RP- 3.1-1, 058a Plenum cover Stainless Reactor Loss of material due AMP XI.M16A, Yes 251a assembly: plenum steel coolant and to wear; loss of "PWR Vessel cover weldment rib neutron flux preload (wear) Internals" pads and, plenum cover support flange, plenum cover support ring Page 11 of 14

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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 LRA/SLRA- alloy neutron flux assisted SCCIASCC, management specified reactor cyclic loading, fatigue program, or vessel internal AMP XI.M16A, component "PWR Vessel Internals," and AMP XI.M2, "Water Chemistry" (SCC and IASCC only),

with an adjusted site-specific or component-specific aging management basis for a specified reactor vessel internal component Page 12 of 14

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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 LRA/SLRA- alloy, stellite neutron flux neutron irradiation management specified reactor (as a wear- embrittlement or program, or vessel internal resistant thermal aging AMP XI.M16A, component surface embrittlement; PWR Vessel material) changes in Internals, with dimensions due to an adjusted void swelling or site-specific or distortion; loss of component-preload due to specific aging thermal and management irradiation-enhanced basis for a stress relaxation or specified creep; loss of reactor vessel material due to wear internal component 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 material Inspection, Category B-N-3 core stainless due to wear Subsections support structure steel 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 SCC AMP XI.M16A, Yes alloy X-750 dowel-to- coolant and "PWR Vessel upper grid fuel neutron flux Internals," and assembly support AMP XI.M2, Page 13 of 14 pad welds (all plants, "Water except including Chemistry" (for alternate weld SCC configuration at mechanisms Davis-Besse) only)

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM SLR-ISG-2021-01-PWRVI: Appendix B.3 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, 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)

Page 14 of 14

APPENDIX B.4 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.

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-2021-01-PWRVI: Appendix B.4 GALL-SLR Report Table IV.E 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)

Page 2 of 2

APPENDIX C 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 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. ML20175A112). 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 and (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-2021-01-PWRVI: 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. ML20175A112ML12017A191 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-2021-01-PWRVI: 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), but identified a new A/LAI.

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-2021-01-PWRVI: 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 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 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 bydue to 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 needed to address an 80-year operating period. In this program, the term MRP-227-A (as

SLR-ISG-2021-01-PWRVI: Appendix D Page 2 of 10 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 MRP-227, Revision 1-A, are identified and justified for each component in the applicable 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-2021-01-PWRVI: Appendix D Page 3 of 10 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-2021-01-PWRVI: 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 bydue to SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material induced bydue to wear; (c) loss of fracture toughness induced bydue 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.

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, Revision 1-A, or 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

SLR-ISG-2021-01-PWRVI: Appendix D Page 5 of 10 also be used as an 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-A, or as modified by a gap analysisMRP-227 (as supplemented).

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, Revision 1-A -A (as supplemented) and its subsections, or MRP-227 (as supplemented). Component reinspection frequencies for Primary and Expansion category components are defined in specific tables in Section 4 of the MRP-227, Revision 1-A report or in MRP-227 (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, Rev. 3 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.

SLR-ISG-2021-01-PWRVI: Appendix D Page 6 of 10 The program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in 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, Revision 1-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, or MRP-227 (as supplemented) 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

SLR-ISG-2021-01-PWRVI: Appendix D Page 7 of 10 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.

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, Revision 1-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.

SLR-ISG-2021-01-PWRVI: Appendix D Page 8 of 10

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, 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.

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

ASME. ASME Code,Section V, Nondestructive Examination. 2004 Edition2. 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). ADAMS Accession No. ML090160206. Palo Alto, California: Electric Power Research Institute.

December 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 Proprietary Topical Report No. 10166093002010399, Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228, Rev. 3). (Non-publicly available ADAMS Accession No. ML092120574ML19081A064). The non-proprietary version of the report may be accessed by members of the public at ADAMS Accession No. ML092750569ML19081A058. Palo Alto, California: Electric Power Research Institute.

July 2009November 2018.

_____. EPRI Topical Report 3002017168, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A). ADAMS 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-2021-01-PWRVI: Appendix D Page 9 of 10 Accession No. ML20175A112. Palo Alto, California: Electric Power Research Institute. June 2020.

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.

SLR-ISG-2021-01-PWRVI: Appendix D Page 10 of 10 GALL-SLR Report Table XI-01 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. prior to the subsequent 30020103991016609 (MRP-228, Rev. 3) period of extended to manage the aging effects on the operation, or reactor vessel internal components, as alternatively, a plant-supplemented by a gap analysis that specific program may identifies enhancements to the program be implemented that are needed to address an 80-year 6 months prior to the operating period. subsequent period to extended operation.

Alternatively, the program relies on 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 due to 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 bydue to 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 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 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 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-2021-01-PWRVI 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-2021-01-PWRVI: 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-2021-01-PWRVI: 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

APPENDIX H Disposition of Public Comments Comments received regarding the draft version of this interim staff guidance are available electronically at the U.S. Nuclear Regulatory Commissions (NRCs) electronic Reading Room at http://www.nrc.gov/reading-rm/adams.html. From this page, the public can access the Agencywide Documents Access and Management System (ADAMS), which provides text and image files of publicly available documents. The following table lists the comments the NRC received regarding the draft version of this ISG.

Letter Number ADAMS Accession No Commenter Affiliation Commenter Name 1 ML20246G654 Nuclear Energy Institute Peter W. Kissinger (NEI) 2 ML20245E539 Electric Power Research Christopher Koehler and Institute, Materials Brian Burgos Reliability Program (EPRI MRP)

As indicated in the table above, the staff received two public comment letters. The comment set submitted by NEI includes a total of three comments on the contents of the draft ISG. The comment set from the EPRI MRP submitted comments of behalf of Mr. J. McKinley (Westinghouse Electric Company, providing a total of 27 comments), Mr. E. Blocher (Dominion Energy Company; submitting one comment), and Mr. M. DeVan (Framatome Corporation, submitting five comments).

The table that follows on the next page of this Appendix provides the comment sources and numbers as listed in the public comment letter, the original comment(s) as written by the commenter, and the NRC staffs response to a specified comment or to a group of comments that the staff compiled as being similar in context. Some comments provided by NEI or the EPRI MRP include justifications for the comments or proposed actions for consideration by the staff for resolution of the specific comments. These justifications and proposed actions may be reviewed through access to ADAMS in the NRCs public electronic Reading Room (https://www.nrc.gov/reading-rm/adams.html) and performing a Web-based ADAMS search (WBA search) for the ML numbers associated with the comment source documents listed in the table above.

Disposition of Public Comments SLR-ISG-2021-01-PWRVI: Appendix H Comment #(s) ISG Section/Page Comment NRC Staff Response Comments From NEI NEI #1 ISG Appendix C, NEI wrote: PWR applicants will use MRP-191, Rev 2, The staff did not accept NEI Comment #1 or NEIs Edits to AMR Screening, Categorization, and Ranking of Reactor recommended revision of the staffs AMR Further Further Evaluation Internals Components, for Westinghouse and Evaluation guidance in SRP-SLR Sections 3.1.2.2.9 and Section 3.1.3.2.9 Combustion Engineering PWR Design, or MRP-189 3.1.3.2.9 (i.e., to include guidance criteria that would (ISG pages 3 and 4 Rev 3, Screening, Categorization, and Ranking of permit use of the proprietary EPRI MRP-191, Rev. 2 or of 4 in the Babcock and Wilcox Designed Pressurized Water MRP-189, Rev. 3 reports for component-specific appendix) Reactor Internals Component Items and Welds, as the screening objectives).

principal input for the aging effects and aging mechanism screening portion (first two bullets of the Specifically, these reports have not been formally Gap Analysis on page 3 of Appendix C) for their Reactor reviewed for acceptance by the NRC staff and were not Vessel Internals (RVI) Gap Analysis. used as the component-specific screening report criteria for the staff-accepted inspection and evaluation (I&E)

Recommend revising the ISG to reference MRP-191 guidelines in MRP-227, Rev. 1-A. Instead, these reports Rev 2 or MRP-189 Rev 3 as one acceptable way to form the screening and ranking bases for what will be screen RVI aging effects and aging mechanisms (first EPRIs updated I&E guidelines for a planned MRP-227, two bullets of the Gap Analysis on page 3 of ISG Rev. 2 report, which has yet to be submitted for staff Appendix C). review. Accordingly, the staff does not find it appropriate to reference MRP-191, Rev. 2 or MRP-189, Rev. 3 in the staffs updates of SRP-SLR Sections 3.1.2.2.9 or 3.1.3.2.9, as provided in Appendix C of this ISG.

However, because the staffs update of AMP XI.M16A, PWR Vessel Internals, in the ISG permits the use of additional reports or methodologies to supplement MRP-227, Rev. 1-A, an SLR applicant would not be precluded from using MRP-191, Rev. 2 or MRP-189, Rev. 3 for component-specific screening objectives if the applicant determines that the use of those reports is appropriate for its RVI management program.

For SLR applicants that decide to use these reports, the staff expects that use of the reports would be discussed, evaluated, and supported in the applicants technical basis document for its RVI management program.

Page 2 of 20

Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H NEI #2 ISG Appendix C, NEI wrote: MRP 2018-022, Transmittal of MRP-191- The staff did not accept NEI Comment #2 or NEIs Edits to AMR SLR Screening, Ranking and Categorization Results and recommended revision of the staffs AMR Further Further Evaluation Interim Guidance in Support of Subsequent License Evaluation guidance in SRP-SLR Sections 3.1.2.2.9 and Section 3.1.3.2.9 Renewal at U.S. PWR Plants, provides expert panel 3.1.3.2.9 (i.e., to state that EPRI Report MRP-2018-022 (ISG pages 3 and 4 results that supplement MRP-227 Inspection and may be used to identify component-specific gap analysis of 4 in the Evaluation guidance for subsequent license renewal changes to the I&E criteria defined for the components in appendix) prior to the publication of MRP-227 Revision 2. the MRP-227, Rev. 1-A report).

Recommend revising the ISG to reference MRP-2018- The staff acknowledges that the MRP-2018-022 report 022 as an acceptable starting point to identify changes was developed by EPRI as an Expert Panel basis to to the existing MRP-227 Rev 1-A program identify those changes in I&E criteria in MRP-277, characteristics or criteria (third bullet of the Gap Analysis Rev. 1-A (as developed under a 60-year assessment on page 3 of ISG Appendix C). basis) that would be necessary for a subsequent period of extended operation (i.e., by further licensed operations for years 60 - 80). The staff also acknowledges that MRP-2018-022 was used and acceptably justified on a case-by-case basis for the RVI gap analyses of the RVI management programs in the first two PWR SLRAs reviewed by the staff (i.e., those for the nuclear units at the Turkey Point and Surry power stations). However, MRP-2018-022 is currently limited in that it only applies to Westinghouse and CE design PWR vessel internals and the scope of the report does not include RVI components in B&W design PWRs. Nor has NRC staff review of MRP-2018-022 been requested. Thus, the staff has not found the report acceptable for generic use by the industry licensees. Based on this rationale, the staff does not find it appropriate to reference MRP-2018-022 in the staffs updates of SRP-SLR Sections 3.1.2.2.9 and 3.1.3.2.9, as provided in Appendix C of the ISG.

However, since the staffs update of GALL-SLR AMP XI.M16A, PWR Vessel Internals, in the ISG permits use of additional reports or methodologies in supplement of MRP-227, Rev. 1-A, this would not preclude an SLR applicant from using MRP-2018-022 for the gap analysis of its PWR internal components if it is determined that the use of the report is appropriate for the RVI management Page 3 of 20 program. For SLR applicants that decide to apply this report as one of the supporting reports for the RVI management program, the staff expectation is that use of the report would be discussed, evaluated, and supported in the applicants technical basis document for the program.

Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H NEI #3 ISG Appendices B1 NEI wrote: Aging effects and mechanisms identified in The staff partially accepted NEIs comment.

through B4, various NUREG-2191 AMR lines and NUREG-2192 Table 1s pages should be consistent with those identified in MRP-227 The staff acknowledges that, for the development of Rev 1-A Section 4 I&E Tables. SRP-SLR and GALL-SLR Report, the norm was to include an AMR item for cracking of a specified component with an AMR item on irradiation or thermal embrittlement of the component even if cracking mechanisms were not attributed as being applicable to the component in Chapter 4 of MRP-227, Rev. 1-A. The reason for this is that the inspection methods implemented by the AMP and by MRP-227, Rev. 1-A cannot inspect for direct evidence of embrittlement, and instead are only designed to look for presence of flaw indications that, if detected, may provide indirect evidence of embrittlement occurring in the components.

Besides those listed in Chapter 4 of MRP-227, Rev. 1-A, the staffs identified aging effects and mechanisms in the AMR items of SLR-ISG-2021-01-PWRVI are also based on lessons learned from the gap analyses provided in the Turkey Point and Surry SLRAs or as cited in the SLRAs as being contained in the MRP-2018-022, 80-Year Expert Panel report. Additionally, for some cases, the Chapter 4 AMR item in MRP-227, Rev. 1-A for a given component may cite cracking as an applicable aging effect without listing the cracking mechanisms. For these cases, the staff used its own engineering judgement for the cited cracking mechanisms.

Based on partial acceptance of this comment, the staff has reviewed the relevant AMR items for the accuracy of cited aging effects and mechanisms and adjusted the cited effects and mechanisms in the applicable AMR items on a case-by-case basis. The staff will provide the final basis for the cited aging effects and mechanisms in the staffs official technical basis statements for the specified AMR items.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H Comments from the EPRI MRP EPRI MRP #1 For #1, ISG Collectively, EPRI MRP made the following statements, The staff did not accept EPRI MRP Comments #1 and #8 and #8 Appendix A, as a basis for deleting GALL-SLR IV.B2.RP-296a as a or the EPRI MRPs statements that: (1) cracking should Table 3.1-1, referenced GALL-SLR item for SRP-SLR Table 3.1-1, not be attributed for guide plates (guide cards) in AMR Item 053a Item 053a in Appendix A of the ISG and for deleting Westinghouse-design control rod guide tube (CRGT)

GALL-SLR Item IV.B2.RP-296a from the scope of Table assemblies, and (2) GALL-SLR Report (Page 5 of the IV.B2 in Appendix B1 of the ISG: item IV.B2.RP-296a should not be included as a new AMR appendix) item in Appendix B.1 of the ISG or referenced as a IV.B2.RP296a adds the guide plates/cards for cracking GALL-SLR Report item for the update of SRP-SLR For #8, ISG mechanisms. This has been dispositioned in the past Table 3.1-1, Item 053a in Appendix A of the ISG.

Appendix B.1, (See MRP227 Table 33) and the same is expected for GALL-SLR the future. The mechanisms are not new for SLR, so Specifically, based on lessons learned from the staffs Table IV.B2, Item this does not seem like a necessary addition. approval of the first two (2) SLRAs for the Westinghouse-IV.B2.RP-296a designed PWR units (i.e., the Turkey Point and Surry It is true that SCC and fatigue were screened in for the SLRAs), cracking was identified as potentially applicable (Page 6 of the guide cards/plates in earlier revisions, as well, but to the AMR assessments of the CRGT guide cards in the appendix) Table 33 in MRP227 shows that these mechanisms units. This included reporting of stress corrosion cracking were dispositioned as no additional measures. One (SCC) as a potentially applicable cracking mechanism for major reason for this was the inclusion of the lower the CRGT cards in the Turkey Point units and fatigue as flange welds as leading components that are directly an applicable cracking mechanism for the CRGT guide representative of the guide plates/cards. card in the Surry units. Thus, the staff does not find EPRI MRP Comments #1 and #8 to be consistent with industry lessons learned from the staffs processing of the first two SLRAs for Westinghouse-design PWR units.

GALL-SLR Report Item IV.B2.RP-296a will remain as a new GALL-SLR Report item for managing cracking in Westinghouse-design CRGT guide plates (guide cards),

as cited in Appendix B.1 of the ISG, and will be referenced in the staffs update of SRP-SLR Table 3.1-1, Item 053a in Appendix A of the ISG. However, there is no requirement for an applicant to use the referenced GALL-SLR Report AMR item IV.B2.RP-296a for its SLRA if the applicants integrated plant assessment (IPA) for the SLRA concludes that cracking is not an aging effect that requires management for the specified CRGT guide card components.

Page 5 of 20

Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #2 For #2, ISG Collectively, EPRI MRP made the following statements The staff did not accept EPRI MRP Comments #2 and #10 and #10 Appendix A, as a basis for deleting GALL-SLR Report AMR or the EPRI MRPs statements that: (1) cracking should Table 3.1-1, item IV.B2.RP-345a as a referenced item in SRP-SLR not be attributed for core barrel flanges (base metal of AMR Item 053c Table 3.1-1, item 053c, and for deleting GALL-SLR components) in Westinghouse-design core barrel Report item IV.B2.RP-345a from Table IV.B2 in assemblies and (2) GALL-SLR Report item (Page 6 of the Appendix B1 of the ISG: IV.B2.RP-345a should not be included as a new AMR item appendix) in Appendix B.1 of the ISG or referenced as a GALL-SLR Comment #2: IV.B2.RP345a adds cracking Report AMR item for the update of SRP-SLR Table 3.1-1, For #10, ISG mechanisms for the upper core barrel flange. This has item 053c in Appendix A of the ISG.

Appendix B.1, been dispositioned in past MRP227 revisions because GALL-SLR the UFW is the location of concern for cracking (due to Specifically, based on lessons learned from the staffs Table IV.B2, Item the weld) and is already included as a Primary. processing and approval of the first two SLRAs for the IV.B2.RP-345a Westinghouse-designed PWR units (i.e., the Turkey Point Comment #10: Item IV.B2.RP345a is covered by the and Surry SLRAs), cracking was identified as potentially (Page 8 of the cracking screened in for the UFW in IV.B2.RP276 (see applicable to the AMR assessments of the core barrel appendix) overall comment 2). Cracking is most likely to occur at flanges in the units, including the reporting of fatigue as an the weld and not on the base metal of the flange. applicable cracking mechanism for the core barrel flanges of the Surry units. Thus, the staff does not find EPRI MPR Comments #2 and #10 to be consistent with industry lessons learned from the staffs processing of these SLRAs.

GALL-SLR Item IV.B2.RP-345a will remain as a new GALL-SLR item for managing cracking in Westinghouse-design core barrel flanges, as cited in Appendix B.1 of the ISG and will be referenced in the staffs update of SRP-SLR Table 3.1-1, Item 053c in Appendix A of the ISG.

However, there is no requirement for an applicant to use the referenced GALL-SLR Report AMR item IV.B2.RP-345a for its SLRA if the applicants integrated plant assessment (IPA) for the SLRA concludes that cracking is not an aging effect that requires management for the specified core barrel flange components.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #3 ISG Appendix B.1, EPRI MRP wrote the following as a basis for requesting The staff accepted EPRI MRP Comment #3. Upon further GALL-SLR Table adjustments of the component descriptions in GALL-SLR review of the past Turkey Point and Surry SLRAs, the staff IV.B2, Items Items IV.B2.RP-299 and IV.B2.RP-301: confirmed the fuel alignment pins in the upper internals IV.B2.RP-299 and assembly are not the same components as the core plate IV.B2.RP-301 The addition of "(fuel alignment pins)" to this line is not alignment pins in Westinghouse-design PWRs. The staff correct, since the UCP alignment pins are different than has deleted the parenthetical clause (fuel alignment (Page 2 of the the fuel alignment pins. The UCP alignment pins are pins) from the Component column entries in the appendix) attached to the core barrel and interface with the updated versions of GALL-SLR Items IV.B2.RP-299 and alignment slots on the edge of the upper core plate. It is IV.B2.RP-301, as updated in Appendix B.1 of the ISG.

possible that the inclusion of reference to TB164 here in the Existing table (Table 49) has caused some In addition, loss of fracture toughness due to neutron confusion. This was intended to incorporate the latest irradiation embrittlement has been removed as an OE to MRP227 by reference. However, that technical applicable aging effect and mechanism combination for bulletin does not provide requirements for the UCP the core plate alignment pins in GALL-SLR Report alignment pins themselves. (See SLR new existing Item IV.B2.RP-299, as loss of material due to wear is component wear MRP 2018022). listed as the only non-cracking aging effect and mechanism combination for the core plate alignment pins in Item W15 of Table 4.9 in the MRP-227, Rev. 1-A report.

Thus, the modification of GALL-SLR Report item IV.B2.RP-299 has been deleted from the final ISG, and no change is proposed to IV.B2.RP-299.

EPRI MRP #4 ISG Appendix B.1, EPRI MRP wrote the following as a basis for requesting The staff accepted EPRI MRP Comment #4.

GALL-SLR Table adjustments of the component descriptions in GALL-SLR IV.B2, Items Items IV.B2.RP-275 and IV.B2.RP-354: Based on the staffs acceptance of EPRIs rationale made IV.B2.RP-275 and in Comment #4, the staff deleted reference of corner bolts IV.B2.RP-354 The modified text deleted "all plants with baffleedge and the component-related parenthetical explanation from bolts" and replaced it with "corner bolts". This is not the Structure and/or Component column entries in the (Page 4 of the correct. Corner bolts are a subset of baffleformer bolts, final versions of the GALL-SLR IV.B2.RP-275 and appendix) not baffleedge bolts. Note that Bracket bolts are a IV.B2.RP-354 items in the ISG. The modified Structure subset of baffleedge bolts. (See MRP227 Table 43 and/or Component column entries for the RP-275 and W7Baffle Former Assembly (Includes: Baffle plates, RP-354 items now read as Baffle-to-former assembly:

baffle edge bolts, corner bolts). baffle-edge bolts.

Based on the statements in EPRI MPR Comment #4 and the staffs discussions of the comment with members of the industry during the public meeting on the ISG of November 19, 2020, the staff also modified the Structure Page 7 of 20 and/or Component column entries of the GALL-SLR Report AMR items IV.B2.RP-271 and IV.B2.RP-272 for baffle-to-former bolts to include the parenthetical clause (includes corner bolts) as part of the component descriptions.

Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #5 For #5, ISG EPRI MRP wrote the following as a basis for requesting The staff accepted EPRI MRP Comments #5 and #6 for and #6 Appendix B.1, deletion of the mechanism IASCC from GALL-SLR the final adjustment adjustments of GALL-SLR Report GALL-SLR Table Report AMR item IV.B2.RP-293 and the aging effect and items IV.B2.RP-292 and IV.B2.RP-293 in Appendix B.1 of IV.B2, Item mechanism combination of changes in dimension due to the ISG. Specifically, the staff confirmed that irradiation-IV.B2.RP-293 void swelling or distortion from GALL-SLR Report AMR assisted stress corrosion cracking (IASCC) and void item IV.B2.RP-292: swelling (VS) are not listed as applicable irradiation-For #6, ISG induced aging mechanisms for BMI column bodies in Appendix B.1, Not sure where IASCC comes from for the BMI column item W2.2 of Table 4-6 in the MRP-227, Rev. 1-A reports.

GALL-SLR Table bodies. These are only screened in for SCC, wear, and Although the staff confirmed that the Surry SLRA had IV.B2, Item fatigue in MRP191, Rev. 2. They are only included for indicated that IASCC and VS were identified as applicable IV.B2.RP-292 fatigue and IE in MRP227, Rev. 1A. mechanisms for the BMI column bodies in MRP-191, Rev. 1, it also identified that the IASCC and VS (Page 5 of the Similar to Item IV.B2.RP293, the reason for adding mechanisms were moved from the assessment of the appendix) void swelling, distortion, and changes in dimension [i.e., components based on the 80-Year Expert Panel basis for to Item IV.B2.RP-292] is not clear. This does not show the column bodies performed in MRP-2018-022.

up in MRP191 or MRP227, Rev. 1A.

Thus, the staff finds this basis to be sufficient for removing IASCC as a listed aging mechanism for the update of GALL-SLR Item IV.B2.RP-293 in Appendix B.1 of the ISG and for removing changes in dimension due to void swelling or distortion as a listed aging effect and mechanism combination for the components in GALL-SLR Item IV.B2.RP-292. The Aging Effect/Mechanism column entry for GALL-SLR Item IV.B2.RP-293 will now read as Cracking due to SCC or fatigue for the final version of the line item in Appendix B.1 of the ISG.

Similarly, the Aging Effect/Mechanism column entry for GALL-SLR Report item IV.B2.RP-292 will now read as Loss of fracture toughness due to neutron irradiation embrittlement; loss of material due to wear.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #7 ISG Appendix B.1, EPRI MRP wrote the following as a basis for requesting The staff accepted EPRI MRP Comment #7 for the final GALL-SLR Table deletion of loss for fracture toughness due to irradiation adjustment of GALL-SLR Report item IV.B2.RP-296 in the IV.B2, Item embrittlement from GALL-SLR Item IV.B2.RP-296: ISG and the EPRI MRPs statement in the comment that IV.B2.RP-296 GALL-SLR Report item IV.B2.RP-296 should not list Not sure where the IE comes from here for the guide irradiation embrittlement (IE) as a listed loss of fracture (Page 6 of the cards/plates. This mechanism does not show up in toughness mechanism for the guide plates (guide cards) appendix) MRP227, Rev. 1A or MRP191, Rev. 2. It was included in Westinghouse-design control rod guide tube (CRGT) for the CASS guide cards in MRP191, Rev. 1, but the assemblies.

updated and refined analysis in MRP191, Revision 2 did not show IE for these. Guide cards made from CASS will still be listed as being susceptible to loss of fracture toughness due to thermal aging embrittlement (TE). Based on acceptance of EPRI MRP Comment #7, the staff has amended Aging Effect/Mechanism column entry for GALL-SLR Report item IV.B2.RP-296 to read as Loss of material due to wear; loss of fracture toughness due to thermal aging embrittlement (CASS only).

EPRI MRP #9 ISG Appendix B.1, EPRI wrote the following in relation to the staffs The staff accepted EPRI MRP Comment #9 for the final GALL-SLR Table proposed revision of Aging Management Program adjustment of GALL-SLR Report item IV.B2.RP-355 in the IV.B2, Item (AMP)/TLAA column entry for GALL-SLR Item ISG and the EPRI MRPs recommended editorial change IV.B2.RP-355 IV.B2.RP-355 in the ISG: of GALL-SLR Report item IV.B2.RP-355 as stated in the comment. Section 4.5 in the MRP-227, Rev. 1-A report (Page 7 of the Should the statement "using componentspecific uses the words plant-specific, so the staff has amended appendix) evaluation per MRP guidelines" be "using plantspecific the wording to be consistent with those stated in the aging management program per MRP227 guidelines" to MRP-227, Rev. 1-A report.

be consistent with MRP227, Revision 1A, Section 4.5? The Aging Management Program (AMP)/TLAA column entry of GALL-SLR Report item IV.B2.RP-355 has been amended to read as AMP XI.M16A, "PWR Vessel Internals," and AMP XI.M2, "Water Chemistry" (for SCC mechanisms only) - using plant-specific evaluation per MRP guidelines.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #11 ISG Appendix B.1, EPRI wrote the following as a basis for requesting The staff did not accept EPRI MRP Comments #11 and and #12 GALL-SLR Table removal of irradiation stress corrosion cracking (IASCC) #12 or the EPRI MRPs collective statements in the two IV.B2, Items as a listed aging mechanism for core barrel lower flange comments that neither irradiation-assisted stress corrosion IV.B2.RP-280 and welds (LFWs) in the updated ISG version of GALL-SLR cracking (IASCC) nor irradiation embrittlement (IE) should IV.B2.RP-280a Items IV.B2.RP-280 and for deleting GALL-SLR be attributed to Westinghouse design core barrel lower IV.B2.RP-280a from the scope of the ISG: flange welds (LFWs).

(Page 9 of the appendix) The grouping of the LGW and LFW in MRP191, Based on lessons learned from the staffs review of the Revision 2 have likely caused confusion here. The LGW Surry SLRA, the past applicant indicated that IASCC, IE, is in the core beltline and is subject to irradiation effects, and void swelling (VS) were all attributed as being such as IASCC. The LFW is located near the bottom of screened-in irradiation mechanisms for Westinghouse-the core barrel where IASCC is not an effect. These design core barrel LFWs per EPRIs 80-Year Expert Panel were grouped originally because they are both in the analysis of the components in MRP-2018-022.

lower core barrel. Note that this type of detail will be Specifically, the Surry SLRA indicated that the core barrel addressed in MRP232 and MRP227, Rev. 2. LFWs are within 80-year fluence exposure zones high enough to screen the welds in for IASCC, IE and VS Similar to overall comment 11 above, the LFW is far mechanisms. Thus, the final version of GALL-SLR Report from the core and not susceptible to irradiation effects. item IV.B2.RP-280 has been amended to include IASCC as a potential aging mechanism for the core barrel LFWs.

Similarly, the final version of GALL-SLR Report item IV.B2.RP-280a in the ISG has been amended further to include both IE and VS as potential aging mechanisms for the core barrel LFWs based on the inclusion of IE and VS in the Surry gap analysis. The staff cannot rely on MRP-227, Rev. 2 report as the basis for providing updated aging mechanisms for the LFWs, as the report has not yet to be docketed for staff approval in ADAMS or accepted by the staff for implementation.

Given these comment considerations, the Aging Effect/Mechanism column entry for GALL-SLR Report item IV.B2.RP-280 will remain as amended in the ISG to state Cracking due to SCC, IASCC (lower flange weld only), or fatigue, and the corresponding column entry in GALL-SLR Report item IV.B2.RP-280a has been further amended in the final ISG to state Loss of fracture toughness due to neutron irradiation embrittlement; changes in dimension due to void swelling or distortion.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #13 ISG Appendix B.1, EPRI wrote the following as a basis for deleting the The staff did not accept EPRI MRP Comment #13 or the GALL-SLR Table additional aging mechanisms cited in the staffs EPRI MRPs statement in the comment that void swelling IV.B2, Items proposed updated of GALL-SLR Item IV.B2.RP-287: or distortion should not be included as cited aging IV.B2.RP-287, mechanisms for the non-cracking based AMR line items IV.B2.RP-290, A problem in multiple places is the addition of all for Westinghouse lower support column bodies made from IV.B2.RP-295 degradation mechanisms screened in under MRP191, either cast or wrought stainless steel materials (i.e., in Rev. 2. For example, the LSC bolts and support columns GALL-SLR Report items IV.B2.RP-290 or IV.B2.RP-295)

(Pages 12 and 13 were screened in for VS under MRP191, Rev. 1; and for lower support column bolts made from stainless of the appendix) however, those were dispositioned for MRP227 based steel (i.e., in GALL-SLR Report item IV.B2.RP-287).

on more detailed evaluation of those locations. Based on that, VS was assigned to "no additional measures" in Specifically, based on lessons learned from the staffs Table 33 of MRP227, Revision 1A. That same logic review of the Surry SLRA, the past applicant indicated that applies here, too. void swelling (VS) was attributed as being a screened-in an irradiation aging mechanism for Westinghouse-design The comment also applied to GALL-SLR Items lower support column bodies and lower support column IV.B2.RP-290 and IV.B2.RP-295, as updated in bolts per EPRIs 80-Year Expert Panel analysis of the Appendix B.1 of the ISG. components in MRP-2018-022. Specifically, the Surry SLRA indicates that the lower support column bodies and lower support column bolts are within 80-year fluence exposure zones high enough to screen the components in for applicable irradiation-mechanisms (i.e., irradiation-assisted stress corrosion cracking (IASCC), irradiation embrittlement (IE), and VS, and for the bolts, irradiation-enhanced stress relaxation/irradiation-enhanced creep (ISR/IC) as a potential loss of preload mechanism).

Based on these considerations, the staff finds it appropriate for the updates of the GALL-SLR Report items IV.B2.RP-290 and IV.B2.RP-295 for the cast and non-cast lower support column bodies to cite VS and IE as applicable non-cracking, irradiation-based mechanisms for the column bodies, and for the update of GALL-SLR Report item IV.B2.RP-287 for the lower support column bolts to include VS, IE, and ISR/IC as the listed non-cracking, irradiation-based mechanisms for the bolts.

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EPRI MRP For #14 and #16, For the updates of GALL-SLR Items IV.B2.R-423 and The staff partially accepted these comments. The staff SLR-ISG-2021-01-PWRVI: Appendix H

1. 14, #15, #16, ISG Appendix B.1, IV.B2.R-424 in ISG Appendix B.1 and GALL-SLR Items applied these comments generically to the staffs updates

1. 17, #22 and GALL-SLR Table IV.B3.R-423 and IV.B3.R-424 in ISG Appendix B.2, of the SRP-SLR Table 3.1-1, items 118 and 119 in

1. 23 IV.B2, Items EPRI wrote: Appendix A of the ISG and to the staffs updates of all IV.B2.R-423 and R-423 and R-424 type line items in GALL-SLR Report IV.B2.R-424 This is a catchall AMR line. Revise the AMP column to Tables IV.B2, IV.B3, and IV.B4, as cited and revised in delete "or specified reactor internal componentspecific Appendices B.1, B.2, and B.3 of the ISG. Upon further (Page 14 of ISG aging management basis". The Structure or component review, the staff agrees with the EPRI MRP that further Appendix B.1) column already indicates the need for a sitespecific or changes can be made to SRP-SLR items 118 and 119 component specific aging management basis. and the GALL-SLR Report type R-423 and R-424 items For #15 and #17, for consistency objectives, but not necessarily in the exact ISG Appendix A, For the updates of SRP-SLR Table 3.1-1 Items 118 and manner that EPRI recommended in this set of comments.

SRP-SLR Table 119 in ISG Appendix A, EPRI wrote:

3.1-1, Items 118 Since the set of components in a specified GALL-SLR and 119 Revise Table 3.11 item 118 Component column and Report item is commonly only a subset of the components the Aging Management Program/TLAA column to be listed in a related SRP-SLR item, the component (Pages 10 and 11 consistent with revision of AMR IV.B2.R423. descriptions in the GALL-SLR Report R-423 and R-424 of ISG Appendix A) items only need to be similar to the component Revise Table 3.11 item 119 Component column and descriptions for the analogous items in SRP-SLR For #22 and #23, the Aging Management Program/TLAA column to be Table 3.1-1, which is the correlation of SRP-SLR ISG Appendix B.2, consistent with revision of AMR IV.B2.R424. item 3.1-1-118 to the R-423 items for cracking GALL-SLR Table mechanisms and the correlation of SRP-SLR item IV.B3, Items 3.1-1-119 to the R-424 items for non-cracking IV.B3.R-423 and mechanisms.

IV.B3.R-424 Therefore, the staff updated the Component column (Pages 14 and 15 entries of SRP-SLR Table 3.1-1, items 118 and 119, to of ISG Appendix state: Stainless steel, nickel alloy . . . PWR reactor B.2) vessel internals components or LRA/SLRA-specified reactor vessel internal component exposed to reactor coolant, neutron flux. Stellite was also included in the adjustment of the component description in SRP-SLR item 3.1-1-119.

The staff also updated the Structure and/or Component column entries GALL-SLR Report items IV.B2.R-R-423, IV.B3.R-423, IV.B4.R-423, IV.B2.R-424, IV.B3.R-424, and IV.B4.R-424 to state: Reactor vessel internal components or LRA/SLRA specified reactor vessel Page 12 of 20 internal component. The staff updated the Aging Management Program (AMP)/TLAA column entries of the R-423 items to be consistent with the corresponding column entry in SRP-SLR item 3.1-1-118, and updated the Aging Management Program (AMP)/TLAA column entries of the R-424 items to be consistent with the

Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP corresponding column entry in SRP-SLR item 3.1-1-119.

1. 14, #15, #16, The staff also updated the Material column entries of the

1. 17, #22 and R-424 items to state: Stainless steel, nickel alloy, stellite

1. 23 (Cont.) (as a wear-resistant surface).

Stellite needed to be added as part of the final adjustment of the component description of SRP-SLR item 3.1-1-119 and the material descriptions in the GALL-SLR Report R-424 items in order to be consist with industry comments that stellite has been used as a material wear-resistant surface for some RVI components.

EPRI MRP #18 ISG Appendix B.1, EPRI wrote the following as a basis for requesting The staff accepted EPRI MRP Comment #18 and the GALL-SLR Table deletion of loss of fracture toughness due to neutron EPRI MRPs statement that the update of GALL-SLR IV.B2, Item irradiation embrittlement and loss of preload due to Report item IV.B2.RP-302a in the ISG should not include IV.B2.RP-302a irradiation-enhanced stress relaxation or creep as additional cited irradiation effects for Westinghouse-design additional cited non-cracking effect and mechanism thermal shield flexures.

(Page 15 of the combinations for GALL-SLR Item IV.B2.RP-302a in ISG appendix) Appendix B.1: Specifically, based on lessons learned from the staffs review of the Surry SLRA gap analysis, the past applicant The irradiation effects added for this component appear indicated that the thermal shield flexures for the Surry to originate from MRP191, Revision 1. They show up in units did not screen in for void swelling (VS), irradiation that revision, but the more refined analysis of MRP191, embrittlement (IE),or irradiation-enhanced stress Revision 2 showed that these irradiation effects were not relaxation or irradiation-enhanced creep (ISR/IC) per a concern for the thermal shield flexures. EPRIs 80-Year Expert Panel analysis of the components in MRP-2018-022. Specifically, the gap analysis indicates that the projected 80-year fluence exposures of the thermal shield flexures are in a fluence zone lower than the threshold for screening the thermal shield flexures in for the referenced irradiation mechanisms.

Since this ISG only reports changes to the existing guidance for PWR RVI components in the GALL-SLR Report and SRP-SLR, and since the staff has determined as a result of this comment that there is no change to GALL-SLR Report item IV.B2.RP-302a, the staff removed this item from the final version of this ISG. The item in the GALL-SLR Report stands unrevised.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #19 ISG Appendix B.2, EPRI wrote the following regarding the staffs basis for The staff did not accept Comment #19. Based on the GALL-SLR Table adding CE design upper internals assembly guide lug staffs rejection of NEI Comment #1, the staff cannot use IV.B3, Item inserts to the draft update of GALL-SLR Item MRP-191, Rev. 2 as a basis for citing component-specific IV.B3.RP-320 IV.B3.RP-320 in ISG Appendix B.2: aging effects or mechanisms in this ISG.

(Page 2 of the It is unclear why the guide lug inserts were added to this Furthermore, in MRP-227, Rev. 1-A, Table 3-2, the guide appendix) line for cracking effects. They are only screened in for lugs and guide lug inserts and bolts, are screened in as wear and IE in MRP191, Rev. 2. Existing Program components (i.e., X designations) for the mechanisms of fatigue, wear, and irradiation-enhanced stress relaxation or creep (ISR/IC), even though the EPRI MRP did not specifically reflect fatigue as a cited mechanism in the Existing Program line items for the components in Line Items C13 and C14 in Table 4-8 of the MRP-227, Rev. 1-A. In Footnote 3 of Table 3-2 in MRP-227, Rev. 1-A, EPRI makes the following statement relative to potential degradations that may occur in the guide lug fixtures:

Bolt deterioration may lead to degradation in the lug fixtures. Inspection recommendations relate to the entire guide lug fixture.

Thus, the inclusion and updated version of GALL-SLR Report item IV.B3.RP-320 in Appendix B.2 of the draft ISG remains valid for the final version of the line item in the final ISG and it is appropriate for the staff to include the guide lug inserts in the scope of the line item. Based on Footnote 3 in Table 3-2 of MRP-227, Rev. 1-A, and for simplicity of the RP-320 line item, the staff conservatively applied fatigue to all of the specified guide lug fixture components that are within the scope of the RP-320 item. The staff assumes the ASME Section XI VT-3 examinations credited for the guide lug fixtures in Table 4-8 of MRP-227, Rev. 1-A, are sufficient to detect any potential cracking or wear that may occur in the fixtures, and additionally, any loss of preload that may occur in the lug bolts. However, there is no requirement that forces an applicant to use the RP-320 item for its Page 14 of 20 SLRA if the IPA concludes that cracking is not any aging effect that requires management for the specified guide luge fixture components. The applicant could also apply the RP-320 item as a consistent-with-GALL item for only some of the referenced guide lug fixture components in the line item if that is consistent with the IPA basis.

Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #20 ISG Appendix B.2, EPRI wrote the following regarding the staffs basis for The staff did not accept EPRI MRP Comments #20 and and #21 GALL-SLR Table adding IASCC as a listed aging mechanism for CE #21 or the EPRI MRPs collective statements in the two IV.B3, Items design core support barrel lower flange welds (LFWs) in comments that neither irradiation-assisted stress corrosion IV.B3.RP-333 and the staffs draft updates of GALL-SLR Items IV.B3.RP- cracking (IASCC) nor irradiation embrittlement (IE) should IV.B3.RP-333a 333 and for deleting GALL-SLR item IV.B3.RP-333a on be attributed to CE design core support barrel lower girth the topic of loss of material due to neutron irradiation welds (LGWs)/lower flange welds (LFWs).

(Pages 8 and 9 of embrittlement in the LFWs:

the appendix) In EPRI MRPs response to Request for Additional For Comment #20 Like with overall comments 11 and Information (RAI) #26, Item a on EPRI Report MRP-227, 12, IASCC was included in MRP191, Rev. 2 because Rev. 1 (as submitted in EPRI Letter No. MRP 2017-27 the LGW/LFW and MGW were combined on the same dated October 16, 2017, ADAMS Accession line. IASCC is not expected at the elevation of the No. ML17305A056), EPRI identified that IASCC and IE LGW/LFW. are applicable irradiation mechanisms for CE-design LGWs/LFWs. Although the EPRI MRP may have For Comment #21 made in relation to IE of the LFWs, performed further studies to exclude IASCC and IE as Similar discussion to overall comments 11, 12, and 20. applicable mechanisms for CE design core support barrel LGWs/LFWs in the upcoming MRP-227, Rev.2, report, the report has yet to docketed with the NRC or reviewed or accepted by the staff.

Thus, for the final issuance of the ISG, IASCC will remain as a cited irradiation-based cracking mechanism for CE-design core support barrel LGWs/LFWs in the ISGs update of GALL-SLR Report item IV.B3.RP-333 and IE will remain as a cited non-cracking, irradiation-based aging mechanism for CE-design core support barrel LGWs/LFWs in the ISGs update of GALL- SLR Report item IV.B3.RP-333a.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #24, For #24, ISG EPRI made the following editorial comments in regard to The staff did not accept EPRI Comments #24, #30, and

1. 30, and #33 Appendix B.4, the staffs update of GALL-SLR Report item IV.E.R-44 in #33.

GALL-SLR Appendix B.4 of the ISG and updates of SRP-SLR Table Table IV.E, Item 3.1-1, items 114, 118, and 119 in Appendix A of the ISG: The staff defines and discusses its criteria for New (N),

IV.E.R-44 Modified (M), Edited (E), or Deleted (D) AMR item Should this line be marked as "M" for modified rather designations in Section 1.2 of NUREG-2192 (the SRP-(Page 2 of ISG than "N"? Change to M. SLR). For comparisons to the Table 1 AMR summary Appendix B.4) items for PWR RVI components in Table 3.1-1 (as updated in Appendix A of the ISG), the N, M, E, or D For #30, ISG designations for AMR summary items are made in Appendix A, SRP- comparison to the versions of the AMR summary items in SLR Table 3.1-1, Table 3.1-1 in the prior license renewal SRP, (i.e.,

Items 114, 118, SRP-LR, Rev. 2), not in comparison to the version of and 119 these items in the SRP-SLR Report.

(Pages 10 and 11 Similarly, for comparisons to the Table 2 type AMR items of ISG Appendix A) for PWR RVI components in Tables IV.B2, IV.B3, IV.B4, and IV.E (as updated in Appendices B.1, B.2, B.3 and B.4 For #33, ISG of the ISG, respectively), the N, M, E, or D Appendix B.3, designations for AMR items are made in comparison to various AMR Items the versions of the AMR items in Tables IV.B2, IV.B3, IV.B4.R-423 (as IV.B4, and IV.E in the previous license renewal GALL referenced to SRP- Report (i.e., GALL Report, Rev. 2), and not the versions of SLR Item 118) and these AMR items in the corresponding tables of the IV.B4.R-424 (as GALL-SLR Report.

referenced to SRP-SLR Item 119)

(Pages 12 and 13 in ISG Appendix B.3)

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #25 ISG Appendix D, EPRI made the following editorial comment in regard to The staff accepted EPRI MRP Comment #25.

GALL-SLR AMP MRP-227 references in GALL AMP XI.M16A, PWR Specifically, the staff performed a review of the individual XI.M16A, PWR Vessel Internals, as updated in ISG Appendix D: MRP-227, Revision 1-A references in the update of Vessel Internals - GALL-SLR Report AMP XI.M16A, PWR Vessel Internals, Program The references to MRP227 are mixed between "MRP in Appendix D of the ISG versus those stating MRP-227 Description and 227, Revision 1A" and "MRP227 (as supplemented)". (as supplemented) in the updated version of the AMP.

Program Elements In multiple cases, it seems like this should be the second The staff found at least one case where the referenced in the AMP one. MRP-227 terminology in the updated version of GALL-SLR AMP XI.M16A, PWR Vessel Internals, should be switched to either MRP-227, Rev. 1-A or MRP-227 (as supplemented) consistent with the comment statement.

But to clarify, the AMP program description and program element criteria reference MRP-227, Revision 1-A directly if the context of the sentence is referring to criteria in MRP-227, Rev. 1-A when used as a starting point for the AMP. In contrast, the revisions of the AMP specify MRP-227 (as supplemented) if the I&E criteria in MRP-227, Rev. 1-A for a given component are being amended or supplemented by criteria in supplemental methodologies. Examples of the latter case are the citing and use of MRP-2018-022 for the gap analysis of the past Turkey Point and Surry SLRAs, or the past citing in the Surry SLRA of MRP-2019-023 as a supplemental one-time inspection protocol for the core barrel middle and lower axials welds in the units.

For implementation criteria, the staff should be referencing Chapter 7 of the MRP-227, Rev. 1-A because that section establishes the most up-to-date staff-approved implementation criteria for these types of living programs.

The staffs understanding is that Chapter 7 in MRP-227, Rev. 1-A would allow supplemental methodologies to be incorporated and used by the licensee for the living programs on an as-needed basis.

The staff has updated the MRP-227 terminology references in the updated version of the AMP accordingly.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #26 For #26, ISG EPRI made the following comments for these AMP The staff accepted Comments #26 and #27. The staff has and #27 Appendix D, GALL- sections in ISG Appendix D: updated the MRP-228 references in the staffs update of SLR AMP XI.M16A, GALL-SLR Report AMP XI.M16A, PWR Vessel Internals, PWR Vessel MRP-228 is currently at Revision 3. Update reference and FSAR Supplement Example for the AMP in Internals, to Revision 3. Appendix D of the ISG to cite Revision 3 of the MRP-228 References Section report.

(Page 8 of the This includes the staffs correction of the MRP-228 appendix) reference in the References section of AMP XI.M16A to be that for MRP-228, Rev. 3 report, and to specify it as For #27, ISG EPRI Proprietary Topical Report No. 3002010399, Appendix D, Materials Reliability Program: Inspection Standard for GALL-SLR PWR Internals (MRP-228, Rev. 3) (Non-publicly available Table X-01 FSAR in ADAMS Accession No. ML19081A064),

Supplement November 2018. A non-proprietary version of the Example for GALL- MRP-228, Revision 3 report may be accessed by SLR AMP XI.M16A members of the public at ADAMS Accession No. ML19081A058.

(Page 10 of the appendix)

EPRI MRP #28 ISG and ISG EPRI made the following generic comment for the The staff accepted EPRI MRP Comment #28. The staff Appendices contents of the ISG: found five instances in the ISG where the words EPRI Technical Report should be replaced with the words

" EPRI Technical Report was changed to EPRI Topical EPRI Topical Report. The staff has made the Report in two locations in Appendix D, but not changed appropriate adjustments of the terminology in the ISG in several other locations within the document. Change consistent with EPRI MRPs request in Comment #28.

EPRI Technical Report to EPRI Topical Report.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #29 ISG Appendix B.3, EPRI made the following generic comments for the EPRI MRP Comments #29 and #32 are somewhat and #32 Various AMR line contents of the updated AMR line items for GALL-SLR analogous to NEIs comment made in NEI Comment #3.

items for Report Table IV.B4 in Appendix B.3 and SRP-SLR Table The staff partially accepts the comments and rationale GALL-SLR Table 3.1-1 in Appendix A of the ISG: made in EPRI MRP Comments #29 and #32. Consistent IV.B4 in the with the staffs basis for responding to NEI Comment #3, appendix ISG Comment 29: There are several 'GALLSLR Item' the staff does not always use aging effect and mechanism Appendix A, and entries in this table for the B&W units (IV.B4.RP) that bases in the MRP-277, Revision 1-A report as the sole seem to identify aging effects/mechanisms that are not basis for establishing the aging effects and mechanisms Various AMR line consistent with MRP227, Rev. 1A. For example, GALL for referenced SRP-SLR AMR line items that were items for SRP-SLR SLR Item IV.B4.RP242a is an entry for the CRGT updated in Appendix A of the ISG or GALL-SLR Report Table 3.1-1 in the spacer castings (ID 51a). Per MRP227, Rev. 1A, the AMR line items updated in Appendices B.1, B.2, B.3 or appendix. CRGT spacer castings are only potentially susceptible to B.4 of the ISG.

TE, but the Table 3.11 entry for IV.B4.RP242a (ID 51a) states that the 'aging effect/mechanism' is 'cracking due Unlike the staffs review of the AMR line items for to SCC, IASCC, fatigue'. As this is Objective 1 for this Westinghouse-designed PWR RVI components in which SLR ISG, review the entries in Table 3.11 to ensure the staff used lessons learned from previous SLRA they identify aging effects/mechanisms consistent with reviews, the staff does not have any current 80-year MRP227, Rev. 1A. lessons learned criteria that the staff can apply to the assessment of the AMR line items for B&W-designed RVI Comment 32: IV Table B4 provides various items and components, as the staff has yet to approve any docketed lists their applicable aging effect/mechanism in one of SLRAs for B&W PWRs.

the columns in the table. It appears there are some aging effects/mechanisms that are not consistent with For this reason, the staff adjusted the AMR items for B&W MRP227, Rev. 1A in several table entries. For designed components to reflect the guidance in MRP-227, example, item IV.B4.RP249a (baffle plates) identifies Rev. 1-A.

the aging effect/mechanism of 'cracking due to IASCC or fatigue'. Per MRP227, Rev. 1A, the baffle plates are For the example in Comment #29 regarding GALL-SLR only potentially susceptible to IE. Report item IV.B4.RP-242a, the adjustment reverted the item to what is listed in table IV.B4 of the original GALL-SLR Report. Therefore, these items and other B&W AMR items that were adjusted in this final ISG to reflect the original GALL-SLR Report have been removed from the final version of this ISG.

For the example in Comment #32 regarding GALL-SLR Report items IV.B4.RP-249a, the adjustment resulted in deletion of the AMR item from the GALL-SLR Report, as reflected in this final ISG.

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Comment #(s) ISG Section/Page Comment NRC Staff Response SLR-ISG-2021-01-PWRVI: Appendix H EPRI MRP #31 ISG Appendix B.3, EPRI provided the following comment as the for The staff did not accept the comment statements made by GALL-SLR Table modifying the listed materials in GALL-SLR Item EPRI MRP in EPRI MRP Comment #31. The staff IV.B4, Items IV.B4.RP-247 and IV.B4.RP-247c. acknowledges that the line item entry for B&W-design IV.B4.RP-247 and lower core barrel (LCB) bolts in Table 3-1 of MRP-227, IV.B4.RP-247c ONS1 has 12 additional LCB bolts, which are Rev. 1-A reports cites the bolt material as being an A-286 fabricated from Type 304 stainless steel (see type stainless steel or X-750 nickel alloy material.

(Page 2 of the ISG ML003708443, pages 212 and 223). These entries in However, the staff has decided to list the more generic appendix) IV Table B4 for the LCB bolts state the material is stainless steel and Nickel alloy material designations for "stainless steel, nickel alloy". As the additional 12 Type B&W design lower core barrel (LCB) bolts in GALL-SLR 304 LCB bolts at ONS1 are not included in MRP227, Report items IV.B4.RP-247 and IV.B4.RP-247c, or other Rev. 1A for examination (see Table 31, page 321 types of B&W-design bolts that EPRI had indicated are where LCB bolts are specifically identified as Alloy A286 made from the specific materials (e.g., UCB bolts covered or Alloy X750), this material description of 'stainless by item IV.B4.RP-248, etc.).

steel' could be confusing and/or misinterpreted.

The first reason for citing the more generic classifications of the materials is that the Type A-286 stainless steel materials do not have an individual line item in GALL-SLR Report Table IX.C, Materials, and are instead a subset of the stainless steel material definition in the table; similarly, Type X-750 materials do not have an individual material definition in GALL-SLR Report Table IX.C and are identified as a subset of the generic Nickel alloys material definition in the table.

Secondly, the AMR item for inspecting the LCB bolts and their bolt locking devices is provided in Item B8 in Table 4-1 of the MRP-227, Rev. 1-A report. Item B8 includes Note 6 on the line item, which states: A minimum of 75% of the total population (examined + unexamined),

including coverage consistent with the Expansion criteria in Section 5.3 of this document, must be examined for inspection credit. The footnote on the B8 line item does not define total population in terms of a specified bolt material type. In addition, the AMR line item criteria in the GALL-SLR report are included only for potential LRA or SLRA AMR identification and application objectives and do not have any bearing on how an applicant or licensee would interpret the MRP-227, Rev. 1-A inspection or evaluation (I&E) guidelines under the scope of the Page 20 of 20 licensees PWR Vessel Internals Program. Thus, for the LCB bolt example, it is the licensees responsibility to determine what is appropriate for LCB bolt inspections in order to achieve EPRIs specified 75% bolt population criterion under Item B8 of Table 4-1 in MRP-227, Rev. 1-A report. The same logic applies to other B&W bolt types.