L-2021-147, Subsequent License Renewal Application - Aging Management Supplement 3 Revision 1

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Subsequent License Renewal Application - Aging Management Supplement 3 Revision 1
ML21207A066
Person / Time
Site: Point Beach  NextEra Energy icon.png
Issue date: 07/26/2021
From: Maher W
Point Beach
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2021-147
Download: ML21207A066 (51)
v • d • e


Text

NEXTera ENERGY ~

POINT BEACH July 26, 2021 L-2021-147 10 CFR 54.17 U.S. Nuclear Regulatory Commission Attention: Document Control Desk 11545 Rockville Pike One \Vhite Flint North Rockville, MD 20852-2746 Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR 27 SUBSEQUENT LICENSE RENEWAL APPLICATION -AGING MANAGEMENT SUPPLEMENT 3REVISION1

References:

1. NextEra Energy Point Beach, LLC (NEPB) Letter NRC 2020-0032 dated November 16, 2020, Application for Subsequent Renewed Facility Operating Licenses (ADAMS Package Accession No. ML20329A292)
2. U.S. Nuclear Regulatory Commission (NRC) Letter dated Janua1y 15, 2021, Point Beach Nuclear Plant, Units 1 and 2 - Determination of Acceptability and Sufficiency for Docketing, Proposed Review Schedule, and Notice of Opportunity to Request a Hearing Regarding the NextEra Energy Point Beach, LLC Application for Subsequent License Renewal (EPID No. L-2020-SLR-0002)

(ADAMS Accession No. IvIL21006A417)

3. NRC Letter dated January 15, 2021, Point Beach Nuclear Plant, Units 1 and 2 - Aging Management Audit Plan Regarding the Subsequent License Renewal Application Review (ADAMS Accession No. ML21007A260)
4. NEPB Letter L-2021-113 dated May 27, 2021, Subsequent License Renewal Application -Aging Management Supplement 3 (ADAMS Accession No. IvIL21147A115)

NEPB, owner and licensee for Point Beach Nuclear Plant (PBN) Units 1 and 2, has submitted a subsequent license renewal application (SLRA) for the Facility Operating Licenses for PBN Units 1 and 2 (Reference 1).

On Januaty 15, 2021, the NRC determined that NEPB's SLRA was acceptable and sufficient for docketing and issued the regulato1y audit plan for the aging management portion of the SLRA review (References 2 and 3). During this audit conducted between January 19, 2021 to March 26, 2021, NEPB agreed to supplement the SLRA (Enclosure 3, Attachment 1 of Reference 1) with new or clarifying information. The attachment to this letter revises the information provided in Reference 4.

For ease of reference, the attachments topic index is provided on page 3 of this letter. The attachments contain changes being made the to the Point Beach SLRA that were identified after submittal of the SLRA.

The changes are described along with the affected section(s) and page number(s) of the docketed SLRA (Enclosure 3 Attachment 1) where the changes are to apply. For clarity, revisions to the SLRA are provided with deleted text by strikethroughs and inserted text by bold red underline.

NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk L-2021-147 Page 2 Should you have any questions regarding this submittal, please contact me at (561) 304-6256 or William.Maher@fpl.com.

I declare under penalty of perjury that the foregoing is trne and correct.

Executed on the 26 1" day of July 2021.

Sincerely, William Digitally signed by William Maher DN: cn=Willlam Maher, o=Nuclear, ou=Nuclear licensing Projects, emall=willlam.mah er@fpl.com, c=US Maher Date: 2021.07.26 10:23:45 *04'00' William D. Maher Licensing Director - Nuclear Licensing Projects Cc: Administrator, Region III, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC Public Service Commission \X!isconsin

Document Control Desk L-2021 -147 Page 3 Attachments Index Attachment Topic No.

Revision 1 to SLRA Supplement 3 - Incotporation of Interim Staff guidance SLR-ISG-2021-1 01-P\WVI 2 Editorial Revision to SLRA Section 4.2.5 3 Addition of Station Blackout (SBO) Recovery Path Drawings

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 1 of 41 Revision 1 to SLRA Supplement 3 - Incorporation of SLR-ISG-2021-01-PWRVI This attachment supersedes in its entirety Revision 0 of Supplement 3 to the SLRA which was included as an attachment to NEPB Letter L-2021-113 dated, May 27, 2021 (ADAMS Accession No. ML21147A115).

Affected SLRA Sections/Tables: Section 2.1.6, Section 2.1.6.4 (new), Table 2.3.1-2, Section 3.1.2.1.2, Section 3.1.2.2.9, Table 3.1-1, Table 3.1.2-2, Table 4.1.5-2, Table 16-3, Section 8.1.1, Table 8-4, Section 8.2.3.7, Section C.1.0 SLRA Page Numbers: 2.1-31, 2.1-32, 2.1-35, 2.3-7, 3.1-3, 3.1-14, 3.1-15, 3.1-32, 3.1-33, 3.1-41through3.1-43, 3.1-46, 3.1-56 through 3.1-58, 3.1-75 through 3.1-86, 4.1-6, A-67, 8-5, 8-18, 8-73, C-3 Description of Change:

The SLR-ISG provides interim guidance to subsequent license renewal applicants which is incorporated into the following SLRA Sections.

  • Section 2 Section 2 is revised to identify the Interim Staff Guidance which is being incorporated by this Supplement and outline the changes made to NUREG-2191 and NUREG-2192 .

Table 2.3.1-2 is updated to reflect accurate component names.

  • Section 3 Section 3.1.2.2 is updated to reflect accurate aging effects and aging management programs .

Section 3.1.2.2.9 is updated to incorporate the changes made to the further evaluation by the Interim Staff Guidance.

Table 3.1-1 is revised to account for changes in inspection and examination (l&E) criteria for PWR reactor vessel internals (RVI) components made in MRP-227, Revision 1-A, and in other relevant industry documents.

Table 3.1.2-2 is revised to incorporate the Interim Staff Guidance, recognize the role of the Water Chemistry AMP in reactor vessel internals aging management, and make editorial changes .

  • Section 4 Table 4.1.5 2 is revised to incorporate the Interim Staff Guidance.
  • Appendix A Table 16-3 row 11 is revised to make editorial changes .
  • Appendix B Section B.1 .1 is revised to reflect that there is no longer an exception to the Reactor Vessel Internals AMP due to incorporation of the Interim Staff Guidance.

Table B-4 is revised to reflect that there is no longer an exception to the Reactor Vessel Internals AMP due to incorporation of the Interim Staff Guidance.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 2 of 41 Section B.2.3.7 is revised to reflect that there is no longer an exception to the Reactor Vessel Internals AMP due to incorporation of the Interim Staff Guidance.

  • Appendix C Section C.1.0 is revised to state that the Interim Staff Guidance has been incorporated.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page3of41 SLRA Section 2.1.6, Pages 2.1 -31 and 2.1-32, is revised as follows :

2.1.6. Interim Staff Guidance Discussion As discussed in NEI 17-01, the NRC has encouraged applicants to address Subsequent License Renewal Interim Staff Guidance (SLR-ISG) documents in the Subsequent License Renewal Applications (SLRA) . The following final SLR-ISGs have been issued for use and comment but have not been incorporated in NUREG-2191 or NUREG-2192 at the time of submittal :

  • SLR-ISG-2021-01 -PWRVI Updated Aging Management Criteria for (Reference ML20217L203) Reactor Vessel Internal Components for Pressurized-Water Reactors The following sub-sections provide summaries of how each of the SLR-ISGs are addressed in the SLRA.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 4 of 41 New SLRA Section 2.1.6.4 is added on page 2.1-35:

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

This SLR-ISG provides interim guidance to subsequent license renewal applicants for the following NUREG-2191 and NUREG-2192 Sections:

  • NUREG-2192, Table 3.1-1 The SLR-ISG revises NUREG-2192, Table 3.1-1 to account for changes in inspection and examination (l&E) criteria for PWR reactor vessel internals (RVI) components made in MRP-227, Revision 1-A, and in other relevant industry documents. The PBN RVI further evaluation items in Table 3.1-1 incorporate the guidance presented in this SLR-ISG.
  • NUREG-2191, Tables IV.82, IV.83 and IV.84 The SLR-ISG revises NUREG-2191, Tables IV.82, IV.83 and IV.84 to update the staff's guidance for RVI components to account for changes in l&E criteria for PWR RVI components made in MRP-227, Revision 1-A, and in other relevant industry documents. Tables IV.83 and IV.84 are revised to reflect changes for Combustion Engineering and Babcock & Wilcox designed RVI components, respectively, and are not applicable to PBN.

Table IV.82 is revised to reflect changes for Westinghouse designed RVI components and is applicable to PBN. The revisions in Table IV.82 have been incorporated into the PBN RVI AMR in Table 3.1.2-2.

  • NUREG-2192 Further Evaluation items 3.1.2.2.9 and 3.1.3.2.9 The SLR-ISG revises NUREG-2192 Further Evaluation items 3.1.2.2.9 and 3.1.3.2.9 to provide staff guidance for the acceptance criteria and review procedures, respectively, related to aging management of PWR RVI components. The revisions to item 3.1.2.2.9 have been incorporated into the PBN RVI AMR. Item 3.1.3.2.9 is not applicable to the PBN SLRA. Item 3.1.3.2.9 provides NRC staff review procedures and is not meant to be incorporated into an application.
  • NUREG-2191 AMP Xl.M16A, PWR Vessel Internals The SLR-ISG revises the AMP to incorporate the changes included in MRP-227, Revision 1-A. The PBN PWR Vessel Internals AMP (8.2.3.7) incorporates the guidance presented in this SLR-ISG.
  • NUREG-2191, Table IX.C The SLR-ISG revises NUREG-2191, Table IX.C to add "Stellite" material and its usage. This revision has been incorporated into the PBN RVI AMR, as appropriate.
  • NUREG-2192, Table 4.7-1 The SLR-ISG revises NUREG-2192, Table 4.7-1 to add "EPRI MRP cycle-based and fluence-based analyses in support of MRP-227" as an example of a plant-specific TLAA topic. Cycle-based fatigue for the PBN RVI is

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-14 7 Attachment 1 Page 5 of 41 included with the generic industry TLAA "Metal Fatigue of Class 1 Components" in SLRA Table 4.1.5.3 and Section 4.3.1. A PBN plant-specific RVI fluence-based analysis is not part of the PBN CLB and therefore does not meet the TLAA definition for SLR.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 6 of 41 SLRA Table 2.3.1-2, page 2.3-7, is revised as follows:

Table 2.3.1-2 Reactor Vessel Internals Components Subject to Aging Management Review Component Type Component Intended Function(s)

Alignment and interfacing components (clevis bearing Structural support Stellite wear surfaces)

Alignment and interfacing components (clevis insert Structural support bolts)

Alignment and interfacing components (clevis insert Structural support dowels)

Alignment and interfacing components (upper core plate Structural support alignment pins)

ASME Section Xl 1 examination catego~ B-N-3 Structural sui;mort reactor vessel internals components Baffle-former assembly (baffle plates, baffle edge bolts, Structural support former plates) Flow distribution Baffle-former assembly (baffle plates, former plates) Structural support Flow distribution Baffle-former assembly (baffle-edge bolts) Structural support Baffle-former assembly (baffle-former bolts) Structural suooort Bottom mounted instrumentation (column bodies) Structural support Bottom mounted instrumentation (flux thimble tubes) Structural support Pressure boundary Control rod guide tube assembly (guide cards) Structural support Control rod guide tube assembly (lower flange welds in Structural support peripheral assemblies)

Control rod guide tube assembl)l {lower flange welds Structural sui;mort in non-peripheral assemblies)

Core barrel assembly (barrel former bolts) Structural support Core barrel assembly (core barrel flange) Structural support Flow distribution GeFe 9aFFel asseA=191y EseFe 9aFFel el:ltlet R82:2:le welEl~ :;:;tFl:lStl:lFal Sl:lJ3J38R Core barrel assembly (lower axial welds) Structural support Core barrel assembly (lower flange weld) Structural support Core barrel assembly (lower girth weld) Structural support Core barrel assembly (middle axial welds) Structural support Core barrel assembly (upper axial weld) Structural support Core barrel assembly (upper flange weld) Structural support Core barrel assembly (upper girth weld) Structural support Lower core plate (fuel alignment pins) Structural support Lower internals assembly (lower core plate) Structural support Flow distribution

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 7 of 41 Table 2.3.1-2 Reactor Vessel Internals Components Subject to Aging Management Review Component Type Component Intended Function(s)

Lower internals assembly (lower support forging) Structural support Lower support assembly (lower support column bodies) Structural support Lower support assembly (lower support column bolts) Structural support Reactor vessel internals com12onents Structural su1212ort

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021 -14 7 Attachment 1 Page 8 of 41 SLRA Section 3.1.2.1.2, page 3.1 -3, is revised as follows :

Aging Effects Requiring Management The following aging effects associated with the reactor vessel internals require management:

  • Changes in dimensions
  • Cracking
  • Cumulative fatigue damage
  • Loss of fracture toughness
  • Loss of material
  • Loss of preload

--wear-Aging Management Programs The following AMPs manage the aging effects for the reactor vessel internals components:

  • Flux Thimble Tube Inspection (B.2 .3.24)
  • Reactor Vessel Internals (B.2.3.7)
  • Water Chemistry (B.2.3.2)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 9 of 41 SLRA Section 3.1.2.2.9, pages 3.1-14 through 3.1-15, is revised as follows :

3.1.2.2.9 Aging Management of PWR 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)" (Agency wide Documents Access and Management System (ADAMS) Accession Nos. ML12017A191 through ML12017A197 and ML12017A199), prov-ides provided the industry's current aging management initial set of aging management inspection and evaluation O&EJ recommendations for the reactor vessel internal (RV/)

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 l&E guidelines for the PWR RV/ 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 industry's 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 staff's 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 Ro/.iabi!ity Program

{MRP} identified that the following aging mechanisms may be applicable to the design of the RV/ components in these types of facilities: (a) stress corrosion cracking rscq , (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 ,

er-and (h) thermal or irradiation-enhanced stress relaxation or irradiation enhanced creep. The methodology in MRP 227 A vms approved by the NRG in a safety evaluation dated December 16, 2011 (ADAMS Accession No.

PAL11308A770), which includes those plant specific app/.icantl/icensee 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 MRP's functionality analysis and failure modes, effects, and criticality analysis bases for grouping Westinghouse-designed, B& W-designed and Combustion Engineering (CE)-designed RV/ components into these-the applicable inspection categories (as evaluated in MRP-227, Revision 1-A) was

~based on an assessment of aging effects and relevant 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

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 10 of 41 SLRA Section 3.1.2.2.9, pages 3.1-14 through 3.1-15, is revised as follows:

period of extended operation) . The EPRI MRP's has not assessed assessment 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 RV/ components in these designs, as defined in MRP-227, Revision 1-A orif!.&-the applicable MRP background documents (e.g., MRP-191 , Revision 1, for Westinghouse-designed or CE-designed RV/ components or MRP-189, Revision 2, for B& W-designed components) .

As described in GALL-SLR Report AMP Xl.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 RV/ components for the subsequent period of extended operation. However, to use this alternative basis, GALL-SLR Report AMP Xl.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 be £!!!1 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 Xl.M16A. If a gap analysis is needed to establish the appropriate aging management criteria for the RV/ components, the applicant has the option of including the gap analysis in the SLRA f&F..#5 reaetor 1:1Rit(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 AILA/s previously issued on MRP-227-A because the AJLA/s were resolved and closed by the staff in the April 25, 2019, safety evaluation for MRP-227, Revision 1-A. The sole AILA/ issued by the staff in the safety evaluation dated April 25, 2019, relates to an applicant's 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 applicant's resolution of this AILA/ can be appropriately addressed in the "Operating Experience" program element discussion for the AMP and in the applicant's basis document for the AMP, a separate SLRA Appendix C response for the AILA/ is unnecessary.

Alternatively, the PWR SLRA may define a plant-specific AMP for the RV/

components to demonstrate that the RV/ 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,

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page11of41 SLRA Section 3.1.2.2.9, pages 3.1-14 through 3.1-15, is revised as follows:

expansion criteria, and acceptance criteria are justified in the SLRA. +Re-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.

The PBN Reactor Vessel Internals AMP is based on the current MRP-227 Revision 1-A framework modified by an 80-year gap analysis. Appendix C of this application provides a detailed discussion of the RVI gap analysis. As enhanced, this program will continue to manage the effects of stress corrosion cracking, irradiation-assisted stress corrosion cracking, wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling, thermal and irradiation-induced stress relaxation, and irradiation creep, including any combined effects .

As a condition monitoring program, the PBN Reactor Vessel Internals AMP specifies inspection methods that are sufficient to detect aging effects, such as cracking, whether from a single aging mechanism or combination of mechanisms, prior to a component approaching a condition in which it may not be able to fulfill its intended functions ; and if such aging effects are detected, the evaluation and corrective action is required to consider the effects from any applicable mechanism in order to provide reasonable assurance that the component will continue to perform its intended function.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1Page12of41 SLRA Table 3.1-1, pages 3.1-32 and 3.1-33 , is revised as follows:

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Evaluation Discussion Number Effect/Mechanism Program Recommended (AMP)/TLAA 3.1-1 , 025 Steel (with nickel alloy cladding) Cracking due to primary AMP Xl.M2 , "Water Yes (SRP-SLR Not applicable.

or nickel alloy steam generator water sec Chemistry," and AMP Sections 3.1.2.2. 11 .1 Further evaluation is primary side components: Xl.M19 , and documented in divider plate and tube-to-tube "Steam Generators." In 3.1 .2.2.11.2) subsection 3.1.2.2.11 .

sheet welds exposed to reactor addition , a plant- specific coolant program is to be evaluated.

3.1-1 , 028 Westinghouse-specific Loss of material due to AMP Xl.M16A, "PWR Yes (SRP-SLR GeAsisteAt witt:i "Existing Programs" wear; cracking due to Vessel Internals," and Section 3.1.2.2.9) ~Jl::JREG ~~ 9~ . +Ae PE!~J components : Stainless steel, SCC , irradiatieA assisted AMP Xl.M2, "Water Reaster Vessel IAtemals nickel alloy VVestiAgAeuse, and ~IASCC , fatigue Chemistry" (for SCC t B - ~ . 3 . 7 j aAd l/lJater X-750 control rod guide tube mechanisms only) GAeA'listry t El - ~ . 3 - ~ j AMPs support pins {split pins},...aRG are used te A'laAage tt:le GeA'leustieA EAgiAeeriAg reaster vessel iAtemals tt:ierffial st:iield 13esitieAiAg 13iAs; u1313er eere 13late aAd lirsaley 4 GeffieustieA aligAA'leAt 13iAs. Not EAgiAeeriAg iAeere applicable. The control iAstruffieAtatieA thiA'lele tuees rod guide tubes are not exposed to reactor coolant and an "Existing Programs" neutron flux component. Further evaluation is documented in subsection 3.1.2.2.9.

3.1-1 , 029 Not applicable. This line item only applies to BWRs.

3.1-1 , 030 Not applicable. This line item only applies to BWRs.

3.1-1 , 031 Not applicable. This line item only applies to BWRs.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page13of41 SLRA Table 3.1-1 , pages 3.1-32 and 3.1-33 , is revised as follows :

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Evaluation Discussion Number Effect/Mechanism Program Recommended (AMP)/TLAA 3.1 1, 032 ~taiAless steel, AiGkel alley, eF CmskiAg, less ef mateFial AMP Xl.M1 , "A~Me Ne CeAsisteAt >NitR CA~£ FeasteF vessel iAternals , due te 'Nem £estieA XI IAsei:vise ~JldReG 2191 . +Re 12El~J GeFe su1313eFt stFUGtUFe tAet IAs13estieA, £u9sestieAs A£Me £estieA XI IAseFVise alFeady FefeFeAGed as IWEl , IWC , aAd IWD" IAs13eGtieA , £u9seGtieAs A£Me £estieA XI examiAatieA ll#El , p,11,1c , a Ad ll/lJD CategeFJ< El ~J 3 GeFe su1313eFt tEl.2.3.1 j AMP is used te stFUGtUFe Gem13eAeAts iA maAage GFaGkiAg aAd less MRP 227 Aj , ex13esed te ef mateFial iA FeaGteF i,tessel rnaGteF GeelaAt aAd AeutmA flux iAternal GeFe su1313eFt stFusturns ex13esed te FeaGteF seelaAt aAd AeutmA flt!*.-

3.1-1, 033 Stainless steel, steel with Cracking due to SCC AMP Xl.M1 , "ASME No Consistent with stainless steel cladding Class 1 Section XI lnservice NUREG-2191 . The PBN reactor coolant pressure Inspection, Subsections ASME Section XI lnservice boundary components exposed IWB , IWC, and IWD ," Inspection, Subsections to reactor coolant and AMP Xl.M2, "Water IWB , IWC, and IWD Chemistry" (B.2.3.17) and Water Chemistry (B.2.3.2) AMPs are used to manage sec in Class 1 reactor coolant pressure boundary components exposed to reactor coolant.

3.1 -1 , 034 Stainless steel, steel with Cracking due to SCC AMP Xl.M1 , "ASME No Not applicable. The PBN stainless steel cladding Section XI lnservice pressurizer relief tank is not pressurizer relief tank (tank shell Inspection , Subsections an ASME Section XI and heads, flanges , nozzles) IWB , IWC, and IWD," component. Cracking due exposed to treated borated and AMP Xl.M2 , "Water to sec in the stainless water >60°C (>140°F) Chemistry" steel pressurizer relief tank exposed to treated borated water >140°F is managed with item number 3.1-1 , 080.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 14 of 41 SLRA Table 3.1-1 , pages 3.1-41 through 3.1-43, is revised as follows :

Table 3.1-1: Summary of A~in~ Mana~ement Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program Evaluation (AMP)/TLAA Recommended 3.1-1 , 053a Stainless steel, nickel alloy Cracking due to SCC, AMP Xl.M16A, Yes (SRP-SLR Consistent with NUREG-2191 Westinghouse reactor internal irradiation assisted "PWR Vessel Internals," Section 3.1.2.2.9) as modified by "Primary" components exposed ~IASCC , fatigue and AMP Xl.M2, SLR-ISG-2021-01-PWRVI.

to reactor coolant, neutron flux "Water Chemistry" (for The Reactor Vessel Internals sec mechanisms only) (B.2.3.7) and Water Chemistry (B.2.3.2) AMPs are used to manage cracking due to sec, irradiation assisted sec and fatigue in reactor vessel internals "Primary" components exposed to reactor coolant and neutron flux. Note that many aging effects managed by the Reactor Vessel Internals (B.2.3.7) AMP are dispositioned through FMECA analysis and not inspected.

Further evaluation is documented in subsection 3.1.2.2.9.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page15 of41 SLRA Table 3.1-1 , pages 3.1-41 through 3.1-43 , is revised as follows :

Table 3.1-1: Summary of Ac:iinc:i Manac:iement Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program Evaluation (AMP)/TLAA Recommended 3.1-1 , 053b Stainless steel Westinghouse Cracking due to SCC , AMP Xl.M16A, Yes (SRP-SLR Consistent with NUREG-2191 reactor internal "Expansion" irradiation assisted "PWR Vessel Internals, Section 3.1.2.2.9) as modified by components exposed to reactor .£.GGIASCC, fatigue and AMP Xl.M2, SLR-ISG-2021-01-PWRVI.

coolant and neutron flux "Water Chemistry" (for The Reactor Vessel Internals sec mechanisms only) (B.2.3.7) and Water Chemistry (B.2.3.2) AMPs are used to manage cracking due to sec, irradiation assisted sec and fatigue in reactor vessel internals "Expansion" components exposed to reactor coolant and neutron flux. Note that many aging effects managed by the Reactor Vessel Internals (B.2.3.7) AMP are dispositioned through FMECA analysis and not inspected .

Further evaluation is documented in subsection 3.1.2.2.9.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page16of41 SLRA Table 3.1-1, pages 3.1-41 through 3.1-43 , is revised as follows:

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program Evaluation (AMP)/TLAA Recommended 3.1-1 , 053c Stainless steel, nickel alloy....QI Cracking due to SCC, AMP Xl.M16A, Yes (SRP-SLR Consistent with NUREG-2191 stellite Westinghouse reactor irradiation assisted "PWR Vessel Internals," Section 3.1 .2.2.9) as modified by internal "Existing Programs" SGGIASCC, fatigue and AMP Xl.M2, SLR-ISG-2021-01-PWRVI.

components exposed to reactor "Water Chemistry" (for The Reactor Vessel Internals coolant, neutron flux sec mechanisms only) (B.2.3 .7) and Water Chemistry (B.2.3.2) AMPs are used to manage cracking due to sec, irradiation assisted sec and fatigue in reactor vessel internals "Existing Programs" components exposed to reactor coolant and neutron flux. Note that many aging effects managed by the Reactor Vessel Internals (B.2.3.7) AMP are dispositioned through FMECA analysis and not inspected .

Further evaluation is documented in subsection 3.1.2.2.9.

3.1-1 , 054 Stainless steel Westinghouse- Loss of material due to AMP Xl.M37, No Consistent with NUREG-2191 design bottom mounted wear "Flux Thimble Tube as modified by instrument system flux thimble Inspection" SLR-ISG-2021-01-PWRVI.

tubes (with or without chrome The Flux Thimble Tube plating) exposed to reactor Inspection (B.2.3 .24) AMP is coolant and neutron flux used to manage loss of material due to wear in stainless steel bottom mounted instrument system flux thimble tubes exposed to reactor coolant and neutron flux.

3.1-1 , 055a Not applicable. This line item only applies to Babcock and Wilcox designs.

3.1-1 , 055b Not applicable. This line item only applies to Combustion Engineering designs.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 17 of 41 SLRA Table 3.1-1 , page 3.1-52 , is revised as follows:

Table 3.1-1: Summary of A~in~ Mana~ement Evaluations for the Reactor Vessel , Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program Evaluation (AMP)/TLAA Recommended

3. 1-1 , 059c Stainless steel (SS , including Loss of fracture AMP Xl.M16A, Yes (SRP-SLR Consistent with CASS , PH SS or martensitic toughness due to neutron "PWR Vessel Internals" Section 3.1.2.2.9) NUREG-2191 as modified b~

SS),, ei: nickel alloy, or stellite irradiation embrittlement SLR-ISG-2021-01-PWRVI. The Westinghouse reactor internal and for CASS, martensitic Reactor Vessel Internals "Existing Programs" SS, and PH SS due to (B.2.3.7) AMP is used to components exposed to thermal aging manage reactor vessel reactor coolant and neutron embrittlement; changes in internals "Expansion" flu x dimensions due to void components exposed to swelling , distortion; loss of reactor coolant and neutron preload due to thermal flu x. Note that many aging and irradiation-enhanced effects managed by the stress relaxation, creep ; Reactor Vessel Internals loss of material due to (B.2.3.7) AMP are wear dispositioned through FMECA analysis and not inspected.

Further evaluation is documented in subsection 3.1.2.2.9.

3.1-1 , 060 Not applicable. Th is line item only applies to BWRs.

3.1-1 , 061 Steel steam generator steam Wall thinning due to AMP Xl.M17 , No Consistent with NUREG-2191 .

nozzle and safe end , feedwater flow-accelerated corrosion "Flow-Accelerated The Flow-Accelerated nozzle and safe end , AFW Corrosion" Corrosion (B.2.3.8) AMP is nozzles and safe ends used to manage wall thinning exposed to secondary due to flow accelerated feedwater/steam corrosion in the steam generator feedwater nozzle and steam outlet nozzle exposed to secondary feedwater/steam .

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page18of41 SLRA Table 3.1-1 , page 3.1-52, is revised as follows:

Table 3.1-1 : Summary of Aging Management Evaluations for the Reactor Vessel , Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program Evaluation (AMP)/TLAA Recommended

3. 1-1, 087 Stainless steel , nickel alloy Loss of material due to AMP Xl.M2, "Water No Consistent with NU REG-PWR reactor internal pitting , crevice corrosion Chemistry" 2191. The Water Chemistty components exposed to reactor (B.2.3.2} AMP is used to coolant, neutron flux manage loss of material due to gitting and crevice corrosion in stainless steel and nickel allo:l£ reactor vessel internal comgonents exgosed to reactor coolant and neutron flux. Net a13131ica91e. bess ef ffiateFial feF FeacteF i.1essel iAternal ceffiJ:leAeAts ex13esed te rnacteF ceelaAt aAd Aet1trnA flt1x is addFessed iA rnws :u 1, Ga9a, :u 1, Ga99 , :u 1, Ga9c, aAd a.1 1, 119.

Wi=lile ti=lese iteffis addFess less ef ffiateFial dt1e te weaF, ti=le Feactm vessel iAternals AMR de13eAds eA ti=le scrneAiAg 13eFfeFffied iA MRP 191 'NAiCA dees Aet distiAgt1 isi=l less ef Ff!ateFial Ff!eci=laAisffis.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1Page19of41 SLRA Table 3.1-1 , pages 3.1-56 through 3.1-58 , is revised as follows:

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program Evaluation (AMP)/TLAA Recommende d

3.1 -1 , 11 3 Not applicable. Th is line item only appl ies to BWRs.

3. 1-1, 114 Reactor coolant system Cracking due to SCC, AMP XJ.M1 , "ASME No Consistent with NUREG-2191 as components defined as IGSCCI PWSCCI IASCC Section XI lnservice modified by ASME Section XI Code Class (SCC mechanisms for Inspection , Subsections SLR-ISG-2021-01-PWRVI. The components (ASME Code stainless steel, nickel IWB , IWC , and IWD," ASME Section XI lnservice Class 1 reactor coolant alloy components only)1 and AMP Xl.M2, lnsQection 1 Subsections IWB 1 pressure boundary fatigue 1 or cyclic loading ; "Water Chemistry" IWC 1 and IWD (B.2.3.1} and components 1 reactor loss of material due to (water Water Chemistrt (B.2.3.2} AMPs vessel interior general corros ion (steel chemistry- related or are used to manage cracking attachments 1 or core support only), pitting corros ion , corrosion- related aging due to sec IGSCC 1 PWSCC 1 structure components, .;__or crevice corrosion , or wear effect mechan isms IASCC (SCC mechanisms for ASME Class 2 or 3 only) stainless steel 1 nickel alloy components - including comQonents only} 1 fatigue 1 or ASME defined appurtenances, cyclic loadingj loss of material component supports, and due to Qitting corrosion 1 crevice associated pressure boundary corrosion 1 and wear in reactor welds , or components subject vessel internal comQonents to plant-specific equivalent exQosed to reactor coolant and classifications for these neutron flux. ~Jet l:lse9.

ASME code classes) All Felei,,<aAt a§l iA§l FAeGhaAisFAs FeE11:l iFiA§l FAaAa§leFAeAt ey A~Me ~eGtieA :XI IAseF\<iGe IAsr:ieGtieA, ~l:leseGti eAs IWB , IWG ,

aA9 IWQ ( B . 2 . 3 . ~ j 9F l,IVateF GheFAistry (B.2.3.2j arn FeG9§lAize9 l:JSiA§l liAe iteFAS FA9Fe Sf38GifiG te

  • '- - . ,;,..J

~ ;, I - - - . ...... . . .

  • - ~

~

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 20 of 41 SLRA Table 3.1-1 , pages 3.1-56 through 3.1-58, is revised as follows :

Table 3. 1-1: Summary of A!'.tin!'.1 Mana!'.lement Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Meehan ism Program Evaluation (AMP)/TLAA Recommende d

3.1-1 , 118 Stainless steel , nickel alloy Cracking due to SCC, Plant-specific aging Yes (SRP-SLR Not applicable. Cracking due to PWR reactor vessel internal irradiation assisted management program Section 3.1.2.2.9) sec, irrad iation-assisted sec, components or ~IASCC , cyclic or AMP Xl.M16A, cyclic load ing, and fatigue of LRA/SLRA-s12ecified load ing , fatigue "PWR Vessel stainless steel, nickel alloy PWR reactor vessel internal Internals," and AMP reactor vessel internal components com12onent exposed to reactor Xl.M2, "Water exposed to reactor coolant, coolant, neutron flux Chemistrt" {SCC and neutron flu x is addressed in rows IASCC onl~}, with an 3.1-1 , 053a, 3.1-1 , 053b, and adjusted site-s12ecific 3.1-1 , 053c. The associated or com12onent-s12ecific NUREG-2191 aging items are not aging management used.

basis for a s12ecified reactor vessel internal comoonent

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page21of41 SLRA Table 3.1-1, pages 3.1-56 through 3.1-58, is revised as follows:

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Aging Management Further Discussion Number Effect/Mechanism Program Evaluation (AMP)/TLAA Recommende d

3.1-1 , 119 Stainless steel, nickel alloyi Loss of fracture Plant-specific aging Yes (SRP-SLR Consistent with NUREG-2191 as stellite PWR reactor vessel toughness due to neutron management program Section 3.1.2.2.9) modified by internal components or irradiation embrittlement or AMP Xl.M16A 1 SLR-ISG-2021-01-PWRVI for le&&

LRA/SLRA-sQecified reactor or thermal aging "PWRVessel ef fFaGtl:JFe tel:J§lARess aAEl GAaA§les vessel internal comQonent embrittlement; changes in Internals," with an iA EliFfleAsieAs loss of material in exposed to reactor coolant, dimensions due to void adjusted site-sQecific the stainless steel UQQer and neutron flux swelling or distortion ; loss or comQonent-sQecific lower core Qlate fuel alignment of preload due to thermal aging management Qins as well as the stellite radial and irradiation-enhanced basis for a SQecified SUQQOrt keys and UQQer core stress relaxation or creep ; reactor vessel internal Qlate alignment Qins. bess ef loss of material due to comQonent fFaGtl:lFe te1:J§lAAess aAEl GAaA§les iA wear EliFfleAsieA foF staiAless steel FeasteF >Jessel iAternals G9Ffl~9AeAtS is Ffl8Aa§leEl SJ' tAe ReasteF Vessel IAternals t B - ~ . 3 . 7 ~

AMP . bess ef ~Feleae is Aet a~~lisasle. bess ef FflateFial is aElElrnsseEl witt:i iteFfl Al:JFflBeF 3.1 1, Ga4- Note that many aging effects managed by the Reactor Vessel Internals (B.2.3.7) AMP are dispositioned through FMECA analysis and not inspected.

Further evaluation is documented in subsection 3.1.2.2.9.

3.1-1, 120 Not applicable. This line item only applies to BWRs.

3.1-1 , 121 Not applicable. This line item only applies to BWRs.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1Page22 of41 SLRA Table 3.1.2-2, pages 3.1-75 through 3.1-86, is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of A~in :i Mana~ement Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management Alignment and Structural ~ Reacter ceelant Cracking Reacter Vessel Internals NeRe NeRe ~

interfacing Sl:lppert Ne1:Jtren flux (B.2.3 .7) cempenents (clevis bearing Stellite wear

-" ,\

Alignment and Structural Stellite Reactor coolant Loss of Reactor Vessel Internals NefielV.82.RP- Nefie3.1-1, ~A interfacing support Neutron flux material (8 .2.3.7) 285 059c components Changes in (clevis bearing dimension Stellite wear surfaces)

Alignment and Structural Nickel alloy Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-399 3.1-1 , 053c gA , 1 interfacing support Neutron flux (8.2.3.7) components Water Chemistry § (clevis insert bolts) (8 .2.3.2)

Alignment and Struct1:Jral Nickel alley Reacter ceelant Cracking ASMe Sectien XI IV.B2.RP 382 3.1 1, 032 A interfacing suppert Neutren f11:Jx Less ef lnservice lnspectien, cempenents material Si:Jbsectiens IWB, IWC, r-*- *'-*- .. *-- ... h--- " - \

~--

--rl llMl""I I C "l '>

-*-*-* ~ \

Alignment and Structural Nickel alloy Reactor coolant Loss of Reactor Vessel Internals IV.82.RP-285 3.1-1 , 059c gA , 1 interfacing support Neutron flux material (8 .2.3.7) components Loss of preload (clevis insert bolts) Changes in dimension Alignment and Structmal Nickel alley Reacter ceelant Cracking ASMe Sectien XI IV.B2.RP 382 3.1 1, 032 A interfacing sup pert Neutren flux Less ef lnservice lnspectien ,

cempenents material Subsectiens IWB, PNC, (clevis insert and IVVD (B.2.3.1 )

...i~. *~1 Alignment and Structural Nickel alloy Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-399 3.1-1 , 053c gA , 1 interfacing support Neutron flux (8.2.3.7) components Water Chemist!:Y (clevis insert (8.2.3.2) dowels)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 23 of 41 SLRA Table 3.1.2-2, pages 3.1-75 through 3.1-86, is revised as follows :

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management Alignment and Structural Nickel Reactor Loss of Reactor Vessel IV.82.RP-285 3.1-1, 059c c, 1 interfacing su1212ort alloy coolant material Internals {B.2.3.7}

com12onents Neutron flux Changes in

{clevis insert dimension dowels\

Al ignment and Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP- 3.1-1 , 053c DA interfacing support steel Neutron flux (B.2.3.7) ~301 components Water Chemistry § (upper core plate (B.2 .3.2) aliqnment pins)

Al ign ment and Structural Stainless Reactor coolant Loss of Reactor Vessel Internals IV.82.RP-299 3.1-1 , 059c BA interfacing support steel Neutron flux material (B.2.3.7) components Water Chemistry § (upper core plate (B.2.3.2) aliqnment pins)

Alignment and Structural Stellite Reactor coolant Loss of Reactor Vessel Internals ~leRe l).(B2.R NeAe3.1-1, J;-JA interfacing support Neutron flux material (B.2.3.7) 424 119 components (upper core plate aliqnment pins)

ASME Section XI, Structural Nickel Reactor Cracking ASME Section XI IV.E.R-444 3.1-1, 114 A examination su1212ort alloy coolant Loss of lnservice lns12ection, catego!:Y B-N-3 Neutron flux material Subsections IWB, IWC, reactor vessel and IWD {B.2.3.1}

internals Water Chemist!'.¥ comoonents lB.2.3.2\

Baffle-former Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.B2.RP-270a 3.1-1 , 053a BA assembly (baffle support steel Neutron flux (B.2.3.7) plates, baffle edge Flow Water Chemistry § bolts, former distribution (B.2.3.2) plates)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 24 of 41 SLRA Table 3.1.2-2, pages 3.1-75 through 3.1-86 , is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management Baffle-former Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.B2.RP-387 3.1-1 , 053a Qf assembly (baffle support steel Neutron flux (B.2.3.7) plates, baffle edge Flow Water Chemistry Q bolts, former distribution (B.2.3.2) plates) 8a#le feFrneF StFUctuml Stainless ReacteF ceelant Cmcking ASMe Sectien XI IV.82.RP 382 3.1 1, 032 A assernely (ea#le suppert steel Neutmn flux Less ef lnsePJice lnspectien, plates, eaffie eElge F-lew rnateFial Suesectiens PJIJ8 , PlVG, eelts, ferrnm Elistrieutien anEl IVVD (B.2.3.1)

-*-*- -\

Baffle-former Structural Stainless Reactor coolant Changes in Reactor Vessel Internals IV.B2.RP-270 3.1-1 , 059a t!A assembly (baffle support steel Neutron flux dimensions (B.2.3 .7) plates, former Flow Loss of plates) distribution fracture touahness 8a#le ferrner Structuml Stainless Reacter ceelant Less ef fractuFe ReacteF Vessel Internals IV.82.RP 388 3.1 1, 059a Q assernely (ea#le suppert steel Neutmn flux teughness (B.2.3.7) plates, termer F-lew

~ Elistrieutien Baffle-former Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV. B2. RP-354 3.1-1 , 059a t!A assembly (baffle- support steel Neutron flux toughness (B.2.3.7) edge bolts) Changes in dimensions Loss of preload Loss of material Baffle-former Structural Stainless Reactor coolant Loss of Reactor Vessel Internals IV.B2.RP-296 3.1-1 , 059a Qf assembly (baffle- support steel Neutron flux material (B.2.3.7) edQe bolts)

Baffle-former Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.B2.RP-271 3.1-1 , 053a A assembly (baffle- support steel Neutron flux (B.2.3.7) former bolts) Water Chemistry B (B.2.3.2)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 25 of 41 SLRA Table 3.1 .2-2 , pages 3.1-75 through 3.1-86, is revised as follows:

Table 3.1 .2-2: Reactor Vessel Internals - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Mana!:tement l:fa#le feFFfleF StFuctural Stainless ReacteF Geelant CraGking ASMe SeGtien XI IV.B2.HP 382 3.1 1, 032 A asseFflbly (baffle suppert ~ Neutrnn flux Less ef lnservice lnspectien ,

feFFfleF belts) FflateFial SubseGtiens IWB , PAIC ,

--...J l\Mf""\ I D "> ".> * \

Baffle-former Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV.B2 .RP- 3.1-1 , 059a EiA assembly (baffle- support steel Neutron flux toughness (B.2.3.7) JM.272 former bolts) Changes in dimensions Loss of preload Loss of material Baffle feFFfleF StFuctural Stainless ReaGteF ceelant Less ef ReaGteF Vessel Internals IV.B2.RP 296 3.1 1, 059a Q asseFflbly (baffle suppert ~ Neutrnn flu x Fflaterial (B.2.3.7)

~---- - h-1 + ~ \

Bottom mounted Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.B2.RP-293 3.1-1 , 053b EiA instrumentation support steel Neutron flux (B.2.3 .7)

(column bodies) Water Chemistry .§ (B.2.3.2)

Bottom mounted Structural Stainless Reactor coolant Loss of Reactor Vessel Internals IV. B2.RP-29~GS 3.1-1 , 059b GA instrumentation support steel Neutron flux fracture (B.2.3.7)

(column bod ies) toughness Loss of material BetteFfl FfleunteEI StFUGtural Stainless ReaGteF Geelant CraGking ReaGteF Vessel Internals IV.B2.RP 355 3.1 1, 053c Q instFUFflentatien suppert ~ Neutrnn flux (B.2.3.7)

(flux th iFflble __

....Prnssure

_ , V'kiter CheFfl istry tHBest --- -- I D">".>"> \

BetteFfl FfleunteEI StruGtural Stainless ReaGter Geelant Less ef fraGture ReaGteF Vessel Internals IV.B2.R 424 3.11 , 119 ,-2 instruFflentatien suppert ~ Neutrnn flux teughness (B.2.3.7)

(flux thiFflble Prnssure Changes in tHBest -. . _ -...!--

--* El iFflensiens Bottom mounted Structural Stainless Reactor coolant Loss of Flux Thimble Tube IV.B2.RP-284 3.1-1 , 054 A instrumentation support steel Neutron flux material Inspection (B.2.3.24)

(flux th imble Pressure tubes) boundary

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page26of41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86, is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management Control rod guide Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-296ag 3.1-1 , 053a GA tube assembly support steel Neutron flux (8 .2.3.7)

(guide cards) Cast Water Chemistry § austenitic (8 .2.3 .2) stainless steel GeAtFel FeEl §bliEle StFblCtblFal Gast ReacteF ceelaAt bess ef fFactblFe ReacteF Vessel IAtemals IV.82.RP 297 3.1 1, 059a Q tblee asseFflely Sblppert ablsteAitic NebltF9A flblx tebl§hAess (8 .2.3.7)

(§bliEle cmEls) stainless steei Control rod guide Structural Stainless Reactor coolant Loss of Reactor Vessel Internals IV.82.RP-296 3.1-1 , 059a gA tube assembly support steel Neutron flux material (8 .2.3.7)

(guide cards) Cast Loss of austenitic fracture stainless toughness steel Gentml mEl §bliEle Strnctb!Fal Stainless ReacteF ceelant Gmckin§ ReacteF Vessel IAtemals IV.82.RP 298 3.1 1, 053a Q tblee asseFflely Sblppert steei Nebltmn flb!X (8.2.3.7)

(§bliEle caFEls) V\latm GheFflistry 10 .., '2 ') \

Gentml mEl §bliEle Strnctb!Fal Stainless ReacteF ceelaAt bess ef ReacteF Vessel IAtemals IV.82.RP 296 3.1 1, 059a g tblee asseFflely Sblppert steei NebltmA flblx FflateFial (8 .2.3.7) r~ ,;...J~ - , * ...J~ \

Control rod guide Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-298 3.1-1 , 053a gA tube assembly support steel Neutron flux (8 .2.3.7)

(lower flange Cast Water Chemistry § welds in austenitic (8 .2.3.2)

~eri~heral stainless assemblies) steel Control rod guide Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV.82.RP-297 3.1-1 , 059a gA tube assembly support steel Neutron flux toughness (8.2.3.7)

(lower flange Cast welds in austen itic QeriQheral stainless assemblies) steel

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-1 .4 7 Attachment 1 Page 27 of 41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86, is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aain :'.1 Manaaement Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management GeAtFOI F9EI §l::liEle StF1::1otmal Stai Riess ReaoteF oeelaAt GmokiA§ Reaotm Vessel IAtemals IV.B2.RP 298 3.1 1, 053a B tl::ll:;ie assem91y Sl::lppert steei Ne1::1tFOA f11::1x tB.2.3.7j Oov,ceF flaA§e VVateF Ghemistry ID '"> ">'"> \

welG} .-*- *- *-

GoAtml FOEI §l::liEle StF1::1ot1::1ml StaiAless ReaotoF ooolaAt boss of fFaot1::1Fe ReaotoF Vessel IAtemals IV.B2.RP 297 3.11 , 059a B t1::18e assem91y Sl::lpport steei Ne1::1tmA f11::1x te1::1§hAess tB.2.3.7j tleweF flaA§e welG}

Control rod Structural Stainless Reactor Loss of Reactor Vessel IV.82.RP-297a 3.1-11 059b A guide tube suggort steel coolant fracture Internals {B.2.3. 7}

assembl~ {lower Cast Neutron flux toughness flange welds in austenitic non-gerigheral stainless assemblies) steel Control rod Structural Stainless Reactor Cracking Reactor Vessel IV .82.RP-298a 3.1-1, 053b A guide tube suggort steel coolant Internals {B.2.3.7}

assembl~ {lower Cast Neutron flux Water Chemist!Y flange welds in austenitic {B.2.3.2}

non-gerigheral stainless assemblies) steel Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-273 3.1-1 , 053b BA assembly (barrel support steel Neutron flux (8.2.3.7) former bolts) Water Chemistry ~

(8 .2.3.2)

Core barrel Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV.82.RP-274 3.1 -1, 059b BA assembly (barrel support steel Neutron flux toughness (8.2.3.7) former bolts) Changes in dimensions Loss of preload Loss of material GoFe 9aFFel StF1::1ot1::1Fal Stai Riess ReaoteF ooolaAt boss ef ReaotoF Vessel IAtemals IV.B2.RP 299 3.1 1, 059a Q, assem91y t9aFFel Sl::lpport steei ~Je1::1tmA fl1::1x mateFial tB.2.3.7j

- -- .\

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page28of41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86, is revised as follows :

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP- 3.1-1 , 053c GA assembly (core support steel Neutron flux (B.2.3.7) ~45a barrel flange) Flow Water Chemistry § distribution (B.2.3.2)

GeFe 9aFFel StFLict1:JFal Stainless ReacteF ceelant Gmcking ASMe Sectien XI IV.B2.RP 382 3.1 1, 032 A assern91y (ceFe sup pert ~ Neutrnn flux Less ef lnser:vice lnspectien ,

9mml flange) F-lew mate Fial Su9sectiens IVVB , IVVG ,

,.J II A II""'\ I D "> ') ~ \

9istFi9utien -* - - - *-* -

  • Core barrel Structural Stainless Reactor coolant Loss of Reactor Vessel Internals IV.82.RP-345 3.1-1 , 059c gA assembly (co re support steel Neutron flux material (B.2.3.7) barrel flange) Flow distribution GeFe 9aFFel StFUctuml Stainless ReacteF ceelant Gmcking ReacteF Vessel Internals IV.B2.RP 278 3.11 , 053b g assern91y (cme suppert ~ Neutrnn flux (B.2.3.7) 9aFFel ,_ eutlet
  • -*ri \

VVateF GAern isti:y r e "> -:i "> \

GeFe 9aFFel StFUcturnl Stainless ReacteF ceelant Grncking ASMe Sectien XI IV.B2RP 382 3.1 1, 032 A assern91y (ceFe sup pert ~ Neutrnn flu x Less ef lnser:vice lnspectien ,

9aFFel eutlet mate Fial Su9sectiens IVVB , l'NG ,

~ __ , .. -Ir! \ -~,.J llAll""'\ I D">')~ \

~*-

~

GeFe 9aFml StFucturnl Stainless ReacteF ceelant Less ef Reactm Vessel Internals IV.B2RP 2909 3.1 1, 0599 G assern91y (ceFe suppert ~ Neutrnn flux rnateFial (B.2.3.7) 9aFFel eutlet Core barrel Structural Stainless Reactor coolant GAanges in Reactor Vessel Internals IV.82.RP-274 3.1 -1, 059b gg_

assembly (lower support steel Neutron flux eirnensiens (B.2.3.7) axial welds) Loss of material Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.B2.RP-387a 3.1-1, 053b gA assembly (lower support steel Neutron flux (B.2.3 ..7) axial welds) Water Chem istry § (B.2.3.2)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page29of41 SLRA Table 3.1 .2-2 , pages 3.1-75 through 3.1-86, is revised as follows :

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management GeFe eaFFel StFUcturnl Stainless ReacteF ceelant Grncking ASMe Sectien X:I IV.82.RP 382 3.1 1, 032 A assemely (lm¥eF suppert steel Neutmn flux Less ef lnseFVice lnspectien ,

axial wel9s) mate Fial Suesectiens IW8, IWG ,

~-..I 11/\H""I I C">'> 1 \

Core barrel Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV.82.RP-388a 3.1-1 , 059b SA assembly (lower support steel Neutron flux toughness (8 .2.3.7) axial welds) Changes in dimensions Core barrel Structural Stainless Reacto r coolant Gl=!angesin Reactor Vessel Internals IV.82.RP-274 3.1-1 , 059b Qf_

assembly (lower support steel Neutron flux Ei imensiens (8 .2.3.7) flange weld) Loss of material Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-280 3.1-1 , g.,:t.A assembly (lower support steel Neutron flux (8 .2.3.7) 053ah flange weld) Water Chemistry § (8 .2.3.2)

Core barrel Structural Stainless Reactor Loss of Reactor Vessel IV.82.RP-280a 3.1-11 059b A assembl~ {lower SUE!E!Ort steel coolant fracture Internals {B.2.3.7}

flange weld} Neutron flux toughness Changes in dimensions GeFe eaFFel StFUcturnl Stainless ReacteF ceelant Grncking ASMe Sectien X:I IV.82.RP 382 3.1 1, 032 A assemely (leweF sup pert steel Neutmn flux Less ef lnseFVice lnspectien ,

flange wel9) mate Fial Suesectiens 1\1\18, IVVG ,

__ _, 11~/r"I I C">'> 1 \

GeFe eaFFel StFucturnl Stainless ReacteF ceelant Less ef fFactuFe Reactm Vessel lntemals IV.82.RP 388a 3.1 1, 0599 Q assemely (lev,reF suppert steel Neutmn flux teugl=!ness (B.2.3.7) n * -1..J \

GeFe eaFFel StFUctuFal Stainless ReacteF ceelant Gl=!anges in ReacteF Vessel lntemals IV.82.RP 270 3.1 1, 059a Q assemely (leweF suppert steel Neutmn flux Eiimensiens (B.2.3.7)

~:...LL. 1..J \

Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-387 3.1-1 , 053a SA assembly (lower support steel Neutron flux (8 .2. 3.7) girth weld) Water Chemistry § (B.2.3.2)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 30 of 41 SLRA Table 3.1.2-2, pages 3.1-75 through 3.1-86, is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of AC1inC1 Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management GeFe saFFel StFuctural Stainless geacteFceelant Gracking ASMe Sectien XI IVJ32 _gp 382 3.1 1, 032 A assemsly (le'NeF suppert steel Neutrnn flux Less ef lnsewice lnspectien, girth weld) mateFial Sussectiens l'NB , IWG ,

~-...1 11f\1r. r e...,')~ \

Core barrel Structural Stainless Reactor coolant Loss of fracture Reacto r Vessel Internals IV.82.RP-388 3.1-1 , 059a SA assembly (lower support steel Neutron flux toughness (8 .2.3.7) girth weld) Changes in dimensions GeFe saFFel StFUctural Stainless ReacteF ceelant Ghanges in geacteF Vessel lntemals IV.B2RP 274 3.1 1, 05913 Q assemsly (middle sup pert steel Neutrnn flux dimensiens (B.2.3 .7)

Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-387a 3.1-1 , 053b SA assembly (middle support steel Neutron flux (8 .2.3 .7) axial welds) Water Chemistry § (8 .2.3.2)

GeFe saFFel StFUctural Stainless geacteF ceelant Gracking ASMe Sectien XI IV.B2RP 382 3.1 1, 032 A assemsly (midele suppert steel Neutrnn flux Less ef lnsewice lnspectien ,

axial welds) material Sussectiens IVVB, IVVC ,

--* - l\A/rl- ( [)...,..,

-1 \

Core barrel Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV. 82. RP-388a 3.1-1 , 059b SA assembly (middle support steel Neutron flux toughness (8 .2.3.7) axial welds) Changes in dimensions Core barrel Structural Stainless Reactor coolant Cracking Reacto r Vessel Internals IV.82.RP- 3.1-1 , 053b SA assembly (upper support steel Neutron flux (8 .2.3.7) 280~

axial weld) Water Chemistry § (8 .2.3.2)

Gere sarrel Structural Stainless geacter ceelant Gracking ASMe Sectien XI IV.B2.gp 382 3.1 1, 032 A assemsly (upper suppert steel Neutrnn flux Less ef lnsewice lnspectien ,

axial welEl) material Sussectiens IWB , IWG,

~~,.., l\A/rl ( [) ..., .., .. \

Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-276 3.1-1 , 053a SA assembly (upper support steel Neutron flux (8 .2.3.7) flange weld) Water Chemistry § (8 .2.3.2)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021- 147 Attachment1Page31of41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86, is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aginq Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management GeFe saFFel StrnctuFal Stainless ReacteF ceelant Grncking ASMe Sectien XI IV.132.RP 382 3.1 1, 032 A asseA181y (uppeF sup pert steel Neutrnn flux Less ef lnseFVice lnspectien ,

flange weld) Ai ate Fial Sussectiens IV\113 , IWG ,

l\A/r'\ I D "l '> . ,

Core barrel Struct ural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP- 3.1-1 , AB; assembly (upper support steel Neutron flux (8 .2.3.7) ~280 053.Qa girth weld) Water Chemistry § (8.2.3.2)

GeFe saFFel Strnctural Stainless ReacteF ceelant Grncking ASMe Sectien XI IV.132.RP 382 3.1 1, 032 A asseA'lsly (uppm sup pert steel Neutrnn flux Less ef lnseFVice lnspectien ,

girth weld) Ai ate Fial Sussectiens IW13 , IVVG ,

__ _, 11/\/r"\ 1 n ') -:i 1 \

LeweF ceFe plate StFucturnl Stainless ReacteF ceelant Ghangesin ReacteF Vessel Internals IV.132.RP 270 3.1 1, 059a G; (fuel alignffient suppert steel Neutrnn flux diA'lens iens (13.2.3.7)

~

LeweF ceFe plate StFucturnl Stainless ReacteF ceelant Grncking Reactm Vessel Internals IV.132.RP 289 3.1 1, 053c Q (fuel alignA'lent suppert steel Neutrnn flux (13 .2.3.7)

~ VVatm GheA'l istry ID '> '> '> \

Lower core plate Structural Stainless Reactor coolant Less ef fFactuFe Reactor Vessel Internals IV.82.RJ::L 3.1-1 , GA (fuel al ignment support steel Neutron flux teughness (8 .2.3.7) ~ 24 119GWG pins) Loss of material LeweF internals Strncturnl Stainless ReacteF ceelant Ghanges in ReacteF Vessel Internals IV.132.RP 270 3.1 1, 059a G; asseA'lsly (leweF sup pert steel Neutrnn flux diA'lensiens (13 .2.3.7) cern plate) .i;.Jew distFisutien Lower internals Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-289 3.1-1 , 053c BA assembly (lower support steel Neutron flux (8.2.3.7) core plate) Flow Water Chemistry § distribution (8.2.3.2)

LeweF internals Strncturnl Stainless ReacteF ceelant Gracking ASMe Sectien XI IV.132.RP 382 3.1 1, 032 A asseA'lsly (leweF sup pert steel Neutrnn flu x Less ef lnseFVice lnspectien ,

cere plate) .i;.Jew A'laterial Sussectiens l'.'\113 , l'NG ,

- ' l\Mr'\ I D "l '> * \

distrisutien - * - - - *- *-

  • Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 32 of 41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86 , is revised as follows :

Table 3.1.2-2: Reactor Vessel Internals - Summary of A~in :i Mana~ement Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management Lower internals Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV.82.RP-288 3.1-1 , 059c GA assembly (lower support steel Neutron flux toughness (8 .2.3.7) core plate) Flow Loss of distribution material Changes in dimensions Lower internals Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-291 a 3.1-1 , 053b SA assembly (lower support steel Neutron flux (8 .2.3.7) support forging) Water Chemistry § (8 .2.3.2) beweF internals StFuctuml Stainless ReaGteF Geelant CmGking ASMe SeGtien XI IV.B2.RP 382 3.1 1, 032 A assemely (leweF support steei Neutrnn flux bess ef lnsePv'iGe lnspeGtien ,

support feFging) mateFial Suesectiens IVVB , PNC ,

~-.J llA/I"\ / C 'J 'l

~ \

Lower support Structural Stainless Reactor coolant Loss of Reactor Vessel Internals IV.82.RP- 3.1-1, 059b GA assembly (lower support steel Neutron flux fracture (8 .2.3.7) :2+4295 support column toughness bodies) Changes in dimensions Lower support Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV. 82. RP-29~ 3.1-1, 053b GA assembly (lower support steel Neutron flux (8 .2.3.7) support column Water Chemistry bodies) (8.2.3.2) be>A<eF suppert StFuctuml Stainless ReacteF ceelant Cmcking ASMe Section XI IV.B2.RP 382 3.1 1, 032 A assemely (le>NeF suppert steei Neutrnn flux bess ef lnsePo<ice lnspeGtien, suppert celumn mateFial Suesectiens IWB , PNC ,

..J llMI"\ I D 'I'>~ \

beweF suppert StFUcturnl Stainless ReacteF ceelant bess ef fFactuFe ReactoF Vessel Internals IV.B2.RP 290a 3.1 1, 0599 s assemely (leweF sup pert steei Neutron flux toughness (B.2.3.7) support celumn

..J:

Lower support Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-286 3.1-1, 053b SA assembly (lower support steel Neutron flux (8 .2.3 .7) support column Water Chemistry § bolts) (8.2.3.2)

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1Page33of41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86 , is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of Aain :J Manaaement Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management be>NeF su1313eFI: StFucturnl Stainless ReacteF ceelant Crncking ASMe Sectien XI IV.82.RP 382 3.1 1, 032 A assembly tleweF su1313eFI: steel Neutrnn flux bess ef lnseF¥ice lns13ectien, su1313eFI: celumn mateFial Subsectiens IW8 , l'JVC ,

OOlt&t -

l\Mr"\/ D'> '>~ \

Lower support Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals IV.82.RP-287 3.1-1 , 059b EA assembly (lower support steel Neutron flux toughness (8 .2 .3.7) support col umn Loss of preload bolts) Changes in dimensions Loss of material beweF su1313eFI: StFUcturnl Stainless ReacteF ceelant bess ef ReacteF Vessel lntemals IV.82.RP 290b 3.1 1, 059b G assembly Oew eF su1313eFI: steel Neutrnn flux mate Fial Oil .2.3.7) su1313eFI: celumn OOlts}

No additional Structural Nickel alloy Reactor coolant None Reactor Vessel Internals IV.82 .RP-265 3.1- 1, 055c EA measures support Stainless Neutron flux (8 .2.3.7) componen ts Flow steel distribution Radial su1313eFI: StFUcturnl .st:eIDte ReacteF ceelant Crncking ASMe Sectien XI IV.82.RP 382 3.1 1, 032 ~

~ su1313eFI: Neutrnn flux bess ef lnseFVice lns13ectien ,

matmial Subsectiens l'/lJ8 , l'AIG ,

-* ,_ 11*/n

--...1 / 0 '"l ':l 1 \

Radial support Structural Stellite Reactor coolant Wea!:Loss of Reactor Vessel Internals NGAelV.82.R- NGA83.1-1, ~A keys support Neutron flu x material (8 .2.3.7) 424 119 Reactor vessel Structural Stainless Reactor Loss of Water Chemist!)£ IV.82.RP-24 3.1-1, 087 A internal su1212ort steel coolant material (8.2.3.2}

com12onents Flow Nickel Neutron flux distribution allov

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147Attachment1Page34of41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86, is revised as follows:

Table 3.1.2-2: Reactor Vessel Internals - Summary of Agin~ Management Evaluation Component Intended Material Environment Aging Effect Aging Management NUREG-219 Table 1 Notes Type Function Requiring Program 1 Item Item Management Reactor vessel Structural Cast Reactor coolant Cumulative TLAA - Section 4.3.1, IV.8 2. RP-303 3.1-1 , 003 SA internal support austenitic Neutron flux fatigue damage Metal Fatigue of Class 1 components with a stainless Components fatigue analysis steel Nickel alloy Stainless steel Thermal sh ield Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-302 3.1-1, 053a SA assembly (thermal support steel Neutron flux (8 .2.3.7) shield flexures) Water Chemistry § (8 .2.3.2) bl1313eF eeFe 13late StFUetuml Stainless ReaeteF eeelant Cmeking ReaeteF Vessel Internals IV.B2.RP 289 3.1 1, 053e G (fu el alignment SUJ3f39rt steel Neutmn flux (B.2.3.7}

~ 1/Vater Chemistry f D ') ':> ') \

Upper core plate Structural Stainless Reactor coolant bess ef fraeture Reacto r Vessel Internals IV.82.RP- 3.1-1, GA (fuel alignment support steel Neutron flux teughness (8.2.3.7) ~ 24 0Ws119 pins) Loss of material Upper internals Structural Stainless Reactor coolant Cracking Reactor Vessel Internals IV.82.RP-291 b 3.1-1, 053b SA assembly (upper support steel Neutron flux (8 .2.3.7) core plate) Water Chemistry § (8 .2.3.2) bl1313er internals Struetural Stainless Reaeter eeelant Craeking ASM5 Seetien XI IV.B2.RP 382 3.1 1, 032 A assembly (u1313er SUJ3f39rt steel Neutmn flux bess ef lnseFViee lns13eetien, eere 13latej material Subseetiens PJIJB , IWC ,

-Reactor

~--'

l\Mr\ f D

') ':> 1 \

  • - * -
  • I Upper internals Structural Stainless Reactor coolant Loss of Vessel Intern als IV.82.RP-290b5 3.1-1 , 059b GA assembly (upper support steel Neutron flux material (8 .2.3.7) core plate) bess ef fraeture

.~ ....

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1Page35of41 SLRA Table 3.1.2-2 , pages 3.1-75 through 3.1-86 , is revised as follows:

Generic Notes A. Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description.

B. Consistent with component, material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP has exceptions to NUREG-2191 AMP description.

C. Component is different, but consistent with material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description.

D. Component is different, but consistent with material, environment, aging effect and aging management program listed for NUREG-2191 line item. AMP has exceptions to NUREG-2191 AMP description.

E. Consistent with NU REG 2191 material, environment, and aging effect but a different aging management program is credited or NU REG 2191 identifies a plant specific aging management program.

F. Material not in NU REG 2191 for this component.

J. Neither the component nor the material and environment combination is evaluated in NU REG 2191.

Plant Specific Notes

1. Component inspection category is not consistent with the inspection category cited in Table 3.1-1 .
2. The PWR Vessel Internals program manages loss of fracture toughness and changes in dimension for stainless steel flux thimble tubes through FM EGA analysis described further in Appendix C. Loss of preload is not applicable to flux thimble tubes, and loss of material is addressed by NUREG 2191 item IV.B2.RP 284. Flux thimble tubes are existing program components.

~ V'iear surfaces for the upper core plate alignment pins, clevis inserts, and radial support keys are Stellite. Aging effects identified in the Appendix C RVI gap analysis for these components are managed by the Reactor Vessel Internals program .

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 36 of 41 SLRA Table 4.1.5-2, page 4.1-6, is revised as follows :

Table 4.1.5-2 Review of Plant-Specific TLAAs Listed in NUREG-2192, Table 4.7-1 Applies SLRA Table 4.7-1 Examples of Potential Plant-Specific TLAA Topics to PBN Section PW Rs No Reactor pressure vessel underclad cracking N/A (Note 1) 4.7.1 Leak-before-break Yes 4 .7.2 Reactor coolant pump flywheel fatigue crack growth Yes 4.7.4 Response to NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Yes 4.3.1 Stratification" Response to NRC Bulletin 88-08, "Thermal Stresses in Piping Yes 4.3.1 Connected to Reactor Cooling Systems" EPRI MRP cycle-based and fluence-based analyses in No N/A su121~ort of MRP-227 (Note 3}

BWRs and PWRs Fatigue of cranes (crane cycle limits) Yes 4.7 .6 No Fatigue of the spent fuel pool liner N/A (Note 2)

No Corrosion allowance calculations N/A (Note 2)

No Flaw growth due to stress corrosion cracking N/A (Note 2)

Predicted lower limit Yes 4.3.5 Note 1: Refer to Section 3.1.2.2.5.

Note 2: Refer to Notes 3, 4, and 5 of Table 4.1 .5-1 .

Note 3: Cycle-based fatigue for the PBN RVI is included with the generic industry TLAA "Metal Fatigue of Class 1 Components" in SLRA Table 4.1.5.3 and Section 4.3.1. A PBN plant-specific RVI fluence-based analysis is not part of the PBN CLB and therefore does not meet the TLAA definition for SLR.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1Page37of41 SLRA Appendix A Table 16-3, page A-67, is revised as follows:

Table 16-3 List of SLR Commitments and Implementation Schedule No. Aging NUREG-2191 Commitment Implementation Management Section Schedule Program or Activity (Section) 11 Reactor Vessel Xl.M16A Continue the existing PBN Reactor Vessel Internals AMP, including No later than 6 months Internals enhancement to: prior to the SPEO , i.e. :

(16.2.2.7) a) Implement the guidance in MRP 227 Rev. 1: A as supplemented by PBN1 : 04/05/2030 the gap analysis, or the latest NRC approved version of MRP 227 PBN2: 09/08/2032 which addresses 80 years of operation if one is available prior to the subsequent period of extended operation.

b) Implement the results of the gap analysis in the Reactor Vessel Internals Program unless it is superseded by the latest NRC approved version of MRP 227 which addresses 80 years of operation . If so , the AMP may be implemented directly without the use of a gap analysis .

c) Incorporate the updated examination acceptance criteria , Primary I Expansion links, expansion criteria , and expansion item examination criteria in MRP 227 Rev. 1: A as supplemented by the gap analysis.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 38 of 41 SLRA Appendix B Section B.1 .1, page B-5, is revised as follows:

These new AMPs will be consistent with the 10 elements of their respective NUREG-2191 AMPs. The following programs each have exception(s) justified by technical data:

  • the PBN Water Chemistry AMP (Section B.2.3.2),
  • the PBN Reactor Head Closure Stud Bolting AMP (Section B.2.3.3),
  • the PBN Reactor Vessel Internals AMP (Section B.2.3.7),
  • the PBN Open-Cycle Cooling Water System AMP (Section B.2.3.11 ),
  • the PBN Closed Treated Water Systems AMP (Section B.2.3.12),
  • the PBN Fuel Oil Chemistry AMP (Section B.2.3.18),
  • the PBN Reactor Vessel Material Surveillance AMP (Section B.2.3.19),
  • the PBN Buried and Underground Piping and Tanks AMP (Section B.2.3.27),
  • the PBN Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks AMP (Section B.2.3.28),

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021 -147 Attachment 1 Page 39 of 41 SLRA Appendix B Table B-4, page B-18, is revised as follows: '

Table 8-4 Point Beach Aging Management Program Consistency with NUREG-2191 PBN Aging Section PBN NUREG-2191 Comparison Management Plant-Specific?

Program NUREG-2191 Enhancements? Exceptions?

Section Cracking of B.2.3.5 No Xl.M11B Yes No Nickel-Alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components Thermal Aging B.2.3.6 No Xl.M12 New No Embrittlement of Cast Austenitic Stainless Steel Reactor Vessel B.2.3.7 No Xl.M16A Yes ¥esNo Internals Flow-Accelerate B.2.3.8 No Xl.M17 Yes No d Corrosion Bolting Integrity B.2.3.9 No Xl.M18 Yes No Steam B.2.3.10 No Xl.M19 Yes Yes Generators Open-Cycle B.2.3.11 No Xl.M20 Yes Yes Cooling Water System Closed Treated B.2.3.12 No Xl.M21A Yes Yes Water Systems Inspection of B.2.3.13 No Xl.M23 Yes No Overhead Heavy Load Handling Systems

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 1 Page 40 of 41 SLRA Appendix B Section B.2.3.7, page B-73, is revised as follows :

NUREG-2191 Consistency The PBN RVI AMP, with enhancements, will be consistent with an exception with the 10 elements program described in of NUREG-2191, Section Xl.M16A, "PWR Vessel Internals" as modified by the Interim Staff Guidance SLR-ISG-2021 PWRVI .

Exceptions to NUREG-2191 The program described in NUREG 2191, Section Xl.M16/\ provides MRP 227 A as the basis for a site specific RVI program. The scope of the PBN Reactor Vessel Internals /\MP applies the methodology and guidance in MRP 227 Revision 1 A (as supplemented by a gap analysis). MRP 227 Revision 1 /\is the most recent NRG approved guidance for managing PWR vessel internals and incorporates significant recent operating experience .None.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021 -147Attachment1Page41of41 SLRA Appendix C Section C.1.0, page C-3, is revised as follows:

In accordance with Interim Staff Guidance SLR-ISG-2021-01-PWRVI, +! he PBN subsequent license renewal (SLR) RVI gap analysis uses the most recent guidelines provided in EPRI Technical Report No. 3002017168, MRP-227 Rev. 1-A (Reference C.9.1) as the baseline to address an 80-year operating period , consistent with the NRC SE dated April 15, 2019 (Reference ML19081A001) indicating that MRP-227 Rev. 1 can be used as a starting point for performing a gap analysis in order to develop an RVI AMP for the 60-80-year subsequent period of extended operation (SPEO), and the NRC SE dated February 19, 2020 (Reference ML20006D152) indicating that MRP-227 Rev . 1-A is acceptable to the extent delineated in the April 151 2019 SE. Revision 1 of the guidelines provides updates based on Revision 1 of the NRC SE fo r MRP-227 Revision 0 (Reference ML11308A770) and includes operating experience and new knowledge gained from materials testing , modeling, and research . MRP-227 Rev. 1-A is the acceptance version incorporating changes from the NRC SE approving MRP-227 Revision 1. Note that MRP-227 Rev . 1-A still only addresses an operating period of 60 years and will be implemented at PBN for the current period of extended operation by January 1, 2022.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 2 Page 1 of 2 Pressure-Temperature Limits and Low Temperature Overpressure Protection (L TOP) Setpoints - Clarification Affected SLRA Sections: Section 4.2.5 SLRA Page Numbers: 4.2-23 Description of Change:

Revised the last paragraph on the page regarding update of P-T limit curves and LTOP PORV setpoints .

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 2 Page 2 of 2 SLRA Enclosure 3 Attachment 1 Section 4.2.5, page 4.2-23, last paragraph is revised as follows:

The P-T limit curves and LTOP PORV setpoints will be updated (if required) and a Technical Specification change request will be submitted for approval prior to exceeding the current 50 EFPY limits.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 3 Page 1 of 5 Electrical Scoping and Screening - Figures showing off-site power recovery paths following an SBO Affected SLRA Sections: Table of Contents "List of Figures" , Section 2.5.1.4 SLRA Page Numbers: xviii, 2.5-4 Description of Change:

SLRA Section 2.5.1.4 is revised to add Figures 2.5-1 and 2.5-2 indicating Station Blackout (SBO) recovery paths for PBN Units 1 and 2, respectively.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021 -147 Attachment 3 Page 2 of 5 SLRA Enclosure 3 Attachment 1 Table of Contents "List of Figures" page xviii is revised as follows :

LIST OF FIGURES Figure 2.5-1 PBN Unit 1 Restoration of Offsite Power Following a 580.............................. 2.5-4 Figure 2.5-2 PBN Unit 2 Restoration of Offsite Power Following a SBO ............................ .. 2.5-4 Figure 3.5.2.2-1 Elevation View of the PSW and BSW ................. .. .............. .......... .. .......... .... .. .... .. . 3.5-34 Figure 4.5-1 Time-Dependent Tendon Force Curves for Unit 1 Dome Tendons .... .... ........ .. .. .. .. 4.5-9 Figure 4.5-2 Time-Dependent Tendon Force Curves for Unit 1 Hoop Tendons .. ...... ..... .... .. .... .. 4.5-9 Figure 4.5-3 Time-Dependent Tendon Force Curves for Unit 1 Vertical Tendons .. .... ... ... ..... .... . 4.5-10 Figure 4.5-4 Time-Dependent Tendon Force Curves for Unit 2 Dome Tendons .... ... ..... ....... ... .. 4.5-10 Figure 4.5-5 Time-Dependent Tendon Force Curves for Unit 2 Hoop Tendons .... .... .............. . .. 4.5-11 Figure 4.5-6 Time-Dependent Tendon Force Curves for Unit 2 Vertical Tendon .. ........ .. ...... .. .. . 4.5-11

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 3 Page 3 of 5 SLRA Enclosure 3 Attachment 1 Section 2.5.1.4, page 2.5-4, last paragraph is revised as follows:

The electrical interconnection between PBN Units 1 and 2 and the offsite transmission network and the off-site power recovery paths following an SBO are highlighted on electrical boundary drawing SLR-ELECTRICAL-E1 . Figures 2.5-1 and 2.5-2 identify major components or commodities associated with restoration of off-site power following an SBO event.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 3 Page 4 of 5 SLRA Enclosure 3 Attachment 1 Section 2.5.1.4, page 2.5-4, the following two (2) figures are added after the last paragraph :

Figure 2.5 PBN Unit 1 Restoration of Offsite Power Following an SBO F"89-112 BVS lli'ANSAllSSICW CCW/}(/C!CW ..VAIPl'R CAOt.l" HIGH-VCI.. TAGE STATIDN AUXILIARY TRANSf'CRM[R IX03 AIETAL ENCZOSEO eus ----c VNO£~NO NEOIVV-W TAC£ CABl.E 13.8KV BUS 1H02

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-147 Attachment 3 Page 5 of 5 Figure 2.5 PBN Unit 2 Restoration of Offsite Power Following an SBO ra9-152 evs HIGH-VCL TAGE STATION AUXILIARY TRANSF"DRMER 2X03

>---VETAl /"NG. OSEIJ eus

>--- llNIJERGROVNIJ .VEO/ll.V- ~TAC£ CABLE 13.SKV BUS 2H03

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