L-2021-113, Subsequent License Renewal Application - Aging Management Supplement 3

Subsequent License Renewal Application - Aging Management Supplement 3

ML21147A115
Person / Time
Site:	Point Beach
Issue date:	05/27/2021
From:	Maher W Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2021-113
Download: ML21147A115 (41)

Text

May 27, 2021 U.S. Nuclear Regulatory Commission Attention: Document Control Desk 11545 Rockville Pike One White Flint North Rockville, JvID 20852-2746 Point Beach Nuclear Plant Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR 27 NEXTera ENERGY ~

POINT BEACH L-2021-113 10 CFR 54.17 SUBSEQUENT LICENSE RENEWAL APPLICATION - AGING MANAGEMENT SUPPLEMENT 3

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 January 15, 2021, Point Beach Nuclear Plant, Un.its 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. ML21006A417)

3.

NRC Letter dated January 15, 2021, Point Beach Nuclear Plant, Units 1 and 2 - Aging lvianagement Audit Plan Regarding the Subsequent License Renewal Application Review (ADAJvIS Accession No. ML21007A260)

4. NEPB Letter L-2021-081 dated April 21, 2021, Subsequent License Renewal Application - Aging Management Supplement 1 (ADAMS Accession No. ML211l1A155)

5.

NEPB Letter L-2021-102 dated May 6, 2021, Subsequent License Renewal Application -Aging Management Supplement 2 (ADAMS Accession No. 1vIL21126A239)

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 January 15, 2021, the NRC determined that NEPB's SLRA was acceptable and sufficient for docketing and issued the regulatory audit plan for the aging management portion of the SLRA review (References 2 and 3). During this audit conducted betweenJanua1y 19, 2021 to Jviarch 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 provides that information and does not incorporate or otherwise affect any new or clarifying information provided in References 4 and 5.

For ease of reference, the attachment topic index is provided on page 3 of this letter. In the attachment, changes are described along with the affected section(s) and page number(s) of the docketed SLRA NextEra Energy Point Beach, LLC 661 0 Nuclear Road, Two Rivers, WI 54241

Document Control Desk L-2021-113 Page 2 (Enclosure 3 Attachment 1) where the changes are to apply. For clarity, revisions to the SLRA are provided with deleted text by stri:kethroughs and inserted text by bold red underline.

Pursuant to 10 CFR 50.91 (b)(1 ), a copy of this letter is being forwarded to the State of \\\\lisconsin.

Should you have any questions regarding this submittal, please contact me at (561) 304-6256 or

\\Xlilliam.Maher@fpl.com.

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

Executed on the 27 1h day of May 2021.

Sincerely, DkJlt.IUyslg~byWil i..am /,Uhfr Dfl:tn:\\'lilhm M.i l'ler,o: l/1.1Clur, William Maher ::~~::~~~:9:,~~';<~:

e=US 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 \\Xlisconsin

Document Control Desk L-2021-113 Page 3 Attachment No.

Attachment Index PBN SLRA Enclosure 3 Attachment 1 Topic Incorporation of Interim Staff Guidance SLR-ISG-2021-01-P\\VRVI

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 1 of 38 Incorporation of SLR-ISG-2021-01-PWRVI 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 B.1.1, Table B-4, Section B.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, 3.1-41 through 3.1 -43, 3.1-46, 3.1 -56 through 3.1-58, 3.1-75 through 3.1 -86, 4.1 -6, A-67, B-5, B-18, B-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 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.

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-2021-113 Attachment Page 2 of 38 SLRA Enclosure 3Attachment1Section2.1.6, Pages 2.1-31and2.1-32, 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-Electrical-2020-XX (Reference ML20156A324)

SLR-ISG-Structures-2020-XX (Reference ML20156A338)

SLR-ISG-Mechanical-2020-XX (Reference ML20156A330)

SLR-ISG-2021-01-PWRVI (Reference ML20217L203)

Updated Aging Management Criteria for Electrical Portions of Subsequent License Renewal Guidance Updated Aging Management Criteria for Structures Portions of Subsequent License Renewal Guidance Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance Updated Aging Management Criteria for 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-2021-113 Attachment Page 3 of 38 New Section 2.1.6.4 is added on SLRA 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-2021-113 Attachment Page 4 of 38 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-2021-113 Attachment Page 5 of 38 SLRA Enclosure 3 Attachment 1 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 Stellite Structural support wear surfaces)

Alignment and interfacing components (clevis insert bolts)

Structural support Alignment and interfacing components (clevis insert dowels)

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

Baffle-former assembly (baffle plates, baffle edge bolts, former Structural support plates)

Flow distribution Baffle-former assembly (baffle plates, former plates)

Structural support Flow distribution Baffle-former assembly (baffle-edqe bolts)

Structural suooort Baffle-former assembly (baffle-former bolts)

Structural suooort Bottom mounted instrumentation (column bodies)

Structural suooort Bottom mounted instrumentation (flux thimble tubes)

Structural support Pressure boundary Control rod quide tube assembly (quide cards)

Structural support Control rod guide tube assembly (lower flange welds in Structural support oerioheral assemblies)

Control rod guide tube assembl~ {lower flange welds in Structural su1mort non-oerioheral assemblies\\

Core barrel assembly (barrel former bolts)

Structural support Core barrel assembly (core barrel flange)

Structural support Flow distribution GeFe saFFel assernsly (seFe saFFel e1:Jtlet R9llle--welti1

~trnGtl:JFal Sl:Jl31300 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 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

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 6 of 38 SLRA Enclosure 3 Attachment 1 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

~'Vear Aging Management Programs The following AMPs manage the aging effects for the reactor vessel internals components:

.L.-,1\\SME Section XI lnservice Inspection (B.2.3.1)

• 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-2021-113 Attachment Page 7 of 38 SLRA Enclosure 3Attachment1 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), provides provided the industry's current aging management initial set of aging management inspection and evaluation (l&E) 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 Reliability Program fMRP} 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 (SCCl, (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. Tho methodology in MRP 227 A was approved by tho NRG in a safety evaluation dated December 16, 2011 (ADAMS AccessfOR-Ale-:

ML11308A770), 11/hich includes those plant specific app!icantlticonsoo action items that a liconsoo or applicant applying tho MRP 227 A report would need to address and resoh'fJ 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 f.hese--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

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 8 of 38 SLRA Enclosure 3Attachment1Section3.1.2.2.9, pages 3.1-14 through 3.1-15, is revised as follows:

for the initial operating license period plus an additional 20 years during the initial 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 are 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.UU 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 feF-ils reaGtor uRitfs) 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 AILA/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

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 9 of 38 SLRA Enclosure 3 Attachment 1 Section 3.1.2.2.9, pages 3.1-14 through 3.1-15, is revised as follows:

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. +IIB-lf the AMP is a plant-specific program, the UUNRC 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-2021-113 Attachment Page 10 of 38 SLRA Enclosure 3 Attachment 1 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 Effect/Mechanism Aging Management Further Evaluation Number Program (AMP)/TLAA Recommended 3.1-1, 025 Steel (with nickel alloy cladding)

Cracking due to primary AMP Xl.M2, "Water Yes (SRP-SLR or nickel alloy steam generator water sec Chemistry," and AMP Sections 3.1.2.2.11.1 primary side components:

Xl.M19, and divider plate and tube-to-tube "Steam Generators." In 3.1.2.2.11.2) 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 "Existing Programs" wear; cracking due to Vessel Internals," and Section 3.1.2.2.9) components: Stainless steel, SCC, irradiatieA assisted AMP Xl.M2, "Water nickel alloy WestiAgheuse, and

-SGGIASCC, fatigue Chemistry" (for SCC X-750 control rod guide tube mechanisms only) support pins {split pins},--aM GeFA9ustieA EAgiAeeriAg therFAal shield 13esitieAiAg 13iAs;

~irealey 4 GeFA9ustieA EAgiAeeriAg iAeere iAstruFAeAtatieA thiFAble tubes exposed to reactor coolant and neutron flux 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.

Discussion Not applicable.

Further evaluation is documented in subsection 3.1.2.2.11.

GeAsisteAt wiU1 t>Jl::JREG ~~ 9~. +Re 1=1Elt>J Reaeter Vessel IAtemals

~ B. ~. 3. 7 j aAd VlJater GheFAistry ~ El. ~. d - ~ j AMF!s are used te FAaAage the reaeter vessel iAtemals u1313er sere 13late aAd aligAFAeAt 13iAs. Not a1;mlicable. The control rod guide tubes are not an "Existing Programs" component. Further evaluation is documented in subsection 3.1.2.2.11.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 11 of 38 SLRA Enclosure 3 Attachment 1 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 EffecUMechanism Aging Management Further Evaluation Number Program (AMP)/TLAA Recommended 3.1 1, 032 StaiRless steel, Rieke! alley, eF CmekiRg, less ef mateFial AMP XlM1, "ASMe Ne CASS FeaeteF vessel iRteFRals, due te weaF SeetieR XI IRseFviee eeFe su1313eFt stFuetuFe tRet IRs13eetieR, Sul3seetieRs alFeady FefeFeReed as IWEl, PNC, aRd IVVD" ASMe SeetieR XI e:x:amiRatieR CategeFy El

~J 3 eeFe su1313eFt stFuetuFe eem13eReRts iR MRP 22+ A~, e:l~13esed te maetoF eoolaRt aRd ReutFOR flu:x:

3.1-1, 033 Stainless steel, steel with Cracking due to SCC AMP Xl.M1, "ASME No stainless steel cladding Class 1 Section XI lnservice reactor coolant pressure Inspection, Subsections boundary components exposed IWB, IWC, and IWD,"

to reactor coolant and AMP Xl.M2, "Water Chemistry" 3.1 -1, 034 Stainless steel, steel with Cracking due to SCC AMP Xl.M1, "ASME No stainless steel cladding Section XI lnservice pressurizer relief tank (tank shell Inspection, Subsections and heads, flanges, nozzles)

IWB, IWC, and IWD,"

exposed to treated borated and AMP Xl.M2, "Water water >60°C (>140°F)

Chemistry" Discussion CeRsisteRt witl=i

~Jl:::IReG 2191. +Re PEl~J ASMe SeetieR XI IRSeFViee IRs13eetieR, Sul3seetieRs IVVEl, IVVC, aRd IWD tEl.2.3.1 ~ AMP is used te maRage eFaekiRg aRd less ef mateFial iR FeaeteF vessel iRternal eoFe su13130Ft stFUetums e:x:13osed to FeaetoF eoolaRt aRd ReutFoR flt!*.-

Consistent with NUREG-2191. The PBN ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD (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.

Not applicable. The PBN pressurizer relief tank is not an ASME Section XI component. Cracking due to sec in the stainless 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-2021-113 Attachment Page 12 of 38 SLRA Enclosure 3Attachment1Table3.1-1, pages 3.1-41through3.1-43, is revised as follows:

Table 3.1-1: Summary of AQinQ ManaQement Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Effect/Mechanism Aging Management Further Evaluation Number ProQram (AMP)/TLAA Recommended 3.1-1, 053a Stainless steel, nickel alloy Cracking due to SCC, AMP Xl.M16A, Yes (SRP-SLR Westinghouse reactor internal irradiation assisted "PWR Vessel Internals,"

Section 3.1.2.2.9)

"Primary" components exposed SGGIASCC, fatigue and AMP Xl.M2, to reactor coolant, neutron flux "Water Chemistry" (for sec mechanisms only)

Discussion Consistent with NUREG-2191. The Reactor Vessel Internals (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-2021-113 Attachment Page 13 of 38 SLRA Enclosure 3Attachment1Table3.1-1, pages 3.1-41through3.1-43, is revised as follows:

Table 3.1-1: Summarv of Aqinq Manaqement Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Effect/Mechanism Aging Management Further Evaluation Number Proqram (AMP)/TLAA Recommended 3.1-1, 053b Stainless steel Westinghouse Cracking due to SCC, AMP Xl.M16A, Yes (SRP-SLR reactor internal "Expansion" irradiation assisted "PWR Vessel Internals,"

Section 3.1.2.2.9) components exposed to reactor

~IASCC, fatigue and AMP Xl.M2, coolant and neutron flux "Water Chemistry" (for sec mechanisms only)

Discussion Consistent with NUREG-2191. The Reactor Vessel Internals (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-2021-113 Attachment Page 14 of 38 SLRA Enclosure 3 Attachment 1 Table 3.1-1, pages 3.1-41 through 3.1-43, is revised as follows:

Table 3.1-1: Summary of AQinQ ManaQement Evaluations for the Reactor Vessel, Internals, and Reactor Coolant System Item Component Aging Effect/Mechanism Aging Management Further Evaluation Number ProQram (AMP)/TLAA Recommended 3.1-1, 053c Stainless steel, nickel alloy...m:

Cracking due to SCC, AMP Xl.M16A, Yes (SRP-SLR stellite Westinghouse reactor irradiation assisted "PWR Vessel Internals,"

Section 3.1.2.2.9) internal "Existing Programs"

.sGGIASCC, fatigue and AMP Xl.M2, components exposed to reactor "Water Chemistry" (for coolant, neutron flux sec mechanisms only) 3.1-1, 054 Stainless steel Westinghouse-Loss of material due to AMP Xl.M37, No design bottom mounted wear "Flux Thimble Tube instrument system flux thimble Inspection" tubes (with or without chrome plating) 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.

Discussion Consistent with NUREG-2191. The Reactor Vessel Internals (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.

Consistent with NUREG-2191. The Flux Thimble Tube Inspection (B.2.3.24) AMP is 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.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 15 of 38 SLRA Enclosure 3 Attachment 1 Table 3.1-1, page 3.1-46, is revised as follows:

Table 3.1-1: Summary of Aging Management Evaluations for the Reactor Vessel, Internals, and Reactor Coolant Svstem Item Component Aging Effect/Mechanism Aging Management Further Evaluation Number ProQram (AMP)/TLAA Recommended 3.1-1, 059c Stainless steel (SS, including Loss of fracture toughness AMP Xl.M16A, Yes (SRP-SLR CASS, PH SS or martensitic due to neutron irradiation "PWR Vessel Internals" Section 3.1.2.2.9)

SS)~ m nickel alloy, or stellite embrittlement and for Westinghouse reactor internal CASS, martensitic SS, and "Existing Programs" PH SS due to thermal components exposed to reactor aging embrittlement; coolant and neutron flux changes in dimensions due to void swelling, distortion; loss of preload due to thermal and irradiation-enhanced stress relaxation, creep; loss of material due to wear 3.1-1, 060 Not applicable. This line item only applies to BWRs.

3.1-1, 061 Steel steam generator steam Wall thinning due to AMP Xl.M17, No nozzle and safe end, feedwater flow-accelerated corrosion "Flow-Accelerated nozzle and safe end, AFW Corrosion" nozzles and safe ends exposed to secondary feedwater/steam Discussion The Reactor Vessel Internals (B.2.3.7) AMP is used to manage 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.

Consistent with NUREG-2191. The Flow-Accelerated Corrosion (B.2.3.8) AMP is used to manage wall thinning due to flow accelerated corrosion in the steam generator feedwater nozzle and steam outlet nozzle exposed to secondary feedwate r/ steam.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 16 of 38 SLRA Enclosure 3 Attachment 1 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 Effect/Mechanism Aging Management Further Evaluation Number Program (AMP)/TLAA Recommended 3.1-1, 113 Not applicable. This line item only applies to BWRs.

3.1-1, 114 Reactor coolant system Cracking due to SCC, AMP Xl.M1, "ASME No components defined as IGSCC, PWSCC, IASCC Section XI lnservice ASME Section XI Code Class (SCC mechanisms for Inspection, Subsections components (ASME Code Class stainless steel, nickel alloy IWB, IWC, and IWD,"

1 reactor coolant pressure components only)~

and AMP Xl.M2, boundary components, reactor fatigue, or cyclic loading; "Water Chemistry" vessel interior loss of material due to (water chemistry-related attachments, or core support general corrosion (steel or corrosion-related structure components,-,;_or only), pitting corrosion, aging effect mechanisms ASME Class 2 or 3 crevice corrosion, or wear only) 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) 3.1-1, 115 Stainless steel piping, piping None None Yes (SRP-SLR components exposed to Section 3.1.2.2.1 5) concrete Discussion Not used.

All relevant aging mechanisms requiring management by ASME Section XI lnservice Inspection, Subsections IWB, IWC, and IWD (B.2.3.1) or Water Chemistry (B.2.3.2) are recognized using line items more specific to the individual component type.

Not applicable.

There are no PBN stainless steel reactor coolant system piping or piping components exposed to concrete. Further evaluation is documented in subsection 3.1.2.2.15.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021 -113 Attachment Page 17 of 38 SLRA Enclosure 3Attachment1Table3.1-1, 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 Effect/Mechanism Aging Management Further Evaluation Number Program (AMP)/TLAA Recommended 3.1-1, 116 Nickel alloy control rod drive Loss of material due to Plant-specific aging Yes (SRP-SLR penetration nozzles exposed to wear management program Section 3.1.2.2.10.1) reactor coolant 3.1 -1, 11 7 Stainless steel, nickel alloy Loss of material due to Plant-specific aging Yes (SRP-SLR control rod drive penetration wear management program Section 3.1.2.2.10.2) nozzle thermal sleeves exposed to reactor coolant 3.1-1, 118 Stainless steel, nickel alloy Cracking due to SCC, Plant-specific aging Yes (SRP-SLR PWR reactor vessel internal irradiation assisted management program Section 3.1.2.2.9) components or

&GGIASCC, cyclic or AMP Xl.M1 GA, "PWR LRA/SLRA-specified loading, fatigue Vessel Internals," and reactor vessel internal AMP Xl.M2, "Water component exposed to reactor Chemist!Y" (SCC and coolant, neutron flux IASCC onl~}, with an adjusted site-specific or component-specific aging management basis for a specified reactor vessel internal comoonent Discussion Consistent with NUREG-2191. The ASME Section XI lnservice Inspection, Subsections IWB, !WC, and IWD (B.2.3.2) AMP is used to manage loss of material due to wear in the control rod drive mechanism head penetration housings exposed to reactor coolant.

Further evaluation is documented in subsection 3.1.2.2.10.

Not applicable. Further evaluation is documented in subsection 3.1.2.2.10.2.

Not applicable. Cracking due to sec, irradiation-assisted sec, cyclic loading, and fatigue of stainless steel, nickel alloy PWR reactor vessel internal components exposed to reactor coolant, neutron flux is addressed in rows 3.1-1, 053a, 3.1-1,

053b, and 3.1-1, 053c. The associated NUREG-2191 aoinq items are not used.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 18 of 38 SLRA Enclosure 3 Attachment 1 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 Effect/Mechanism Aging Management Further Evaluation Number Program (AMP)/TLAA Recommended 3.1-1, 119 Stainless steel, nickel alloy..

Loss of fracture toughness Plant-specific aging Yes (SRP-SLR stellite PWR reactor vessel due to neutron irradiation management program Section 3.1.2.2.9) internal components or embrittlement or thermal or AMP Xl.M1 SA. "PWR LRA/SLRA-specified reactor aging embrittlement; Vessel Internals," with vessel internal component changes in dimensions an adjusted site-exposed to reactor coolant, due to void swelling or specific or component-neutron flux distortion; loss of preload specific aging due to thermal and management basis for irradiation-enhanced a specified reactor stress relaxation or creep; vessel internal loss of material due to component wear 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.

Discussion Consistent with NUREG-2191 for~

fFastl:lFe tel:l§lRAess aA8 6RaA§leS iA simeAsieAS loss of material in the stainless steel upper and lower core plate fuel alignment pins as well as the stellite radial support ke)ls and upper core plate alignment pins. Less ef fFactl:lFe tel:l§lAAess aA8 GAaA§leS iA simeAsieA feF staiAless steel FeacteF vessel iAtemals cem13eAeAtS is maAa§leS ey tt:ie ReasteF Vessel IAtemals ~ B. ~. d.7 ) AMP.

Less ef f3Felea8 is Aet a1313lisaele. Less ef mateFial is a88rnsse8 witt:i item Al:lmeeF d. ~ ~

, Ge4. 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-2021-113 Attachment Page 19 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management Ali§JAmeAt aAe Structural

~ Reacter ceelaAt CrackiA§l Reacter Vessel IAternals iAterfaciA§l suppert NeutreA flux (B.2.3.7) cempeAeAts (clevis seariA§l Stellite 1.vear

.r Alignment and Structural Ste I lite Reactor coolant Loss of Reactor Vessel Internals interfacing support Neutron flux material (8.2.3.7) components (clevis bearing Stellite wear surfaces)

Alignment and Structural Nickel alloy Reactor coolant Cracking Reactor Vessel Internals interfacing support Neutron flux (8.2.3.7) components Water Chemistry (clevis insert bolts)

(8.2.3.2)

Ali§lAmeAt aAe Structural Nickel alley Reacter ceelaAt CrackiA§l ASME: SectieA XI iAterfaciA§l sup pert NeutreA flux Less ef IAsePJice IAspectieA, cempeAeAts material SussectieAS IWB, IWC, l\\"11"\\ / 0

~ \\

Alignment and Structural Nickel alloy Reactor coolant Loss of Reactor Vessel Internals interfacing support Neutron flux material (8.2.3.7) components Loss of preload (clevis insert bolts)

,1\\li§lAmeAt aAe Structural

~Jickel alley Reacter ceelaAt CrackiA§l ASME: SectieA XI iAterfaciA§l suppert NeutroA flux Less ef IAsePJice IAspectieA, cempeAeAts material SussectieAS IWB, IVVC, (clevis iAsert aAd IWD (B.2.3.1)

Alignment and Structural Nickel alloy Reactor coolant Loss of Reactor Vessel Internals interfacing support Neutron flux material (8.2.3.7) components (clevis insert dowels)

NUREG-2191 Table 1 Notes Item Item Nooe Nooe

~

NooelV.82.RP-Nooe3.1-1,

~A 285 059c IV.82.RP-399 3.1-1, 053c BA. 1

§ IV.B2.RP 382 3.1 1, 032 A

IV. 82. RP-285 3.1-1, 059c BA, 1 IV.B2.RP 382 3.1 1, 032 A

IV. 82. RP-399 3.1-1, 053c BA. 1

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 20 of 38 SLRA Enclosure 3 Attachment 1 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 Ac::tinc::t Manac::iement Evaluation Component Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management Alignment and Structural Stainless Reactor coolant Cracking Reactor Vessel Internals interfacing support steel Neutron flux (B.2.3.7) components Water Chemistry (upper core plate (B.2.3.2) aliQnment pins)

Alignment and Structural Stainless Reactor coolant Loss of Reactor Vessel Internals interfacing support steel Neutron flux material (B.2.3.7) components Water Chemistry (upper core plate (B.2.3.2) alignment pins)

Alignment and Structural Ste I lite Reactor coolant Loss of Reactor Vessel Internals interfacing support Neutron flux material (B.2.3.7) components (upper core plate alionment pins)

Baffle-former Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (baffle support steel Neutron flux (B.2.3.7) plates, baffle edge Flow Water Chemistry bolts, former distribution (B.2.3.2) plates)

Baffle-former Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (baffle support steel Neutron flux (B.2.3.7) plates, baffle edge Flow Water Chemistry bolts, former distribution (B.2.3.2) plates)

Baffle feFFfleF Structuml Stai A less geacter ceelaAt CrackiAg ASMe SectieA XI asseFflely (eaffle sup pert steet NeutreA flux Less ef IAservice IAspectieA, plates, eaffle eElge

~

Ff!ateFial SuesectieAS IWB, IWC, eelts, ferFfler ElistrieutieA aAEI IWD (B.2.3.1)

~

Baffle-former Structural Stainless Reactor coolant Changes in Reactor Vessel Internals assembly (baffle support steel Neutron flux dimensions (B.2.3.7) plates, former Flow Loss of plates) distribution fracture touahness NUREG-2191 Table 1 Notes Item Item IV.B2.RP-3.1-1, 053c GA

~01

§ IV. B2. RP-299 3.1-1, 059c BA

§ NeAel¥.B2.R Noo-e3.1-1,

~A 424 119 IV. B2. RP-270a 3.1-1, 053a BA

§ IV. B2. RP-387 3.1-1, 053a QG g

IV.82.gp 382 3.11, 032 A

IV. B2. RP-270 3.1-1, 059a BA

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 21 of 38 SLRA Enclosure 3 Attachment 1 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 Aqinq Manaqement Evaluation Component Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Manaqement Baffle feFFfleF StFuctuml Stainless ReacteF ceelant bess ef fFactuFe ReacteF Vessel Internals asseffiely (eaffle suppert steel Neutrnn flux teughness (B.2.3.7) plates, foFFfleF P-lew

~

EiistFieutien Baffle-former Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals 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 assembly (baffle-support steel Neutron flux material (B.2.3.7) edQe bolts)

Baffle-former Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (baffle-support steel Neutron flux (B.2.3.7) former bolts)

Water Chemistry (B.2.3.2)

Baffle foFFfleF StFUctuml Stainless ReacteF ceelant Gmcking ASMe Sectien :XI asseffil:lly (eaffle suppert steel Neutrnn flux bess ef lnseFVice lnspectien, foFFfleF eelts)

FflateFial Suesectiens IWB, PNC,

,,.,,..., 1n"...,.,

Baffle-former Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals assembly (baffle-support steel Neutron flux toughness (B.2.3.7) former bolts)

Changes in dimensions Loss of preload Loss of material Baffle foFFfleF StFuctuml Stainless ReacteF ceelant bess ef ReacteF Vessel Internals assefflely (eaffle sup pert steel Neutrnn flux FflateFial (B.2.3.7) 1'

- -- *~-

NUREG-2191 Table 1 Notes Item Item IV.B2.RP 388 3.1 1, 059a Qf IV.B2.RP-354 3.1-1, 059a gA IV.B2.RP-296 3.1-1, 059a Qf IV.B2.RP-271 3.1-1, 053a A

B IV.B2.RP 382 3.1 1, 032 A

IV.B2.RP-354 3.1-1, 059a gA IV.B2.RP 296 3.11, 059a Q

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 22 of 38 SLRA Enclosure 3 Attachment 1 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 - Summarv of Ac:iinc:i Manac:iement Evaluation Component Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management Bottom mounted Structural Stainless Reactor coolant Cracking Reactor Vessel Internals 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 instrumentation support steel Neutron flux fracture (B.2.3.7)

(column bodies) toughness Loss of material BetteFfl Ffl91::1Ate8 8tF1::1ct1::1ral Stai Riess ReacteF ceelaAt CrackiAg Reactm Vessel IAtemals iAStFl::!FfleAtatieA Sl::lflfl9Ft 5teel Ne1::1trnA fl1::1x (B.2.3.7)

(f11::1x thiFflble PFeSSl::IFO WateF CheFflistrr tH9e-&}

I D 'l '> 'l \\

BetteFfl Ffl91::1Ate8 8tF1::1ct1::1ral Stai Riess ReacteF ceelaAt bess ef fFact1::1Fe ReacteF Vessel IAtemals iAStFl::!FfleAtatieA Sl::lflfl9Ft 5teel Ne1::1trnA fl1::1x te1::1ghAeSS (B.2.3.7)

(f11::1x th iFflble PFOSSl::IFO ChaAges iA tH9e-&}

EliFfleASi9AS Bottom mounted Structural Stainless Reactor coolant Loss of Flux Thimble Tube instrumentation support steel Neutron flux material Inspection (B.2.3.24)

(flux thimble Pressure tubes) boundary Control rod guide Structural Stainless Reactor coolant Cracking Reactor Vessel Internals tube assembly support steel Neutron flux (B.2.3.7)

(guide cards)

Cast Water Chemistry austenitic (B.2.3.2) stainless steel CeAtrnl Fee g1::1iEle 8tF1::1ct1::1ral Gast ReacteF ceelaAt bess ef fract1::1Fe Reactm Vessel IAtemals t1::1be asseFflbly Sl::lflfl9Ft a1::1steAitic Ne1::1trnA fl1::1x te1::1ghAeSS (B.2.3.7)

(g1::1iEle caFEls) staiAless 5teel NUREG-2191 Table 1 Notes Item Item IV.B2.RP-293 3.1-1, 053b 8A

§ IV. B2. RP-29~QB 3.1-1, 059b GA IV.B2.RP 355 3.1 1, 053c G

IV.B2.R 424 3.11, 119

~

IV. B2. RP-284 3.1-1, 054 A

IV. B2. RP-296ag 3.1-1, 053a GA

§ IV.B2.RP 297 3.1 1, 059a G

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 23 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management Control rod guide Structural Stainless Reactor coolant Loss of Reactor Vessel Internals tube assembly support steel Neutron flux material (8.2.3.7)

(guide cards)

Cast Loss of austenitic fracture stainless toughness steel GeAtrnl Fee g1:1iee StF1:1ctmal Stai Riess ReacteF ceelaAt GrnckiAg ReacteF Vessel IAtemals t1:18e asseFflsly s1:1ppert

~

Ne1:1trnA f11:1x (B.2.3.7)

(g1:1iee caFes)

WateF GheFflistry I D

'"I ".l '"I \\

GeAtFel Fee g1:1iee StFl:lctmal Stai Riess ReacteF ceelaAt Less ef ReacteF Vessel IAtemals t1:18e asseFflsly s1:1ppert

~

Ne1:1tFeA f11:1x Fflatmial (B.2.3.7) 1-

, \\

.. -i- --

Control rod guide Structural Stainless Reactor coolant Cracking Reactor Vessel Internals tube assembly support steel Neutron flux (8.2.3.7)

(lower flange Cast Water Chemistry welds in austenitic (8.2.3.2) 12eri12heral stainless assemblies) steel Control rod guide Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals tube assembly support steel Neutron flux toughness (8.2.3.7)

(lower flange Cast welds in austenitic 12eri12heral stainless assemblies) steel GeAtrnl Fee g1:1iee StF1:1ct1:1rnl Stai Riess ReacteF ceelaAt GrnckiAg ReacteF Vessel IAtemals t1:18e asseFflsly Si:!ppert

~

Ne1:1trnA fl1:1x (B.2.3.7)

(ie'..veF flaAge VVatm GheFflistry WBW}

I D

'"I ".l '"I \\

GeAtFel Fee g1:1iee StF1:1ct1:1rnl Stai Riess ReacteF ceelaAt Less ef fract1:1Fe ReacteF Vessel IAtemals t1:18e asseFflsly s1:1ppert

~

Ne1:1tFeA fl1:1x te1:1ghAess (B.2.3.7)

(leweF flaAge WBW}

NUREG-2191 Table 1 Notes Item Item IV. 82. RP-296 3.1-1, 059a gA IV.B2.RP 298 3.1 1, 053a G

IV.B2.RP 296 3.1 1, 059a g

IV. 82. RP-298 3.1-1, 053a gA

.§ IV. 82. RP-297 3.1-1, 059a gA IV.B2.RP 298 3.1 1, 053a g

IV.B2.RP 297 3.1 1, 059a g

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 24 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management Control rod Structural Stainless Reactor Loss of Reactor Vessel guide tube SUQQOrt steel coolant fracture Internals {B.2.3.7}

assembl~ {lower Cast Neutron flux toughness flange welds in austenitic non-12eriQheral stainless assemblies) steel Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals 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 assembly (barrel support steel Neutron flux toughness (8.2.3.7) former bolts)

Changes in dimensions Loss of preload Loss of material Gem l::laFFel StF1:JGt1:JFal Stai A less ReaGteF GeelaAt Less ef ReaGteF Vessel IAteFAals asseml::lly tl::laFFel Sl:Jppert steel

~Je1:JtrnA flw~

mate Fial tl9.2.a.7)

~~

h Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (core support steel Neutron flux (8.2.3.7) barrel flange)

Flow Water Chemistry distribution (B.2.3.2)

GeFe l::laFFel StF1:JGt1:JFal Stai A less ReaGteF GeelaAt GFaGkiAg ASME: SeGtieA :XI asseml::lly tGeFe Sl:Jppert steel Ne1:JtrnA f11:Jx Less ef IAseFViGe IAspeGtieA, l::laFFel flaAge) f.lew matmial Si:Jl::lSeGtieAS IW19, IWG, ElistFil::l1:JtieA

--rl 1111/n I C '"> ') ~ \\

Core barrel Structural Stainless Reactor coolant Loss of Reactor Vessel Internals assembly (core support steel Neutron flux material (8.2.3.7) barrel flange)

Flow distribution NUREG-2191 Table 1 Notes Item Item IV.82.RP-297a 3.1-1,059b A

IV.82.RP-273 3.1-1, 053b BA

§ IV.82.RP-274 3.1-1, 059b BA IV.192.RP 299 a.1 1, Qa9a G,4 IV.82.RP-3.1-1, 053c GA 2-W345a

§ IV.192.RP ag2 a.1 1, Ga2 A

IV. 82. RP-345 3.1-1, 059c BA

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021 -113 Attachment Page 25 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management GeFe 13aFFel StFuctuml Stainless ReacteF ceelant Gmcking ReacteF Vessel lntemals assernl31y (ceFe suppert

~

Neutrnn flux (8.2.d.7) 13aFrnl eutlet VVatm Ghernistr; I D'"> 'l ') \\

GeFe 9aFFel StFuctuml Stainless ReacteF ceelant Gmcking ASME: Sectien XI assernl31y (ceFe suppert

~

Neutrnn flux Less ef lnseFVice lnspectien, 13aFFel eutlet rnateFial Sul3sectiens IW8, IVVG,

.J l\\M,-, I D

'"> 'l 1 \\

GeFe 13aFFel StFUctuml Stainless ReacteF ceelant Less ef ReacteF Vessel lntemals assernl31y (cern sup pert

~

Neutrnn flux mate Fial (B.2.d.7) 13aFFel eutlet

- 1,J \\

Core barrel Structural Stainless Reactor coolant Ghanges in Reactor Vessel Internals assembly (lower support steel Neutron flux dirnensiens (8.2.3.7) axial welds)

Loss of material Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (lower support steel Neutron flux (8.2.3.7) axial welds)

Water Chemistry (8.2.3.2)

GeFe 13aFFel StFUctuml Stainless ReacteF ceelant Grncking ASME: Sectien XI assernl31y (leweF sup pert

~

Neutrnn flux Less ef lnseFVice lnspectien, axial welds) mate Fial Sul3sectiens IW8, IVl/G,

~--'I\\/\\/,-, I D'"> 'l 1 \\

I Core barrel Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals assembly (lower support steel Neutron flux toughness (8.2.3.7) axial welds)

Changes in dimensions Core barrel Structural Stainless Reactor coolant Ghanges in Reactor Vessel Internals assembly (lower support steel Neutron flux dirnensiens (8.2.3.7) flange weld)

Loss of material NUREG-2191 Table 1 Notes Item Item IV.82.RP 278 d.11, 05dl3 s

IV.82.RP ;:ig2

i.1 1, o;

i2 A

IV.82.RP 29013 d.1 1, 05913 G

IV. 82. RP-27 4 3.1-1, 059b Gf IV.82.RP-387a 3.1-1, 053b SA

~

IV.82.RP ;:ig2 d.1 1, Od2 A

IV. 82. RP-388a 3.1-1, 059b SA IV.82.RP-274 3.1-1, 059b Gf

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 26 of 38 SLRA Enclosure 3 Attachment 1 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 - Summarv of Aaina Manaaement Evaluation Component Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Manaaement Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (lower support steel Neutron flux (8.2.3.7) flange weld)

Water Chemistry (B.2.3.2)

Core barrel Structural Stainless Reactor Loss of Reactor Vessel assembl)l {lower su1;mort steel coolant fracture Internals {B.2.3.7}

flange weld}

Neutron flux toughness Changes in dimensions GeFe 9aFFel StFUctural Stainless ReacteF ceelant Gracking ASME Sectien XI asseFA91y (leweF suppert stecl Neutron flux Less ef lnseFVice lnspectien, flange weld)

FAateFial Su9sectiens IW8, IWG,

'""',.., "" ~ \\

~

~ -*-*

GeFe 9aFFel StFUctural Stainless ReacteF ceelant Less ef fFactuFe ReacteF Vessel Internals asseFA91y (leweF suppert stecl Neutron flux teughness (8.2.3.7)

.Cl---.,-.,

• -1.J\\

.~-

GeFe 9aFFel StFuctural Stainless ReacteF ceelant Ghanges in ReacteF Vessel Internals asseFA91y (leweF suppert stecl Neutron flux diFAensiens (8.2.3.7)

.~..

Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (lower support steel Neutron flux (8.2.3.7) girth weld)

Water Chemistry (8.2.3.2)

GeFe 9aFFel StFuctural Stainless Reacter ceelant Gracking ASME Sectien XI asseFA91y (levier suppert stecl Neutron flux Less ef lnseFVice lnspectien, girth weld)

FA ate rial Su9sectiens l'N8, IVVG,

.J ""' rn,.,.., ~ \\

ur "\\;;,(.

v L.J

--c:T.~

Core barrel Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals assembly (lower support steel Neutron flux toughness (8.2.3.7) girth weld)

Changes in dimensions Gem earrel Structural Stainless Reacter ceelant Ghanges in Reacter Vessel Internals asseFA91y (FAiddle suppert stecl Neutron flux diFAensiens (8.2.3.7)

---'~I

• -*.J-\\

NUREG-2191 Table 1 Notes Item Item IV. 82. RP-280 3.1-1, B.4A 053a.Q

§ IV.B2.RP-280a 3.1-1, 059b A

IV.82.RP 382 3.1 1, 032 A

IV.82.RP 388a 3.1 1, 0599 Q

IV.82.RP 270 3.1 1, 059a Q

IV. 82. RP-387 3.1-1, 053a gA

§ IV.82.RP 382 3.1 1, 032 A

IV. 82. RP-388 3.1-1, 059a gA IV.82.RP 274 3.1 1, 0599 Q

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 27 of 38 SLRA Enclosure 3 Attachment 1 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 - Summarv of Aciinci Manaciement Evaluation Component Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Manaciement Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (middle support steel Neutron flux (8.2.3.7) axial welds)

Water Chemistry (8.2.3.2)

Gem 9aFFel StFuctuml Stainless ReacteF ceelant Gmcking ASMe Sectien XI assern91y tfflielelle suppert stecl Neutrnn flux Less ef lnseFviee lnspeetien, axial welels) rnateFial Su9sectiens IWB, IWG,

~--' l\\MI""\\ I D '1 ".l

~ \\

I Core barrel Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals assembly (middle support steel Neutron flux toughness (8.2.3.7) axial welds)

Changes in dimensions Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (upper support steel Neutron flux (8.2.3.7) axial weld)

Water Chemistry (B.2.3.2)

GeFe 9aFFel StFuctuml Stainless ReaeteF ceelant Gmcking ASMe Sectien XI assern91y tuppeF suppert stecl Neutrnn flux Less ef lnseFVice lnspectien, axial welel) rnateFial Su9seetiens IWB, IWG,

,,..J llMI""\\ I D '1 ".l

~ \\

Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (upper support steel Neutron flux (B.2.3.7) flange weld)

Water Chemistry (8.2.3.2)

GeFe 9aFFel StFUctuml Stainless ReacteF ceelant Gmcking ASMe Sectien XI assern91y tuppeF suppert stecl Neutrnn flux Less ef lnseFvice lnspectien, flange weld) rnatmial Su9sectiens IVVB, IVVG,

~~,..J 11*11""\\ I D '1 ".l

~

\\

Core barrel Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (upper support steel Neutron flux (B.2.3.7) girth weld)

Water Chemistry (8.2.3.2)

NUREG-2191 Table 1 Notes Item Item IV.82.RP-387a 3.1-1, 053b SA

§ IV.B2.RP 382 3.1 1, 032 A

IV. 82. RP-388a 3.1-1, 059b SA IV.82.RP-3.1-1, 053b SA 280Jg+a

§ IV.B2.RP 382 3.1 1, 032 A

IV.82.RP-276 3.1-1, 053a SA

§ IV.B2.RP 382 3.11,032 A

IV.82.RP-3.1-1, As.,+

~280 053.Qa

§

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 28 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management GeFe eaFFel StFUctuml Stainless ReacteF ceelant Gmcking ASMe Sectien XI assemely (uppeF sup pert

~

Neutrnn flux Less ef I nseFvice I nspectien, girth weld) mateFial Suesectiens II/VB, IWG,

~ -' 1111/n / D '> '} ~ \\

LeweF ceFe plate StFuctuml Stainless ReacteF ceelant Ghanges in ReacteF Vessel Internals (fuel alignment suppert

~

Neutrnn flux dimensiens (B.2.3.7)

~

LeweF ceFe plate StFUctuml Stainless ReacteF ceelant Gmcking ReacteF Vessel Internals (fuel alignment sup pert

~

Neutrnn flux (B.2.3.7) fAAsj WateF Ghemistry I D '> '} ') \\

Lower core plate Structural Stainless Reactor coolant Less ef fFacturn Reactor Vessel Internals (fuel alignment support steel Neutron flux teughness (8.2.3.7) pins)

Loss of material LeweF internals StFuctuml Stainless ReacteF ceelant Ghanges in ReacteF Vessel Internals asseFAely (leweF suppert

~

Neutrnn flux dimensiens (B.2.3.7) cern plate)

~

distFieutien Lower internals Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (lower support steel Neutron flux (8.2.3.7) core plate)

Flow Water Chemistry distribution (8.2.3.2)

LeweF internals StFUctuml Stainless ReacteF ceelant Gmcking ASMe Sectien :XI assemely (leweF sup pert

~

Neutrnn flux Less ef lnseFVice lnspectien, cern plate)

~

mate Fial Suesectiens IWB, IWG, distFibutien

~--' 1111/n / D '> '} ~ \\

Lower internals Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals assembly (lower support steel Neutron flux toughness (8.2.3.7) core plate)

Flow Loss of distribution material Changes in dimensions NUREG-2191 Table 1 Notes Item Item IV.B2.RP 382 3.1 1, 032 A

IV.B2.RP 270 3.1 1, 059a G;-4 IV.B2.RP 289 3.1 1, 053G f)

IV.82.RP-3.1-1, GA

~24 119GWG IV.B2.RP 270 3.1 1, 059a G;-4 IV.82.RP-289 3.1-1, 053c gA

§ IV.B2.RP 382 3.1 1, 032 A

IV. 82. RP-288 3.1-1, 059c GA

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 29 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management Lower internals Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (lower support steel Neutron flux (8.2.3.7) support forging)

Water Chemistry (8.2.3.2) bev.1eF iAtemals StFUctural Stai A less ReacteF ceelaAt CrackiAg ASMe SectieA XI asseFAely (le1A<eF sup pert steel NeutrnA flux bess ef IAseFViee IAspeetieA, suppert feFgiAg)

FAateFial SuesectieAS IV\\!8, PNC,

-~...i 11/\\ln re..,'.! 1 \\

Lower support Structural Stainless Reactor coolant Loss of Reactor Vessel Internals assembly (lower support steel Neutron flux fracture (8.2.3.7) support column toughness bodies)

Changes in dimensions Lower support Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (lower support steel Neutron flux (8.2.3.7) support column Water Chemistry bodies)

(8.2.3.2) bev.1er suppert Structural Stai A less ReacteF ceelaAt CrackiAg ASMe SectieA XI asseFAely (leweF sup pert steel NeutrnA flux bess ef IAseFViee IAspeetieA, suppert celuFAA FAateFial SuesectieAS IW8, PNC, h-..J;~~\\

'""' / [")....,., ~ \\

'"-~

beweF suppert StFUctural Stai A less ReacteF eeelaAt bess effFaetuFe ReacteF Vessel IAtemals asseFAely (lev.1eF sup pert steel NeutrnA flux teughAess (8.2.3.7) suppert celuFAA

,\\

Lower support Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (lower support steel Neutron flux (8.2.3.7) support column Water Chemistry bolts)

(8.2.3.2) beweF suppert StFuctural Stai A less ReacteF ceelaAt CrackiAg ASMe SectieA XI asseFAely (le1A1eF suppert steel NeutrnA flux bess ef IAseFVice IAspectieA, suppert eeluFAA FAateFial SueseetieAS IW8, IVVC, OOltsi

~-_, 11~1n re..,'.! 1 \\

NUREG-2191 Table 1 Notes Item Item IV.82.RP-291 a 3.1-1, 053b gA

~

IV.82.RP 382 3.1 1, 032 A

IV.82.RP-3.1-1, 059b GA 2-74295 IV. 82. RP-29-1-a1 3.1-1, 053b GA IV.82.RP 382 3.11, 032 A

IV.82.RP 290a 3.1 1, 0599 g

IV. 82. RP-286 3.1-1, 053b gA

~

IV.82.RP 382 3.1 1, 032 A

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 30 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management Lower support Structural Stainless Reactor coolant Loss of fracture Reactor Vessel Internals assembly (lower support steel Neutron flux toughness (8.2.3.7) support column Loss of preload bolts)

Changes in dimensions Loss of material beweF suppeA:

StFUetuml Stainless ReaeteF eeelant bess ef Reaetm Vessel Internals assembly (leweF suppeA:

~

Neutrnn flux mateFial (B.2.3.7) suppeA: eelumn

~

No additional Structural Nickel alloy Reactor coolant None Reactor Vessel Internals measures support Stainless Neutron flux (8.2.3.7) components Flow steel distribution Radial suppeA:

StFuetuml

~ ReaeteF eeelant Cmcking ASMe Sectien XI keys suppeA:

Neutrnn flux bess ef lnsei=vice lnspectien, mateFial Subsectiens IWB, PNC,

~--' 11~1n 1 0,.., ".l

~

\\

Radial support Structural Ste I lite Reactor coolant Weafloss of Reactor Vessel Internals keys support Neutron flux material (8.2.3.7)

Reactor vessel Structural Cast Reactor coolant Cumulative TLAA - Section 4.3.1, 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 shield Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (thermal support steel Neutron flux (8.2.3.7) shield flexures)

Water Chemistry (8.2.3.2)

NUREG-2191 Table 1 Notes Item Item IV. 82. RP-287 3.1-1, 059b BA IV.B2.RP 290b 3.1 1. 059b G

IV.82.RP-265 3.1-1, 055c BA IV.B2.RP 382 3.1 1, 032 F-NooelV.82.R-Nooe3.1-1,

~A 424 119 IV. 82. RP-303 3.1-1, 003 BA IV. 82. RP-302 3.1-1, 053a BA

§

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 31 of 38 SLRA Enclosure 3 Attachment 1 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 Type Intended Material Environment Aging Effect Aging Management Function Requiring Program Management blppeF seFe plate Strustural Stainless geaster seelant Crasking geasteF Vessel lntemals (fuel alignment sup pert

-steel Neutren flux (B.2.3.7)

~

VVatOF Chemistry r o..., '>...,,

Upper core plate Structural Stainless Reactor coolant bess ef fFasture Reactor Vessel Internals (fuel alignment support steel Neutron flux teughness (8.2.3.7) pins)

Loss of material Upper internals Structural Stainless Reactor coolant Cracking Reactor Vessel Internals assembly (upper support steel Neutron flux (B.2.3.7) core plate)

Water Chemistry (8.2.3.2) blpper intemals StFUstural Stainless geaster seelant Crasking ASME Sestien XI assemely (upper sup pert

-steel Neutren flux bess ef lnsePJise lnspestien, sere plate) mateFial Suesestiens IWB, IWC,

..J '"tr"\\ r o..., '> ~ '

Upper internals Structural Stainless Reactor coolant Loss of Reactor Vessel Internals assembly (upper support steel Neutron flux material (B.2.3.7) core plate) bess ef fraeture

.........,,..h,.,,--.l""

Generic Notes NUREG-2191 Table 1 Notes Item Item IV.B2RP 289 3.11, 0536 G

IV.82.RP-3.1-1, GA

~24 GW6119 IV.82.RP-291 b 3.1-1, 053b EA

~

IV.B2RP 382 3.1 1, 032 A

IV. 82. RP-290ba 3.1-1, 059b GA 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.

8. 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 Censistent vvith NbJgEG 2191 material, envirenment, and aging effest eut a different aging management pregram is sredited er NbJgEG 21 91

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 32 of 38 SLRA Enclosure 3 Attachment 1 Table 3.1.2-2, pages 3.1-75 through 3.1-86, is revised as follows:

identifies a plant speoifio aging management program.

F.

Material not in NU REG 2191 for this oomponent.

J.

Neither the oomponent nor the material and environment oombination is evaluated in NUREG 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 FMECi\\ analysis described further in Appendix C. Loss of preload is not applicable to flux thimble tubes, and loss of material is addressed by NU REG 2191 item IV.B2.RP 284. Flux thimble tubes are existing program oomponents.

3. \\'Vear 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-2021-113 Attachment Page 33 of 38 SLRA Enclosure 3 Attachment 1 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 Table4.7-1 Examples of Potential Plant-Specific TLAA Topics Applies to PBN PW Rs Reactor pressure vessel underclad cracking No (Note 1)

Leak-before-break Yes Reactor coolant pump flywheel fatigue crack growth Yes Response to NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification" Yes Response to NRC Bulletin 88-08, "Thermal Stresses in Piping Yes Connected to Reactor Cooling Systems" EPRI MRP Cl£cle-based and fluence-based anall£ses in su1mort No of MRP-227

{Note 3}

BWRs and PWRs Fatigue of cranes (crane cycle limits)

Yes Fatigue of the spent fuel pool liner No (Note 2)

Corrosion allowance calculations No (Note 2)

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

Predicted lower limit Yes 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.

SLRA Section N/A 4.7.1 4.7.2 4.7.4 4.3.1 4.3.1 N/A 4.7.6 N/A N/A N/A 4.3.5 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-2021-113 Attachment Page 34 of 38 SLRA Enclosure 3 Attachment 1 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 Management Section Program or Activity (Section) 11 Reactor Vessel Xl.M16A Continue the existing PBN Reactor Vessel Internals AMP, including enhancement Internals (16.2.2.7) to:

a) Implement the guidance in MRP 227 Rev. 1:A as supplemented by the gap analysis, or the latest NRC approved version of MRP 227 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.

Implementation Schedule No later than 6 months prior to the SPEO, i.e.:

PBN1 : 04/05/2030 PBN2: 09/08/2032

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 35 of 38 SLRA Enclosure 3 Attachment 1 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 PBt>J Reactor Vessel Internals AMP (Section B.2.3.7),

the PBN Steam Generators AMP (Section B.2.3.10),

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),

the PBN ASME Section XI, Subsection IWF AMP (Section B.2.3.31 ).

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 36 of 38 SLRA Enclosure 3 Attachment 1 Appendix B Table B-4, page B-18, is revised as follows:'

Table B-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-Accelerated B.2.3.8 No Xl.M17 Yes No Corrosion Bolting Integrity B.2.3.9 No Xl.M18 Yes No Steam Generators B.2.3.10 No Xl.M19 Yes Yes Open-Cycle Cooling B.2.3.11 No Xl.M20 Yes Yes 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 -113 Attachment Page 37 of 38 SLRA Enclosure 3 Attachment 1 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 l'JUREG 2191, Section Xl.M16A provides MRP 227 /\\as the basis for a site specific RVI program. The scope of the PBl'J Reactor Vessel Internals AMP applies the methodology and guidance in MRP 227 Revision 1 A (as supplemented by a gap analysis). MRP 227 Revision 1 Ais the most recent NRG approved guidance for managing PVVR vessel internals and incorporates significant recent operating experience.None.

Point Beach Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2021-113 Attachment Page 38 of 38 SLRA Enclosure 3 Attachment 1 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 15 2019 SE. Revision 1 of the guidelines provides updates based on Revision 1 of the NRC SE for 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.