L-MT-23-025, Subsequent License Renewal Application Supplement 2

Subsequent License Renewal Application Supplement 2

ML23177A218
Person / Time
Site:	Monticello
Issue date:	06/26/2023
From:	Domingos C Northern States Power Co, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-MT-23-025
Download: ML23177A218 (1)

Text

(l Xcel Energy* 2807 West County Road 75 Monticello, MN 55362 June 26, 2023 L-MT-23-025 10 CFR 54.17 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Subsequent License Renewal Application Supplement 2

References:

1) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, Monticello Nuclear Generating Plant Docket No. 50-263, Renewal License Number DPR-22 Application for Subsequent Renewal Operating License dated January 9, 2023, ML23009A353

2) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, Monticello Nuclear Generating Plant Subsequent License Renewal Application Supplement 1 dated April 3, 2023, ML23094A136 Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy hereafter "NSPM", is submitting a supplement to the Subsequent License Renewal Application, listed in Reference 1.

Clarifying information regarding Tables 4.2.3-1 and 4.2.3-2 and an updated reference was provided in Supplement 1, listed in Reference 2. Clari"cations to sections of the SLRA discussed in the breakout audits occurring April through June of 2023 are being provided in this Supplement. Further clari"cations will be provided in subsequent supplements.

In the enclosures, changes are described along with the aected section(s) and page number(s) of the docketed SLRA (Reference 1) where the changes are to apply. For clarity, revisions to the SLRA are provided with deleted text by strikethrough and inserted text by bold red underline.

Docum ent Contro l Desk L-MT-23 -025 Page 2 Summary of Commitments This letter makes new commitments and revisions to existing commitments as explained in the enclosures. Commitments 10, 12, 19, 30, 32, 33, 35, 36, and 37 include additions and revisions.

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

Executed on June 2b , 2023.

Ch~ ,D~o¥m~i-n~ go

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Site Vice President, Monticello and Prairie Island Nuclear Generating Plants Northern States Power Company - Minnesota cc: Administrator, Region 111, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce

Document Control Desk L-MT-23-025 Page 3 Enclosures Index Enclosure Subject No.

01 Resolve Crane TLAA Disposition Inconsistency 02a Revise Table 2.5-1 to Omit Cable Bus 02b Clarify Summary of Aging Management Review Results For Fuse Holders 02c Deletion of Fuse Holders From Item Number 3.6.1-008 02d Deletion of ALE Statement in Plant-Speci"c Note 03 Resolve the TLAA Disposition Inconsistency 04 BWR Vessel Internals Supplements 05a Clarify the Structures for the Fire Protection Barrier Commodity Group 05b Corrections to Fire Protection Screening, AMR, and AMP Items 05c Revise Fire Barrier Penetration Seals to Electrical Penetration Assemblies 06a Upgrade Cathodic Protection System 06b Exception for Existing Back"ll 07 Revise SLRA Section 4.2.4 to Cite 'PTLR' 08 Revise References from 40.40(a) to 50.55a Revise Discussion of Item Numbers 3.2.1-107 and 3.2.1-108 to State Not 09 Applicable

Document Control Desk L-MT-23-025 Page 4 Enclosures Index Enclosure Subject No.

Applicability of SCC and LOM Aging Mechanisms for Stainless Steel in the O-10 Gas Condensate System 11 Separation of Buried Piping Materials Subject to Selective Leaching 12 Revise Fluence Values to 3.68E21 13 Updated Reference to Ranganath Analysis 14 Revise the TLAA Disposition Title to Re"ect 10 CFR 54.21(c)(1)(iii) 15 Inclusion of Discussion of Impact of Deposits on Downstream Components RPV Fatigue TLAA Inconsistency Between Original and Rerate Temperature 16 Changes 17 Submit HELB as a TLAA in Section 4.3.6 18 Flex Power Versus Load Following 19a Selective Leaching Supplements 19b Fire Water Piping Coating 20 Section B.2.3.4 Inclusion of Future Approvals 21 Consistency with Section A.2.2.19 and Table A-3 22 IWF Supplements

Document Control Desk L-MT-23-025 Page 5 Enclosures Index Enclosure Subject No.

Stress Corrosion Cracking in Copper Alloy with Greater Than 15% Zinc 23a Components Exposed to Raw Water Clari"cation that the Requirement in Footnote 7 of Table XI.M27-1 in GALL-23b SLR is Satis"ed Clari"cation of the Trending Process for Flow Testing and Wall Thickness 23c Measurements 23d Removal of Wording Implying that MNGP Has More Than One Unit 24a Update Section A.2.2.9 to Remove Reference to the Erosion Module 24b Add EPRI 3002023786 Guidance to FAC 25 Clari"cation of Sample Main Drain Testing at Risers and Standpipes 26 Masonry Wall Voluntary Supplements 27 ASME Section XI, Subsection IWE AMP Clari"cations Correct Casings and Housings with a Leakage Boundary Intended Function 28a for the Turbine Generator System 28b Addition of EPR Components 29 Corrosion Structural Supplement 30a External Surfaces Monitoring of Mechanical Components Heat Exchangers Revised To Cite Correct Environments And Aging 30b Management Programs

Document Control Desk L-MT-23-025 Page 6 Enclosures Index Enclosure Subject No.

31a Addition of Joint and Penetration Seals Commodity Group 31b Railroad Bay Roo"ng Supplement 31c Add Acceptance Criteria for Element 6 31d Enhancement Consistency 31e Provide Clari"cation for Mislabeled Enhancement Element 31f Clari"cation of 115/345 kV Substation Control House 31g Structures Monitoring Program Inspection Frequency 31h Line Item 3.3.1-111 Clari"ed 31i Removal of Grouted Penetration Seals 32 Correct Drawing from SLR-11929 to SLR-119259 Clarify the Source of the Maximum 7000 Cycles and Clarify Operating 33a Cycles is Equivalent Full Temperature Thermal Cycles 33b Updated Reference to Bellows Fatigue Analysis 34a Revise to Include Trash Rack 34b Item 3.5.1-079 Piles/Plates Clari"cation 34c Revise Table 3.5.2-9 to Cite Correct NUREG-2191 Item

Document Control Desk L-MT-23-025 Page 7 Enclosures Index Enclosure Subject No.

34d Supplement for INS Structural Steel and Structural Bolting 35a Concrete Aging Management Review-Groundwater/Soil 35b Concrete Aging Management Review-Add ASR Detail 35c Concrete Aging Management Review-Settlement 35d Concrete Aging Management Review-Correction Of Omitted Line Item 35e Concrete Aging Management Review-Operating Experience 35f Concrete Aging Management Review-Clari"cation of Freeze-thaw Evaluation 35g Concrete Aging Management Review-Clari"cation of Inconsistencies 35h Concrete Aging Management Review-Clari"cation of Operating Experience

Enclosure 01 Resolve Crane TLAA Disposition Inconsistency

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 1 of 6 Resolve Crane TLAA Disposition Inconsistency Revise SLRA Section 4.6.1 to resolve crane TLAA disposition and other typos Affected SLRA Sections: 3.1.2.2.14, 3.3.2.2.1, Table 3.3-1, Table 4.1-1, and 4.6.1 SLRA Page Numbers: 3.1-18, 3.3-21, 3.3-32, 4.1-3, and 4.6-2 Description of Change:

SLRA Section 3.1.2.2.14 is updated with the correct SRP-SLR Item of 3.1.2.2.14.

SLRA Section 3.3.2.2.1 is updated to add the TLAA Disposition of 10 CFR 54.21(c)(1)(ii).

SLRA Table 3.3-1, Item Number 3.3.1-001 is updated with the correct TLAA Disposition of 10 CFR 54.21(c)(1)(ii).

SLRA Table 4.1-1 is updated with the correct section of 4.6.1 for Fatigue of Cranes SLRA Section 4.6.1 Fatigue of Cranes is updated with the correct TLAA Disposition. The crane load cycle limits have been projected through the SPEO and should be disposition as 10 CFR 54.21(c)(1)(ii).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 2 of 6 SLRA Section 3.1.2.2.14 on page 3.1-18 is revised as follows:

Loss of preload due to thermal or irradiationenhanced stress relaxation in core plate rim holddown bolts, as described in SRPSLR Item 3.3.2.2.143.1.2.2.14, is addressed as a TLAA in Section 4.2.9, Loss of Preload for Core Plate Rim Holddown Bolts.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 3 of 6 SLRA Section 3.3.2.2.1 on page 3.3-21 is revised as follows:

Identification of components subject to this aging effect are addressed in Sections 4.3 and 4.6.1 only and not in AMR Tables 3.3.2X because all Auxiliary Systems components have been dispositioned as 10 CFR 54.21(c)(1)(i) and 10 CFR 54.21(c)(ii) respectively and do not require aging management.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 01 Page 4 of 6 SLRA Table 3.3-1, Item Number 3.3.1-001, on page 3.3-32 is revised as follows:

Table 3.31: Summary of Aging Management Evaluations for the Auxiliary Systems Aging Further Item Aging Effect Management Component Evaluation Discussion Number / Mechanism Program Recommended (AMP)/TLAA 3.3.1001 Steel cranes: Cumulative TLAA, Yes (SRPSLR Consistent with NUREG2191.

bridges, fatigue SRPSLR Section 3.3.2.2.1) structural damage due Section 4.7, The Crane Cycle Limits TLAA is used to manage cumulative members, to fatigue "Other fatigue damage of steel cranes and associated components. This structural PlantSpecific line item is used to evaluate structural items in components TLAAs" Section 3.5. Identification of components subject to this aging exposed to effect are addressed in Section 4.6.1 only and not in AMR any Tables 3.3.2X because all Auxiliary Systems components have environment been dispositioned as 10 CFR 54.21(c)(1)(i)(ii) and do not require aging management.

Further evaluation is documented in Section 3.3.2.2.1.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 5 of 6 SLRA Table 4.1-1 on page 4.1-3 is revised as follows:

Table 4.11 Review of Generic TLAAs Listed in NUREG2192, Tables 4.12 and 4.71 NUREG2192, Table 4.71 - Examples of Potential PlantSpecific TLAA Topics (BWRs, BWRs and PWRs)

Fatigue of Cranes (Crane Cycle Limits) Yes 4.5.14.6.1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 6 of 6 SLRA Section 4.6.1 on page 4.6-2 is revised as follows:

TLAA Disposition: 10 CFR 54.21(c)(1)(i)(ii)

The MNGP crane load cycle limits have been projected through the SPEO in accordance with 10 CFR 54.21(c)(1)(i)(ii).

Enclosure 02a Revise Table 2.5-1 to Omit Cable Bus

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2a Page 1 of 2 Revise Table 2.5-1 to Omit Cable Bus Table 2.5-1, Electrical and I&C Component Commodity Groups Installed at MNGP for In-Scope Systems, Incorrectly Lists Cable Bus.

Affected SLRA Sections: Table 2.5-1 SLRA Page Numbers: 2.5-6 Description of Change:

There are no cable buses within the scope of SLR installed at MNGP. Therefore, cable bus is not an Electrical and I&C Component Commodity Group subject to AMR. SLRA, Table 2.5-1 Electrical and I&C Component Commodity Groups Installed at MNGP for In-Scope Systems incorrectly shows cable bus as a listing in the commodity group for an in-scope system. This change will reflect that cable bus is not installed at MNGP.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2a Page 2 of 2 SLRA Table 2.5-1 on page 2.5-6 is revised as follows:

Table 2.51 Electrical and I&C Component Commodity Groups Installed at MNGP for InScope Systems Alarm Units Electrical/I&C Light Bulbs Signal Conditioners Penetration Load Centers Solenoid Operators Analyzers Assemblies Loop Controllers Solid State Devices Annunciators Elements Meters Splices Motor Control Batteries Fuse Holders Surge Arresters Centers Cable Bus*

Cable Connections Fuses (Metallic Parts) Motors Switches Cable TieWraps Generators Chargers Power Distribution Circuit Breakers Electric Heaters Switchgear Panels Converters Heat Tracing Power Supplies Switchyard Bus Communication HighVoltage Radiation Monitors Terminal Blocks Equipment Insulators Electrical Bus Indicators Recorders Thermocouples (aka MetalEnclosed Insulated Cables Regulators Transducers Bus) and Connections Relays Transformers Transmission Electrical Controls Inverters RTDs Conductors and and Panel Internal Connections Component Transmitters Assemblies Isolators Sensors Uninsulated Ground Conductors

• Cable bus is not installed at MNGP.

Enclosure 02b Clarify Summary of Aging Management Review Results For Fuse Holders

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2b Page 1 of 2 Clarify Summary of Aging Management Review Results For Fuse Holders Revise SLRA Section 3.6.2.3.1 to clarify the summary of aging management review results for fuse holders.

Affected SLRA Sections: 3.6.2.3.1 SLRA Page Numbers: 3.6-13 Description of Change:

SLRA Section 3.6.2.3.1 is revised to clarify why there are no aging effects to be managed for fuse holder insulation material.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2b Page 2 of 2 SLRA Section 3.6.2.3.1 on page 3.6-13 is revised as follows:

MNGP fuse holders (not part of active equipment): The insulation material that may be subject to an ALE that may affect insulation resistance are addressed as part of Electrical Insulation for Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements of the fuse holders in this section are located in areas not subject to adverse localized environments (reactor building EL. 935, turbine building EL. 931, and switchyard). Fuse holder insulation material that is not subject to an adverse environment does not have aging effects requiring management. Conservatively, fuse holder insulation material was added to the XI.E1 AMP.

Enclosure 02c Deletion of Fuse Holders From Item Number 3.6.1-008

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2c Page 1 of 2 Deletion of Fuse Holders From Item Number 3.6.1-008 Revise SLRA Table 3.6-1 to delete fuse holders from Item Number 3.6.1-008.

Affected SLRA Sections: Table 3.6-1 SLRA Page Numbers: 3.6-21 Description of Change:

Revise SLRA Table 3.6-1 to delete fuse holders from the list of components in Item Number 3.6.1-008. There are no fuse holders or their associated insulation material that are required to have their aging managed by the XI.E1 AMP. Fuse holders are also not included in the components listed in NUREG-2192, Table 3.6-1, Item 008. Therefore, fuse holders are not applicable to Item Number 3.6.1-008.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 02c Page 2 of 2 SLRA Table 3.6-1 on page 3.6-21 is revised as follows:

Table 3.6-1: Summary of Aging Management Evaluations for Electrical Commodities Aging Further Item Aging Discussion Component Management Evaluation Number Effects/Mechanism Program Recommended 3.6.1008 Electrical Reduced insulation AMP XI.E1, No Consistent with NUREG2191.

insulation for resistance due to "Electrical electrical cables thermal/ Insulation for The Electrical Insulation for Electrical Cables and and connections thermoxidative Electrical Cables Connections Not Subject to 10 CFR 50.49 (including terminal degradation of and Connections Environmental Qualification Requirements (B.2.3.36) blocks, fuse organics, radiolysis, Not Subject to AMP will manage the effects of aging. This AMP holders, etc.) and photolysis (UV 10 CFR 50.49 includes inspection of nonEQ electrical and I&C composed of sensitive materials Environmental penetration cables and connections.

various organic only) of organics; Qualification polymers (e.g., radiationinduced Requirements" MNGP EQ electrical and I&C penetration assemblies EPR, SR, EPDM, oxidation; moisture are covered under the Environmental Qualification of XLPE) exposed to intrusion Electric Equipment program.

an ALE caused by heat, radiation, or moisture

Enclosure 02d Deletion of ALE Statement in Plant-Speci"c Note

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2d Page 1 of 2 Deletion of ALE Statement in Plant-Specific Note Revise Table 3.6.2-1 to delete the sentence regarding fuse holder insulation material in an ALE from Plant-Specific Notes 1.

Affected SLRA Sections: Table 3.6.2-1, Plant-Specific Notes 1 SLRA Page Numbers: 3.6-36 Description of Change:

Revise Table 3.6.2-1 to delete the sentence Fuse holder insulation material in an ALE is managed via the XI.E1 AMP. from Plant-Specific Notes 1. None of the fuse holders within the scope of and screened in for SLR were identified as having exposure to moisture, heat, or radiation that would qualify as an ALE.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2d Page 2 of 2 Table 3.6.2-1 on page 3.6-36 is revised as follows:

PlantSpecific Notes

1. In alignment with GALLSLR, no AMP is required when fuse holders are located in an environment that does not subject them to environmental aging mechanisms. Fuse holder insulation material in an ALE is managed via the XI.E1 AMP. MNGP fuse holders (not in active components) insulation material and environment combination has no aging effects requiring management. See SLRA Section 3.6.2.3.1 for additional information.

Enclosure 03 Resolve the TLAA Disposition Inconsistency

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3 Page 1 of 2 Resolve the TLAA Disposition Inconsistency For SLRA Section 4.3.4, revise the TLAA Disposition: heading to reflect 10 CFR 54.21(c)(1)(i).

Affected SLRA Sections: 4.3.4 SLRA Page Numbers: 4.3-12 Description of Change:

The TLAA disposition heading of SLRA Section 4.3.4 is incorrect. The TLAA was dispositioned as remaining valid and bounding for the SPEO in accordance with 10 CFR 54(c)(1)(i) not 10 CFR 54(c)(1)(ii). The heading is revised to reflect 10 CFR54(c)(1)(i). SLRA Table 4.1-2 has the correct disposition of 10 CFR 54.21(c)(1)(i).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3 Page 2 of 2 SLRA Section 4.3.4 on page 4.3-12 is revised as follows:

TLAA Disposition: 10 CFR 54.21(c)(1)(ii)(i)

The RVI component fatigue analysis for 60 years remains valid and is bounding for the SPEO in accordance with 10 CFR 54.21(c)(1)(i).

Enclosure 04 BWR Vessel Internals Supplements

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4 Page 1 of 5 BWR Vessel Internals Supplements Revise commitments to reflect enhancements to BWRVIP-315 Affected SLRA Sections: Section A.2.2.7, Table A-3, Section B.2.3.7 SLRA Page Numbers: A-15, A-57, B-60, B-63 Description of Change:

The discussion regarding BWRVIP-315 in Sections A.2.2.7 and B.2.3.7 are revised to note that there are comments to BWRVIP-315 safety evaluation submitted by letter dated January 20, 2022 (ML22025A113) that are currently being reviewed by the NRC. Language concerning continuing to implement the most recent NRC-approved versions of the BWRVIP guidance is also added explicitly to Table A-3, Commitment 10 and B.2.3.7 enhancements.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4 Page 2 of 5 SLRA Appendix A.2.2.7 on page A-15 is revised as follows:

A.2.2.7 BWR Vessel Internals The MNGP BWR Vessel Internals program is an existing program that includes inspections and flaw evaluations in conformance with the guidelines of applicable staff-approved BWRVIP documents and provides reasonable assurance of the long-term integrity and safe operation of BWR vessel internal components that are fabricated of nickel alloy and stainless steel.

Available industry guidance includes time-dependent assumptions regarding component degradation mechanisms which have only been evaluated for 60 years of operation. To address this, NUREG-2192 includes three further evaluation items for an SLR applicant to address regarding BWR reactor vessel internals components aging mechanisms (3.1.2.2.12 through 3.1.2.2.14). In response, the BWRVIP developed BWRVIP-315 to disposition these further evaluations and identify any necessary plant-specific evaluations. For MNGP, there are no additional components subject to degradation mechanisms for SLR. However, to implement the guidance in BWRVIP-315, some BWRVIP guidance documents require enhancement and revision (as shown in BWRVIP-315) in order to address operation beyond 60 years. These are documented in Appendix C. The MNPG BWR Vessel Internals AMP recognizes the BWRVIP SLR guidance continues to develop and will continue to implement the most recent NRC-approved versions of the BWRVIP guidance. Note that the NRC is reviewing a revised version of BWRVIP-315 and requested comments on a draft safety evaluation, which the BWRVIP provided in a letter dated January 20, 2022 (Reference ML22025A113).

The BWR water chemistry is controlled per the EPRI guidelines of BWRVIP-190 Revision 1 (TR-3002002623), Volume 1: BWR Water Chemistry Guidelines -

Mandatory, Needed, and Good Practice Guidance.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 04 Page 3 of 5 Table A-3 on page A-57 is revised as follows:

Aging Management NUREG-2191 Implementation No. Program or Activity Commitment Section Schedule (Section) 10 BWR Vessel XI.M9 The BWR Vessel Internals AMP is an existing program that will be enhanced No later than 6 months Internals to: prior to the SPEO, or no (A.2.2.7) later than the last a) Include implementation of BWRVIPs A, R4-A, A, refueling outage prior to and -183-A as indicated in BWRVIP-315. the SPEO b) Implement BWRVIP-315-A and subsequent revisions approved by the NRC for MNGP to use during SPEO.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4 Page 4 of 5 SLRA Section B.2.3.7 on page B-60 is revised as follows:

B.2.3.7 BWR Vessel Internals The BWR Vessel Internals AMP is an existing condition monitoring and mitigative program that includes inspections and flaw evaluations in conformance with the guidelines of applicable staff-approved BWRVIP documents and provides reasonable assurance of the long-term integrity and safe operation of BWR vessel internal components that are fabricated of nickel alloy and stainless steel.

Available industry guidance includes time-dependent assumptions regarding component degradation mechanisms which have only been evaluated for 60 years of operation. To address this, NUREG-2192 includes three further evaluation items for an SLR applicant to address regarding BWR reactor vessel internals components aging mechanisms (3.1.2.2.12 through 3.1.2.2.14). In response, the BWRVIP developed BWRVIP-315 to disposition these further evaluations and identify any necessary plant-specific evaluations. For MNGP, there are no additional components subject to degradation mechanisms for SLR. However, to implement the guidance in BWRVIP-315, some BWRVIP guidance documents require enhancement and revision (as shown in BWRVIP-315) in order to address operation beyond 60 years.

These are documented in Appendix C. The MNPG BWR Vessel Internals AMP recognizes the BWRVIP SLR guidance continues to develop and will continue to implement the most recent NRC-approved versions of the BWRVIP guidance. Note that the NRC is reviewing a revised version of BWRVIP-315 and requested comments on a draft safety evaluation, which the BWRVIP provided in a letter dated January 20, 2022 (Reference ML22025A113).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4 Page 5 of 5 SLRA Section B.2.3.7 on page B-63 is being revised as follows:

Enhancements Element Affected Enhancement

1. Scope of Program Continue to implement the most recent NRC-approved versions of BWRVIP guidance, specifically BWRVIPs A, R4-A, A, and -183-A as indicated in BWRVIP-315.

1. Scope of Program Implement BWRVIP-315-A and subsequent revisions approved by the NRC for MNGP to use during SPEO.

Enclosure 05a Clarify the Structures for the Fire Protection Barrier Commodity Group

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5a Page 1 of 2 Clarify the Structures for the Fire Protection Barrier Commodity Group Revise SLRA Section 2.4.6 to clarify the structures in the sentence Curbs, dikes, concrete component other than barriers are evaluated in the structure where they are located.

Affected SLRA Sections: 2.4.6 SLRA Page Numbers: 2.4-15 Description of Change:

SLRA Section 2.4.6 on page 2.4-15 is revised to clarify the structures in the sentence Curbs, dikes, concrete components other than barriers are evaluated as part of the structure where they are located. The sentence is clarified by adding (e.g., walls, ceilings, floors) at the end of the sentence. This revision is supported by SLRA Section 2.1.4.2.1. The word and is added between dikes and other concrete components in the list at the beginning of the sentence.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5a Page 2 of 2 SLRA Section 2.4.6 on page 2.4-15 is revised as follows:

Curbs, dikes, and concrete components other than barriers are evaluated as part of the structure where they are located (e.g., walls, ceilings, floors). Fire detection and alarm system (e.g., smoke detectors), and fire suppression (e.g., automatic sprinklers, automatic halon systems) are evaluated in the FIR System (Section 2.3.3.9). The dieseldriven fire pump is evaluated in the FIR System (Section 2.3.3.9). Information on the Fire Protection Barriers Commodity Group is found in Section 10.3 and Appendix J of the USAR.

Enclosure 05b Corrections to Fire Protection Screening, AMR, and AMP Items

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 1 of 8 Corrections to Fire Protection Screening, AMR, and AMP Items Revise Table 2.4-6 , Table 3.3-1 , Section 3.5.2.1.6, Table 3.5.2-6, and Section B.2.3.15 to correct items associated with FP Screening, AMR, and AMP items.

Affected SLRA Sections: Table 2.4-6, Table 3.3-1, Section 3.5.2.1.6, Table 3.5.2-6, and Section B.2.3.15 SLRA Page Numbers: 2.4-16, 3.3-90, 3.5-7, 3.5-96, 3.5-97 and B-108, Description of Change:

SLRA Table 2.4-6 on page 2.4-16 is revised to include FP guard pipe and rigid board (gypsum walls, etc.). This change is supported by SLRA Section 2.4.6.

SLRA Table 3.3-1 on page 3.3-90 is revised to omit HELB Barrier in the discussion column.

HELB Barrier was incorrectly cited for item number 3.3.1-268 as HELB Barrier was not cited as an intended function for any components in SLRA Table 3.5.2-6. No components with a HELB barrier intended function in the other SLRA AMR Tables cites Item 3.3.1-268.

Section 3.5.2.1.6 on page 3.5-7 is revised to include delamination and separation as aging affects. It is also revised to include gypsum as a material.

SLRA Table 3.5.2-6 on page 3.5-96 is revised to cite separation as an Aging Effect Requiring Management for Fireproofing and Non-Metallic Fireproofing. The Table is revised to cite cracking, delamination, and separation as Aging Effect Requiring Management for Thermal Fiber. The redundant rows for Fireproofing and Thermal Fibers were deleted. The Table is also revised to list FP Guard Pipe and Rigid Board (gypsum walls, etc.) as component types.

Additionally, the General Notes for this table on page 3.5-97 is revised to add notes C and F to the general notes and a plant-specific note 1. The plant-specific note states Gypsum drywall is utilized to provide fire barriers at MNGP. The material is not addressed in NUREG-2191, but aging is managed by the Fire Protection AMP consistent with silicate fire barriers SLRA Section B.2.3.15 on page B-108 is revised to include aluminum as a fire protection component material and FP guard pipe as a fire barrier component. The inclusion of aluminum is supported by SLRA Section 3.5.2.1.6 and SLRA Table 3.5.2-6. The inclusion of FP guard pipe is supported by SLRA Section 2.4.6.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 2 of 8 SLRA Table 2.4-6 on page 2.4-16 is revised as follows:

Table 2.46 Fire Protection Barriers Commodity Group Components Subject to Aging Management Review Component Type Component Intended Function(s)

Cable Tray Cover Fire Barrier Fire Barrier Penetration Seals Fire Barrier Fire Damper Housing Fire Barrier Fire Rated Doors Fire Barrier Fireproofing Fire Barrier FP Guard Pipe Fire Barrier Masonry (Block) Walls Fire Barrier Nonmetallic Fireproofing Fire Barrier Rigid Board (Gypsum Walls, etc.) Fire Barrier Structural Fire Barriers (Walls, Ceilings Fire Barrier and Floors)

Thermal fiber Fire Barrier

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 3 of 8 SLRA Table 3.3-1 on page 3.3-90 is revised as follows:

Table 3.31: Summary of Aging Management Evaluations for the Auxiliary Systems Item Aging Effect / Aging Management Further Evaluation Component Discussion Number Mechanism Program (AMP)/TLAA Recommended 3.3.1268 Cementitious Loss of material, AMP XI.M26, Fire No Consistent with NUREG2191.

coating change in Protection fireproofing/fire material The Fire Protection (B.2.3.15) AMP is barriers (Pyrocrete, properties, used to manage loss of material, BIO' K10 Mortar, cracking, change in material properties, cracking, Cafecote, and delamination, and delamination, and separation for other similar separation cementitious coating fireproofing/fire materials) exposed barriers/HELB barriers exposed to air to air indoor uncontrolled.

This line item is used to evaluate structural items in Section 3.5.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 4 of 8 SLRA Section 3.5.2.1.6 on page 3.5-7 is revised as follows:

3.5.2.1.6 Fire Protection Barrier and Commodity Group Materials The materials of construction for the Fire Protection Barriers Commodity Group structural components are:

Aluminum Cementitious Concrete Block Concrete (Reinforced)

Elastomer Gypsum Silicates Steel Aging Effects Requiring Management The following aging effects associated with the Fire Protection Barriers Commodity Group structural components require management:

Change In Material Properties Cracking Delamination Hardening Loss of Material Loss of Strength Shrinkage Separation

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 5 of 8 SLRA Table 3.5.2-6 on page 3.5-96 is revised as follows:

Table 3.5.26: Fire Protection Barriers Commodity Group - Summary of Aging Management Evaluation Aging Component Intended Aging Effect Requiring NUREG- Table 1 Material Environment Management Notes Type Function Management 2191 Item Item Program Fireproofing Fire Cementitious Air - Indoor Cracking Fire Protection VII.G.A806 3.3.1268 A Barrier Uncontrolled Change in Material (B.2.3.15)

Properties Delamination Loss of Material Separation Fireproofing Fire Cementitious Air - Indoor Loss of Material Fire Protection VII.G.A806 3.3.1268 A Barrier Uncontrolled (B.2.3.15)

FP Guard Fire Steel Air - Indoor Loss of Material Fire VII.G.A-789 3.3.1-255 C Pipe Barrier Uncontrolled Protection (B.2.3.15)

Masonry Fire Concrete Air - Indoor Cracking Loss of Fire Protection VII.G.A-626 3.3.1179 A (Block) Walls Barrier Block Uncontrolled Material (B.2.3.15)

Masonry Walls (B.2.3.32)

NonMetallic Fire Cementitious Air - Indoor Cracking Fire Protection VII.G.A806 3.3.1268 A Fireproofing Barrier Uncontrolled Loss of Material (B.2.3.15)

Change in Material Properties Delamination Separation Rigid Board Fire Gypsum Air - Indoor Change in Material Fire None None F, 1 (gypsum Barrier Uncontrolled Properties Protection walls, etc.) Cracking (B.2.3.15)

Delamination Loss of Material Separation Structural Fire Fire Concrete Air - Indoor Cracking Fire Protection VII.G.A90 3.3.1060 A Barriers Barrier (Reinforced) Uncontrolled Loss of Material (B.2.3.15)

(Walls, Air - Outdoor Structures Ceilings and Monitoring

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 6 of 8 Floors) (B.2.3.33)

Thermal Fiber Fire Silicate Air - Indoor Change in Material Fire Protection VII.G.A.807 3.3.1269 A Barrier Uncontrolled Properties (B.2.3.15)

Cracking Delamination Lost of Material Separation Thermal Fiber Fire Silicate Air - Indoor Lost of material Fire Protection VII.G.A.807 3.3.1269 A Barrier Uncontrolled (B.2.3.15)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 7 of 8 SLRA Table 3.5.2-6 on page 3.5-97 is revised as follows:

General Notes A. Consistent with component, material, environment, aging effect, and AMP listed for NUREG2191 line item. AMP is consistent with NUREG2191 AMP description.

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

F. Material not in NUREG-2191 for this component.

PlantSpecific Notes

1. Gypsum drywall is utilized to provide fire barriers at MNGP. The material is not addressed in NUREG-2191, but aging is managed by the Fire Protection AMP consistent with silicate fire barriers.None

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 8 of 8 SLRA Section B.2.3.15 on page B-108 is revised as follows:

B.2.3.15 Fire Protection The MNGP Fire Protection Program is an existing condition and performance monitoring program that manages the identified aging effects for fire barrier penetration seals, fire barriers, structural steel fire proofing materials, fire damper assemblies, fire rated doors and a halon fire suppression system installed in air/gas environments through the use of inspections/testing to detect aging prior to loss of intended function(s). The fire protection components materials include aluminum, carbon steel, cast iron, concrete (masonry) block, cementitious fireproofing (thermal insulation mastic), thermal fiber (silicate), fire stop sealant (silicone, silicone foam, caulk), galvanized steel, gray cast iron, reinforced concrete, rigid board (gypsum walls, etc.), and stainless steel.

The program is effective in detecting the applicable aging effects and as such includes a fire barrier visual inspection program, and a halon fire suppression system inspection. The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals, FP guard pipe, fire barriers (e.g., walls, ceilings, and floors), fireproofing materials, fire damper assemblies, and periodic visual inspection and functional tests of associated fire rated doors to ensure that their functionality is maintained. The Fire Protection Program includes periodic visual inspection and testing of the Cable Spreading Room halon fire suppression system.

Enclosure 05c Revise Fire Barrier Penetration Seals to Electrical Penetration Assemblies

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 1 of 2 Revise Fire Barrier Penetration Seals to Electrical Penetration Assemblies Revise SLRA Section 3.5.2.2.2.4 to state electrical penetration assemblies instead of fire barrier penetration seals.

Affected SLRA Sections: 3.5.2.2.2.4 SLRA Page Numbers: 3.5-35 Description of Change:

Revise SLRA Section 3.5.2.2.2.4 to state electrical penetration assemblies instead of fire barrier penetration seals in the discussion regarding stainless steel supports or anchorage in the air environment (and underground environment in manholes). SLRA Section 3.5.2.2.2.4 incorrectly list fire barrier penetration seals; stainless steel is, instead, used in electrical penetration assemblies associated with primary containment. This is supported by SLRA Table 2.4-1 and Table 3.5.2-1.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 2 of 2 Revise SLRA Section 3.5.2.2.2.4 on page 3.5-35 as follows:

Table 3.51, Item Number 3.5.1100: For stainless steel and aluminum components or connections exposed to air environments, this item number evaluates the components aligned to this item number for loss of material due to pitting and crevice corrosion and cracking due to SCC. The Structures Monitoring (B.2.3.33) AMP will continue to be used to examine the structural components and connections aligned to this item number. The Structures Monitoring (B.2.3.33) AMP requires periodic monitoring of ground/lake water chemistry to verify that it remains nonaggressive. Also, the air environment (and underground environment in manholes) for stainless steel supports or anchorage is not expected to be aggressive enough to cause cracking or localized loss of material for components (stainless steel new fuel storage racks, refueling cavity liner, component supports, anchorages, fire barrier penetration sealselectrical penetration assemblies, insulation jacketing inside containment, aluminum insulation jacketing outside containment, and aluminum manway covers) exposed to indoor, outdoor, or underground air in the presence of wetting.

Enclosure 06a Upgrade Cathodic Protection System

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6a Page 1 of 7 Upgrade Cathodic Protection System Cathodic Protection System is Upgraded Affected SLRA Sections: A.2.2.27, Table A-3, Commitment 30, B.2.3.27 SLRA Page Numbers: A-26, A-27, A-81, A-82, A-84, B-197, B-198, B-199, and B-201 Description of Change:

Upgrade the Cathodic Protection System in accordance with NACE SP0169-2007 for buried steel piping within the scope of the program. The commitment schedule supports this upgrade 5 years before SPEO in order to credit for pre-SPEO inspections. The upgrade to the Cathodic Protection System will meet the -850 to -1200 mV acceptance criteria.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6a Page 2 of 7 Section A.2.2.27 on pages A-26 and A-27 is being revised as follows:

The MNGP Buried and Underground Piping and Tanks AMP, previously known as the Buried Piping and Tanks Inspection Program, is an existing AMP that manages the aging effects associated with the external surfaces of buried and underground piping and tanks such as loss of material and cracking. This AMP addresses piping and tanks composed of metallic (steel and stainless steel) materials that are within the scope of SLR in the CST, EDGs, ESW, fire water, offgas, secondary containment, service and seal water, and wells and domestic water systems. Loss of material is monitored by visual inspection of the exterior surface and wall thickness measurements of the piping. Wall thickness is determined by a nondestructive examination technique such as UT. For steel components, where the acceptance criteria for the effectiveness of the cathodic protection is other than -850 mV instant off, loss of material rates are measured.

This AMP also manages aging through preventive actions (e.g., coatings or wrapping, cathodic protection, and quality of backfill). Annual cathodic protection surveys are conducted. MNGP currently meets the conditions of Preventive Action Category F for inspections of buried steel piping, unless a reevaluation of cathodic protection system performance, future OE, or soil conditions determines that another preventive action category is more applicable. MNGP will refurbish its Cathodic Protection System 5 years prior to the SPEO to meet the acceptance criteria of -850 mV relative to a copper/copper sulfate reference electrode (CSE) (instant off), or acceptance criteria alternatives as outlined in NUREG-2191,Section XI.M41, Subsection 6.m, for buried steel components. The intent is to satisfy conditions of Preventive Action Category C for inspections of buried steel and stainless steel piping during the SPEO, unless a reevaluation of future OE and soil conditions determines that another Preventive Action Category is more applicable. The number of inspections for each 10year inspection period, commencing 10 years prior to the SPEO, are based on the effectiveness of the preventive actions above.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6a Page 3 of 7 Table A-3, Commitment 30 on pages A-81 and A-82 is being revised as follows:

No. Aging NUREG2191 Commitment Implementation Management Section Schedule Program or Activity (Section) 30 Buried and XI.M41 e) Perform inspections of buried and No later than 6 Underground underground piping and tanks in months prior to Piping and accordance with NUREG-2191 Table the SPEO, or no Tanks XI.M41-2 Preventive Action Category later than the last (A.2.2.27) F C for buried steel and stainless refueling outage steel components, unless a prior to the SPEO reevaluation of cathodic protection system performance, future OE, or and Implement the soil conditions determines that another AMP and start Preventive Action Category is more 10-year interval applicable. In the 10-year period prior inspections no to and during SPEO for each 10-year earlier than 10 interval, perform buried and years prior to the underground piping and tanks SPEO.

inspections in accordance with the Preventive Action Category FC as outlined in NUREG-2191 Table XI.M41-2. When the inspections for a given material type is based on percentage of length and results in an inspection quantity of less than 10 feet, then 10 feet of piping is inspected. If the entire run of piping of that material type is less than 10 feet in total length, then the entire run of piping is inspected.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6a Page 4 of 7 Table A-3, Commitment 30 on page A-84 is being revised as follows:

No. Aging NUREG2191 Commitment Implementation Management Section Schedule Program or Activity (Section) q) Refurbish the Cathodic Commitment Protection System to meet the 30q will be recommendations of GALL-SLR implemented 5 Section XI.M41, including the - years prior to 850 mV polarized potential the SPEO in criteria of NUREG-2191, or order to credit acceptance criteria alternatives, the system for and annual system monitoring. pre-SPEO The cathodic protection system inspections.

for buried piping shall also include a limiting critical potential of -1,200 mV to prevent overprotection.

r) The acceptance criterion for the MNGP Cathodic Protection System is -850 mV relative to a CSE (instant off). For locations where the refurbished Cathodic Protection System cannot meet the -850 mV criterion, the acceptance criteria alternatives to the -850 mV criteria will be implemented as outlined in NUREG-2191,Section XI.M41, Subsection 6.m.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6a Page 5 of 7 Section B.2.3.27 on page B-197 and B-198 is being revised as follows:

surveyed locations met the 100 mV polarization criterion. Therefore, the cathodic protection system does not currently meet the acceptance criteria of NACE SP0169-2007 or NACE RP0285-2002 and is not credited as a preventive measure at MNGP. MNGP will refurbish its Cathodic Protection System at least 5 years prior to the SPEO to meet the acceptance criteria of -850 mV relative to a copper/copper sulfate reference electrode (CSE) (instant off), or acceptance criteria alternatives, for buried steel components.

SCC of steel piping can occur in steel piping exposed to a carbonate and/or bicarbonate environment. Soil sample results from MNGP have shown the presence of carbonate and/or bicarbonate in certain locations and non-detectable in other locations. Figure 2 of NACE SP0169-2007 indicates the susceptibily susceptibility of buried steel to SCC is based on temperature and polarized potential of the cathodic protection system. Exposure of steel to this environment also requires a degraded coating, and since steel piping at MNGP is coated in accordance with plant procedures, and plant-specific OE indicates that buried coatings have shown little degradation; this exposure is not expected. Based on soil temperature at the site and good OE with coatings of buried piping, SCC of steel piping is not expected but is conservatively assumed to be applicable at MNGP and will therefore be managed by this AMP.

The cathodic protection system will be upgraded to meet -850 mV acceptance criteria from NUREG-2191. In the event the refurbished Cathodic Protection System cannot meet the -850 mV criterion for all surveyed locations, the acceptance criteria alternatives to the

-850 mV criteria will be implemented as outlined in NUREG-2191,Section XI.M41, Subsection 6.m. These alterative acceptance criteria will be based on the level of cathodic protection that can be achieved for the specific surveyed area(s).

The number of inspections for each 10 year inspection period, commencing 10 years prior to the start of SPEO, are based on the inspection quantities noted in NUREG-2191, Table XI.M41 2 for Category FC. However, changes in plant specific conditions can result in transitioning to a differenthigher number of inspections than originally planned at the beginning of a 10 year period. For example, refurbishmentdegradation of the cathodic protection system, coatings, backfill, or the condition of exposed piping that does not to meet NACE acceptance criteria could result in transitioning to a lower from Preventive Action Category C from to Preventive Action Category F Material No. of Inspections Notes Steel piping 6 1 inspections The smaller of 10.5% of the piping (buried) length or 61 inspections.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6a Page 6 of 7 Section B.2.3.27 on page B-199 is being revised as follows:

Element Affected Enhancement

2. Preventive Actions Update MNGP BUPT AMP procedures as appropriate:

State that new and replacement backfill shall meet the requirements of NACE SP0169-2007 Section 5.2.3 or NACE RP0285-2002 Section 3.6.

Refurbish the Cathodic Protection System at least 5 years prior to the SPEO and perform effectiveness reviews in accordance with Table XI.M41-2 in NUREG-2191,Section XI.M41. The cathodic protection system for buried piping shall also include a limiting critical potential of

-1,200 mV to prevent overprotection.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6a Page 7 of 7 Section B.2.3.27 on page B-201 is being revised as follows:

6. Acceptance Criteria Update MNGP BUPT AMP procedures as appropriate:

For coated piping or tanks, there is either no evidence of coating degradation, or the type and extent of coating degradation is evaluated as insignificant by an individual:

(a) possessing a NACE Coating Inspector Program Level 2 or 3 inspector qualification; (b) who has completed the Electric Power Research Institute Comprehensive Coatings Course and completed the EPRI Buried Pipe Condition Assessment and Repair Training Computer Based Training Course; or (c) a coatings specialist qualified in accordance with an ASTM standard endorsed in RG 1.54, Revision 2, Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants.

Specify that degradation (e.g., coating condition, wall thickness) is projected until the next scheduled inspection. Results are evaluated against acceptance criteria to confirm that the sampling bases (e.g., selection, size, frequency) will maintain the components intended functions throughout the SPEO based on the projected rate and extent of degradation.

Indications of cracking in metallic pipe are managed in accordance with the CAP.

Backfill is acceptable if the inspections do not reveal evidence that the backfill caused damage to the components coatings or the surface of the component (if not coated).

For pressure tests, the test acceptance criteria are that there are no visible indications of leakage, and no drop in pressure within the isolated portion of the piping that is not accounted for by a temperature change in the test media or by quantified leakage across test boundary valves.

Cracks in cementitious backfill that could admit groundwater to the surface of the component are not acceptable.

Specify that the acceptance criteria for the MNGP Cathodic Protection System is -850 mV relative to a CSE (instant off). For locations where the refurbished Cathodic Protection System cannot meet the -850 mV criterion, the acceptance criteria alternatives to the

-850 mV criteria will be implemented as outlined in NUREG-2191,Section XI.M41, Subsection 6.m.

Enclosure 06b Exception for Existing Back"ll

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6b Page 1 of 29 Exception for Existing Backfill Exception for Existing Backfill Affected SLRA Sections: Tables 3.2-1, 3.2.2-6, 3.3-1, 3.3.2-6, 3.3.2-8, 3.3.2-9, 3.3.2-16, 3.3.2-18, 3.4-1, 3.4.2-1, 3.4.2-5, B-4, Sections B.1.1 and B.2.3.27 SLRA Page Numbers: 3.2-28, 3.2-34, 3.2-99, 3.2-100, 3.3-54, 3.3-65, 3.3-132, 3.3-144, 3.3-149, 3.3-167, 3.3-173, 3.3-175, 3.3-179, 3.3-180, 3.3-181, 3.3-191, 3.3-192, 3.3-201, 3.3-203, 3.3-204, 3.3-293, 3.3-298, 3.3-311, 3.3-314, 3.4-27, 3.4-31, 3.4-48, 3.4-95, 3.4-97, B-6, B-21, B-199 Description of Change:

SLRA Section B.2.3.27 is revised to add an exception for backfill in Element 2. Original construction backfill does not meet the guidance in NUREG-2191,Section XI.M41. Impacted Section 3 tables are changed to reflect this exception.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 2 of 29 SLRA Table 3.2-1 on Page 3.2-28 is revised as follows:

Table 3.2-1: Summary of Aging Management Evaluations for the Engineered Safety Features Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program / TLAA Recommended 3.2.1052 Steel piping, piping Loss of material AMP XI.M41, "Buried and No Consistent with NUREG2191 with components exposed due to general, Underground Piping and exception for the Buried and to soil, concrete, pitting, crevice Tanks" Underground Piping and Tanks underground corrosion, MIC (soil (B.2.3.27) AMP.

only)

The Buried and Underground Piping and Tanks (B.2.3.27) AMP is used to managed loss of material of steel piping and piping components exposed to soil.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 3 of 29 SLRA Table 3.2-1 on Page 3.2-34 is revised as follows:

Table 3.2-1: Summary of Aging Management Evaluations for the Engineered Safety Features Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program / TLAA Recommended 3.2.1078 Stainless steel, steel, Cracking due to AMP XI.M41, "Buried and No Consistent with NUREG2191 with aluminum piping, SCC (steel in Underground Piping and exception for the Buried and piping components, carbonate/bicarbon Tanks" Underground Piping and Tanks tanks exposed to ate environment (B.2.3.27) AMP.

soil, concrete only)

The Buried and Underground Piping and Tanks (B.2.3.27) AMP is used to manage cracking in carbon steel piping, piping components exposed to soil.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 4 of 29 SLRA Table 3.2.2-6 on Page 3.2-99 is revised as follows:

Table 3.2.2-6: Secondary Containment - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Carbon Steel Soil (External) Loss of Material Buried and V.E.EP111 3.2.1052 AB Components Boundary Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 5 of 29 SLRA Table 3.2.2-6 on Page 3.2-100 is revised as follows:

Table 3.2.2-6: Secondary Containment - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Carbon Steel Soil (External) Cracking Buried and V.E.E420 3.2.1078 AB Components Boundary Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 6 of 29 SLRA Table 3.3-1 on Page 3.3-54 is revised as follows:

Table 3.3-1: Summary of Aging Management Evaluations for the Auxiliary Systems Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program / TLAA Recommended 3.3.1109 Steel piping, piping Loss of material AMP XI.M41, "Buried and No Consistent with NUREG2191 with components, closure due to general, Underground Piping and exception for the Buried and bolting exposed to pitting, crevice Tanks" Underground Piping and Tanks soil, concrete, corrosion, MIC (soil (B.2.3.27) AMP.

underground only)

Buried and Underground Piping and Tanks (B.2.3.27) AMP is used to manage loss of material of steel piping, piping components, and closure bolting exposed to soil or concrete.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 7 of 29 SLRA Table 3.3-1 on Page 3.3-65 is revised as follows:

Table 3.3-1: Summary of Aging Management Evaluations for the Auxiliary Systems Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program / TLAA Recommended 3.3.1144 Stainless steel, steel, Cracking due to AMP XI.M41, "Buried and No Consistent with NUREG2191 with aluminum piping, SCC (steel in Underground Piping and exception for the Buried and piping components, carbonate/bicarbon Tanks" Underground Piping and Tanks tanks exposed to ate environment (B.2.3.27) AMP.

soil, concrete only)

The Buried and Underground Piping and Tanks (B.2.3.27) AMP is used to manage cracking of steel piping and piping components exposed to soil.

The only component exposed to concrete that is susceptible to cracking is the SLC tank, which is addressed by item 3.3.1230.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 8 of 29 SLRA Table 3.3.2-6 on Page 3.3-132 is revised as follows:

Table 3.3.2-6: Emergency Diesel Generators - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Bolting (Closure) Mechanical Carbon and Soil (External) Loss of Material Buried and VII.I.AP241 3.3.1109 AB Closure Low Alloy Underground Piping Steel Bolting and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 9 of 29 SLRA Table 3.3.2-6 on Page 3.3-144 is revised as follows:

Table 3.3.2-6: Emergency Diesel Generators - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Carbon Steel Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Components Boundary Underground Piping and Tanks (B.2.3.27)

Piping, Piping Pressure Carbon Steel Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Components Boundary Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 10 of 29 SLRA Table 3.3.2-6 on Page 3.3-149 is revised as follows:

Table 3.3.2-6: Emergency Diesel Generators - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Tanks (DG Fuel Oil Pressure Carbon Steel Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Storage Tank) Boundary Underground Piping and Tanks (B.2.3.27)

Tanks (DG Fuel Oil Pressure Carbon Steel Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Storage Tank) Boundary Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 11 of 29 SLRA Table 3.3.2-8 on Page 3.3-167 is revised as follows:

Table 3.3.2-8: Emergency Service Water - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Bolting (Closure) Mechanical Carbon and Soil (External) Loss of Material Buried and VII.I.AP241 3.3.1109 AB Closure Low Alloy Underground Piping Steel Bolting and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 12 of 29 SLRA Table 3.3.2-8 on Page 3.3-173 is revised as follows:

Table 3.3.2-8: Emergency Service Water - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Carbon Steel Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Components Boundary Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 13 of 29 SLRA Table 3.3.2-8 on Page 3.3-175 is revised as follows:

Table 3.3.2-8: Emergency Service Water - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Carbon Steel Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Components Boundary Underground Piping and Tanks (B.2.3.27)

SLRA Table 3.3.2-8 on Page 3.3-179 is revised as follows:

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

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 14 of 29 SLRA Table 3.3.2-9 on Page 3.3-180 is revised as follows:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Bolting (Closure) Mechanical Carbon and Soil (External) Loss of Material Buried and VII.I.AP241 3.3.1109 AB Closure Low Alloy Underground Piping Steel Bolting and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 15 of 29 SLRA Table 3.3.2-9 on Page 3.3-181 is revised as follows:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Fire Hydrant Pressure Ductile Iron Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Boundary Underground Piping and Tanks (B.2.3.27)

Fire Hydrant Pressure Ductile Iron Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Boundary Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 16 of 29 SLRA Table 3.3.2-9 on Page 3.3-191 is revised as follows:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Gray Cast Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Components Boundary Iron Underground Piping and Tanks (B.2.3.27)

Piping, Piping Pressure Gray Cast Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Components Boundary Iron Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 17 of 29 SLRA Table 3.3.2-9 on Page 3.3-192 is revised as follows:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Gray Cast Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Components Boundary Iron (with Underground Piping Internal and Tanks (B.2.3.27)

Coating)

Piping, Piping Pressure Gray Cast Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Components Boundary Iron (with Underground Piping Internal and Tanks (B.2.3.27)

Coating)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 18 of 29 SLRA Table 3.3.2-9 on Page 3.3-201 is revised as follows:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Valve Body Pressure Carbon Steel Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Boundary Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 19 of 29 SLRA Table 3.3.2-9 on Page 3.3-203 is revised as follows:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Valve Body Pressure Ductile Iron Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Boundary Underground Piping and Tanks (B.2.3.27)

Valve Body Pressure Gray Cast Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Boundary Iron Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 20 of 29 SLRA Table 3.3.2-9 on Page 3.3-204 is revised as follows:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Valve Body Pressure Carbon Steel Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Boundary Underground Piping and Tanks (B.2.3.27)

Valve Body Pressure Ductile Iron Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Boundary Underground Piping and Tanks (B.2.3.27)

Valve Body Pressure Gray Cast Soil (External) Cracking Buried and VII.I.A425 3.3.1144 AB Boundary Iron Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 21 of 29 SLRA Table 3.3.2-16 on Page 3.3-293 is revised as follows:

Table 3.3.2-16: Service and Seal Water - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Carbon Steel Soil (External) Loss of Material Buried and VII.I.AP198 3.3.1109 AB Components Boundary Underground Piping and Tanks (B.2.3.27)

SLRA Table 3.3.2-16 on Page 3.3-298 is revised as follows:

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

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 22 of 29 SLRA Table 3.3.2-18 on Page 3.3-311 is revised as follows:

Table 3.3.2-18: Wells and Domestic Water - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Carbon Steel Concrete Loss of Material Buried and VII.I.AP198 3.3.1109 AB Components Boundary (External) Underground Piping and Tanks (B.2.3.27)

SLRA Table 3.3.2-18 on Page 3.3-314 is revised as follows:

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

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 23 of 29 SLRA Table 3.4-1 on Page 3.4-27 is revised as follows:

Table 3.4-1: Summary of Aging Management Evaluations for the Steam and Power Conversion Systems Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program / TLAA Recommended 3.4.1047 Stainless steel Loss of material AMP XI.M41, "Buried and No Consistent with NUREG2191 with piping, piping due to pitting, Underground Piping and exception for the Buried and components, tanks, crevice corrosion, Tanks" Underground Piping and Tanks closure bolting MIC (soil only) (B.2.3.27) AMP.

exposed to soil, concrete The Buried and Underground Piping and Tanks (B.2.3.27) AMP is used to manage loss of material of stainless steel piping and piping components exposed to soil.

3.4.1050 Steel piping, piping Loss of material AMP XI.M41, "Buried and No Consistent with NUREG2191 with components, tanks, due to general, Underground Piping and exception for the Buried and closure bolting pitting, crevice Tanks" Underground Piping and Tanks exposed to soil, corrosion, MIC (soil (B.2.3.27) AMP.

concrete, only) underground The Buried and Underground Piping and Tanks (B.2.3.27) AMP is used to manage loss of material of steel piping and piping components exposed to soil.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 24 of 29 SLRA Table 3.4-1 on Page 3.4-31 is revised as follows:

Table 3.4-1: Summary of Aging Management Evaluations for the Steam and Power Conversion Systems Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program / TLAA Recommended 3.4.1072 Stainless steel, steel, Cracking due to AMP XI.M41, "Buried and No Consistent with NUREG2191 with aluminum piping, SCC (steel in Underground Piping and exception for the Buried and piping components, carbonate/ Tanks" Underground Piping and Tanks tanks exposed to bicarbonate (B.2.3.27) AMP.

soil, concrete environment only)

The Buried and Underground pPiping and Tanks (B.2.3.27) AMP is used to manage cracking in carbon steel and stainless steel piping and piping components exposed to soil.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 25 of 29 SLRA Table 3.4.2-1 on Page 3.4-48 is revised as follows:

Table 3.4.2-1: Condensate Storage - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Stainless Soil (External) Cracking Buried and VIII.H.S420 3.4.1072 AB Components Boundary Steel [Pipe segment Underground Piping C1120HK] and Tanks (B.2.3.27)

Piping, Piping Pressure Stainless Soil (External) Loss of Material Buried and VIII.H.SP145 3.4.1047 AB Components Boundary Steel [Pipe segment Underground Piping C1120HK] and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 26 of 29 SLRA Table 3.4.2-5 on Page 3.4-95 is revised as follows:

Table 3.4.2-5: Off-Gas - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Holdup and Carbon Steel Soil (External) Loss of Material Buried and VIII.H.SP161 3.4.1050 AB Components Plateout Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 06b Page 27 of 29 SLRA Table 3.4.2-5 on Page 3.4-97 is revised as follows:

Table 3.4.2-5: Off-Gas - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Holdup and Carbon Steel Soil (External) Cracking Buried and VIII.H.S420 3.4.1072 AB Components Plateout Underground Piping and Tanks (B.2.3.27)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6b Page 28 of 29 SLRA Section B.1.1 on Page B-6 is revised as follows:

The following programs each have exception(s) justified with a technical basis:

MNGP Water Chemistry AMP (B.2.3.2)

MNGP Reactor Head Closure Stud Bolting AMP (B.2.3.3),

MNGP BWR Vessel Internals AMP (B.2.3.7)

MNGP Fire Water System (B.2.3.16)

MNGP Fuel Oil Chemistry AMP (B.2.3.18)

MNGP Buried and Underground Piping and Tanks (B.2.3.27)

MNGP ASME Section XI, Subsection IWE AMP (B.2.3.29)

SLRA Table B-4 on page B-21 is revised as follows:

Buried and Underground B.2.3.27 XI.M41 Yes No Yes Piping and Tanks

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6b Page 29 of 29 Section B.2.3.27 on page B-199 is being revised as follows:

NUREG-2191 Consistency The MNGP Buried and Underground Piping and Tanks AMP, with enhancements, will be consistent without one exception, to the 10 elements of NUREG-2191,Section XI.M41, Buried and Underground Piping and Tank.

Exceptions to NUREG2191 None.

New and replacement backfill is to be consistent with SP0169-2007 Section 5.2.3 or NACE RP0285-2002. The maximum allowable backfill size per ASTM D448-08 (size number 67) is one inch.

Existing backfill for buried components at MNGP was installed per site design specifications. The general requirement for earthwork material is that materials containing brush, roots, peat, sod, or other organic, perishable or deleterious matter, snow, ice, or frozen soil is not used for backfilling. Structural backfill is well graded, sound, dense and durable material. No more than 10 percent by weight shall pass the No.

200 sieve. The maximum size of structural backfill is two inches in confined areas where hand tamping is required and four inches in other areas. This is an exception to what is required by SP0169-2007 Section 5.2.3 or NACE RP0285-2002.

This exception is acceptable because operating experience demonstrates that surveys/tests are capable of detecting and identifying degraded conditions in backfill.

These degraded conditions are entered into the CAP and then addressed. If the backfill does not meet acceptance critieria, the degraded condition is evaluated or repaired.

MNGP has been exposed to severe winter conditions for many years, and has, to date, shown no signs of significant freeze-thaw damage. An enhancement has been added to ensure compliance with NACE SP01692007 Section 5.2.3 or NACE RP02852002 Section 3.6 for new and replacement backfill.

Enclosure 07 Revise SLRA Section 4.2.4 to Cite 'PTLR'

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 1 of 2 Revise SLRA Section 4.2.4 to Cite 'PTLR' For SLRA Section 4.2.4, revise to cite PTLR instead of Technical Specification.

Affected SLRA Sections: 4.2.4, A.3.2.4 SLRA Page Numbers: 4.2-21, A-41 Description of Change:

Sections 4.2.4 and A.3.2.4 are revised to cite PTLR instead of Technical Specification. Updates to the PT limit curves are provided to the NRC in a PTLR change request.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 2 of 2 SLRA Section 4.2.4 on page 4.2-21 is revised as follows:

The PT limit curves will be updated and a Technical Specification PTLR change request will be submitted to the NRC prior to exceeding the current 54 EFPY limit.

SLRA Section A.3.2.4 on page A-41 is revised as follows:

The P-T limit curves will be updated, and a Technical Specification PTLR change request will be submitted for approval prior to exceeding the current 54 EFPY limit.

Enclosure 08 Revise References from 40.40(a) to 50.55a

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8 Page 1 of 3 Revise References from 40.40(a) to 50.55a Revise SLRA Section 4.2.5 and A.3.2.5 to resolve error in reference to 10 CFR 40.40(a)(z)(1)

Affected SLRA Sections: 4.2.5 and A.3.2.5 SLRA Page Numbers: 4.2-22 and A-42 Description of Change:

Sections 4.2.5 page 4.2-22 and A.3.2.5 page A-42 are updated with the correct ASME Code.

The MNGP RPV meets the applicability criteria of BWRVIP-329-A, and therefore is justified for a request for alternative pursuant to 10 CFR 50.55a(z)(1) from the ASME Code.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8 Page 2 of 3 SLRA Section A.3.2.5 on page A-42 is revised as follows:

A.3.2.5 RPV Circumferential Weld Examination Relief MNGP has previously applied for and been granted relief from RPV circumferential weld inspections. The relief from inspection is based on assessment of the probability of failure of the limiting circumferential weld. This assessment is based on 54 EFPY fluence values associated with 60 years of operation and has therefore been identified as a TLAA requiring evaluation for the SPEO.

BWRVIP329A and the associated NRC safety evaluation report (SER) provide technical basis for reduction in inspection of RPV circumferential welds and an assessment of axial weld integrity for extended operations of up to 80 years. BWRVIP329A provides criteria for applicability based on plantspecific data. Evaluation for applicability to MNGP confirms that the RPV dimensions are within the limits of the enveloping RPV dimensions in BWRVIP329A.

Using plantspecific data for the RPV dimensions and limiting ARTs for the RPV plates and welds, the evaluation shows that the MNGP RPV meets the applicability criteria of BWRVIP329A. As such, on the technical basis of BWRVIP329A and as stated in the BWRVIP329A SER, MNGP is justified for request for alternative pursuant to 10 CFR 40.4050.55a(a)(z)(1) from the ASME Code,Section XI examinations for RPV circumferential weld for up to 80 years of plant operation. These analyses will be managed in accordance with 10 CFR 54.21(c)(1)(iii) by requesting relief from circumferential weld inspection using the 10 CFR 50.55a process.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8 Page 3 of 3 SLRA Section 4.2.5 on page 4.2-22 is revised as follows:

TLAA Evaluation Plantspecific RPV dimensions and material chemistry for MNGP were evaluated for the applicability criteria in BWRVIP329A. This confirmed that the MNGP RPV dimensions are within the limits of the enveloping RPV dimensions in BWRVIP329A.

The limiting maximum reference temperatures (RTMAX) for the RPV surface (0T) and 72 EFPY was calculated using plantspecific material chemistry (copper content, nickel content, chemistry factor, and RTNDT(U) (referred to as initial RTNDT)) and neutron fluence for the MNGP RPV plates and welds. The endofinterval (EOI) for MNGP is defined as 80 years, which is equivalent to the 72 EFPY for the neutron fluence. The 0T values were calculated for the fluence at the RPV inner surface. The EOI RTMAX values for all MNGP RPV plates and welds meet the acceptability criteria for limiting plate, circumferential weld, and axial weld in BWRVIP329A.

Using plantspecific data for the RPV dimensions and limiting ARTs for the RPV plates and welds, the evaluation shows that the MNGP RPV meets the applicability criteria of BWRVIP329A. As such, on the technical basis of BWRVIP329A and as stated in the BWRVIP329A SER, MNGP is justified for request for alternative pursuant to 10 CFR 40.4050.55a(a)(z)(1) from the ASME Code,Section XI examinations for RPV circumferential weld for up to 80 years of plant operation.

Enclosure 09 Revise Discussion of Item Numbers 3.2.1-107 and 3.2.1-108 to State Not Applicable

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9 Page 1 of 4 Revise Discussion of Item Numbers 3.2.1-107 and 3.2.1-108 to State Not Applicable Revised SLRA Table 3.2-1 to state in the discussion of Item Numbers 3.2.1-107 and 3.2.1-108 are not applicable.

Affected SLRA Sections: Section 3.2.2.2.2, Section 3.2.2.2.4, Table 3.2-1 SLRA Page Numbers: 3.2-9, 3.2-10, 3.2-12, 3.2-39, and 3.2-40 Description of Change:

Item Numbers 3.2.1-107 and 3.2.1-108 from Table 3.2-1 revised to state in discussion that they are not applicable. Both items were incorrectly stated to be consistent with NUREG-2191.

SLRA supports this as Items 3.2.1-107 and 3.2.1-108 are not used in any Summary of Aging Management Evaluation tables.

SLRA Sections 3.2.2.2.2 and 3.2.2.2.4 are revised to reflect the changes made in discussion of Item Numbers 3.2.1-107 and 3.2.1-108 from Table 3.2-1.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9 Page 2 of 4 SLRA Table 3.2-1, Items 3.2.1-107 and 3.2.1-108 on pages 3.2-39 and 3.2-40 respectively are revised as follows:

Table 3.21: Summary of Aging Management Evaluations for the Engineered Safety Features Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program / TLAA Recommended 3.2.1107 Insulated stainless Loss of material AMP XI.M29, "Outdoor and Yes (SRPSLR Consistent with NUREG2191. Not steel, nickel alloy due to pitting, Large Atmospheric Metallic Section 3.2.2.2.2) applicable.

piping, piping crevice corrosion Storage Tanks," AMP components, tanks XI.M32, "OneTime There are no insulated stainless steel or exposed to air, Inspection," AMP XI.M36, nickel alloy piping, piping components, or condensation "External Surfaces tanks exposed to air or condensation in Monitoring of Mechanical the ESF systems.

Components," or AMP The OneTime Inspection (B.2.3.20) AMP is XI.M42, "Internal used to manage loss of material of insulated Coatings/Linings for stainless steel piping and piping components InScope Piping, Piping exposed to air.

Components, Heat Exchangers, and Tanks" Further evaluation is documented in Section 3.2.2.2.2.

3.2.1108 Insulated stainless Cracking due to AMP XI.M29, "Outdoor and Yes (SRPSLR Consistent with NUREG2191.Not steel piping, piping SCC Large Atmospheric Metallic Section 3.2.2.2.4) applicable.

components, tanks Storage Tanks," AMP exposed to air, XI.M32, "OneTime There are no insulated stainless steel condensation Inspection," AMP XI.M36, piping, piping components, or tanks "External Surfaces exposed to air or condensation in the ESF Monitoring of Mechanical systems.

Components," or AMP The OneTime Inspection (B.2.3.20) AMP is XI.M42, "Internal used to manage cracking of insulated Coatings/Linings for stainless steel piping and piping components InScope Piping, Piping exposed to air.

Components, Heat Exchangers, and Tanks" Further evaluation is documented in Section 3.2.2.2.4.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9 Page 3 of 4 SLRA Section 3.2.2.2.2 on pages 3.2-9 and 3.2-10 is revised as follows:

Ambient air at MNGP is not subject to a marine atmosphere. MNGP is located in the vicinity of a major road that is routinely salted for snow and ice. A review of the over 69,000 ARs created during the 01/01/2010 to 07/29/2021 period was performed to determine if the proximity to the salted road has resulted in any plantspecific OE for loss of material of the susceptible materials to chlorides in an air environment. The results of this review show that the ambient air environments do not contain sufficient halides (e.g., chlorides) in the presence of moisture to result in loss of material. As such, stainless steel and nickel alloy components exposed to air indoor uncontrolled, air outdoor, or condensation in the ESF are not susceptible to loss of material. There are no insulated stainless steel or nickel alloy piping, piping components, or tanks exposed to air or condensation in the ESF systems.

Consistent with the recommendation of GALLSLR, the OneTime Inspection AMP will confirm that loss of material is not occurring in stainless steel or nickel alloy components exposed to air indoor uncontrolled, air outdoor, or condensation, and, in insulated stainless steel components exposed to condensation. Deficiencies will be documented in accordance with the sites 10 CFR Part 50, Appendix B, Section XVI, CAP. The OneTime Inspection AMP is described in Section B.2.3.20.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9 Page 4 of 4 SLRA Section 3.2.2.2.4 on pages 3.2-12 is revised as follows:

Ambient air at MNGP is not subject to a marine atmosphere. MNGP is located in the vicinity of a major road that is routinely salted for snow and ice. A review of the over 69,000 ARs created during the 01/01/2010 to 07/29/2021 period was performed to determine if the proximity to the salted road has resulted in any plantspecific OE for cracking of the susceptible materials to chlorides in an air environment. The results of this review show that the ambient air environments do not contain sufficient halides (e.g., chlorides) in the presence of moisture to result in SCC. As such, stainless steel components exposed to air indoor uncontrolled, air outdoor, or condensation in the ESF are not susceptible to cracking due to SCC. There are no insulated stainless steel piping, piping components, or tanks exposed to air or condensation in the ESF systems.

Consistent with the recommendation of GALLSLR, the OneTime Inspection AMP will confirm that cracking is not occurring in stainless steel components exposed to air indoor uncontrolled, air outdoor, or condensation, and, in insulated stainless steel components exposed to condensation. Deficiencies will be documented in accordance with the sites 10 CFR Part 50, Appendix B, Section XVI, CAP. The OneTime Inspection AMP is described in Section B.2.3.20.

Enclosure 10 Applicability of SCC and LOM Aging Mechanisms for Stainless Steel in the O-Gas Condensate System

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0 Page 1 of 4 Applicability of SCC and LOM Aging Mechanisms for Stainless Steel in the Off-Gas Condensate System Update to include applicability of SCC and LOM for stainless steel components in a condensate environment Affected SLRA Sections: Table 3.4.2-5 SLRA Page Numbers: 3.4-91, 3.4-96, 3.4-100 Description of Change:

SLRA Table 3.4.2-5 (Off-Gas - Summary of Aging Management Evaluation) included six table entries with None as both the aging effect requiring management and the aging management program for Off-Gas system stainless steel components exposed internally to a condensate environment. The row items in the Table are corrected to include Cracking and Loss of Material as the aging effects and that the One-Time Inspection (B.2.3.20) program will manage aging.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 10 Page 2 of 4 SLRA Table 3.4.2-5 on pages 3.4-91 is revised as follows:

Table 3.4.2-5: Off-Gas - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG- Table 1 Component Type Material Environment Requiring Notes Function Program 2191 Item Management Item Heat Exchanger - Holdup and Stainless Steel Condensation None Loss of None One-Time VIII.E.SP-127a 3.4.1-003 I, 1 A (H2O2 Sample Plateout (Internal) Material Inspection (B.2.3.20)

Coolers) Tubes Heat Exchanger - Holdup and Stainless Steel Condensation None Cracking None One-Time VIII.E.SP-118a 3.4.1-002 I, 1 A (H2O2 Sample Plateout (Internal) Inspection (B.2.3.20)

Coolers) Tubes

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 10 Page 3 of 4 SLRA Table 3.4.2-5 on pages 3.4-96 is revised as follows:

Table 3.4.2-5: Off-Gas - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG- Table 1 Component Type Material Environment Requiring Notes Function Program 2191 Item Management Item Piping, Piping Holdup and Stainless Steel Condensation None Loss of None One-Time VIII.E.SP-127a 3.4.1-003 I, 1 A Components Plateout (Internal) Material Inspection (B.2.3.20)

Piping, Piping Holdup and Stainless Steel Condensation None Cracking None One-Time VIII.E.SP-118a 3.4.1-002 I, 1 A Components Plateout (Internal) Inspection (B.2.3.20)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 10 Page 4 of 4 SLRA Table 3.4.2-5 on pages 3.4-100 is revised as follows:

Table 3.4.2-5: Off-Gas - Summary of Aging Management Evaluation Aging Effect Aging Intended NUREG-2191 Table 1 Component Type Material Environment Requiring Management Notes Function Item Item Management Program Valve Body Holdup and Stainless Steel Condensation None Cracking None One-Time VIII.E.SP-118a 3.4.1-002 I, 1 A Plateout (Internal) Inspection (B.2.3.20)

Valve Body Holdup and Stainless Steel Condensation None Loss of None One-Time VIII.E.SP-127a 3.4.1-003 I, 1 A Plateout (Internal) Material Inspection (B.2.3.20)

Enclosure 11 Separation of Buried Piping Materials Subject to Selective Leaching

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 1 of 5 Separation of Buried Piping Materials Subject to Selective Leaching Update the material column of SLRA Table 3.3.2-9 to differentiate ductile iron from gray cast iron.

Affected SLRA Sections: Section 3.3.2.1.9, Table 3.3.2-9 SLRA Page Numbers: 3.3-11, 3.3-190, 3.3-205 Description of Change:

Add Ductile Iron and Ductile Iron (with Internal Coating) as materials of piping and piping components in the Fire System AMR Table 3.3.2-9. Update Plant-Specific Notes 2, 3, and 7 to include ductile iron and ductile iron (with internal coating) as appropriate. Additionally, update Section 3.3.2.1.9 to include Ductile Iron (with Internal Coating) as a material.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 2 of 5 SLRA Section 3.3.2.1.9, on Page 3.3-11, is revised to add the following:

3.3.2.1.9 Fire System Materials The materials of construction for the FIR System components are:

Carbon Steel Carbon and Low Alloy Steel Bolting Copper Alloy with Greater Than 15% Zinc Copper Alloy with 15% Zinc or Less Ductile Iron Ductile Iron with Internal Coating Elastomer Galvanized Steel Glass Gray Cast Iron Gray Cast Iron with Internal Coating Polymer PVC Stainless Steel Stainless Steel Bolting

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 11 Page 3 of 5 SLRA Table 3.3.2-9, on Page 3.3-190, is revised to insert the following:

Table 3.3.2-9: Fire System - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Piping, Piping Pressure Ductile Iron Air - Indoor Loss of Fire Water System VII.G.A-412 3.3.1-136 D Components Boundary Uncontrolled Material (B.2.3.16)

(External)

Piping, Piping Pressure Ductile Iron Raw Water Flow Blockage Fire Water System VII.G.A-33 3.3.1-064 B Components Boundary (Internal) (B.2.3.16)

Piping, Piping Pressure Ductile Iron Raw Water Long-Term One-Time Inspection VII.G.A-532 3.3.1-193 A Components Boundary (Internal) Loss of (B.2.3.20)

Material Piping, Piping Pressure Ductile Iron Raw Water Loss of Fire Water System VIII.G.A-33 3.3.1-064 B Components Boundary (Internal) Material (B.2.3.16)

Piping, Piping Pressure Ductile Iron Raw Water Loss of Selective Leaching VII.G.A-51 3.3.1-072 A Components Boundary (Internal) Material (B.2.3.21)

Piping, Piping Pressure Ductile Iron Raw Water Wall Thinning Fire Water System VII.C1.A-409 3.3.1-126 E, 7 Components Boundary (Internal) (B.2.3.16)

Piping, Piping Pressure Ductile Iron Soil (External) Cracking Buried and VII.I.A425 3.3.1144 B Components Boundary Underground Piping and Tanks (B.2.3.27)

Piping, Piping Pressure Ductile Iron Soil (External) Loss of Buried and VII.I.AP198 3.3.1109 B Components Boundary Material Underground Piping and Tanks (B.2.3.27)

Piping, Piping Pressure Ductile Iron Soil (External) Loss of Selective Leaching VII.G.A02 3.3.1072 A Components Boundary Material (B.2.3.21)

Piping, Piping Pressure Ductile Iron Raw Water Cracking Internal VII.G.A416 3.3.1138 A Components Boundary (with Internal (Internal) Coatings/Linings for Coating) InScope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.28)

Piping, Piping Pressure Ductile Iron Raw Water Flow Blockage Fire Water System VII.G.A647 3.3.1195 B Components Boundary (with Internal (Internal) (B.2.3.16)

Coating)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 11 Page 4 of 5 Piping, Piping Pressure Ductile Iron Raw Water Loss of Internal VII.G.A416 3.3.1138 A Components Boundary (with Internal (Internal) Coating or Coatings/Linings for Coating) Lining Integrity InScope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.28)

Piping, Piping Pressure Ductile Iron Raw Water Loss of Fire Water System VII.G.A414 3.3.1139 E, 2 Components Boundary (with Internal (Internal) Material (B.2.3.16)

Coating)

Piping, Piping Pressure Ductile Iron Raw Water Loss of Internal VII.G.A416 3.3.1138 A Components Boundary (with Internal (Internal) Material Coatings/Linings for Coating) InScope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.28)

Piping, Piping Pressure Ductile Iron Raw Water Loss of Selective Leaching VII.G.A415 3.3.1140 E, 3 Components Boundary (with Internal (Internal) Material (B.2.3.21)

Coating)

Piping, Piping Pressure Ductile Iron Soil (External) Cracking Buried and VII.I.A425 3.3.1144 B Components Boundary (with Internal Underground Piping Coating) and Tanks (B.2.3.27)

Piping, Piping Pressure Ductile Iron Soil (External) Loss of Buried and VII.I.AP198 3.3.1109 B Components Boundary (with Internal Material Underground Piping Coating) and Tanks (B.2.3.27)

Piping, Piping Pressure Ductile Iron Soil (External) Loss of Selective Leaching VII.G.A02 3.3.1072 A Components Boundary (with Internal Material (B.2.3.21)

Coating)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 5 of 5 SLRA Table 3.3.2-9 on Page 3.3-205 is revised as follows:

PlantSpecific Notes

1. The Fire Water System (B.2.3.16) program is being substituted for the OpenCycle Cooling Water System (B.2.3.11) program to manage cracking in copper alloy with greater than 15% zinc piping with an internal environment of raw water.

2. The Fire Water System (B.2.3.16) program is being substituted for the Internal Coatings/Linings for InScope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.28) program to manage loss of material of the base metal due to general, pitting and crevice corrosion and MIC in the cement lined gray cast iron and ductile iron (with internal coating) fire water piping with an internal environment of raw water.

3. The Selective Leaching (B.2.3.21) program is being substituted for the Internal Coatings/Linings for InScope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.28) program to manage loss of material of the base metal due to selective leaching in the cement lined gray cast iron and ductile iron (with internal coating) fire water piping with an internal environment of raw water.

4. The Fire Protection (B.2.3.15) program will be used to manage flow blockage in stainless steel halon system spray nozzles with an internal environment of condensation.

5. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Component (B.2.3.24) program is being substituted for the Closed Treated Water Systems (B.2.3.12) program to manage the reduction of heat transfer due to fouling in copper alloy with greater than 15% zinc heat exchanger tubes with an internal environment of closed cycle cooling water.

6. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Component (B.2.3.24) program is being substituted for the Closed Treated Water Systems (B.2.3.12) program to manage loss of material and cracking in gray cast iron and copper alloy with greater than 15%

zinc heat exchanger components with internal and external environments of closed cycle cooling water.

7. The Fire Water System (B.2.3.16) program is being substituted for the FlowAccelerated Corrosion (B.2.3.9) program to manage wall thinning due to erosion in gray cast iron and ductile iron piping and piping components exposed to raw water.

Enclosure 12 Revise Fluence Values to 3.68E21

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 1 of 3 Revise Fluence Values to 3.68E21 Revise SLRA Sections 4.2.8 and A.3.2.8 to cite the correct fluence values of 3.68E21.

Affected SLRA Sections: 4.2.8 and A.3.2.8 SLRA Page Numbers: 4.2-28 and A-44 Description of Change:

Section 4.2.8 and Section A.3.2.8 are revised to cite the correct fluence value of 3.68E21. The fluence value was incorrectly cited as 5.68E21.The fluence value of 3.68E21 is supported by SLRA Table 4.2.1.2-1 on page 4.2-10.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 2 of 3 SLRA Section 4.2.8 on page 4.2-28 is revised as follows:

The fluence for the most irradiated point on the core shroud was calculated to be 5.68 x 1021 n/cm23.68 x 1021 n/cm2 (E >1 MeV) for 80 years. This can be compared to the test data for control blade handles at 8 x 1021 n/cm2 (E >1 MeV) described in BWRVIP66. The lowest measured value of percent elongation for stainless steel weld metal is 4 percent for a temperature of 297 C (567 F) with a neutron fluence of 8 x 1021 n/cm2 (E >1 MeV), while the average value of base metal at 290 C (554 F) is 20 percent (Reference 4.7.21).

Since the most irradiated point on the core shroud for 80 years of operation is calculated to be 5.68 x 1021 n/cm23.68 x 1021 n/cm2 (E >1 MeV), below the 8 x 1021 n/cm2 fluence threshold for which elongation test data are available, the measured value of elongation bounds the calculated thermal shock strain amplitude of 0.57 percent. The calculated thermal shock strain at the most irradiated location is acceptable considering the loss of ductility effects for an 80year operating period.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 3 of 3 SLRA Section A.3.2.8 on page A-44 is revised as follows:

The fluence for the most irradiated point on the core shroud was calculated to be 5.68 x 1021 n/cm23.68 x 1021 n/cm2 (E >1 MeV) for 80 years. This can be compared to the test data for control blade handles at 8 x 1021 n/cm2 (E >1 MeV) described in BWRVIP66. The lowest measured value of percent elongation for stainless steel weld metal is 4 percent for a temperature of 297 C (567 F) with a neutron fluence of 8 x 1021 n/cm2 (E >1 MeV), while the average value of base metal at 290 C (554 F) is 20 percent.

Since the most irradiated point on the core shroud for 80 years of operation is calculated to be 5.68 x 1021 n/cm23.68 x 1021 n/cm2 (E >1 MeV), below the 8 x 1021 n/cm2 fluence threshold for which elongation test data are available, the measured value of elongation bounds the calculated thermal shock strain amplitude of 0.57 percent. The calculated thermal shock strain at the most irradiated location is acceptable considering the loss of ductility effects for an 80year operating period. Therefore, this analysis has been projected through the SPEO in accordance with 10 CFR 54.21(c)(1)(ii).

Enclosure 13 Updated Reference to Ranganath Analysis

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3 Page 1 of 3 Updated Reference to Ranganath Analysis Update SLRA Sections 4.2.7 and A.3.2.7 to Provide the Correct Reference Affected SLRA Sections: 4.2.7, A.3.2.7 SLRA Page Numbers: 4.2-26, A-43 Description of Change:

SLRA Sections 4.2.7 and A.3.2.7 are updated to reference the Ranganath analysis (Reference 4.7.18).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3 Page 2 of 3 SLRA Section 4.2.7 on Page 4.2-26 is revised as follows:

The critical location for the fracture mechanics analysis is at 1/4T. The peak stress intensity factor, K, at 1/4T has a value of approximately 100 ksiin. A maximum KI of 105 ksiin was utilized per Section XI IWB-3612 Reference 4.7.18. The acceptability of this K on a plant-specific basis for MNGP can be determined by considering a revised allowable fracture toughness applicable to the MNGP vessel for 72 EFPY. Based on a 0T ART of 197.8°F, the fracture toughness KIC of 174.4°F is above the upper shelf value of 200 ksiin.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3 Page 3 of 3 SLRA Section A.3.2.7 on Page A-43 is revised as follows:

Further, the thermal stress evaluation determined the peak stress intensity factor, K, at 1/4T has a value of approximately 100 ksiin. A maximum KI of 105 ksiin was utilized per Section XI IWB3612 Reference 4.7.18. The acceptability of this K on a plantspecific basis for MNGP can be determined by considering a revised allowable fracture toughness applicable to the MNGP vessel for 72 EFPY. Based on a 0T ART of 197.8°F, the fracture toughness KIC of 174.4°F is above the upper shelf value of 200 ksiin.

Enclosure 14 Revise the TLAA Disposition Title to Re"ect 10 CFR 54.21(c)(1)(iii)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4 Page 1 of 2 Revise the TLAA Disposition Title to Reflect 10 CFR 54.21(c)(1)(iii)

For SLRA Section 4.3.2, revise the TLAA Disposition title to reflect 10 CFR 54.21(c)(1)(iii).

Affected SLRA Sections: 4.3.2 SLRA Page Numbers: 4.3-6 Description of Change:

The TLAA disposition title of SLRA Section 4.3.2 is incorrect. The TLAA was dispositioned as being managed through SPEO in accordance with 10 CFR54(c)(1)(iii) not 10 CFR54(c)(1)(ii).

Revised the title to reflect 10 CFR54(c)(1)(iii). SLRA Table 4.1-2 has the correct disposition of 10 CFR 54.21(c)(1)(iii).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4 Page 2 of 2 SLRA Section 4.3.2 on page 4.3-6 is revised as follows:

TLAA Disposition: 10 CFR 54.21(c)(1)(ii)(iii)

The ASME Code,Section III, Class 1 component fatigue waivers will be managed by the Fatigue Monitoring (B.2.2.1) AMP through the SPEO in accordance with 10 CFR 54.21(c)(1)(iii). The Fatigue Monitoring (B.2.2.1) AMP will monitor the transient cycles which are the inputs to the fatigue waiver reevaluations and require action prior to exceeding design limits that would invalidate their conclusions.

Enclosure 15 Inclusion of Discussion of Impact of Deposits on Downstream Components

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5 Page 1 of 3 Inclusion of Discussion of Impact of Deposits on Downstream Components Inclusion of discussion of impact of deposits on downstream components Affected SLRA Sections: A.4, Table A-3, B.2.3.16 SLRA Page Numbers: A-67, B-118 Description of Change:

The Fire Water System AMP basis document includes discussion on impact of identified deposits on downstream components in the enhancement to the Monitoring and Trending program element, whereas Commitment 19 in Table A-3 and Section B.2.3.16 in the SLRA do not include this discussion. These SLRA sections are revised to include this discussion.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5 Page 2 of 3 SLRA Table A-3, page A-67 is revised as follows:

Table A3 List of SLR Commitments and Implementation Schedule No. Aging NUREG2191 Commitment Implementation Schedule Management Section Program or Activity (Section) 19 Fire Water System XI.M27 e) Update spray and sprinkler system flushing procedures to No later than 6 months prior (A.2.2.16) document and trend deposits (scale or foreign to the SPEO, or no later than material). Incorporate acceptance criteria that no loose the last refueling outage prior fouling products can exist in the systems that could cause to the SPEO flow blockage in the sprinklers or deluge nozzles.

Implement the AMP and start Include steps in flushing procedures to compare the amount the preSPEO inspections of deposits to the previous inspections results, and if the and tests no earlier than 5 trend shows increasing deposits, then the CAP will be years prior to the SPEO.

utilized to drive improvement. Additionally, identified deposits will be evaluated for potential impact on downstream components, such as sprinkler heads or spray nozzles. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g.,

attached to the completed work order). However, if a trend is negative, a CAP item is initiated to evaluate the trend and determine any follow-up corrective actions.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5 Page 3 of 3 The enhancements table in SLRA Section B.2.3.16, page B-118 is revised as follows:

Element Affected Enhancement

5. Monitoring and Trending Update spray and sprinkler system flushing procedures to enable trending of data. Specifically, the existing flushing procedures and preventive maintenance activities will be revised to document and trend deposits (scale or foreign material).

Incorporate acceptance criteria that no loose fouling products can exist in the systems that could cause flow blockage in the sprinklers or deluge nozzles.

Existing flushing procedures, as well as new flushing procedures, will include steps to compare the amount of deposits to the previous inspections results, and if the trend shows increasing deposits, then the MNGP CAP will be utilized to drive improvement. Additionally, identified deposits will be evaluated for potential impact on downstream components, such as sprinkler heads or spray nozzles. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g., attached to the completed work order).

However, if a trend is negative, a CAP item is initiated to evaluate t the trend and determine any follow-up corrective actions.

Enclosure 16 RPV Fatigue TLAA Inconsistency Between Original and Rerate Temperature Changes

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6 Page 1 of 2 RPV Fatigue TLAA Inconsistency Between Original and Rerate Temperature Changes Original versus rerate temperature changes in the TLAA.

Affected SLRA Sections: 4.3.5 SLRA Page Numbers: 4.3-14 Description of Change:

MNGP SLRA Section 4.3.5 will be clarified to include the rerate scaling thermal transient temperature range and to list the original and rerate temperature ranges.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6 Page 2 of 2 SLRA Section 4.3.5 on page 4.3-14 is revised to add the following information:

Core Spray Line Rerate/EPU scaling is calculated for the load set pairs used in the analysis. These are based on a thermal transient and OBE. The original thermal transient from 546°F to 80°F is the largest temperature delta evaluated and is used as the basis for EPU scaling. The EPU thermal transient from 549°F to 80°F is also used as a basis for scaling factor and is bounding. The calculation of EPU scaling factors is based on original and rerate conditions, which bound EPU conditions. The pressure scaling factor is bounding.

RHR Intertie Line Rerate/EPU scaling is calculated for the load set pairs used in the analysis. The only evaluated thermal transient affected by EPU is the first thermal change during shutdown when flow is initiated in the RHR intertie line and temperature goes from 150°F to 546°F. The second temperature change, from 546°F to 375°F, is a slow ramp and is not used in the analysis. The last thermal transients represent a 375°F to 50°F step and a 50°F to 300°F step. The calculation of EPU scaling factors is based on the original (150°F to 546°F) and rerate (150°F to 549°F) conditions, which bound EPU conditions. The pressure scaling factor is bounding.

Enclosure 17 Submit HELB as a TLAA in Section 4.3.6

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 1 of 8 Submit HELB as a TLAA in Section 4.3.6 Submit HELB analysis as a TLAA in Section 4.3.6 of the SLRA.

Affected SLRA Sections: Table of Contents, Tables 4.1-1, 4.1-2, and A-2; Sections 4.3, 4.3.6, A.3.3, and A.3.3.6.

SLRA Page Numbers: viii, 4.1-3, 4.1-4, 4.1-7, A-8 & A-9; 4.3-1, 4.3-15 to 4.3-17, A-45, and A-47 & A-48 Description of Change:

Per NUREG-1865 and the original MNGP LRA, HELB was not considered a TLAA for MNGP operation to sixty years. However, USAR Appendix I,Section I.3.1.1 defines the selection of break locations for Class I high energy lines outside containment, which in turn references guidance taken from MEB 3-1, B.1.c(2)(3) contained in GL 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements. Furthermore, USAR Appendix I,Section I.3.1.2 defines the selection of break locations for Class II high energy lines outside containment as including terminal ends and each intermediate location of potential high stress or fatigue such as pipe fittings, valves, flanges and welded on attachments .

The maximum allowable stress criterion (SA) is used to determine break locations. This in turn is based on time-dependent fatigue design cycles; as such HELB analyses are considered to be TLAAs for SLR. The HELB analyses are generally described and dispositioned for the SPEO.

There is no impact to the Fatigue Monitoring AMP in discussions in Appendix A or B based on a disposition of 10 CFR 54.21(c)(1)(i).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 2 of 8 The SLRA Table of Contents entry for Section 4.3.6 on page viii is revised as follows:

4.3.6 ASME Section III, Class 2 and 3 and ANSI B31.1 & High Energy ............. 4.3-15 Line Break Analyses The High Energy Line Break Analyses Example TLAA in SLRA Table 4.1-1 on page 4.1-3 is revised as follows:

Table 4.11 Review of Generic TLAAs Listed in NUREG2192, Tables 4.12 and 4.71 NUREG2192 Example TLAA Applies to MNGP? SLRA Section NUREG2192, Table 4.12 - Generic TLAAs High Energy Line Break Analyses No Yes(1) N/A4.3.6 Note (1) of SLRA Table 4.1-1 on page 4.1-4 is revised as follows:

Table 4.11 Review of Generic TLAAs Listed in NUREG2192, Tables 4.12 and 4.71 NUREG2192 Example TLAA Applies to MNGP? SLRA Section NUREG2192, Table 4.12 - Generic TLAAs Notes:

(1) High energy line break based on fatigue cumulative usage factor is not considered a TLAA for MNGP SLR, and as such is not included with metal fatigue. Break locations were not postulated based on fatigue criteria (i.e., CUF) stress criteria for high-energy lines outside containment.

The Metal Fatigue TLAA item in SLRA Table 4.1-2 on page 4.1-7 is revised as follows:

Table 4.12 Summary of Results - MNGP TimeLimited Aging Analysis SLRA TLAA Description Disposition Section Metal Fatigue 4.3 80Year Transient Cycle Projections N/A 4.3.1

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 3 of 8 ASME Section III, Class 1 Fatigue Waivers 10 CFR 54.21(c)(1)(iii) 4.3.2 RPV Fatigue Analyses 10 CFR 54.21(c)(1)(iii) 4.3.3 Fatigue Analysis of RPV Internals 10 CFR 54.21(c)(1)(i) 4.3.4 ASME Section III, Class 1 Fatigue Analysis 10 CFR 54.21(c)(1)(iii) 4.3.5 ASME Section III, Class 2 and 3 and ANSI B31.1 10 CFR 54.21(c)(1)(i) 4.3.6

& High Energy Line Break Analyses EnvironmentallyAssisted Fatigue 10 CFR 54.21(c)(1)(iii) 4.3.7 SLRA Section 4.3 on page 4.3-1 is revised as follows:

4.3 METAL FATIGUE Fatigue analyses are required for components designed to ASME Code,Section III, Class 1. Also, certain other codes such as ASME Code,Section III, Class 2 and 3, American National Standards Institute (ANSI) B31.1, Power Piping, and ASME Section VIII, Rules for Construction of Pressure Vessels, Fatigue analyses are required for components designed to ASME Code,Section III, Class 1. Also, certain other codes such as ASME Code,Section III, Class 2 and 3, ANSI B31.1, Power Piping, and ASME Section VIII, Rules for Construction of Pressure Vessels, Division 2, may require a fatigue analysis or assume a stated number of fullrange thermal and displacement transient cycles. NUREG2192 also provides examples of components likely to have fatigue TLAA within the CLB that would require evaluation for the SPEO. Searches were performed to identify these and any other potential fatigue TLAAs within the current licensing bases for MNGP. Each of the potential TLAAs were evaluated against the six elements of the TLAA definition specified in 10 CFR 54.3. Those that were identified as fatigue TLAAs are evaluated using 80year transient cycle and cumulative usage projections, summarized in the following subsections:

80Year Transient Cycle Projections (Section 4.3.1)

ASME Section III, Class 1 Fatigue Waivers (Section 4.3.2)

RPV Fatigue Analyses (Section 4.3.3)

Fatigue Analysis of RPV Internals (Section 4.3.4)

ASME Section III, Class 1 Fatigue Analysis (Section 4.3.5)

ASME Section III, Class 2 and 3 and ANSI B31.1 & High Energy Line Break Analyses (Section 4.3.6)

EnvironmentallyAssisted Fatigue (Section 4.3.7)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 4 of 8 SLRA Section 4.3.6 on page 4.3-15 to 4.3-17 is revised as follows:

4.3.6 ASME Section III, Class 2 and 3 and ANSI B31.1 & High Energy Line Break Analyses TLAA Description A metal component may progressively degrade and lose its structural integrity when it is subjected to fluctuating loads, even at magnitudes less than the design static loads, due to metal fatigue. This mechanism of degradation can occur in flaw free components by developing cracks during service. Implicit fatiguebased maximum allowable stress calculations are performed for piping components designed to USAS / ANSI B31.1 requirements. ASME Section III Code Class 2 and 3 components are designed to requirements that are similar to the guidance in ANSI B31.1.

In addition, process piping that is subject to significant thermal expansion and contraction includes those that penetrate the drywell shell. Typically, these penetrations, which were designed to the ASME Code,Section III, Class B requirements, are a triple flued head design which has a guard pipe between the process piping and the penetration nozzle. The penetration assembly which provides the interface between the exterior of the process piping with the containment liner is typically known as a bellows. This permits the penetration to be vented to the drywell should a rupture of the hot line occur within the penetration. These containment penetration process bellows have been designed for a maximum of 7,000 operating cycles.

Although the code of construction for MNGP did not invoke fatigue analyses, a stress range reduction factor which is applied to the allowable stress range for expansion stresses (SA) is required to account for cyclic thermal conditions. The allowable secondary stress range is 1.0 SA for 7,000 equivalent full temperature thermal cycles or less and is incrementally reduced to 0.5 SA for greater than 100,000 cycles.

In addition, USAR Appendix I, Subsection I.3 indicates that postulated break locations for each high energy line outside containment were evaluated for the Main Steam, HPCI, RCIC, Feedwater, Condensate, RWCU, and miscellaneous sensing and sample lines. USAR Appendix I, Subsection I.3.1 indicates that the postulation of high-energy line break (HELB) locations for high energy lines outside containment is, in part, based on the allowable stress criterion (SA),

consistent with MEB 3-1, B.1.c(2)(3) contained in Generic Letter (GL) 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements:

At terminal ends and any intermediate location where normal operating and seismic stresses exceeds GL 87-11 criteria for (seismic) Class I lines At terminal ends and each intermediate location of potential high stress or fatigue such as pipe fittings, valves, flanges and welded on attachments for (seismic) Class II lines.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 5 of 8 The maximum allowable stress criterion used to determine break locations is, in turn, based at least in part on time-dependent fatigue design cycles. Therefore, HELB analyses are also in part based on a set of anticipated design transients and considered TLAAs for SLR that must be evaluated for the SPEO.

TLAA Evaluation As stated in MNGP USAR Supplement K, MNGP piping systems were originally designed in accordance with ASA B31.1, 1955 Edition and USAS B31.1.0, 1967 Edition which did not require that an explicit fatigue analysis be performed. Also, reconciliation for the use of later editions of construction codes for modifications to or replacement of piping and components has been performed in accordance with Section IWA 7210(c),

Section XI of the ASME Code. The governing code for design, materials, fabrication and erection of piping, piping components, and pipe support modifications or replacements is ANSI B31.1, 1977 Edition including Addenda up to and including the Winter of 1978.

NonASME Class 1 components are excluded from the scope of this evaluation if they are in systems that may have normal/upset condition operating temperature that do not exceed 220 F. This is based on recommended values of 220 F for carbon steel or 270 F for austenitic stainless steel in the EPRI Fatigue Management Handbook (Reference 4.7.30).

Piping & instrument diagrams (P&IDs) were used to identify affected systems for this evaluation. In addition, specific station procedures were used to aid in this evaluation.

NUREG2192, Table 2.16, provides examples of structures, components and commodity groups associated with nonClass 1 piping components. This includes component types such as piping, tubing, expansion joints, fittings, couplings, reducers, elbows, thermowells, flanges, fasteners, and welded attachments.

Section 4.3.2.1.1 of NUREG2192 provides guidance for piping and components evaluated for fatigue parameters other than cumulative usage factor (CUFen) including fatiguebased maximum allowable stress calculations for components evaluated to B31.1 or ASME Code Class 2 and 3 requirements.

Table 4.3.61 provides a summary of the review performed to estimate 80 year cycles.

As described in the USAR supplement for initial license renewal, a conservative estimate of the number of thermal cycles experienced by the piping systems not analyzed to ASME Section III Class 1 requirements was approximated by using the maximum number of thermal cycles assumed in the reactor nozzle fatigue analyses. For MNGP the bounding number of cycles used for the qualification of a vessel nozzle is 1,500 for the feedwater nozzle. Table 4.3.61 was created to validate the approach used for initial license renewal.

Transient cycles on the bellows are composed of thermal cycles experienced by the associated system piping. The conservatively estimated cycles are provided in

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 6 of 8 Table 4.3.61. Conservatively estimated cycles for the systems and penetration bellows not analyzed to Class 1 requirements show significant margin to the 7,000 cycle value used for these piping systems and containment process penetration bellows.

For MNGP the limiting system from a total cycle standpoint is feedwater, which has as its design basis 1,500 applied thermal cycles to the nozzles for a 40year operating period. For the 80year extended operating period, the number of cycles was estimated by multiplying the 40year value times 2 which results in an estimated operating cycle expectation of 3,000 cycles. Since projected 80year cycles (Table 4.3.1-1) are less thant design cycles, this is a conservative estimate. This is less than half of the original requirement of 7,000 cycles.

Consequently, the current Class 2/3 piping and containment penetration bellows fatigue design criteria remain valid with significant margin for the 80 year SPEO.

The same is true for the HELB locations selected, implicitly based in part on fatigue, due to reliance on SA for postulated locations. Therefore, with the stress range reduction factor remaining at 1.0 as described above, the original locations identified through the HELB screening process are expected to remain unchanged and applicable during the SPEO.

TLAA Disposition: 10 CFR 54.21(c)(1)(i)

There are no inscope systems that are projected to experience more than 3,000 full range temperature cycles for a period of 80 years based on plant operation to date. This provides significant margin to the 7,000 cycle value which would require further evaluation and additionally support potential HELB locations outside of containment. Therefore, all of these systems at MNGP are suitable for extended operation without further evaluation and can be dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 7 of 8 The Metal Fatigue category in SLRA Table A-2 on pages A-8 and A-9 is revised as follows:

Table A2 List of TimeLimited Aging Analyses Category (Section) TimeLimited Aging Analyses Name Section 80Year Transient Cycle Projections A.3.3.1 ASME Section III, Class 1 Fatigue Waivers A.3.3.2 RPV Fatigue Analysis A.3.3.3 Fatigue Analysis of RPV Internals A.3.3.4 Metal Fatigue (A.3.3) ASME Section III, Class 1 A.3.3.5 ASME Section III, Class 2 and 3 and A.3.3.6 ANSI B31.1 & High-Energy Line Break Analyses EnvironmentallyAssisted Fatigue A.3.3.7 SLRA Section A.3.3 on page A-45 is revised as follows:

A.3.3 Metal Fatigue Fatigue is an age-related degradation mechanism caused by cyclic stressing of a component by either mechanical or thermal stresses. The thermal and mechanical fatigue analyses of plant mechanical components have been identified as TLAAs for MNGP. Specific components have been designed considering transient cycle assumptions, as listed in vendor specifications and the USAR. Fatigue analyses are considered TLAA for Class 1 and non-Class 1 mechanical components requiring evaluation for the SPEO in accordance with 10 CFR 54.21(c).

The following metal fatigue evaluations are documented in the following sections:

80-Year Transient Cycle Projections (Section A.3.3.1)

ASME Section III, Class 1 Fatigue Waivers (Section A.3.3.2)

RPV Fatigue Analyses (Section A.3.3.3)

Fatigue Analysis of RPV Internals (Section A.3.3.4)

ASME Section III, Class 1 (Section A.3.3.5)

ASME Section III, Class 2 and 3 and ANSI B31.1 & High-Energy Line Break Analyses (Section A.3.3.6)

Environmentally-Assisted Fatigue (Section A.3.3.7)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 8 of 8 SLRA Section A.3.3.6 on pages A-47 and A-48 is revised as follows:

A.3.3.6 ASME Section III, Class 2 and 3 and ANSI B31.1 & High-Energy Line Break Analyses The MNGP nonClass 1 Reactor Coolant System (RCS) piping and balanceofplant piping systems within the scope of SLR are designed to the requirements of the ANSI B31.7 and ANSI B31.1 Codes. Piping and components designed in accordance with these Codes are not required to have an explicit analysis of cumulative fatigue usage, but cyclic loading is considered in a simplified manner in the design process. These nonClass 1 piping Codes first require prediction of the overall number of thermal and pressure cycles expected during the lifetime of these components. Then a stress range reduction factor is determined for that number of cycles using a table from the applicable design code. If the total number of cycles is 7,000 or less, the stress range reduction factor is 1.0, which when applied, would not reduce the allowable stress value.

A review of the ANSI B31.7 and ANSI B31.1 piping within the scope of SLR was performed in order to identify those systems that operate at elevated temperature and to establish their cyclic operating practices. NonClass 1 components are excluded from the scope of this evaluation if they are in systems that may have normal/upset condition operating temperature that do not exceed 220 F. This is based on recommended values of 220 F for carbon steel or 270 F for austenitic stainless steel in the EPRI Fatigue Management Handbook. Piping & Instrument Diagrams (P&IDs) were used to identify affected systems for this evaluation.

The current Class 2/3 piping fatigue design criteria remain valid with significant margin for the 80 year SPEO. Therefore, all of these systems at MNGP are suitable for extended operation without further evaluation and can be dispositioned in accordance with 10 CFR 54.21(c)(1)(i).

USAR Appendix I, Subsection I.3.1 indicates that the postulation of HELB locations for lines outside containment is based on the allowable stress (SA).

The maximum allowable stress criterion used to determine break locations is, in turn, based at least in part on time-dependent fatigue design cycles. Similar to fatigue of ASME Class 2 and 3 and ANSI B31.1 lines, where the stress range reduction factor remains 1.0, the HELB analyses also remain valid for the SPEO.

Enclosure 18 Flex Power Versus Load Following

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8 Page 1 of 2 Flex Power Versus Load Following Flex power versus load following operational methods.

Affected SLRA Sections: B.2.2.1 SLRA Page Numbers: B-23 Description of Change:

MNGP SLRA Section B.2.2.1 is updated to discuss flex power versus load following operational methods.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8 Page 2 of 2 SLRA Section B.2.2.1 on page B-23 is revised to add a paragraph about flexible power operations as follows:

B.2.2.1 Fatigue Monitoring The MNGP Fatigue Monitoring AMP is an existing preventive AMP that manages fatigue damage of RPV components, RCPB piping components, and other components. This AMP provides an acceptable basis for managing fatigue of components that are subject to fatigue or cycle-based TLAAs or other analyses that assess fatigue or cyclical loading.

The Fatigue Monitoring AMP monitors and tracks the number of critical thermal, pressure, and seismic transients to ensure that the CUF and CUFen for each analyzed component does not exceed the applicable limit through the SPEO. The program monitors and tracks the number of thermal and pressure transients as specified in USAR Table 4.2-1.

Load-following operation is a design option, as described in USAR Sections 3.2.5 and 3.3.3.2.2. MNGP is primarily run as a baseload unit at 100% power. Flexible (flex) power operation at MNGP started in 2019.

Flexible power operation includes reducing power to 80%, allowing for windmills to operate when wind generation is predicted to be greater than demand. Flexible power operations and load-following changes in reactor power have minor impact on temperature (<50F) and pressure and have negligible impact on fatigue analyses.

Enclosure 19a Selective Leaching Supplements

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9a Page 1 of 3 Selective Leaching Supplements Remove reference to copper alloy components containing greater than 8 percent aluminum Affected SLRA Sections: A.2.2.21, B.2.3.21 SLRA Page Numbers: A-23, B-161 Description of Change:

SLRA sections A.2.2.21 and B.2.3.21 have been revised to remove reference to copper alloy components containing greater than 8 percent aluminum from the scope of the Selective Leaching AMP. The copper alloy components susceptible to selective leaching within the scope are copper alloys containing greater than 15 percent zinc.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9a Page 2 of 3 SLRA Section A.2.2.21 on page A-23 is revised as follows:

A.2.2.21 Selective Leaching The MNGP Selective Leaching AMP is an existing AMP that includes inspections of components that may be susceptible to loss of material due to selective leaching by demonstrating the absence of selective leaching (dealloying) of materials. The scope of this AMP includes components constructed of gray cast iron, ductile iron, and copper alloys (except for inhibited brass) containing greater than 15 percent zinc or greater than 8 percent aluminum in susceptible environments. One-time inspections for components exposed to a closed-cycle cooling water or treated water environment will be conducted, based on MNGP plant-specific OE which has not revealed selective leaching in these environments. Opportunistic and periodic inspections will be conducted for raw water, waste water, soil, and groundwater environments. Visual inspections coupled with mechanical examination techniques such as chipping or scraping are conducted. Periodic destructive examinations of components for physical properties (i.e., degree of dealloying, depth of dealloying, through-wall thickness, and chemical composition) are conducted for components exposed to raw water, waste water, soil, and groundwater environments. Inspections and tests will be conducted to determine whether loss of material will affect the ability of the components to perform their intended function for the SPEO. Inspections will be conducted in accordance with plant-specific procedures including inspection parameters such as lighting, distance, offset and surface conditions.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9a Page 3 of 3 SLRA Section B.2.3.21 on page B-161 is revised as follows:

B.2.3.21 Selective Leaching The MNGP Selective Leaching AMP is an existing AMP that has the principal objective to manage the aging effect of loss of material due to selective leaching.

The MNGP Selective Leaching AMP includes inspections of components made of gray cast iron, ductile iron, and copper alloys (except for inhibited brass) that contain greater than 15 percent zinc or greater than 8 percent aluminum exposed to a raw water, closed-cycle cooling water, treated water, waste water, or soil environment. For closed-cycle cooling water and treated water environments, the AMP includes one-time visual inspections of selected components that are susceptible to selective leaching, coupled with mechanical examination techniques (e.g., chipping, scraping).

For raw water, waste water, and soil environments, the AMP includes opportunistic and periodic visual inspections of selected components that are susceptible to selective leaching, coupled with mechanical examination techniques. Destructive examinations of components to determine the presence of and depth of dealloying through-wall thickness are also conducted. These techniques can determine whether loss of material due to selective leaching is occurring and whether selective leaching will affect the ability of the components to perform their intended function for the SPEO.

Enclosure 19b Fire Water Piping Coating

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9b Page 1 of 2 Fire Water Piping Coating Provide Clarification on Coating of Fire Water Piping Affected SLRA Sections: B.2.3.16 SLRA Page Numbers: B-114 Description of Change:

SLRA Section B.2.3.16 is updated to discuss the coatings and thickness for the Fire Water System buried piping.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9b Page 2 of 2 SLRA Section B.2.3.16 on Page B-114 is revised as follows:

The MNGP Fire Water System AMP also utilizes biocide as a preventive measure to prevent MIC. As another preventive measure, many fire water components are provided with a protective external coating to minimize the potential for external degradation. The external coating of buried cast iron or ductile iron piping is a bituminous coating of either coal-tar or asphalt base approximately 1 mil thick.

Additionally, the main fire water header is internally coated with a cementitious lining. Coatings minimize corrosion by limiting exposure to the environment, however, coatings are not credited for eliminating the aging effects/mechanisms.

Enclosure 20 Section B.2.3.4 Inclusion of Future Approvals

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0 Page 1 of 2 Section B.2.3.4 Inclusion of Future Approvals Update Section B.2.3.4 for Consistency with Future Approvals of Relief Request Affected SLRA Sections: B.2.3.4 SLRA Page Numbers: B-51 Description of Change:

Revise paragraph two of SLRA Section B.2.3.4 to include future approvals of relief requests under 10 CFR 50.55a.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0 Page 2 of 2 SLRA Section B.2.3.4 on Page B-51 is revised as follows:

B.2.3.4 BWR Vessel ID Attachment Welds The MNGP BWR Vessel ID Attachment Welds AMP is an existing condition monitoring program that manages the aging effect of cracking due to cyclic loading, SCC, and IGSCC of the BWR vessel ID attachment welds exposed to a reactor coolant environment. The MNGP BWR Vessel ID Attachment Welds AMP is implemented through station procedures that provide for mitigation of cracking through management of water chemistry and condition monitoring through examinations of reactor vessel interior attachment welds. The examination categories include volumetric, surface, and visual examination methods.

Under a relief request granted in accordance with 10 CFR 50.55a relief, examinations of BWR vessel ID attachment welds during ISI inspection interval #5 were are completed exclusively under consistent with the guidance of the BWRVIP program documents in lieu of ASME Section XI requirements, including schedule, extent, frequency, sequence of exams, reexaminations, and additional examinations (Reference ML16208A462). For inspections in the 7th and 8th intervals, that would occur during the SPEO, the ISI program at MNGP will, in accordance with 10 CFR 50.55a(g)(4), be updated each successive 120-month inspection interval to comply with the requirements of the latest edition of the ASME Code specified 18 months before the start of the inspection interval. Any deviation from the ASME Code,Section XI requirements (including those documented in a BWRVIP) will be approved by the NRC per a relief request prior to use. The exams are completed during general overview exams performed on the associated components attached to the RPV. The MNGP BWR Vessel ID Attachment Welds AMP incorporates the inspection and flaw evaluation recommendations of BWRVIP48A, Vessel ID Attachment Weld and Inspection and Flaw Evaluation Guidelines, and the recommendations for reactor water chemistry as described in the MNGP Water Chemistry AMP (B.2.3.2).

The MNGP BWR Vessel ID Attachment Welds AMP monitors the effects of cracking due to cyclic loading, SCC, and IGSCC by requiring inspections of the reactor vessel interior attachment welds as part of the ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection program and BWRVIP reports. A description of the ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection program, including the controlling edition of ASME Code,Section XI, is provided in the ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD AMP (B.2.3.1).

Enclosure 21 Consistency with Section A.2.2.19 and Table A-3

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 1 of 2 Consistency with Section A.2.2.19 and Table A-3 Update Section B.2.3.19 to include and subsequent NRC approved revisions Affected SLRA Sections: B.2.3.19 SLRA Page Numbers: B-153 Description of Change:

Include and subsequent NRC approved revisions in the Enhancement to Section B.2.3.19.

This will make Section B.2.3.19 consistent with Section A.2.2.19 and Table A-3.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1 Page 2 of 2 Section B.2.3.19 on page B-153 is revised as follows:

Element Affected Enhancement

1. Scope of Program; Implement BWRVIP321A, Boiling Water Reactor Vessel and Internals Project, Plan for

3. Parameters Monitored or Extension of the BWR Integrated Surveillance Inspected; (ISP) Through the Second License Renewal (SLR), and subsequent NRC approved

4. Detection of Aging Effects; and revisions upon obtaining NRC approval for MNGP to use BWRVIP321A to maintain

5. Monitoring and Trending compliance with 10 CFR Part 50, Appendix H.

2 IWF Supplements

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 1 of 9 IWF Supplements IWF Supplements to Address Editorial Inconsistencies Affected SLRA Sections: Table 3.5-1, Table A-3, A.2.2.30, and B.2.3.30 SLRA Page Numbers: 3.5-66, A-28, A-29, A-87, A-88, A-89, B-224, B-225, and B-226 Description of Change:

Table A-3, commitment item 33a, 33f and Section B.2.3.30, Element 1 and Element 4 are revised to include MC supports to make the language consistent with other areas of the NUREG-2191 Section XI.S3.

Table A-3, commitment item 33g and Section B.2.3.30 are clarified to add or polymeric to the vibration isolation elements referenced in the sections. This will ensure consistency with NUREG-2191 Section XI.S3 and Table A-3, commitment item 33d and 33j, and Section B.2.3.30.

Table A-3, commitment item 33h was revised to include MC supports and to make it consistent with Section B.2.3.30, Element 4.

Table A-3, commitment item 33i and Section B.2.3.30, Element 5 are revised to clarify that MNGP will increase or modify the component inspection sample (not population) when a component support is repaired to as-new condition by including another support that is representative of the remaining population of supports that were not repaired. This will ensure consistency with NUREG-2191 Section XI.S3, Element 5 (Monitoring and Trending).

Table A-3, implementation schedule for the one-time inspection cites the wrong commitment number. The correct commitment is 33f. This is a typographical error.

Section A.2.2.30 revises the language of the USAR supplement to use includes and will instead of recommends and should. This revision will make the language consistent with Table A-3, commitments 33f and 33h.

Table 3.5-1, Item Number 3.5.1-068 discussion states Not Applicable. This is revised to Not used and additional details will be provided in the discussion to make it reflect other details associated with the aging management of high-strength structural bolts.

Table A-3, commitment item 33e and Section B.2.3.30, Element 3 are revised to use the word excessive instead of significant loss of material. Significant implies there is a loss of intended function and that is not the case for these surfaces.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 2 of 9 SLRA Table 3.5-1 on page 3.5-66 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Aging Further Item Component Requiring Management Evaluation Discussion Number Management Program Recommended 3.5.1-068 High-strength steel Cracking due to SCC AMP XI.S3, "ASME No Not applicable used.

structural bolting Section XI, Subsection IWF There is no high-strength steel structural bolting used in MNGP structures or component supports.

Preventive actions and guidance for high-strength steel structural bolting is included for the ASME Section XI, Subsection IWF AMP to ensure proper aging management will be proceduralized to address the potential to use high-strength steel bolting at MNGP in the future.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 3 of 9 SLRA Appendix A.2.2.30 on pages A-28 and A-29 is revised as follows:

A.2.2.30 ASME Section XI, Subsection IWF The MNGP ASME Section XI, Subsection IWF AMP is an existing AMP and part of the MNGP ASME Section XI In-Service Inspection program. Inspections provide for condition monitoring of Class 1, 2, 3, and MC component supports. Component supports are selected for inspection in accordance with the ASME code classification. The quantity of component supports selected for examination is increased as a result of discovered support deficiencies. The program is updated periodically as required by 10 CFR 50.55a.

This AMP consists of periodic visual examination of piping and component supports for signs of degradation, evaluation of the examination results, and corrective actions for any identified deficiencies. This AMP recommends includes additional inspections beyond the inspections required by ASME Code Section XI, Subsection IWF. This consists of a one-time inspection of an additional 5 percent of the sample size specified in Table IWF-2500-1 for Class 1, 2, and 3, and MC piping supports. This one-time inspection is conducted within 5 years prior to entering the SPEO. For high-strength bolting in sizes greater than 1-inch nominal diameter, volumetric examination comparable to that of ASME Code Section XI, Table IWB-2500-1, Examination Category B-G-1 should will be performed to detect cracking in addition to the VT-3 examination.

If a component support does not exceed the acceptance standards of IWF-3400 but is electively repaired to as-new condition, the sample is increased or modified to include another support that is representative of the remaining population of supports that were not repaired.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 22 Page 4 of 9 SLRA Table A-3 on page A-87 is revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Implementation Schedule Program or Activity Section (Section) 33 ASME Section XI, XI.S3 The ASME Section XI, Subsection IWF AMP is an existing program No later than 6 months prior Subsection IWF that will be enhanced to: to the SPEO, or no later than (A.2.2.30) the last refueling outage prior a) Revise procedures to evaluate the acceptability of to the SPEO.

inaccessible areas (e.g., portions of ASME Class 1, 2, and 3, and MC supports encased in concrete, buried Start the one-time inspection underground, or encapsulated by guard pipe) when in commitment 33-gf) no conditions are identified in accessible areas that could earlier than 5 years prior to indicate the presence of, or result in, degradation to such the SPEO.

inaccessible areas.

b) Revise procedures to clarify that in addition to molybdenum disulfide (MoS2), other lubricants containing sulfur will be prohibited from use on structural bolting.

c) Revise procedures to specify the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of Research Council for Structural Connections publication Specification for Structural Joints Using High-Strength Bolts, for structural bolting consisting of ASTM A325, ASTM A490, and equivalent bolts.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 22 Page 5 of 9 SLRA Table A-3 on page A-88 is revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Implementation Schedule Program or Activity Section (Section) d) Revise procedures to specify that elastomeric or polymeric vibration isolation elements are monitored for cracking, loss of material, and hardening.

e) Revise procedures to specify that accessible sliding surfaces are monitored for significant excessive loss of material due to wear and accumulation of debris or dirt.

f) Perform and document a one-time inspection of an additional 5% of the sample populations for Class 1, 2, and 3, and MC piping supports. The additional supports will be selected from the remaining population of IWF piping supports and will include components that are most susceptible to age-related degradation.

g) Revise procedures to include tactile inspection (feeling, prodding) of elastomeric or polymeric vibration isolation elements to detect hardening if the vibration isolation function is suspect.

h) Revise procedures to specify that, for component supports with high-strength bolting greater than one-in. nominal diameter, volumetric examination comparable to that of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-G-1 will be performed to detect cracking in addition to the VT-3 examination. A representative sample of bolts will be inspected during the inspection interval prior to the start of the SPEO and in each 10-year period during the SPEO. Identify the population of ASME Class 1, 2, and 3, and MC high-strength structural bolting greater than one-in. nominal diameter within the boundaries of IWF-1300 and establish a sample to be 20% of the population (for a material/

environment combination) up to a maximum of 25 bolts.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 22 Page 6 of 9 SLRA Table A-3 on page A-89 is revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Implementation Schedule Program or Activity Section (Section) i) Revise procedures to increase or modify the component support inspection population sample when a component support is repaired to as-new condition by including another support that is representative of the remaining population of supports that were not repaired.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 7 of 9 SLRA Section B.2.3.30 on page B-224 is revised as follows:

B.2.3.30 ASME Section XI, Subsection IWF The ASME Section XI, Subsection IWF AMP is an existing AMP that consists of periodic visual examination of supports for ASME Class 1, 2, 3, and MC piping and components for signs of degradation such as corrosion; cracking; deformation; misalignment of supports; missing, detached, or loosened support items; loss of integrity of welds; improper clearances of guides and stops; and improper hot or cold settings of spring supports and constant load supports.

Bolting for Class 1, 2, and 3, and MC piping and component supports is also included and inspected for corrosion, loss of integrity of bolted connections due to self-loosening, and material conditions that can affect structural integrity.

The ASME Section XI, Subsection IWF AMP provides inspection and acceptance criteria and meets the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, 2007 edition with addenda through 2008, and 10 CFR 50.55a(b)(2) for Class 1, 2, 3, and MC piping and components and their associated supports. The primary inspection method employed is visual examination. NDE indications are evaluated against the acceptance standards of ASME Code Section XI. Examinations that reveal indications are evaluated. Examinations that reveal flaws or relevant conditions that exceed the referenced acceptance standard are expanded to include additional examinations during the current outage. The scope of inspection for supports is based on sampling of the total support population. The sample size varies depending on the ASME Code classification.

This AMP emphasizes proper selection of bolting material, lubricants, and installation torque or tension to prevent or minimize loss of bolting preload for structural bolting. As noted below in the enhancement discussion, the AMP also includes preventive actions for storage requirements of high-strength bolts and ensuring that molybdenum disulfide (MoS2) and other lubricants containing sulfur are not used for structural bolting. The requirements of ASME Code Section XI, Subsection IWF are supplemented to include volumetric examination of high-strength bolting for cracking. This AMP will also include a one-time inspection within 5 years prior to the SPEO of an additional 5 percent of piping supports from the remaining IWF population that are considered most susceptible to age-related degradation. Inspections of elastomeric or polymeric vibration isolation elements to detect hardening are also included if the vibration isolation function is suspect.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 8 of 9 SLRA Section B.2.3.30 on page B-225 is revised as follows:

Element Affected Enhancement

1. Scope of Program Revise procedures to evaluate the acceptability of inaccessible areas (e.g., portions of ASME Class 1, 2, and 3, and MC supports encased in concrete, buried underground, or encapsulated by guard pipe) when conditions are identified in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas.

2. Preventive Actions Revise procedures to clarify that in addition to molybdenum disulfide (MoS2), other lubricants containing sulfur will be prohibited from use on structural bolting.

2. Preventive Actions Revise procedures to specify the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of Research Council for Structural Connections publication Specification for Structural Joints Using High-Strength Bolts for structural bolting consisting of ASTM A325, ASTM A490, and equivalent bolts.

3. Parameters Monitored or Revise procedures to specify that elastomeric or polymeric Inspected vibration isolation elements are monitored for cracking, loss of material, and hardening.

3. Parameters Monitored or Revise procedures to specify that accessible sliding Inspected surfaces are monitored for significant excessive loss of material due to wear and accumulation of debris or dirt.

4. Detection of Aging Effects Perform and document a one-time inspection of an additional 5% of the sample populations for Class 1, 2, and 3, and MC piping supports. The additional supports will be selected from the remaining population of IWF piping supports and will include components that are most susceptible to age-related degradation. The one-time inspection will occur within five years prior to entering the SPEO.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 9 of 9 SLRA Section B.2.3.30 on page B-226 is revised as follows:

Element Affected Enhancement

4. Detection of Aging Effects Revise procedures to include tactile inspection (feeling, prodding) of elastomeric or polymeric vibration isolation elements to detect hardening if the vibration isolation function is suspect.

4. Detection of Aging Effects Revise procedures to specify that, for component supports with high-strength bolting greater than one-in. nominal diameter, volumetric examination comparable to that of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-G-1 will be performed to detect cracking in addition to the VT-3 examination. A representative sample of bolts will be inspected during the inspection interval prior to the start of the SPEO and in each 10-year period during the SPEO. Identify the population of ASME Class 1, 2, 3, and MC high-strength structural bolting greater than one-in.

nominal diameter within the boundaries of IWF-1300 and establish a sample to be 20% of the population (for a material/environment combination) up to a maximum of 25 bolts.

5. Monitoring and Trending Revise procedures to increase or modify the component support inspection population sample when a component support is repaired to as-new condition by including another support that is representative of the remaining population of supports that were not repaired.

Enclosure 23a Stress Corrosion Cracking in Copper Alloy with Greater Than 15% Zinc Components Exposed to Raw Water

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 1 of 5 Stress Corrosion Cracking in Copper Alloy with Greater Than 15% Zinc Components Exposed to Raw Water Stress corrosion cracking in copper alloy greater than 15% zinc components exposed to raw water Affected SLRA Sections: Tables 3.3-1 and 3.3.2-9 SLRA Page Numbers: 3.3-68, 3.3-183, 3.3-185, 3.3-190, and 3.3-205 Description of Change: The AMP managing the aging effect of cracking for copper alloy with greater than 15% zinc components with an internal or external environment of raw water is changed from the Fire Water System program to the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program. SLRA Table 3.3-1 line item 160 is revised to remove mention of the Fire Water System program in the discussion, and the relevant line items in Table 3.3.2-9 are revised to replace the Fire Water System program with the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components program as the AMP. Additionally, Plant-Specific Note 1 for Table 3.3.2-9 is revised to reflect this change.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 2 of 5 SLRA Table 3.3-1 on page 3.3-68 is revised as follows:

Table 3.31: Summary of Aging Management Evaluations for the Auxiliary Systems Further Item Aging Effect / Aging Management Component Evaluation Discussion Number Mechanism Program (AMP)/TLAA Recommended 3.3.1160 Copper alloy (>15% Cracking due to SCC AMP XI.M20, No Consistent with NUREG2191 Zn or >8% Al) piping, "OpenCycle Cooling with exception for the Fire Water piping components, Water System," AMP System (B.2.3.16) AMP.

heat exchanger XI.M21A, "Closed Treated components exposed Water Systems," or AMP The OpenCycle Cooling Water to closedcycle XI.M38, "Inspection of System (B.2.3.11), Closed cooling water, raw Internal Surfaces in Treated Water Systems water, waste water Miscellaneous Piping and (B.2.3.12) and Inspection of Ducting Components" Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

AMPs are used to manage cracking in copper alloy >15%

Zn components exposed to closedcycle cooling water or raw water. Additionally, the Fire Water System (B.2.3.16) AMP is used to manage cracking in copper alloy components exposed to raw water in the FIR System. This line item is also applied to components in the S&PC Systems.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 3 of 5 SLRA Table 3.3.2-9 on page 3.3-183 is revised as follows:

Table 3.3.29: Fire System - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Notes Type Function Requiring Management Item Item Management Program Heat Pressure Copper Alloy Raw Water Cracking Fire Water VII.C1.A473b 3.3.1160 E, 1 Exchanger Boundary with Greater (External) System (B.2.3.16)

(Diesel Fire Than 15% Inspection of Pump) Tube Zinc Internal Surfaces Sheet in Miscellaneous Piping and Ducting Components (B.2.3.24)

SLRA Table 3.3.2-9 on page 3.3-185 is revised as follows:

Table 3.3.29: Fire System - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Notes Type Function Requiring Management Item Item Management Program Heat Pressure Copper Alloy Raw Water Cracking Fire Water VII.C1.A473b 3.3.1160 E, 1 Exchanger Boundary with Greater (Internal) System (B.2.3.16)

(Diesel Fire Than 15% Inspection of Pump) Tubes Zinc Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 4 of 5 SLRA Table 3.3.2-9 on page 3.3-190 is revised as follows:

Table 3.3.29: Fire System - Summary of Aging Management Evaluation Component Intended Material Environment Aging Effect Aging NUREG-2191 Table 1 Notes Type Function Requiring Management Item Item Management Program Piping, Piping Pressure Copper Alloy Raw Water Cracking Fire Water VII.C1.A473b 3.3.1160 E, 1 Components Boundary with Greater (Internal) System (B.2.3.16)

Than 15% Inspection of Zinc Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 5 of 5 The plant-specific notes for SLRA Table 3.3.2-9 on page 3.3-205 are revised as follows:

PlantSpecific Notes

1. The Fire Water System (B.2.3.16) Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) program is being substituted for the OpenCycle Cooling Water System (B.2.3.11) program to manage cracking in copper alloy with greater than 15%

zinc piping components with an internal or external environment of raw water.

Enclosure 23b Clari"cation that the Requirement in Footnote 7 of Table XI.M27-1 in GALL-SLR is Satis"ed

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3b Page 1 of 3 Clarification that the Requirement in Footnote 7 of Table XI.M27-1 in GALL-SLR is Satisfied Clarification that the requirement in Footnote 7 of Table XI.M27-1 in GALL-SLR is satisfied Affected SLRA Sections: A.2.2.16, B.2.3.16 SLRA Page Numbers: A-20, B-113 Description of Change:

Footnote 7 of Table XI.M27-1 in GALL-SLR AMP XI.M27 requires of the SLRA one of three possible solutions to demonstrate acceptable wet pipe sprinkler systems. MNGP fulfills the requirement in the first bullet of Footnote 7 in demonstrating that the water is not corrosive to the sprinklers. The only environments to which the wet pipe sprinklers are exposed are plant indoor air, uncontrolled (external); raw water (internal); and condensation (internal). Therefore, per NFPA 25 Section 5.3.1.1.2 and Section A.5.3.1.1.2 in Annex A of NFPA 25, the wet pipe sprinklers are not exposed to any harsh environments, including corrosive atmospheres and corrosive water supplies. SLRA Sections A.2.2.16 and B.2.3.16 are revised to address this.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3b Page 2 of 3 SLRA Section A.2.2.16 on page A-20 is revised as follows:

wall-thickness examinations. Preventive actions (i.e., periodic flushes and biocide utilization) as well as periodic maintenance, testing, and inspection activities of the water-based fire protection systems are implemented to provide reasonable assurance that the fire water systems are capable of performing their intended functions. Inspections and testing are performed in accordance with guidance of applicable NFPA codes and standards with the following exception. An exception is taken that instead of performing the main drain tests on all standpipes and rises, the main drain tests will be performed on 20 percent of all standpipes and risers.

The MNGP wet pipe sprinkler systems are not exposed to any harsh or corrosive environments as defined in NFPA 25 Section 5.3.1.1.2 and Section A.5.3.1.1.2 of Annex A of NFPA 25. The wet pipe sprinklers are exposed only to an external environment of plant indoor air and internal environments of raw water and condensation.

The water-based fire protection system is normally maintained at required operating pressure and is monitored such that loss of system pressure is immediately detected and corrective actions are initiated. Piping wall thickness measurements are conducted when visual inspections detect surface irregularities indicative of unexpected levels of degradation. When the presence of organic or inorganic material sufficient enough to obstruct piping or sprinklers is detected, the material is removed, the source of the material is identified, and the source is corrected.

Inspections and tests follow site procedures that include inspection parameters for items such as lighting, distance, offset, presence of protective coatings, and cleaning processes for an adequate examination.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3b Page 3 of 3 SLRA Section B.2.3.16 on page B-113 is revised as follows:

B.2.3.16 Fire Water System The MNGP Fire Water System AMP is an existing AMP, that manages the aging effects of loss of material, wall thinning, cracking, and flow blockage due to fouling for water-based fire protection system components. This objective is achieved through conducting periodic visual inspections, tests, and flushes performed in accordance with the 2011 Edition of the National Fire Protection Association Code, NFPA 25 (Reference 1.6.47).

MNGP Fire Water System AMP applies to water-based fire protection system components, including closed head sprinklers; open head sprinklers and spray nozzles; fittings; valve bodies; fire pump casings; metallic equipment hoses (not fire hoses); hydrants; hose stations; standpipes; diesel fire pump heat exchanger; and aboveground, buried, and underground piping and components that are tested in accordance with the NFPA codes and standards. Full-flow testing and visual inspections are conducted in order to provide reasonable assurance that loss of material, cracking, and flow blockage are adequately managed. In addition to NFPA codes and standards, portions of the water-based fire protection system that are: (a) normally dry but periodically are subject to flow (e.g., dry-pipe or preaction sprinkler system piping and valves) and (b) that cannot be drained or allow water to collect, are subjected to augmented testing or inspections. Also, portions of the system (e.g., fire service main, standpipe) are normally maintained at required operating pressure and monitored such that loss of system pressure is immediately detected and corrective actions are initiated.

The MNGP wet pipe sprinkler systems are not exposed to any harsh or corrosive environments as defined in NFPA 25 Section 5.3.1.1.2 and Section A.5.3.1.1.2 of Annex A of NFPA 25. The wet pipe sprinklers are exposed only to an external environment of plant indoor air and internal environments of raw water and condensation.

The MNGP Buried and Underground Piping and Tanks AMP (B.2.3.27) is used to manage aging of the external surfaces of buried and underground fire water system piping. The MNGP Bolting Integrity AMP (B.2.3.10) will manage loss of preload, cracking, and loss of material for fire water system closure bolting. MNGP External Surfaces Monitoring of Mechanical Components AMP (B.2.3.23) will manage cracking of air-exposed copper alloy (>15 percent Zn) valve bodies, sprinklers, spray nozzles, and piping components through the SPEO. The MNGP Selective Leaching AMP (B.2.3.21) is used to manage aging of surfaces within the fire water system that have a material-environment combination susceptible to selective leaching.

Enclosure 23c Clari"cation of the Trending Process for Flow Testing and Wall Thickness Measurements

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3c Page 1 of 5 Clarification of the Trending Process for Flow Testing and Wall Thickness Measurements Clarification of the trending process for flow testing and wall thickness measurements Affected SLRA Sections: Table A-3, B.2.3.16 SLRA Page Numbers: A-66, A-67, B-118 Description of Change:

The MNGP Fire Water System program enhancements for Elements 5 and 6, Monitoring and Trending and Acceptance Criteria, respectively, are revised to clarify the trending process for flow testing and wall thickness measurements. Specifically, SLRA Table A-3 and the table of enhancements in SLRA Section B.2.3.16 are revised to clarify this.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3c Page 2 of 5 SLRA Table A-3, commitments 19d, 19e, and 19f on pages A-66 and A-67 are revised as follows:

Table A3 List of SLR Commitments and Implementation Schedule No. Aging NUREG2191 Commitment Implementation Schedule Management Section Program or Activity (Section) 19 Fire Water System XI.M27 d) Clarify that, where practical, degradation identified will be No later than 6 months prior (A.2.2.16) projected until the next scheduled inspection. Results will to the SPEO, or no later than be evaluated against acceptance criteria to confirm that the the last refueling outage prior timing of subsequent inspections will maintain the to the SPEO components intended functions throughout the SPEO based on the projected rate of degradation. Results of flow Implement the AMP and start testing (e.g., buried and underground piping, fire mains, and the preSPEO inspections sprinklers/spray nozzles), flushes, and wall thickness and tests no earlier than 5 measurements will be monitored and trended per the years prior to the SPEO.

instructions of the specific test/inspection procedure. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g., attached to the completed work order). However, if a trend is negative, a CAP item is initiated to evaluate the trend and determine any follow-up corrective actions.

Degradation identified by flow testing, flushes, and inspections will be evaluated. If the condition of the piping/component does not meet acceptance criteria, then the issue will be entered into the corrective action program, and the component will be evaluated for cleaning, recoating, repair, or replacement. For samplingbased inspections, results will be evaluated against acceptance criteria to confirm that the sampling bases (e.g., selection, size, frequency) will maintain the components intended functions throughout the SPEO based on the projected rate

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3c Page 3 of 5 and extent of degradation.

e) Update spray and sprinkler system flushing procedures to document and trend deposits (scale or foreign material). Incorporate acceptance criteria that no loose fouling products can exist in the systems that could cause flow blockage in the sprinklers or deluge nozzles.

Include steps in flushing procedures to compare the amount of deposits to the previous inspections results, and if the trend shows increasing deposits, then the CAP will be utilized to drive improvement. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g., attached to the completed work order).

However, if a trend is negative, a CAP item is initiated to evaluate the trend and determine any follow-up corrective actions.

f) Clarify that identified wall loss greater than the manufacturers tolerance will be entered into the CAP for engineering evaluation and trending to determine when minimum wall thickness will be reached and what corrective actions are required. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g.,

attached to the completed work order). However, if a trend is negative, a CAP item is initiated to evaluate the trend and determine any follow-up corrective actions.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3c Page 4 of 5 Section B.2.3.16 on page B-118, is revised as follows:

Element Affected Enhancement

5. Monitoring and Trending Update inspection and test procedures and preventive maintenance activities to state that, where practical, degradation identified will be projected until the next scheduled inspection.

Results will be evaluated against acceptance criteria to confirm that the timing of subsequent inspections will maintain the components intended functions throughout the SPEO based on the projected rate of degradation. Results of flow testing (e.g.,

buried and underground piping, fire mains, and sprinklers/spray nozzles), flushes, and wall thickness measurements will be monitored and trended per the instructions of the specific test/inspection procedure. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g., attached to the completed work order). However, if a trend is negative, a CAP item is initiated to evaluate the trend and determine any follow-up corrective actions. Degradation identified by flow testing, flushes, and inspections will be evaluated. If the condition of the piping/component does not meet acceptance criteria, then the issue will be entered into the MNGP CAP, and the component will be evaluated for cleaning, recoating, repair, or replacement. For sampling-based inspections, results will be evaluated against acceptance criteria to confirm that the sampling bases (e.g., selection, size, frequency) will maintain the components intended functions throughout the SPEO based on the projected rate and extent of degradation.

5. Monitoring and Trending Update spray and sprinkler system flushing procedures to enable trending of data. Specifically, the existing flushing procedures and preventive maintenance activities will be revised to document and trend deposits (scale or foreign material).

Incorporate acceptance criteria that no loose fouling products can exist in the systems that could cause flow blockage in the sprinklers or deluge nozzles.

Existing flushing procedures, as well as new flushing procedures, will include steps to compare the amount of deposits to the previous inspections results, and if the trend shows increasing deposits, then the MNGP CAP will be utilized to drive improvement. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g., attached to the completed work order). However, if a trend is negative, a CAP item is initiated to evaluate the trend and determine

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3c Page 5 of 5 Element Affected Enhancement any follow-up corrective actions.

6. Acceptance Criteria Clarify within the new internal inspection procedure and relevant existing preventive maintenance activities which inspect wall thickness that identified wall loss greater than the manufacturers tolerance will be entered into the MNGP CAP for engineering evaluation and trending to determine when minimum wall thickness will be reached and what corrective actions are required. For inspections and testing, the inspection and testing results data will be documented and accessible for future use or trending, regardless of whether the trend is positive, negative, or neutral (e.g., attached to the completed work order). However, if a trend is negative, a CAP item is initiated to evaluate the trend and determine any follow-up corrective actions.

Enclosure 23d Removal of Wording Implying that MNGP Has More Than One Unit

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3d Page 1 of 4 Removal of Wording Implying that MNGP Has More Than One Unit Removal of wording implying that MNGP has more than one Unit Affected SLRA Sections: Table A-3, Commitment 19i, B.2.3.16 SLRA Page Numbers: A-68, B-120 Description of Change:

Commitment 19i and Element 7 of the Fire Water System AMP are revised to remove including consideration to the other unit systems since MNGP is a single unit site.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3d Page 2 of 4 SLRA Table A-3 commitment 19i on page A-68 is revised as follows:

Table A3 List of SLR Commitments and Implementation Schedule No. Aging NUREG2191 Commitment Implementation Schedule Management Section Program or Activity (Section) 19 Fire Water System XI.M27 i) Clarify that for ongoing degradation mechanisms such as No later than 6 months (A.2.2.16) MIC or recurring internal corrosion, the frequency and extent prior to the SPEO, or no of wall thickness inspections are increased commensurate later than the last refueling with the significance of the degradation. The number of outage prior to the SPEO increased inspections is determined in accordance with the CAP; however, no fewer than 5 additional inspections are Implement the AMP and conducted for each inspection that did not meet acceptance start the preSPEO criteria, or 20% of each applicable material, environment, and inspections and tests no aging effect combination is inspected, whichever is less. The earlier than 5 years prior to additional inspections will occur at least every 24 months until the SPEO.

the rate of recurring internal corrosion occurrences no longer meets the criteria for loss of material due to recurring internal corrosion as defined in NUREG 2192. The selected inspection locations will be periodically reviewed to validate their relevance and usefulness and adjusted as appropriate.

Evaluation of the inspection results will include (1) a comparison to the nominal wall thickness or previous wall thickness measurements to determine rate of corrosion degradation; (2) a comparison to the design minimum allowable wall thickness to determine the acceptability of the component for continued use; and (3) a determination of reinspection interval. If a failure occurs (e.g., a throughwall leak or blockage impacting operability), the failure mechanism shall be identified and used to determine the most susceptible system locations for additional inspections, including consideration to the other unit systems as driven by the corrective action program. When piping is replaced prior to failure, due to concerns with wall thinning or blockage,

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3d Page 3 of 4 inspections are considered for similar areas of the system to determine the presence and extent of degradation.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3d Page 4 of 4 Section B.2.3.16 on page B-120 is revised as follows:

Element Affected Enhancement

7. Corrective Actions Clarify within the new internal inspection procedure(s) and relevant preventive maintenance activities that for ongoing degradation mechanisms such as MIC or recurring internal corrosion, the frequency and extent of wall thickness inspections are increased commensurate with the significance of the degradation. The number of increased inspections is determined in accordance with the MNGP CAP; however, no fewer than five additional inspections are conducted for each inspection that did not meet acceptance criteria, or 20 percent of each applicable material, environment, and aging effect combination is inspected, whichever is less. The additional inspections will occur at least every 24 months until the rate of recurring internal corrosion occurrences no longer meets the criteria for loss of material due to recurring internal corrosion as defined in NUREG 2192. The selected inspection locations will be periodically reviewed to validate their relevance and usefulness and adjusted as appropriate. Evaluation of the inspection results will include (1) a comparison to the nominal wall thickness or previous wall thickness measurements to determine rate of corrosion degradation; (2) a comparison to the design minimum allowable wall thickness to determine the acceptability of the component for continued use; and (3) a determination of reinspection interval. If a failure occurs (e.g., a throughwall leak or blockage impacting operability), the failure mechanism shall be identified and used to determine the most susceptible system locations for additional inspections, including consideration to the other unit systems as driven by the corrective action program. When piping is replaced prior to failure, due to concerns with wall thinning or blockage, inspections are considered for similar areas of the system to determine the presence and extent of degradation.

Enclosure 24a Update Section A.2.2.9 to Remove Reference to the Erosion Module

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4a Page 1 of 2 Update Section A.2.2.9 to Remove Reference to the Erosion Module Section A.2.2.9 is revised to remove reference to the Erosion Module.

Affected SLRA Sections: A.2.2.9 SLRA Page Numbers: A-17 Description of Change:

Section A.2.2.9 is revised to remove the reference to the erosion module of the software tool, FAC Manager', as the erosion module is not used. The change to Section A.2.2.9 is consistent with the discussion in Section B.2.3.9 of the SLRA.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4a Page 2 of 2 SLRA Section A.2.2.9 on page A-17 is revised as follows:

A.2.2.9 Flow-Accelerated Corrosion The MNGP Flow-Accelerated Corrosion (FAC) AMP is an existing AMP that manages wall thinning caused by flow-accelerated corrosion, as well as wall thinning due to erosion mechanisms. This AMP is based on industry guidelines (Nuclear Safety Analysis Center document, (NSAC) 202L R4) and industry OE.

A predictive analytical software EPRI computer program CHECWORKS' is used to predict component wear rates and remaining service life in the systems susceptible to FAC which provides reasonable assurance that structural integrity will be maintained between inspections. Additionally, the software tool, FAC Manager', with the erosion module, is used to evaluate components for both FAC and erosion. The software QA classification for CHECWORKS' and FAC Manager' are Classification Level 2, which is important to compliance with regulatory requirements/commitments, required by nuclear laws or regulations, or whose failure to operate as expected may have an indirect effect on nuclear plant safety, in accordance with the Software QA Program.

Enclosure 24b Add EPRI 3002023786 Guidance to FAC

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4b Page 1 of 3 Add EPRI 3002023786 Guidance to FAC Add Enhancement to provide guidance from EPRI 3002023786 Affected SLRA Sections: Table A-3, Commitment 12 and Section B.2.3.9 SLRA Page Numbers: A-57 and B-71 Description of Change:

The flow acceleration corrosion program is enhanced to provide guidance consistent with the erosion remaining service life safety factor from EPRI 3002023786 for known erosion mechanisms. Changes from the recommended safety factor of 2.0 will be documented in the FAC program as required by EPRI 3002023786.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4b Page 2 of 3 Table A-3 on page A-57 is revised as follows:

No. Aging NUREG Commitment Implementation Management -2191 Schedule Program or Section Activity (Section) 12 Flow- XI.M17 The Flow-Accelerated Corrosion AMP is an existing program that will be No later than 6 Accelerated enhanced to: months prior to Corrosion the SPEO, or no (A.2.2.9) a) Perform a re-assessment of piping systems that have been excluded later than the last from wall thickness monitoring due to operation less than 2 percent of refueling outage plant operating time (as allowed by NSAC-202L-R4) to ensure that prior to the SPEO adequate bases exist to justify this exclusion for the SPEO.

b) Provide guidance to evaluate inspection results to determine if assumptions in the extent of condition review remain valid. If degradation is associated with infrequent operational alignments, such as surveillances or pump starts/stops, then trending activities should consider the number or duration of these occurrences.

c) Provide guidance consistent with the erosion remaining service life safety factor provided in EPRI 3002023786 for known erosion mechanisms and changes from the recommended safety factor of 2.0 will be documented in the FAC program as required by EPRI 3002023786.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4b Page 3 of 3 SLRA Section B.2.3.9 on page B-71 is revised as follows:

Element Affected Enhancement

6. Acceptance Criteria Provide guidance consistent with the erosion remaining service life safety factor provided in EPRI 3002023786 for known erosion mechanisms and changes from the recommended safety factor of 2.0 will be documented in the FAC program as required by EPRI 3002023786.

Enclosure 25 Clari"cation of Sample Main Drain Testing at Risers and Standpipes

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5 Page 1 of 4 Clarification of Sample Main Drain Testing at Risers and Standpipes Clarification of sample main drain testing at risers and standpipes Affected SLRA Sections: A.2.2.16, B.2.3.16 SLRA Page Numbers: A-20, B-116, B-131 Description of Change:

The exception to the Fire Water System program Element 4 in Section B.2.3.16 of the SLRA is revised to clarify that the main drain tests of a 20 percent sample of the water-based risers and standpipes will be conducted at different locations each refueling outage so that all riser and standpipes will undergo main drain testing within a 10-year period. Section A.2.2.16 is also revised to clarify this.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5 Page 2 of 4 SLRA Section A.2.2.16, page A-20 is revised as follows:

wall-thickness examinations. Preventive actions (i.e., periodic flushes and biocide utilization) as well as periodic maintenance, testing, and inspection activities of the water-based fire protection systems are implemented to provide reasonable assurance that the fire water systems are capable of performing their intended functions. Inspections and testing are performed in accordance with guidance of applicable NFPA codes and standards with the following exception. An exception is taken that instead of performing the main drain tests on all standpipes and rises, the main drain tests will be performed on 20 percent of all standpipes and risers. The 20 percent sample testing will be performed at different locations each refueling outage so that all risers and standpipes will undergo main drain testing within a 10-year period.

The water-based fire protection system is normally maintained at required operating pressure and is monitored such that loss of system pressure is immediately detected and corrective actions are initiated. Piping wall thickness measurements are conducted when visual inspections detect surface irregularities indicative of unexpected levels of degradation. When the presence of organic or inorganic material sufficient enough to obstruct piping or sprinklers is detected, the material is removed, the source of the material is identified, and the source is corrected. Inspections and tests follow site procedures that include inspection parameters for items such as lighting, distance, offset, presence of protective coatings, and cleaning processes for an adequate examination.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5 Page 3 of 4 SLRA Section B.2.3.16, page B-116 is revised as follows:

valves associated with fire suppression systems at MNGP due to flow blockage has not occurred.

MNGP will take the following exception to the NUREG-2191 guidance in the MNGP Fire Water System AMP:

(1) MNGP will perform the main drain tests on 20 percent of the standpipes and risers every refueling cycle. The 20 percent sample testing will be performed at different locations each refueling outage so that all risers and standpipes will undergo main drain testing within a 10-year period.

This is an exception to NUREG-2191, Table XI.M27-1 and NFPA 25, Section 13.2.5.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5 Page 4 of 4 The table in SLRA Section B.2.3.16 on page B-131 is revised as follows:

Description NFPA 25 Required Enhancements Section water supply piping and CVs and any time the CV is closed and reopened at system riser.

As identified in the exception to the MNGP Fire Water System AMP, MNGP will perform the main drain tests on 20 percent of the standpipes and risers every refueling cycle. The 20 percent sample testing will be performed at different locations each refueling outage so that all risers and standpipes will undergo main drain testing within a 10-year period.

13.2.5.1: In systems where the sole water supply is through a backflow preventer and/or pressure reducing valves, the main drain test of at least one system downstream of the device shall be conducted on a quarterly basis.

13.2.5.2: When there is a 10 percent reduction in full flow pressure when compared to the original acceptance test or previously performed tests, the cause of the reduction shall be identified and corrected if necessary.

Per NUREG-2191 Table XI.M27-1, the following notes also apply:

Items in areas that are inaccessible because of safety considerations such as those raised by continuous process operations, radiological dose, or energized electrical equipment are inspected during each scheduled shutdown but not more often than every refueling outage interval.

Calibration of measuring and test equipment is conducted in accordance with plant-specific procedures in lieu of NFPA 25 requirements.

Enclosure 26 Masonry Wall Voluntary Supplements

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6 Page 1 of 4 Masonry Walls Voluntary Supplements Masonry Walls AMP Commitment Revisions Affected SLRA Sections: Table A-3 Commitment 33 and Section B.2.3.32 SLRA Page Numbers: A-90 and B-236 Description of Change:

Revise Table A-3, Commitment 35 and the enhancement to element 1 in SLRA Section B.2.3.32 (Masonry Walls) to delete the EFB from the inspection. The EFB has no masonry walls.

Revise Table A-3, Commitment 35 and the enhancements to element 5 in SLRA Section B.2.3.32 (Masonry Walls) to include the comparison of inspection results with previous inspections to identify changes or trends in the condition of masonry walls.

Revise Table A-3, Commitment 35 and the enhancements to element 7 in SLRA Section B.2.3.32 (Masonry Walls) to add a corrective action option to develop a new analysis or evaluation basis that accounts for the degraded condition of the wall (i.e., acceptance by further evaluation).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 26 Page 2 of 4 SLRA Table A-3 on page A-90 is revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Implementation Schedule Program or Activity Section (Section) 35 Masonry Walls XI.S5 The Masonry Walls Amp is an existing program that will be enhanced to: No later than 6 months prior to (A.2.2.32) the SPEO, or no later than the a) Update the implementing procedure to include the inspection of last refueling outage prior to the masonry walls in the EFB and Radwaste Building. SPEO.

b) Update the implementing procedure to monitor and inspect for gaps between the supports and masonry walls that could potentially impact the intended function or potentially invalidate its evaluation basis.

c) Update the implementing procedure for more frequent inspections in areas where significant loss of material, cracking, or other signs of degradation are projected or observed to provide reasonable assurance than there is no loss of intended function between inspections.

d) Update the implementing procedure for trending of crack widths and lengths and gaps between supports and masonry walls that approach or exceed acceptance criteria.

e) Update the implementing procedure will include projected degradation until the next scheduled inspection where it is practical.

f) Update the implementing procedure to include acceptance criteria to ensure observed aging effects do not invalidate the evaluation basis of the wall or impact its intended function.

g) Update the implementing procedure to state that if any projected inspection results will not meet acceptance criteria prior to the next scheduled inspection, inspection frequencies are adjusted as determined by the MNGP CAP.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 6 Page 3 of 4 h) Update the implementing procedure to include a corrective action option to develop a new analysis or evaluation basis that accounts for the degraded condition of the wall (i.e., acceptance by further evaluation).

i) Update the implementing procedure to include the comparison of inspection results with previous inspections to identify changes or trends in the condition of masonry walls.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 26 Page 4 of 4 SLRA Section B.2.3.32 on page B-236 is revised as follows:

Element Affected Enhancement

1. Scope of Program Enhance the implementing procedure to include the inspection of masonry walls in the EFB and Radwaste Building.

3. Parameters Monitored or Inspected Enhance the implementing procedure to monitor and inspect for gaps between the supports and masonry walls that could potentially impact the intended function or potentially invalidate its evaluation basis.

4. Detection of Aging Effects Enhance the implementing procedure to include provisions for more frequent inspections in areas where significant loss of material, cracking, or other signs of degradation are projected or observed to provide reasonable assurance than there is no loss of intended function between inspections.

5. Monitoring and Trending Enhance the implementing procedure to include trending of widths and lengths of cracks and gaps between supports and masonry walls that approach or exceed acceptance criteria.

5. Monitoring and Trending Enhance the implementing procedure to include projected degradation until the next scheduled inspection where it is practical.

5. Monitoring and Trending Enhance the implementing procedure to include the comparison of inspection results with previous inspections to identify changes or trends in the condition of masonry walls.

6. Acceptance Criteria Enhance the implementing procedure to include acceptance criteria for masonry wall inspections that will be used to ensure observed aging effects (cracking, loss of material, or gaps between the structural steel supports and masonry walls) do not invalidate the evaluation basis of the wall or impact its intended function.

7. Corrective Actions Enhance the implementing procedure to ensure that if any projected inspection results will not meet acceptance criteria prior to the next scheduled inspection, inspection frequencies are adjusted as determined by the MNGP CAP.

7. Corrective Actions Enhance the implementing procedure to initiate a corrective action option to develop a new analysis or evaluation basis that accounts for the degraded condition of the wall (i.e., acceptance by further evaluation)

Enclosure 27 ASME Section XI, Subsection IWE AMP Clari"cations

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 1 of 16 ASME Section XI, Subsection IWE AMP Clarifications ASME Section XI, Subsection IWE AMP Clarifications Affected SLRA Sections: Sections 3.5.2.2.1.3 and 3.5.2.2.1.6, Table 3.5-1, Table A-3 #32, Sections B.2.3.29 and B.2.3.30 SLRA Page Numbers: 3.5-21, 3.5-25, 3.5-26, 3.5-44, 3.5-54, A-85, A-86, A-87, B-213, B-214, B-215, B-216, B-217, B-220, and B-228 Description of Change:

Introductory text in SLRA Section 3.5.2.2.1.3 and data in Table 3.5-1 are revised to more accurately reflect applicability of AMR items presented in NUREG-2191 to the MNGP Mark I steel containment.

Introductory text in SLRA Section 3.5.2.2.1.6 and data in Table 3.5-1 are revised to more accurately reflect applicability of the AMR items presented in NUREG-2191 to the MNGP Mark I steel containment. The section is also clarified to state that 8 vent line bellows are included in the population of penetration bellows from which the sample to be examined for cracking will be drawn. A concluding statement is added to reflect that review of plant-specific OE over the past ten years has yielded no evidence of SCC for the subject components.

Provided further clarification that one-time supplemental volumetric examinations will be triggered by plant-specific OE after the date of issuance of the first renewed license. Also clarified implementation schedules for pre-SPEO inspections.

Updated commitments and enhancements to more clearly define "high temperature" as used to categorize piping penetrations.

Additional information is provided to more clearly describe a sample expansion strategy to confirm the absence of the SCC aging effect.

Included additional text from the further evaluation for cracking due to SCC (SLRA Section 3.5.2.2.1.6) in the AMP program description (SLRA Section B.2.3.29).

Clarified text to justify 'Exceptions to NUREG-2191' by (a) using phrasing that better aligns with discussion of fatigue waiver analysis in Appendix A to SLR-ISG-2021-03-STRUCTURES and (b) deleting details about components that are NOT addressed by the fatigue waiver analysis performed for MNGP.

The ambiguous term exceptional was removed from descriptions of the overall ISI program performance status. The Green program health color is described in the governing plant procedure for program health and status reporting.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 2 of 16 SLRA Section 3.5.2.2.1.3 on page 3.5-21 is revised as follows:

As summarized in items 3.5.1-004, 3.5.1-005, and 3.5.1-035, loss of material due to general, pitting, and crevice corrosion of steel elements in accessible and inaccessible areas is not applicable to the MNGP Mark I steel containment.

The MNGP primary containment design includes an accessible moisture barrier at the concrete floor to drywell shell interface perimeter and includes an inaccessible sheet metal cover and joint sealing compound above the sand pocket region on the exterior of the drywell shell. The ASME Section XI, Subsection IWE (B.2.3.29) AMP performs an examination of the accessible moisture barrier at the concrete to shell interface for wear, damage, erosion, tears, cracks, or other defects that may violate the leak-tight integrity. There has been no corrosion detected at the moisture barrier at the bottom of the drywell interior.

The MNGP primary containment design includes an inaccessible 2-inch air gap between the exterior steel drywell surface and the concrete sacrificial shield. MNGP has three drainage paths for removing leakage into the drywell air gap. The first path prevents leakage past the refueling bellows from entering the air gap. This consists of a drain line located below the bellows bellows non-wetted side (i.e., FPW7-8 to FPW7-4 with flow switch FS-2792 to light panel C65 located on operating floor to alarm on panel C04 located in the MCR). The second path is the air gap to sand pocket interface where there is a galvanized steel plate which is sealed to the drywell shell. Four-inch drain lines are provided to remove water that might collect on the plate from above. The third pathway is from the sand pocket itself, which is provided with four, 2-inch drain lines. To provide reasonable assurance that moisture is not present in the air gap region of the steel drywell, the ASME Section XI, Subsection IWE (B.2.3.29) AMP monitors for blockage and leakage of the drywell air gap and sand pocket drain line outlets during each outage when the refueling cavity is flooded. The drywell to reactor building refueling seal is addressed as part of the Reactor Building.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 3 of 16 SLRA Section 3.5.2.2.1.6 on page 3.5-25 is being revised as follows:

As summarized in items 3.5.1-010, 3.5.1-038, and 3.5.1-039, cracking due to SCC is an applicable aging effect when stainless steel or nickel alloy components are exposed to temperatures in excess of 140 F. The suppression chamber and drywell shells at MNGP, as well as penetration nozzles, sleeves, etc., are made of carbon steel and not susceptible to SCC. Stainless steel or nickel alloy components of the primary containment include: Torus thermowells, electrical penetration canisters, certain piping penetration manifold plates/spare penetration nozzles/TIP drive penetration nozzles, personnel airlock leakage test connections, and hot piping penetration double ply expansion bellows, and the vent line bellows.

Connection of these stainless steel or nickel alloy components to the steel drywell or torus involve dissimilar metal welds (DMWs). Of these, only expansion bellows may experience temperatures in excess of 140 F during normal operation.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 4 of 16 SLRA Section 3.5.2.2.1.6 on page 3.5-26 is being revised as follows:

• MNGP replaced the double ply bellows on core spray penetration X16B after finding cracks during local leakage rate testing in the 1996 refueling outage. The replacement was done during the 1998 refueling outage. The original bellows was fabricated from austenitic stainless steel; the replacement is Inconel.

The suppression chamber and drywell shells at MNGP, as well as penetration nozzles, sleeves, etc., are made of carbon steel and not susceptible to SCC.

Cracking due to SCC of stainless steel (SS) or nickel alloy (NA) penetration bellows (hot fluid penetrations), and associated DMWs will be managed by the ASME Section XI, Subsection IWE (B.2.3.29) AMP and the 10 CFR Part 50, Appendix J (B.2.3.31) AMP, as clarified below. Additionally, the vent system is relied upon as a pathway for steam between the drywell and the torus in the event of a pipe rupture. Furthermore, the vent system also provides support for a portion of the SRV piping inside the vent line and suppression chamber. Loads which act on the SRV piping are transferred to the vent system by the penetration assembly which is welded to the vent. Each of the 8 vent lines includes a stainless steel expansion bellows assembly.

Cracking due to SCC of stainless steel (SS) or nickel alloy (NA) penetration bellows (hot fluid penetrations), and associated DMWs will be managed by the ASME Section XI, Subsection IWE (B.2.3.29) AMP and the 10 CFR Part 50, Appendix J (B.2.3.31) AMP. As such, the ASME Section XI, Subsection IWE (B.2.3.29) AMP will be enhanced to include one-time volumetric/surface examination of 20 percent of these 24 penetration bellows (i.e., 5 inspections). In addition, due to being higher temperature, these penetrations are also leading indicators for cyclic load cracking of other susceptible drywell shell, penetration sleeve or other locations.

Note that plant-specific OE (i.e., IWE examination results and results from leak rate testing performed under the 10 CFR Part 50, Appendix J AMP) over the past ten years have yielded no evidence of SCC for the subject components.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 27 Page 5 of 16 SLRA Table 3.5-1 on page 3.5-44 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Item Aging Effect Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1-004 Steel elements Loss of material due to XI.S1, Yes (SRP-SLR Not applicable.Consistent with (inaccessible areas): general, pitting, crevice ASME Section XI, Section 3.5.2.2.1.3.1) NUREG-2191, as clarified.

liner; liner anchors; corrosion Subsection Section integral attachments, IWE, and XI.S4, 10 CFR This item number is not applicable to steel elements Part 50, the MNGP Mark I steel containment.,

(inaccessible areas): Appendix J even though it specifically suppression chamber; addresses This item number is drywell; drywell head; applicable only to BWR Mark III embedded shell; containments.

region shielded by The 10 CFR Part 50, Appendix J diaphragm (B.2.3.31) and ASME Section XI, floor (as applicable) Subsection IWE (B.2.3.29) programs will be used to manage loss of material due to general, pitting, and crevice corrosion in steel elements of inaccessible areas of the MNGP Mark I steel containment drywell shell and drywell head are addressed in items 3.5.1-035 and 3.5.1-041.

Further evaluation is documented in Section 3.5.2.2.1.3.1.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 27 Page 6 of 16 SLRA Table 3.5-1 on page 3.5-44 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1-005 Steel elements Loss of material due to XI.S1, Yes (SRP-SLR Not applicable.

(inaccessible areas): general pitting corrosion "ASME Section XI, Section 3.5.2.2.1.3.1) liner; liner anchors; Subsection IWE" and This item number is not applicable to integral attachments, XI.S4 "10 CFR Part 50, the MNGP Mark I steel containment.

steel elements Appendix J" This item number is applicable only (inaccessible areas): to PWR concrete and steel suppression chamber; containments, BWR Mark II drywell; drywell head; containments, and BWR Mark I and embedded shell; Mark III concrete containments.

region shielded by diaphragm floor (as Further evaluation is documented in applicable) Section 3.5.2.2.1.3.1.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 27 Page 7 of 16 SLRA Table 3.5-1 on page 3.5-54 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1-038 Steel elements: Cracking due to SCC XI.S1, Yes (SRP-SLR Not applicable.

suppression chamber "ASME Section XI, Section 3.5.2.2.1.6) shell (interior surface) Subsection IWE" and This item number is not applicable to XI.S4 "10 CFR Part 50, the MNGP Mark I steel containment.

Appendix J" This item number is applicable only to BWR Mark III containments.

Further evaluation is documented in Section 3.5.2.2.1.6.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 27 Page 8 of 16 SLRA Table A-3 on page A-85 is being revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Schedule Program or Activity Section (Section) 32 ASME Section XI, XI.S1 The ASME Section XI, Subsection IWE AMP is an existing program that No later than 6 months prior Subsection IWE (A.2.2.29) will be enhanced to: to the SPEO, or no later than the last refueling outage prior to the SPEO.

a) Revise procedures to specify the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of Research Council for Structural Connections Start the one-time publication Specification for Structural Joints Using High-Strength supplemental inspections in Bolts, for structural bolting consisting of ASTM A325, ASTM commitments 32-c), 32-d) no A490, and equivalent bolts. earlier than 5 years prior to the SPEO.

b) Revise procedures to specify that accessible noncoated surfaces (including those comprising the torus vent system) are monitored Complete one-time for arc strikes. inspection in commitment 32-e (of metal shell locations if degradation from the inaccessible side is identified) on a schedule established by the MNGP corrective action program.

Inspection will be scheduled to provide reasonable assurance that the metal shell intended function is maintained consistent with the CLB through the SPEO.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 27 Page 9 of 16 SLRA Table A-3 on page A-86 is being revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Schedule Program or Activity Section (Section) c) Implement periodic supplemental surface or enhanced visual examinations, in addition to visual examinations, at intervals no greater than 10 years to detect cracking on accessible portions of high-temperature (temperatures above 140oF) drywell piping penetrations that are not pressurized during local leak rate testing and have no CLB fatigue analysis. Cracking is corrected by repair or replacement or accepted by engineering evaluation.

d) Conduct supplemental one-time surface or enhanced visual examinations, performed by qualified personnel using methods capable of detecting cracking, comprising a representative sample 5 of the stainless steel penetrations or dissimilar metal welds associated with high-temperature (temperatures above 140 F) stainless steel piping systems in frequent use. These inspections are intended to confirm the absence of SCC aging effects.

e) Revise procedures to specify a one-time volumetric examination of metal shell surfaces that are inaccessible from one side if triggered by plant-specific OE identified after the date of issuance of the initial renewed license. If triggered, this inspection will be performed by sampling randomly selected, as well as focused, metal shell locations susceptible to corrosion that are inaccessible from one side. The trigger for this one-time examination is plant-specific occurrence or recurrence of metal shell corrosion (base metal material loss exceeding 10% of nominal plate thickness) that is determined to originate from the inaccessible side. Any such instance would be identified through code inspections performed since November 8, 2006. Guidance provided in EPRI TR-107514 will be considered when establishing a sampling plan.

This sampling is conducted to demonstrate, with 95% confidence, that 95% of the accessible portion of the metal shell is not experiencing greater than 10% wall loss.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 27 Page 10 of 16 SLRA Table A-3 on page A-87 is being revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Schedule Program or Activity Section (Section) f) If SCC is identified as a result of the supplemental one-time inspections, additional inspections will be conducted in accordance with the sites corrective action process. This will include incrementing sample size by one additional penetration with at a time from the uninspected population of stainless steel penetrations or dissimilar metal welds associated with high-temperature (greater than 140 F) stainless steel piping systems in frequent use until cracking is no longer detected. Periodic inspection of subject penetrations with dissimilar metal welds for cracking will be added to the ASME Section XI, Subsection IWE AMP if necessary, depending on the inspection results.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 11 of 16 SLRA Section B.2.3.29 on page B-213 is revised as follows:

If plant-specific OE identified after the date of issuance of the initial renewed license triggers the requirement to implement a one-time supplemental volumetric examination, then this inspection is performed by sampling randomly selected, as well as focused, metal shell locations susceptible to corrosion that are inaccessible from one side. Guidance provided in EPRI TR-107514 (Reference 1.6.51) will be considered for sampling determinations. The trigger for this one-time examination is plant-specific occurrence or recurrence of metal shell corrosion (base metal material loss exceeding 10 percent of nominal plate thickness) that is determined to originate from the inaccessible side. Any such instance would be identified through code inspections performed since November 8, 2006. Based on a review of current MNGP OE, no such triggers have occurred.

SLRA Section B.2.3.29 on page B-214 is revised as follows:

Cumulative fatigue damage for the MNGP drywell penetration bellows is addressed in the Containment Liner Plate, Metal Containments, and Penetrations Fatigue Analysis TLAA for SLR (Section 4.5). Cracking due to cyclic loading for portions of high-temperature drywell piping penetrations that are not pressurized during local leak rate testing and have no current licensing bases fatigue analysis will be managed by periodic supplemental surface or enhanced visual examinations incorporated into and consistent with the frequency of this AMP. This AMP will also include supplemental one-time inspections within 5 years prior to the SPEO for a representative sample of stainless steel penetrations and dissimilar metal welds that may be susceptible to SCC. In addition, due to being higher temperature, these penetrations are also leading indicators for cyclic load cracking of other susceptible drywell shell, penetration sleeve or other locations.

SLRA Section B.2.3.29 on pages B-214 and B-215 is revised as follows:

Exceptions to NUREG-2191 The Evaluation and Technical Basis discussions for the XI.S1 AMP in NUREG-2191 state that steel, stainless steel, and DMW pressure-retaining components that are subject to cyclic loading but have no CLB fatigue analysis are monitored for cracking (Element 3) and are supplemented with surface examination (or other applicable technique) in addition to visual examination to detect cracking (Element 4). The MNGP ASME Section XI, Subsection IWE AMP will take exception to this NUREG-2191 guidance as summarized below:

The MNGP primary containment was designed to the requirements of ASME Code Section III, Subsection B, 1965 Edition with 1965 Winter Addenda. A fatigue evaluation was not required by the 1965 Edition or by original MNGP construction specifications. An assessment fatigue waiver analysis was performed demonstrating that the six criteria for cyclic loading in paragraph NE-3222.4(d) of ASME Code,Section III, Division 1 (1974 Edition with Addenda through Winter 1975) are satisfied for the drywell shell and Class MC portions of the drywell penetrations. to address Tthe following design inputs were addressed for component materials comprising the MNGP

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 12 of 16 primary containment that could be subject to cyclic loading but have no CLB fatigue analysis:

(1) Atmospheric-to-operating pressure cycle, (2) Normal operation pressure fluctuation, (3) Temperature difference - startup and shutdown, (4) Temperature difference - normal operation, (5) Temperature difference - dissimilar metals, and (6) Mechanical loads.

The assessment fatigue waiver analysis concluded that the drywell shell, non-high temperature drywell penetrations, and penetration sleeves are subjected to a small and acceptable amount of fatigue such that neither detailed fatigue analysis nor a fatigue waiver is not required. As such, cracking due to cyclic loading does not require aging management for the drywell shell, non-high temperature drywell penetrations, and penetration sleeves.

MNGP does The fatigue waiver analysis did not monitor for cracking utilizing supplemental surface examinations except at accessible portions of certain steel and stainless steel penetrations associated with high temperature systems. Original design and installation specifications for containment penetration components such as bellows, welds, and include penetration adapters of high-temperature drywell mechanical penetrations (i.e., those associated with greater than 140oF stainless steel piping systems in frequent use).required surface examinations to ensure no flaws existed as part of initial installation. Appropriate testing is conducted for pressure boundary These components per the 10 CFR Part 50, Appendix J AMP (B.2.3.31). Through-wall cracking would be detected by the Type A integrated leak rate test. Additionally, visual examinations are performed on (i.e., accessible portions of the containment high temperature drywell piping penetrations in accordance with the MNGP IWE Plan.

Since issuance of the initial renewed license, MNGP has that are not pressurized during local leak rate testing) are subject to aging management as described by the 10 elements of NUREG-2191,Section XI.S1experienced a failure of the subject containment components and integrated leak rate test results have been within the overall limits. Industry OE has also shown strong performance of the subject primary containment components. Thus, existing 10 CFR Part 50, Appendix J leak testing and ASME Section XI, Subsection IWE examinations at MNGP remain adequate for the drywell shell, non-high temperature drywell penetrations, and penetration sleeves without supplemental surface examination to detect cracking.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 13 of 16 Enhancements The MNGP ASME Section XI, Subsection IWE AMP will be enhanced as follows, for alignment with NUREG-2191. The one-time inspection will be started no earlier than five years prior to the SPEO. The enhancements will be implemented, and one-time pre-SPEO inspections completed no later than six months prior to entering the SPEO, or no later than the last refueling outage prior to the SPEO.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 14 of 16 SLRA Section B.2.3.29 on page B-216 is being revised as follows:

Element Affected Enhancement

2. Preventive Actions Revise procedures to specify the preventive actions for storage, lubricants, and stress corrosion cracking potential discussed in Section 2 of Research Council for Structural Connections publication Specification for Structural Joints Using High-Strength Bolts, for structural bolting consisting of ASTM A325, ASTM A490, and equivalent bolts.

3. Parameters Monitored or Inspected Revise procedures to specify that accessible Inspected noncoated surfaces (including those comprising the torus vent system) are monitored for arc strikes.

3. Parameters Monitored or Implement periodic supplemental surface or enhanced Inspected visual examinations, in addition to visual examinations, at intervals no greater than 10 years to detect cracking on

4. Detection of Aging Effects accessible portions of high-temperature (temperatures above 140 F) drywell piping penetrations that are not

6. Acceptance Criteria pressurized during local leak rate testing and have no CLB fatigue analysis. Cracking is corrected by repair or replacement or accepted by engineering evaluation. The supplemental inspections will start no earlier than five years prior to entering the SPEO.

4. Detection of Aging Effects Conduct supplemental one-time surface or enhanced visual examinations, performed by qualified personnel using methods capable of detecting cracking, comprising a representative sample (five) of the stainless steel penetrations or DMWs associated with high-temperature (temperatures above 140 F) stainless steel piping systems in frequent use. These inspections are intended to confirm the absence of SCC aging effects. The one-time supplemental inspections will occur within start no earlier than five years prior to entering the SPEO.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 15 of 16 SLRA Section B.2.3.29 on page B-217 is being revised as follows:

Element Affected Enhancement

4. Detection of Aging Effects Revise procedures to specify a one-time volumetric examination of metal shell surfaces that are inaccessible from one side if triggered by plant-specific OE identified after the date of issuance of the initial renewed license. If triggered, this inspection will be performed by sampling randomly selected, as well as focused, metal shell locations susceptible to corrosion that are inaccessible from one side.

The trigger for this one-time examination is plant-specific occurrence or recurrence of metal shell corrosion (base metal material loss exceeding 10% of nominal plate thickness) that is determined to originate from the inaccessible side. Any such instance would be identified through code inspections performed since November 8, 2006. Guidance provided in EPRI TR-107514 will be considered when establishing a sampling plan. This sampling is conducted to demonstrate, with 95% confidence, that 95% of the accessible portion of the metal shell is not experiencing greater than 10% wall loss.

7. Corrective Actions If SCC is identified as a result of the supplemental one-time inspections, additional inspections will be conducted in accordance with the sites corrective action process. This will include incrementing sample size by one additional penetration with at a time from the uninspected population of stainless steel penetrations or DMWs associated with high-temperature (greater than 140 F) stainless steel piping systems in frequent use until cracking is no longer detected. Periodic inspection of subject penetrations with DMWs for cracking will be added to the MNGP ASME Section XI, Subsection IWE AMP if necessary, depending on the inspection results.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 7 Page 16 of 16 SLRA Section B.2.3.29 on page B-220 is revised as follows:

The ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection program health report (July 2020) was reviewed. The overall program performance health color was reported as exceptional (GREEN).

SLRA Section B.2.3.30 on page B-228 is revised as follows:

The ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection program health report (July 2020) was reviewed. The overall program performance health color was reported as exceptional (GREEN).

Enclosure 28a Correct Casings and Housings with a Leakage Boundary Intended Function for the Turbine Generator System

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8a Page 1 of 7 Correct Casings and Housings with a Leakage Boundary Intended Function for the Turbine Generator System Revise Leakage Boundaries Within the Turbine Generator System Affected SLRA Sections: Section 3.4.2.1.6, Tables 2.3.4-6 and 3.4.2-6 SLRA Page Numbers: 2.3-81; 2.3-82; 3.4-7; 3.4-103; 3.4-111; 3.4-113; 3.4-114; 3.4-115 Description of Change:

Update SLRA Table 2.3.4-6 to include the Blower Housing (Vapor Extractor) component type with the intended function of a leakage boundary. SLRA Table 3.4.2-6 is also to be updated to include the Blower Housing (Vapor Extractor) component type. Update SLRA Section 3.4.2.1.6 to include Condensation as an environment.

Update SLRA Table 2.3.4-6 to remove the Pump Casing (Emergency Bearing Oil Pump), Pump Casing (Turbine Bearing Lift Pump), Pump Casing (Turb Aux Oil Pump), and Pump Casing (Turning Gear Oil Pump). These pump casings are internal to the Turbine Generator Oil Reservoir, and therefore do not have a leakage boundary intended function. SLRA Table 3.4.2-6 is also updated to remove the following component types: Pump Casing (Emergency Bearing Oil Pump), Pump Casing (Turbine Bearing Lift Pump), Pump Casing (Turb Aux Oil Pump), and Pump Casing (Turning Gear Oil Pump).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8a Page 2 of 7 SLRA Table 2.3.4-6 on pages 2.3-81 and 2.3-82 is revised as follows:

Table 2.3.46 Turbine Generator System Components Subject to Aging Management Review Component Type Component Intended Function Blower Housing (Vapor Extractor) Leakage Boundary Bolting (Closure) Mechanical Closure Heat Exchanger (Exciter Air Cooler) Shell Side Leakage Boundary Components Heat Exchanger (Exciter Air Cooler) Tube Side Leakage Boundary Components Heat Exchanger (Generator Hydrogen Cooler) Leakage Boundary Shell Side Components Heat Exchanger (Generator Hydrogen Cooler) Leakage Boundary Tube Side Components Heat Exchanger (Isophase Bus Cooler) Shell Leakage Boundary Side Components Heat Exchanger (Isophase Bus Cooler) Tube Leakage Boundary Side Components Heat Exchanger (Stator Water Cooling Heat Leakage Boundary Exchanger) Shell Side Components Heat Exchanger (Stator Water Cooling Heat Leakage Boundary Exchanger) Tube Side Components Heat Exchanger (Steam Packing Exhauster) Holdup and Plateout Shell Side Components Heat Exchanger (Steam Packing Exhauster) Holdup and Plateout Tube Side Components Heat Exchanger (Turbine Lubricating Oil Leakage Boundary Cooler) Shell Side Components Heat Exchanger (Turbine Lubricating Oil Leakage Boundary Cooler) Tube Side Components Piping Elements Leakage Boundary Piping, Piping Components Holdup and Plateout Leakage Boundary Pump Casing (Emergency Bearing Oil Pump) Leakage Boundary Pump Casing (Emergency H2 Seal Oil Pump) Leakage Boundary Pump Casing (EPR Oil Pump) Leakage Boundary Pump Casing (Main H2 Seal Oil Pump) Leakage Boundary Pump Casing (Recirc H2 Seal Oil Pump) Leakage Boundary Pump Casing (Seal Oil Vacuum Pump) Leakage Boundary Pump Casing (Stator Liquid Cooling Pump) Leakage Boundary Pump Casing (Steam Packing Exhauster Holdup and Plateout Blower)

Pump Casing (Turb Aux Oil Pump) Leakage Boundary Pump Casing (Turbine Lubricating Oil Purifier Leakage Boundary Pump)

Pump Casing (Turbine Bearing Lift Pump) Leakage Boundary Pump Casing (Turning Gear Oil Pump) Leakage Boundary

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8a Page 3 of 7 SLRA Section 3.4.2.1.6, Page 3.4-7 is revised as follows:

3.4.2.1.6 Turbine Generator Materials The materials of construction for the TGS components are:

Carbon and Low Alloy Steel Bolting Carbon Steel Copper Alloys with 15% Zinc or Less Copper Alloy Greater Than 15% Zinc Glass Gray Cast Iron Stainless Steel Stainless Steel Bolting Environments The TGS components are exposed to the following environments:

Air Indoor Uncontrolled Condensation Gas Lubricating Oil Raw Water Treated Water Treated Water >140°F Due to steam quality, the environment of steam is identified as Treated Water >140°F Aging Effects Requiring Management The following aging effects associated with the TGS require management:

Cracking LongTerm Loss of Material Loss of Material Loss of Preload Wall Thinning

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 28a Page 4 of 7 SLRA Table 3.4.2-6 on page 3.4-103 is revised as follows:

Table 3.4.26: Turbine Generator - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Blower Housing Leakage Carbon Air-Indoor Loss of Material External Surfaces VIII.H.S-29 3.4.1-034 A (Vapor Extractor) Boundary Steel Uncontrolled Monitoring of (External) Mechanical Components (B.2.3.23)

Blower Housing Leakage Carbon Condensation Loss of Material Inspection of Internal VIII.E.SP-60 3.4.1-037 A (Vapor Extractor) Boundary Steel (Internal) Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Bolting (Closure) Mechanical Carbon and Air Indoor Loss of Material Bolting Integrity VIII.H.S02 3.4.1009 A Closure Low Alloy Uncontrolled (B.2.3.10)

Steel Bolting (External)

Bolting (Closure) Mechanical Carbon and Air Indoor Loss of Preload Bolting Integrity VIII.H.SP142 3.4.1006 A Closure Low Alloy Uncontrolled (B.2.3.10)

Steel Bolting (External)

Bolting (Closure) Mechanical Carbon and Lubricating Oil Loss of Material Bolting Integrity VIII.H.S418 3.4.1070 A Closure Low Alloy (External) (B.2.3.10)

Steel Bolting Bolting (Closure) Mechanical Carbon and Lubricating Oil Loss of Preload Bolting Integrity VIII.H.SP142 3.4.1006 A Closure Low Alloy (External) (B.2.3.10)

Steel Bolting Bolting (Closure) Mechanical Stainless Air Indoor Cracking Bolting Integrity VIII.H.S421 3.4.1073 A Closure Steel Bolting Uncontrolled (B.2.3.10)

(External)

Bolting (Closure) Mechanical Stainless Air Indoor Loss of Material Bolting Integrity VIII.H.S02 3.4.1009 A Closure Steel Bolting Uncontrolled (B.2.3.10)

(External)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 28a Page 5 of 7 SLRA Table 3.4.2-6 on page 3.4-111 is revised as follows:

Table 3.4.26: Turbine Generator - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Emergency Boundary (External) (B.2.3.25)

Bearing Oil Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Emergency Boundary (External) (B.2.3.20)

Bearing Oil Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Emergency Boundary (Internal) (B.2.3.25)

Bearing Oil Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Emergency Boundary (Internal) (B.2.3.20)

Bearing Oil Pump)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 28a Page 6 of 7 SLRA Table 3.4.2-6 on pages 3.4-113, 3.4-114, and 3.4-115 is revised as follows:

Table 3.4.26: Turbine Generator - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Turb Aux Oil Boundary (External) (B.2.3.25)

Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Turb Aux Oil Boundary (External) (B.2.3.20)

Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Turb Aux Oil Boundary (Internal) (B.2.3.25)

Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Turb Aux Oil Boundary (Internal) (B.2.3.20)

Pump)

Pump Casing Leakage Carbon Steel Air Indoor Loss of Material External Surfaces VIII.H.S29 3.4.1034 A (Turb Lubricating Boundary Uncontrolled Monitoring of Oil Purifier Pump) (External) Mechanical Components (B.2.3.23)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Turb Lubricating Boundary (Internal) (B.2.3.25)

Oil Purifier Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Turb Lubricating Boundary (Internal) (B.2.3.20)

Oil Purifier Pump)

Pump Casing Leakage Carbon Steel Air Indoor Loss of Material External Surfaces VIII.H.S29 3.4.1034 A (Turbine Bearing Boundary Uncontrolled Monitoring of Lift Pump) (External) Mechanical Components (B.2.3.23)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Turbine Bearing Boundary (Internal) (B.2.3.25)

Lift Pump)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 28a Page 7 of 7 Table 3.4.26: Turbine Generator - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Turbine Bearing Boundary (Internal) (B.2.3.20)

Lift Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Turning Gear Oil Boundary (External) (B.2.3.25)

Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Turning Gear Oil Boundary (External) (B.2.3.20)

Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C (Turning Gear Oil Boundary (Internal) (B.2.3.25)

Pump)

Pump Casing Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C (Turning Gear Oil Boundary (Internal) (B.2.3.20)

Pump)

Enclosure 28b Addition of EPR Components

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8b Page 1 of 6 Addition of EPR Components Update Turbine Generator AMR Tables to Add Components of EPR System.

Affected SLRA Sections: Tables 2.3.4-6, 3.4-1, and 3.4.2-6 SLRA Page Numbers: 2.3-81, 2.3-82, 3.4-26, 3.4-103, 3.4-115 Description of Change:

Update SLRA Tables 2.3.4-6 and 3.4.2-6 to include Electrohydraulic (also called Electrical or Electric) Pressure Regulator (EPR) components not already in the Turbine Generator AMR.

EPR Oil Pump Casings are already included within the AMR; however this update adds the EPR Oil Tank (and associated sight glass), Accumulator, Piping (including oil filter housings),

and Valve Bodies. Update SLRA Table 3.4-1 Item 40 to include tanks in the discussion.

Of these newly added components, which provide a Leakage Boundary intended function, only the EPR Oil Tank and Accumulator add new component types to SLRA Tables 2.3.4-6 and 3.4.2-6. The sight glass for the EPR Oil Tank is encompassed by the existing items for Piping Elements, EPR Piping (including oil filter housings) is encompassed by existing items for Piping, Piping Components, and EPR Valve Bodies are encompassed by existing items for Valve Body.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8b Page 2 of 6 SLRA Table 2.3.4-6 on Page 2.3-81 is revised as follows:

Component Type Component Intended Function(s)

Accumulator (EPR) Leakage Boundary Bolting (Closure) Mechanical Closure Heat Exchanger (Exciter Air Cooler) Shell Side Leakage Boundary Components Heat Exchanger (Exciter Air Cooler) Tube Side Leakage Boundary Components Heat Exchanger (Generator Hydrogen Cooler) Leakage Boundary Shell Side Components Heat Exchanger (Generator Hydrogen Cooler) Leakage Boundary Tube Side Components Heat Exchanger (Isophase Bus Cooler) Shell Leakage Boundary Side Components

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8b Page 3 of 6 SLRA Table 2.3.4-6 on Page 2.3-82 is revised as follows:

Component Type Component Intended Function(s)

Tanks (Clean Lubricating Oil Storage Tank) Leakage Boundary Tanks (Dirty Lubricating Oil Storage Tank) Leakage Boundary Tanks (EPR Oil Tank) Leakage Boundary Tanks (LO Purifier Tank) Leakage Boundary Tanks (Lubricating Oil Dump Overflow Tank) Leakage Boundary Tanks (Lubricating Oil Tank) Leakage Boundary Tanks (Moisture Separator Drain Tank) Leakage Boundary Tanks (Moisture Separator) Leakage Boundary Tanks (Oily Water Separator Tank) Leakage Boundary Tanks (Seal Oil Detraining Tank) Leakage Boundary Tanks (Stator Cooling Surge Tank) Leakage Boundary Tanks (Turbine Lubricating Oil Purifier Drain Leakage Boundary Tank)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 8b Page 4 of 6 SLRA Table 3.4-1 on page 3.4-26 is revised as follows:

Table 3.4-1: Summary of the Aging Management Evaluations for the Steam and Power Conversion Systems Item Aging Effect / Aging Management Further Evaluation Component Discussion Number Mechanism Programs Recommended 3.4.1-040 Steel piping, Loss of material AMP XI.M39, No Consistent with NUREG-2191.

piping due to general, "Lubricating Oil components pitting, crevice Analysis," and AMP This line item is also applied to exposed to corrosion, MIC XI.M32, "One-Time heat exchanger components and lubricating Inspection" tanks. The Lubricating Oil oil Analysis (B.2.3.25) and One-Time Inspection (B.2.3.20) AMPs are used to manage loss of material of steel piping, piping components, tanks, and heat exchanger components exposed to lubricating oil.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 28b Page 5 of 6 SLRA Table 3.4.2-6 on page 3.4-103 is revised as follows:

Table 3.4.26: Turbine Generator - Summary of Aging Management Evaluation Aging Effect Component Intended Aging Management NUREG-2191 Table 1 Material Environment Requiring Notes Type Function Program Item Item Management Accumulator Leakage Carbon Air - Indoor Loss of Material External Surfaces VIII.H.S-29 3.4.1-034 A (EPR) Boundary Steel Uncontrolled Monitoring of (External) Mechanical Components (B.2.3.23)

Accumulator Leakage Carbon Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP-91 3.4.1-040 C (EPR) Boundary Steel (Internal) (B.2.3.25)

Accumulator Leakage Carbon Lubricating Oil Loss of Material One-Time Inspection VIII.A.SP-91 3.4.1-040 C (EPR) Boundary Steel (Internal) (B.2.3.20)

Bolting Mechanical Carbon and Air Indoor Loss of Material Bolting Integrity (B.2.3.10) VIII.H.S02 3.4.1009 A (Closure) Closure Low Alloy Uncontrolled Steel Bolting (External)

Bolting Mechanical Carbon and Air Indoor Loss of Preload Bolting Integrity (B.2.3.10) VIII.H.SP142 3.4.1006 A (Closure) Closure Low Alloy Uncontrolled Steel Bolting (External)

Bolting Mechanical Carbon and Lubricating Oil Loss of Material Bolting Integrity (B.2.3.10) VIII.H.S418 3.4.1070 A (Closure) Closure Low Alloy (External)

Steel Bolting Bolting Mechanical Carbon and Lubricating Oil Loss of Preload Bolting Integrity (B.2.3.10) VIII.H.SP142 3.4.1006 A (Closure) Closure Low Alloy (External)

Steel Bolting

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 28b Page 6 of 6 SLRA Table 3.4.2-6 on page 3.4-115 is revised as follows:

Table 3.4.2-6: Turbine Generator - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Tanks (Dirty Leakage Carbon Steel Air Indoor Loss of Material External Surfaces VIII.H.S29 3.4.1034 A Lubricating Oil Boundary Uncontrolled Monitoring of Mechanical Storage Tank) (External) Components (B.2.3.23)

Tanks (Dirty Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C Lubricating Oil Boundary (Internal) (B.2.3.25)

Storage Tank)

Tanks (Dirty Leakage Carbon Steel Lubricating Oil Loss of Material OneTime Inspection VIII.A.SP91 3.4.1040 C Lubricating Oil Boundary (Internal) (B.2.3.20)

Storage Tank)

Tanks (EPR Oil Leakage Carbon Air-Indoor Loss of External Surfaces VIII.H.S-29 3.4.1-034 A Tank) Boundary Steel Uncontrolled Material Monitoring of Mechanical (External) Components (B.2.3.23)

Tanks (EPR Oil Leakage Carbon Lubricating Oil Loss of Lubricating Oil Analysis VIII.A.SP-91 3.4.1-040 C Tank) Boundary Steel (Internal) Material (B.2.3.25)

Tanks (EPR Oil Leakage Carbon Lubricating Oil Loss of One-Time Inspection VIII.A.SP-91 3.4.1-040 C Tank) Boundary Steel (Internal) Material (B.2.3.20)

Tanks (LO Purifier Leakage Carbon Steel Air Indoor Loss of Material External Surfaces VIII.H.S29 3.4.1034 A Tank) Boundary Uncontrolled Monitoring of Mechanical (External) Components (B.2.3.23)

Tanks (LO Purifier Leakage Carbon Steel Lubricating Oil Loss of Material Lubricating Oil Analysis VIII.A.SP91 3.4.1040 C Tank) Boundary (Internal) (B.2.3.25)

Enclosure 29 Corrosion Structural Supplement

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 9 Page 1 of 2 Corrosion Structural Supplement Correction of Structural Support Corrosion Discussion Affected SLRA Sections: Table 3.5-1 SLRA Page Numbers: 3.5-74 Description of Change:

The Discussion for Table 3.5-1, Item 3.5.1-098 is revised to reflect that the item is Not Applicable at MNGP. This is because there are no stainless steel or aluminum alloy support members, welds, bolted connections, or support anchorage to building structures exposed to an air with borated water leakage environment at MNGP.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 29 Page 2 of 2 SLRA Table 3.5-1 on page 3.5-74 is being revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1-098 Stainless steel, None None No Not used applicable.

aluminum alloy support members; There are no stainless steel or welds; bolted aluminum alloy support members, connections; support welds, bolted connections, or support anchorage to building anchorages to building structures structure exposed to an air with borated water leakage environment.This component, material, and environment combination is addressed by item number 3.5.1-099.

Enclosure 30a External Surfaces Monitoring of Mechanical Components

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0a Page 1 of 4 External Surfaces Monitoring of Mechanical Components Clarification of Heating Coils Affected SLRA Sections: Table 3.3.2-11 SLRA Page Numbers: 3.3-225, 3.3-227, 3.3-230, 3.3-231, and 3.3-242 Description of Change:

The Aging Effects Requiring Management, Aging Management Programs, NUREG-2191 Items, Table 1 items, and Notes for the following component types in Table 3.3.2-11 are revised: Heat Exchanger - (Reactor Building Heating Coils) Tube Side Components, Heat Exchanger -

(Reactor Building Main Supply HVAC Unit Heating Coil) Tube Side Components, Heat Exchanger - (Turbine Building Reheaters) Tube Side Components, and Heat Exchanger - (Unit Heaters) Tube Side Components. The revisions are:

Revise Aging Effect Requiring Management to None; Revise Aging Management Program to None; Revise NUREG-2191 item to VII.J.AP-144; Revise Table 1 item to 3.3.1-114; and Revise Notes to reflect plant specific note 3.

In Plant Specific Notes, add #3 to indicate SCC is not applicable because the copper alloy heat exchanger components are in an air - indoor uncontrolled environment.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0a Page 2 of 4 Table 3.3.2-11 on page 3.3-225 is revised as follows:

Table 3.3.2-11: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-Component Type Material Environment Requiring Table 1 Item Notes Function Program 2191 Management Item Heat Exchanger - (Reactor Leakage Copper Air - Indoor Cracking External Surfaces VIII.H.S-454 3.4.1-106 C, 3 Building Heating Coils) Boundary Alloy Uncontrolled None Monitoring of 3.3.1-114 Tube Side Components with (External) Mechanical VII.J.AP-144 Greater Components (B.2.3.23)

Than None 15%

Zinc Table 3.3.2-11 on page 3.3-227 is revised as follows:

Table 3.3.2-11: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-Component Type Material Environment Requiring Table 1 Item Notes Function Program 2191 Management Item Heat Exchanger - (Reactor Leakage Copper Air - Indoor Cracking External Surfaces VIII.H.S-454 3.4.1-106 C, 3 Building Main Supply Boundary Alloy Uncontrolled None Monitoring of 3.3.1-114 HVAC with (External) Mechanical VII.J.AP-144 Unit Heating Coil) Tube Greater Components (B.2.3.23)

Side Components Than None 15%

Zinc

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0a Page 3 of 4 Table 3.3.2-11 on page 3.3-230 is revised as follows:

Table 3.3.2-11: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-Component Type Material Environment Requiring Table 1 Item Notes Function Program 2191 Management Item Heat Exchanger - (Turbine Leakage Copper Air - Indoor Cracking External Surfaces VIII.H.S-454 3.4.1-106 C, 3 Building Reheaters) Tube Boundary Alloy Uncontrolled None Monitoring of 3.3.1-114 Side Components with (External) Mechanical VII.J.AP-144 Greater Components (B.2.3.23)

Than None 15%

Zinc Table 3.3.2-11 on page 3.3-231 is revised as follows:

Table 3.3.2-11: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-Component Type Material Environment Requiring Table 1 Item Notes Function Program 2191 Management Item Heat Leakage Copper Air - Indoor Cracking External Surfaces VIII.H.S-454 3.4.1-106 C, 3 Exchanger - (Unit Heaters) Boundary Alloy Uncontrolled None Monitoring of 3.3.1-114 Tube Side Components with (External) Mechanical VII.J.AP-144 Greater Components (B.2.3.23)

Than None 15%

Zinc

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0a Page 4 of 4 Table 3.3.2-11 on page 3.3-242 is revised as follows:

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

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

E. Consistent with NUREG-2191 material, environment, and aging effect, but a different AMP is credited or NUREG-2191 identifies a plant-specific AMP.

Plant-Specific Notes

1. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) program has been substituted for the Open-Cycle Cooling Water System (B.2.3.11) program and will be used to manage loss of material in carbon steel chiller components exposed to raw water.

2. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) program has been substituted for the Open-Cycle Cooling Water System (B.2.3.11) program and will be used to manage cracking in copper alloy with greater than 15% zinc chiller components exposed to raw water.

3. The surface of the component is not buried, subject to prolonged/cyclic wetting, humidity, concentration of contaminants, or an aggressive environment. As such, no aging effects requiring management are identified.

Enclosure 30b Heat Exchangers Revised To Cite Correct Environments And Aging Management Programs

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 1 of 11 Heat Exchangers Revised To Cite Correct Environments And Aging Management Programs Revise SLRA Tables 3.3.2-6 and 3.3.2-11 to cite correct environments and aging management programs for heat exchangers.

Affected SLRA Sections: Table 3.3.2-6 and Table 3.3.2-11 SLRA Page Numbers: 3.3-134, 3.3-135, 3.3-136, 3.3-153, 3.3-219, 3.3-222, 3.3-223, 3.3-224, 3.3-228, 3.3-229, and 3.3-242 Description of Change:

SLRA Table 3.3.2-6 is updated for Heat Exchanger - (After Cooler) Fins, Heat Exchanger -

(After Cooler) Shell Side Components, Heat Exchanger - (After Cooler) Tube Sheet, and Heat Exchanger - (After Cooler) Tubes. The environment is revised from Condensation (External) to Air - Indoor Uncontrolled (External). The aging management program is revised from External Surfaces Monitoring of Mechanical Components (B.2.3.23) to Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) for Heat Exchanger (After Cooler) Fins with material type Aluminum . Heat Exchanger (After Cooler) Tubes with material type Copper Alloy with Greater Than 15% Zinc is revised to cite none for aging effect requiring management and aging management program. A Plant-Specific note 4 is added to page 3.3-153.

SLRA Table 3.3.2-11 is updated for Heat Exchanger - (Area Air Cooling Units) Tube Side Components, Heat Exchanger - (HPCI/RHR/CS Room Air Cooling Unit) Tube Side Components, Heat Exchanger - (HPCI/RHR/CS Room Air Cooling Unit) Tubes, Heat Exchanger

- (Reactor Building Main Supply HVAC Unit, Cooling Coil) Tube Side Components, and Heat Exchanger - (Steam Chase Supply Cooling Coil) Tube Side Components.The aging management program is revised from External Surfaces Monitoring of Mechanical Components (B.2.3.23) to None for all, except for Heat Exchanger (HPCI/RHR/CS Room Air Cooling Unit)

Tubes with a material type of stainless steel. Heat Exchanger (HPCI/RHR/CS Room Air Cooling Unit) Tubes with a material type of stainless steel was revised to cite Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24). for the aging management program. A Plant Specific Note 3 is applied to the changes made in this table.

This plant-specific note is incorporated into the SLRA by enclosure 30a and shown in bold, black print in this enclosure.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 2 of 11 Heat Exchanger - (After Cooler) sub-components in SLRA Table 3.3.2-6 on pages 3.3-134 to 3.3-136 are revised as follows:

Table 3.3.26: Emergency Diesel Generators - Summary of Aging Management Evaluation Aging Effect Aging Component Intended Material Environment Requiring Management NUREG-2191 Table 1 Item Notes Type Function Management Program Heat Heat Aluminum Condensation Reduction of External VII.I.A716VII.F4.A- 3.3.11513.3.1- C Exchanger Transfer (External)Air Heat Transfer Surfaces 419 096a (After - Indoor Monitoring of Cooler) Fins Uncontrolled Mechanical (External) Components (B.2.3.23)

Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Heat Heat Aluminum Condensation Loss of OneTime VII.F4.A771a 3.3.1242 A Exchanger Transfer (External)Air Material Inspection (After - Indoor (B.2.3.20)

Cooler) Fins Uncontrolled (External)

Heat Heat Aluminum Condensation Cracking OneTime VII.F4.A788a 3.3.1254 A Exchanger Transfer (External)Air Inspection (After - Indoor (B.2.3.20)

Cooler) Fins Uncontrolled (External)

Heat Pressure Carbon Air Indoor Loss of External VII.I.AP41 3.3.1080 A Exchanger Boundary Steel Uncontrolled Material Surfaces (After (External) Monitoring of Cooler) Mechanical Shell Side Components Components (B.2.3.23)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 3 of 11 Heat Pressure Carbon Condensation Loss of Inspection of VII.H2.A26VII.F4.A- 3.3.10553.3.1- C Exchanger Boundary Steel (External)Air Material Internal 778 249 (After - Indoor Surfaces in Cooler) Uncontrolled Miscellaneous Shell Side (External) Piping and Components Ducting Components (B.2.3.24)

Heat Pressure Carbon Closed Cycle Loss of Closed VII.H2.AP202 3.3.1045 A Exchanger Boundary Steel Cooling Material Treated Water (After Water Systems Cooler) (Internal) (B.2.3.12)

Tube Sheet Heat Pressure Carbon Condensation Loss of Inspection of VII.H2.A26VII.F4.A- 3.3.10553.3.1- C Exchanger Boundary Steel (External)Air Material Internal 778 249 (After - Indoor Surfaces in Cooler) Uncontrolled Miscellaneous Tube Sheet (External) Piping and Ducting Components (B.2.3.24)

Heat Pressure Carbon Air Indoor Loss of External VII.I.AP41 3.3.1080 A Exchanger Boundary Steel Uncontrolled Material Surfaces (After (External) Monitoring of Cooler) Mechanical Tube Side Components Components (B.2.3.23)

Heat Pressure Carbon Closed Cycle Loss of Closed VII.H2.AP202 3.3.1045 C Exchanger Boundary Steel Cooling Material Treated Water (After Water Systems Cooler) (Internal) (B.2.3.12)

Tube Side Components Heat Heat Copper Closed Cycle Reduction of Closed VII.C2.AP205 3.3.1050 A Exchanger Transfer Alloy with Cooling Heat Transfer Treated Water (After Greater Water Systems Cooler) Than (Internal) (B.2.3.12)

Tubes 15% Zinc

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 4 of 11 Heat Heat Copper Condensation Reduction of External VII.I.A716VII.J.AP- 3.3.11513.3.1- AC, 4 Exchanger Transfer Alloy with (External)Air Heat Surfaces 144 114 (After Greater - Indoor TransferNone Monitoring of Cooler) Than Uncontrolled Mechanical Tubes 15% Zinc (External) Components (B.2.3.23)

None Heat Pressure Copper Closed Cycle Cracking Closed VII.C2.A473a 3.3.1160 A Exchanger Boundary Alloy with Cooling Treated Water (After Greater Water Systems Cooler) Than (Internal) (B.2.3.12)

Tubes 15% Zinc Heat Pressure Copper Closed Cycle Loss of Closed VII.H2.AP199 3.3.1046 C Exchanger Boundary Alloy with Cooling Material Treated Water (After Greater Water Systems Cooler) Than (Internal) (B.2.3.12)

Tubes 15% Zinc Heat Pressure Copper Closed Cycle Loss of Selective VII.H2.AP43 3.3.1072 C Exchanger Boundary Alloy with Cooling Material Leaching (After Greater Water (B.2.3.21)

Cooler) Than (Internal)

Tubes 15% Zinc Heat Pressure Copper Condensation CrackingNone External VIII.H.S454VII.J.AP- 3.4.11063.3.1- C, 4 Exchanger Boundary Alloy with (External)Air Surfaces 144 114 (After Greater - Indoor Monitoring of Cooler) Than Uncontrolled Mechanical Tubes 15% Zinc (External) Components (B.2.3.23)

None

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 5 of 11 SLRA Plant-Specific Notes on page 3.3-153 are revised as follows:

PlantSpecific Notes

1. The OpenCycle Cooling Water System (B.2.3.11) AMP has been substituted for the FlowAccelerated Corrosion (B.2.3.9) AMP for wall thinning in raw water environments.

2. The OpenCycle Cooling Water System (B.2.3.11) AMP is being substituted for the Internal Coatings/Linings for InScope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.28) program to manage loss of material in the base metal of the gray cast iron (with internal coating) heat exchanger tube side components exposed to raw water.

3. The Selective Leaching (B.2.3.21) AMP is being substituted for the Internal Coatings/Linings for InScope Piping, Piping Components, Heat Exchangers, and Tanks (B.2.3.28) program to manage selective leaching in the base metal of the gray cast iron (with internal coating) heat exchanger tube side components

4. The surface of the component is not buried, subject to prolonged/cyclic wetting, humidity, concentration of contaminants, or an aggressive environment. As such, no aging effects requiring management are identified.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 30b Page 6 of 11 The Heat Exchanger Tube Side Components item in an air - indoor uncontrolled environment in SLRA Table 3.3.2-11 on pages 3.3-219 is revised as follows:

Table 3.3.211: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NURE-2191 Material Environment Requiring Management Table 1 Item Notes Type Function Item Management Program Heat Leakage Copper Air Indoor CrackingNone External VIII.H.S454VII. 3.4.11063.3.1- AC, 3 Exchanger Boundary Alloy Uncontrolled Surfaces J.AP-144 114 (Area Air with (External) Monitoring of Cooling Greater Mechanical Units) Tube Than Components Side 15% (B.2.3.23)

Components Zinc None

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 7 of 11 The Heat Exchanger Tube Side Components item in an air - indoor uncontrolled environment in SLRA Table 3.3.2-11 on pages 3.3-222 is revised as follows:

Table 3.3.211: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NURE-2191 Material Environment Requiring Management Table 1 Item Notes Type Function Item Management Program Heat Pressure Copper Air Indoor CrackingNone External Surfaces VIII.H.S454VII. 3.4.11063.3.1- C, 3 Exchanger Boundary Alloy Uncontrolled Monitoring of J.AP-144 114 (HPCI/RHR/ with (External) Mechanical CS Room Air Greater Components Cooling Unit) Than (B.2.3.23) None Tube Side 15%

Components Zinc

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 8 of 11 The Heat Exchanger Tubes item in an air - indoor uncontrolled environment in SLRA Table 3.3.2-11 on pages 3.3-223 is revised as follows:

Table 3.3.211: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NURE-2191 Material Environment Requiring Management Table 1 Item Notes Type Function Item Management Program Heat Heat Copper Air Indoor Reduction of External V.E.E424VII.J. 3.2.10813.3.1 AC, 3 Exchanger Transfer Alloy with Uncontrolled Heat Surfaces AP-144 -114 (HPCI/RHR/ Greater (External) TransferNone Monitoring of CS Room Air Than 15% Mechanical Cooling Unit) Zinc Components Tubes (B.2.3.23)

None

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 9 of 11 The Heat Exchanger Tubes items in an air - indoor uncontrolled environment in SLRA Table 3.3.2-11 on pages 3.3-224 are revised as follows:

Table 3.3.211: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-2191 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Heat Heat Stainless Air Indoor Reduction of External Surfaces V.E.E424 3.2.1081 A Exchanger Transfer Steel Uncontrolled Heat Transfer Monitoring of VII.F3.A-419 3.3.1-096a (HPCI/RHR/ (External) Mechanical CS Room Air Components Cooling Unit) (B.2.3.23)

Tubes Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Heat Pressure Copper Air Indoor CrackingNone External Surfaces VIII.H.S454 3.4.1106 C, 3 Exchanger Boundary Alloy with Uncontrolled Monitoring of VII.J.AP-144 3.3.1-114 (HPCI/RHR/ Greater (External) Mechanical CS Room Air Than 15% Components Cooling Unit) Zinc (B.2.3.23) None Tubes

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 10 of 11 The Heat Exchanger Tube Side Components item in an air - indoor uncontrolled environment in SLRA Table 3.3.2-11 on pages 3.3-228 is revised as follows:

Table 3.3.211: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Component Intended Aging Management NURE-2191 Table 1 Material Environment Requiring Notes Type Function Program Item Item Management Heat Leakage Copper Air Indoor CrackingNone External Surfaces VIII.H.S454 3.4.1106 C, 3 Exchanger Boundary Alloy Uncontrolled Monitoring of VII.J.AP-144 3.3.1-114 (Reactor with (External) Mechanical Building Greater Components Main Supply Than (B.2.3.23) None HVAC Unit, 15%

Cooling Coil) Zinc Tube Side Components

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 0b Page 11 of 11 The Heat Exchanger Tube Side Components item in an air - indoor uncontrolled environment in SLRA Table 3.3.2-11 on pages 3.3-229 is revised as follows:

Table 3.3.211: Heating and Ventilation - Summary of Aging Management Evaluation Aging Effect Component Intended Aging Management NURE-2191 Table 1 Material Environment Requiring Notes Type Function Program Item Item Management Heat Leakage Copper Air Indoor CrackingNone External Surfaces VIII.H.S454 3.4.1106 C, 3 Exchanger Boundary Alloy Uncontrolled Monitoring of VII.J.AP-144 3.3.1-114 (Steam with (External) Mechanical Chase Greater Components Supply Than (B.2.3.23) None Cooling Coil) 15%

Tube Side Zinc Components The Plant-Specific Notes for Table 3.3.2-11 on page 3.3-242 are revised to add the following additional note:

Plant-Specific Notes

1. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) program has been substituted for the Open-Cycle Cooling Water System (B.2.3.11) program and will be used to manage loss of material in carbon steel chiller components exposed to raw water.

2. The Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24) program has been substituted for the Open-Cycle Cooling Water System (B.2.3.11) program and will be used to manage cracking in copper alloy with greater than 15% zinc chiller components exposed to raw water.

3. The surface of the component is not buried, subject to prolonged/cyclic wetting, humidity, concentration of contaminants, or an aggressive environment. As such, no aging effects requiring management are identified.

Enclosure 31a Addition of Joint and Penetration Seals Commodity Group

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1a Page 1 of 4 Addition of Joint and Penetration Seals Commodity Group Addition of joint and penetration seals Commodity Group Affected SLRA Sections: Tables 2.4-13, 2.4-14, 2.4-15, 3.5.2-13, 3.5.2-14, and 3.5.2-15 SLRA Page Numbers: 2.4-30, 2.4-32, 2.4-34, 3.5-126, 3.5-130, 3.5-137 Description of Change:

Seismic joint fillers are in scope of SLR and are included within the generic commodity group joint and penetration seals. The structures monitoring program inspects these in the existing implementing procedure. SLRA structures associated with Tables 2.4-13, 2.4-14, 2.4-15, 3.5.2-13, 3.5.2-14, and 3.5.2-15 include seismic gaps in their design but the Tables list seals using different terminology than joints.

This change adds the commodity group joint and penetration seals to Tables 2.4-13, 2.4-14, 2.4-15, 3.5.2-13, 3.5.2-14, and 3.5.2-15 to reflect the presence of seismic joint fillers associated with the Plant Control and Cable Spreading Structure, Radioactive Waste Building, and Reactor Building.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1a Page 2 of 4 SLRA Table 2.4-13 on page 2.4-30 is revised to add the following between the Control Room Seals and Masonry (Block) Walls Component Types:

Table 2.4-13 Plant Control and Cable Spreading Structure Components Subject to Aging Management Review Component Type Component Intended Function(s)

Joint and Penetration Seals Shelter, Protection SLRA Table 2.4-14 on page 2.4-32 is revised to add the following between the Concrete:

Interior Walls, Ceiling, and Floor and Masonry (Block) Walls Component Types:

Table 2.4-14 Radioactive Waste Building Components Subject to Aging Management Review Component Type Component Intended Function(s)

Joint and Penetration Seals Shelter, Protection SLRA Table 2.4-15 on page 2.4-34 is revised to add the following between the Fuel Storage Racks: Neutron Absorbing Sheets and Masonry (Block) Walls Component Types:

Table 2.4-15 Reactor Building Components Subject to Aging Management Review Component Type Component Intended Function(s)

Joint and Penetration Seals Shelter, Protection

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1a Page 3 of 4 SLRA Table 3.5.2-13 on page 3.5-126 is revised to add the following between the Control Room Seals and Masonry (Block) Walls Component Types:

Table 3.5.2-13: Plant Control and Cable Spreading Structure - Summary of Aging Management Evaluation Intended Aging Effect Aging Management NUREG-2191 Table 1 Component Type Function Material Environment Requiring Program Item Notes Item Management Joint and Shelter, Elastomer Air-Indoor Loss of Sealing Structures III.A6.TP-7 3.5.1-072 A Penetration Seals Protection Uncontrolled Monitoring Air - Outdoor (B.2.3.33)

SLRA Table 3.5.2-14 on page 3.5-130 is revised to add the following between the Concrete: Interior Walls, Ceiling, and Floor and Masonry (Block) Walls Component Types:

Table 3.5.2-14: Radioactive Waste Building - Summary of Aging Management Evaluation Intended Aging Effect Aging Management NUREG-2191 Table 1 Component Type Function Material Environment Requiring Program Item Notes Item Management Joint and Shelter, Elastomer Air-Indoor Loss of Sealing Structures III.A6.TP-7 3.5.1-072 A Penetration Seals Protection Uncontrolled Monitoring Air - Outdoor (B.2.3.33)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1a Page 4 of 4 SLRA Table 3.5.2-15 on page 3.5-137 is revised to add the following between the Fuel Storage Racks: Neutron Absorbing Sheets and Masonry (Block) Walls Component Types:

Table 3.5.2-15: Reactor Building - Summary of Aging Management Evaluation Intended Aging Effect Aging Management NUREG-2191 Table 1 Component Type Function Material Environment Requiring Program Item Notes Item Management Joint and Shelter, Elastomer Air-Indoor Loss of Sealing Structures III.A6.TP-7 3.5.1-072 A Penetration Seals Protection Uncontrolled Monitoring Air - Outdoor (B.2.3.33)

Enclosure 31b Railroad Bay Roo"ng Supplement

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1b Page 1 of 3 Railroad Bay Roofing Supplement Railroad Bay Roofing Component Type Clarification Affected SLRA Sections: Table 2.4-14 and Table 3.5.2-14 SLRA Page Numbers: 2.4-32 and 3.5-131 Description of Change:

Tables 2.4-14 and 3.5.2-14 are revised to clarify that the Roofing Railroad Bay is actually Seal/Moisture Barrier (Railroad Bay Roofing).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1b Page 2 of 3 SLRA Table 2.4-14 on page 2.4-32 is revised as follows:

Component Type Component Intended Function(s)

Roofing Railroad Bay Seal/Moisture Barrier Pressure Boundary (Railroad Bay Roofing) Shelter, Protection

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1b Page 3 of 3 SLRA Table 3.5.2-14 on page 3.5-131 is revised as follows:

Table 3.5.2-14: Radioactive Waste Building - Summary of Aging Management Evaluation Aging Effect Aging Component Intended Requiring Management NUREG-2191 Table 1 Type Function Material Environment Item Notes Management Program Item Roofing Railroad Pressure Elastomer Air - Outdoor Loss of Sealing Structures III.A6.TP-7 3.5.1-072 A Bay Boundary Monitoring Seal/Moisture Shelter, (B.2.3.33)

Barrier (Railroad Protection Bay Roofing)

Enclosure 31c Add Acceptance Criteria for Element 6

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1c Page 1 of 3 Add Acceptance Criteria for Element 6 Structures Monitoring Acceptance Criteria Enhancement Affected SLRA Sections: Table A-3 Commitment 36 and Section B.2.3.33 SLRA Page Numbers: A-92 and B-240 Description of Change:

Revise Table A-3, Commitment 36 and the enhancements to element 6 in SLRA Section B.2.3.33 (Structures Monitoring) to include an enhancement to the implementing procedure to add the acceptance criteria for the following components and commodities:

• Expansion plugs

• Fuel Storage Racks (New Fuel)

• Manhole covers, supports

• Supports

• Concrete Diesel Fuel Oil Storage Tank Deadmen

• Vibration Isolation Elements

• Electrical Enclosures

• RPV to Drywell Refueling Seal

• Exterior Surfaces of Roofing This will provide consistency with GALL-SLR Section XI.S6, Element 6.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1c Page 2 of 3 SLRA Table A-3, commitment number 36 on Page A-92 is revised as follows:

No. Aging Management NUREG-2191 Commitment Implementation Schedule Program or Activity Section (Section) 36 j) Revise the implementing procedure to include acceptance criteria for inspections of the following components and commodities:

Expansion plugs Fuel Storage Racks (New Fuel)

Manhole covers, supports Supports Concrete Diesel Fuel Oil Storage Tank Deadmen Vibration Isolation Elements Electrical Enclosures RPV to Drywell Refueling Seal Exterior Surfaces of Roofing

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1c Page 3 of 3 SLRA Section B.2.3.33 on page B-240 is revised as follows:

Element Affected Enhancement

6. Acceptance Criteria Revise the implementing procedure to include acceptance criteria for inspections of the following components and commodities:

Expansion plugs Fuel Storage Racks (New Fuel)

Manhole covers, supports Supports Concrete Diesel Fuel Oil Storage Tank Deadmen Vibration Isolation Elements Electrical Enclosures RPV to Drywell Refueling Seal Exterior Surfaces of Roofing

Enclosure 31d Enhancement Consistency

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1d Page 1 of 7 Enhancement Consistency Ensuring Consistency between Structures Monitoring and Water-Control Structures Enhancements Affected SLRA Sections: Table A-3 Commitments 36 and 37, Sections B.2.3.33 and B.2.3.34 SLRA Page Numbers: A-91, A-92, A-93, B-239, B-240, and B-244 Description of Change:

Commitments 36 and 37 and the associated enhancements are being revised to ensure consistency between the Structures Monitoring AMP and the Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP. The first added commitment in this supplement is given letter k because commitment j is being proposed for addition in enclosure 31c of this supplement.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1d Page 2 of 7 SLRA Table A-3, commitment number 36 on pages A-91 and A-92 is revised to correct a typographical error in commitment 36c and add two additional commitments (36k and 36l) as follows:

Aging Management NUREG-Implementation No. Program or 2191 Commitment Schedule Activity Section (Section) 36 Structures XI.S6 c) Revise the implementing procedure to include provisions for more frequent inspections in areas where significant signs of degradation Monitoring are projected or observed to provide reasonable assurance than (A.2.3.33) that there is no loss of intended function between inspections.

k) Ensure that the implementing procedure states that visual inspections of inaccessible concrete for evidence of leaching of calcium hydroxide and carbonation are performed if the area becomes accessible or if inspections in an accessible area identifies a condition that would be a leading indicator for the inaccessible area.

l) Include trending of quantitative measurements and qualitative information for findings exceeding the acceptance criteria for all applicable parameters monitored or trended.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1d Page 3 of 7 SLRA Table A-3, commitment number 37 on page A-93 (exclusive of commitment 37a, which is not changed) is revised as follows:

Aging Management NUREG-Implementation No. Program or 2191 Commitment Schedule Activity Section (Section) 37 Inspection of XI.S7 b) State that further evaluation of evidence of groundwater infiltration or Water-Control through-concrete leakage may also include destructive testing of Structures affected concrete to validate existing concrete properties, including Associated with concrete pH levels, and that assessments may include analysis of the Nuclear Power leakage pH, along with mineral, chloride, sulfate, and iron content in Plants (A.2.2.34) the leakage water if leakage volumes allow.

Revise the implementing procedure to include evidence of water in-leakage as a finding requiring further evaluation. This may include engineering evaluation, more frequent inspections, or destructive testing of affected concrete to validate existing concrete properties, including concrete pH levels. When leakage volumes allow, assessment may include analysis of the leakage pH, along with mineral, chloride, sulfate, and iron content in the water.

c) Ensure that the implementing procedure states that visual inspections of inaccessible concrete for evidence of leaching of calcium hydroxide and carbonation are performed if the area becomes accessible or if inspections in an accessible area identifies a condition that would be a leading indicator for the inaccessible area.

d) Include qualification requirements for both inspection and evaluation personnel that is in accordance with ACI 349.3R.

e) Include trending of quantitative measurements and qualitative information for findings exceeding the acceptance criteria for all applicable parameters monitored or trended.

f) Revise the implementing procedure to include that if any projected inspection results will not meet acceptance criteria prior to the

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1d Page 4 of 7 next scheduled inspection, inspection frequencies are adjusted as determined by the CAP.

g) Revise the implementing procedure to include acceptance criteria for concrete surfaces based on the second-tier evaluation criteria provided in ACI 349.3R-02.

h) Revise the implementing procedure to include monitoring and trending of leakage volumes and chemistry for signs of concrete or steel reinforcement degradation if active through-wall leakage or groundwater infiltration is identified.

i) Revise the implementing procedure to include provisions for more frequent inspections in areas where significant signs of degradation are projected or observed to provide reasonable assurance that there is no loss of intended function between inspections.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1d Page 5 of 7 SLRA Section B.2.3.33 on pages B-239 and B-240 is revised to correct the one typographical error and add two new enhancements (to be inserted between the last enhancement for element affected 4 and the enhancement for element affected 6). The revised and added enhancements are as follows:

Element Affected Enhancement

4. Detection of Aging Effects Revise the implementing procedure to include provisions for more frequent inspections in areas where significant signs of degradation are projected or observed to provide reasonable assurance thanthat there is no loss of intended function between inspections.

4. Detection of Aging Effects Revise the implementing procedure to state that visual inspections of inaccessible concrete for evidence of leaching of calcium hydroxide and carbonation are performed if the area becomes accessible or if inspections in an accessible area identifies a condition that would be a leading indicator for the inaccessible area.

5. Monitoring and Trending Revise the implementing procedure to include trending of quantitative measurements and qualitative information for findings exceeding the acceptance criteria for all applicable parameters monitored or trended.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1d Page 6 of 7 SLRA Section B.2.3.34 on page B-244 is revised to add four new enhancements and revise one of the existing enhancements in the SLRA. The new enhancements for element affected 3 and 4 will be inserted between the existing element affected 2 enhancement and the first element affected 4 enhancement (which is the enhancement being revised in this supplement as well).

The other two enhancements will be appended to the end of the enhancement table for this SLRA section. The added and revised enhancements are as follows:

Element Affected Enhancement

3. Parameters Monitored or Revise the implementing procedure to include monitoring Inspected and trending of leakage volumes and chemistry for signs of concrete or steel reinforcement degradation if active through-wall leakage or groundwater infiltration is identified.

4. Detection of Aging Effects Revise the implementing procedure to include provisions for more frequent inspections in areas where significant signs of degradation are projected or observed to provide reasonable assurance that there is no loss of intended function between inspections.

4. Detection of Aging Effects Enhance the implementing procedure to state that further evaluation of evidence of groundwater infiltration or through-concrete leakage may also include destructive testing of affected concrete to validate existing concrete properties, including concrete pH levels, and that assessments may include analysis of the leakage pH, along with mineral, chloride, sulfate, and iron content in the leakage water if leakage volumes allow.

Revise the implementing procedure to include evidence of water in-leakage as a finding requiring further evaluation.

This may include engineering evaluation, more frequent inspections, or destructive testing of affected concrete to validate existing concrete properties, including concrete pH levels. When leakage volumes allow, assessment may include analysis of the leakage pH, along with mineral, chloride, sulfate, and iron content in the water.

4. Detection of Aging Effects Enhance the implementing procedure to state that visual inspections of inaccessible concrete for evidence of leaching of calcium hydroxide and carbonation are performed, when exposed if the area becomes accessible or if inspections in an accessible area identifies a condition that would be a leading indicator for the inaccessible area.

6. Acceptance Criteria Revise the implementing procedure to include acceptance criteria for concrete surfaces based on the second-tier evaluation criteria provided in ACI 349.3R-02.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1d Page 7 of 7

7. Corrective Actions Revise the implementing procedure to include that if any projected inspection results will not meet acceptance criteria prior to the next scheduled inspection, inspection frequencies are adjusted as determined by the MNGP CAP.

Enclosure 31e Provide Clari"cation for Mislabeled Enhancement Element

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1e Page 1 of 3 Provide Clarification for Mislabeled Enhancement Element Revise SLRA Section B.2.3.33 to correct the mislabeled enhancement element affected from 4. Detection of Aging Effects to 1. Scope of Program.

Affected SLRA Sections: Section B.2.3.33 SLRA Page Numbers: Pages B-239 and B-240 Description of Change:

Section B.2.3.33 enhancement Revise the implementing procedure to explicitly include inspection of the following components and commodities: incorrectly labels the element affected as 4. Detection of Aging Effects. This enhancement is being revised to correct the element affected from 4. Detection of Aging Effects to 1. Scope of Program. This change will provide consistency with NUREG-2191, Vol. 2.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1e Page 2 of 3 SLRA Section B.2.3.33 on page B-239 is revised as follows:

Element Affected Enhancement

1. Scope of Program Revise the implementing procedure to explicitly include inspection of the following components and commodities:

Expansion plugs Fuel Storage Racks (New Fuel)

Manhole covers, supports Supports Concrete Diesel Fuel Oil Storage Tank Deadmen Vibration Isolation Elements Electrical Enclosures RPV to Drywell Refueling Seal Exterior Surfaces of Roofing

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1e Page 3 of 3 SLRA Section B.2.3.33 on page B-240 is revised as follows:

Element Affected Enhancement

4. Detection of Aging Effects Revise the implementing procedure to explicitly include inspection of the following components and commodities:

Expansion plugs Fuel Storage Racks (New Fuel)

Manhole covers, supports Supports Concrete Diesel Fuel Oil Storage Tank Deadmen Vibration Isolation Elements Electrical Enclosures RPV to Drywell Refueling Seal Exterior Surfaces of Roofing

Enclosure 31f Clari"cation of 115/345 kV Substation Control House

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1f Page 1 of 8 Clarification of 115/345 kV Substation Control House Revise SLRA to Clarify Identity of 115/345 kV Substation Control House.

Affected SLRA Sections: 2.4.10, Table 2.4-10, Table 3.5-1, Table 3.5.2-10, and B.2.3.39 SLRA Page Numbers: 2.4-23, 2.4-24, 3.5-58, 3.5-114, 3.5-115, and B-270.

Description of Change:

Throughout the SLRA, the 115/345 kV Substation Control House is not denoted consistently.

The SLRA is revised for clarity, changing any reference of this control house to 115/345 kV Substation Control House.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1f Page 2 of 8 SLRA Section 2.4.10, Page 2.4-23 is revised as follows:

2.4.10 Miscellaneous Station Blackout Yard Structures Description The miscellaneous SBO yard structures are those yard structures that provide support for equipment relied upon for recovery from a station blackout. These structures are listed below:

The foundations and transformer structures for 1R, 2R, 1AR and 2RS Transformers The 115/345 kV Substation Control House The towers/foundations for the 1N2, 1N6, 5N5, 5N7, 8N4, 8N7, 8N10, and 8N11 breakers The towers/foundations for the bus bars between the 2RS transformer and the 8N4, 8N7, 8N10, and 8N11 breakers, this includes the towers/foundations for the 3N4 breaker, 3N5 fused disconnect and the towers/foundations to the 1ARS motor operated disconnect (MOD).

The towers/foundations for the bus bars for the 5N5 and 5N7 breakers. This includes the west four rows of columns and the beams that connect them together.

The Trenwa trenches connecting the control house to the 115 kV ring bus.

The Trenwa trenches connecting the control house to the 345 kV breaker-and-a-half bus.

The electrical duct bank from the 1N2 breaker to the 1AR transformer.

The tower/foundation for the 115 kV bus 1 potential transformer.

The three 115 kV transmission towers along the west Owner Control Area (OCA) fence between the switchyard and the 1R transformer and the first transmission tower northwest of the plant.

The block walls surrounding the 1R and 2R transformers.

The foundation for the CST tanks.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1f Page 3 of 8 SLRA Page 2.4-24, Table 2.4-10 is revised as follows:

Table 2.410 Miscellaneous SBO Yard Structures Components Subject to Aging Management Review Component Type Component Intended Function(s)

Concrete: 345 kV House 115/345 kV Structural Support Substation Control House, Foundations, Trenches, and Duct Banks

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 31f Page 4 of 8 SLRA Table 3.5-1, Item 3.5.1-047 on Page 3.5-58 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Aging Management Further Evaluation Discussion Item Number Component Requiring Program Recommended Management 3.5.1047 Groups 15, Increase in porosity Plantspecific aging Yes (SRPSLR Group 7 and group 8 structures are not 79: concrete and permeability; management program Section 3.5.2.2.2.1.4) applicable to MNGP.

(inaccessible loss of strength due or AMP XI.S6, areas): exterior to leaching of "Structures Monitoring" The Structures Monitoring (B.2.3.33) AMP will above and calcium hydroxide be used to manage increase in porosity and below grade; and carbonation. permeability, loss of strength of the reinforced foundation concrete basemat, foundation, subfoundation, belowgrade exterior concrete, pedestal, walls, slabs (inaccessible areas), diesel fuel oil storage tank deadmen, 345 kV house 115/345 kV Substation Control House, trenches, and duct bank exposed to water flowing in Groups 2, 3, and 9 structures.

Further evaluation is documented in Section 3.5.2.2.2.1, item 4.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 31f Page 5 of 8 SLRA Table 3.5.2-10 on Pages 3.5-114 and 3.5-115 is revised as follows:

Table 3.5.210: Miscellaneous Station Blackout Yard Structures - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Concrete: 345 kV Structural Concrete Air - Outdoor Cracking Structures Monitoring III.A3.TP25 3.5.1054 A House 115/345 kV Support (Reinforced) (B.2.3.33)

Substation Control House, Foundations, Trenches, Duct Bank (Accessible)

Concrete: 345 kV Structural Concrete Air - Outdoor Cracking Structures Monitoring III.A3.TP26 3.5.1066 A House 115/345 kV Support (Reinforced) Loss of (B.2.3.33)

Substation Control Bond House, Foundations, Loss of Trenches, Duct Bank Material (Accessible)

Concrete: 345 kV Structural Concrete Air - Outdoor Cracking Structures Monitoring III.A3.TP23 3.5.1064 A House 115/345 kV Support (Reinforced) Loss of (B.2.3.33)

Substation Control Material House, Foundations, Trenches, Duct Bank (Accessible)

Concrete: 345 kV Structural Concrete Groundwater/Soil Cracking Structures Monitoring III.A3.TP27 3.5.1065 A House 115/345 kV Support (Reinforced) Loss of (B.2.3.33)

Substation Control Bond House, Foundations, Loss of Trenches, Duct Bank Material (Accessible)

Concrete: 345 kV Structural Concrete Water - Flowing Increase in Structures Monitoring III.A3.TP24 3.5.1063 A, 1 House 115/345 kV Support (Reinforced) Porosity and (B.2.3.33)

Substation Control Permeability; House, Foundations, Loss of Trenches, Duct Bank Strength (Accessible)

Concrete: 345 kV Structural Concrete Air Outdoor Cracking Structures Monitoring III.A3.TP204 3.5.1043 A

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 31f Page 6 of 8 House 115/345 kV Support (Reinforced) (B.2.3.33)

Substation Control House, Foundations, Trenches, Duct Bank (Inaccessible)

Concrete: 345 kV Structural Concrete Air - Outdoor Increase in Structures Monitoring III.A3.TP28 3.5.1067 A House 115/345 kV Support (Reinforced) Porosity and (B.2.3.33)

Substation Control Permeability; House, Foundations, Loss of Trenches, Duct Bank Strength (Inaccessible)

Concrete: 345 kV Structural Concrete Groundwater/Soil Cracking Structures Monitoring III.A3.TP30 3.5.1044 A House 115/345 kV Support (Reinforced) Distortion (B.2.3.33)

Substation Control House, Foundations, Trenches, Duct Bank (Inaccessible)

Concrete: 345 kV Structural Concrete Groundwater/Soil Cracking Structures Monitoring III.A3.TP212 3.5.1065 A House 115/345 kV Support (Reinforced) Loss of (B.2.3.33)

Substation Control Bond House, Foundations, Loss of Trenches, Duct Bank Material (Inaccessible)

Concrete: 345 kV Structural Concrete Groundwater/Soil Cracking Structures Monitoring III.A3.TP108 3.5.1042 A House 115/345 kV Support (Reinforced) Loss of (B.2.3.33)

Substation Control Material House, Foundations, Trenches, Duct Bank (Inaccessible)

Concrete: 345 kV Structural Concrete Groundwater/Soil Increase in Structures Monitoring III.A3.TP29 3.5.1067 A House 115/345 kV Support (Reinforced) Porosity and (B.2.3.33)

Substation Control Permeability; House, Foundations, Loss of Trenches, Duct Bank Strength (Inaccessible)

Concrete: 345 kV Structural Concrete Water - Flowing Increase in Structures Monitoring III.A3.TP67 3.5.1047 A, 1 House 115/345 kV Support (Reinforced) Porosity and (B.2.3.33)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 31f Page 7 of 8 Substation Control Permeability House, Foundations, Loss of Trenches, Duct Bank Strength (Inaccessible)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1f Page 8 of 8 SLRA Appendix B, Section B.2.3.39, Page B-270 is revised as follows:

In April 2014, during a transformer CT feedback test, the technician performed a megger test. The megger results identified degraded cables from a breaker in the breaker cabinet to the 345 kV house 115/345 kV Substation Control House. The results were much lower than usual, and a work order was generated to locate the cables via ground penetrating radar, hydro excavating, and replacement of the degraded cables. Replacements were completed in May 2015.

Enclosure 31g Structures Monitoring Program Inspection Frequency

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1g Page 1 of 2 Structures Monitoring Program Inspection Frequency Inspection frequency of inaccessible structures Affected SLRA Sections: B.2.3.33 SLRA Page Numbers: B-238 Description of Change:

Appendix B of the SLRA has been revised to provide information on the inspection frequency for the Structures Monitoring program.

The normally inaccessible areas monitored on an interval that may exceed its five-year interval are high radiation areas such as primary containment, condenser room steam chase, and air ejector room. In general, all structures are monitored on an interval not to exceed 5 years, consistent with GALL-SLR XI.S6. However, if a normally inaccessible area only becomes accessible during an outage in which the 5-year inspection frequency is exceeded, it is reasonable to exceed the 5-year frequency depending on safety significance and the condition of the structure as specified in NRC RG 1.160 consistent with GALL-SLR XI.S6. There is no exception to NUREG-2191 and the Structures Monitoring AMP does not require enhancement.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1g Page 2 of 2 Section B.2.3.33 on page B-238 is revised to add sentences to the end of the second paragraph as follows:

B.2.3.33 Structures Monitoring The Structures Monitoring AMP is an existing AMP based on the requirements of 10 CFR 50.65 (the Maintenance Rule) and NRC RG 1.160 (Reference 1.6.52), and Nuclear Management and Resources Council 9301 (Reference 1.6.53). These documents provide guidance for development of site/fleetspecific programs to monitor the condition of structures and structural components within the scope of the SLR rule, such that there is no loss of structure or structural component intended function.

The MNGP Structures Monitoring AMP consists primarily of periodic visual inspections of plant SCs for evidence of deterioration or degradation, such as described in the American Concrete Institute (ACI) Standards 349.3R (Reference 1.6.54), ACI 201.1R (Reference 1.6.55), and Structural Engineering Institute/American Society of Civil Engineers Standard (SEI/ASCE) 11 (Reference 1.6.56). Quantitative acceptance criteria for concrete inspections are based on ACI 349.3R. In accordance with the guidance in NUREG-2191, Chapter XI.S6, Element 4, the inspection frequency depends on safety significance and the condition of the structure as specified in NRC RG 1.160.

In general, structures are monitored on an interval of 5 years.

Enclosure 31h Line Item 3.3.1-111 Clari"ed

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1h Page 1 of 2 Line Item 3.3.1-111 Clarified SLRA Item 3.3.1-111 is revised from Not Used to Not Applicable and an explanation is provided.

Affected SLRA Sections: Table 3.3-1 SLRA Page Numbers: 3.3-55 Description of Change:

SLRA Table 3.3-1 Item 3.3.1-111 is revised to state not applicable because MNGP new fuel storage racks are made of aluminum and were evaluated under Item 3.5.1-100.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1h Page 2 of 2 SLRA Table 3.3-1 on page 3.3-55 is revised as follows:

Table 3.3-1: Summary of Aging Management Evaluations for the Auxiliary Systems Item Aging Effect / Aging Management Further Evaluation Component Discussion Number Mechanism Program (AMP)/TLAA Recommended 3.3.1111 Steel Loss of AMP XI.S6, "Structures No Not applicable used.

structural material due to Monitoring" steel general, This item applies to new fuel storage exposed to pitting, crevice racks made of steel. MNGPs new fuel air - indoor corrosion storage racks are made of aluminum and uncontrolled evaluated in Item 3.5.1-100.Structural steel is addressed as part of structural items in Section 3.5.

Enclosure 31i Removal of Grouted Penetration Seals

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1i Page 1 of 5 Removal of Grouted Penetration Seals Removal of Grouted Penetration Seals from the Turbine Building and Emergency Diesel Generator Building Aging Management Evaluations.

Affected SLRA Sections: 3.5.2.1.4, 3.5.2.1.17, and Tables 3.5.2-4 and 3.5.2-17 SLRA Page Numbers: 3.5-5, 3.5-6, 3.5-17, 3.5-18, 3.5-91, and 3.5-146 Description of Change:

The Joint and Penetration Seals Component Type in Tables 3.5.2-4 and 3.5.2-17 are deleted.

Grout that is not used for concrete anchors can be treated as a sub-part of the associated concrete wall. Sections 3.5.2.1.4 and 3.5.2.1.17 are revised to remove grout from materials and Reduction in Concrete Anchor Capacity from the aging effects requiring management.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1i Page 2 of 5 SLRA Section 3.5.2.1.4 on Pages 3.5-5 and 3.5-6 is revised as follows:

3.5.2.1.4 Emergency Diesel Generator Building Materials The materials of construction for the DGB structural components are:

Concrete Block Concrete (Reinforced)

Elastomer Grout Steel Environments The DGB structural components are exposed to the following environments:

Air Indoor Uncontrolled Air Outdoor Groundwater/Soil Water Flowing Aging Effects Requiring Management The following aging effects associated with the DGB structural components require management:

Cracking Distortion Increase in Porosity and Permeability Loss of Bond Loss of Material Loss of Preload Loss of Sealing Loss of Strength Reduction In Concrete Anchor Capacity Aging Management Programs The following AMPs manage the aging effects for the DGB structural components:

Masonry Walls (B.2.3.32)

Structures Monitoring (B.2.3.33)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 1i Page 3 of 5 SLRA Section 3.5.2.1.17 on Pages 3.5-17 and 3.5-18 is revised as follows:

3.5.2.1.17 Turbine Building Materials The materials of construction for the Turbine Building structural components are:

Aluminum Concrete Block Concrete (Reinforced)

Elastomer Grout Steel Environments The Turbine Building structural components are exposed to the following environments:

Air Indoor Uncontrolled Air Outdoor Groundwater/Soil Water Flowing Aging Effects Requiring Management The following aging effects associated with the Turbine Building structural components require management:

Cracking Distortion Increase in Porosity and Permeability Loss of Bond Loss of Material Loss of Preload Loss of Sealing Loss of Strength Reduction In Concrete Anchor Capacity Aging Management Programs The following AMPs manage the aging effects for the Turbine Building structural components:

Masonry Walls (B.2.3.32)

Structures Monitoring (B.2.3.33)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 31i Page 4 of 5 SLRA Table 3.5.2-4 on Page 3.5-91 is revised as follows:

Table 3.5.2-4 Emergency Diesel Generator Building Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Joint and Flood Barrier Grout Air - Indoor Reduction in Structures Monitoring III.B2.TP42 3.5.1055 A Penetration Seals Uncontrolled Concrete Anchor (B.2.3.33)

Air - Outdoor Capacity

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 31i Page 5 of 5 SLRA Table 3.5.2-17 on Page 3.5-146 is revised as follows:

Table 3.5.2-17: Turbine Building - Summary of Aging Management Evaluation Aging Effect Component Intended Aging Management NUREG-2191 Table 1 Material Environment Requiring Notes Type Function Program Item Item Management Joint and Flood Barrier Grout Air - Indoor Reduction in Concrete Structures Monitoring III.B2.TP42 3.5.1055 A Penetration HELB Barrier Uncontrolled Anchor Capacity (B.2.3.33)

Seals Air - Outdoor

Enclosure 32 Correct Drawing from SLR-11929 to SLR-119259

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 1 of 2 Correct Drawing from SLR-11929 to SLR-119259 Correct SLRA Section 2.3.4.2 Boundary list on page 2.3-70 to list SLR-119259 instead of SLR-11929.

Affected SLRA Sections: 2.3.4.2 SLRA Page Numbers: 2.3-70 Description of Change:

Revise SLRA CFW System Section 2.3.4.2 Boundary list on page 2.3-70 to correct the listed drawing to SLR-119259 instead of SLR-11929.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 2 Page 2 of 2 SLRA Section 2.3.4.2 on page 2.3-70 is revised as follows:

Boundary The CFW System boundaries are shown on the following SLRBDs:

SLR36034 SLR36035 SLR36036 SLR360372 SLR360373 SLR36037 SLR360381 SLR360382 SLR360383 SLR36038 SLR36039 SLR36041 SLR36044 SLR360471 SLR36241 SLR548174 SLR85509 SLR100320 SLR11929SLR-119259 SLR252182 SLR236609

Enclosure 33a Clarify the Source of the Maximum 7000 Cycles and Clarify Operating Cycles is Equivalent Full Temperature Thermal Cycles

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 1 of 3 Clarify the Source of the Maximum 7000 Cycles and Clarify Operating Cycles is Equivalent Full Temperature Thermal Cycles Clarifying the source of the maximum 7000 cycles and clarifying operating cycles is equivalent full temperature thermal cycles Affected SLRA Sections: 4.3.6, 4.5.5, and A.3.3.5 SLRA Page Numbers: 4.3-15, 4.5-5, and A-51 Description of Change:

Revised to clarify the source of the maximum 7000 cycles and changed the wording from operating cycles to equivalent full temperature thermal cycles to keep the wording consistent with the ANSI B31.1.0, Pressure Piping, 1977 Edition with Addenda through Winter 1978.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 2 of 3 SLRA Section 4.3.6 on page 4.3-15 is revised as follows:

4.3.6 ASME Section III, Class 2 and 3 and ANSI B31.1 TLAA Description A metal component may progressively degrade and lose its structural integrity when it is subjected to fluctuating loads, even at magnitudes less than the design static loads, due to metal fatigue. This mechanism of degradation can occur in flaw free components by developing cracks during service. Implicit fatiguebased maximum allowable stress calculations are performed for piping components designed to USAS / ANSI B31.1 requirements. ASME Section III Code Class 2 and 3 components are designed to requirements that are similar to the guidance in ANSI B31.1.

In addition, process piping that is subject to significant thermal expansion and contraction includes those that penetrate the drywell shell. Typically, these penetrations, which were designed to the ASME Code,Section III, Class B requirements, are a triple flued head design which has a guard pipe between the process piping and the penetration nozzle. The penetration assembly which provides the interface between the exterior of the process piping with the containment liner is typically known as a bellows. This permits the penetration to be vented to the drywell should a rupture of the hot line occur within the penetration. These containment penetration process bellows have been designed for a maximum of 7,000 operating cycles equivalent full temperature thermal cycles. The maximum 7,000 cycles is an ANSI B31.1.0 Pressure Piping, 1977 Edition with Addenda through Winter 1978 and ASME Code requirement for when no stress range reduction is applicable.

SLRA Section 4.5.5 on page 4.5-5 is revised as follows:

4.5.5 Primary Containment Process Penetration Bellows Fatigue Analysis TLAA Description Containment pipe penetrations that are required to accommodate thermal movement have expansion bellows. The bellows are designed for a minimum number of operating equivalent full temperature thermal cycles over the design life of the plant.

Consequently, the primary containment process penetrations bellows cycle basis is a TLAA.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3a Page 3 of 3 SLRA Section A.3.5.5 on page A-51 is revised as follows:

A.3.5.5 Primary Containment Process Penetration Bellows Fatigue Analysis Containment pipe penetrations that are required to accommodate thermal movement have expansion bellows. The bellows are designed for a minimum number of operating equivalent full temperature thermal cycles over the design life of the plant. Consequently, the primary containment process penetrations bellows cycle basis is a TLAA

Enclosure 33b Updated Reference to Bellows Fatigue Analysis

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3b Page 1 of 3 Updated Reference to Bellows Fatigue Analysis Revise Sections A.3.5.5 & A.3.3.6 to provide additional information and correct a typo Affected SLRA Sections: A.3.3.6, A.3.5.5 SLRA Page Numbers: A-47, A-48, A-51 Description of Change:

The cross-reference in Section A.3.5.5 is corrected from Section A.3.3.5 to Section A.3.3.6. Sections A.3.3.6 and A.3.5.5 are updated to add discussion of the bellows.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3b Page 2 of 3 SLRA Section A.3.3.6 on pages A-47 and A-48 is revised to add two sentences as follows:

A.3.3.6 ASME Section III, Class 2 and 3 and ANSI B31.1 The MNGP nonClass 1 Reactor Coolant System (RCS) piping and balanceofplant piping systems within the scope of SLR are designed to the requirements of the ANSI B31.7 and ANSI B31.1 Codes. Piping and components designed in accordance with these Codes are not required to have an explicit analysis of cumulative fatigue usage, but cyclic loading is considered in a simplified manner in the design process. Containment pipe penetrations that are required to accommodate thermal movement have expansion bellows. Transient cycles on the bellows are composed of thermal cycles experienced by the associated system piping. These nonClass 1 piping Codes first require prediction of the overall number of thermal and pressure cycles expected during the lifetime of these components. Then a stress range reduction factor is determined for that number of cycles using a table from the applicable design code. If the total number of cycles is 7,000 or less, the stress range reduction factor is 1.0, which when applied, would not reduce the allowable stress value.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 3b Page 3 of 3 SLRA Section A.3.5.5 on Page A-51 is revised as follows:

A.3.5.5 Primary Containment Process Penetration Bellows Fatigue Analysis Containment pipe penetrations that are required to accommodate thermal movement have expansion bellows. The bellows are designed for a minimum number of operating cycles over the design life of the plant. Consequently, the primary containment process penetrations bellows cycle basis is a TLAA.

This evaluation was performed as part of the ASME Section III, Class 2 and 3 and ANSI B31.1 fatigue evaluation and is described in Section A.3.3.5 A.3.3.6. The limiting pipe penetration expansion bellows was evaluated and was determined to have a thermal cycle count below the 7000 cycle limit with considerable margin.

The containment penetration bellows fatigue design criteria remains valid for the SPEO in accordance with 10 CFR 54.21(c)(1)(i).

Enclosure 34a Revise to Include Trash Rack

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4a Page 1 of 5 Revise to Include Trash Rack Revise SLRA Tables 2.4-9, 3.5-1, and 3.5.2-9 to explicitly include trash rack.

Affected SLRA Sections: Table 2.4-9, Table 3.5-1, and Table 3.5.2-9 SLRA Page Numbers: 2.4-23, 3.5-71, and 3.5-113 Description of Change:

Revise SLRA Tables 2.4-9, 3.5-1, and 3.5.2-9 to explicitly include trash rack. The General Notes associated with Table 3.5.2-9 is revised to include note C.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4a Page 2 of 5 SLRA Table 2.4-9 on page 2.4-23 is revised as follows:

Table 2.49 Intake Structure Components Subject to Aging Management Review Component Type Component Intended Function(s)

Sheet Piles Structural Support Stored Steel Plates, Hatch Covers, and Bin Flood Barrier Wall Structural Bolting Structural Support Trash Rack Filter Structural Support

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4a Page 3 of 5 SLRA Table 3.5-1, Item Number 3.5.1-083, on page 3.5-71 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Further Item Aging Management Discussion Component Requiring Evaluation Number Program Management Recommended 3.5.1082 Structural Loss of AMP XI.S6, No Consistent with NUREG2191.

bolting material due "Structures Monitoring" to general, The Structures Monitoring (B.2.3.33) AMP is pitting, credited with managing loss of material for crevice structural bolting exposed to uncontrolled indoor corrosion air and outdoor air.

3.5.1083 Structural Loss of AMP XI.S7, No Consistent with NUREG2191.

bolting material due "Inspection of Water to general, Control Structures The Inspection of WaterControl Structures pitting, Associated with Nuclear Associated with Nuclear Power Plants (B.2.3.34) crevice Power Plants" or the AMP is credited with managing loss of material for corrosion FERC/US Army Corp of structural bolting and trash rack exposed to Engineers dam inspections outdoor air and water - flowing or standing in the and maintenance programs INS.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 34a Page 4 of 5 SLRA Table 3.5.2-9 and General Notes on page 3.5-113 is revised as follows:

Table 3.5.29: Intake Structures - Summary of Aging Management Evaluation Intended Aging Effect Aging Management NUREG-2191 Table 1 Component Type Function Material Environment Requiring Program Item Item Notes Management Miscellaneous Structural Steel Air - Indoor Loss of Material Structures Monitoring III.B5.TP-43 3.5.1-092 A Structural Support Uncontrolled (B.2.3.33)

Components Miscellaneous Flood Steel Air - Indoor Loss of Material Structures Monitoring III.A3.TP-302 3.5.1-077 A Structural Barrier Uncontrolled (B.2.3.33)

Components Air - Outdoor Sheet Piles Structural Steel Groundwater/ Loss of Material Structures Monitoring III.A3.TP-219 3.5.1-079 A Support Soil (B.2.3.33)

Stored Steel Flood Steel Air - Indoor Loss of Material Structures Monitoring III.A3.TP-302 3.5.1-077 A Plates, Hatch Barrier Uncontrolled (B.2.3.33)

Covers, and Bin Air - Outdoor Wall Structural Bolting Structural Steel Air - Indoor Loss of Material Inspection of Water- III.A6.TP-221 3.5.1-083 A Support Uncontrolled Control Structures Air - Outdoor Associated with Nuclear Power Plants (B.2.3.34)

Structural Bolting Structural Steel Air - Indoor Loss of Preload Structures Monitoring III.A6.TP-261 3.5.1-088 A Support Uncontrolled (B.2.3.33)

Air - Outdoor Trash Rack Filter Steel Air - Outdoor Loss of Material Inspection of III.A6.TP-221 3.5.1-083 C Structural Water - Water-Control Support Flowing Structures Associated with Nuclear Power Plants (B.2.3.34)

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 34a Page 5 of 5 General Notes A. Consistent with component, material, environment, aging effect, and AMP listed for NUREG-2191 line item. AMP is consistent with NUREG-2191 AMP description.

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

Enclosure 34b Item 3.5.1-079 Piles/Plates Clari"cation

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4b Page1 of 2 Item 3.5.1-079 Piles/Plates Clarification Clarifying Which Structures Include Steel Plates and Steel Piles Affected SLRA Sections: Table 3.5-1, Item 3.5.1-079 SLRA Page Numbers: 3.5-69 Description of Change:

The Item 3.5.1-079 further evaluation discussion is clarified to specify which structures include steel plates and which structures include steel piles.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4b Page2 of 2 SLRA Table 3.5-1, Item 3.5.1-079 on page 3.5-69 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Requiring Aging Management Further Evaluation Number Component Program Recommended Discussion Management 3.5.1-079 Steel components: Loss of material due to AMP XI.S6, No Consistent with NUREG-2191.

piles corrosion "Structures Monitoring" The Structures Monitoring (B.2.3.33) AMP is credited with managing loss of material of steel plates in the HPCI building, INS, and Underground Duct Bank and steel piles in the INS. HPCI building, INS, and Underground Duct Bank.

Enclosure 34c Revise Table 3.5.2-9 to Cite Correct NUREG-2191 Item

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4c Page 1 of 2 Revise Table 3.5.2-9 to Cite Correct NUREG-2191 Item Revise SLRA Table 3.5.2-9 Table 1 Item 3.5.1-096 to cite III.A6.T-34 for NUREG 2191 Item.

Affected SLRA Sections: Table 3.5.2-9 SLRA Page Numbers: 3.5-110 Description of Change:

SLRA Table 3.5.2-9 Table 1 Item 3.5.1-096 is revised to change the NUREG-2191 Item for the component, Concrete Exterior Wall and Roof (Accessible) from III.A6.TP-34 to III.A6.T-34.

This is a typo and the change is supported by NUREG-2191, Vol. 1.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4c Page 2 of 2 SLRA Table 3.5.2-9 on page 3.5-110 is revised as follows:

Table 3.5.29: Intake Structures - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG2191 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Concrete: Flood Barrier Concrete Air - Indoor Cracking Inspection of III.A6.TP34III. 3.5.1096 A, 1 Exterior Missile Barrier (Reinforced) Uncontrolled WaterControl A6.T-34 Walls and Shelter, Air - Outdoor Structures Roof Protection Water Flowing Associated with (Accessible) Structural Nuclear Power Support Plants (B.2.3.34)

Enclosure 34d Supplement for INS Structural Steel and Structural Bolting

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4d Page 1 of 3 Supplement for INS Structural Steel and Structural Bolting Clarifying the SLRA to specify that structural steel and structural bolting is included.

Affected SLRA Sections: A.2.2.34 and B.2.3.34 SLRA Page Numbers: A-30 and B-243 Description of Change:

In conjunction with the changes made to the SLRA by Enclosure 34a of this supplement, sections A.2.2.34 and B.2.3.34 are revised to add additional detail that states that structural steel and structural bolting associated with the INS is included in the Inspection of Water-Control Structures Associated with Nuclear Power Plants program.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4d Page 2 of 3 SLRA Section A.2.2.34 on page A-30 is revised as follows:

Inspection of Water-Control Structures Associated with Nuclear Power Plants The MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP is an existing AMP that is currently implemented as part of the MNGP Structures Monitoring Program.

The MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP was evaluated as a portion of the MNGP Systems and Structures Monitoring AMP in the initial license renewal application (LRA). The MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP is evaluated separately in the SLRA, and it is compared to the NUREG-2191,Section XI.S7 program. This condition monitoring AMP addresses age-related deterioration, degradation due to environmental conditions, and the effects of natural phenomena that may affect water-control structures.

The MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP consists of inspection and surveillance of water control structures, including associated structural steel and structural bolting. The only structure within the scope of the MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP is the INS. Parameters monitored are in accordance with RG 1.127 and quantitative measurements are recorded for findings that exceed the acceptance criteria for applicable parameters monitored or inspected. Inspections occur at least once every 5 years. Evaluation of ground water chemistry is performed under the scope of the MNGP Structures Monitoring AMP (Section A.2.2.33).

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 4d Page 3 of 3 SLRA Section B.2.3.34 on page B-243 is revised as follows:

B.2.3.34 Inspection of Water-Control Structures Associated with Nuclear Power Plants The MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP is an existing AMP that is currently implemented as part of the MNGP Structures Monitoring Program. The MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP was evaluated as a portion of the MNGP Systems and Structures Monitoring AMP in the initial LRA. The MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP is evaluated separately in the SLRA, and it is compared to the NUREG-2191,Section XI.S7 program.

This condition monitoring AMP addresses age-related deterioration, degradation due to environmental conditions, and the effects of natural phenomena that may affect water-control structures. The program consists of inspection and surveillance of water control structures, including associated structural steel and structural bolting. The only structure within the scope of the MNGP Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP is the INS. Parameters monitored are in accordance with RG 1.127 and quantitative measurements are recorded for findings that exceed the acceptance criteria for applicable parameters monitored or inspected.

Inspections occur at least once every five years. Evaluation of ground water chemistry is performed under the scope of the MNGP Structures Monitoring AMP periodically to assure the groundwater remains non-aggressive.

Enclosure 35a Concrete Aging Management Review-Groundwater/Soil

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5a Page 1 of 2 Concrete Aging Management Review-Groundwater/Soil Correction of omitted line item from SLRA Table 3.5.2-11 Affected SLRA Sections: Table 3.5.2-11 SLRA Page Numbers: 3.5-117 Description of Change:

Table 3.5.2-11: Off-Gas Stack-Summary of Aging Management Evaluation is revised to reflect that concrete in a groundwater/soil environment is inaccessible with the aging effect associated with NUREG-2192 Item number 3.5.1-044.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5a Page 2 of 2 Table 3.5.2-11 page 3.5-117 is revised to reflect that concrete in a groundwater/soil environment is inaccessible:

Table 3.5.211: Off-Gas Stack - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG2191 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Concrete: Flood Barrier Concrete Groundwater/Soil Cracking Structures III.A9.TP30 3.5.1044 A Pedestal, Shelter, (Reinforced) Distortion Monitoring Walls, Slabs Protection (B.2.3.33)

(Accessible) Structural (Inaccessible) Support

Enclosure 35b Concrete Aging Management Review-Add ASR Detail

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 1 of 4 Concrete Aging Management Review-Add ASR Detail Enhancement for alkali-silica reaction inspection Affected SLRA Sections: 3.5.2.2.2.3, Appendix A Table A-3, Appendix B Section B.2.3.33 SLRA Page Numbers: 3.5-32, 3.5-33, A-92 and B-240 Description of Change:

Section 3.5.2.2.2.3 Item 2 is revised to add additional details related to how inaccessible areas of Group 6 structures are managed for alkali-silica reaction. Appendix A and Appendix B are revised to add an enhancement to the acceptance criteria in the Structures Monitoring AMP to provide additional characteristics that may indicate the presence of the alkali-silica reaction aging mechanism. The Appendix A enhancement is listed as commitment 36m because new commitments 36j, 36k, and 36l are being added in enclosures 31c and 31d.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 2 of 4 Section 3.5.2.2.2.3 Item 2 pages 3.5-32 and 3.5-33 is revised as follows:

(2) Cracking due to expansion and reaction with aggregates could occur in inaccessible concrete areas of Group 6 structures. Further evaluation is recommended to determine the need for a plantspecific AMP or plantspecific enhancements to Structures Monitoring AMP, to manage this aging effect. Acceptance criteria are described in BTP RLSB1 (Appendix A.1 of this SRPSLR).

Table 3.5.1, item number 3.5.1050: The Group 6 structures at MNGP are designed and constructed in accordance with ACI 201.2R77 using ingredients/materials conforming to ACI and ASTM standards. Concrete aggregates conform to the requirements of ASTM C33, Standard Specification of Concrete Aggregates. Water used for mixing concrete or processing concrete aggregates is free from any injurious amounts of acid, alkali, salts, oil, sediment, and organic matter. Tests and petrographic examinations performed according to ASTM C28964 and ASTM C295 verified that aggregates used are not reactive. For initial LR, the NRC determined that cracking due to reaction with aggregates would be adequately managed. OE has not identified any evidence of reaction with aggregates at MNGP. However, based on industry/fleet OE, cracking due to expansion and reaction with aggregates is an applicable aging effect in belowgrade inaccessible concrete areas for MNGP Group 6 structures and will be managed by the MNGP Inspections of WaterControl Structures Associated with Nuclear Power Plants AMP through the Structures Monitoring AMP (B.2.3.33). The Structures Monitoring AMP (B.2.3.33) will manage cracking due to the alkali-silica reaction (ASR) aging mechanism for inaccessible areas by monitoring accessible areas and opportunistically examining inaccessible areas. A plantspecific AMP is not required.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 3 of 4 SLRA Table A-3, commitment number 36 on Page A-92 is revised as follows:

No. Aging NUREG2191 Commitment Implementation Management Section Schedule Program or Activity (Section) 36 Structures XI.S6 m) Revise the implementing procedure Monitoring to include enhanced acceptance (A.2.2.33) criteria for detection of alkali-silica reactions in concrete to include:

Alkali-silica gel exudations Surface staining Expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5b Page 4 of 4 SLRA Section B.2.3.33 on page B-240 is revised to include an additional enhancement to the acceptance criteria as follows:

Element Affected Enhancement

6. Acceptance Criteria Revise the implementing procedure to include enhanced acceptance criteria for detection of alkali-silica reactions in concrete to include:

• Alkali-silica gel exudations

• Surface staining

• Expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components

Enclosure 35c Concrete Aging Management Review-Settlement

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 1 of 8 Concrete Aging Management Review-Settlement Settlement Affected SLRA Sections: 3.5.2.2.2.1, Table 3.5-1, A.2.2.33, B.2.3.33 SLRA Page Numbers: 3.5-29, 3.5-30, 3.5-57, 3.5-58, A-30, B-238, B-241 Description of Change:

Discussion of settlement in the SLRA has been revised to provide additional information on how the Structures Monitoring program manages settlement and provide additional details on operating experience with settlement for the Diesel Fuel Oil Transfer House and other structures that are monitored for settlement.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 2 of 8 Section 3.5.2.2.2.1, Item 3 on pages 3.5-29 and 3.5-30 is revised as follows:

3. Cracking and distortion due to increased stress levels from settlement could occur in belowgrade inaccessible concrete areas of structures for all Groups, and reduction in foundation strength, and cracking due to differential settlement and erosion of porous concrete sub foundations could occur in belowgrade inaccessible concrete areas of Groups 1-3, 5-9 structures. The existing program relies on structure monitoring programs to manage these aging effects. Some plants may rely on a dewatering system to lower the site groundwater level. If the plants CLB credits a dewatering system, verification is recommended of the continued functionality of the dewatering system during the subsequent period of extended operation. No further evaluation is recommended if this activity is included in the scope of the applicants structures monitoring program.

Table 3.5-1, item number 3.5.1044: MNGP does not rely on a dewatering system; thus, the Structures Monitoring AMP (B.2.3.33) will be used to manage cracking and distortion of the reinforced concrete elements of the MNGP structures founded on soil and/or exposed to a soil environment.

Table 3.5-1, item number 3.5.1046: The Structures Monitoring (B.2.3.33) AMP manages the aging effects in addition to monitoring for settlement and potential cracking. For SLR, groundwater is considered to be flowing water.

Therefore, the identification of indications of settlement is included in the Structures Monitoring (B.2.3.33) AMP for MNGP Group 1 through 3, 5, and 9 structures using item 3.5.1-044 to identify where this is required; 3.5.1-046 is not used. Additionally, as part of the Structures Monitoring (B.2.3.33) AMP, an annual inspection of the Diesel Fuel Oil Transfer House, Diesel Fuel Oil Storage Tank, and the Offgas Storage Building HTV exhaust pipe for settlement is performed to manage the aging effects of cracks, distortion, and increase in component stress level due to settlement.

With the exception of For structures other than the Diesel Fuel Oil Transfer House, Diesel Fuel Oil Storage Tank and OffGas Storage Building HTV exhaust pipe, no significant settlement has been observed on any major structure and dewatering systems are not used. Therefore, visual inspections are sufficient This satisfies to satisfy NUREG2192 requirements on for managing aging effects associated with concrete settlement, and therefore, with the exception of the Diesel Fuel Oil Transfer House, cracks, distortion, and increase in component stress levels due to settlement do not require aging management.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 35c Page 3 of 8 Table 3.5-1 Item 3.5.1-044 on page 3.5-57 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1044 All Groups: Cracking and AMP XI.S6, Structures Yes (SRPSLR Group 7 structures are not applicable to concrete: all distortion due to Monitoring" Section 3.5.2.2.2.1.3) MNGP.

increased stress levels from Consistent with NUREG2191. MNGP settlement does not rely on a dewatering system; thus, the Structures Monitoring(B.2.3.33)

AMP will be used to manage cracking and distortion of the reinforced concrete elements of the MNGP structures founded on soil and/or exposed to a soil environment. The Structures Monitoring AMP (B.2.3.33) includes periodic settlement monitoring of the Diesel Fuel Oil Transfer House, the Diesel Fuel Oil Storage Tank and Offgas Storage Building HTV exhaust pipe.

Further evaluation is documented in Section 3.5.2.2.2.1, item 3.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 35c Page 4 of 8 Table 3.5-1 Item 3.5.1-046 on page 3.5-58 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1046 Groups 13, Reduction of AMP XI.S6, "Structures Yes (SRPSLR Not Used Applicable.

59: concrete: foundation Monitoring" Section 3.5.2.2.2.1.3) foundation; strength and sub foundation cracking due to The foundation designs do not differential incorporate porous concrete in the settlement and subfoundation. The aging mechanism erosion of of settlement for MNGP is evaluated porous concrete under item 3.5.1-044. Since the sub foundation magnitude of the total settlements is small, differential settlement distortion is insignificant for Monticello structures.

Further evaluation is documented in Section 3.5.2.2.2.1, item 3.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 5 of 8 Section A.2.2.33 on page A-30 is revised as follows:

A.2.2.33 Structures Monitoring The MNGP Structures Monitoring AMP is an existing AMP that consists of periodic visual inspection and monitoring of the condition of concrete and steel structures, structural components, component supports, and structural commodities to ensure that aging degradation (such as those described in ACI 349.3R, ACI 201.1R, SEI/ASCE 11, and other documents) will be detected, the extent of degradation determined and evaluated, and corrective actions taken prior to loss of intended functions. Structures are monitored on an interval not to exceed 5 years. Inspections also include seismic joint fillers, elastomeric materials; steel edge supports, and bracings associated with masonry walls, and periodic evaluation of ground water chemistry and opportunistic inspections for the condition of below grade concrete. The program includes annual survey measurement of settlement for the Diesel Fuel Oil Transfer House, the Diesel Fuel Oil Storage Tank and Offgas Stroage Building HTV exhaust pipe to provide early indication of potential stress increases that could result in cracking or deflection of the structural components associated with these structures. Quantitative results (measurements) and qualitative information from periodic inspections are trended with sufficient detail, such as photographs and surveys for the type, severity, extent, and progression of degradation, to ensure that corrective actions can be taken prior to a loss of intended function. The acceptance criteria are derived from applicable consensus codes and standards. For concrete structures, the program includes personnel qualifications and quantitative evaluation criteria of ACI 349.3R.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 6 of 8 Section B.2.3.33 on page B-238 third paragraph is revised as follows:

Inspections and evaluations are performed by personnel qualified in accordance with GALLSLR requirements using criteria derived from industry codes and standards contained in the plant CLB including but not limited to ACI 349.3R, ACI 318, SEI/ASCE 11, and the American Institute of Steel Construction (AISC) specifications, as applicable. The AMP includes preventive actions to ensure structural bolting integrity. The program also includes periodic sampling and testing of ground water and the need to assess the impact of any changes in its chemistry on below grade concrete structures. In addition to routine inspections for cracking of concrete, as a result of plant operating experience, the program includes annual survey measurement of settlement for the Diesel Fuel Oil Transfer House, the Diesel Fuel Oil Storage Tank and Offgas Storage Building HTV exhaust pipe to provide early indication of potential stress increases that could result in cracking or deflection of the structural components associated with these structures.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 7 of 8 Section B.2.3.33 on page B-241 is revised to add additional plant specific operating experience and clarify the structure name for the Diesel Fuel Oil Transfer house as follows:

PlantSpecific Operating Experience Significant settlement of the Diesel Fuel Oil Transfer House occurred shortly after initial construction.

o 1970-Following heavy rainfall significant settlement of the Diesel Fuel Oil Transfer House was observed. The condition of the structure and internal equipment was reviewed and found acceptable for continued use.

o 1991-In the summer of 1991 the settlement of the Diesel Fuel Oil Transfer House was reevaluated. The elevations of various locations in the structure were measured and compared to reference drawings and documented that the east side of the building had experienced approximately 3/4 to one inch of settlement and the west side of the building had experienced approximately 5 1/4 to 5 1/2 inches of settlement. The Diesel Fuel Oil Transfer House provides shelter to two diesel fuel oil pumps that are connected by buried piping to a diesel fuel oil storage tank and to equipment in the Diesel Generator Building. To ensure the buried piping was not damaged by the building settlement, the area around the piping was excavated and detailed piping examinations performed. The piping was found to be in acceptable condition. The review indicated the piping is fabricated to meet ASTM A-106 Grade B pipe specifications with ASTM A-105 II fittings and concluded that any stresses related to the settlement were self-limiting due to the ductile nature of the material and the fact that the excavation relieved any significant interaction with the soil. Periodic surveillance of future settling was established.

o 2006-During performance of the annual settlement surveillance, a weakness in the procedure was noted with respect to specifying the locations where survey measurements were taken to ensure repeatability. The procedure was revised to include controlled points and to establish a new baseline for the measurements.

Acceptance criteria were defined relative to the new baseline measurements that ensure even small amounts of settlement that exceed the established acceptance criteria will result in entering the condition into the site corrective action program, thus ensuring the condition is evaluated well in advance of any potential loss of function for the Diesel Fuel Oil Transfer House or the associated diesel fuel oil pumps and piping.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5c Page 8 of 8 o 2006 Review for License Renewal-In October 2006 the NRC issued a Safety Evaluation Report (SER) in NUREG-1865. Included in this SER was a review of how the site manages settlement of the Diesel Fuel Oil Transfer House. The NRC found that The Structures Monitoring Program manages the aging effects for the diesel fuel oil transfer house. As part of the Structures Monitoring Program the applicant performs an annual inspection for settlement to manage the aging effects of cracking, distortion, and increase in component stress level due to settlement. The inspection adds assurance that the aging effects do not occur or progress so slowly that the components intended function will be maintained during the period of extended operation.

o 2006-2022-Annual surveillance of settlement associated with the Diesel Fuel Oil Transfer building have continued consistent with the program described for the initial License Renewal. All settlement values observed for the Diesel Fuel Oil Transfer House, Diesel Fuel Oil Storage Tank, and the Offgas Storage Building HTV exhaust pipe have been within the specified range with no exceedance of the specified acceptance criteria. Structures Monitoring inspections of the concrete in the Diesel Fuel Oil Transfer Building have found no significant cracking.

During a structures walkdown, blistering and peeling paint was found around a grouted pipe penetration. The paint was determined to not affect the grouted penetration or impact any SSCs. The peeling paint was removed, and the area was repainted and sealed.

Corrosion was found on the offgas stack supports and hangers. The vents on the lower part of the stack were sealed. The lack of air flow and presence of moisture could potentially be accelerating the corrosion. A condition evaluation was performed which determined there were no adverse conditions and no resulting actions were required.

Leaking was identified in the Diesel Fuel Oil Pump Transfer House after a large concrete slab was placed on the roof as a missile shield. The water intrusion was determined to be from a pipe penetration. There was evidence that the flow of runoff water had changed and that recent rains around the building resulted in leakage into the transfer pump house. Sampling was performed that confirmed no indications of oil leakage. Minor maintenance was performed to seal the penetration and apply fill material to grading near the pump to direct water away from the structure.

Enclosure 35d Concrete Aging Management Review-Correction Of Omitted Line Item

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5d Page 1 of 2 Concrete Aging Management Review-Correction Of Omitted Line Item Correction of omitted line item from SLRA Table 3.5.2-4 Affected SLRA Sections: Table 3.5.2-4 SLRA Page Numbers: 3.5-89 Description of Change:

Table 3.5.2-4: Emergency Diesel Generator Building-Summary of Aging Management Evaluation is revised to include an additional line for the aging effect associated with NUREG-2192 Item number 3.5.1-042.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 Enclosure 35d Page 2 of 2 Table 3.5.2-4 page 3.5-89 is revised to add the following line:

Table 3.5.24: Emergency Diesel Generator Building - Summary of Aging Management Evaluation Aging Effect Intended Aging Management NUREG-2191 Table 1 Component Type Material Environment Requiring Notes Function Program Item Item Management Concrete: Basemat, Structural Concrete Groundwater/Soil Cracking Structures Monitoring III.A3.TP108 3.5.1042 A Foundation Support (Reinforced) Loss of Material (B.2.3.33)

(Inaccessible)

Enclosure 35e Concrete Aging Management Review-Operating Experience

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5e Page 1 of 2 Concrete Aging Management Review-Operating Experience Clarification of operating experience for leaching Affected SLRA Sections: 3.5.2.2.2.1 Item 4 SLRA Page Numbers: 3.5-30 Description of Change:

Section 3.5.2.2.2.1 Item 4 is revised to clarify that MNGP has not experienced a loss of function as a result of leaching of calcium hydroxide or carbonation.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5e Page 2 of 2 Section 3.5.2.2.2.1 Item 4 on page 3.5-30 is revised as follows:

4. Increase in porosity and permeability, and loss of strength due to leaching of calcium hydroxide and carbonation could occur in belowgrade inaccessible concrete areas of Groups 1-5 and 7-9 structures. Further evaluation is recommended to determine the need for a plantspecific AMP or plantspecific enhancements to Structures Monitoring AMP, to manage these aging effects if leaching is observed in accessible areas that impact intended functions. Acceptance criteria are described in BTP RLSB1 (Appendix A.1 of this SRPSLR).

Table 3.5.1, item number 3.5.1047: Group 1 through 3, 5, and 9 structures at MNGP are designed and constructed in accordance with ACI 201.2R77 using ingredients/materials conforming to ACI and ASTM standards. Concrete aggregates conform to the requirements of ASTM C3364 (fine and coarse aggregate). Materials for concrete used in MNGP concrete SCs were specifically investigated, tested, and examined in accordance with pertinent ASTM standards. MNGP foundation materials do not contain any porous layers. The concrete base or lean concrete fill material used beneath major building foundations did not include highalumina cement. MNGP does not rely on a dewatering system to lower site ground water. While some instances of water seepage through concrete have occurred (see B.2.3.33), MNGP OE does not indicate leaching or carbonation has been observed on accessible concrete areas that would impact intended functions of the structure.

However, that notwithstanding, the foundations of MNGP groups 1 through 3, 5, and 9 plant structures are considered to be exposed to groundwater, which for SLR is considered to be flowing water. Periodic ground water level measurements and chemical analysis of ground water are performed to verify the associated chemistry remains nonaggressive as described in the Structures Monitoring AMP (B.2.3.33). The frequency of monitoring ground water chemistry (pH, chlorides, and sulfates) is monthly. In addition, the Structures Monitoring AMP (B.2.3.33) includes opportunistic inspection of inaccessible concrete surfaces, when excavation for other reasons permits access. Accessible areas of concrete structures exposed to an outdoor air environment can be used as an indicator of concrete condition in a soil or groundwater environment. Any significant leaching or carbonation that is observed in accessible areas will be evaluated for the potential impact on the function of concrete in inaccessible areas.

Enclosure 35f Concrete Aging Management Review-Clari"cation of Freeze-thaw Evaluation

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5f Page 1 of 4 Concrete Aging Management Review-Clarification of Freeze-thaw Evaluation Clarification of Freeze-thaw Evaluation Affected SLRA Sections: 3.5.2.2.2.3 Item 1, B.2.3.34 SLRA Page Numbers: 3.5-32, B-245 Description of Change:

Section 3.5.2.2.2.3 Item 1 is revised to state that the air content of concrete associated with MNGP Group 6 structures is within the limits defined by NUREG-2192 for freeze-thaw resistance and clarify operating experience.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5f Page 2 of 4 Section 3.5.2.2.2.3 Item 1 on page 3.5-32 is revised as follows:

(1) Loss of material (spalling, scaling) and cracking due to freezethaw could occur in belowgrade inaccessible concrete areas of Group 6 structures.

Further evaluation is recommended to determine the need for a plantspecific AMP or plantspecific enhancements to Structures Monitoring AMP to manage these aging effects for inaccessible areas for plants located in moderate to severe weathering conditions. Acceptance criteria are described in BTP RLSB1 (Appendix A1 of this SRPSLR).

Further evaluation for Inaccessible Areas for Group 6 Structures is provided below:

Table 3.5.1 Item Number 3.5.1049: Group 6 structures at MNGP are located in a severe weathering region per Figure 1 of ASTM C33, Location of Weathering Regions. As such, consistent with the initial LR, the MNGP Structures Monitoring AMP (B.2.3.33) would detect concrete aging effects related to freezethaw, should it occur. Cracking, spalling and disintegration of concrete due to freezethaw cycling are concrete aging effects requiring management in the atmosphere/weather environment. In addition, lLoss of material (disintegration of concrete) and corrosion of reinforcing steel (with consequent spalling of concrete) due to the action of deicing salts is an aging effect requiring management in the localized area of the Intake Structure and Tunnel roof slabs based on operating experience. The degradation was corrected prior to a loss of function for the Intake Structure.

The principal tool for managing these effects is examinations performed as required by the Structures Monitoring AMP (B.2.3.33). A significant portion of the Group 6 structures at MNGP are accessible and provide indication of the condition of inaccessible portions of the structure.

In accordance with NUREG2192, concrete located exterior and above grade in accessible areas (i.e., exposed to atmosphere/weather environment) is managed for the aging effects:

Loss of material (spalling, scaling) and cracking due to freeze thaw Loss of material (spalling, scaling) and cracking due to deicing salts Increase in porosity, permeability, and loss of strength due to leaching of calcium hydroxide Expansion and cracking due to reaction with aggregates Cracking, loss of bond, loss of material (spalling, scaling) due to corrosion of embedded steel

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5f Page 3 of 4 Cracking, loss of bond, loss of material (spalling, scaling) due to deicing salts Increase in porosity and permeability, cracking, loss of material due to aggressive chemical attack Reduction in concrete anchor capacity due to local concrete degradation/

Serviceinduced cracking or other concrete aging mechanisms Additionally, the MNGP Structures Monitoring AMP (B.2.3.33) opportunistic inspections confirm the absence of aging effects by examining normally inaccessible structural components, when scheduled maintenance work and planned plant modifications permit access and will evaluate observed aging effects in accessible areas that could be indicative of degradation in inaccessible areas.

The air content of the concrete associated with Group 6 structures is within the bounds of 3% to 8% specified in NUREG-2192 and therefore a plant specific program for managing the aging effect of freeze-thaw is not warranted.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5f Page 4 of 4 Section B.2.3.34 on page B-245 is revised as follows:

A study was completed on the roof of the INS by American Engineering and Testing (AET). Repairs were recommended in order to maintain the roof. There were no immediate concerns that would affect the equipment within the INS. Deicing salt was frequently used on the intake structure roof. This salt water mixture infiltrated the concrete, which was further aggravated by freeze-thaw cycling; ultimately resulting in cracking. An evaluation discussed a number of actions including calculations that were performed to confirm the structural integrity of the intake building roof. The roof concrete was repaired, and a new roof membrane installed. Use of deicing salt on the intake structure roof was prohibited.

Enclosure 35g Concrete Aging Management Review-Clari"cation of Inconsistencies

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 1 of 10 Concrete Aging Management Review-Clarification of Inconsistencies Clarification of inconsistencies in Table 3.5-1 Affected SLRA Sections: Table 3.5-1 SLRA Page Numbers: 3.5-56, 3.5-57, 3.5-58, 3.5-62, 3.5-64, 3.5-65, 3.5-66 Description of Change:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports, is revised to consistently reflect that NUREG-2191 Group 7 and 8 line items are not applicable to MNGP.

Note that information provided in bold, black font are changes incorporated from Enclosures 31f and 35c.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 2 of 10 Table 3.5-1 Item 3.5.1-042 on page 3.5-56 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Aging Item Further Evaluation Component Requiring Management Discussion Number Recommended Management Program 3.5.1042 Groups 13, 5, 7 9: Loss of material Plantspecific aging Yes (SRPSLR Group 7 structures are not applicable to concrete (spalling, scaling) management Section 3.5.2.2.2.1.1) MNGP. Group 7 and Group 8 (inaccessible areas): and cracking due to program or AMP structures are not applicable to foundation freeze thaw XI.S6, "Structures MNGP. Concrete associated with Monitoring" missile barriers are evaluated with the associated structure and the Condensate Storage Tank foundations are evaluated with Group 3 structures.

MNGP is located in a severe weathering region, where freezing conditions are occasionally experienced. However, a plantspecific AMP is not required to manage loss of material, cracking in inaccessible areas.

Consistent with the current renewed licenses, the Structures Monitoring (B.2.3.33) AMP would detect degradation of concrete due to freezethaw, should it occur, and includes opportunistic examination of normally inaccessible components when excavated for other reasons.

Further evaluation is documented in Section 3.5.2.2.2.1, item 1.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 3 of 10 Table 3.5-1 Items 3.5.1-043 and 3.5.1-044 on page 3.5-57 are revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1043 All Groups Cracking due to Plantspecific aging Yes (SRPSLR Group 7 structures are not applicable to except Group expansion from management program Section 3.5.2.2.2.1.2) MNGP. Group 7 and Group 8 structures 6: concrete reaction with or AMP XI.S6, are not applicable to MNGP. Concrete (inaccessible aggregates "Structures Monitoring" associated with missile barriers are areas): all evaluated with the associated structure and the Condensate Storage Tank foundations are evaluated with Group 3 structures.

Consistent with the current renewed licenses, a plantspecific AMP is not required to manage cracking in inaccessible areas. The Structures Monitoring (B.2.3.33) AMP includes examination for unique map or cracking. The Structures Monitoring (B.2.3.33) AMP also includes opportunistic examination of belowgrade inaccessible concrete areas.

Further evaluation is documented in Section 3.5.2.2.2.1, item 2.

3.5.1044 All Groups: Cracking and AMP XI.S6, Structures Yes (SRPSLR Group 7 structures are not applicable to concrete: all distortion due to Monitoring" Section 3.5.2.2.2.1.3) MNGP. Group 7 and Group 8 structures increased stress are not applicable to MNGP. Concrete levels from associated with missile barriers are

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 4 of 10 settlement evaluated with the associated structure and the Condensate Storage Tank foundations are evaluated with Group 3 structures.

Consistent with NUREG2191. MNGP does not rely on a dewatering system; thus, the Structures Monitoring (B.2.3.33) AMP will be used to manage cracking and distortion of the reinforced concrete elements of the MNGP structures founded on soil and/or exposed to a soil environment. The Structures Monitoring AMP (B.2.3.33) includes periodic settlement monitoring of the Diesel Fuel Oil Transfer House, the Diesel Fuel Oil Storage Tank and Offgas Storage Building HTV exhaust pipe.

Further evaluation is documented in Section 3.5.2.2.2.1, item 3.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 5 of 10 Table 3.5-1 Item 3.5.1-047 on page 3.5-58 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1047 Groups 15, Increase in Plantspecific aging Yes (SRPSLR Group 7 and Group 8 structures are not 79: concrete porosity and management program or Section 3.5.2.2.2.1.4) applicable to MNGP. Concrete associated (inaccessible permeability; AMP XI.S6, "Structures with missile barriers are evaluated with the areas): loss of Monitoring" associated structure and the Condensate exterior strength due Storage Tank foundations are evaluated above and to leaching of with Group 3 structures.

below grade; calcium foundation hydroxide The Structures Monitoring (B.2.3.33) AMP will and be used to manage increase in porosity and carbonation.

permeability, loss of strength of the reinforced concrete basemat, foundation, subfoundation, belowgrade exterior concrete, pedestal, walls, slabs (inaccessible areas), diesel fuel oil storage tank deadmen, 115/345 kV Substation Control House, trenches, and duct bank exposed to waterflowing in Groups 2, 3, and 9 structures.

Further evaluation is documented in Section 3.5.2.2.2.1, item 4.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 6 of 10 Table 3.5-1 Item 3.5.1-054 on page 3.5-62 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Aging Item Further Evaluation Component Requiring Management Discussion Number Recommended Management Program 3.5.1054 All groups Cracking due AMP XI.S6 No Group 7 and Group 8 structures are not except 6: to expansion Structures applicable to MNGP. Concrete associated with concrete from reaction Monitoring missile barriers are evaluated with the (accessible with associated structure and the Condensate areas):all aggregates Storage Tank foundations are evaluated with Group 3 structures.

Consistent with NUREG-2191.

The Structures Monitoring (B.2.3.33) AMP is credited with managing cracking of accessible concrete exposed to uncontrolled indoor air and outdoor air environments.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 7 of 10 Table 3.5-1 Item 3.5.1-063 on page 3.5-64 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Aging Item Further Evaluation Component Requiring Management Discussion Number Recommended Management Program 3.5.1063 Groups 13, Increase in AMP XI.S6, No Group 7 and Group 8 structures are not applicable to 5, 79: porosity and "Structures MNGP. Concrete associated with missile barriers concrete permeability; Monitoring" are evaluated with the associated structure and (accessible loss of the Condensate Storage Tank foundations are areas): strength due evaluated with Group 3 structures.

exterior to leaching of above and calcium Consistent with NUREG2191.

below hydroxide and grade; carbonation foundation The Structures Monitoring (B.2.3.33) AMP is credited with managing leaching or carbonation of exterior plant structure concrete and foundations where groundwater or precipitation runoff forms a flowing water environment.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 8 of 10 Table 3.5-1 Items 3.5.1-064, 3.5.1-065, and 3.5.1-066 on page 3.5-65 are revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5.1064 Groups 13, 5, 79: Loss of material AMP XI.S6, No Group 7 and Group 8 structures are not concrete (accessible (spalling, "Structures applicable to MNGP. Concrete areas): exterior scaling) and Monitoring associated with missile barriers are above and below cracking due to evaluated with the associated structure grade; foundation freeze thaw and the Condensate Storage Tank foundations are evaluated with Group 3 structures.

Consistent with NUREG2191.

The Structures Monitoring (B.2.3.33) AMP is credited with managing loss of material and cracking for accessible plant structure concrete exposed to outdoor air.

3.5.1065 Groups 13, 5, 79: Cracking; loss of AMP XI.S6, No Group 7 and Group 8 structures are not concrete bond; and loss "Structures applicable to MNGP. Concrete (inaccessible areas): of material Monitoring associated with missile barriers are belowgrade exterior; (spalling, evaluated with the associated structure foundation, Groups scaling) due to and the Condensate Storage Tank 13, 5, 79: concrete corrosion of foundations are evaluated with Group 3 (accessible areas): embedded steel structures.

belowgrade exterior; foundation, Groups 6: concrete Consistent with NUREG2191.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 9 of 10 (inaccessible areas):

all The Structures Monitoring (B.2.3.33) AMP is credited with managing cracking, loss of bond, loss of material for inaccessible plant structure concrete exposed to groundwater/soil.

3.5.1066 Groups 15, 7, 9: Cracking, Loss AMP XI.S6 No Group 7 and Group 8 structures are not concrete (accessible of bond, Loss of "Structures applicable to MNGP. Concrete areas): interior and material Monitoring" associated with missile barriers are abovegrade exterior (spalling, evaluated with the associated structure scaling) due to and the Condensate Storage Tank corrosion of foundations are evaluated with Group 3 embedded steel structures.

Consistent with NUREG2191 The Structures Monitoring (B.2.3.33) AMP is credited with managing cracking, loss of bond, and loss of material for accessible plant structure concrete exposed to uncontrolled indoor air, and outdoor air environments.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5g Page 10 of 10 Table 3.5-1 Item 3.5.1-067 on page 3.5-66 is revised as follows:

Table 3.5-1: Summary of Aging Management Evaluations for Plant Structures and Component Supports Aging Effect Aging Item Further Evaluation Component Requiring Management Discussion Number Recommended Management Program 3.5.1067 Groups 15, Increase in AMP XI.S6 No Group 7 and Group 8 structures are not applicable to 7, 9: porosity and "Structures MNGP. Concrete associated with missile barriers Concrete: permeability, Monitoring" are evaluated with the associated structure and interior; Cracking, Loss the Condensate Storage Tank foundations are abovegrade of material evaluated with Group 3 structures.

exterior, (spalling, Groups 13, scaling) due to 5, 79 aggressive Consistent with NUREG2191.

concrete: chemical belowgrade attack The Structures Monitoring (B.2.3.33) AMP is credited exterior; with managing potential increase in porosity and foundation, permeability, cracking, and loss of material due to Group 6: aggressive chemical attack for inaccessible plant concrete: all structure concrete in uncontrolled indoor air, outdoor air, and groundwater/soil environments.

Enclosure 35h Concrete Aging Management Review-Clari"cation of Operating Experience

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5h Page 1 of 2 Concrete Aging Management Review-Clarification of Operating Experience Clarification of operating experience with leaching at Intake Structure Affected SLRA Sections: 3.5.2.2.2.3 Item 3 SLRA Page Numbers: 3.5-33 Description of Change:

Section 3.5.2.2.2.3 Item 3 is revised to discuss operating experience and state that operating experience with leaching for the MNGP Intake Structure has not resulted in a loss of function.

Monticello Nuclear Generating Plant Docket 50-263 L-MT-23-025 5h Page 2 of 2 Section 3.5.2.2.2.3 Item 3 on page 3.5-33 is revised as follows:

(3) Increase in porosity and permeability and loss of strength due to leaching of calcium hydroxide and carbonation could occur in inaccessible areas of concrete elements of Group 6 structures. Further evaluation is recommended to determine the need for if a plantspecific AMP or plantspecific enhancements to Structures Monitoring AMP, to manage these aging effects if leaching is observed in accessible areas that impact intended functions. Acceptance criteria are described in BTP RLSB1 (Appendix A.1 of this SRPSLR).

Table 3.5-1, item number 3.5.1051: The Groups 6 structures at MNGP are designed and constructed in accordance with ACI 201.2R77 using ingredients/materials conforming to ACI and ASTM standards. Concrete aggregates conform to the requirements of ASTM C3364 (fine and coarse aggregate). Materials for concrete used in MNGP concrete SCs were specifically investigated, tested, and examined in accordance with pertinent ASTM standards. MNGP foundation materials do not contain any porous layers. The concrete base or lean concrete fill material used beneath major building foundations did not include highalumina cement. MNGP does not rely on a dewatering system to lower site ground water.

However, that notwithstanding, the foundations of MNGP Group 6 plant structures are considered to be exposed to groundwater, which for SLR is considered to be flowing water. Periodic ground water level measurements and chemical analysis of ground water are performed to verify the associated chemistry remains nonaggressive as described in the Structures Monitoring (B.2.3.33) AMP. The frequency of monitoring ground water chemistry (pH, chlorides, and sulfates) is monthly. In addition, the Structures Monitoring (B.2.3.33) AMP includes opportunistic inspection of inaccessible concrete surfaces, when excavation for other reasons permits access, and evaluation of impact to inaccessible area intended functions if degradation, such as leaching or carbonation, is observed in an accessible area.

As discussed in the Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.3.34) AMP, the MNGP Intake Structure has experienced some instances of water seepage through concrete. In each case the condition was corrected prior to a loss of an intended function for the intake structure, therefore a plant specific program for managing increase in porosity and permeability and loss of strength due to leaching of calcium hydroxide and carbonation in inaccessible areas of concrete elements of Group 6 structures is not warranted.