CP-202300322, Response to Requests for Additional Information Regarding Review of the License Renewal Application Sets 2 and 3

Response to Requests for Additional Information Regarding Review of the License Renewal Application Sets 2 and 3

ML23208A193
Person / Time
Site:	Comanche Peak
Issue date:	07/27/2023
From:	John Lloyd Luminant, Vistra Operations Company
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
CP-202300322, TXX-23048
Download: ML23208A193 (1)

Text

II Comanche Peak Jay Lloyd Nuclear Power Plant Senior Director, Engineering (Vistra Operations

& Regulatory Affairs Company LLC)

Luminant P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 T 254.897.5337 CP-202300322 TXX-23048 July 27, 2023 U. S. Nuclear Regulatory Commission Ref 10 CFR 54 ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

COMANCHE PEAK NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NUMBERS 50-445 AND 50-446 FACILITY OPERATING LICENSE NUMBERS NPF-87 and NPF-89 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION REGARDING REVIEW OF THE LICENSE RENEWAL APPLICATION - SETS 2 AND 3

REFERENCES:

1. Letter TXX-22077, from Steven K. Sewell to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application, October 3, 2022 (ADAMS Accession No. ML22276A082)

2. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

3. Letter TXX-23022, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 2, April 24, 2023 (ADAMS Accession No. ML23114A377)

4. Electronic Communications, from M. Yoo (NRC) to K. Peters (Vistra), "Comanche Peak LRA -

Request for Additional Information - Set 2," June 29, 2023 (ADAMS Accession Nos.

ML23181A020 and ML23181A021)

5. Electronic Communications, from M. Yoo (NRC) to K. Peters (Vistra), "Comanche Peak LRA -

Request for Additional Information - Set 3," July 7, 2023 (ADAMS Accession Nos. ML23188A044 and ML23188A045)

Dear Sir or Madam:

In Reference 1, as supplemented by References 2 and 3, Vistra Operations Company LLC (Vistra OpCo) submitted a license renewal application (LRA) for the Facility Operating Licenses for Comanche Peak Nuclear Power Plant (CPNPP) Units 1 and 2. The NRC issued Requests for Additional Information (RAIs) to Vistra OpCo via Reference 4 and Reference 5. Vistra OpCos responses to the Reference 4 RAIs are provided in Attachments A through L of this letter. Vistra OpCos response to the Reference 5 RAI is provided in Attachment M of this letter.

During recent discussions with the NRC Staff, Vistra OpCo agreed to provide further clarifying information. Attachment N to this letter provides that information.

TXX-23048 Page 2 of 3 Lastly, a typographical inconsistency was noted in the header of the left column of the enhancement tables in LRA Appendix B. Typically, the column header appears as "Element Affected," however, in several instances the column header appears as "Element Effected." In Attachments A and B of this submittal, the column header is changed to "Element Affected" using the conventions established for added and deleted text. To ensure consistency, the column header is hereby changed to "Element Affected" on the following pages: B-21 (B.2.2.1), B-25 (B.2.2.2), B-33 (B.2.3.2), B-68 and B-69 (B.2.3.8), B-73 and B-74 (B.2.3.9), B-89 (B.2.3.11), B-97 (B.2.3.13), B-110 (B.2.3.16), B-189 (B.2.3.33), and B-201 (B.2.3.36).

Markups of these pages are not provided as this is an editorial change.

For ease of reference, an index of the attachments topics is provided on page 3 of this letter.

Should you have any questions, please contact Todd Evans at (254) 897-8987 or Todd.Evans@luminant.com.

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

Executed on July 27, 2023 Sincerely, Jay Lloyd c: (email) w / o Attachments Mahesh Chawla, NRR, DORL [Mahesh.Chawla@mc.gov]

Victor Dricks, RGN IV /OPA [Victor.Dricks@mc.gov]

John Ellegood, RGN IV, CPNPP [John.Ellegood@mc.gov]

Dennis Galvin, NRR [Dennis.Galvin@mc.gov]

Lauren Gibson, NRR/DNRL [Lauren.Gibson.me.gov]

Jessica Hammock, NRR, DNRL [Jessica.Hammock@mc.gov]

Jim Melfi, RGN IV [Jim.Melfi@mc.gov]

Robert Lewis, RGN IV [Robert.Lewis@mc.gov]

Greg Pick RGN IV [Greg.Pick@mc.gov]

David Proulx, NRR/RGN IV [David.Proulx@mc.gov]

Mark Yoo, NRR/DNRL [Mark.Yoo@mc.gov]

Chris Smith, RGN IV [Chris.Smith@mc.gov]

Theodore Smith, NMSS/REFS [Theodore.Smith@mc.gov]

Tam Tran, NMSS/REFS [Tam.Tran@mc.gov]

Nick Taylor, RGN IV [Nick.Taylor@mc.gov]

Greg Werner, RGN IV [Greg.Werner@mc.gov]

TXX-23048 Page 3 of 3 Attachments Index Attachment CPNPP LRA Information Topics No.

A B.2.3.16-1 Diesel Engine Fire Pump Heat Exchanger B B.2.3.16-2 Fire Water Storage Tank Bottoms C B.2.3.16-3 Coating Corrective Actions D B.2.3.16-4 Fire Water System - Inspection and Testing 3.2.2.2.3.2-1 Loss of Material Due to Pitting and Crevice Corrosion - Stainless Steel E

Exposed to Outdoor Air F B.2.3.15-1 Masonry Walls and Removal Concrete Block Fire Barriers G B.2.3.15-2 Fire Protection - Concrete Curbs, Concrete Berm/Dike, Fire Barrier Function 3.5.2.2.1.2-1 Reduction of Strength and Modulus Due to Elevated Temperature (Stainless H

Steel Insulation)

I 4.2.1-1 Least Squares Adjustment of RPV Neutron Exposures J 4.2.1-2 Neutron Fluence Methodology B.2.3.34-1 Enhancements Related to Cracking Due to Expansion from Reactions with K

Aggregates L 4.3.3-1 ANSI B31.1 Piping M 3.3.2.2.2-1 Radiation Monitoring in the Closed Cooling Water System Voluntary Clarification of B.2.3.27 Enhancements Relative to Fire Water Storage Tank N

Level Monitoring for Greater Consistency Between the Elements.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 1 of 15 LRA Section: B.2.3.16, Fire Water System NRC RAI No: B.2.3.16-1 (Diesel Engine Fire Pump Heat Exchanger)

Regulatory Basis:

Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S.

Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described in the requests for additional information.

Background:

License Renewal Application (LRA) Supplement 1 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23096A302) added the following aging management review (AMR) items associated with the diesel engine fire pump heat exchangers to LRA Table 3.3.2-7:

Loss of material of copper alloy tubes exposed internally to treated water (AMR Item 3.4-1, 016)

Reduction of heat transfer of copper alloy tubes exposed internally to treated water (AMR Item 3.4-1, 018)

Loss of material of carbon steel tubesheet exposed internally to treated water (AMR Item 3.4-1, 015)

Loss of material of carbon steel channel head exposed internally to treated water (AMR Item 3.4-1, 015)

Each of the above AMR items cite generic note E (consistent material, environment, and aging effect, but different aging management program (AMP) credited) and plant-specific note 2. Plant-specific note 2 states, The Fire Water System (B.2.3.16) AMP is substituted for the Water Chemistry (B.2.3.2) AMP and One-Time Inspection (B.2.3.19) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 2 of 15 Issue:

Revision 2 of NUREG-1801, Generic Aging Lessons Learned (GALL) Report (ML103490041), and Revision 2 of NUREG-1800, Standard Review Plan for Review of License Renewal [SRP-LR] Applications for Nuclear Power Plants (ML103490036), do not include AMR items for heat exchanger components (e.g., heat exchanger tubes) that credit the Fire Water System AMP as the program to manage the effects of aging.

LRA Supplement 1 and LRA Supplement 2 (ML23114A377) did not make changes to the Discussion column of AMR Items 3.4-1, 015, 016 (only refers to copper alloy piping components in the Fire Protection System), and 018 in LRA Table 3.4-1 to indicate that these AMR items are used to manage diesel engine fire pump heat exchanger components.

LRA Supplements 1 and 2 did not make changes (e.g., enhancements) to Section A.2.2.16 in LRA Appendix A or Section B.2.3.16 in LRA Appendix B related to the Fire Water System AMP managing loss of material and reduction of heat transfer of the copper alloy heat exchanger tubes, and managing loss of material of the carbon steel heat exchanger tubesheet and channel head. Specifically, neither the LRA nor Revision 2 of LUM00020-REPT-053, Fire Water System Aging Management Program Basis Document, describe how the Fire Water System AMP will manage these aging effects for the diesel engine fire pump heat exchanger components.

The staff notes that one subsequent license renewal applicant stated that inspection of the heat exchanger tube bundle for degradation is not practical due to the small tube diameter (ML21091A187).

Request:

Please provide information discussing how the Fire Water System AMP will manage loss of material and reduction of heat transfer of the copper alloy heat exchanger tubes and managing loss of material of the carbon steel heat exchanger tubesheet and channel head.

At a minimum, the discussion should include whether inspections of the heat exchanger tubes are practical, what inspections will be performed (i.e., scope, frequency, acceptance criteria, corrective actions), and whether the program will be enhanced to include these inspections in an existing or new procedure.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 3 of 15 Luminant Response:

The Fire Water System AMP will manage loss of material and reduction of heat transfer of the copper alloy heat exchanger tubes and loss of material of the carbon steel heat exchanger tubesheet and channel head. Aging effects will be managed via periodic cleaning and visual inspection of the tubes and associated sub-components, which is practical and performed based on maintenance history. An existing periodic maintenance procedure for the fire pump diesel engines includes cleaning and visual inspection of heat exchanger internals, including tubes. The Fire Water System AMP will be enhanced so that the existing procedure will also include cleaning and visual inspection of the heat exchanger tubesheet and channel head for evidence that loss of material has not occurred.

Each time the heat exchanger is opened for maintenance, the heat exchanger tubes are cleaned. In addition to periodic cleaning and inspection of heat exchanger components, engine parameters are monitored during routine performance runs. This includes monitoring of jacket water temperature. Monitoring of jacket water temperature provides additional assurance that the heat exchangers continue to perform their intended function. An increase in jacket water temperature would indicate that fouling may be occurring.

The scope of the Fire Water System AMP is enhanced to include the tubes, tubesheet, and channel head of each fire pump diesel engine heat exchanger.

The Fire Water System AMP is clarified to state that heat exchanger inspections are performed based on maintenance history and at least once every ten years, and will continue through the period of extended operation. In addition, heat exchanger tubes are cleaned whenever the heat exchanger is opened for maintenance.

With respect to acceptance criteria, the Fire Water System AMP will be enhanced to state that any indication of fouling or corrosion of heat exchanger components is evaluated.

With respect to corrective actions, the Fire Water System AMP will be enhanced to specify that conditions that do not meet acceptance criteria are entered into the CAP for evaluation.

References:

1. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

2. Letter TXX-23022, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 2, April 24, 2023 (ADAMS Accession No. ML23114A377)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 4 of 15 LRA Table 3.4-1 (page 3.4-21), is revised as follows:

Table 3.4-1: Summary of Aging Management Programs for Steam and Power Conversion Systems Aging Aging Further Evaluation Item Component Effect/Mechanis Management Discussion Recommended m Programs 3.4-1, Steel Heat Loss of material Chapter XI.M2, No Consistent with exception to NUREG-1801.

015 exchanger due to general, Water Chemistry, Loss of material of steel heat exchanger components exposed components pitting, crevice, and Chapter to treated water in the Main Steam and Auxiliary Steam exposed to and galvanic XI.M32, One-Time Systems is managed by the Water Chemistry (B.2.3.2) AMP, Treated water corrosion Inspection which takes exception to NUREG-1801. The One-Time Inspection (B.2.3.19) AMP will verify the effectiveness of the Water Chemistry (B.2.3.2) AMP to manage loss of material.

Loss of material of cast iron heat exchanger components exposed to treated water in the AFWS is managed by the Water Chemistry (B.2.3.2) AMP, which takes exception to NUREG-1801. The One-Time Inspection (B.2.3.19) AMP will verify the effectiveness of the Water Chemistry (B.2.3.2) AMP to manage loss of material.

Furthermore, loss of material for steel heat exchanger components exposed to treated water in the FPS is managed by the Fire Water System (B.2.3.16) AMP. A generic note E and a plant-specific note are used.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 5 of 15 LRA Table 3.4-1 (page 3.4-22), is revised as follows:

Table 3.4-1: Summary of Aging Management Programs for Steam and Power Conversion Systems Aging Aging Further Evaluation Item Component Effect/Mechanis Management Discussion Recommended m Programs 3.4-1, Copper alloy, Loss of material Chapter XI.M2, No Consistent with exception to NUREG-1801.

016 Stainless steel, due to pitting and Water Chemistry, Loss of material of stainless steel piping, piping components, Nickel alloy, crevice corrosion and Chapter heat exchanger components and tubes exposed to treated Aluminum Piping, XI.M32, One-Time water, including steam, in the S&PC Systems is managed by piping Inspection the Water Chemistry (B.2.3.2) AMP, which takes exception to components, and NUREG-1801. The One-Time Inspection (B.2.3.19) AMP will piping elements, verify the effectiveness of the Water Chemistry (B.2.3.2) AMP Heat exchanger to manage loss of material.

components and tubes, PWR heat Loss of material of stainless steel piping and piping exchanger components exposed to treated water, including steam, in the components Chemical Feed System (specifically the pump casings and exposed to tanks) is managed by the Internal Surfaces in Miscellaneous Treated water, Piping and Ducting Components AMP. A generic note E and Steam plant-specific note are used.

Loss of material of stainless steel and nickel alloy steam generator (secondary side) components; stainless steel piping, piping components, and heat exchanger components exposed to treated water in the Control Room Ventilation System, CVCS, DRMWS, EDG and Auxiliary Systems, and Process and Effluent Radiation Monitoring and Sampling System is also managed by the Water Chemistry (B.2.3.2) AMP, which takes exception to NUREG-1801. The One-Time Inspection (B.2.3.19) AMP will verify the effectiveness of the Water Chemistry (B.2.3.2) AMP to manage loss of material.

There are no copper alloy, nickel alloy, or aluminum piping, piping components, or heat exchanger components exposed to treated water or steam in the S&PC Systems.

Furthermore, loss of material for copper alloy heat exchanger tubes and piping components in the FPS is managed by the Fire Water System (B.2.3.16) AMP. A generic note E and a plant-specific mnote are used.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 6 of 15 LRA Table 3.4-1 (page 3.4-23), is revised as follows:

Table 3.4-1: Summary of Aging Management Programs for Steam and Power Conversion Systems Aging Aging Further Evaluation Item Component Effect/Mechanis Management Discussion Recommended m Programs 3.4-1, Copper alloy Heat Reduction of heat Chapter XI.M2, No Not applicable.

017 exchanger tubes transfer due to Water Chemistry, There are no copper alloy heat exchanger tubes exposed to exposed to fouling and Chapter treated water in the S&PC Systems.

Treated water XI.M32, One-Time Inspection 3.4-1, Copper alloy, Reduction of heat Chapter XI.M2, No Consistent with exception to NUREG-1801.

018 Stainless steel transfer due to Water Chemistry, Reduction of heat transfer of stainless steel heat exchanger Heat exchanger fouling and Chapter tubes exposed to treated water in the AFWS is managed by tubes exposed to XI.M32, One-Time the Water Chemistry (B.2.3.2) AMP, which takes exception to Treated water Inspection NUREG-1801. The One-Time Inspection (B.2.3.19) AMP will verify the effectiveness of the Water Chemistry (B.2.3.2) AMP to manage fouling.

There are no copper alloy heat exchanger tubes exposed to treated water in the S&PC Systems.

Furthermore, reduction of heat transfer for copper alloy heat exchanger tubes in the FPS is managed by the Fire Water System (B.2.3.16) AMP. A generic note E and a plant-specific note are used.

3.4-1, Stainless steel, Loss of material Chapter XI.M20, No Not applicable.

019 Steel Heat due to general, "Open-Cycle There are no stainless steel or steel heat exchanger exchanger pitting, crevice, Cooling Water components exposed to raw water in the S&PC Systems.

components galvanic, and System" exposed to Raw microbiologically-water influenced corrosion; fouling that leads to corrosion

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 7 of 15 LRA Section A.2.2.16, (pages A-18 and A-19), is revised as follows:

A.2.2.16. Fire Water System The Fire Water System AMP is an existing condition and performance monitoring AMP that manages loss of material and flow blockage due to fouling for in-scope long-lived passive water-based fire suppression system components using periodic flow testing and visual inspections. The Fire Water System AMP will also manage loss of coating integrity for the fire water storage tanks (FWSTs), via periodic internal tank inspections. When visual inspections are used to detect loss of material and fouling, the inspection technique will be capable of detecting surface irregularities that could indicate wall loss due to corrosion, corrosion product deposition, and flow blockage due to fouling. There are no fire pump suction strainers/screens for the main fire protection (electric and diesel-driven) pumps, or foam water sprinkler systems within the scope of LR. Only the infrequently operated emergency fill pump for the FWSTs includes a suction strainer/screen.

Testing or replacement of sprinkler heads that have been in service for 50 years will be performed in accordance with the 2011 Edition of NFPA 25. Portions of the water-based fire water system that (a) are normally dry, but periodically subject to flow (e.g., dry-pipe or downstream of the deluge valve in a deluge system); and (b) cannot be drained or allow water to collect are subject to augmented examination beyond that specified in NFPA 25. The augmented examinations for the portions of normally dry piping that are periodically wetted or experience recurring internal corrosion include (a) periodic full flow tests at the design pressure and flow rate, or internal inspections; and (b) volumetric wall thickness evaluations.

Water system pressure is continuously monitored such that loss of pressure is detected and corrective action initiated. A drop of system pressure would cause automatic start of the electric fire pump and/or the emergency diesel driven fire pumps. Upon actuation, associated alarms would annunciate locally, in the FB, and in the Control Room, indicating pump start and potential loss of system pressure. Low level leakage from the fire water system would lead to a reduction in a FWST level and an associated low level alarm. Makeup to the affected FWST would be initiated by automatically opening an isolation valve from the treated water system which would close when the tank is refilled to an acceptable level. The frequency of these makeup revolutions and associated time that the makeup valves are open will be trended to determine overall leakage from the fire water system. Trending these revolutions is used in lieu of jockey pump monitoring because the jockey pump is usually in continuous operation with relief valves discharging back into the FWSTs.

The training and qualification of individuals involved in FWST coating inspections is conducted in accordance with ASTM International standards endorsed in RG 1.54, including limitations (if any) identified in RG 1.54 on a particular standard.

• Additional change associated with Attachment C

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 8 of 15 Fire pump diesel engine heat exchanger tubes are cleaned whenever the heat exchanger is opened for maintenance. The tubes, tubesheet, and channel head of the heat exchanger are cleaned and visually inspected based on maintenance history and at least once every ten years, continuing through the period of extended operation. Any indication of fouling or corrosion is evaluated. Conditions that do not meet acceptance criteria are entered into the CAP for evaluation.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 9 of 15 LRA Section A.4, Table A-3, (page A-60, A-61, and A-62), is revised as follows:

Aging Management NUREG-1801 No. Program or Activity Commitment Implementation Schedule Section (Section) 18 Fire Water System XI.M27 Continue the existing Fire Water System) AMP, including enhancements No later than 6 months prior to (A.2.2.16) to: the PEO, i.e.:

a) Inspect the fire water storage tank internal linings. The internal U1: 08/08/2029 linings will be inspected for blistering, cracking, flaking, peeling, U2: 08/02/2032, delamination, and rusting. The training and qualification of individuals involved in tank lining inspections and evaluation of or no later than the last refueling degraded conditions will be conducted in accordance with an outage prior to the PEO.

ASTM International standard endorsed in RG 1.54 including staff limitations associated with a particular standard. The following Perform the pre-PEO coating/lining acceptance criteria will be applied: inspections within the 5-year period prior to the PEO.

Indications of peeling and delamination are not acceptable.

Blisters will be evaluated by a qualified coating specialist.

Blisters should be limited to a few intact small blisters that are completely surrounded by sound coating/lining bonded to the substrate. Blister size and frequency should not be increasing between inspections (e.g., reference ASTM D714-02, Standard Test Method for Evaluating Degree of Blistering of Paints).

As applicable, wall thickness measurements, projected to the next inspection, meet design minimum wall requirements.

For fire water storage tank linings inspected by the procedure that do not meet acceptance criteria, appropriate corrective measures will be taken, consistent with LR-ISG-2013-01 Appendix C Element 7, with the exception of adhesion tests.

b) Ensure that visual inspections for loss of material use inspection techniques capable of detecting surface irregularities that could indicate an unexpected level of degradation due to corrosion and

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 10 of 15 corrosion product deposition. Where such irregularities are detected, follow-up volumetric wall thickness examinations will be performed.

c) Perform augmented tests and inspections on piping segments that cannot be drained or piping segments that allow water to collect.

In each 5-year interval, beginning 5 years prior to the PEO, either a flow test or flush sufficient to detect potential flow blockage will be conducted, or a visual inspection of 100 percent of the internal surface of piping segments that cannot be drained or piping segments that allow water to collect will be performed. In each 5-year interval of the PEO, 20 percent of the length of piping segments that cannot be drained or piping segments that allow water to collect will be subject to volumetric wall thickness inspections. Measurement points will be obtained to the extent that each potential degraded condition can be identified (e.g.,

general corrosion, MIC). The 20 percent of piping that is inspected in each 5-year interval will be in different locations than previously inspected piping. If the results of a 100-percent internal visual inspection are acceptable, and the segment is not subsequently wetted, no further augmented tests or inspections will be necessary. For portions of the normally dry piping that are configured to drain, the above augmented tests and inspections are not required.

d) Perform testing and visual inspections in accordance with Table 4a of LR-ISG-2012-02 Appendix L. This table is based on NFPA 25, 2011 edition. Unless recommended otherwise, external visual inspections are to be conducted on a refueling outage interval.

e) Update procedures to state that minimum design wall thickness must be maintained for in-scope fire protection piping.

f) Caulking or sealant shall be installed at the interface between the steel FWSTs and the respective concrete foundation ring.

The caulking/sealant will be visually inspected on a refueling outage interval with acceptance criteria of no drying, cracking, or missing caulking/sealant. Flaws in the caulking/sealant are repaired/replaced.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 11 of 15 g) Measure the tank bottom thickness of each FWST in accordance with LR-ISG-2012-02, Appendix L, using ultrasonic testing (UT) during the first 10-year period of the PEO. UT thickness measurements of the tank bottoms are evaluated against the design thickness and corrosion allowance. Inadequate bottom thickness results are evaluated in accordance with the site Corrective Action Program and, if required, repairs are implemented.

h) Following any activation of the electric motor driven vertical centrifugal fire pump, the suction strainer/screen shall be inspected and cleared of any debris or obstructions.

i) Perform volumetric examinations at five locations on the carbon steel aboveground fire water piping susceptible to recurring internal corrosion (RIC) on a refueling outage interval to identify loss of material. Continue these examinations until RIC, as defined by LR-ISG-2012-02, of the aboveground carbon steel fire suppression piping has been arrested. Additional locations will be examined if these volumetric examinations or plant operating experience identify significant degradation. For through-wall leaks and material loss greater than 50 percent of nominal wall, four additional locations will be examined. Where the identified material loss is 30 percent to 50 percent of nominal wall thickness and the calculated remaining life is less than two years, two additional locations will be examined.

j) Clean and visually inspect the tubesheet and channel head of each fire pump diesel engine heat exchanger based on maintenance history and at least once every ten years.

Evaluate any indication of fouling or corrosion. Enter conditions that do not meet acceptance criteria into the CAP for evaluation.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 12 of 15 LRA Section B.2.3.16, (page B-106), is revised as follows:

B.2.3.16 Fire Water System Program Description The Fire Water System AMP is an existing condition and performance monitoring AMP that manages loss of material and flow blockage due to fouling for in-scope long-lived passive water-based fire suppression system components using periodic flow testing and visual inspections. Reduction of heat transfer of diesel engine fire pump heat exchanger tubes is also managed using periodic visual inspections. In addition, the tubes are cleaned whenever the heat exchanger is opened for maintenance. The Fire Water System AMP will also manage loss of coating integrity for the FWSTs, via periodic internal tank inspections. When visual inspections are used to detect loss of material and fouling, the inspection technique will be capable of detecting surface irregularities that could indicate wall loss due to corrosion, corrosion product deposition, and flow blockage due to fouling. There are no fire pump suction strainers/screens for the main fire protection (electric and diesel-driven) pumps, or foam water sprinkler systems within the scope of LR. Only the infrequently operated emergency fill pump for the FWSTs includes a suction strainer/screen.

The Fire Water System AMP cleans and inspects the tubes of each fire pump diesel engine heat exchanger based on maintenance history and at least once every ten years. The AMP will also clean and inspect the tubesheet and channel head on the same frequency. Any indication of fouling or corrosion will be evaluated. Conditions that do not meet acceptance criteria are entered into the CAP for evaluation.

Testing or replacement of sprinkler heads that have been in service for 50 years will be performed in accordance with the 2011 Edition of NFPA 25. Portions of the water-based fire water system that (a) are normally dry, but periodically subject to flow (e.g., dry-pipe or downstream of the deluge valve in a deluge system) and (b) cannot be drained or allow water to collect are subject to augmented examination beyond that specified in NFPA 25. The augmented examinations for the portions of normally dry piping that are periodically wetted or experiencing recurring internal corrosion include (a) periodic full flow tests at the design pressure and flow rate, or internal inspections, and (b) volumetric wall thickness evaluations. These augmented tests and inspections are described in the enhancement section below.

A review of CPNPP operating experience identified that recurring internal corrosion has occurred in the fire water system. Therefore, recurring internal corrosion will be managed by this AMP.

Water system pressure is continuously monitored such that loss of pressure is detected and corrective action initiated. A drop of system pressure would cause automatic start of the electric main fire pump and/or the emergency diesel driven main fire pumps. Upon actuation, associated alarms would annunciate locally, in the FB, and in the Control Room, indicating pump start and potential loss of system pressure. Low level leakage from the fire water system would lead to a

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 13 of 15 reduction in a FWST level and an associated low level alarm. Makeup to the affected FWST would be initiated by automatically opening an isolation valve from the treated water system which would close when the tank refilled to an acceptable level. The frequency of these makeup revolutions and associated time that the makeup valves are open will be trended to determine overall leakage from the fire water system. Trending these revolutions is used in lieu of jockey pump monitoring because the jockey pump is usually in continuous operation with relief valves discharging back into the FWSTs.

Volumetric wall thickness inspections will be performed if visual inspections detect surface irregularities that could indicate an unexpected level of degradation due to corrosion and corrosion product deposition. If the presence of organic or inorganic material sufficient to obstruct piping or sprinklers is detected, the material will be removed, and the source determined.

The training and qualification of individuals involved in FWST coating inspections is conducted in accordance with ASTM International standards endorsed in RG 1.54, including limitations (if any) identified in RG 1.54 on a particular standard.

Program acceptance criteria include (a) the water-based fire protection system can maintain required pressure and flow rates, (b) minimum design wall thickness of piping and FWST bottoms is maintained, (c) no fouling exists in the sprinkler systems that could cause corrosion in the sprinklers and (d) no drying, cracking, or missing FWST caulking/sealant. Additionally, if the presence of sufficient foreign organic or inorganic material to obstruct pipe or sprinklers is detected during pipe inspections, the material is removed and its source is determined and corrected. In the event surface irregularities are identified, testing is performed to ensure minimum design pipe or tank wall thickness is maintained.

Caulking/sealant is repaired/replaced if defective or missing. In the event the fire water tank fails to meet the acceptance criteria for coating of the tank (e.g., peeling, delamination, blistering, flaking, cracking, or rust), the program requires corrective actions consistent with LR-ISG-2013-01 Appendix C Element 7.

With respect to the fire pump diesel engine heat exchangers, any indication of fouling or corrosion will be evaluated, and conditions that do not meet acceptance criteria will be entered into the CAP for evaluation.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 14 of 15 LRA Section B.2.3.16, (page B-109), is revised as follows:

Element Affected Enhancement Effected

1. Scope A new procedure will be created to inspect the fire water storage tank internal linings. The internal linings will be inspected for blistering,

3. Parameters cracking, flaking, peeling, delamination, and rusting. The training and Monitored or qualification of individuals involved in tank lining inspections and Inspected evaluation of degraded conditions will be conducted in accordance with an ASTM International standard endorsed in RG 1.54 including staff

4. Detection of limitations associated with a particular standard. The procedure will Aging Effects include the following coating/lining acceptance criteria:

6. Acceptance a. Indications of peeling and delamination are not acceptable.

Criteria

b. Blisters will be evaluated by a qualified coating specialist.

7. Corrective Actions c. Blisters should be limited to a few intact small blisters that are completely surrounded by sound coating/lining bonded to the substrate. Blister size and frequency should not be increasing between inspections (e.g., reference ASTM D714-02, Standard Test Method for Evaluating Degree of Blistering of Paints).

d. As applicable, wall thickness measurements, projected to the next inspection, meet design minimum wall requirements.

For fire water storage tank linings inspected by the procedure that do not meet acceptance criteria, appropriate corrective measures will be taken, consistent with LR-ISG-2013-01 Appendix C Element 7, with the exception of adhesion tests.

1. Scope Clean and inspect the tubesheet and channel head of each fire pump diesel engine heat exchanger based on maintenance history

4. Detection of and at least once every ten years. Evaluate any indication of Aging Effects fouling or corrosion. Enter conditions that do not meet acceptance criteria into the CAP for evaluation.

6. Acceptance Criteria

7. Corrective Actions

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-1 TXX-23048 and CP-202300322 Attachment A Page 15 of 15 Element Affected Enhancement Effected

2. Preventive Caulking or sealant shall be installed at the interface between the Actions steel FWSTs and the respective concrete foundation ring. The caulking/sealant will be visually inspected on a refueling outage

3. Parameters interval with acceptance criteria of no drying, cracking, or missing Monitored or caulking/sealant. Flaws in the caulking/sealant are Inspected repaired/replaced.

4. Detection of Aging Effects

6. Acceptance Criteria

7. Corrective Actions

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 1 of 8 LRA Section: B.2.3.16, Fire Water System NRC RAI No: B.2.3.16-2 (Fire Water Storage Tank Bottoms)

Regulatory Basis:

Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S.

Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described in the requests for additional information.

Background:

The Preventative Actions program element for GALL Report AMP XI.M29, Aboveground Metallic Tanks, in Appendix M of License Renewal (LR) Interim Staff Guidance (ISG), LR-ISG-2012-02, Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion under Insulation (ML13227A361), states, in part, For outdoor tanks sealant or caulking is applied at the external interface between the tank and concrete or earthen foundation to mitigate corrosion of the bottom surface of the tank by minimizing the amount of water and moisture penetrating the interface, which could lead to corrosion of the bottom surface.

LRA Supplement 1 (ML23096A302) revised Table A-3 in LRA Appendix A and Section B.2.3.16 in LRA Appendix B to add the following enhancement: Caulking or sealant shall be installed at the interface between the steel FWSTs and the respective concrete foundation ring. The caulking/sealant will be visually inspected on a refueling outage interval with acceptance criteria of no drying, cracking, or missing caulking/sealant. Flaws in the caulking/sealant are repaired/replaced.

The Description of the Program for GALL Report AMP XI.M29 in Table 3.0-1 of Appendix J in LR-ISG-2012-02 states, "External visual examinations are sufficient to monitor degradation of the protective paint, coating, and caulking, or sealant (when supplemented with physical manipulation), or uncoated surfaces." However, the enhancement added in LRA Supplement 1 did not discuss whether the visual examinations would be supplemented with physical manipulation.

The Acceptance Criteria program element for GALL Report AMP XI.M29 in Appendix M of LR-ISG-2012-02 state, Drying, cracking, or missing sealant and caulking are unacceptable

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 2 of 8 and need to be evaluated using the corrective action program. The evaluation will determine the need to repair the sealant and caulking." However, the enhancement added in LRA Supplement 1 did not discuss unacceptable sealant or caulking inspections being evaluated using the corrective action program. In addition, because flaws in sealant and caulking can allow intrusion of moisture and potential degradation, the staff notes that the Technical Bases for Changes column in Table 2-29 of NUREG-2221, "Technical Bases for Changes in the Subsequent License Renewal Guidance Documents NUREG-2191 and NUREG-2192" (ML17362A126), states, in part, "...the corrective action should include a determination for the need for examinations [of the tank's surface]..." which would be applicable for the initial period of extended operation if sealant or caulking were found to have flaws.

Issue:

LRA Supplement 1 did not discuss whether the condition of the FWST bottoms would be assessed prior to installing caulking or sealant at the interface between the steel FWSTs and the respective concrete foundation ring. The staff notes that the current absence of caulking or sealant may have allowed water and moisture to penetrate the interface, which could lead to corrosion of the bottom surface.

LRA Supplement 1 did not discuss whether the visual examinations of the caulking or sealant would be supplemented with physical manipulation as noted in Appendix J of LR-ISG-2012-02.

Also, LRA Supplement 1 did not discuss unacceptable sealant or caulking inspections being evaluated using the corrective action program as noted in Appendix M of LR-ISG-2012-02, including whether the evaluation would include a determination for the need for examinations of the tank's surface.

Request:

Please address the following:

1. Will the condition of the FWST bottoms be assessed prior to installing caulking or sealant? If so, how will the FWST bottoms be assessed? If corrosion of the FWST bottoms is detected, how may the FWST inspections change for the period of extended operation (e.g., more frequent FWST bottom wall thickness measurements)?

2. Was physical manipulation of the sealant or caulking considered as a supplement to the visual inspections?

3. If degraded sealant or caulking is detected, will they be evaluated using the corrective action program, including a determination for the need for examinations of the tank's surface?

Luminant Response:

1. The timing of the installation of caulk or sealant is not tied to the timing of this inspection and the caulk or sealant may or may not be applied before this inspection is performed. In accordance with LRA Section B.2.3.16, enhancement for NFPA 25 Section 9.2.7, the internal inspection of the tank bottom will include UT and vacuum-box testing in accordance

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 3 of 8 with the guidance of LR-ISG-2012-02 Appendix L. In accordance with LR-ISG-2012-02 Appendix L guidance for elements 7, 8, and 9, inadequate results, such as tank bottom minimum wall thickness violations, will be evaluated in accordance with the site Corrective Action Program and, if required, corrective actions including confirmation processes and administrative controls will be implemented.

2. The visual inspections of the caulking or sealant will be supplemented with physical manipulation to ensure that any potential degradation is identified. Revisions to LRA Sections A.4 and B.2.3.16 are included below for clarity.

3. If degraded sealant or caulking is detected, the condition will be evaluated using the corrective action program, in accordance with the guidance of LR-ISG-2012-02 Appendix L element 7. The evaluation will include a determination of whether inspection of the tank surface is warranted. Revisions to LRA Sections A.4 and B.2.3.16 are included below to clarify that the sites corrective action program is used.

References:

1. LR-ISG-2012-02, NRC License Renewal Interim Staff Guidance, Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion Under Insulation

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 4 of 8 Associated LRA Revisions:

LRA Section A.4, Table A-3 (portion of Commitment #18 on page A-62), is revised as follows:

Table A-3 List of LR Commitments and Implementation Schedule Aging Management Implementation No. NUREG-1801 Program or Activity Commitment Schedule Section (Section) or piping segments that allow water to collect will be performed. In each 5-year interval of the PEO, 20 percent of the length of piping segments that cannot be drained or piping segments that allow water to collect will be subject to volumetric wall thickness inspections. Measurements points will be obtained to the extent that each potential degraded condition can be identified (e.g., general corrosion, MIC). The 20 percent of piping that is inspected in each 5-year interval will be in different locations than previously inspected piping. If the results of a 100-percent internal visual inspection are acceptable, and the segment is not subsequently wetted, no further augmented tests or inspections will be necessary. For portions of the normally dry piping that are configured to drain, the above augmented tests and inspections are not required.

d) Perform testing and visual inspections in accordance with Table 4a of LR-ISG-2012-02 Appendix L. This table is based on NFPA 25, 2011 edition. Unless recommended otherwise, external visual inspections are to be conducted on a refueling outage interval.

e) Update procedures to state that minimum design wall thickness must be maintained for in-scope fire protection piping.

f) Caulking or sealant shall be installed at the interface between the steel FWSTs and the respective concrete foundation ring. The caulking/sealant will be visually inspected and physically manipulated on a refueling outage

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 5 of 8 Table A-3 List of LR Commitments and Implementation Schedule Aging Management Implementation No. NUREG-1801 Program or Activity Commitment Schedule Section (Section) interval with acceptance criteria of no drying, cracking, or missing caulking/sealant. Flaws in the caulking/sealant are evaluated in the Corrective Action Program and are repaired/replaced accordingly.

g) Measure the tank bottom thickness of each FWST in accordance with LR-ISG-2012-02, Appendix L, using ultrasonic testing (UT) during the first 10-year period of the PEO. UT thickness measurements of the tank bottoms are evaluated against the design thickness and corrosion allowance. Inadequate bottom thickness results are evaluated in accordance with the site Corrective Action Program and, if required, repairs are implemented.

h) Following any activation of the electric motor driven vertical centrifugal fire pump, the suction strainer/screen shall be inspected and cleared of any debris or obstructions.

i) Perform volumetric examinations at five locations on the carbon steel aboveground fire water piping susceptible to recurring internal corrosion (RIC) on a refueling outage interval to identify loss of material. Continue these examinations until RIC, as defined by LR-ISG-2012-02, of the aboveground carbon steel fire suppression piping has been arrested. Additional locations will be examined if these volumetric examinations or plant operating experience identify significant degradation. For through-wall leaks and material loss greater than 50 percent of nominal wall, four additional locations will be examined. Where the identified material loss is 30 percent to 50 percent of nominal wall thickness and the calculated remaining life is

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 6 of 8 Table A-3 List of LR Commitments and Implementation Schedule Aging Management Implementation No. NUREG-1801 Program or Activity Commitment Schedule Section (Section) less than two years, two additional locations will be examined.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 7 of 8 LRA Section B.2.3.16, (page B-109), is revised as follows:

Element EAffected Enhancement

1. Scope A new procedure will be created to inspect the fire water storage tank

3. Parameters internal linings. The internal linings will be inspected for blistering, Monitored or cracking, flaking, peeling, delamination, and rusting. The training and Inspected qualification of individuals involved in tank lining inspections and evaluation of degraded conditions will be conducted in accordance with

4. Detection of an ASTM International standard endorsed in RG 1.54 including staff Aging Effects limitations associated with a particular standard. The procedure will

6. Acceptance include the following coating/lining acceptance criteria:

Criteria

a. Indications of peeling and delamination are not acceptable.

7. Corrective

b. Blisters will be evaluated by a qualified coating specialist.

Actions

c. Blisters should be limited to a few intact small blisters that are completely surrounded by sound coating/lining bonded to the substrate. Blister size and frequency should not be increasing between inspections (e.g., reference ASTM D714-02, Standard Test Method for Evaluating Degree of Blistering of Paints).

d. As applicable, wall thickness measurements, projected to the next inspection, meet design minimum wall requirements.

For fire water storage tank linings inspected by the procedure that do not meet acceptance criteria, appropriate corrective measures will be taken, consistent with LR-ISG-2013-01 Appendix C Element 7, with the exception of adhesion tests.

2. Preventive Caulking or sealant shall be installed at the interface between the Actions steel FWSTs and the respective concrete foundation ring. The

3. Parameters caulking/sealant will be visually inspected and physically Monitored or manipulated on a refueling outage interval with acceptance criteria Inspected of no drying, cracking, or missing caulking/sealant. Flaws in the caulking/sealant are evaluated in the Corrective Action Program

4. Detection of and are repaired/replaced accordingly.

Aging Effects

6. Acceptance Criteria

7. Corrective Actions

3. Parameters Measure the tank bottom thickness of each FWST in accordance Monitored or with LR-ISG-2012-02, Appendix L, using ultrasonic testing (UT)

Inspected during the first 10-year period of the PEO. UT thickness

4. Detection of measurements of the tank bottoms are evaluated against the Aging Effects design thickness and corrosion allowance. Inadequate bottom thickness results are evaluated in accordance with the site

6. Acceptance Corrective Action Program and, if required, repairs are Criteria implemented.

7. Corrective Actions

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-2 TXX-23048 and CP-202300322 Attachment B Page 8 of 8 LRA Section B.2.3.16, (page B-113), is revised as follows:

Water Storage Tanks Exterior Inspections 9.2.5.5 Current Compliance None Required Enhancements A new procedure will be created to perform visual inspections of the fire water storage tanks and supporting structures painted or coated exterior surfaces for signs of degradation on an annual interval. This inspection will be in accordance with NFPA 25, Section 9.2.5.5. The inspection will include inspection of the exterior tank coating for protective coating degradation. If degradation is identified, follow-up volumetric examinations will be performed to ensure wall thickness is equal to or exceeds nominal wall thickness.

The FWST to concrete interface caulking/sealant will be visually inspected and physically manipulated on a refueling outage interval with acceptance criteria of no drying, cracking, or missing caulking/sealant. Flaws in the caulking/sealant are evaluated in the Corrective Action Program and are repaired/replaced accordingly.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-3 TXX-23048 and CP-202300322 Attachment C Page 1 of 6 LRA Section: B.2.3.16, Fire Water System NRC RAI No: B.2.3.16-3 (Coating Corrective Actions)

Regulatory Basis:

Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S.

Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described in the requests for additional information.

Background:

License Renewal (LR) Interim Staff Guidance (ISG), LR-ISG-2013-01, Aging Management of Loss of Coating or Linting Integrity for Internal Coatings/Linings on In-Scope Piping, Piping Components, Heat Exchangers, and Tanks (ML14225A059), added AMP XI.M42, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks. In Appendix C of LR-ISG-2013-01, the Scope of Program element for AMP XI.M42 states, The aging effects associated with fire water tank internal coatings/linings are managed by GALL Report AMP XI.M27, Fire Water System, instead of this AMP. However, where the fire water storage tank internals are coated, the Fire Water System Program and FSAR [Final Safety Analysis Report] Summary Description of the Program should be enhanced to include the recommendations associated with training and qualification of personnel and the corrective actions program element.

The Corrective Actions program element in Appendix C of LR-ISG-2013-01 includes the corrective actions that the Fire Water System AMP should be enhanced to include.

Table A-3 in LRA Appendix A and Section B.2.3.16 of LRA Appendix B include an enhancement to the Fire Water System AMP to include recommendations from AMP XI.M42 related to corrective actions.

The last sentence in the sixth paragraph under Program Description in Section B.2.3.16 of LRA Appendix B states, In the event the fire water tank fails to meet the acceptance criteria for coating or the tank (e.g., peeling, delamination, blistering, flaking, cracking, or rust), the

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-3 TXX-23048 and CP-202300322 Attachment C Page 2 of 6 program requires corrective actions consistent with LR-ISG-2013-01 Appendix C Element 7.

Issue:

The FSAR summary description of the Fire Water System program in LRA Appendix A, Section A.2.2.16 does not include the recommendations associated with the Corrective Actions program element in AMP XI.M42, as provided in LR-ISG-2013-01.

During the Fire Water System program audit, it was noted that the corrective action statement was omitted from Section A.2.2.16 in LRA Appendix A because it is an enhancement in Table A-3 in LRA Appendix A.

The staff notes that this appears to be a different approach than that taken for the recommendations associated with training and qualification of personnel. Consistent with LR-ISG-2013-01, the applicant included the recommendations associated with training and qualification of personnel in both Section A.2.2.16 in LRA Appendix A and Section B.2.3.16 of LRA Appendix B, even though there is an enhancement. As previously noted, the applicant included the recommendations associated with corrective actions in Section B.2.3.16 of LRA Appendix B, even though there is an enhancement.

The staff notes that guidance in SRP-LR for FSAR Supplements (e.g., 3.3.2.5) states that the program description should be sufficiently comprehensive (i.e., contain future aging management activities, including enhancements and commitments), such that later changes can be controlled by 10 CFR 50.59. The staff also notes that including the information in the FSAR summary description now, will ensure the information is not inadvertently omitted if the commitments in LRA Appendix A, Table A-3 are deleted once the enhancements are implemented.

Request:

If the FSAR summary description of the Fire Water System program (Section A.2.2.16 in LRA Appendix A) will not include the corrective action information, as prescribed in LR-ISG-2013-01, please provide information regarding the constraints to ensure that later changes to the program, associated with enhancements contained in LRA Appendix A, Table A-3, can be controlled by 10 CFR 50.59.

Luminant Response:

The corrective action statement, with respect to fire water storage tank internal coatings, was omitted from Section A.2.2.16 in LRA Appendix A because it is an enhancement in Table A-3 in LRA Appendix A. However, LR-ISG-2013-01 Appendix D states, in part:

The aging effects associated with fire water tank internal coatings/linings are managed by this AMP (XI.M27) in lieu of AMP XI.M42, Internal Coatings/Linings for In-Scope Piping, Piping Components, Heat Exchangers, and Tanks. However, where the fire water storage tank internals, are coated, the Fire Water System Program and FSAR

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-3 TXX-23048 and CP-202300322 Attachment C Page 3 of 6 Summary Description of the Program should be enhanced to include the recommendations associated with training and qualification of personnel and the corrective actions program element of AMP XI.M42.

Therefore, LRA Section A.2.2.16 is revised to explicitly include the fire water storage tank coating corrective actions that will be taken once the associated enhancement is implemented. Additionally, the phrasing with respect to training and qualification of individuals involved in FWST coating inspections is revised to clarify it is an enhancement.

In addition, and separate from this RAI, LRA Section 3.3.3, Table 3.3-1, Table Item 3.3-1, 128 is revised to correct an editorial error removing a duplicated statement within the Discussion column.

References:

1. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-3 TXX-23048 and CP-202300322 Attachment C Page 4 of 6 Associated LRA Revisions:

LRA Table 3.3-1, (page 3.3-87), is revised as follows:

Table 3.3-1 Summary of Aging Management Programs for Auxiliary Systems Item Number Component Aging Aging Management Further Evaluation Discussion Effect/Mechanism Programs Recommended 3.3-1, 128 Steel, stainless steel, or Loss of material due Chapter XI.M29, No Consistent with exception to NUREG-aluminum tanks (within to general (steel Aboveground Metallic 1801, as supplemented by LR-ISG-the scope of Chapter only), pitting, or Tanks 2012-02.

XI.M29, Aboveground crevice corrosion; As described in Sections 2.3.3.7 and Metallic Tanks) exposed cracking due to 2.4.11, the carbon steel FWSTs in the to soil or concrete, or the stress corrosion FPS are located outdoors with a sand following external cracking (stainless (soil) and concrete foundation. The environments steel and aluminum Fire Water System (B.2.3.16) AMP, air-outdoor, air-indoor only) which takes exception to uncontrolled, moist air, NUREG-1801, includes enhancement condensation to address the steel/concrete interface. A generic note E and plant-specific note are used.

Loss of material for the steel FWSTs exposed to outdoor air is addressed in item 3.3-1, 136 below.

Other than the RMWSTs, FOSTs, and FWSTs, metallic tanks in the auxiliary systems are located indoors and not exposed to soil or concrete. Also, tank capacities are less than 100,000 gallons.

As described in Section 2.4.11, the RMWST in the Demineralized and Reactor Water Makeup System is a reinforced concrete tank. Aging

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-3 TXX-23048 and CP-202300322 Attachment C Page 5 of 6 Table 3.3-1 Summary of Aging Management Programs for Auxiliary Systems Item Number Component Aging Aging Management Further Evaluation Discussion Effect/Mechanism Programs Recommended management review of the concrete is listed in Table 3.5.2-11.

As described in Section 2.3.3.5, the carbon steel FOST is buried in soil.

Loss of material for the steel surface exposed to soil is addressed in item 3.3-1, 106.

Loss of material for the steel exposed to outdoor air is addressed in item 3.3-1, 136 below.

Therefore, there are no CPNPP tanks in the auxiliary systems that fit the scope of NUREG-1801, AMP XI.M29, Aboveground Metallic Tanks, as supplemented by LR-ISG-2012-02.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-3 TXX-23048 and CP-202300322 Attachment C Page 6 of 6 LRA Section A.2.2.16, (page A-19), is revised as follows:

corrosion include (a) periodic full flow tests at the design pressure and flow rate, or internal inspections; and (b) volumetric wall thickness evaluations.

Water system pressure is continuously monitored such that loss of pressure is detected and corrective action initiated. A drop of system pressure would cause automatic start of the electric fire pump and/or the emergency diesel driven fire pumps. Upon actuation, associated alarms would annunciate locally, in the FB, and in the Control Room, indicating pump start and potential loss of system pressure. Low level leakage from the fire water system would lead to a reduction in a FWST level and an associated low level alarm. Makeup to the affected FWST would be initiated by automatically opening an isolation valve from the treated water system which would close when the tank refilled to an acceptable level. The frequency of these makeup revolutions and associated time that the makeup valves are open will be trended to determine overall leakage from the fire water system. Trending these revolutions is used in lieu of jockey pump monitoring because the jockey pump is usually in continuous operation with relief valves discharging back into the FWSTs.

For the internal coated/lined surfaces of the FWST determined to not meet acceptance criteria, physical testing will be performed where physically possible (i.e., sufficient room to conduct testing) in conjunction with repair or replacement of the coating/lining, with the exception of adhesion testing.

Regardless of physical constraints lightly tapping the coating/lining surrounding a blister may be performed as an alternative to adhesion testing in order to determine whether the remaining coating/lining is tightly bonded to the base metal.

The training and qualification of individuals involved in FWST coating inspections iswill be conducted in accordance with ASTM International standards endorsed in RG 1.54, including limitations (if any) identified in RG 1.54 on a particular standard.

• Additional change associated with Attachment A

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4 TXX-23048 and CP-202300322 Attachment D Page 1 of 6 LRA Section: B.2.3.16, Fire Water System NRC RAI No: RAI B.2.3.16-4 Fire Water System AMP (Inspection and Testing)

Regulatory Basis:

Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S.

Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described in the requests for additional information.

Background:

Table 4a, Fire Water System Inspection and Testing Recommendations, in Appendix D of LR- ISG-2012-02, recommends that operational testing of water spray fixed systems follow Section 10.3.4.3, Discharge Patterns, of National Fire Protection Association (NFPA) 25, Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems. Section 10.3.4.3 of NFPA 25 requires (1) the water discharge patterns from open spray nozzles be observed, (2) testing with air if water cannot be discharged on protected property, and (3) cleaning and retesting of nozzles if obstructions occur.

Section B.2.3.16 in LRA Appendix B includes an exception related to deluge systems that cannot be tested with water and have no provisions to be tested with air. For Section 10.3.4.3 of NFPA 25, the Fire Water System Inspections and Tests table in Section B.2.3.16 in LRA Appendix B that provides additional detail on the enhancements states that no enhancements are required and Due to the nature of the protected property such that water cannot be discharged, in addition to the inspections above, nozzles are tested with air to ensure the nozzles are not obstructed. This appears to indicate that all the operational tests performed in accordance with Section 10.3.4.3 of NFPA 25 are performed with air.

Issue:

The staff noted that Revision 1 of LUM00020-REPT-053, Fire Water System Aging Management Program Basis Document, states water cannot be discharged through some spray nozzles when referring to the operational tests. This appears to indicate that water may be discharged through some spray nozzles.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4 TXX-23048 and CP-202300322 Attachment D Page 2 of 6 Therefore, it is unclear whether there are operational tests performed in accordance with Section 10.3.4.3 of NFPA 25 that are performed with water.

GALL Report AMP XI.M27 in Appendix L of LR-ISG-2012-02 states, Results of flow testing (e.g., buried and underground piping, fire mains, sprinkler) are monitored and trended.

However, during the audit of the Fire Water System AMP, the applicant stated that sprinkler system inspections are not trended because there are no applicable items to trend.

Request:

Please provide information discussing whether some operational tests performed in accordance with Section 10.3.4.3 of NFPA 25 are performed with water. If so, please provide information discussing whether items such as pressure, discharge time, deposits, etc. could be monitored and trended to provide timely indication of changes in the system that could impact its ability to perform its intended function.

Luminant Response:

As stated in LRA Appendix B, Section B.2.3.16, Fire Water System Inspections and Tests Table Verification of spray nozzle discharge patterns and obstructions is addressed under Operational Tests below. As noted within that section, due to the nature of the protected property water cannot be discharged through some spray nozzles. In these instances, nozzles are inspected for correct orientation and the system is tested with air to ensure that the nozzles are not obstructed. In all cases where deluge valves are not trip tested at full flow, nozzles are air tested, except for charcoal filter deluge valves which are addressed below.

Those fire water deluge systems that are performed with water are the transformer fire protection systems which receive full flow testing. Those fire protection systems where water cannot be discharged applies to certain locations that allowing such a discharge is not practical due to possible water damage and cleanup of the components and areas (e.g., main generator bearing fire protection, Turbine lube oil fire protection, Main Feedwater pump fire protection). These systems are tested with air to ensure that the nozzles are not obstructed and inspected for correct orientation per the applicable test procedure. Charcoal filter fire protection system testing is documented as an exception and utilizes other inspection methods for the verification of spray nozzle discharge orientation and obstruction.

Also, as described in LRA Appendix B, Section B.2.3.16, the Fire Water System AMP manages loss of material and flow blockage due to fouling for in-scope long-lived passive water-based fire suppression system components using periodic flow testing and visual inspections. If existing procedures do not provide the requirements to monitor and trend pressure and flow test results, enhancements have been added where required to the Fire Water System AMP.

LRA Appendix B, Section B.2.3.16 Fire Water System Inspections and Tests Table Operational Tests is revised, in order to further clarify some deluge system operational tests are performed with water, those that cannot be performed with water are tested with air, and also the deluge systems for the filter units that cannot be tested with either water or air.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4 TXX-23048 and CP-202300322 Attachment D Page 3 of 6 Additionally, it is revised to provide procedural enhancements for monitoring and trending of results including pump performance, run and discharge time, pressure and other aspects that would give indication of aging such as deposits or sediment.

References:

LR-ISG-2012-02 Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks and Corrosion Under Insulation

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4 TXX-23048 and CP-202300322 Attachment D Page 4 of 6 Associated LRA Revisions:

LRA Appendix B, Section B.2.3.16, page B-117, is revised as follows:

Description NFPA 25 Current Compliance and Required Enhancements Section(s)

Water Spray Fixed Systems Strainers 10.2.1.6, Current Compliance 10.2.1.7, 10.2.7 Plant procedures require periodic strainer maintenance, with instructions provided for strainer maintenance/replacement.

Plant procedures perform mainline strainer maintenance every 5 years, ensuring strainers are removed and inspected for damaged or corroded parts.

Required Enhancements Procedures performing periodic system flow tests will be enhanced to ensure strainers are flushed after each flow test, until the flow stream is clear with no observed entrained debris.

Operational 10.3.4.3 Current Compliance Tests Plant procedures require visual inspection of headers/ nozzle spray patterns every 18 months. In addition, plant procedures also require trip tests of deluge system valves every 18 months.

Plant procedures perform testing/inspections of deluge spray nozzles and ensure that water discharge patterns are not impeded by plugged nozzles, ensure nozzles are correctly positioned, and ensure obstructions do not prevent discharge patterns from wetting surfaces to be protected. Deluge systems which receive full flow testing are the transformer fire protection systems.

Due to the nature of the protected property such that water cannot be discharged through spray nozzles in some systems, in addition to the inspections above, those nozzles are tested with air to ensure the nozzles are not obstructed (e.g., main generator bearing fire protection, Turbine lube oil fire protection, Main Feedwater pump fire protection).

Deluge systems for the Containment Preaccess Filtration System charcoal filter units (2 per Unit) and the Primary Plant Ventilation Engineered Safety Feature (ESF) filter units (18 total), which cannot be tested with water and have no provisions to perform an air test to verify that the spray

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4 TXX-23048 and CP-202300322 Attachment D Page 5 of 6 Description NFPA 25 Current Compliance and Required Enhancements Section(s) openings are not obstructed, are addressed in the above Deluge Valve (NFPA 25 Section 13.4.3.2.2 through 13.4.3.2.5) row.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4 TXX-23048 and CP-202300322 Attachment D Page 6 of 6 LRA Appendix B, Section B.2.3.16, page B-118, is revised as follows:

Description NFPA 25 Current Compliance and Required Enhancements Section(s)

Required Enhancements None. Procedures performing periodic testing of deluge systems will be enhanced to ensure that test results are monitored and trended and include monitoring pump performance, run and discharge time, pressure and other aspects that would give indication of aging such as deposits or sediment.

Foam Water Sprinkler Systems Strainers, 11.2.7.1 N/A. CPNPP does not use foam water sprinkler systems Operational Tests, and 11.3.2.6 Storage Tanks Obstruction Investigation Obstruction, 14.2, 14.3 Current Compliance internal inspection of None piping Required Enhancements A new implementing document will be prepared, or an existing implementing document will be revised to incorporate the instructions and requirements for internal inspection of piping and obstruction investigation from NFPA 25, Sections 14.2, 14.3, and subsections. The alternative nondestructive examination methods permitted by NFPA 25 Sections 14.2.1.1 and 14.3.2.3 will be limited to those that can ensure that flow blockage will not occur. Existing procedures will also be enhanced to require obstruction investigations, if the presence of sufficient foreign organic or inorganic material to obstruct pipe or sprinklers is detected during pipe inspections.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 1 of 27 LRA Section: 3.2.2.2.3, Loss of Material due to Pitting and Crevice Corrosion NRC RAI No: 3.2.2.2.3.2-1 (Stainless Steel Components Exposed to Outdoor Air)

Regulatory Basis:

10 CFR 54.21(a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information about the matter described in the request for additional information.

Background:

The following Table 1 items address loss of material (LOM) by pitting and crevice corrosion, or stress corrosion cracking (SCC), for stainless steel exposed to outdoor air in Engineered Safety Features, Auxiliary, and Steam and Power Conversion Systems:

3.2-1, 004 3.3-1, 004 3.4-1, 002 3.2-1, 007 3.3-1, 006 3.4-1, 003 License Renewal Application (LRA) further evaluation Sections 3.2.2.2.3.2, 3.3.2.2.5, and 3.4.2.2.3 identify select instances for which LOM due to pitting and crevice corrosion are considered possible and managed for the Table 1 items listed above using the External Surfaces Monitoring of Mechanical Components Aging Management Program. These instances are for areas of the Condensate Storage Tank, Reactor Makeup Water Storage Tank, and Refueling Water Storage Tank. The discussions for these instances note that contaminants in outdoor air could collect on these component surfaces. Other than for these select instances, however, these Table 1 items are identified as aging effects not requiring management. This is based on location criteria for judging the amount of halide likely to be present, along with having no expectation of the halide amount increasing. Three aging management review items associated with item 3.4-1, 003 cite generic note I (aging effect not applicable). The items addressing SCC (3.2-1, 007; 3.3-1, 004; 3.4-1, 002) are all identified as not applicable.

Issue:

Guidance for initial license renewal (LR), such as NUREG-1800 (Reference 1), allows an applicant to conclude that managing LOM or SCC is not applicable based on the components proximity to outdoor air with the potential for halogens to be present (e.g.,

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 2 of 27 distance from a saltwater coastline or highway treated with salt). However, this was changed in more recent guidance for subsequent license renewal (SLR), which can be considered as relevant operating experience during the first period of extended operation, rather than specific to a subsequent period of extended operation. As described in supplemental guidance for SLR (Reference 2), the staff concluded that the LR screening may not account for all relevant factors. The staff therefore concluded that the most accurate and practical method for determining the susceptibility to LOM and SCC of these materials to the plant-specific environments is to review the available plant-specific operating experience and perform a one-time inspection. In addition to the changes to guidance for stainless steel, the staff explained in NUREG-2221 (Reference 3) that LOM due to pitting and crevice corrosion should also be managed for nickel alloys exposed to air.

Given these changes to the guidance, the staff seeks further clarification on the basis for excluding LOM due to pitting corrosion, crevice corrosion, and SCC as aging effects requiring management at Comanche Peak for stainless steel components exposed to outdoor air. The staff also seeks clarification on whether there are nickel alloy components exposed to outdoor air in Engineered Safety Features Systems, Auxiliary Systems, or Steam and Power Conversion Systems that would be susceptible to LOM based on the further evaluation applied to stainless steel.

Request:

1. With respect to LRA Sections 3.2.2.2.3.2, 3.3.2.2.5, and 3.4.2.2.3, please state the basis for concluding that LOM due to pitting and crevice corrosion for stainless steel in outdoor air is not an applicable aging effect requiring management, other than for the instances identified in the LRA, considering the changes in the staff guidance under SLR for these reviews. Alternatively, if appropriate, revise the LRA to reflect that that LOM due to pitting and crevice corrosion will be managed for additional components.

2. With respect to LRA Sections 3.2.2.2.6, 3.3.2.2.3, and 3.4.2.2.2, please state the basis for not treating SCC of stainless steel in outdoor air as an applicable aging effect in cases for which LOM due to pitting or crevice corrosion is applicable. Alternatively, if appropriate, revise the LRA to include SCC as an applicable aging effect for these components.

3. Please discuss whether aging management for LOM due to pitting and crevice corrosion for nickel alloy components exposed to outdoor air is applicable considering the guidance for stainless steel in LRA Sections 3.2.2.2.3.2, 3.3.2.2.5, and 3.4.2.2.3 and the changes to this guidance for SLR. If appropriate, revise the LRA to reflect that this aging effect will be managed for nickel alloys.

Luminant Response:

1. Loss of Material due to pitting and crevice corrosion for stainless steel in outdoor air:

There are a limited number of stainless steel components exposed to outdoor air at CPNPP for which loss of material is not identified as an applicable aging effect. The LRA had conservatively identified certain components associated with Containment Spray System

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 3 of 27 RWST, Demineralized and Reactor Makeup Water System RMWST, and Auxiliary Feedwater System CST as exposed to outdoor air. However, upon further review it was determined that these components are located within the piping tunnels associated with these tanks, and are therefore not exposed to an outdoor air environment. The remaining stainless steel components associated with these tanks exposed to outdoor air are tank liners and internal tank vents above the water line. Loss of material is an applicable aging effect for these components due to periodic wetting that could increase halogens/contaminants above the non-aggressive ambient air. The LRA is revised based on these findings.

In consideration of SLR OE and the above, the basis for concluding LOM for uninsulated stainless steel in outdoor air is not an applicable aging effect (except in select instances) is as follows:

a) Stainless steel components exposed to outdoor air within the scope of license renewal where loss of material is not identified as an applicable aging effect are limited to Fire Protection System flow elements and orifices.

b) As discussed in CPNPP LRA further evaluation sections, where contaminants could concentrate, due to periodic wetting that could increase halogens/contaminants above the non-aggressive ambient air, SLR OE is considered, and LOM is conservatively identified as an applicable aging effect managed by the External Surfaces Monitoring of Mechanical Components AMP or Bolting Integrity AMP, regardless of halide concentrations.

c) SLR guidance indicates that air environments can be aggressive if halides are present. CPNPP is not near a source of halides or other contaminants that could lead to accelerated local corrosion of stainless steel components located outdoors.

CPNPP is located a significant distance from any saltwater sources and the site is not located in an industrial area, as the surrounding area is primarily agricultural.

As such the outdoor air is not aggressive to stainless steel components.

d) A review of plant specific OE has not identified instances of degradation of stainless steel surfaces in outdoor air, including degradation due to LOM. The CPNPP LRA is revised to include this finding, consistent with SLR recommendations to document the results of the plant-specific OE review more fully in the application.

e) The Structures Monitoring Program manages LOM for structural stainless steel components exposed to outdoor air. Instances of LOM has not been identified for stainless steel components exposed to outdoor air.

f) SLR guidance recommends a one-time inspection to demonstrate that the LOM aging effect is not occurring or is occurring so slowly that it will not affect the intended function of the components during the subsequent period of extended operation. However, due to the fact that this aging effect has not been identified during the current period of operation, and the above basis provided, this part of the SLR operating experience was found to be not applicable for initial license renewal.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 4 of 27 For the above reasons, LOM is not identified as an aging effect for uninsulated stainless steel components exposed to outdoor air, except for the instances already identified in further evaluation LRA Sections 3.2.2.2.3, item 2; 3.3.2.2.5; and 3.4.2.2.3.

2. SCC of stainless steel in outdoor air:

With respect to the basis for not identifying SCC as an applicable aging effect for the select instances in which LOM is identified for uninsulated stainless steel components exposed to outdoor air, the select instances are the RWST tank liner and vent, RMWST tank liner and vent, and CST tank liner and vent. As discussed previously, these components are exposed to contaminant concentration from internal tank water or the vents to atmosphere. SLR guidance discusses halides and elevated temperatures as contributing factors leading to SCC of stainless steel in outdoor air. LOM is conservatively identified as an aging effect for these components due to the possibility for concentration of contaminants. However, SCC is not identified as an applicable aging effect due to the lack of ambient environment halides, lack of CPNPP OE identifying past instances of LOM or SCC for stainless steel components exposed to outdoor air, and the absence of elevated temperatures for the components in question. The CPNPP LRA is revised to include this finding, consistent with SLR recommendations to document the results of the plant-specific OE review more fully in the application.

3. LOM due to pitting and crevice corrosion for nickel alloy components exposed to outdoor air.

Aging management for LOM due to pitting and crevice corrosion of nickel alloy components exposed to outdoor air is not applicable for CPNPP. CPNPP does not have any nickel alloy components exposed to outdoor air within the scope of license renewal as reflected in the ESF, Auxiliary, Steam & Power Conversion and Structural AMR Results Tables.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 5 of 27 Associated LRA Revisions:

LRA Section 3.2.2.2.3, (page 3.2-8), is revised as follows:

Rose, the nearest community. Hood and Somervell counties are essentially rural, sparsely populated areas. The sparsely settled rural character extends well beyond the 10-mile radius. The area extending from 10 to 20 miles out from the CPNPP site is even more sparsely populated than the 0 to 10-mile area. As such, CPNPP is located inland in a rural area and away from potentially chloride heavy coastal environments or sulfate heavy industrial environments. As confirmed by OE, samples of groundwater and rainwater show chloride and sulfate concentrations less than 500 parts per million (ppm) and 1,500 PPM, respectively, with adequate pH. Furthermore, the CPNPP does not incorporate cooling towers.

Also, the closest highway is U.S. Highway 67 that passes through Glen Rose (as shown on google/maps) and the use of salt/ash to de-ice roadways is a rare occurrence in the moderate north-central Texas environs.

As such, the general ambient outdoor and indoor air is considered to be benign, non-aggressive lacking sufficient halides for corrosion of stainless steel.

A review of plant-specific OE does not reveal a history of loss of material due to pitting or crevice corrosion for stainless steel components exposed to outdoor air.

Therefore, stainless steel components exposed to air environments (including condensation) in the ESF Systems are not susceptible to loss of material and do not require management.

As summarized in item 3.2-1, 004, there are select instances where corrosion of stainless steel in air is considered possible and managed by the External Surfaces Monitoring of Mechanical Components (B.2.3.22) AMP. These instances within ESF systems are the stainless steel vents and liner inside the concrete RWST. There are no other stainless steel components exposed to outdoor air within ESF Systems. A plant-specific note is used for these instances.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 6 of 27 LRA Section 3.2.2.2.6, (page 3.2-9), is revised as follows:

3.2.2.2.6 Cracking due to Stress Corrosion Cracking Cracking due to stress corrosion cracking could occur for stainless steel piping, piping components, piping elements and tanks exposed to outdoor air. The possibility of cracking also extends to components exposed to air which has recently been introduced into buildings, i.e., components near intake vents.

Cracking is only known to occur in environments containing sufficient halides (primarily chlorides) and in which condensation or deliquescence is possible.

Condensation or deliquescence should generally be assumed to be possible.

Applicable outdoor air environments (and associated indoor air environments) include, but are not limited to, those within approximately 5 miles of a saltwater coastline, those within 1/2 mile of a highway which is treated with salt in the wintertime, those areas in which the soil contains more than trace chlorides, those plants having cooling towers where the water is treated with chlorine or chlorine compounds, and those areas subject to chloride contamination from other agricultural or industrial sources. This item is applicable for the environments described above.

GALL AMP XI.M36, External Surfaces Monitoring, is an acceptable method to manage the aging effect. The applicant may demonstrate that this item is not applicable by describing the outdoor air environment present at the plant and demonstrating that external chloride stress corrosion cracking is not expected.

The GALL Report recommends further evaluation to determine whether an aging management program is needed to manage this aging effect based on the environmental conditions applicable to the plant and requirements applicable to the components.

The outdoor environment at CPNPP is described in Section 3.2.2.2.3, item 2 above.

As described there, the outdoor air does not contain sufficient halides for degradation of stainless steel components in air. A review of CPNPP OE and evaluation of the location and surroundings of the plant have determined that the air at CPNPP does not contain sufficient halides nor are there events that would likely increase the halide content in the air to make SCC an AERM. A review of plant-specific OE does not reveal a history of stress corrosion cracking for stainless steel components exposed to outdoor air. As such, and as summarized in item 3.2-1, 007, stainless steel components exposed to air environments (including condensation) in the ESF Systems are not susceptible to cracking and do not require management for SCC. The RWSTs are outdoor reinforced concrete tanks with stainless steel liners as described in Section 2.4.11. The internal liners and vents of the tanks are not subject to elevated temperatures.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 7 of 27 LRA Section 3.3.2.2.3, (page 3.3-23), is revised as follows:

Sufficient halides nor are there events that would likely increase the halide content in the air to make cracking (due to stress corrosion cracking) an AERM for stainless steel exposed to air. A review of plant-specific OE does not reveal a history of stress corrosion cracking for stainless steel components exposed to outdoor air. As such, stainless steel components exposed to air environments in the auxiliary systems, as well as the in the structures and structural commodities are not susceptible to (halide induced) cracking, which is not expected and does not require management. The RMWSTs are outdoor reinforced concrete tanks with stainless steel liners as described in Section 2.4.11. The internal liners and vents of the tanks are not subject to elevated temperatures.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 8 of 27 LRA Section 3.3.2.2.5, (page 3.3-24), is revised as follows:

requirements Quality Assurance for Aging Management of Nonsafety-Related Components.

As described in Section 3.2.2.2.3, the outdoor air does not contain sufficient halides for corrosion of stainless steel in air. A review of CPNPP OE and evaluation of the location and surroundings of the plant have determined that the air at CPNPP does not contain sufficient halides nor are there events that would likely increase the halide content in the air to make loss of material (due to pitting and crevice corrosion) an AERM for stainless steel exposed to air except in select instances. A review of plant-specific OE does not reveal a history of loss of material due to pitting or crevice corrosion for stainless steel components exposed to outdoor air. As such, stainless steel components exposed to air environments in the auxiliary systems are not susceptible to (halide-induced) loss of material which does not require management. As summarized in item 3.3-1, 006; and 3.3-1, 012 there are select instances where corrosion of stainless steel in air is considered possible and managed by the External Surfaces Monitoring of Mechanical Components (B.2.3.22)

AMP or Bolting Integrity (B.2.3.9) AMP. These include the RMWST tank liner and internal vent (where contaminants could concentrate in the air-space that is vented to the outdoors, closure bolting (due to potential for local leakage) and insulation jacketing of components located outdoors (per LR-ISG-2012-02). Additionally, as summarized in item 3.3-1, 081 (per LR-ISG-2012-01), aluminum insulation jacketing is also susceptible to a loss of material and managed by the same AMP.

Furthermore, as summarized in item 3.5-1, 093, stainless steel structural components located outdoors in above-grade areas prone to frequent or prolonged wetting are susceptible to corrosion that is managed by the Structures Monitoring (B.2.3.34) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 9 of 27 LRA Table 3.3.2-4, (page 3.3-158), is revised as follows:

Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Orifice Pressure Stainless steel Gas (Internal) None None VII.J.AP-22 3.3-1, 120 A boundary Orifice Pressure Stainless steel Treated Loss of material Water Chemistry VIII.E.SP-87 3.4-1, 016 B boundary water (B.2.3.2) A (internal) One-Time Inspection (B.2.3.19)

Orifice Throttle Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A uncontrolled (external)

Orifice Throttle Stainless steel Gas (Internal) None None VII.J.AP-22 3.3-1, 120 A Orifice Throttle Stainless steel Treated Loss of material Water Chemistry VIII.E.SP-87 3.4-1, 016 B water (B.2.3.2) A (internal) One-Time Inspection (B.2.3.19)

Piping Leakage Carbon steel Air - indoor Loss of material External Surfaces VII.I.A-77 3.3-1, 078 A boundary uncontrolled Monitoring of (spatial) (external) Mechanical Components (B.2.3.22)

Piping Leakage Carbon steel Air - outdoor Loss of material External Surfaces VII.I.A-77 3.3-1, 078 A boundary (external) Monitoring of (spatial) Mechanical Components (B.2.3.22)

Piping Leakage Carbon steel Air with Loss of material Boric Acid VII.I.A-79 3.3-1, 009 A boundary borated water Corrosion (B.2.3.4)

(spatial) leakage (external)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 10 of 27 Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Piping Leakage Carbon steel Treated Loss of material Water Chemistry VIII.E.SP-73 3.4-1, 014 B boundary water (B.2.3.2) A (spatial) (internal) One-Time Inspection (B.2.3.19)

LRA Table 3.3.2-4, (page 3.3-159), is revised as follows:

Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Piping Leakage Carbon steel Treated Wall thinning Flow-Accelerated VII.E1.A-407 3.3-1, 126 B boundary water Corrosion (B.2.3.8)

(spatial) (internal)

Piping Leakage Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A boundary uncontrolled (spatial) (external)

Piping Leakage Stainless steel Air - outdoor None None VII.J.AP-123 3.3-1, 120 A, 5, 6 boundary (external)

(spatial)

Piping Leakage Stainless steel Treated Loss of material Water Chemistry VIII.E.SP-87 3.4-1, 016 B boundary water (B.2.3.2) A (spatial) (internal) One-Time Inspection (B.2.3.19)

Piping Leakage Stainless steel Treated Wall thinning Flow-Accelerated VII.E1.A-407 3.3-1, 126 B boundary water Corrosion (B.2.3.8)

(spatial) (internal)

Piping Pressure Carbon steel Air - indoor Loss of material External Surfaces VII.I.A-77 3.3-1, 078 A boundary uncontrolled Monitoring of (external) Mechanical Components (B.2.3.22)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 11 of 27 Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Piping Pressure Carbon steel Air with Loss of material Boric Acid VII.I.A-79 3.3-1, 009 A boundary borated water Corrosion (B.2.3.4) leakage (external)

Piping Pressure Carbon steel Treated Loss of material Water Chemistry VIII.E.SP-73 3.4-1, 014 B boundary water (B.2.3.2) A (internal) One-Time Inspection (B.2.3.19)

Piping Pressure Carbon steel Treated Wall thinning Flow-Accelerated VII.E1.A-407 3.3-1, 126 B boundary water Corrosion (B.2.3.8)

(internal)

LRA Table 3.3.2-4, (page 3.3-160), is revised as follows:

Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Piping Pressure Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A boundary uncontrolled (external)

Piping Pressure Stainless steel Air - outdoor None None VII.J.AP-123 3.3-1, 120 A, 5, 67 boundary (external) Loss of External Surfaces VII.E1.AP-221 3.3-1, 006 material Monitoring of Mechanical Components (B.2.3.22)

Piping Pressure Stainless steel Concrete None None VII.J.AP-19 3.3-1, 120 C, 4 boundary (external)

Piping Pressure Stainless steel Gas (Internal) None None VII.J.AP-22 3.3-1, 120 A boundary

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 12 of 27 Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Piping Pressure Stainless steel Treated Loss of material Water Chemistry VIII.E.SP-87 3.4-1, 016 B boundary water (B.2.3.2) A (internal) One-Time Inspection (B.2.3.19)

Piping Pressure Stainless steel Treated Wall thinning Flow-Accelerated VII.E1.A-407 3.3-1, 126 B boundary water Corrosion (B.2.3.8)

(internal)

Piping Structural Carbon steel Air - indoor Loss of material External Surfaces VII.I.A-77 3.3-1, 078 A, 5 integrity uncontrolled Monitoring of (attached) (external) Mechanical Components (B.2.3.22)

Piping Structural Carbon steel Air with Loss of material Boric Acid VII.I.A-79 3.3-1, 009 A integrity borated water Corrosion (B.2.3.4)

(attached) leakage (external)

Piping Structural Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A integrity uncontrolled (attached) (external)

LRA Table 3.3.2-4, (page 3.3-163), is revised as follows:

Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Tank liner Pressure Stainless steel Treated Loss of material Water Chemistry VIII.E.SP-75 3.4-1, 012 B (RMWST) boundary water (B.2.3.2) A, 1 (internal) One-Time Inspection (B.2.3.19)

Tubing Leakage Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A boundary uncontrolled (spatial) (external)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 13 of 27 Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Tubing Leakage Stainless steel Air - outdoor None None VII.J.AP-123 3.3-1, 120 A, 5, 6 boundary (external)

(spatial)

Tubing Leakage Stainless steel Treated Loss of material Water Chemistry VIII.E.SP-87 3.4-1, 016 B boundary water (B.2.3.2) A (spatial) (internal) One-Time Inspection (B.2.3.19)

Tubing Structural Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A integrity uncontrolled (attached) (external)

Valve body Leakage Carbon steel Air - indoor Loss of material External Surfaces VII.I.A-77 3.3-1, 078 A boundary uncontrolled Monitoring of (spatial) (external) Mechanical Components (B.2.3.22)

Valve body Leakage Carbon steel Air with Loss of material Boric Acid VII.I.A-79 3.3-1, 009 A boundary borated water Corrosion (B.2.3.4)

(spatial) leakage (external)

Valve body Leakage Carbon steel Treated Loss of material Water Chemistry VIII.E.SP-73 3.4-1, 014 B boundary water (B.2.3.2) A (spatial) (internal) One-Time Inspection (B.2.3.19)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 14 of 27 LRA Table 3.3.2-4, (page 3.3-164), is revised as follows:

Table 3.3.2-4: Demineralized and Reactor Makeup Water System - Summary of Aging Management Evaluation Aging Effect Aging Component Intended NUREG-1801 Table 1 Material Environment Requiring Management Notes Type Function Item Item Management Program Valve body Leakage Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A boundary uncontrolled (spatial) (external)

Valve body Leakage Stainless steel Air - outdoor None None VII.J.AP-123 3.3-1, 120 A, 5, 6 boundary (external)

(spatial)

Valve body Leakage Stainless steel Treated Loss of material Water Chemistry VIII.E.SP-87 3.4-1, 016 B boundary water (B.2.3.2) A (spatial) (internal) One-Time Inspection (B.2.3.19)

Valve body Pressure Carbon steel Air - indoor Loss of material External Surfaces VII.I.A-77 3.3-1, 078 A boundary uncontrolled Monitoring of (external) Mechanical Components (B.2.3.22)

Valve body Pressure Carbon steel Air with Loss of material Boric Acid VII.I.A-79 3.3-1, 009 A boundary borated water Corrosion (B.2.3.4) leakage (external)

Valve body Pressure Carbon steel Treated Loss of material Water Chemistry VIII.E.SP-73 3.4-1, 014 B boundary water (B.2.3.2) A (internal) One-Time Inspection (B.2.3.19)

Valve body Pressure Stainless steel Air - indoor None None VII.J.AP-123 3.3-1, 120 A boundary uncontrolled (external)

Valve body Pressure Stainless steel Air - outdoor None None VII.J.AP-123 3.3-1, 120 A, 5, 6 boundary (external)

Valve body Pressure Stainless steel Gas (Internal) None None VII.J.AP-22 3.3-1, 120 A boundary

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 15 of 27 LRA Table 3.3.2-4, (page 3.3-166), is revised as follows:

Generic Notes A. Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.

B. Consistent with component, material, environment, aging effect, and AMP listed for NUREG-1801 line item. AMP has exceptions to NUREG-1801 AMP description.

C. Component is different, but consistent with NUREG-1801 for material, environment, aging effect, and AMP. AMP is consistent with NUREG-1801 AMP.

G. Environment not in NUREG-1801 for this component and material.

Plant-Specific Notes

1. The internal stainless steel liner associated with the RMWST is evaluated in this AMR as a stainless steel tank component type.

2. The external concrete portion of the RMWST is evaluated with the Yard structures in Table 3.5.2-11.

3. The RMWST diaphragm separates the water and air spaces inside the RMWST. RMWST diaphragm membranes are rubber-type material with fiber-reinforcement (i.e. elastomer). Only the loss of material aging effect in the treated water environment requires management since the diaphragm is not exposed directly to UV or heat inside the tank and so is not susceptible to hardening or loss of strength.

4. This item does not require an AMP provided that certain concrete conditions are met. RMWST concrete and associate building wall AMR results are contained in Table 3.5.2-11.

5. Internal and external environment are the same for vent and drain components that are empty downstream of closed valves, such that the condition of external surface is representative of the internal surface condition.

6. Aging effects not applicable, for more information see Sections 3.3.2.2.3 and 3.3.2.2.5.Not Used.

7. The RMWST is vented to the outdoor air and contaminants/moisture may conservatively collect in the air space above the liner and connected diaphragm.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 16 of 27 LRA Section 3.4.2.2.2, (page 3.4-10), is revised as follows:

Applicable outdoor air environments (and associated indoor air environments) include, but are not limited to, those within approximately 5 miles of a saltwater coastline, those with1/21/2 mile of a highway which is treated with salt in the wintertime, those areas in which the soil contains more than trace chlorides, those plants having cooling towers where the water is treated with chlorine or chlorine compounds, and those areas subject to chloride contamination from other agricultural or industrial sources. This item is applicable for the environments described above. GALL AMP XI.M36, External Surfaces Monitoring, is an acceptable method to manage the aging effect. The applicant may demonstrate that this item is not applicable by describing the outdoor air environment present at the plant and demonstrating that external chloride stress corrosion cracking is not expected. The GALL Report recommends further evaluation to determine whether an adequate aging management program is used to manage this aging effect based on the environmental conditions applicable to the plant and ASME Code Section XI requirements applicable to the components.

The outdoor environment at CPNPP is described in Section 3.2.2.2.3, item 2 above.

As described there, the outdoor air does not contain sufficient halides for degradation of stainless steel components in air. A review of CPNPP OE and evaluation of the location and surroundings of the plant have determined that the air at CPNPP does not contain sufficient halides nor are there events that would likely increase the halide content in the air to make stress corrosion cracking an AERM. A review of plant-specific OE does not reveal a history of stress corrosion cracking for stainless steel components exposed to outdoor air. As such, as summarized in item 3.41, 002, stainless steel components exposed to air environments in the S&PC Systems are not susceptible to SCC and do not require management for SCC except as noted. The CSTs are outdoor reinforced concrete tanks with stainless steel liners as described in Section 2.4.11. The internal liners and vents of the tanks are not subject to elevated temperatures. As summarized in item 3.41, 063 (per LRISG201202), SCC of insulated Feedwater System piping located outdoors is considered susceptible to cracking (water through or under insulation) and is managed by the External Surfaces Monitoring of Mechanical Components (B.2.3.22)

AMP. The CSTs are outdoor reinforced concrete tanks with stainless steel liners as described in Section 2.4.11.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 17 of 27 LRA Section 3.4.2.2.3, (page 3.4-11), is revised as follows:

GALL AMP XI.M36, External Surfaces Monitoring, is an acceptable method to manage the aging effect. The applicant may demonstrate that this item is not applicable by describing the outdoor air environment present at the plant and demonstrating that external pitting or crevice corrosion is not expected. The GALL Report recommends further evaluation to determine whether an adequate aging management program is used to manage this aging effect based on the environmental conditions applicable to the plant and ASME Code Section XI requirements Quality Assurance for Aging Management of Nonsafety-Related Components.

The outdoor environment at CPNPP is described in Section 3.2.2.2.3, item 2 above.

As described there, the outdoor air does not contain sufficient halides for degradation of stainless steel components in air. A review of CPNPP OE and evaluation of the location and surroundings of the plant have determined that the air at CPNPP does not contain sufficient halides nor are there events that would likely increase the halide content in the air to make loss of material due to pitting and crevice corrosion an AERM. A review of plant-specific OE does not reveal a history of loss of material due to pitting or crevice corrosion for stainless steel components exposed to outdoor air. As such, stainless steel components exposed to air environments in the S&PC Systems are not susceptible to SCC or loss of material and do not require management for SCC or loss of material except as noted. As summarized in items 3.4-1, 003; 3.4-1, 008; and 3.4-1, 063, there are select instances where corrosion of stainless steel in air is considered possible and managed by the External Surfaces Monitoring of Mechanical Components (B.2.3.22)

AMP or Bolting Integrity (B.2.3.9) AMP. These instances includeare: liner and vent piping inside the CST (where contaminants could concentrate in the air space that is vented to the outdoors), closure bolting (due to the potential for local leakage), and through or under insulation of insulated Feedwater System piping (per LR-ISG-2012-02). Note that other than these instances, there are no other stainless steel components exposed to outdoor air within S&PC Systems.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 18 of 27 LRA Section 3.4.3, Table 3.4-1, (page 3.4-14), is revised as follows:

Table 3.4-1: Summary of Aging Management Programs for Steam and Power Conversion Systems Aging Aging Further Evaluation Item Component Effect/Mechanis Management Discussion Recommended m Programs 3.4-1, Steel Piping, Cumulative Fatigue is a Yes, TLAA (See Consistent with NUREG-1801.

001 piping fatigue damage time-limited aging subsection 3.4.2.2.1)

Cumulative fatigue damage is an aging effect addressed by a components, and due to fatigue analysis (TLAA) to fatigue TLAA (Section 4.3).

piping elements be evaluated for the exposed to Steam period of extended Further evaluation is documented in subsection 3.4.2.2.1.

or Treated water operation. See the SRP, Section 4.3 Metal Fatigue, for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

3.4-1, Stainless steel Cracking due to Chapter XI.M36, Yes, environmental Not applicable.

002 Piping, piping stress corrosion "External Surfaces conditions need to be Air does not contain sufficient halides to make SCC a concern.

components, cracking Monitoring of evaluated (See Further evaluation is documented in subsection 3.4.2.2.2.

and piping Mechanical subsection 3.4.2.2.2) elements; tanks Components" exposed to Air - outdoor 3.4-1, Stainless steel Loss of material Chapter XI.M36, Yes, environmental Consistent with NUREG-1801, as clarified.

003 Piping, piping due to pitting and "External Surfaces conditions need to be Air does not contain sufficient halides to make loss of material components, and crevice corrosion Monitoring of evaluated (See due to pitting and crevice corrosion a concern. A generic note piping elements; Mechanical subsection 3.4.2.2.3)

I and plant-specific note are used for vent piping and tanks exposed to Components" components inside the concrete CST as listed in Table 3.4.2-1.

Air - outdoor Loss of material for the top portion of the liner in the concrete CST and internal vent above the water-line and exposed to outdoor air through vents is manageare managed by the External Surfaces Monitoring of Mechanical Components (B.2.3.22) AMP.

Further evaluation is documented in subsection 3.4.2.2.3.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 19 of 27 LRA Section 3.4.3, Table 3.4-1, (page 3.4-15), is revised as follows:

Table 3.4-1: Summary of Aging Management Programs for Steam and Power Conversion Systems Aging Aging Further Evaluation Item Component Effect/Mechanis Management Discussion Recommended m Programs 3.4-1, Steel External Loss of material Chapter XI.M10, No Consistent with NUREG-1801.

004 surfaces, Bolting due to boric acid Boric Acid Loss of material of steel external surfaces and bolting exposed exposed to Air corrosion Corrosion to air with borated water leakage in the S&PC Systems is with borated water managed by the Boric Acid Corrosion (B.2.3.4) AMP.

leakage 3.4-1, Steel Piping, Wall thinning due Chapter XI.M17, No Consistent with exception to NUREG-1801.

005 piping to flow-accelerated Flow-Accelerated Wall thinning of steel piping and piping components exposed components, corrosion Corrosion to treated water, including steam, in the S&PC Systems is and piping managed by the Flow-Accelerated Corrosion (B.2.3.8) AMP, elements which takes exception to NUREG-1801.

exposed to Steam, Treated water 3.4-1, Steel, Stainless Loss of preload Chapter XI.M18, No Not applicable.

006 Steel Bolting "Bolting Integrity " There are no steel or stainless steel bolting exposed to soil in exposed to Soil the S&PC Systems.

3.4-1, High-strength Cracking due to Chapter XI.M18, No Not applicable.

007 steel Closure cyclic loading, "Bolting Integrity " There are no high strength steel closure bolting exposed to air bolting exposed to stress corrosion with steam or water leakage in the S&PC Systems.

Air with steam or cracking water leakage 3.4-1, Steel; stainless Loss of material Chapter XI.M18, No Consistent with NUREG-1801.

008 steel Bolting, due to general "Bolting Integrity "

Loss of material of steel and stainless steel bolting exposed to Closure bolting (steel only), pitting, air - outdoor (external) and air - indoor, uncontrolled exposed to and crevice (external) in the S&PC Systems is managed by the Bolting Air - outdoor corrosion Integrity (B.2.3.9) AMP.

(External),

Air - indoor, Loss of material of steel bolting exposed to air - indoor, uncontrolled uncontrolled (external) in SGs secondary side is also managed (External) by the Bolting Integrity (B.2.3.9) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 20 of 27 LRA Table 3.4.2-1, (page 3.4-47), is revised as follows:

Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Piping Pressure boundary Carbon steel Air - outdoor Loss of External VIII.H.S-41 3.4-1, 034 A (external) material Surfaces Monitoring of Mechanical Components (B.2.3.22)

Piping Pressure boundary Carbon steel Air with Loss of Boric Acid VIII.H.S-30 3.4-1, 004 A borated water material Corrosion leakage (B.2.3.4)

(external)

Piping Pressure boundary Carbon steel Lubricating oil Loss of Lubricating Oil VIII.G.SP-91 3.4-1, 040 A (internal) material Analysis (B.2.3.25)

One-Time Inspection (B.2.3.19)

Piping Pressure boundary Carbon steel Treated water Loss of Water VIII.G.SP-74 3.4-1, 013 B (internal) material Chemistry A (B.2.3.2)

One-Time Inspection (B.2.3.19)

Piping Pressure boundary Stainless steel Air - dry None Compressed VII.J.AP-20 3.3-1, 120 E, 4 (internal) Air Monitoring (B.2.3.14)

Piping Pressure boundary Stainless steel Air - indoor None None VIII.I.SP-12 3.4-1, 058 A uncontrolled (external)

Piping Pressure boundary Stainless steel Air - indoor None None VIII.I.SP-86 3.4-1, 058 A uncontrolled (internal)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 21 of 27 Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Piping Pressure boundary Stainless steel Air - outdoor None None VIII.G.SP-127 3.4-1, 003 I, 2 (external) Loss of External A material Surfaces Monitoring of Mechanical Components (B.2.3.22)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 22 of 27 LRA Table 3.4.2-1, (page 3.4-48), is revised as follows:

Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Piping Pressure boundary Stainless steel Treated water Loss of Water VIII.G.SP-87 3.4-1, 016 B (internal) material Chemistry A (B.2.3.2)

One-Time Inspection (B.2.3.19)

Piping Pressure boundary Stainless steel Treated water Wall thinning Flow- VIII.G.S-408 3.4-1, 060 B (internal) Accelerated Corrosion (B.2.3.8)

Piping Structural integrity Carbon steel Air - dry None Compressed VII.J.AP-4 3.3-1, 121 E, 4 (attached) (internal) Air Monitoring (B.2.3.14)

Piping Structural integrity Carbon steel Air - indoor Loss of External VIII.H.S-29 3.4-1, 034 A (attached) uncontrolled material Surfaces (external) Monitoring of Mechanical Components (B.2.3.22)

Piping Structural integrity Carbon steel Air - indoor Loss of Inspection of V.A.E-29 3.2-1, 044 C (attached) uncontrolled material Internal (internal) Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Piping Structural integrity Carbon steel Air - outdoor Loss of External VIII.H.S-41 3.4-1, 034 A (attached) (external) material Surfaces Monitoring of Mechanical Components (B.2.3.22)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 23 of 27 LRA Table 3.4.2-1, (page 3.4-49), is revised as follows:

Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Piping Structural integrity Carbon steel Air with Loss of Boric Acid VIII.H.S-30 3.4-1, 004 A (attached) borated water material Corrosion leakage (B.2.3.4)

(external)

Piping Structural integrity Stainless steel Air - dry None Compressed VII.J.AP-20 3.3-1, 120 E, 4 (attached) (internal) Air Monitoring (B.2.3.14)

Piping Structural integrity Stainless steel Air - indoor None None VIII.I.SP-12 3.4-1, 058 A (attached) uncontrolled (external)

Piping Structural integrity Stainless steel Air - indoor None None VIII.I.SP-86 3.4-1, 058 A (attached) uncontrolled (internal)

Piping Structural integrity Stainless steel Air - outdoor None None VIII.G.SP-127 3.4-1, 003 I, 2 (attached) (external)

Pump casing Leakage boundary Stainless steel Air - indoor None None VIII.I.SP-12 3.4-1, 058 C (condensate transfer) (spatial) uncontrolled (external)

Pump casing Leakage boundary Stainless steel Treated water Loss of Water VIII.G.SP-87 3.4-1, 016 D (condensate transfer) (spatial) (internal) material Chemistry C (B.2.3.2)

One-Time Inspection (B.2.3.19)

Pump casing (MDAFW) Pressure boundary Carbon steel Air - indoor Loss of External VIII.H.S-29 3.4-1, 034 A uncontrolled material Surfaces (external) Monitoring of Mechanical Components (B.2.3.22)

Pump casing (MDAFW) Pressure boundary Carbon steel Air with Loss of Boric Acid VIII.H.S-30 3.4-1, 004 A borated water material Corrosion leakage (B.2.3.4)

(external)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 24 of 27 LRA Table 3.4.2-1, (page 3.4-54), is revised as follows:

Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Valve body Leakage boundary Stainless steel Air - indoor None None VIII.I.SP-12 3.4-1, 058 A (spatial) uncontrolled (external)

Valve body Leakage boundary Stainless steel Treated water Loss of Water VIII.G.SP-87 3.4-1, 016 B (spatial) (internal) material Chemistry A (B.2.3.2)

One-Time Inspection (B.2.3.19)

Valve body Pressure boundary Carbon steel Air - dry None Compressed VII.J.AP-4 3.3-1, 121 E, 4 (internal) Air Monitoring (B.2.3.14)

Valve body Pressure boundary Carbon steel Air - indoor Loss of External VIII.H.S-29 3.4-1, 034 A uncontrolled material Surfaces (external) Monitoring of Mechanical Components (B.2.3.22)

Valve body Pressure boundary Carbon steel Air - outdoor Loss of External VIII.H.S-41 3.4-1, 034 A (external) material Surfaces Monitoring of Mechanical Components (B.2.3.22)

Valve body Pressure boundary Carbon steel Air with Loss of Boric Acid VIII.H.S-30 3.4-1, 004 A borated water material Corrosion leakage (B.2.3.4)

(external)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 25 of 27 Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Valve body Pressure boundary Carbon steel Treated water Loss of Water VIII.G.SP-74 3.4-1, 013 B (internal) material Chemistry A (B.2.3.2)

One-Time Inspection (B.2.3.19)

LRA Table 3.4.2-1, (page 3.4-55), is revised as follows:

Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Valve body Pressure boundary Stainless steel Air - dry None Compressed VII.J.AP-20 3.3-1, 120 E, 4 (internal) Air Monitoring (B.2.3.14)

Valve body Pressure boundary Stainless steel Air - indoor None None VIII.I.SP-12 3.4-1, 058 A uncontrolled (external)

Valve body Pressure boundary Stainless steel Air - indoor None None VIII.I.SP-86 3.4-1, 058 A uncontrolled (internal)

Valve body Pressure boundary Stainless steel Air - outdoor None None VIII.G.SP-127 3.4-1, 003 I, 2 (external)

Valve body Pressure boundary Stainless steel Treated water Loss of Water VIII.G.SP-87 3.4-1, 016 B (internal) material Chemistry A (B.2.3.2)

One-Time Inspection (B.2.3.19)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 26 of 27 Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Valve body Structural integrity Carbon steel Air - indoor Loss of External VIII.H.S-29 3.4-1, 034 A (attached) uncontrolled material Surfaces (external) Monitoring of Mechanical Components (B.2.3.22)

Valve body Structural integrity Carbon steel Air - indoor Loss of Inspection of V.A.E-29 3.2-1, 044 C (attached) uncontrolled material Internal (internal) Surfaces in Miscellaneous Piping and Ducting Components (B.2.3.24)

Valve body Structural integrity Carbon steel Air with Loss of Boric Acid VIII.H.S-30 3.4-1, 004 A (attached) borated water material Corrosion leakage (B.2.3.4)

(external)

LRA Table 3.4.2-1, (page 3.4-56), is revised as follows:

Table 3.4.2-1: Auxiliary Feedwater System - Summary of Aging Management Evaluation Aging Effect Aging NUREG-1801 Component Type Intended Function Material Environment Requiring Management Table Item Notes Item Management Programs Valve body Structural integrity Stainless steel Air - indoor None None VIII.I.SP-12 3.4-1, 058 A (attached) uncontrolled (external)

Valve body Structural integrity Stainless steel Air - indoor None None VIII.I.SP-86 3.4-1, 058 A (attached) uncontrolled (internal)

Generic Notes A. Consistent with component, material, environment, aging effect, and AMP listed for NUREG-1801 line item. AMP is consistent with NUREG-1801 AMP description.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1 TXX-23048 and CP-202300322 Attachment E Page 27 of 27 B. Consistent with component, material, environment, aging effect, and AMP listed for NUREG-1801 line item. AMP has exceptions to NUREG-1801 AMP description.

C. Component is different, but consistent with NUREG-1801 for material, environment, aging effect, and AMP. AMP is consistent with NUREG-1801 AMP.

D. Component is different, but consistent with NUREG-1801 item for material, environment, and aging effect. AMP takes some exceptions to NUREG-1801 AMP.

E. Consistent with NUREG-1801 item for material, environment, and aging effect, but a different AMP is credited or NUREG-1801 identifies a plant-specific AMP.

I. Aging effect in NUREG-1801 for this component, material and environment combination is not applicable.

Plant-Specific Notes

1. The external concrete portion of the CST is evaluated with the Yard structures in Table 3.5.2-10.

2. Aging effects not applicable, for more information see Sections 3.4.2.2.2 and 3.4.2.2.3.Not Used.

3. Structural support functions of moment restraints are managed by the ASME Section XI, Subsection IWF (B.2.3.31) AMP and evaluated in Table 3.5.2-13.

4. Compressed (dry) air is supplied to the air accumulators and attached components from the Instrument Air System for air-operated AFW control valves.

5. Conservatively, the liner and makeup/vent lines inside the CST are exposed to outdoor air above the waterline where contaminants could concentrate.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-1 TXX-23048 and CP-202300322 Attachment F Page 1 of 6 LRA Section: B.2.3.15, Fire Protection NRC RAI No: RAI B.2.3.15-1 Fire Protection AMP Regulatory Basis:

Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S.

Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described in the requests for additional information.

Background:

Section 4.1.2 in Revision 1 of LUM00020-REPT-070, Comanche Peak Nuclear Power Plant Units 1 and 2 License Renewal Masonry Walls Aging Management Program [AMP] Basis Document, states, Masonry walls that perform a fire barrier intended function are also managed by the Fire Protection Program [Ref. 9.3], consistent with the operating experience reflected in subsequent license renewal guidance NUREG-2191 [Ref. 9.23] Item VII.G.A-626.

Aging Management Review (AMR) item VII.G.A-626, 3.3-1, 179 in Volume 1 of NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR)

Report (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17187A031), addresses cracking and loss of material of masonry walls that are structural fire barrier walls exposed to air managed by both the Fire Protection AMP and the Masonry Walls AMP.

Section B.2.3.33, Masonry Walls, of Appendix B of the License Renewal Application (LRA) states, Masonry walls that perform a fire barrier intended function are also managed by the Fire Protection (B.2.3.15) AMP. This statement is consistent with both NUREG-2191 and Revision 2 of NUREG-1801, Generic Aging Lessons Learned (GALL) Report (ML103490041).

Issue:

License Renewal Application (LRA) Tables 3.3-1, Summary of Aging Management Programs for Auxiliary Systems, and 3.5.2-15, Fire Barrier Commodity Group - Summary of Aging Management Evaluation, do not cite AMR item VII.G.A-626, 3.3-1, 179 for the

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-1 TXX-23048 and CP-202300322 Attachment F Page 2 of 6 masonry block concrete block (removable) for opening, and wall, floor, and ceiling, commodities.

LRA Table 3.5.2-15 cites AMR item III.A3.T-12, 3.5-1, 070 with plant-specific note 3, which states, The Masonry Walls (B.2.3.33) AMP and Fire Protection (B.2.3.15) AMP credit and communicate with each other, is cited for the masonry block commodities noted above.

The staff notes that stating two AMPs credit and communicate with each other does not clearly reflect that the commodity is managed by both the AMPs.

During the audit of the Fire Protection AMP, the applicant stated that visual inspections of masonry walls that perform a fire barrier intended function are performed by the Masonry Walls AMP and the Fire Protection AMP. Therefore, the LRA does not appear to accurately reflect the management of masonry walls that perform a fire barrier intended function because the Fire Protection AMP is not cited. In addition, since the Fire Protection AMP is not cited, the LRA does not appear to be consistent with the operating experience reflected in NUREG-2191 (AMR item VII.G.A-626, 3.3-1, 179).

Request:

Given that it appears that the Fire Protection AMP, along with the Masonry Walls AMP, manages masonry block concrete block (removable) for opening, and wall, floor, and ceiling, commodities, please provide the basis for not crediting the Fire Protection AMP in the LRA. Alternatively, revise the LRA to also credit the Fire Protection AMP.

Luminant Response:

NUREG-1801, Revision 2, includes cracking due to restraint shrinkage, creep, and aggressive environment as an aging effect requiring management for concrete block/masonry walls in both an air indoor or outdoor environment and specifies the Masonry Walls AMP for management of this aging effect. The NUREG-1800, XI.S5 Masonry Walls AMP description includes inspections for managing cracking of masonry walls. In order to address the industry operating experience discussed in NUREG-2192 (which is considered as relevant operating experience during the period of extended operation), masonry walls that perform a fire barrier intended function are also managed by the Fire Protection AMP, XI.M26.

As described in LRA Table 3.5-1, Item 3.5-1, 070 and Table 3.5.2-15, plant-specific notes 1 and 3, the Masonry Walls (B.2.3.33) AMP and Fire Protection (B.2.3.15) AMP credit and communicate with each other. In addition, LRA Sections A.2.2.15 and B.2.3.15, Fire Protection AMP, reflect inspection of fire barrier walls, ceilings, floors, and other fire resistant material component types and LRA Sections A.2.2.33 and B.2.3.33, Masonry Walls AMP, refer to the Fire Protection AMP for masonry walls that perform a fire barrier function.

LRA Table 3.5-1 (Item 3.5-1, 070) and Table 3.5.2-15 are revised, in order to further clarify that masonry walls (and removable concrete blocks) that perform a fire barrier function are managed by the Fire Protection AMP as well as the Masonry Walls AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-1 TXX-23048 and CP-202300322 Attachment F Page 3 of 6

References:

None.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-1 TXX-23048 and CP-202300322 Attachment F Page 4 of 6 Associated LRA Revisions:

LRA Table 3.5-1, page 3.5-62, is revised as follows:

Table 3.5-1 Summary of Aging Management Programs for Containments, Structures and Commodities Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5-1, 070 Masonry walls: all Cracking due to Chapter XI.S5, No Consistent with NUREG-1801.

restraint shrinkage, Masonry Walls The Masonry Walls (B.2.3.33) AMP will be creep, and used to manage cracking of masonry walls aggressive exposed to indoor air and outdoor air.

environment As described in Table 3.5.2-15, some masonry walls and block openings include fire barriers.

Both the Masonry Walls (B.2.3.33) AMP and the Fire Protection (B.2.3.15) AMP will be used to manage cracking of masonry walls and block openings in fire barriers exposed to indoor air and outdoor air, and the Masonry Walls (B.2.3.33) AMP credits and communicates with the Fire Protection (B.2.3.15) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-1 TXX-23048 and CP-202300322 Attachment F Page 5 of 6 LRA Table 3.5.2-15, page 3.5-185, is revised as follows:

Table 3.5.2-15: Fire Barrier Commodity Group - Summary of Aging Management Evaluation Commodity Intended Material Environment Aging Effect Aging NUREG-1801 Table 1 Notes Function Requiring Management Item Item Management Program Concrete block Fire barrier Masonry block Air - indoor Cracking Masonry Walls III.A3.T-12 3.5-1, 070 A, 1 (removable) for uncontrolled (B.2.3.33) opening Concrete block Fire barrier Masonry block Air - indoor Cracking; loss of Structures III.A3.TP-26 3.5-1, 066 C, 6 (removable) for uncontrolled bond; and loss of Monitoring opening material (B.2.3.34)

Concrete block Fire barrier Masonry block Air - indoor Cracking Fire Protection III.A3.T-12 3.5-1, 070 E, 1 (removable) for uncontrolled (B.2.3.15) opening Damper housing Fire barrier Galvanized steel Air - indoor None None III.B4.TP-8 3.5-1, 095 C uncontrolled Damper housing Fire barrier Galvanized steel Air with borated Loss of Material Boric Acid III.B4.TP-3 3.5-1, 089 C water leakage Corrosion (B.2.3.4)

Door Fire barrier Carbon steel Air - indoor Loss of material Fire Protection VII.G.A-21 3.3-1, 059 A uncontrolled (B.2.3.15)

Door Fire barrier Galvanized steel Air - indoor Loss of material Fire Protection VII.G.A-21 3.3-1, 059 A uncontrolled (B.2.3.15)

Door Fire barrier Carbon steel Air - outdoor Loss of material Fire Protection VII.G.A-22 3.3-1, 059 A (B.2.3.15)

Door Fire barrier Galvanized steel Air - outdoor Loss of material Fire Protection VII.G.A-22 3.3-1, 059 A (B.2.3.15)

Door Fire barrier Carbon steel Air with borated Loss of material Boric Acid III.B2.T-25 3.5-1, 089 C water leakage Corrosion (B.2.3.4)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-1 TXX-23048 and CP-202300322 Attachment F Page 6 of 6 LRA Table 3.5.2-15, page 3.5-189, is revised as follows:

Table 3.5.2-15: Fire Barrier Commodity Group - Summary of Aging Management Evaluation Commodity Intended Material Environment Aging Effect Aging NUREG-1801 Table 1 Notes Function Requiring Management Item Item Management Program Wall, floor, and Fire barrier Masonry block Air - indoor Cracking Fire Protection III.A3.T-12 3.5-1, 070 E, 3 ceiling uncontrolled (B.2.3.15)

Wall, floor, and Fire barrier Concrete Air - indoor Cracking Structures VII.G.A-90 3.3-1, 060 A ceiling (reinforced) uncontrolled Monitoring (B.2.3.34)

Wall, floor, and Fire barrier Gypsum Air - indoor Cracking; loss of Fire Protection None None F, 4 ceiling uncontrolled bond; and loss of (B.2.3.15) material Wall, floor, and Fire barrier Carbon steel Air - indoor Loss of material Fire Protection VII.G.A-21 3.3-1, 059 C ceiling uncontrolled (B.2.3.15)

Wall, floor, and Fire barrier Concrete Air - indoor Loss of material Fire Protection VII.G.A-91 3.3-1, 062 A ceiling (reinforced) uncontrolled (B.2.3.15)

Wall, floor, and Fire barrier Concrete Air - indoor Loss of material Structures VII.G.A-91 3.3-1, 062 A ceiling (reinforced) uncontrolled Monitoring (B.2.3.34)

Wall, floor, and Fire barrier Masonry block Air - outdoor Cracking Fire Protection III.A3.T-12 3.5-1, 070 E, 3 ceiling (B.2.3.15)

Wall, floor, and Fire barrier Masonry block Air - outdoor Cracking Masonry Walls III.A3.T-12 3.5-1, 070 C, 3 ceiling (B.2.3.33)

Wall, floor, and Fire barrier Concrete Air - outdoor Cracking; Loss of Fire Protection VII.G.A-92 3.3-1, 061 A ceiling (reinforced) material (B.2.3.15)

Wall, floor, and Fire barrier Concrete Air - outdoor Cracking; Loss of Structures VII.G.A-92 3.3-1, 061 A ceiling (reinforced) material Monitoring (B.2.3.34)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 1 of 30 LRA Section: B.2.3.15, Fire Protection NRC RAI No: RAI B.2.3.15-2 Fire Protection AMP (Concrete curbs, concrete berm/dike, fire barrier function.)

Regulatory Basis:

Section 54.21(a)(3) of Title 10 of the Code of Federal Regulations (10 CFR) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the U.S.

Nuclear Regulatory Commission (NRC) staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described in the requests for additional information.

Background:

Revision 2 of NUREG-1800 includes AMR items for managing reinforced concrete structural fire barriers by both the Fire Protection AMP and the Structures Monitoring AMP (VII.G.A-90, 3.3-1, 060; VII.G.A-92, 3.3-1, 061; and VII.G.A-91, 93, 3.3-1, 062).

LRA Supplement 1 (ML23096A302) revised the following tables to add concrete curbs with intended functions of direct flow and fire barrier.

Table 2.4-3, Diesel Generator Building Components Subject to Aging Management Review Table 3.5.2-3, Diesel Generator Buildings - Summary of Aging Management Evaluation Table 2.4-4, Electrical and Control Building Components Subject to Aging Management Review Table 3.5.2-4, Electrical and Control Building - Summary of Aging Management Evaluation Tables 3.5.2-3 and 3.5.2-4 cite AMR Items III.A3.TP-25, 3.5-1, 054; III.A3.TP-26, 3.5-1, 066; and III.A3.TP-28, 3.5-1, 067 for the concrete curbs. These AMR items cite only the Structures Monitoring AMP. The associated plant-specific notes 7 (Table 3.5.2-3) and 8 (Table 3.5.2-4) state, The Structures Monitoring (B.2.3.34) AMP and Fire Protection

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 2 of 30 (B.2.3.15) AMP credit and communicate with each other.

LRA Supplement 1 revised Tables 2.4-11, Yard Structure Components Subject to Aging Management Review, and 3.5.2-11, Yard Structures - Summary of Aging Management Evaluation, to add reinforced concrete berm/dike with an intended function of fire barrier.

AMR items VII.G.A-92, 3.5-1, 061 and VII.G.A-93, 3.5-1, 062 were cited. Generic note E and plant- specific note 9 were cited, which states, The Fire Protection (B.2.3.15) AMP alone manages the aging of the berm/dike around the auxiliary boiler fuel oil storage tank. The berm/dike is located outside the protected area and has a conservative fire barrier intended function. However, LRA Supplement 1 revised the Discussions for AMR items VII.G.A-92, 3.3-1, 061 and VII.G.A-93, 3.3-1, 062 in LRA Table 3.3-1 to indicate that the Fire Protection AMP is used to manage the reinforced concrete berm/dike. Therefore, it appears that LRA Table 3.5.2-11 may have cited the incorrect AMR items for the reinforced concrete berm/dike.

Issue:

Not crediting both the Structures Monitoring AMP and the Fire Protection AMP for the concrete curbs and the reinforced concrete berm/dike is not consistent with the recommendation in Revision 2 of NUREG-1800 (VII.G.A-90, 3.3-1, 060; VII.G.A-92, 3.3-1, 061; and VII.G.A-91, 93, 3.3-1, 062), and it does not appear to be consistent with the applicants treatment of concrete and reinforced concrete commodities in LRA Table 3.5.2-15 that credit both the Structures Monitoring AMP and the Fire Protection AMP.

Request:

Please provide the basis for not also crediting the Fire Protection AMP in the LRA for managing the reinforced concrete curbs and for not also crediting the Structures Monitoring AMP in the LRA for managing the reinforced concrete berm/dike. Alternatively, revise the LRA, consistent with the recommendation in Revision 2 of NUREG-1800, to credit both the Fire Protection AMP and the Structures Monitoring AMP for the reinforced concrete curbs and the reinforced concrete berm/dike.

Luminant Response:

The reinforced concrete curbs and the reinforced concrete berm/dike both have a fire barrier intended function of containing flammable liquids/oil spills and preventing the spread of fire.

These functions are identified in the LRA as discussed below. In order to provide better clarity, and conciseness these fire barrier intended functions for the concrete curb and concrete berm/dike components are being moved from the Diesel Generator Building, the Electrical and Control Building and the Yard Structures AMR tables to the Fire Barrier Commodity Group AMR table. These changes are as described below.

As defined in LRA Section 2.1.6.2, Table 2.1-1 the structure and component intended function Direct Flow includes curbs for directing flow or providing a means for fluid flow diversion.

These curbs are subject to AMR with the component intended function managed by the Structures Monitoring AMP (B.2.3.34). Concrete curbs are part of the reinforced concrete floor in various locations at floor openings or other items. In areas where these curbs are part of a

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 3 of 30 floor or wall which serves as a fire barrier, the curbs are fire rated equal to that of the fire barrier and have a fire barrier intended function (e.g., oil spills, containing flammable liquids, etc.) managed by the Fire Protection AMP (B.2.3.15) and as clarified by LRA Supplement 1 (Reference 1).

As described in LRA Table 3.5-1 (Items 3.5-1, 054, 3.5-1, 066 and 3.5-1, 067), LRA Table 3.5.2-3, plant-specific note 7 and LRA Table 3.5.2-4 plant-specific note 8, the Structures Monitoring (B.2.3.34) AMP and the Fire Protection (B.2.3.15) AMP credit and communicate with each other.

LRA Table 2.4-3, Table 2.4-4, Table 3.5-1 (Items 3.5-1, 054; 3.5-1, 066; and 3.5-1, 067), and Table 3.5.2-15 are revised, in order to further clarify that concrete curbs that perform a fire barrier function are managed by the Fire Protection AMP as well as the Structures Monitoring AMP.

The Auxiliary Boiler Fuel Oil Storage Tank reinforced concrete berm/dike is described in FSAR Section 9.5.1.6.1 (APCSB 9.5-1 App A F.18) which states The CPNPP design includes a fuel oil fired auxiliary boiler including associated above ground fuel oil tank which is diked to contain the entire content of the tank. Based on the FSAR, the intended function of the berm/dike is a fire barrier function for fuel oil containment. Consistent with NUREG-1800, Section 2.4.3.2 which states Structural components having intended functions are within the scope of license renewal (i.e., if the FSAR indicates that a dike within the fire pump house prevents a fuel oil fire from spreading to the electrically driven fire pump), this dike is identified as being within the scope of license renewal., as a result, the berm/dike is managed by the Fire Protection AMP (B.2.3.15).

Furthermore, the following conditions apply to the CPNPP Auxiliary Boiler building and Auxiliary Boiler Fuel Oil Storage Tank:

Auxiliary Boiler building and the Auxiliary Boiler Fuel Oil Storage Tank reinforced concrete berm/dike are not maintenance rule structures and not included in the Structures Monitoring program; They are located well away from other plant scope structures, as shown on LR-STRUCT-01, and as such there are no adverse effects on other structures within the scope of license renewal; Inclusion of the berm/dike within the scope of license renewal based on the FSAR statement is conservative.

LRA Sections 2.4.11 and 2.4.15; Table 2.4-11, Table 3.5-1 (Items 3.5-1, 061; and 3.5-1, 062) and Table 3.5.2-15 are revised, in order to further clarify that the auxiliary boiler berm/dike is managed by the Fire Protection AMP. LRA Appendix A, Table A-3, Item 17 and LRA Appendix B, Section B.2.3.15, Element 4 are revised to add an enhancement to revise the Fire Rated Assembly Visual Inspection procedure to include a requirement to inspect the Auxiliary Boiler Fuel Oil Storage tank Concrete berm/dike. Additionally, a clarification was added in Element 4 for the enhancement added in Supplement 2.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 4 of 30

References:

1. Letter TXX-23012, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 1, April 6, 2023 (ADAMS Accession No. ML23096A302)

2. Letter TXX-23022, from Jay Lloyd to the NRC, submitting Comanche Peak Nuclear Power Plant License Renewal Application Revision 0, Supplement 2, April 24, 2023 (ADAMS Accession No. ML23114A377)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 5 of 30 Associated LRA Revisions:

LRA Table 2.4-3, page 2.4-12, is revised as follows:

Table 2.4-3 Diesel Generator Building Components Subject to Aging Management Review Component type Intended function Anchorage to building structure Structural support Bolting (structural) Structural support Concrete curb Direct flow Fire barrier Concrete: Above-grade exterior (accessible) Missile barrier Shelter, protection Structural support Concrete: Above-grade exterior (inaccessible) Missile barrier Shelter, protection Structural support Concrete: Foundation, sub-foundation (inaccessible) Structural support Concrete: Interior Missile barrier Shelter, protection Structural support Door (watertight) Flood barrier Missile barrier Shelter, protection Door seal Flood barrier Hatch/plug Missile barrier Shelter, protection Structural support Louver housing Pressure relief Shelter, protection Masonry wall Shelter, protection Steel component: All structural members Shelter, protection Structural support Tornado/missile shield Missile barrier Shelter, protection

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 6 of 30 LRA Table 2.4-4, page 2.4-15, is revised as follows:

Table 2.4-4 Electrical and Control Building Components Subject to Aging Management Review Component type Intended function Anchorage to building structure Structural support Bolting (structural) Structural support Compressible joint and seal Shelter, protection Concrete Curb Direct flow Fire barrier Concrete: Above-grade exterior (accessible) Missile barrier Pressure barrier Shelter, protection Structural support Concrete: Above-grade exterior (inaccessible) Missile barrier Pressure barrier Shelter, protection Structural support Concrete: Below-grade exterior (inaccessible) Missile barrier Pressure barrier Shelter, protection Structural support Concrete: Foundation, sub-foundation (inaccessible) Structural support Concrete: Interior Missile barrier Pressure barrier Shelter, protection Shielding Structural support Door (air tight, missile resisting, tornado) Missile barrier Pressure barrier Pressure relief Shelter, protection Door seal Pressure barrier Hatch / Removable slab Missile barrier Shelter, protection Structural support Masonry wall: Interior Shelter, protection Structural support Moisture barrier Flood barrier Steel component: All structural members Shelter, protection Structural support Tornado blowout panel Pressure relief Shelter, protection Tornado pressure relief damper housing Pressure relief Shelter, protection

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 7 of 30 LRA Section 2.4.11, page 2.4-29, is revised as follows:

Fire Barrier Walls/Enclosure Pits The main transformers, unit auxiliary, and startup/alternate startup transformers are oil-cooled and are located outdoors adjacent to the TB. The main transformers are separated from each other, as well as from the TB, by a three-hour rated fire wall.

The unit auxiliary transformers, startup transformers and one alternate (spare) startup transformer are separated from the TB by a three-hour rated fire wall.

Penetrations in these walls within 50-0 from each side of the center line of the transformer are protected, by seals or closures having a fire resistance rating equal to the rating assigned to the barrier, to maintain the fire-resistant integrity of the wall.

Additional walls are provided extending out from the TB wall to protect the ventilation openings located in the exterior TB wall. In addition, the main, unit auxiliary, startup and one spare startup transformers are surrounded by reinforced concrete enclosures/pits and the auxiliary boiler fuel storage tank is surrounded by a concrete berm/dike to contain oil spills and prevent the spread of fire. The fire barrier function for the fire barrier walls/enclosure pits and the concrete berm/dike is addressed in Section 2.4.15.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 8 of 30 LRA Table 2.4-11, page 2.4-34, is revised as follows:

Table 2.4-11 Yard Structure Components Subject to Aging Management Review Component type Intended function Anchorage to building structure Structural support Berm/Dike Fire barrier Bolting (structural) Structural support Concrete: Above-grade exterior (accessible) Missile barrier Shelter, protection Structural support Concrete: Below-grade exterior (inaccessible) Shelter, protection Structural support Concrete: Foundation, sub-foundation (inaccessible) Structural support Concrete: Interior Flood barrier Missile barrier Shelter, protection Structural support Elastic joint filler Shelter, protection Manhole, handhole & duct bank Shelter, protection Structural support Manway seal Shelter, protection Masonry wall: Interior/Exterior Shelter, protection Structural support Metal deck Shelter, protection Metal siding Shelter, protection Steel component: structural member, grating, manhole Missile barrier and manway cover, etc. Shelter, protection Structural support

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 9 of 30 LRA Section 2.4.15, page 2.4-45, is revised as follows:

Fire Barrier Commodity Group Description Fire barriers are those components of construction (walls, floors, or protective covering) that are rated by approved laboratories or are constructed in accordance with the requirements stated by authorities having jurisdiction. Barriers are rated in hours of resistance to fire and are used to prevent the spread of fire. The integrity of a fire barrier may also be substantiated by either a fire test or an FHA (fire hazards analysis) evaluation. Similarly, a fire break is a physical barrier that prevents fire propagation, that is, the spreading of a fire from one component to another or direct exposure of a component to the heat and flames of a fire, or both.

Fire barriers are part of the plant FPS and perform an intended function and are included within the scope for LR per the Commissions regulations for FP (10 CFR 50.48). For LR considerations, fire breaks are considered as fire barriers.

Fire barriers and associated components within the scope of LR include:

Concrete berm/dike The auxiliary boiler fuel oil storage tank is surrounded by a concrete berm/dike. The function of the berm/dike is to contain fuel oil spills and prevent the spread of fire. It does not provide structural support or other protection. Additionally, the concrete berm/dike is not a maintenance rule structure and not included in the structures monitoring program as it is located well away from other plant scope structures, as shown on LR-STRUCT-01, and as such there are no adverse effects on other structures within the scope of license renewal.

Concrete block (removable) for opening Removable concrete block openings are provided in certain walls to facilitate equipment removal and replacement. Thus, in areas where a removable concrete block opening exists in a fire wall, an FHA evaluation justifies the as-built design, and the concrete blocks are installed in such a way that there are no through openings from one side of the barrier to the other. The concrete blocks are also restrained in the openings so that they will remain in place during a fire or seismic condition.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 10 of 30 Concrete curb Concrete curbs are part of the reinforced concrete floor in various locations at floor openings or other items. In areas where these curbs are part of a floor which serves as a fire barrier, the curbs are fire rated equal to or greater than that of the fire barrier. Additionally, containing flammable liquids is part of the fire barrier function of concrete curbs.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 11 of 30 LRA Table 2.4-15, page 2.4-48, is revised as follows:

Table 2.4-15 Fire Barrier Commodity Group Subject to Aging Management Review Commodity Intended Function Berm/Dike Fire barrier Concrete block (removable) for opening Fire barrier Concrete Curb Fire barrier Damper housing Fire barrier Door Fire barrier Hatch Fire barrier Insulation and wrap Fire barrier Penetration seal Fire barrier Penetration sleeve Fire barrier Wall, floor, and ceiling Fire barrier FP hose station (rack, reel, and support) Structural support

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 12 of 30 LRA Table 3.3-1, page 3.3-50, is revised as follows:

Table 3.3-1 Summary of Aging Management Programs for Auxiliary Systems Item Number Component Aging Aging Management Further Evaluation Discussion Effect/Mechanism Programs Recommended 3.3-1, 061 Reinforced concrete Cracking, loss of Chapter XI.M26, Fire No Consistent with NUREG-1801.

Structural fire barriers: material due to Protection, and As listed in Table 3.5.2-15, cracking, walls, ceilings and freeze-thaw, Chapter XI.S6, and loss of material of outdoor concrete floors exposed to aggressive Structures fire barrier hatches, walls, floors, and Air - outdoor chemical attack, Monitoring ceilings are managed by the Fire and reaction with Protection (B.2.3.15) AMP and aggregates Structures Monitoring (B.2.3.34) AMP.

The auxiliary boiler fuel oil storage tank berm/dike is not a maintenance rule structure and not included in the Structures Monitoring AMP (B.2.3.34). The function of the berm/dike is to contain oil spills and prevent the spread of fire. As listed in Table 3.5.2-115, cracking and loss of material of the outdoor reinforced concrete berm/dike around the auxiliary boiler fuel oil storage tank is managed by the Fire Protection (B.2.3.15) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 13 of 30 LRA Table 3.3-1, page 3.3-51, is revised as follows:

Table 3.3-1 Summary of Aging Management Programs for Auxiliary Systems Item Number Component Aging Aging Management Further Evaluation Discussion Effect/Mechanism Programs Recommended 3.3-1, 062 Reinforced concrete Loss of material due Chapter XI.M26, Fire No Consistent with NUREG-1801.

Structural fire barriers: to corrosion of Protection, and As listed in Table 3.5.2-15, loss of walls, ceilings and embedded steel Chapter XI.S6, material of indoor concrete fire barrier floors exposed to Air - Structures hatches, walls, floors, and ceilings is indoor, uncontrolled, Monitoring managed by the Fire Protection Air - outdoor (B.2.3.15) AMP and Structures Monitoring (B.2.3.34) AMP.

The auxiliary boiler fuel oil storage tank berm/dike is not a maintenance rule structure and not included in the Structures Monitoring AMP (B.2.3.34). The function of the berm/dike is to contain oil spills and prevent the spread of fire. As listed in Table 3.5.2-115, loss of material of the outdoor reinforced concrete berm/dike around the auxiliary boiler fuel oil storage tank is managed by the Fire Protection (B.2.3.15) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 14 of 30 LRA Table 3.5-1, page 3.5-55, is revised as follows:

Table 3.5-1 Summary of Aging Management Programs for Containments, Structures and Commodities Aging Effect Item Aging Management Further Evaluation Component Requiring Discussion Number Program Recommended Management 3.5-1, 054 All groups except 6: Cracking due to Chapter XI.S6, No Consistent with NUREG-1801 for Group 1, 3, concrete (accessible): expansion from Structures 4, 5, and 7 structures, as well as accessible all reaction with Monitoring areas of the SWIS (Group 6 structure) that are aggregates above-grade/water-line.

Group 2, and 9 structures are not present at CPNPP. No concrete foundations were evaluated with Group 8 (metallic) structures at CPNPP. The FWST foundation is a Yard Structure (Group 3).

The Structures Monitoring (B.2.3.34) AMP is credited with managing cracking due to expansion from reaction with aggregates.

The RG 1.127 Inspection of Water-Control Structures Associated with Nuclear Power Plants (B.2.3.35) AMP is credited with managing cracking due to expansion from reaction with aggregates for accessible concrete areas of the SWIS and Service Water Discharge Canal (Group 6 structures) exposed to a water-flowing environment.

As described in Table 3.5.2-15, concrete curbs include fire barriers. Both the Structures Monitoring (B.2.3.34) AMP and the Fire Protection (B.2.3.15) AMP will be used to manage cracking of concrete curbs in fire barriers exposed to indoor air, and the Structures Monitoring (B.2.3.34) AMP credits and communicates with the Fire Protection (B.2.3.15) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 15 of 30 LRA Table 3.5-1, page 3.5-60, is revised as follows:

Table 3.5-1 Summary of Aging Management Programs for Containments, Structures and Commodities 3.5-1, 066 Groups 1-5, 7, 9: Cracking; loss of Chapter XI.S6, No Consistent with NUREG-1801.

concrete (accessible): bond; and loss of Structures Group 2 and 9 structures are not present at interior and material (spalling, Monitoring CPNPP. No concrete foundations were above-grade exterior scaling) due to evaluated with Group 8 (metallic) structures corrosion of at CPNPP. The FWST foundation is a Yard embedded steel Structure (Group 3).

The Structures Monitoring (B.2.3.34) AMP will be used to manage cracking, loss of bond, and loss of material of accessible concrete in Groups 1, 3-7 structures exposed to air indoor, and air outdoor environments.

The Structures Monitoring (B.2.3.34) AMP also manages loss of bond and loss of material due to corrosion of masonry wall restraints, and reinforcements.

As described in Table 3.5.2-15, concrete curbs include fire barriers. Both the Structures Monitoring (B.2.3.34) AMP and the Fire Protection (B.2.3.15) AMP will be used to manage cracking of concrete curbs in fire barriers exposed to indoor air, and the Structures Monitoring (B.2.3.34) AMP credits and communicates with the Fire Protection (B.2.3.15) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 16 of 30 LRA Table 3.5-1, page 3.5-61, is revised as follows:

Table 3.5-1 Summary of Aging Management Programs for Containments, Structures and Commodities 3.5-1, 067 Groups 1-5, 7, 9: Increase in porosity Chapter XI.S6, No Consistent with NUREG-1801.

Concrete: Interior; and permeability; Structures Group 2 and 9 structures are not present at above-grade exterior, cracking; loss of Monitoring CPNPP. No concrete foundations were Groups 1-3, 5, material (spalling, evaluated with Group 8 (metallic) structures 7 concrete scaling) due to at CPNPP. The FWST foundation is a Yard (inaccessible): aggressive Structure (Group 3).

below-grade exterior; chemical attack foundation, Group 6: The Structures Monitoring (B.2.3.34) AMP will concrete be used to manage increase in porosity and (inaccessible): all permeability, cracking, and loss of material of inaccessible concrete in Groups 1, 3 through 7 structures.

As described in Table 3.5.2-15, concrete curbs include fire barriers. Both the Structures Monitoring (B.2.3.34) AMP and the Fire Protection (B.2.3.15) AMP will be used to manage cracking of concrete curbs in fire barriers exposed to indoor air, and the Structures Monitoring (B.2.3.34) AMP credits and communicates with the Fire Protection (B.2.3.15) AMP.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 17 of 30 LRA Table 3.5.2-3, page 3.5-95, is revised as follows:

Table 3.5.2-3: Diesel Generator Buildings - Summary of Aging Management Evaluation Aging Effect Aging Intended NUREG-1801 Table 1 Component Type Material Environment Requiring Management Notes Function Item Item Management Program Anchorage to Structural Carbon steel Air - indoor Loss of Structures III.A3.TP-248 3.5-1, 080 A building structure support uncontrolled material Monitoring (B.2.3.34)

Anchorage to Structural Carbon steel Air - indoor Loss of preload Structures III.A3.TP-261 3.5-1, 088 A building structure support uncontrolled Monitoring (B.2.3.34)

Anchorage to Structural Concrete Air - indoor Reduction in Structures III.B3.TP-42 3.5-1, 055 A building structure support (reinforced) uncontrolled concrete Monitoring anchor (B.2.3.34) capacity Anchorage to Structural Grout Air - indoor Reduction in Structures III.B3.TP-42 3.5-1, 055 A building structure support uncontrolled concrete Monitoring anchor (B.2.3.34) capacity Bolting (structural) Structural Carbon steel Air - indoor Loss of Structures III.A3.TP-248 3.5-1, 080 A support uncontrolled material Monitoring (B.2.3.34)

Bolting (structural) Structural Carbon steel Air - indoor Loss of preload Structures III.A3.TP-261 3.5-1, 088 A support uncontrolled Monitoring (B.2.3.34)

Bolting (structural) Structural Galvanized Air - indoor Loss of preload Structures III.B2.TP-261 3.5-1, 088 A support steel uncontrolled Monitoring (B.2.3.34)

Bolting (structural) Structural Galvanized Air - indoor None None III.B1.2.TP-8 3.5-1, 095 A support steel uncontrolled Concrete curb Direct flow Concrete Air - indoor Cracking Structures III.A3.TP-25 3.5-1, 054 A, 7 Fire barrier (reinforced) uncontrolled Monitoring (B.2.3.34)

Concrete curb Direct flow Concrete Air - indoor Cracking; loss Structures III.A3.TP-26 3.5-1, 066 A, 1, 7 Fire barrier (reinforced) uncontrolled of bond; and Monitoring loss of material (B.2.3.34)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 18 of 30 LRA Table 3.5.2-3, page 3.5-96, is revised as follows:

Table 3.5.2-3: Diesel Generator Buildings - Summary of Aging Management Evaluation Aging Effect Aging Intended NUREG-1801 Table 1 Component Type Material Environment Requiring Management Notes Function Item Item Management Program Concrete curb Direct flow Concrete Air - indoor Increase in Structures III.A3.TP-28 3.5-1, 067 A, 7 Fire barrier (reinforced) uncontrolled porosity and Monitoring permeability, (B.2.3.34) cracking, loss of material Concrete: Above Missile barrier Concrete Air - outdoor Cracking Structures III.A3.TP-25 3.5-1, 054 A grade exterior Structural (reinforced) Monitoring (accessible) support (B.2.3.34)

Shelter, protection Concrete: Above Missile barrier Concrete Air - outdoor Cracking; loss Structures III.A3.TP-26 3.5-1, 066 A, 1 grade exterior Structural (reinforced) of bond; and Monitoring (accessible) support loss of material (B.2.3.34)

Shelter, protection Concrete: Above Missile barrier Concrete Air - outdoor Increase in Structures III.A3.TP-28 3.5-1, 067 A grade exterior Structural (reinforced) porosity and Monitoring (accessible) support permeability, (B.2.3.34)

Shelter, cracking, loss protection of material Concrete: Above Missile Concrete Air - outdoor Loss of Structures III.A3.TP-23 3.5-1, 064 A grade exterior barrier (reinforced) material Monitoring (accessible) Structural (spalling, (B.2.3.34) support scaling) and Shelter, cracking protection Concrete: Above Missile barrier Concrete Water-flowing Increase in Structures III.A3.TP-24 3.5-1, 063 A, 2, 3 grade exterior Structural (reinforced) porosity and Monitoring (accessible) support permeability; (B.2.3.34)

Shelter, loss of strength protection

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 19 of 30 Table 3.5.2-3: Diesel Generator Buildings - Summary of Aging Management Evaluation Aging Effect Aging Intended NUREG-1801 Table 1 Component Type Material Environment Requiring Management Notes Function Item Item Management Program Concrete: Above Missile barrier Concrete Air - outdoor Cracking Structures III.A3.TP-204 3.5-1, 043 E grade exterior Structural (reinforced) Monitoring (inaccessible) support (B.2.3.34)

Shelter, protection Concrete: Above Missile barrier Concrete Air - outdoor Increase in Structures III.A3.TP-28 3.5-1, 067 A, 1 grade exterior Structural (reinforced) porosity and Monitoring (inaccessible) support permeability, (B.2.3.34)

Shelter, cracking, loss protection of material Concrete: Above Missile Concrete Air - outdoor Loss of Structures III.A3.TP-108 3.5-1, 042 E, 6 grade exterior barrier (reinforced) material Monitoring (inaccessible) Structural (spalling, (B.2.3.34) support scaling) and Shelter, cracking protection

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 20 of 30 LRA Table 3.5.2-3, page 3.5-100, is revised as follows:

Generic Notes A. Consistent with NUREG-1801 item for component, material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.

C. Component is different, but consistent with NUREG-1801 item for material, environment, and aging effect. AMP is consistent with NUREG-1801 AMP.

E. Consistent with NUREG-1801 item for material, environment, and aging effect, but a different AMP is credited or NUREG-1801 identifies a plant-specific AMP.

Plant-Specific Notes

1. Locations with prolonged or frequent water pooling that may contain aggressive chemicals or absorb into the concrete, and corrode embedded steel, are the most susceptible.

2. Bottom of the DGB foundation is located above the water table such that groundwater is limited to absorbed surface water and perched water.

3. Rainwater, particularly periodic heavy north central Texas rains, is considered water-flowing.

4. At locations of frequent or prolonged water pooling, such as the lower portion of the watertight doors.

5. Masonry (bricks) and mortar between silencer and DGB roof.

6. OE reflected in SLR-ISG-2021-03-STRUCTURES indicates that the Structures Monitoring (B.2.3.34) AMP may be credited in place of a plant specific program.

7. The Structures Monitoring (B.2.3.34) AMP and Fire Protection (B.2.3.15) AMP credit and communicate with each other.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 21 of 30 LRA Table 3.5.2-4, page 3.5-102, is revised as follows:

Table 3.5.2-4: Electrical and Control Building - Summary of Aging Management Evaluation Aging Effect Aging Intended NUREG-1801 Table 1 Component Type Material Environment Requiring Management Notes Function Item Item Management Program Concrete curb Direct flow Concrete Air - indoor Cracking Structures III.A3.TP-25 3.5-1, 054 A, 8 Fire barrier (reinforced) uncontrolled Monitoring (B.2.3.34)

Concrete curb Direct flow Concrete Air - indoor Cracking; loss Structures III.A3.TP-26 3.5-1, 066 A, 1, 8 Fire barrier (reinforced) uncontrolled of bond; and Monitoring loss of material (B.2.3.34)

Concrete curb Direct flow Concrete Air - indoor Increase in Structures III.A3.TP-28 3.5-1, 067 A, 8 Fire barrier (reinforced) uncontrolled porosity and Monitoring permeability, (B.2.3.34) cracking, loss of material Concrete: Above Missile Concrete Air - outdoor Cracking Structures III.A3.TP-25 3.5-1, 054 A grade exterior barrier (reinforced) Monitoring (accessible) Pressure (B.2.3.34) barrier Shelter, protection Structural support Concrete: Above Missile Concrete Air - outdoor Cracking; loss of Structures III.A3.TP-26 3.5-1, 066 A, 1 grade exterior barrier (reinforced) bond; and loss Monitoring (accessible) Pressure of material (B.2.3.34) barrier Shelter, protection Structural support

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 22 of 30 Table 3.5.2-4: Electrical and Control Building - Summary of Aging Management Evaluation Aging Effect Aging Intended NUREG-1801 Table 1 Component Type Material Environment Requiring Management Notes Function Item Item Management Program Concrete: Above Missile Concrete Air - outdoor Increase in Structures III.A3.TP-28 3.5-1, 067 A grade exterior barrier (reinforced) porosity and Monitoring (accessible) Pressure permeability, (B.2.3.34) barrier cracking, loss of Shelter, material protection Structural support Concrete: Missile Concrete Air - outdoor Loss of Structures III.A3.TP-23 3.5-1, 064 A Above grade barrier (reinforced) material Monitoring exterior Pressure (spalling, (B.2.3.34)

(accessible) Barrier scaling) and Shelter, cracking protection Structural support Concrete: Above Missile Concrete Water-flowing Increase in Structures III.A3.TP-24 3.5-1, 063 A, 3 grade exterior barrier (reinforced) porosity and Monitoring (accessible) Pressure permeability; (B.2.3.34) barrier loss of strength Shelter, protection Structural support

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 23 of 30 LRA Table 3.5.2-4, page 3.5-110, is revised as follows:

4. Concrete aging effect conservatively applied to masonry walls.

5. Exterior walls adjacent to other Seismic Category 1 structures are inaccessible due to the elastic joint filler in the gap.

6. Aging effect of cracking due to cracking, loss of bond and loss of material in concrete below grade exterior (inaccessible) will be evaluated and monitored under the Structures Monitoring (B.2.3.34) AMP.

7. OE reflected in SLR-ISG-2021-03-STRUCTURES indicates that the Structures Monitoring (B.2.3.34) AMP may be credited in place of a plant specific program.

8. The Structures Monitoring (B.2.3.34) AMP and Fire Protection (B.2.3.15) AMP credit and communicate with each other.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 24 of 30 LRA Table 3.5.2-11, page 3.5-150, is revised as follows:

Table 3.5.2-11: Yard Structures - Summary of Aging Management Evaluation Aging Effect Aging Intended NUREG-1801 Table 1 Component Type Material Environment Requiring Management Notes Function Item Item Management Program Anchorage to Structural Carbon steel Air - indoor Loss of material Structures III.A3.TP-248 3.5-1, 080 A building structure support uncontrolled Monitoring (B.2.3.34)

Anchorage to Structural Carbon steel Air - indoor Loss of preload Structures III.A3.TP-261 3.5-1, 088 A building structure support uncontrolled Monitoring (B.2.3.34)

Anchorage to Structural Carbon steel Air - outdoor Loss of material Structures III.A3.TP-274 3.5-1, 082 A building structure support Monitoring (B.2.3.34)

Anchorage to Structural Carbon steel Air - outdoor Loss of preload Structures III.A3.TP-261 3.5-1, 088 A building structure support Monitoring (B.2.3.34)

Anchorage to Structural Concrete Air - outdoor Reduction in Structures III.B3.TP-42 3.5-1, 055 A building structure support (reinforced) concrete anchor Monitoring capacity (B.2.3.34)

Anchorage to Structural Grout Air - outdoor Reduction in Structures III.B3.TP-42 3.5-1, 055 A building structure support concrete anchor Monitoring capacity (B.2.3.34)

Berm/Dike Fire barrier Concrete Air - outdoor Cracking. Loss Fire Protection VII.G.A-92 3.5-1, 061 E, 9 (reinforced) of material (B.2.3.15)

Berm/Dike Fire barrier Concrete Air - outdoor Loss of Fire Protection VII.G.A-93 3.5-1, 062 E, 9 (reinforced) material (B.2.3.15)

Bolting Structural Carbon steel Air - indoor Loss of material Structures III.A3.TP-248 3.5-1, 080 A (structural) support uncontrolled Monitoring (B.2.3.34)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 25 of 30 LRA Table 3.5.2-11, page 3.5-156, is revised as follows:

Plant-Specific Notes

1. Locations with prolonged or frequent water pooling that may contain aggressive chemicals or absorb into the concrete, and corroded embedded steel, are the most susceptible.

2. Bottom of the duct banks, enclosure pits, foundations, handholes, manholes, and slabs in the yard are located above the water table such that groundwater is limited to absorbed surface water and perched water.

3. Concrete aging effect conservatively applied to masonry walls.

4. Rainwater, particularly periodic heavy north central Texas rains, is considered water-flowing.

5. Stainless steel liner anchored to the interior wall of the CST, RMWST, and RWST for each unit making the interior wall below the top of the liner inaccessible.

6. Inside the CST, RMWST, or RWST wall and vented roof above the liner and water-line or diaphragm, as applicable.

7. Seismic Category 1 tunnels to the CST, RMWST, and RWST contain piping and conduit, include an above-grade portion for tank penetrations, and are accessible and considered as another SGB room/area.

8. OE reflected in SLR-ISG-2021-03-STRUCTURES indicates that the Structures Monitoring (B.2.3.34) AMP may be credited in place of a plant specific program.

9. The Fire Protection (B.2.3.15) AMP alone manages the aging of the berm/dike around the auxiliary boiler fuel oil storage tank. The berm/dike is located outside the protected area and has a conservative fire barrier intended function.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 26 of 30 LRA Table 3.5.2-15, page 3.5-185, is revised as follows:

Table 3.5.2-15: Fire Barrier Commodity Group - Summary of Aging Management Evaluation Commodity Intended Material Environment Aging Effect Aging NUREG-1801 Table 1 Notes Function Requiring Management Item Item Management Program Berm/Dike Fire barrier Concrete Air - outdoor Cracking. Loss Fire Protection VII.G.A-92 3.3-1, 061 E, 8 (reinforced) of material (B.2.3.15)

Berm/Dike Fire barrier Concrete Air - outdoor Loss of material Fire Protection VII.G.A-93 3.3-1, 062 E, 8 (reinforced) (B.2.3.15)

Concrete block Fire barrier Masonry block Air - indoor Cracking Masonry Walls III.A3.T-12 3.5-1, 070 A, 1 (removable) for uncontrolled (B.2.3.33) opening Concrete block Fire barrier Masonry block Air - indoor Cracking; loss of Structures III.A3.TP-26 3.5-1, 066 C, 6 (removable) for uncontrolled bond; and loss of Monitoring opening material (B.2.3.34)

Concrete block Fire barrier Masonry block Air - indoor Cracking Fire Protection III.A3.T-12 3.5-1, 070 E, 1 (removable) for uncontrolled (B.2.3.15) opening Concrete curb Fire barrier Concrete Air - indoor Cracking Fire Protection III.A3.TP-25 3.5-1, 054 E, 7 (reinforced) uncontrolled (B.2.3.15)

Concrete curb Fire barrier Concrete Air - indoor Cracking; loss Fire Protection III.A3.TP-26 3.5-1, 066 E, 7 (reinforced) uncontrolled of bond; and (B.2.3.15) loss of material Concrete curb Fire barrier Concrete Air - indoor Increase in Fire Protection III.A3.TP-28 3.5-1, 067 E, 7 (reinforced) uncontrolled porosity and (B.2.3.15) permeability, cracking, loss of material Damper housing Fire barrier Galvanized steel Air - indoor None None III.B4.TP-8 3.5-1, 095 C uncontrolled

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 27 of 30 Table 3.5.2-15: Fire Barrier Commodity Group - Summary of Aging Management Evaluation Commodity Intended Material Environment Aging Effect Aging NUREG-1801 Table 1 Notes Function Requiring Management Item Item Management Program Damper housing Fire barrier Galvanized steel Air with borated Loss of Material Boric Acid III.B4.TP-3 3.5-1, 089 C water leakage Corrosion (B.2.3.4)

Door Fire barrier Carbon steel Air - indoor Loss of material Fire Protection VII.G.A-21 3.3-1, 059 A uncontrolled (B.2.3.15)

Door Fire barrier Galvanized steel Air - indoor Loss of material Fire Protection VII.G.A-21 3.3-1, 059 A uncontrolled (B.2.3.15)

Door Fire barrier Carbon steel Air - outdoor Loss of material Fire Protection VII.G.A-22 3.3-1, 059 A (B.2.3.15)

Door Fire barrier Galvanized steel Air - outdoor Loss of material Fire Protection VII.G.A-22 3.3-1, 059 A (B.2.3.15)

Door Fire barrier Carbon steel Air with borated Loss of material Boric Acid III.B2.T-25 3.5-1, 089 C water leakage Corrosion (B.2.3.4)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 28 of 30 LRA Table 3.5.2-15, page 3.5-190, is revised as follows:

Plant-Specific Notes

1. Removable concrete blocks for openings in certain walls to facilitate equipment removal and replacement; in areas where a removable concrete block opening exists in a fire wall, a fire hazards analysis evaluation justifies the as-built design, and the concrete blocks are installed in such a way that there are no through openings from one side of the barrier to the other. Furthermore, the Masonry Walls (B.2.3.33) AMP and Fire Protection (B.2.3.15) AMP credit and communicate with each other.

2. This material is not addressed for fire barriers in NUREG-1801. Consistent with the OE reflected in SLR-ISG-2021-02-MECHANICAL (items VII.G.A-805 to VII.G.A-807; SRP items 3.3-1, 267 to 3.3-1, 269), aging of the component materials is managed by the Fire Protection (B.2.3.15) AMP.

3. The Masonry Walls (B.2.3.33) AMP and Fire Protection (B.2.3.15) AMP credit and communicate with each other.

4. Gypsum drywall is utilized throughout the plant to provide a fire barrier which is lightweight and where unit masonry or concrete is not feasible. This lightweight fire barrier material is not addressed in NUREG-1801; however, aging is managed by the Fire Protection (B.2.3.15) AMP.

5. Relative to stainless-steel components located outdoors, the Structures Monitoring (B.2.3.34) AMP is focused on areas with potential for frequent or prolonged water pooling and communicates with the Fire Protection (B.2.3.15) AMP as warranted. The penetration sleeves located in an Air - outdoor environment are managed for the loss of material aging effect by the listed monitoring AMP.

6. Concrete aging effect conservatively applied to masonry walls.

7. The Structures Monitoring (B.2.3.34) AMP and Fire Protection (B.2.3.15) AMP credit and communicate with each other.

8. The Fire Protection (B.2.3.15) AMP alone manages the aging of the berm/dike around the auxiliary boiler fuel oil storage tank. The berm/dike is located outside the protected area and has a conservative fire barrier intended function.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 29 of 30 LRA Appendix A, Table A-3, page A-60, is revised as follows:

Table A-3 List of LR Commitments and Implementation Schedule Aging Management NUREG-1801 No. Program or Activity Commitment Implementation Schedule Section (Section) b) Ensure procedures performing air quality analysis describe review of analysis results and comparison of previous results.

c) Ensure procedures trend dewpoint temperature readings.

d) Ensure air sampling procedures describe the corrective actions taken if air samples are unsatisfactory.

17 Fire Protection XI.M26 Continue the existing Fire Protection AMP, including enhancements to: No later than 6 months prior to (A.2.2.15) the PEO, i.e.:

a) Expand the sample size of inspected fire penetration seals if any sign of degradation is found in the sample. U1: 08/08/2029 U2: 08/02/2032, b) Require qualified fire protection personnel perform inspections associated with the Fire Protection AMP. or no later than the last refueling outage prior to the PEO.

c) Revise penetration seal inspection procedures to include a requirement to inspect not less than 10% of each type of seal in walkdowns performed at a frequency in accordance with the plants NRC-approved fire protection program or at least once every refueling outage.

d) Revise Fire Rated Assembly Visual Inspection procedure to include a requirement to inspect the Auxiliary Boiler Fuel Oil Storage tank Concrete berm/dike as a part of Section 8.4 -

Fire Walls, Floors and Ceiling Inspections.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-2 TXX-23048 and CP-202300322 Attachment G Page 30 of 30 LRA Appendix B, Section B.2.3.15, page B-104, is revised as follows:

Element Affected Enhancement

4. Detection of Aging Expand the sample size of inspected fire penetration seals if any Effects sign of degradation is found in the sample.

Require qualified fire protection personnel perform inspections associated with the Fire Protection AMP.

Revise penetration seal inspection procedures to include a requirement to inspect not less than 10% of each type of seal in walkdowns performed at a frequency in accordance with the plants NRC-approved fire protection program or at least once every refueling outage.

Revise Fire Rated Assembly Visual Inspection procedure to include a requirement to inspect the Auxiliary Boiler Fuel Oil Storage tank Concrete berm/dike as a part of Section 8.4 -

Fire Walls, Floors and Ceiling Inspections.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.5.2.2.1.2-1 TXX-23048 and CP-202300322 Attachment H Page 1 of 3 LRA Section: 3.5.2.2.1.2, Reduction of Strength and Modulus due to Elevated Temperature NRC RAI No: 3.5.2.2.1.2-1 (Stainless Steel Insulation)

Regulatory Basis:

Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. As described in the SRP-LR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL-LR Report when evaluation of the matter in the GALL-LR Report applies to the plant.

Background:

LRA Section 3.5.2.2.1.2 states that local area temperatures may be elevated above general area temperatures due to process piping that carry high temperatures fluids, and that the Reactor Containment Building (RCB) penetration and reactor coolant piping insulation contributes to keeping the local concrete temperatures of the RCB and Primary Shield Wall (PSW) below 200F during normal plant operation. As such, this insulation has a license renewal intended function as described in Table 2.4-1 and is subject to AMR as listed in Table 3.5.2-1.

Table 2 item associated with AMR item 3.5-1, 095 (GALL-LR Report item III.B1.1.TP-8) in LRA Table 3.5.2-1 lists a component of thermal insulation (high temperature penetration) with material of stainless steel that does not require aging management.

GALL-LR Report item III.B1.1.TP-8 (AMR item 3.5-1, 095) lists a component (Aluminum, galvanized steel and stainless-steel support members; welds; bolted connections; support anchorage to building structure) exposed to air-indoor uncontrolled environment. This component does not include thermal insulation.

Issue:

Thermal insulation (high-temperature penetration) is used to keep the local concrete temperatures of the RCB and PSW below 200F during normal plant operation. It is unclear how the thermal insulation can use GALL-LR Report item III.B1.1.TP-8, and it is unclear why the thermal insulation does not require aging management.

Request:

Evaluate applicability of GALL-LR Report item III.B1.1.TP-8 used for Table 2 item associated with AMR item 3.5-1, 095 in Table 3.5.2-1 and clarify what aging management program is used to adequately manage aging effects for the stainless-steel thermal insulation. Provide justification if aging management of the stainless-steel thermal insulation

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.5.2.2.1.2-1 TXX-23048 and CP-202300322 Attachment H Page 2 of 3 is not needed.

Luminant Response:

Thermal insulation (high-temperature penetration) is used to keep the local concrete temperatures of the RCB and PSW below 200F during normal plant operation, using stainless steel reflective metal insulation as described in plant-specific note 7 of LRA Table 3.5.2-1. NUREG-1801 Item III.B1.1.TP-8 and associated NUREG-1800 Item 3.5-1, 095 addresses the material and environment of stainless steel in indoor-air not requiring aging management. Note C reflects that the component type is different than the component type in NUREG-1801. NUREG-1801 distinguishes aging effects associated with a given material and environment combination that could impact the intended function of the associated component. Since the metallic insulation evaluated in CPNPP Table 3.5.2-1 is a different component type than described for this material/environment combination in NUREG-1801 Item III.B1.1.TP-8, it is appropriate to consider whether the intended function for the differing component type would require consideration of a different aging effect.

LR-ISG-2012-02 defined a unique aging effect associated with the component type of insulation as follows:

IX.E Reduced thermal Impairment of thermal insulations ability to resist insulation resistance the transfer of heat between the ambient environment and the insulated structure or component. This is caused by the degradation of the insulation that typically occurs when insulation is exposed to moisture.

LR-ISG-2012-02 did not identify any unique aging effects for metallic insultation, but did establish the following line items for aging of nonmetallic thermal insulation:

Item Structure Material Environment Aging Aging and/or Effect Management Component /Mechanism Program VII.I.S- Jacketed Calcium Air-indoor Reduced Chapter 403 Insulation silicate, uncontrolled thermal XI.M36 fiberglass or air- insulation External outdoor resistance Surface due to Monitoring of moisture Mechanical intrusion Equipment VII.I.S- Jacketed Foamglas Air-indoor Reduced Chapter

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.5.2.2.1.2-1 TXX-23048 and CP-202300322 Attachment H Page 3 of 3 404 Insulation (glass dust) uncontrolled thermal XI.M36 or air- insulation External outdoor resistance Surface due to Monitoring of moisture Mechanical intrusion Equipment Reflective metallic insulation is not expected to be vulnerable to impairment of its resistance to heat transfer related to moisture intrusion since its design relies on different performance characteristics than those associated with calcium silicate, fiberglass or Foamglas.

Reflective metallic insulation primarily resists the transfer of radiant energy and is generally designed and constructed to be resistant to moisture accumulation. Insulation materials such as calcium silicate, fiberglass and Foamglas are susceptible to moisture absorption or wicking and are designed to be in dry installations that allow them to resist conductive energy. The intrusion of moisture in these materials can increase the potential for conduction of heat through the insulating material. The aging effect of reduced thermal insulation resistance due to moisture intrusion does not similarly apply to metallic reflective insulation as it is unlikely to absorb or wick moisture and any minor moisture that could be present in this insulating material installed on hot piping is unlikely to alter the radiant energy resistance.

A search of available station and industry operating experience for a ten-year period under the keywords metallic insulation, thermal insulation thermal resistance and reflective insulation found no instance of reduced thermal insulation resistance associated with reflective metallic insulation.

Metallic insulation was evaluated consistent with industry guidance which reflects that stainless steel insulation that is not exposed to an aggressive environment does not experience aging requiring management. No unique aging effects for the component type of insulation have been identified for reflective metallic insulation, therefore NUREG-1801 Item III.B1.1.TP-8 and associated NUREG-1800 Item 3.5-1, 095 for stainless steel in indoor-air is appropriate for this component type at CPNPP. Stainless steel reflective metal insulation does not require aging management, as described above.

References:

LR-ISG-2012-02 Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks and Corrosion Under Insulation Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.2.1-1 TXX-23048 and CP-202300322 Attachment I Page 1 of 2 LRA Section: 4.2.1, Neutron Fluence Projections NRC RAI No: 4.2.1-1 (Least-Squares Adjustment of RPV Neutron Exposures)

Regulatory Basis:

10 CFR 54.21(c) requires an applicant to evaluate time-limited aging analyses and demonstrate that the analyses remain valid for the period of extended operation, the analyses have been projected to the end of the period of extended operation, or that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to time-limited aging analyses (TLAAs) that have been identified to require review under 10 CFR 50.21, such that there is reasonable assurance that any changes made to the plants CLB in order to comply with this paragraph are in accord with the Act and the Commissions regulations. In order to complete its review and enable making a finding under 10 CFR 54.29(a), the staff requires additional information in regard to the matters described below.

Background:

SRP-LR Section 4.2.3.1.1 states that for a disposition in accordance with 10 CFR 54.21(c)(1)(ii), the new or updated reactor pressure vessel (RPV) neutron fluence analysis is evaluated to consider whether the applicant identifies the staff-approved methodology used to determine the neutron fluence, and whether the methodology follows the guidance in NRC RG 1.190.

In Section 4.2.1 of the license renewal application (LRA), the applicant dispositions the TLAA by projecting the fluence analyses to the end of the period of extended operation pursuant to 10 CFR 54.21(c)(1)(ii). In Attachment W1 to Supplement 2 to the LRA, the applicant states that fluence projections were performed with a methodology that is consistent with that described in WCAP-18124-NP-A, Revision 0 and WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A, Revision 0.

Issue:

The letter accompanying the NRC staff safety evaluation for WCAP-18124-NP, Revision 0 states that the NRC staff has found that WCAP-18124-NP, Revision 0 is acceptable for referencing in licensing application provided that the limitations and conditions stipulated in the Section 4.0 and applicability defined in the enclosed NRC final SE are met along with the proper documentation. Item 2 in Section 4.0 of the NRC staff SE states: Least squares adjustment is acceptable if adjustments to the M/C ratios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross-sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors.

If errors cannot be found, the values causing the inconsistency should be disqualified.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.2.1-1 TXX-23048 and CP-202300322 Attachment I Page 2 of 2 However, the LRA does not specify whether least-squares adjustment was used in calculating RPV fluence values.

Request:

Clarify whether least-squares adjustment was applied to RPV neutron fluence values used in reactor vessel neutron embrittlement TLAAs. If so, provide the evaluation performed to determine whether individual values should be disqualified.

Luminant Response:

The neutron exposures used in the reactor vessel neutron embrittlement TLAAs performed in support of the Comanche Peak LRA did not include any least-squares adjustments.

References:

None Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.2.1-2 TXX-23048 and CP-202300322 Attachment J Page 1 of 2 LRA Section: A.3.2.1, Neutron Fluence Analysis NRC RAI No: 4.2.1-2 (Neutron Fluence Analysis Methodology)

Regulatory Basis:

10 CFR 54.21(d) requires an applicant to provide an FSAR supplement for the facility that contains a summary description of the evaluation of time-limited aging analyses for the period of extended operation determined by 10 CFR 54.21(c).

Background:

SRP-LR Section 4.2.2.2 states that the specific criterion for meeting 10 CFR 54.21(d) is that the summary description of the evaluation of TLAAs for the period of extended operation in the FSAR supplement is sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59. It also states that the description contains information associated with the TLAAs regarding the basis for determining that the applicant has made the demonstration required by 10 CFR 54.21(c)(1).

Section A.3.2.1, Neutron Fluence Analysis of Appendix A to the license renewal application (LRA) contains the portion of the FSAR supplement pertaining to neutron fluence analysis.

Issue:

Section A.3.2.1 of Appendix A to the LRA does not describe the method of evaluation for the neutron fluence analysis. Although this information is provided in the LRA itself, the FSAR supplement is not sufficiently comprehensive such that later changes can be controlled by 10 CFR 50.59.

Additionally, Section A.3.2.1 of Appendix A to the LRA does not state whether the method has been approved by the NRC or whether it is consistent with RG 1.190. Although this information is provided in the LRA itself, the FSAR supplement does not contain information associated with the TLAAs regarding the basis for determining that the applicant has made the demonstration required by 10 CFR 54.21(c)(1).

Request:

Revise the FSAR supplement to describe the method of evaluating the neutron fluence and state whether the methodology is consistent with RG 1.190 and has been approved by the NRC. Submit the revised FSAR supplement pages on the docket for review.

Luminant Response:

LRA Section A.3.2.1 of Appendix A is revised to state that the methodology used to calculate the 60-year reactor pressure vessel neutron exposures is consistent with RG 1.190, has been

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.2.1-2 TXX-23048 and CP-202300322 Attachment J Page 2 of 2 approved by the USNRC, and is described in detail in WCAP-18124-NP-A (Reference 1) and WCAP-18124-NP-A Revision 0 Supplement 1-NP-A (Reference 2).

References:

1. WCAP-18124-NP-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET, July 2018. (Report Accession No. ML18204A010, Verification Letter Accession No. ML18220A854)

2. WCAP-18124-NP-A Revision 0 Supplement 1-P-A, Revision 0, Fluence Determination with RAPTOR-M3G and FERRET - Supplement for Extended Beltline Materials, May 2022. (Report Accession No. ML22153A139, Verification Letter Accession No. ML22200A029)

Associated LRA Revisions:

LRA Section A.3.2.1 of Appendix A, page A-32, is revised as follows:

The neutron fluence analysis is a TLAA as defined by 10 CFR 54.21(c) and must be evaluated for the increased neutron fluence associated with 60 years of operation. These neutron fluence projections are used as input to the analyses for fracture toughness, or USE, PTS, Reference Temperature - Nil Ductility Transition (RTNDT), ART, Low-Temperature Overpressure Protection (LTOP) limits, and Reactor Vessel Pressure-Temperature Curves.

The methodology used to calculate the 60-year reactor vessel neutron fluence projections is consistent with RG 1.190, has been approved by the USNRC, and is described in detail in WCAP-18124-NP-A and WCAP-18124-NP-A Revision 0 Supplement 1-NP-A. The fluence analyses have been projected to the end of the PEO for CPNPP Units 1 and 2 in accordance with 10 CFR 54.21(c)(1)(ii).

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.34-1 TXX-23048 and CP-202300322 Attachment K Page 1 of 3 LRA Section: B.2.3.34, Structures Monitoring NRC RAI No: B.2.3.34-1 (Enhancements Related to Cracking due to Expansion from Reaction with Aggregates)

Regulatory Basis:

Title 10 of the Code of Federal Regulations Section 54.21(a)(3) requires the applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. As described in the SRP-LR, an applicant may demonstrate compliance with 10 CFR 54.21(a)(3) by referencing the GALL-LR Report when evaluation of the matter in the GALL-LR Report applies to the plant.

Background:

The Structures Monitoring and the Inspection of Water-Control Structures Associated with Nuclear Power Plants aging management programs (AMPs) provide an enhancement to the Parameters Monitored or Inspected program element which relates to visually inspecting concrete structures for unique cracking such as "craze", "mapping" or "patterned" cracking to determine the presence of alkali-silica gel.

LRA Table 3.5-1 indicates that the ASME Section XI, Subsection IWL AMP will continue to inspect and monitor for cracking and indications of Alkali-Silica Reaction (ASR) - induced or similar degradation. The staff reviewed the Westinghouse proprietary procedure TX-ISI-IWL, Revision 6, and finds that the procedure implements the inspection of pattern cracking and exudation, which are related to the ASR.

GALL-LR Report does not clearly define visual indications of aggregate reactions. However, the GALL-SLR Report provides visual indications of aggregate reactions, such as map or patterned cracking, alkali-silica gel exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components, which are applicable to the initial period of extended operation.

Issue:

ASR, or reaction with aggregates, has more characteristics than what has been enhanced or implemented in the Comanche Peak LRA AMPs mentioned above.

The enhanced Structures Monitoring, enhanced Inspection of Water-Control Structures Associated with Nuclear Power Plants, and the ASME Section XI, Subsection IWL AMPs may not be adequate to detect the indications of aggregate reactions, based on operating experience gained since the GALL-LR was developed.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.34-1 TXX-23048 and CP-202300322 Attachment K Page 2 of 3 Request:

1. Evaluate and update the enhancements to the Parameters Monitored or Inspected program element for the Structures Monitoring and the Inspection of Water-Control Structures Associated with Nuclear Power Plants AMPs to ensure that the AMPs are capable of detecting visual indications of aggregate reactions, such as map or patterned cracking, alkali-silica gel exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components.

2. Provide the enhancement to the Parameters Monitored or Inspected program element for the ASME Section XI, Subsection IWL AMP to ensure that the AMP is capable of detecting visual indications of aggregate reactions, such as map or patterned cracking, alkali-silica gel exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components. Otherwise, provide justification why the enhancement is not needed.

Luminant Response:

NUREG-1801 Revision 2 includes cracking due to expansion from reaction with aggregates as an aging effect requiring management for concrete in any environment and specifies the ASME Section XI, Subsection IWL AMP and the Structures Monitoring AMP for management of this aging effect. The NUREG-1801 XI.S2 ASME Section XI, Subsection IWL and XI.S6 Structures Monitoring AMPs include guidance under the Parameters Monitored or Inspected elements related to cracking, but neither NUREG-1801 AMP provides guidance specific to cracking associated with expansion due to reaction with aggregates (alternately described as alkali-silica reaction or ASR). In order to address the industry operating experience discussed in NRC Information Notice (IN) 2011-20 (ADAMs Accession No. ML112241029), as described in LRA Sections B.2.3.34 and B.2.3.35, an enhancement was added to the CPNPP Structures Monitoring AMP and the CPNPP RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP to include additional guidance for identifying the type of cracking associated with aggregate reactions (craze, mapping or patterned cracking). This approach is consistent with NUREG-1801 guidance plus additional guidance from industry operating experience.

NUREG-2192, which is the Standard Review Plan for subsequent license renewal (SRP-SLR), also discusses cracking due to expansion from reaction with aggregates in Section 3.5.3.2.1.8 and states that a plant-specific AMP is necessary if (1) reactivity tests or petrographic examinations of concrete samples identify reaction with aggregates, or (2) accessible concrete exhibits visual indications of aggregate reactions, such as map or patterned cracking, alkali-silica gel exudations, surface staining, expansion causing structural deformation, relative movement or displacement, or misalignment/distortion of attached components. CPNPP has no plant-specific operating experience that indicates aggregate reaction degradation (consistent with the OE reflected in SRP-SLR Sections 3.5.3.2.1.8 and 3.5.3.2.2.1, item 2).

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.34-1 TXX-23048 and CP-202300322 Attachment K Page 3 of 3

1. The implementing procedure for the CPNPP Structures Monitoring AMP and the CPNPP RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP currently includes guidance for visual inspection related to pattern cracking consisting of fine openings in the concrete surface that are in the form of a pattern. Pattern cracking is further defined in the procedure as fine openings on concrete surfaces in the form of a pattern; resulting from a decrease in volume of the material near the surface, or an increase in volume of the material below the surface, or both.

The existing enhancement to the CPNPP Structures Monitoring AMP and the CPNPP RG 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants AMP directly incorporates guidance from NRC IN 2011-20 (ADAMs Accession No. ML112241029), which states that ASR can be identified as a likely cause of degradation during visual inspection by the unique craze, map or patterned cracking and the presence of alkali-silica gel.

Current inspection procedures for the CPNPP Structures Monitoring AMP and the CPNPP RG 1.127, Inspection of Water Control Structures Associated with Nuclear Power Plants AMP includes the following types of indications: (1) deflection, (2) deformation, (3) warping, and (4) discoloration. These indications, along with the existing enhancement, are adequate to identify cracking due to expansion from reaction with aggregates if it occurs in the future.

2. The CPNPP ASME Section XI, Subsection IWL AMP is performed in accordance with ASME Section XI, Subsection IWL. Section IWL-2310 includes requirements for visual examination of concrete containments and states that general visual examination shall be performed in sufficient detail to identify areas of concrete deterioration and distress, such as described in ACI 201.1 and ACI 349.3R. ACI 201.1 identifies map cracking, pattern cracking, deflection, deformation, exudation, warping, and discoloration as types of imperfections and distresses to identify during visual examinations. Following this guidance will ensure that any evidence of cracking due to expansion from reaction with aggregates will be identified and no program enhancements are necessary.

References:

None.

Associated LRA Revisions:

No LRA changes have been identified as a result of this response.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.3-1 TXX-23048 and CP-202300322 Attachment L Page 1 of 4 LRA Section: 4.3.3, ASME Section III, Class 2 and 3 Allowable Stress Analyses and A.3.3.3, ASME Section III, Class 2 and 3 and ANSI B31.1 Allowable Stress Analyses NRC RAI No: 4.3.3-1 (Mention of ANSI B31.1 Missing from Title and Supporting Text of LRA and FSAR)

Regulatory Basis:

Pursuant to 10 CFR 54.21(c), the LRA must include an evaluation of time-limited aging analyses (TLAAs). The applicant must demonstrate that (i) the analyses remain valid for the period of extended operation, (ii) the analyses have been projected to the end of the period of extended operation, or (iii) the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

Background:

LRA Section 4.3.3 addresses the TLAA on allowable stresses for ASME Code Section III, Class 2 and 3 and ANSI B31.1 piping systems. The TLAA evaluation in LRA Section 4.3.3 explains that the TLAA demonstrates the applicable cycles for 60 years of operation for ASME Code Class 2 and 3 and ANSI B31.1 piping systems remain below the 7000 cycle limit.

Issue:

In contrast with the TLAA evaluation, the title and TLAA disposition of LRA Section 4.3.3 do not include ANSI B31.1 piping. In addition, the updated safety analysis report (USAR) supplement description in LRA Section A.3.3.3 does not include a reference to ANSI B31.1 piping.

Given the inclusion of ANSI B31.1 piping in some areas of the application and not in other relevant areas, it is unclear whether the disposition of the TLAA applies to ANSI B31.1 piping as well.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.3-1 TXX-23048 and CP-202300322 Attachment L Page 2 of 4 Request:

Given the inclusion of ANSI B31.1 piping in the TLAA evaluation in LRA Section 4.3.3, provide justification for why the ANSI B31.1 piping is not referenced in the title and TLAA disposition of LRA Section 4.3.3 and the USAR supplement (LRA Section A.3.3.3). If it cannot be justified, revise the LRA to resolve this inconsistency. As part of the response, clarify whether the ANSI B31.1 piping is included in the TLAA on allowable stresses.

Luminant Response:

ANSI B31.1 piping should be referenced in the title and TLAA disposition of LRA Section 4.3.3 and the FSAR supplement (LRA Section A.3.3.3). The LRA is revised to resolve this inconsistency.

References:

None.

Associated LRA Revisions:

LRA Section 4.3.3, Section Title (page 4.3-10), is revised as follows:

ASME Section III, Class 2 and 3 and ANSI B31.1 Allowable Stress Analyses LRA Section 4.3.3, Paragraph 3 (page 4.3-10), is revised as follows:

The total RCS projected cycles were conservatively obtained from the summation of the normal, upset, and test condition transient projected cycles for each unit. Portions of each system under investigation are subject to RCS transients as well as the auxiliary transients.

Therefore, the total projected 60-year cycles from both RCS and auxiliary transients are compared to the ASME Section III allowable of 7,000 cycles for each system.

LRA Section 4.3.3, Paragraph 7 (page 4.3-10), is revised as follows:

Each of the ASME Class 2 and 3 and ANSI B31.1 lines within the scope of this application remain bounded by the ASME Section III 7,000 cycle limit. Therefore, the maximum allowable stress range values for the existing fatigue analyses remain valid because the allowable limit for the number of thermal range transient cycles will not be exceeded during the PEO.

LRA Section 4.3.3, Paragraph 8 (page 4.3-11), is revised as follows:

Table 4.3.1-4 shows that each of the ASME Class 2 and 3 and ANSI B31.1 auxiliary lines within the scope of this application remain bounded by the ASME Section III 7,000 cycle limit.

Therefore, the maximum allowable stress range values for the existing fatigue analyses remain valid because the allowable limit for the number of full thermal range transient cycles

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.3-1 TXX-23048 and CP-202300322 Attachment L Page 3 of 4 will not be exceeded during the PEO.

LRA Section 4.3.3, Paragraph 9 (page 4.3-11), is revised as follows:

Aging Management, 10 CFR 54.21(c)(1)(iii) - The Fatigue Monitoring (B.2.2.1)

Program will be used to monitor transient cycles and ensure the numbers of transients analyzed in the ASME Section III Class 2 and 3 and ANSI B31.1 lines within the scope of this application remain valid for the PEO.

LRA Table 4.3.1-4 in Section 4.3.1 (page 4.3-8) is revised as follows:

Table 4.3.1-4 CPNPP 60-year Projected Transient Cycle Totals for Non-Class 1 Systems and Piping CPNPP Projected Cycles ASME Section III Piping Line/System Unit Unit Class 2 and 3 1 2 Cycle Limit RHR Hot Leg Loops 1 and 4 4001 4167 7000 ECCS/BIT Cold Legs 1, 2, 3, and 4 3961 4116 7000 Accumulator Cold Legs 1, 2, 3, and 4, SI/RHR Return, and SIS 3959 4114 7000 Connection to Accumulator RCP Seal Water Injection Line 4055 4225 7000 Normal and Alternate Charging and 4228 4341 7000 Letdown Branch Line Normal Letdown and Excess 3940 4094 7000 Letdown Branch Line Pressurizer Auxiliary Spray Line 3928 4083 7000 Pressurizer Safety and Relief Valve 3993 4153 7000 Inlet Line Feed Water Piping 897 676 7000 Main Steam Line - Piping Upstream 3926 4081 7000 of MISVs Main Steam Line - Piping 477 271 7000 Downstream of MSIVs Main Steam Line - Admitted for Operation of the Turbine Drive 3225 4759 7000 Auxiliary Feed Water Pump Process Sampling - Hot Leg 4756 3890 7000 Isolation Process Sampling - Pressurizer 5183 5195 7000 Liquid Sample Isolation Liquid Waste Processing 710 690 7000

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 4.3.3-1 TXX-23048 and CP-202300322 Attachment L Page 4 of 4 Steam Generator Blowdown and 5613 5635 7000 Heater Drains Auxiliary Steam 2947 2863 7000 LRA Section A.3.3.3 (page A-36), is revised as follows:

Piping designed in accordance with ASME Section III Class 2 and 3 and ANSI B31.1 design rules is 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 codes first require prediction of the overall number of full thermal range transient 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 of 1.0 is applied that would not reduce the allowable stress value. For high numbers of cycles, a stress range reduction factor of less than 1.0 is applied that limits the allowable stresses applied to the piping, which reduces the likelihood of failure due to cyclic loading.

The ASME Section III 7,000 cycle limit bounds the 60-year projected cycles for the ASME Section III Class 2 and 3 and ANSI B31.1 lines within the scope of this application. Therefore, the allowable stress calculations for ASME Section III Class 2 and 3 and ANSI B31.1 lines within the scope of this application remain valid for the PEO. The Fatigue Monitoring (A.2.1.1)

AMP will be used to monitor transient cycles and ensure the number of transients analyzed in the ASME Section III, Class 2 and Class 3 and ANSI B31.1 fatigue analyses will not be exceeded during the PEO in accordance with 10 CFR 54.21(c)(1)(iii).

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.2.2-1 TXX-23048 and CP-202300322 Attachment M Page 1 of 3 LRA Section: 3.3.2.2.2, Cracking due to Stress Corrosion Cracking and Cyclic Loading NRC RAI No: 3.3.2.2.2-1 (Radiation Monitoring in the Closed Cooling Water System)

Regulatory Basis:

10 CFR 54.21(a)(3) requires an applicant to demonstrate that the effects of aging for structures and components will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation. One of the findings that the staff must make to issue a renewed license (10 CFR 54.29(a)) is that actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21, such that there is reasonable assurance that the activities authorized by the renewed license will continue to be conducted in accordance with the current licensing basis. In order to complete its review and enable making a finding under 10 CFR 54.29(a),

the staff requires additional information in regard to the matters described below.

Background:

Guidance in NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, (SRP-LR) Section 3.3.2.2.2, Cracking due to Stress Corrosion Cracking and Cyclic Loading, addresses cracking in stainless steel non-regenerative heat exchanger components. The SRP-LR guidance states that the GALL Report (NUREG-1801, Generic Aging Lessons Learned Report) recommends that a plant-specific aging management program (AMP) be evaluated to verify the absence of cracking and notes that an acceptable verification program includes radioactivity monitoring of the heat exchanger shell side water. Comanche Peak LRA Section 3.3.2.2.2 addresses this Further Evaluation item by crediting the Water Chemistry AMP and augmenting it with the One-Time Inspection AMP, which will verify the absence of cracking through the use of appropriate visual, surface, or volumetric nondestructive examination techniques. Consistent with the SRP-LR guidance, LRA Section 3.3.2.2.2 further states that absence of cracking of the letdown heat exchanger components is also verified by monitoring radiation levels in the Component Cooling Water (CCW) system, through the Closed Treated Water Systems AMP.

Issue:

Comanche Peak LRA Section B.2.3.12, Closed Treated Water Systems, states that the program will be consistent with corresponding program in NUREG-1801,Section XI.M21A.

The staff notes that the NUREG-1801,Section XI.M21A program does not specifically address radiation monitoring. Consequently, the Comanche Peak Closed Treated Water Systems AMP may be inconsistent with the NUREG-1801 AMP. The staff also notes that, although radioactivity is a monitored parameter in the recommended EPRI guidelines provided in XI.M21A, radioactivity is only classified as a diagnostic parameter with no

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.2.2-1 TXX-23048 and CP-202300322 Attachment M Page 2 of 3 associated limits (as opposed to a control parameter with specified ranges and action levels). In addition, the EPRI guidelines recommend a quarterly monitoring frequency for radioactivity. Although LRA Section 3.3.2.2.2 states that radiation levels in the CCW system are monitored through the Closed Treated Water System AMP to verify the absence of cracking of the letdown heat exchanger components, the associated AMP discussion in the LRA does not include any information about radiation level monitoring in the CCW system.

Request:

Provide information related to the radiation monitoring in the CCW system that is credited in LRA Section 3.3.2.2.2 (e.g., method, frequency (continuous/periodic), radioactivity limits, actions levels) for verifying the absence of cracking of the letdown heat exchanger components. Include a discussion whether the Closed Treated Water Systems AMP, which performs this activity, is consistent with guidance in NUREG-1801,Section XI.M21A. If monitoring is done on a periodic basis, provide a justification for the frequency relative to the maintaining intended function(s).

Luminant Response:

As stated in LRA Section 3.3.2.2.2, the CPNPP Water Chemistry AMP, augmented by the One-Time Inspection AMP, is credited with managing cracking of the stainless steel non-regenerative (letdown) heat exchanger components. In addition, radiation monitoring in the Component Cooling Water (CCW) system provides additional assurance that cracking has not occurred in the letdown heat exchanger tubes and tubesheet, and channel head. Radiation monitoring in the CCW system is performed under the Closed Treated Water Systems AMP.

Total gamma (Ci/mL) is periodically monitored as a diagnostic parameter with an upper alert limit of minimum detectable activity. Any activity above the minimum will cause an alert for investigation and evaluation.

Guidance in Element 4 of NUREG-1801 Rev. 2 Section XI.M21A states that the specific water chemistry parameters monitored and the acceptable ranges of values for closed-cycle cooling water systems as defined in NRC GL 89-13 are in accordance with EPRI 1007820. The CPNPP Closed Treated Water Systems AMP performs water testing in accordance with EPRI 3002000590, Closed Cooling Water Chemistry Guidelines, Revision 2. As discussed in LRA Section B.2.3.12, this is an exception to NUREG-1801 Rev. 2 guidance. However, EPRI 3002000590 is an updated revision to EPRI 1007820. Both EPRI 3002000590 and EPRI 1007820 include radiation as a diagnostic parameter for closed-cooling water systems with guidance to monitor quarterly and evaluate trends. Monitoring of radiation in the CCW system via the CPNPP Closed Treated Water Systems AMP is consistent with the guidance of EPRI 3002000590 and EPRI 1007820.

LRA Section 3.3.2.2.2 is revised as described below to include clarification regarding radiation monitoring in the CCW system.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.3.2.2.2-1 TXX-23048 and CP-202300322 Attachment M Page 3 of 3

References:

None.

Associated LRA Revisions:

LRA 3.3.2.2.2, (page 3.3-22), is revised as follows (text from LRA page 3.3-21 included for clarity):

3.3.2.2.2 Cracking due to Stress Corrosion Cracking and Cyclic Loading Cracking due to SCC and cyclic loading could occur in stainless steel PWR non-regenerative heat exchanger components exposed to treated borated water greater than 60°C (>140°F) in the chemical and volume control system.

The existing aging management program on monitoring and control of primary water chemistry in PWRs manages the aging effects of cracking due to SCC.

However, control of water chemistry does not preclude cracking due to SCC and cyclic loading. Therefore, the effectiveness of the water chemistry control program should be verified to ensure that cracking is not occurring. The GALL Report recommends that a plant-specific aging management program be evaluated to verify the absence of cracking due to SCC and cyclic loading to ensure that these aging effects are managed adequately. An acceptable verification program is to include temperature and radioactivity monitoring of the shell side water, and eddy current testing of tubes.

As summarized in item 3.3-1, 003, cracking due to SCC and cyclic loading in stainless steel PWR non-regenerative heat exchanger (letdown) components exposed to treated borated water greater than 140°F in the CVCS is an AERM.

The Water Chemistry AMP manages cracking of stainless steel non-regenerative heat exchanger components exposed to treated borated water and treated borated water greater than 140°F. The AMP is augmented by the One-Time Inspection (B.2.3.19) AMP which will verify the absence of cracking through the use of appropriate visual, surface, or volumetric NDE techniques. Absence ofAdditional assurance that cracking of the letdown heat exchanger tubes and tubesheet and channel head has not occurred is provided also verified by diagnostic monitoring of radiation levels in the CCW system, through the Closed Treated Water Systems (B.2.3.12) AMP, in accordance with EPRI Closed Cooling Water Chemistry Guidelines. Any level of radiation above the minimum detectable activity will cause an alert for investigation and evaluation. Temperature monitoring is a less sensitive technique and is not used.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 1 of 9 Additional Clarification to Fire Water Storage Tank Level Monitoring Affected LRA Sections: Table A-3 and B.2.3.27.

LRA Page Numbers: A-73 and B-161.

Description of Change:

CPNPP LRA Table A-3 (Commitment No. 29) and Section B.2.3.27 are revised to provide further consistency across the enhancements to the Monitoring and Trending, Acceptance Criteria, and Corrective Action elements of the Buried and Underground Piping and Tank AMP, regarding the use of the fire water storage tank level data as a reliable alternative to monitoring jockey pump activity or header pressure with respect to detecting and monitoring low-level leakage.

References:

None

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 2 of 9 LRA Table A-3 (Commitment No. 29), pages A-71 through A-75, is revised as follows:

Aging Management NUREG-1801 No. Program or Activity Commitment Implementation Schedule Section (Section) 29 Buried and XI.M41 Continue the existing Buried and Underground Piping and Tanks AMP, No later than 6 months prior to Underground Piping including enhancements to: the PEO, i.e.:

and Tanks (A.2.2.27) a) Manage loss of material due to corrosion of piping system bolting U1: 08/08/2029 within the scope of this program. U2: 08/02/2032, b) Implement the requirements of NACE SP0169-2007 or or no later than the last refueling NACE RP0285-2002 for cathodic protection. outage prior to the PEO.

c) Ensure pit depth gages or calipers used for measuring wall Perform the pre-PEO thickness have been demonstrated to be effective for the material, inspections within the 10-year environment, and conditions (e.g., remote methods) during the period prior to the PEO.

examination, and they are capable of quantifying general wall thickness and the depth of pits.

d) Perform inspections of buried and underground piping and tanks within the fire protection, SSW, and emergency diesel generator and auxiliary systems in accordance with LR-ISG-2015-01 Table XI.M41-2 for steel. The inspections will be distributed evenly among the units. Since CPNPP is a two-unit site, the inspection quantities are 50% greater than LR-ISG-2015-01 Table XI.M41-2 and are rounded up to the nearest whole inspection.

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.

e) Ensure a minimum of 25% of the internal surface of the diesel generator fuel oil storage tank, including the upper and lower

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 3 of 9 Aging Management NUREG-1801 No. Program or Activity Commitment Implementation Schedule Section (Section) portion of the tank and tank endbells, is inspected volumetrically.

f) With respect to cathodic protection, use an acceptance criterion equal to or more negative than -850 mV instant off for all in-scope buried components. Trend potential difference and current measurements to identify changes in the effectiveness of the cathodic protection system and/or coatings. Ensure the critical potential limit does not exceed -1200 mV.

g) Trend the main fire pump activity and, for small leaks, the fire water storage tank level indicator alarms and associated makeup from the treated water system (or similar parameter) to identify concerns with buried fire water yard loop header leakage.

h) Ensure type and extent of coating degradation is evaluated by evaluators who:

(a) possess a NACE Coating Inspector Program Level 2 or 3 inspector qualification; (b) who has completed the EPRI 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, Rev. 2, "Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants."

i) Where loss of material is identified, the measured wall thickness is projected to the end of the PEO such that minimum wall thickness requirements are maintained.

j) Revise acceptance criteria to ensure there is no evidence that backfill caused damage to the respective component coatings or the surface of the component (if not coated), and that changes in main fire pump activity or increasing frequency of fire water

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 4 of 9 Aging Management NUREG-1801 No. Program or Activity Commitment Implementation Schedule Section (Section) storage tank level indicator alarms (and associated makeup from the treated water system) (or similar parameter) that, which cannot be attributed to causes other than leakage from buried piping, are not occurring.

k) Conduct an extent of condition evaluation when damage to a coating has been evaluated as significant and the damage was caused by nonconforming backfill to determine the extent of degraded backfill in the vicinity of the observed damage.

l) Unacceptable cathodic protection survey results are entered into the plant corrective action program.

m) When using the option of monitoring the activity of a main fire pump or fire water storage tank level indicator alarms (and associated makeup from the treated water system) instead of inspecting buried fire water system piping, a flow test or system leak rate test is conducted by the end of the next refueling outage or as directed by the current licensing basis, whichever is shorter, when unexplained changes in main fire pump activity, fire water storage tank level indicator alarms, or equivalent equipment or parameter are observed.

n) If coated or uncoated metallic piping or tanks show evidence of corrosion, the remaining wall thickness in the affected area is determined to ensure that the minimum wall thickness is maintained. This may include different values for large area minimum wall thickness and local area wall thickness. If the wall thickness extrapolated to the end of the PEO meets minimum wall thickness requirements, recommendations for expansion of sample size, below do not apply.

o) Where the coatings, backfill, or the condition of exposed piping does not meet acceptance criteria, the degraded condition is repaired, or the affected component is replaced. In addition, where the depth or extent of degradation of the base metal could

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 5 of 9 Aging Management NUREG-1801 No. Program or Activity Commitment Implementation Schedule Section (Section) have resulted in a loss of pressure boundary function when the loss of material is extrapolated to the end of the PEO, an expansion of sample size is conducted. The number of inspections within the affected piping categories are doubled or increased by 5, whichever is smaller. If the acceptance criteria are not met in any of the expanded samples, an analysis shall be conducted to determine the extent of condition and extent of cause.

The timing of the additional examinations is based on the severity of the degradation identified and is commensurate with the consequences of a leak or loss of function. However, in all cases, the expanded sample inspection is completed within the 10-year interval in which the original inspection was conducted or, if identified in the latter half of the current 10-year interval, within 4 years after the end of the 10-year interval. These additional inspections conducted during the four years following the end of an inspection interval cannot also be credited towards the number of inspections in Table XI.M41-2 for the following 10 year interval.

The number of inspections may be limited by the extent of piping or tanks subject to the observed degradation mechanism.

The expansion of sample inspections may be halted in a piping system or portion of system that will be replaced within the 10-year interval in which the inspections were conducted or, if identified in the latter half of the current 10-year interval, within 4 years after the end of the 10-year interval.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 6 of 9 LRA Section B.2.3.27, pages B-160 through B-162, is revised as follows:

Element Affected Enhancement

1. Scope of Program Revise procedures to manage loss of material due to corrosion of piping system bolting within the scope of this program.

2. Preventive Actions Revise cathodic protection procedures to implement the requirements of NACE SP0169-2007 or NACE RP0285-2002.

3. Parameters Monitored and Ensure pit depth gages or calipers used for measuring wall Inspected thickness have been demonstrated to be effective for the material, environment, and conditions (e.g., remote methods) during the examination, and they are capable of quantifying general wall thickness and the depth of pits.

4. Detection of Aging Effects Revise procedures to state that inspections of buried and underground piping and tanks within the fire protection, SSW, and emergency diesel generator and auxiliary systems will be conducted in accordance with LR-ISG-2015-01 Table XI.M41-2 for steel. The inspections will be distributed evenly among the units.

Since CPNPP is a two-unit site, the inspection quantities are 50% greater than LR-ISG-2015-01 Table XI.M41-2 and are rounded up to the nearest whole inspection.

When the inspection 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.

Revise procedures to ensure a minimum of 25% of the internal surface of the diesel generator fuel oil storage tank, including the upper and lower portion of the tank and tank endbells, is inspected volumetrically.

5. Monitoring and Trending Revise cathodic protection procedures to trend potential difference and current measurements to identify changes in the effectiveness of the systems and/or coatings.

Revise procedures to trend the main fire pump activity and, for smaller leaks, fire water storage tank level indicator alarms and associated makeup from the treated water system (or similar parameter) to identify concerns with buried fire water yard loop header leakage.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 7 of 9 Element Affected Enhancement

6. Acceptance criteria Ensure type and extent of coating degradation is evaluated by evaluators who:

(a) possesses a NACE Coating Inspector Program Level 2 or 3 inspector qualification; (b) who has completed the Electric Power Research Institute (EPRI) 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, Rev. 2, "Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants."

Where loss of material is identified, the measured wall thickness is projected to the end of the period of extended operation such that minimum wall thickness requirements are maintained.

Revise acceptance criteria to ensure there is no evidence that backfill caused damage to the respective component coatings or the surface of the component (if not coated), and that changes in main fire pump activity or increasing frequency of fire water storage tank level indicator alarms (and associated makeup from the treated water system) (or similar parameter) that, which cannot be attributed to causes other than leakage from buried piping, are not occurring.

With respect to cathodic protection, use an acceptance criterion equal to or more negative than -850 mV instant off for all in-scope buried components. Ensure the critical potential limit does not exceed -1200 mV.

7. Corrective Actions Revise procedures to conduct an extent of condition evaluation when damage to a coating has been evaluated as significant and the damage was caused by nonconforming backfill to determine the extent of degraded backfill in the vicinity of the observed damage.

Revise procedures to state unacceptable cathodic protection survey results are entered into the plant corrective action program.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 8 of 9 Element Affected Enhancement Revise procedure to state when using the option of monitoring the activity of a main fire pump or fire water storage tank level indicator alarms (and associated makeup from the treated water system) instead of inspecting buried fire water system piping, a flow test or system leak rate test is conducted by the end of the next refueling outage or as directed by the current licensing basis, whichever is shorter, when unexplained changes in main fire pump activity, fire water storage tank level indicator alarms, (or equivalent equipment or parameter) are observed.

If coated or uncoated metallic piping or tanks show evidence of corrosion, the remaining wall thickness in the affected area is determined to ensure that the minimum wall thickness is maintained. This may include different values for large area minimum wall thickness and local area wall thickness. If the wall thickness extrapolated to the end of the period of extended operation meets minimum wall thickness requirements, recommendations for expansion of sample size, below do not apply.

Revise procedures to state where the coatings, backfill, or the condition of exposed piping does not meet acceptance criteria, the degraded condition is repaired, or the affected component is replaced. In addition, where the depth or extent of degradation of the base metal could have resulted in a loss of pressure boundary function when the loss of material is extrapolated to the end of the period of extended operation, an expansion of sample size is conducted. The number of inspections within the affected piping categories are doubled or increased by 5, whichever is smaller. If the acceptance criteria are not met in any of the expanded samples, an analysis shall be conducted to determine the extent of condition and extent of cause.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 TXX-23048 and CP-202300322 Attachment N Page 9 of 9 Element Affected Enhancement The timing of the additional examinations is based on the severity of the degradation identified and is commensurate with the consequences of a leak or loss of function. However, in all cases, the expanded sample inspection is completed within the 10-year interval in which the original inspection was conducted or, if identified in the latter half of the current 10-year interval, within 4 years after the end of the 10-year interval.

These additional inspections conducted during the four years following the end of an inspection interval cannot also be credited towards the number of inspections in Table XI.M41-2 for the following 10-year interval. The number of inspections may be limited by the extent of piping or tanks subject to the observed degradation mechanism.

The expansion of sample inspections may be halted in a piping system or portion of system that will be replaced within the 10-year interval in which the inspections were conducted or, if identified in the latter half of the current 10-year interval, within 4 years after the end of the 10-year interval.