CP-202300432, Response to Request for Additional Information Regarding the Safety Review of the License Renewal Application - Set 4

Response to Request for Additional Information Regarding the Safety Review of the License Renewal Application - Set 4

ML23277A176
Person / Time
Site:	Comanche Peak
Issue date:	10/04/2023
From:	John Lloyd Luminant, Vistra Operations Company
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
CP-202300432, TXX-23069
Download: ML23277A176 (1)

Text

Jay Lloyd Senior Director, Engineering

& Regulatory Affairs Comanche Peak Nuclear Power Plant (Vistra Operations Company LLC)

P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 T 254.897.5337 Ref 10 CFR 54 CP-202300432 TXX-23069 October 4, 2023 U. S. Nuclear Regulatory Commission 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 REQUEST FOR ADDITIONAL INFORMATION REGARDING THE SAFETY REVIEW OF THE LICENSE RENEWAL APPLICATION - SET 4

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 4," September 11, 2023 (ADAMS Accession Nos.

ML23256A149 and ML23256A150)

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. Vistra OpCos responses to these RAIs are provided in the Enclosure of this letter.

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

TXX-23069 Page 2 of 3 This communication contains no new commitments regarding CPNPP Units 1 and 2.

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 October 4, 2023

Enclosure:

Response to RAis c:

(email) w / o Attachments Sincerely, Jay Lloyd Mahesh Chawla, NRR, DORL [Mahesh.Chawla@mc.gov]

Dominic Antonangeli, RGN IV, CPNPP [Dominic.Antonangeli@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 Uessica.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-23069 Page 3 of 3 Enclosure Index Attachment CPNPP LRA RAI CPNPP LRA Request for Additional Information Topics A

B.2.3.16-4a Fire Water System AMP Enhancements B

B.2.3.15-3 Fire Barrier Aging Effects C

3.2.2.2.3.2-1a Stainless Steel in Outdoor Air

Enclosure:

Responses to RAIs 19 pages follow

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4a TXX-23069 and CP-202300432 Attachment A Page 1 of 4 LRA Section: B.2.3.16, Fire Water System NRC RAI No: B.2.3.16-4a, Fire Water System AMP Enhancements 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 as described in the requests for additional information.

Background:

By letter dated July 27, 2023, in response (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23208A193) to Request for Additional Information (RAI) B.2.3.16-4, the applicant revised the table entitled Fire Water System Inspections and Tests in Section B.2.3.16 of Appendix B in the License Renewal Application (LRA) that provides additional detail on the required enhancements based on Table 4a, Fire Water System Inspection and Testing Recommendations, in Appendix L, Revised GALL Report AMP XI.M27 Fire Water System, of License Renewal Interim Staff Guidance, LR-ISG-2012-02, Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion Under Insulation (ML13227A361). Specifically, the table in Section B.2.3.16 of LRA Appendix B that provides additional detail on the required enhancements based on Table 4a in Appendix L of LR-ISG-2012-02 (Operational Tests, NFPA 25 Section 10.3.4.3) was revised to clarify which deluge system operational tests are performed with water, to clarify which deluge systems are tested with air instead of water, and to add an enhancement related to monitoring and trending the results of the deluge system operational tests performed with water (i.e., pump performance, run and discharge time, pressure, deposits or sediment).

Section B.2.3.16 in LRA Appendix B includes a table that provides additional detail on the required enhancements based on Table 4a in Appendix L of LR-ISG-2012-02.

However, the associated enhancement to the Detection of Aging Effects, Monitoring

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4a TXX-23069 and CP-202300432 Attachment A Page 2 of 4 and Trending, and Acceptance Criteria program elements in Table A-3, List of LR Commitments and Implementation Schedule, in LRA Appendix A (No. 18) does not refer to the table for additional detail on the required enhancements.

It is unclear why some enhancements are identified as an enhancement to a particular program element in Table A-3 in LRA Appendix A and Section B.2.3.16 in LRA Appendix B while other required enhancements are identified in the table that provides additional detail on the required enhancements based on Table 4a in Appendix L of LR-ISG-2012-02 in Section B.2.3.16 in LRA Appendix B.

Issue:

Because the table in Section B.2.3.16 of LRA Appendix B that provides additional detail on the required enhancements based on Table 4a in Appendix L of LR-ISG-2012-02 is not referenced in the associated enhancement to the Detection of Aging Effects, Monitoring and Trending, and Acceptance Criteria program elements in Table A-3 in LRA Appendix A, it may be hard to verify implementation of all of the required enhancements.

Request:

Please discuss why some enhancements are identified as an enhancement to a particular program element in Table A-3 in LRA Appendix A and Section B.2.3.16 in LRA Appendix B while other required enhancements are identified in the table that provides additional detail on the required enhancements based on Table 4a in Appendix L of LR-ISG-2012-02 in Section B.2.3.16 in LRA Appendix B. Alternatively, revise the associated enhancement to the Detection of Aging Effects, Monitoring and Trending, and Acceptance Criteria program elements in Table A-3 in LRA Appendix A to clearly reference the table that provides additional detail on the required enhancements.

Luminant Response:

LRA Table A-3, Commitment #18 discusses the enhancements that will be implemented for the existing CPNPP Fire Water System AMP. Part d) of Commitment #18 includes an enhancement to Perform testing and visual inspections in accordance with Table 4a of LR-ISG-2012-02 Appendix L. LR-ISG-2012-02 Appendix L, Revised GALL Report AMP XI.M27 Fire Water System Element 4 states, Testing and visual inspections are performed in accordance with Table 4a, Fire Water System Inspection and Testing Recommendations. Therefore, Commitment #18 part d) meets the guidance of LR-ISG-2012-02 Appendix L.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4a TXX-23069 and CP-202300432 Attachment A Page 3 of 4 Table 4a of LR-ISG-2012-02 Appendix L provides a cross reference to specific sections in NFPA 25 for different categories of recommended inspection and testing.

The table included in LRA Section B.2.3.16, Fire Water System Inspections and Tests includes an evaluation of the existing CPNPP Fire Water Systems AMP as compared to the sections in NFPA 25 and identifies specific actions recommended for the CPNPP Fire Water Systems AMP to meet the NFPA 25 guidance. This table is intended to be an aid for implementation of Commitment #18 part d).

LRA Table A-3, Commitment #18 part d) will be revised as shown in this RAI response for clarity.

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 (ADAMs Accession No. 13227A361)

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.16-4a TXX-23069 and CP-202300432 Attachment A Page 4 of 4 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 No.

Aging Management Program or Activity (Section)

NUREG-1801 Section Commitment Implementation Schedule 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) Update existing procedures and/or develop new procedures, as directed in the table titled, Fire Water System Inspections and Tests in LRA Section B.2.3.16, to state that Perform testing and visual inspections are performed in accordance with Table 4a of LR-ISG-2012-02 Appendix L. Theseis tables areis based on NFPA 25, 2011 edition. Unless recommended otherwise, external visual inspections are to be conducted on a refueling outage interval.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 1 of 8 LRA Section: B.2.3.15, Fire Protection NRC RAI No: RAI B.2.3.15-3 (Fire Barrier Aging Effects)

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 described in the requests for additional information.

Background:

As amended by letter dated April 24, 2023 (ML23114A377), License Renewal Application (LRA) Table 3.5.2-15 states that loss of material of ceramic fiber/blanket insulation and wrap exposed to indoor uncontrolled air and cracking, loss of bond, and loss of material of gypsum walls, floors, and ceilings exposed to indoor uncontrolled air are managed by the Fire Protection program. The items cite plant-specific notes 2 and 4, which state, This material is not addressed for fire barriers in NUREG-1801. Consistent with the OE [operating experience] 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, and 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, respectively.

Subsequent License Renewal (SLR) Interim Staff Guidance (ISG), SLR-ISG-2021 Mechanical, Updated Aging Management Criteria for Mechanical Portions of Subsequent License Renewal Guidance (ML20181A434), added Aging Management Review Items VII.G.A-805, VII.G.A-806, and VII.G.A-807 to Table VII.G in Volume 1 of NUREG-2191, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report (ML17187A031), and Table 3.3-1 in NUREG-2192, Standard Review Plan [SRP] for Review of Subsequent License Renewal Applications for Nuclear Power Plants (ML17188A158).

The aging effects for cementitious coatings, silicates, and subliming compounds used as fireproofing/fire barriers exposed to air are loss of material, change in material properties, cracking, delamination, and separation. These aging effects are consistent with Section 6, Fire Barriers, of EPRI 3002013084, Long-Term Operations: Subsequent License Renewal Aging Affects for Structures and Structural Components (Structural Tools), November 2018.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 2 of 8 Issue:

While LRA Table 3.5.2-15 states that ceramic fiber/blanket insulation and wrap is consistent with the OE 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), it does not cite change in material properties, cracking, delamination, and separation as applicable aging effects. It is unclear why these aging effects were not cited as applicable for ceramic fiber/blanket insulation and wrap since LRA Table 3.5.2015 states it is consistent with the OE in SLR-ISG-2021-02-MECHANICAL and given that these materials are similar to silicate fireproofing/fire barriers.

LRA Table 3.5.2-15 does not cite change in material properties, delamination, and separation as applicable aging effects for gypsum walls, floors, and ceilings; and does cite loss of bond as an applicable aging effect. It is unclear why change in material properties, delamination, and separation were not cited as applicable aging effects for gypsum walls, floors, and ceilings given that gypsum is similar to silicate fireproofing/fire barriers.

Request:

1. Please discuss why change in material properties, cracking, delamination, and separation were not cited as applicable aging effects for the ceramic fiber/blanket insulation and wrap. Alternatively, revise LRA Table 3.5.2-15 to cite these aging effects for the ceramic fiber/blanket insulation and wrap. If change in material properties, cracking, delamination, and separation are added as applicable aging effects for the ceramic fiber/blanket insulation and wrap, please discuss whether plant-specific procedures require updating to address these aging effects.

2. Please discuss why change in material properties, delamination, and separation were not cited as applicable aging effects for the gypsum walls, floors, and ceilings.

Alternatively, revise LRA Table 3.5.2-15 to cite these aging effects for the gypsum walls, floors, and ceilings. If change in material properties, delamination, and separation are added as applicable aging effects for the gypsum walls, floors, and ceilings, please discuss whether plant-specific procedures require updating to address these aging effects. The staff notes, that if LRA Table 3.5.2-15 is revised to cite these aging effects, then plant-specific note 2 may be applicable to the gypsum walls, floors, and ceilings.

Luminant Response:

NUREG-1801, Revision 2, includes aging management for walls, floors and ceilings that perform a fire barrier function and specifies the Fire Protection AMP for management of the identified aging effects. The NUREG-1800, Table 3.0-1, XI.26 Fire Protection AMP description includes inspections for managing fire barriers and states The fire barrier inspection program requires periodic visual inspection of fire barrier penetration seals, fire barrier walls, ceilings, and floors,... In order to address the operating experience reflected in SLR-ISG-2021 Mechanical (ML20181A434), which is considered as relevant operating experience during the period of extended operation, the aging effects for both cementitious fireproofing/fire barriers in SLR-ISG item VII.G.A-806 (i.e. gypsum walls) and ceramic fiber/blanket insulation and wrap

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 3 of 8 in SLR-ISG item VII.G.A-807 is warranted.

The specific aging effects for both components are listed as: Loss of material; cracking/delamination; change in material properties; and separation. These aging effects are managed by the Fire Protection AMP XI.M26.

As described in LRA Table 3.5.2-15, plant-specific notes 2 and 4, the Fire Protection (B.2.3.15) AMP manages aging of the component materials (ceramic fiber/blanket and gypsum, respectively). 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.

LRA Table 3.5.2-15 is revised, in order to further clarify the aging effects of change in material properties, cracking/delamination, and separation are also applicable for ceramic fiber/blanket insulation and wrap and gypsum walls, floors, and ceilings that perform a fire barrier function and are managed by the Fire Protection AMP. These aging effects are 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). Additionally, 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 the specific aging effects identified in LRA Table 3.5.2-15 in the inspection requirements.

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.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 4 of 8 Associated LRA Revisions:

LRA Table 3.5.2-15, page 3.5-187, is revised as follows:

Table 3.5.2-15: Fire Barrier Commodity Group - Summary of Aging Management Evaluation Hatch Fire barrier Concrete Air - outdoor Loss of material Structures Monitoring (B.2.3.34)

VII.G.A-93 3.3-1, 062 A

Hatch Fire barrier Carbon steel Air with borated water leakage Loss of material Boric Acid Corrosion (B.2.3.4)

III.B2.T-25 3.5-1, 089 C

Insulation and wrap Fire barrier Silicate radiant energy shield (Kaowool, Cerafiber, Cera

blanket, Siltemp WR 84 CSR, Avsil 84 CSR fabric or similar)

Air - indoor uncontrolled Change in material properties;

Cracking, Delamination; Loss of material, and Separation Fire Protection (B.2.3.15)

None None F, 2 Insulation and wrap Fire barrier Subliming compound (Thermo-Lag, Carboline, or similar)

Air - indoor uncontrolled Change in material properties;

Cracking, Delamination; Loss of material, and Separation Fire Protection (B.2.3.15)

None None F, 2 Insulation and wrap Fire barrier Ceramic fiber/blanket Air - indoor uncontrolled Change in material properties;

Cracking, Delamination; Loss of material, and Separation Fire Protection (B.2.3.15)

None None F, 2 Insulation and wrap Fire barrier Stainless steel Air - indoor uncontrolled None None III.B5.TP-8 3.5-1, 095 C

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 5 of 8 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 Function Material Environment Aging Effect Requiring Management Aging Management Program NUREG-1801 Item Table 1 Item Notes Wall, floor, and ceiling Fire barrier Masonry block Air - indoor uncontrolled Cracking Fire Protection (B.2.3.15)

III.A3.T-12 3.5-1, 070 E, 3 Wall, floor, and ceiling Fire barrier Concrete (reinforced)

Air - indoor uncontrolled Cracking Structures Monitoring (B.2.3.34)

VII.G.A-90 3.3-1, 060 A

Wall, floor, and ceiling Fire barrier Gypsum Air - indoor uncontrolled Change in material properties;

Cracking, Delamination; lossof bond; and lLoss of material, and Separation Fire Protection (B.2.3.15)

None None F, 4 Wall, floor, and ceiling Fire barrier Carbon steel Air - indoor uncontrolled Loss of material Fire Protection (B.2.3.15)

VII.G.A-21 3.3-1, 059 C

Wall, floor, and ceiling Fire barrier Concrete (reinforced)

Air - indoor uncontrolled Loss of material Fire Protection (B.2.3.15)

VII.G.A-91 3.3-1, 062 A

Wall, floor, and ceiling Fire barrier Concrete (reinforced)

Air - indoor uncontrolled Loss of material Structures Monitoring (B.2.3.34)

VII.G.A-91 3.3-1, 062 A

Wall, floor, and ceiling Fire barrier Masonry block Air - outdoor Cracking Fire Protection (B.2.3.15)

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

III.A3.T-12 3.5-1, 070 C, 3 Wall, floor, and ceiling Fire barrier Concrete (reinforced)

Air - outdoor Cracking; Loss of material Fire Protection (B.2.3.15)

VII.G.A-92 3.3-1, 061 A

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 6 of 8 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 and VII.G.A-807 (subliming compounds, silicates); SRP items 3.3-1, 267 to and 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 (meeting ASTM C-36) 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, consistent with the OE reflected in SLR-ISG-2021-02-MECHANICAL (item VII.G.A-806 (cementitious materials); SRP item 3.3-1, 268), 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.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 7 of 8 LRA Appendix A, Table A-3, page A-60, is revised as follows:

Table A-3 List of LR Commitments and Implementation Schedule No.

Aging Management Program or Activity (Section)

NUREG-1801 Section Commitment Implementation Schedule 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 (A.2.2.15)

XI.M26 Continue the existing Fire Protection AMP, including enhancements to:

a) Expand the sample size of inspected fire penetration seals if any sign of degradation is found in the sample.

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

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.

e) Revise Fire Rated Assembly Visual Inspection procedure to include the following signs of degradation or damage:

Change in material properties; Cracking, Delamination; Loss of material, and Separation with the current list as a part of Section 8.4.3.

No later than 6 months prior to the PEO, i.e.:

U1: 08/08/2029 U2: 08/02/2032, or no later than the last refueling outage prior to the PEO.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. B.2.3.15-3 TXX-23069 and CP-202300432 Attachment B Page 8 of 8 LRA Appendix B, Section B.2.3.15, page B-104, is revised as follows:

Element Affected Enhancement

4. Detection of Aging Effects Expand the sample size of inspected fire penetration seals if any 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.

Revise Fire Rated Assembly Visual Inspection procedure to include the following signs of degradation or damage:

Change in material properties; Cracking, Delamination; Loss of material, and Separation with the current list as a part of Section 8.4.3.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1a TXX-23069 and CP-202300432 Attachment C Page 1 of 7 LRA Section: 3.2.2.2.3, Loss of Material due to Pitting and Crevice Corrosion NRC RAI No: 3.2.2.2.3.2-1a (Stainless Steel in 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:

By letter dated July 27, 2023 (ML23208A193), Attachment E, in response to Request for Additional Information 3.2.2.2.3.2-1, the applicant clarified its aging management approach and revised the License Renewal Application (LRA) regarding loss of material (LOM) and stress corrosion cracking (SCC) of stainless steel components exposed to outdoor air in the Engineered Safety Features, Auxiliary, and Steam and Power Conversion systems. The associated LRA sections are 3.2.2.2.3.2, 3.2.2.2.6, 3.3.2.2.3, 3.3.2.2.5, 3.4.2.2.2, and 3.4.2.2.3.

Issue:

The RAI response provided a discussion of managing LOM (due to pitting or crevice corrosion) and SCC of stainless steel in outdoor air considering differences between the guidance for initial license renewal (Reference 1) and subsequent license renewal (Reference 2). As part of the response, the applicant revised LRA Tables 3.3.2-4, 3.4-1, and 3.4.2-1. The response also discussed the basis for not identifying SCC as an applicable aging effect for stainless steel exposed to outdoor air. It is not clear to the staff why some of the LRA revisions are for carbon steel components. In addition, the criteria applied to stainless steel components and the revisions to the LRA for stainless steel components are unclear to the staff. Finally, the staff determined that the basis for not identifying SCC as an applicable aging effect may not take into account the operating experience used to develop SLR guidance and requires additional technical justification.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1a TXX-23069 and CP-202300432 Attachment C Page 2 of 7 Request:

1. Changes to the LRA in the RAI response include the deletion of carbon steel components from Tables 3.3.2-4 (Demineralized and Reactor Makeup Water System) and 3.4.2-1 (Auxiliary Feedwater System). Please provide the basis for deleting these items from the tables.

2. The first paragraph of the response to Question 1 states that components located within piping tunnels were removed from consideration for pitting and crevice corrosion. Please clarify the aging management treatment for the components in piping tunnels. The staff notes that the GALL-SLR Report (Reference 3),Section XI.M36 states that for underground piping below grade but within a tunnel, aging effects can be managed by the External Surfaces Monitoring of Mechanical Components AMP or the Buried and Underground Piping and Tanks AMP, depending on accessibility.

3. If stainless steel components shown deleted from Tables 3.3.2-4 and 3.4.2-1 were deleted for reasons other than being located in piping tunnels, please provide the justification.

4. Part a of the response to Question 1 notes that aging effects are not considered for stainless steel flow elements and orifices exposed to outdoor air in the Fire Protection System. These components are associated with Note I and Footnote 5 in Table 3.3.2-

7. The footnote refers to LRA Section 3.3.2.2.6 for the justification. However, the staff did not find a justification for this topic in Section 3.3.2.2.6, Quality Assurance for Aging Management of Nonsafety-Related Components, or in Sections 3.3.2.2.3 or 3.3.2.2.5 related to LOM and SCC for stainless steel. If another section was intended, please identify the section and provide the justification.

5. The response to Question 2 provides a basis for not considering SCC, including cases where LOM due to pitting or crevice corrosion is considered. The response cites a lack of ambient environmental halides, lack of OE identifying past instances of LOM or SCC for stainless steel in outdoor air at CPNPP, and absence of elevated temperature for the subject components. However, operating experience includes examples of chloride-induced SCC of austenitic stainless steel occurring at about 100 degrees Fahrenheit or less (References 4 and 5). In addition, in the SLR guidance, the ambient contaminant level and lack of operating experience are no longer considered sufficient for determining the susceptibility because of the potential for unexpected sources of contaminants and concentrating effects. Elevated temperature is not used as a criterion in NUREG-1800 Sections 3.2.2.2.6, 3.3.2.2.3, 3.4.2.2.2 (Reference 1), or corresponding SRP-SLR Sections 3.2.2.2.4, 3.3.2.2.3, 3.4.2.2.2 (Reference 2).

Please clarify whether and how the operating experience in References 4 and 5 was considered. Specifically, the staff requests additional justification for not applying aging management for SCC comparable to that for LOM due to pitting or crevice corrosion (i.e., External Surfaces Monitoring of Mechanical Components AMP).

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1a TXX-23069 and CP-202300432 Attachment C Page 3 of 7

6. Please clarify part f of the response to Question 1, which states that the recommended one-time inspection in SLR guidance was found to be not applicable for initial license renewal. The staff requests clarification because part a of the response identifies specific components for which LOM was not identified as an applicable aging effect, and part b of the response describes components that are addressed with the External Surfaces Monitoring of Mechanical Components AMP. Please clarify whether part f of the response refers to stainless steel components other than those noted in part a for which LOM due to pitting or crevice corrosion was not identified as an applicable aging effect.

Luminant Response:

1. The subject carbon steel components discussed in part (1) of the above request were removed as part of the response to RAI 3.2.2.2.3-1 (ML23208A193), as further research determined these components are not actually exposed to an outdoor air environment.

Instead, the subject carbon steel components removed from LRA Tables 3.3.2-4 and 3.4.2-1 are located indoors within pipe tunnels associated with the Reactor Makeup Water Storage Tank (RMWST) and the Condensate Storage Tank (CST), respectively. These locations are accessed through the basement of the radiological controlled area (RCA) and have designated room numbers (SG-1-085D and SG-2-085D) as accessible portions of the Unit 1 and 2 Safeguards Buildings exposed to indoor air (i.e., an air - indoor uncontrolled environment). Therefore, the correct environment for the subject carbon steel components in LRA Tables 3.3.2-4 and 3.4.2-1 is already covered by existing line items in these tables.

2. The subject stainless steel components discussed in part (2) of the request are contained in the same pipe tunnels discussed in the response to part (1). As stated previously, these pipe tunnels are accessed through the basement of the RCA and are given room numbers (SG-1-085D and SG-2-085D) as accessible portions of the Unit 1 and 2 Safeguards Buildings. The environment inside of these pipe tunnels is indoor air from the Safeguards Buildings (i.e., an air - indoor uncontrolled environment); therefore, this environment is not equivalent to the underground environment, as described in Table IX.D of NUREG-2191. Therefore, because stainless steel components located within these pipe tunnels are exposed to an air - indoor uncontrolled environment, they are not expected to produce any aging effects, consistent with NUREG-1800 items 3.1-1, 107, 3.2-1, 063, 3.3-1, 120, and 3.4-1, 058.

3. See response to part (2) above. Stainless steel components were not removed from LRA Tables 3.3.2-4 and 3.4.2-1 for a reason other than being located indoors within accessible pipe tunnels (i.e., an air - indoor uncontrolled environment).

4. Upon further examination, pursuant to part (4) of the request, all stainless steel flow elements and orifices within the Fire Protection System were found to be located within buildings, in conditioned spaces; therefore, these components are subject to an air -

indoor uncontrolled external environment. Stainless steel components located within an air - indoor uncontrolled environment are not expected to produce any aging effects, consistent with NUREG-1800 item 3.3-1, 120.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1a TXX-23069 and CP-202300432 Attachment C Page 4 of 7 As a result, a revision to LRA Table 3.3.2-7 has been provided below to remove line items for stainless steel flow elements and orifices exposed to outdoor air. Similarly, plant-specific note 5 for this table has also been removed, as it is no longer applicable given the corrected plant configuration. The external air - indoor uncontrolled environment for the subject components is already captured by existing items in LRA Table 3.3.2-7.

5. For the further evaluation of potential aging effects for stainless steel components exposed to outdoor air at CPNPP, the operating experience contained in References 4 and 5 was considered. These industry operating experience items were previously evaluated within the CPNPP corrective action program (CAP) at their respective time of issuance. These evaluations are summarized below.

Reference 4, which details chloride-induced stress corrosion cracking (SCC) events at four sites, was dispositioned within the CPNPP CAP in 2012. The vulnerabilities discussed in Reference 4 were determined to be non-applicable to CPNPP for the following reasons:

a. CPNPP is not located on saltwater. The waters of Squaw Creek Reservoir are brackish but salt films and corrosion from salt is not being observed on site. As a result, SCC from proximity to a saltwater atmosphere is not considered viable.

b. CPNPP uses sand, not salt, on the roads if ice and snow are observed. Typically, an ice or snow event occurs one or two times during the winter. Based upon this the potential for SCC from road salts is not considered viable.

c. At present time, there are no cooling towers in the CPNPP area so the introduction of chlorides from condensed cooling tower waters and subsequent SCC is not considered viable.

d. CPNPP is situated in a portion of Texas where the humidity is low (<50%) most of the year. As a result, there is a minimum amount of condensation on equipment due to environmental conditions.

e. SCC of stainless steel in outdoor air has been reported by the industry and previously evaluated by CPNPP within the CAP. There have been no cases of SCC of stainless steel in outdoor air reported at CPNPP to date (this remains true in the period following the CPNPP evaluation of Reference 4 in 2012, through present day).

Reference 5, which details SCC of stainless steel standby liquid control tanks at another site caused by halogen-containing grout used in external supports positioned against the stainless steel tanks, was dispositioned within the CPNPP CAP in 2006. This industry event was dispositioned as outside of the screening threshold for detailed evaluation, given that the plant configuration described in Reference 5 does not exist at CPNPP. The subject tanks at CPNPP utilize a stainless steel liner which is fully encased in concrete, running from the top of the tank down to a concrete foundation; therefore, there is no potential for leaching into the stainless steel liners from halogen-containing grouts.

The uninsulated stainless steel components at CPNPP which are managed for loss of material are portions of the CST, RWST, and RMWST tank liners and associated piping (internal to the tanks) which are above the waterline of the tank. These tanks are vented to

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1a TXX-23069 and CP-202300432 Attachment C Page 5 of 7 the atmosphere via gooseneck shaped vent pipes that are external to the tanks, which effectively precludes the entrance of rainwater or other moisture into the tanks. Because the area inside these tanks above the waterline are in a confined space which draws in replacement air from the atmosphere, the potential for particulate accumulation in these areas was conservatively assumed to be higher. Therefore, loss of material for the stainless steel components in these areas is conservatively managed using the External Surfaces Monitoring of Mechanical Components AMP.

However, SCC of the stainless steel components located in the air space of the above mentioned tanks is not managed for the following reasons: (1) site operating experience and environmental conditions, as discussed previously, do not indicate any potential for or previous history of chloride-induced SCC at CPNPP, and (2) the subject stainless steel components are not exposed to high operational loads (i.e., are open to atmosphere, with no system pressure), often required to initiate SCC.

Additionally, the generic External Surfaces Monitoring of Mechanical Components AMP manages cracking by performing visual inspections for leakage; however, the subject stainless steel components are exposed to either air or concrete internally and externally, and are not normally flowing, such that visually-observable leakage is not possible.

Therefore, an enhancement to the AMP would be required to perform either a volumetric or enhanced visual examination capable of detecting cracking to inspect for SCC of the subject components. Given that the subject components all reside within confined spaces, involve significant ALARA and personnel contamination concerns (for the Reactor Water Storage Tank), and therefore present elevated risk to plant personnel, the need to perform such an inspection is excessive and overly conservative, considering the lack of site operating experience of SCC of stainless steel in outdoor air and the configuration of the subject components.

6. The one-time inspection recommended in SLR guidance (NUREG-2191 and NUREG-2192) for detection of potential SCC in stainless steel components exposed to outdoor air is inherently performed within the 50-60 year plant life mark (i.e., the typical pre-SPEO implementation period for SLR). Given that CPNPP is a first LR applicant, a commitment to perform a one-time inspection for cracking in the pre-PEO implementation period for LR (i.e., at 30-40 years of plant operation) would be premature, especially considering the lack of applicable site operating experience, as previously discussed.

Additionally, part (a) of the previous response to RAI 3.2.2.2.3-1 (ML23208A193) has been superseded by part (4) of this response, and the subsequent revision to LRA Table 3.3.2-7, as shown below. Following this revision, all the in-scope stainless steel components exposed to outdoor air are managed for loss of material.

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1a TXX-23069 and CP-202300432 Attachment C Page 6 of 7

References:

1. NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, December 2010, ML103490036.

2. NUREG-2192, Standard Review Plan for Review of Subsequent License Renewal Applications for Nuclear Power Plants, July 2017, ML17188A158

3. NUREG-2191, Vol. 2, Generic Aging Lessons Learned for subsequent License Renewal (GALL-SLR) Report, July 2017, ML17187A294

4. Information Notice 2012-20, Potential Chloride-Induced Stress Corrosion Cracking of Austenitic Stainless Steel and Maintenance of Dry Cask Storage System Canisters, 11/14/2012, ML12319A440

5. Licensee Event Report 254-2006-004, Through-wall Leak in Standby Liquid Control Tank Due to the Original Construction Use of Grout with Leachable Halogens, 12/11/2006, ML063530355

Comanche Peak Nuclear Power Plant Units 1 and 2 Dockets 50-445 and 50-446 Luminant Response to NRC RAI No. 3.2.2.2.3.2-1a TXX-23069 and CP-202300432 Attachment C Page 7 of 7 Associated LRA Revisions:

LRA Table 3.3.2-7 (page 3.3-197), is revised as follows:

Table 3.3.2-7: Fire Protection - Summary of Aging Management Evaluation Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs NUREG-1801 Item Table 1 Item Notes Flow element Pressure boundary Stainless steel Air - outdoor (external)

None None VII.J.AP-123 3.3-1, 120 I, 5 LRA Table 3.3.2-7 (page 3.3-198), is revised as follows:

Table 3.3.2-7: Fire Protection - Summary of Aging Management Evaluation Component Type Intended Function Material Environment Aging Effect Requiring Management Aging Management Programs NUREG-1801 Item Table 1 Item Notes Orifice Pressure boundary Stainless steel Air - outdoor (external)

None None VII.J.AP-123 3.3-1, 120 I, 5 Orifice Throttle Stainless steel Air - outdoor (external)

None None VII.J.AP-123 3.3-1, 120 I, 5 LRA Table 3.3.2-7, Plant-Specific Note #5 (page 3.3-210), is revised as follows:

Plant-Specific Notes

5. Stainless steel components in an Air - outdoor (external) environment do not experience aging effects as addressed in LRA Further Evaluation Section 3.3.2.2.6. Not used.