CNL-25-025, Application for Subsequent Renewed Operating Licenses, Response to Proprietary Attached - Request for Confirmation of Information, Set 1
| ML25043A270 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/12/2025 |
| From: | Hulvey K Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| CNL-25-025, EPID L-2024-SLE-0000 | |
| Download: ML25043A270 (1) | |
Text
1101 Market Street, Chattanooga, Tennessee 37402 CNL-25-025 February 12, 2025 10 CFR 54 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296
Subject:
Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Application for Subsequent Renewed Operating Licenses, Response to Proprietary Attached - Request for Confirmation of Information, Set #1 (EPID: L-2024-SLE-0000)
Reference:
- 1. Letter from TVA to NRC, CNL-24-001, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Application for Subsequent Renewed Operating Licenses," dated January 19, 2024 (ML24019A010)
Request for Confirmation of Information - Set #1, dated January 14, 2025 (ML25014A041, ML25014A042, ML25014A043, and ML25014A044)
By Reference 1, the Tennessee Valley Authority (TVA) submitted a subsequent license renewal application (SLRA) for the Browns Ferry Nuclear Plant, Units 1, 2, and 3, Renewed Facility Operating Licenses in accordance with Title 10 of the Code of Federal Regulations (10 CFR),
Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants.
Since March 2024, TVA has been engaged with the Nuclear Regulatory Commission (NRC) staff in the safety audit of the SLRA. This audit has resulted in a request for confirmation of information (RCI) (Reference 2). The enclosure to this letter contains the RCI responses.
There are no new regulatory commitments in this letter. Should you have any questions regarding this submittal, please contact Peter J. Donahue, Director, Subsequent License Renewal, at pjdonahue@tva.gov.
U.S. Nuclear Regulatory Commission CNL-25-025 Page 2 February 12, 2025 I declare under penalty of perjury that the foregoing is true and correct. Executed on this 12th day of February 2025.
Respectfully, Kimberly D. Hulvey General Manager, Nuclear Regulatory Affairs and Emergency Preparedness
Enclosure:
cc:
Response to Proprietary Attached - Browns Ferry SLRA - Request for Confirmation of Information - Set #1 NRC Regional Administrator - Region II NRC Branch Chief - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager, License Renewal Projects Branch (Safety)
State Health Officer, Alabama Department of Public Health (w/o Enclosure)
Digitally signed by Edmondson, Carla Date: 2025.02.12 11:14:49 -05'00'
Enclosure CNL-25-025 E 1 of 5 Response to Proprietary Attached - Browns Ferry SLRA - Request for Confirmation of Information - Set 1 The Nuclear Regulatory Commission (NRC) Request for Confirmation of Information (RCI) is provided in italicized font. The Tennessee Valley Authority (TVA) response is provided in unitalicized font.
Regulatory Basis:
Part 54 of Title 10 of the Code of Federal Regulations (10 CFR), Requirements for Renewal of Operating Licenses for Nuclear Power Plants, is designed to elicit application information that will enable the U.S. Nuclear Regulatory Commission (NRC, staff) to perform an adequate safety review and the Commission to make the necessary findings. Reliability of application information is important and advanced by requirements that license applications be submitted in writing under oath or affirmation and that information provided to the NRC by a license renewal applicant or required to be maintained by NRC regulations be complete and accurate in all material respects. Information that must be submitted in writing under oath or affirmation includes the technical information required under 10 CFR 54.21(a) related to assessment of the aging effects on structures, systems, and components subject to an aging management review.
Thus, both the general submission requirements for license renewal applications and the specific technical application information requirements require that submission of information material to NRCs safety findings (see 10 CFR 54.29 standards for issuance of a renewed license) be submitted by an applicant as part of the application.
Background:
By Subsequent License Renewal Application Appendix E, Applicant’S Environmental Report-Operating License Renewal Stage|letter dated January 19, 2024]] (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML24019A009), as supplemented by letters dated January 22, 2024 (ML24022A292), October 9, 2024 (ML24283A091), November 1, 2024 (ML24306A203), December 17, 2024 (ML24352A216) and January 8, 2025 (ML25008A150),
Tennessee Valley Authority (the applicant) submitted an application for subsequent license renewal of Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant (BFN, Browns Ferry), Units 1, 2, and 3, respectively, to the NRC. TVA submitted the application pursuant to 10 CFR Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, for subsequent license renewal.
The NRC staff is in the process of reviewing the application. Based on our review, the NRC staff has identified the attached request for confirmation of information (RCI).
Request:
During the audit, the staff reviewed several documents that contain information which will likely be used in conclusions documented in the Safety Evaluation (SE). To the best of the staff's knowledge, this information is not on the docket. Any information used to reach a conclusion in
Enclosure CNL-25-025 E 2 of 5 the SE must be included on the docket by the applicant. We request that you submit confirmation that the information gathered from the documents and listed below is correct or provide the associated corrected information.
RCI-SFNB-3.5.2.2.2.6-1 The proprietary request for RCl-SFNB-3.5.2.2.2.6-1 is contained in the email from Jessica Hammock (NRC) to Delson C. Erb (TVA) dated January 14, 2025 (ML25014A041, ML25014A042, ML25014A043, and ML25014A044).
TVA Response TVA confirms the NRC staffs understanding without the need of clarification.
RCI-SFNB-3.5.2.2.2.6-2 The proprietary request for RCl-SFNB-3.5.2.2.2.6-2 is contained in the email from Jessica Hammock (NRC) to Delson C. Erb (TVA) dated January 14, 2025 (ML25014A041, ML25014A042, ML25014A043, and ML25014A044).
TVA Response TVA confirms the NRC staffs understanding and provides the following additional information:
The benchmark results (the percent differences between calculated and computed values) documented in Section 4.0 of the report do not provide an uncertainty term for the values computed using the radiation transport methodology described in Section 3.1 of the report.
RCI-SFNB-3.5.2.2.2.6-3 The proprietary request for RCl-SFNB-3.5.2.2.2.6-3 is contained in the email from Jessica Hammock (NRC) to Delson C. Erb (TVA) dated January 14, 2025 (ML25014A041, ML25014A042, ML25014A043, and ML25014A044).
TVA Response TVA confirms the NRC staffs observation without the need of clarification.
RCI-SFNB-3.5.2.2.2.6-4 The proprietary request for RCl-SFNB-3.5.2.2.2.6-4 is contained in the email from Jessica Hammock (NRC) to Delson C. Erb (TVA) dated January 14, 2025 (ML25014A041, ML25014A042, ML25014A043, and ML25014A044).
Enclosure CNL-25-025 E 3 of 5 TVA Response Bullet 1 - TVA confirms the NRC staffs observation with the following clarification:
The results consider secondary gamma sources from activated ex-core materials which are modeled as described in Section 2.1 of Electric Power Research Institute (EPRI)
Report 3002016055.
Bullet 2 - TVA confirms the NRC staffs observation without the need of clarification.
RCI-NVIB-3.5.2.2.2.6-1 SLRA 3.5.2.2.2.6 states that the initial nil-ductility transition temperature (NDTT) for ASTM A36 structural steel is estimated to be 39°F per Table 4-1 of NUREG-1509. The staff noted that this initial NDTT value of 39°F is the NDT + 1.3 value for carbon-manganese as-hot rolled steel in Table 4-1 of NUREG-1509, Radiation Effects on Reactor Pressure Vessel Supports, which may be used if plant-specific initial NDTT values are not available. Based on the audit review, the staff did not find any initial NDTT values in the plant-specific material certification documents for the ASTM A36 structural steel.
Confirm that plant-specific initial NDTT values are not available for the ASTM A36 structural steel.
TVA Response TVA confirms the NRC staffs understanding with the following additional information:
No plant-specific initial NDTT values for the American Society for Testing and Materials (ASTM) A36 structural steel were provided in the material testing reports for the sacrificial shield wall within contract 69C53-64800 noted on BFN drawing 0-48E445, Structural Steel Sacrificial Shield Wall Plans & Elevation, Revision 005.
RCI-NVIB-3.5.2.2.2.6-2 Based on the audit review of TVA BFN Drawing No. 1-47E225-102, Harsh Environmental Data
- Elevation View, Revision 002, and TVA BFN Drawing No. 1-47E225-100, Harsh Environmental Data - Drawing Series Index, Notes and References, Revision 008 for BFN Unit 1, the staff noted an average drywell temperature of 136°F and minimum drywell temperature of 60°F.
Confirm that the minimum drywell temperature of 60°F is not operating temperature.
Enclosure CNL-25-025 E 4 of 5 TVA Response TVA confirms the NRC staffs understanding with the following additional information:
The minimum drywell temperature of 60°F is not an operating temperature; it is a shutdown temperature that will not occur during reactor operation at power.
RCI-ESEB-3.5.2.2.1.2-1 During a breakout session associated with SLRA Section 3.5.2.2, AMR Results for Which Further Evaluation is Recommended by the GALL-SLR Report, the applicant informed the staff that the One-Time Inspection aging management program (AMP) per GALL-SLR,Section XI.M32 described in SLRA Section B.2.1.20 is used for managing aging effects of the metal reflective metallic thermal insulation; and External Surfaces Monitoring of Mechanical Components AMP per GALL-SLR XI.M36 described in SLRA Section B.2.1.23 is used for managing aging effect of the fiberglass non-metallic thermal insulation.
Confirm that insulations used for maintaining local area concrete temperature are the metal reflective metallic thermal insulation and the fiberglass non-metallic thermal insulation.
Confirm that the One-Time Inspection AMP per GALL-SLR,Section XI.M32 described in SLRA Section B.2.1.20 is used for managing aging effects of loss of material due to pitting and crevice corrosion, cracking due to stress corrosion cracking for the metal reflective metallic thermal insulation exposed to air-indoor uncontrolled environment.
Confirm that External Surfaces Monitoring of Mechanical Components AMP per GALL-SLR XI.M36 described in SLRA Section B.2.1.23 is used for managing the aging effect of reduced thermal insulation resistance due to moisture intrusion for the fiberglass non-metallic thermal insulation exposed to air, condensation environment.
TVA Response Bullet 1 - TVA confirms the stated information without the need of clarification.
Bullet 2 - TVA confirms the stated information without the need of clarification.
Bullet 3 - TVA confirms the stated information without the need of clarification.
RCI-ESEB-3.5.2.2.1.2-2 During a breakout session associated with SLRA Section 3.5.2.2, AMR Results for Which Further Evaluation is Recommended by the GALL-SLR Report, the applicant informed the staff that each drywell is cooled during normal operation of the unit by a closed loop ventilation system designed to hold the average temperature in the drywell to < 150°F, and the Containment and RBCCW (Reactor Building Closed Cooling Water) Systems are used to maintain the function of the drywell coolers in the closed loop ventilation system in removing heat and maintaining the temperature in the drywell. The applicant also informed the staff that the External Surfaces Monitoring of Mechanical Components AMP described in SLRA Section B.2.1.23 and the One-Time Inspection AMP described in SLRA Section B.2.1.20 are used for
Enclosure CNL-25-025 E 5 of 5 managing aging effects of the Containment System; and the Closed Treated Water Systems AMP described in SLRA Section B.2.1.12 and the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Component AMP described in SLRA Section B.2.1.24 are used for managing aging effects of the RBCCW System.
Confirm that the closed loop ventilation system that is described in SLRA Section 3.5.2.2.1.2 is referring to the style of the drywell coolers and drywell ventilation systems being a closed loop arrangement and not a system labeled as the closed loop ventilation system.
Confirm that the function of the drywell coolers in the closed loop ventilation system described in SLRA Section 3.5.2.2.1.2 is maintained by the Containment and RBCCW Systems described in SLRA Sections 2.3.2.1 and 2.3.3.23, respectively.
Confirm that the External Surfaces Monitoring of Mechanical Components AMP described in SLRA Section B.2.1.23 and the One-Time Inspection AMP described in SLRA Section B.2.1.20 are used for managing aging effects of the Containment System.
Confirm that the Closed Treated Water Systems AMP described in SLRA Section B.2.1.12 and the Inspection of Internal Surfaces in Miscellaneous Piping and Ducting Component AMP described in SLRA Section B.2.1.24 are used for managing aging effects of the RBCCW System.
TVA Response Bullet 1 - TVA confirms the stated information without the need of clarification.
Bullet 2 - TVA confirms the stated information without the need of clarification.
Bullet 3 - TVA confirms the stated information without the need of clarification.
Bullet 4 - TVA confirms the stated information without the need of clarification.