ML25219A009

Subsequent Renewed License - DPR-33

ML25219A009
Person / Time
Site:	Browns Ferry
Issue date:	12/11/2025
From:	Jeremy Groom NRC/NRR/DNRL/NLRP
To:	Tennessee Valley Authority
References
Download: ML25219A009 (0)

Text

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-259 BROWNS FERRY NUCLEAR PLANT, UNIT 1 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-33 1.

The Nuclear Regulatory Commission (NRC or the Commission), having previously made the findings set forth in Facility Operating License No. DPR-33 issued June 26, 1973, as superseded by Renewed Facility Operating License No. DPR-33 issued May 4, 2006, has now found that:

A.

The application to subsequently renew Facility Operating License No. DPR-33 filed by the Tennessee Valley Authority (TVA or the licensee) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I and all required notifications to other agencies or bodies have been duly made; B.

Construction of the Browns Ferry Nuclear Plant, Unit 1 (BFN or the facility) has been substantially completed in conformity with Construction Permit No. CPPR-29 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission; C.

Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the subsequent period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1); and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this subsequent renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations; D.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; E.

There is reasonable assurance: (i) that the activities authorized by this subsequent renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; Subsequent Renewed License No. DPR-33 F.

The licensee is technically and financially qualified to engage in the activities authorized by this subsequent renewed operating license in accordance with the rules and regulations of the Commission; G.

The licensee has satisfied the applicable provisions of 10 CFR Part 140, Financial Protection Requirements and Indemnity Agreements, of the Commissions regulations; H.

The issuance of this subsequent renewed operating license will not be inimical to the common defense and security or to the health and safety of the public; I.

After weighing the environmental, economic, technical and other benefits of the facility against environmental and other costs and considering available alternatives, the Commission concludes that the issuance of Subsequent Renewed Facility Operating License No. DPR-33, subject to the conditions for protection of the environment set forth herein, is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied; and J.

The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this subsequent renewed license will be in accordance with the Commissions regulations in 10 CFR Parts 30, 40, and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31.

2.

Based on the foregoing findings, and with the Atomic Safety and Licensing Board having dismissed the proceeding relating to the licensing action in a Memorandum and Order, dated November 27, 1973, Facility Operating License No. DPR-33 (issued June 26, 1973), which was superseded by Renewed Facility Operating License No.

DPR-33 issued May 4, 2006, is hereby superseded by Subsequent Renewed Facility Operating License No. DPR-33, which is issued to TVA to read as follows:

A.

This subsequent renewed license applies to the Browns Ferry Nuclear Plant, Unit 1, a boiling water nuclear reactor and associated equipment (the facility),

owned by TVA. The facility is located in Limestone County, Alabama, and is described in the Final Safety Analysis Report (Amendment 9) as supplemented and amended (Amendments 10 through 52), the licensees Draft Environmental Statement and supplement thereto dated July 1971, and November 8, 1971, respectively, and the licensees Final Environmental Statement dated September 1, 1972.

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses TVA:

(1)

Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess, use, and operate the facility at the designated location in Limestone County, Alabama, in accordance with the procedures and limitations set forth in this subsequent renewed license; (2)

Pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use at any time source and special nuclear material as reactor fuel in Subsequent Renewed License No. DPR-33 accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report as supplemented and amended; (3)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or equipment and instrument calibration or associated with radioactive apparatus or components; (5)

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This subsequent renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 3952 megawatts thermal.

(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 336, are hereby incorporated in this subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

For Surveillance Requirements (SRs) that are new in Amendment 234 to Facility Operating License DPR-33, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 234. For SRs that existed prior to Amendment 234, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the surveillance was last performed prior to implementation of Amendment 234.

(3)

The licensee is authorized to relocate certain requirements included in Appendix A and the former Appendix B to licensee-controlled documents.

Subsequent Renewed License No. DPR-33 Implementation of this amendment shall include the relocation of these requirements to the appropriate documents, as described in the licensees application dated September 6, 1996, as supplemented December 11, 1996; April 11, May 1, August 14, October 15, November 5 and 14, December 3, 4, 11, 22, 23, 29, and 30, 1997; January 23, March 12 and 13, April 16, 20, and 28, May 7, 14, 19, and 27, and June 2, 5, 10 and 19, 1998; evaluated in the NRC staffs Safety Evaluation enclosed with this amendment. This amendment is effective immediately and shall be implemented within 90 days of the date of this amendment.

(4)

Mitigation Strategy License Condition Develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1.

Pre-defined coordinated fire response strategy and guidance 2.

Assessment of mutual aid fire fighting assets 3.

Designated staging areas for equipment and materials 4.

Command and control 5.

Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1.

Protection and use of personnel assets 2.

Communications 3.

Minimizing fire spread 4.

Procedures for implementing integrated fire response strategy 5.

Identification of readily-available pre-staged equipment 6.

Training on integrated fire response strategy 7.

Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:

1.

Water spray scrubbing 2.

Dose to onsite responders (5)

The licensee shall implement and maintain all Actions required by to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

(6)

Deleted.

(7)

Deleted.

(8)

Deleted.

Subsequent Renewed License No. DPR-33 (9)

Deleted.

(10)

Deleted.

(11)(a) The licensee shall fully implement and maintain in effect all provisions of the Commission approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: Browns Ferry Nuclear Plant Physical Security Plan, Training and Qualification Plan, and Contingency Plan, submitted by letter dated April 28, 2006.

(b) The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The licensees CSP was approved by License Amendment No. 279, as amended by changes approved by License Amendment Nos. 286.

(12)

Deleted.

(13)

TVA Browns Ferry Nuclear Plant shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the license amendment request dated March 27, 2013; June 7, 2017; May 3, 2018, October 18, 2018; and July 3, 2019, as supplemented by letters dated May 16, 2013; December 20, 2013; January 10, 2014; January 14, 2014; February 13, 2014; March 14, 2014; May 30, 2014; June 13, 2014; July 10, 2014; August 29, 2014; September 16, 2014; October 6, 2014; December 17, 2014; March 26, 2015; April 9, 2015; June 19, 2015; August 18, 2015; September 8, 2015; October 20, 2015; September 18, 2017; October 23, 2017; February 13, 2019; and March 8, 2019, as approved in the Safety Evaluations dated October 28, 2015; December 19, 2017; October 9, 2018; April 2, 2019; and August 13, 2019. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature Subsequent Renewed License No. DPR-33 and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

(a) Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

(b) Prior NRC review and approval is not required for individual changes that result in a risk increase less than 1x10-7/year (yr) for CDF and less than 1x10-8/yr for LERF. The proposed change must also be consistent with the defense-in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

Other Changes that May Be Made Without Prior NRC Approval 1.

Changes to NFPA 805, Chapter 3, Fundamental Fire Protection Program.

Prior NRC review and approval are not required for changes to the NFPA 805, Chapter 3, fundamental fire protection program elements and design requirements for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3 element is functionally equivalent to the corresponding technical requirement. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant technical requirement or standard.

The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3 elements are acceptable because the alternative is adequate for the hazard. Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the functionality of the component, system, procedure, or physical arrangement, using a relevant Subsequent Renewed License No. DPR-33 technical requirement or standard. The four specific sections of NFPA 805, Chapter 3, are as follows:

Fire Alarm and Detection Systems (Section 3.8);

Automatic and Manual Water-Based Fire Suppression Systems Section 3.9);

Gaseous Fire Suppression Systems (Section 3.10); and Passive Fire Protection Features (Section 3.11).

This License Condition does not apply to any demonstration of equivalency under Section 1.7 of NFPA 805.

2.

Fire Protection Program Changes that Have No More than Minimal Risk lmpact Prior NRC review and approval are not required for changes to the licensees fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in the NRC Safety Evaluation dated October 28, 2015, to determine that certain fire protection program changes meet the minimal criterion. The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.

Transition License Conditions 1.

Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to the licensees fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.

2.

With the exception of Modifications 102 and 106, the licensee shall implement modifications to its facility, as described in Table S-2, Plant Modifications Committed, of TVA letter CNL-18-100, dated October 18, 2018, as supplemented by letter CNL-19-027, dated February 13, 2019, to complete the transition to full compliance with 10 CFR 50.48(c) no later than the end of the second refueling outage (for each unit) following issuance of the NFPA 805 License Amendment dated October 28, 2015. Modifications 102 and 106 as described in Table S-2, shall be implemented no later than the end of BFN Unit 1s Fall 2020 outage, and April 30, 2020, respectively. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

3.

The licensee shall complete Implementation Items 09, 32, 33, and the second part of Implementation Item 47 as listed in Table S-3, Implementation Items, of TVA letter CNL-17-130 dated October 23, 2017. Implementation Item 09 shall be completed by June 29, 2018.

Implementation Items 32, 33, and the second part of Implementation Item 47, i.e., resolving Finding level Facts and Observations, are Subsequent Renewed License No. DPR-33 associated with modifications and will be completed after all procedure updates, modifications, and training are complete.

(14)

The licensee shall maintain the Augmented Quality Program for the Standby Liquid Control System to provide quality control elements to ensure component reliability for the required alternative source term function defined in the Updated Final Safety Analyses Report (UFSAR).

(15)

The licensee is required to confirm that the conclusions made in TVAs letter dated September 17, 2004, for the turbine building remain acceptable using seismic demand accelerations based on dynamic seismic analysis prior to the restart of BFN Unit 1.

(16)

Upon implementation of Amendment No. 275, adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.3.4, in accordance with TS 5.5.13.c(i),

the assessment of the CRE habitability as required by TS 5.5.13.c(ii), and the measure of CRE pressure as required by TS 5.5.13.d, shall be considered met.

Following Implementation:

(a) The first performance of SR 3.7.4.4, in accordance with TS 5.5.13.c(i),

shall be within a specific frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from November 10, 2003, the date of the most recent successful tracer gas test.

(b) The first performance of the periodic assessment of the Control Room Envelope (CRE) Habitability, Technical Specification 5.5.13.c(ii), shall be within 9 months following the initial implementation of the TS Change. The next performance of the periodic assessment will be in a period specified by the CRE Program. That is 3 years from the last successful performance of the Technical Specification 5.5.13.c(ii) tracer gas test.

(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.13.d, shall be within 24 months, plus 180 days allowed by SR 3.0.2 as measured from the date of the most recent successful pressure measurement test.

(17)

The fuel channel bow standard deviation component of the channel bow model uncertainty used by ANP-10307PA, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, Revision 0, (i.e., TS 5.6.5.b.11) to determine the Safety Limit Minimum Critical Power Ratio shall be increased by the ratio of channel fluence gradient to the nearest channel fluence gradient bound of the channel measurement database, when applied to channels with fluence gradients outside the bounds of the measurement database from which the model uncertainty is determined.

This license condition will be effective upon the implementation of Amendment 285.

Subsequent Renewed License No. DPR-33 (18)

Potential Adverse Flow Effects This license condition provides for monitoring, evaluating, and taking prompt action in response to potential adverse flow effects as a result of power uprate operation on plant structures, systems, and components (including verifying the continued structural integrity of the steam dryer) for initial power ascension from 3458 MWt to the extended power uprate (EPU) level of 3952 MWt.

(a) The following requirements are placed on operation of the facility before and during the initial power ascension:

1.

TVA shall provide a Power Ascension Test (PAT) Plan for the BFN Unit 1 steam dryer testing. The PAT plan shall be submitted to the NRC Project Manager no later than 10 days before start-up.

2.

TVA shall monitor the main steamline (MSL) strain gauges at a minimum of three power levels up to 3458 MWt. If the number of active MSL strain gauges is less than two strain gauges (180 degrees apart) at any of the eight MSL locations, TVA will stop power ascension and repair/replace the damaged strain gauges and only then resume power ascension.

3.

At least 60 days prior to exceeding 3458 MWt after the BFN Unit 1 EPU outage, TVA shall revise the BFN Unit 1 replacement steam dryer (RSD) analysis utilizing the BFN Unit 3 on-dryer strain gauge based end-to-end bias and uncertainties (B&Us) at EPU conditions, and submit the information including the updated limit curves and a list of dominant frequencies for BFN Unit 1, to the NRC as a report in accordance with 10 CFR 50.4.

a.

If the on-dryer instrumentation was not available when BFN Unit 3 reached a power level of 3458 MWt and the BFN-specific bias and uncertainty data and transfer function could not be developed, the predicted dryer loads during the BFN Unit 1 power ascension will be calculated with the Plant Based Load Evaluation Method 2 transfer function used in the steam dryer design analyses for EPU. The acceptance limits will be based on BFN Unit 3 steam dryer confirmatory stress analysis results using the MSL strain gauge data collected at EPU conditions. The acceptance limits will ensure the steam dryer stress margins remain above the minimum alternating stress ratio (MASR) determined in the BFN Unit 3 steam dryer EPU confirmatory analyses.

4.

TVA shall evaluate the BFN Unit 1 limit curves prepared in item (a)3 above based on new MSL strain gauge data collected following the BFN Unit 1 EPU outage at or near 3458 MWt. If the limit curves change, the new post-EPU outage limit curves shall be provided to the NRC Project Manager. TVA shall not increase power above 3458 MWt for at least 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after the NRC Subsequent Renewed License No. DPR-33 Project Manager confirms receipt of the reports unless, prior to expiration of the 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> period, the NRC Project Manager advises that the NRC staff has no objections to the continuation of power ascension.

5.

TVA shall monitor the MSL strain gauges during power ascension above 3458 MWt for increasing pressure fluctuations in the steam lines. Upon the initial increase of power above 3458 MWt until reaching 3952 MWt, TVA shall collect data from the MSL strain gauges at nominal 2.5 percent of 3458 MWt (approximately 86 MWt) increments and evaluate steam dryer performance based on this data.

6.

During power ascension at each nominal 2.5 percent power level above 3458 MWt (approximately 86 MWt), TVA shall compare the MSL data to the approved limit curves based on end-to-end B&Us from the BFN Unit 3 benchmarking at EPU conditions and determine the MASR.

7.

TVA shall hold the facility at approximately 3630 MWt and 3803 MWt to perform the following:

a.

Collect strain data from the MSL strain gauges.

b.

Collect vibration data for the locations included in the vibration summary report discussed above.

c.

Evaluate steam dryer performance based on MSL strain gauge data.

d.

Evaluate the measured vibration data (collected in item 7.b above) at that power level, data projected to EPU conditions, trends, and comparison with the acceptance limits.

e.

Provide the steam dryer evaluation and the vibration evaluation, including the data collected, to the NRC Project Manager, upon completion of the evaluation for each of the hold points.

f.

TVA shall not increase power above each hold point until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after the NRC Project Manager confirms receipt of the evaluations unless, prior to the expiration of the 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> period, the NRC Project Manager advises that the NRC staff has no objections to the continuation of power ascension.

8.

If any frequency peak from the MSL strain gauge data exceeds the Level 1 limit curves, TVA shall return the facility to a power level at which the limit curve is not exceeded. TVA shall resolve the discrepancy, evaluate and document the continued structural integrity of the steam dryer, and provide that documentation to the NRC Project Manager prior to further increases in reactor power.

Subsequent Renewed License No. DPR-33 If a revised stress analysis is performed and new limit curves are developed, then TVA shall not further increase power above each hold point until 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> after the NRC Project Manager confirms receipt of the documentation or until the NRC Project Manager advises that the NRC staff has no objections to the continuation of power ascension, whichever comes first. Additional detail is provided in item (b)1 below.

(b) TVA shall implement the following actions for the initial power ascension from 3458 MWt to 3952 MWt condition:

1.

In the event that acoustic signals (in MSL strain gauge signals) are identified that exceed the Level 1 limit curves during power ascension above 3458 MWt, TVA shall re-evaluate dryer loads and stresses, and re-establish the limit curves. In the event that stress analyses are re-performed based on new strain gauge data to address item (a)7 above, the revised load definition, stress analysis, and limit curves shall include:

a.

Application of end-to-end B&Us as determined from BFN Unit 3 EPU measurements.

b.

Use of scaling factors associated with all of the safety relief valve acoustic resonances as estimated in the predictive analysis or in-plant data acquired during power ascension.

2.

After reaching 3952 MWt, TVA shall obtain measurements from the MSL strain gauges and establish the steam dryer flow-induced vibration load fatigue margin for the facility and update the steam dryer stress report. These data will be provided to the NRC staff as described below in item (e).

(c) TVA shall prepare the EPU PAT Plan to include the following.

1.

The MSL strain gauge limit curves to be applied for evaluating steam dryer performance, based on end-to-end B&Us from BFN Unit 3 benchmarking at EPU conditions.

2.

Specific hold points and their durations during EPU power ascension.

3.

Activities to be accomplished during the hold points.

4.

Plant parameters to be monitored.

5.

Inspections and walkdowns to be conducted for steam, feedwater, and condensate systems and components during the hold points.

6.

Methods to be used to trend plant parameters.

Subsequent Renewed License No. DPR-33 7.

Acceptance criteria for monitoring and trending plant parameters, and conducting the walkdowns and inspections.

8.

Actions to be taken if acceptance criteria are not satisfied.

9.

Verification of the completion of commitments and planned actions specified in the application and all supplements to the application in support of the EPU license amendment request pertaining to the steam dryer prior to power increase above 3458 MWt.

10. Identify the NRC Project Manager as the NRC point of contact for providing PAT plan information during power ascension.

11. Methodology for updating limit curves.

(d) The following key attributes of the PAT Plan shall not be made less restrictive without prior NRC approval:

1.

During initial power ascension testing above 3458 MWt, each of the two hold points shall be at increments of approximately 5 percent of 3458 MWt.

2.

Level 1 performance criteria.

3.

The methodology for establishing the limit curves used for the Level 1 and Level 2 performance.

Changes to other aspects of the PAT Plan may be made in accordance with the guidance of NEI 99-04, Guidelines for Managing NRC Commitments, issued July 1999.

(e) Following the data collection and evaluation at the EPU power level, TVA shall provide a final load definition and stress report of the steam dryer, including the results of a complete re-analysis using the end-to-end B&Us from BFN Unit 3 benchmarking at EPU conditions.

The report shall be submitted to NRC within 90 days of the completion of EPU power ascension testing for BFN Unit 1. Should the results of this stress analysis indicate the allowable stress in any part of the steam dryer is exceeded, TVA shall reduce power to a level at which the allowable stress is met, evaluate the steam dryer integrity, and assess any shortcomings in the predictive analysis. The results of this evaluation, including a recommended resolution of any identified issues and a demonstration of steam dryer integrity at EPU conditions, shall be provided to the NRC for review and approval prior to return to EPU conditions.

(f) Following the data collection and evaluation at the EPU power level, TVA shall provide a vibration summary report to the NRC. The summary report shall be submitted to the NRC within 90 days of the completion of EPU power ascension testing for BFN Unit 1. The Subsequent Renewed License No. DPR-33 vibration summary report shall include the information in items (f)1 through (f)3, as follows:

1.

Vibration data for piping and valve locations deemed prone to vibration and vibration monitoring locations identified in 5 to the EPU application dated September 21, 2015, including the identified locations associated with MSLs, Feedwater Lines, Safety Relief Valves and the Main Steam Isolation Valves.

2.

An evaluation of the measured vibration data collected in item (f)1 above compared against acceptance limits.

3.

Vibration values and associated acceptance limits at approximately 3630 MWt, 3803 MWt, and 3952 MWt using the data collected in item (f)1, above.

(g) During the first two scheduled refueling outages after reaching EPU conditions, a visual inspection shall be conducted of the steam dryer as described in the inspection guidelines contained in Boiling Water Reactor Vessel and Internals Project (BWRVIP)-139A (Steam Dryer Inspection and Flaw Evaluation Guidelines) and General Electric (GE) inspection guidelines (SIL 644, BWR Steam Dryer Integrity).

(h) The results of the visual inspections of the steam dryer shall be submitted to the NRC staff in a report in accordance with 10 CFR 50.4. The report shall be submitted to the NRC within 90 days following startup from each of the first two respective refueling outages.

(i) Within 6 months following completion of the second refueling outage, after the implementation of the EPU, the licensee shall submit a long-term steam dryer inspection plan based on industry operating experience along with the baseline inspection results.

The license condition described above shall expire: (1) upon satisfaction of the requirements in items (g) and (h), provided that a visual inspection of the steam dryer does not reveal any new unacceptable flaw(s) or unacceptable flaw growth that is due to fatigue, and (2) upon satisfaction of the requirements specified in item (i).

(19)

Neutron Absorber Monitoring Program The licensee shall, at least once every ten years, withdraw a neutron absorber coupon from the spent fuel pool and perform Boron-10 (B-10) areal density measurement on the coupon. Based on the results of the B-10 areal density measurement, the licensee shall perform any technical evaluations that may be necessary and take appropriate actions using relevant regulatory and licensing processes.

Subsequent Renewed License No. DPR-33 (20)

Radiological Consequences Analyses Using Alternative Source Term TVA shall perform facility and licensing basis modifications to resolve the non-conforming/degraded condition associated with the Alternate Leakage Treatment pathway such that the current licensing basis dose calculations (approved in License Amendment Nos. 251/282 (Unit 1),

290/308 (Unit 2) and 249/267 (Unit 3)) would remain valid. These facility and licensing basis modifications shall be complete prior to initial power ascension above 3458 MWt.

(21)

Prior to extending the frequency for the Integral Leakage Rate Testing described in TS 5.5.12, the licensee shall implement the modifications, that are modeled in the Fire PRA and described in Table S-2, Plant Modifications Committed, of Tennessee Valley Authority letter CNL-18-100, dated October 18, 2018; as supplemented by letter CNL-19-027, dated February 13, 2019.

(22)

Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Special Consideration The licensee shall not operate the facility within the MELLLA+ operating domain with more than a 10°F reduction in feedwater temperature below the design feedwater temperature.

(23)

Maximum Extended Load Line Limit Analysis (MELLLA+) Implementation Prior to the first implementation of MELLLA+, TVA shall perform reload safety analysis using codes that have been corrected for the errors described in TVA letter CNL-19-125, dated December 19, 2019.

(24)

TVA shall close all open Facts and Observations (F&Os) listed in Tables 11 and 13 to Attachment 2 of TVA letter CNL-20-003, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (BFN-TS-516), dated March 27, 2020, prior to implementing any Surveillance Test Interval extensions under the Surveillance Frequency Control Program. The F&O closures will be performed in accordance with the ASME/ANS RA-Sa-2009 PRA Standard, as endorsed by Regulatory Guide 1.200.

(25)

Adoption of 10 CFR 50.69, Risk-Informed Categorization and treatment of structures, systems and components for nuclear power plants (1) TVA is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, internal fire, and seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Subsequent Renewed License No. DPR-33 Class 2 and Class 3 SSCs and their associated supports; and the results of non-PRA evaluations that are based on the Individual Plant Examination of External Events (IPEEE) Screening Assessment for External Hazards, and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; Internal fires and seismic hazards are evaluated with BFN specific PRA models; as specified in License Amendment No. 317.

(2) TVA shall complete the numbered items listed in Enclosure 2, List of Categorization Prerequisites, of TVA letter ML21118B079, dated April 28, 2021, prior to implementation. All issues identified in the enclosure will be addressed and any associated changes will be made, focused scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

(3) Prior NRC approval, under 10 CFR 50.69, is required for a change to the categorization process specified above (e.g., change from a shutdown defense in depth approach to a shutdown probabilistic risk assessment approach).

D.

The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d),

shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4) following the issuance of the renewed operating license.

Until that update is complete, TVA may make changes to the programs and activities described in the supplement without prior Commission approval, provided that TVA evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

E.

The UFSAR supplement, as revised, describes certain future activities to be completed prior to the period of extended operation. TVA shall complete these activities no later than December 20, 2013, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

F.

All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessels and Internals Project (BWRVIP)

Integrated Surveillance Program (ISP) appropriate for the configuration of the specimens in the capsule. Any changes to the BWRVIP ISP capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.

G.

(1) During the power uprate power ascension test program and prior to exceeding 30 days of plant operation above a nominal 3293 megawatts thermal power level (100-percent OLTP) or within 30 days of satisfactory Subsequent Renewed License No. DPR-33 completion of steam dryer monitoring and testing that is necessary for achieving 105-percent OLTP (whichever is longer), with plant conditions stabilized at 105-percent OLTP, TVA shall trip a condensate booster pump, a condensate pump, and a main feedwater pump on an individual basis (i.e., one at a time). Following each pump trip, TVA shall confirm that plant response to the transient is as expected in accordance with previously established acceptance criteria. Evaluation of the test results for each test shall be completed and all discrepancies resolved in accordance with corrective action program requirements and the provisions of the power ascension test program.

(2) Deleted.

H.

The licensee must complete the thirteen (13) Unit 1 restart commitments that are discussed in Appendix F of the license renewal application, dated December 31, 2003, as supplemented by letters dated January 31, 2005, March 2, and April 21, 2006. Completion of these activities must be met prior to power operation of Unit 1.

I.

Subsequent License Renewal License Conditions (1)

The information in the Updated Final Safety Analysis Report Supplement submitted as required by 10 CFR 54.21(d), and revised during the application review process, and the licensees commitments listed in Appendix A of the Safety Evaluation for the SLRA of Browns Ferry Nuclear Plant, Units 1, 2, and 3, dated July 18, 2025, are collectively the Subsequent License Renewal Final Safety Analysis Report Supplement.

This supplement is henceforth part of the Updated Final Safety Analysis Report which will be updated in accordance with 10 CFR 50.71(e).

As such, the licensee may make changes to the programs, activities, and commitments described in the Subsequent License Renewal Updated Final Safety Analysis Report Supplement, provided the licensee evaluates such changes pursuant to 10 CFR 50.59, Changes, Tests and Experiments, and otherwise complies with the requirements in that section.

(2)

This Subsequent License Renewal Final Safety Analysis Report Supplement, as defined in subsequent renewed license condition (1) above, describes programs to be implemented and activities to be completed before the subsequent period of extended operation, which is the period following the expiration of the initial renewed license on December 20, 2033.

a.

The licensee shall implement those new programs and enhancements to existing programs no later than the date 6 months before the subsequent period of extended operation.

b.

The licensee shall complete those activities by the date 6 months prior to the subsequent period of extended operation or by the end of the Subsequent Renewed License No. DPR-33 last refueling outage before the subsequent period of extended operation, whichever occurs later.

c.

The licensee shall notify the NRC in writing within 30 days after having accomplished item (2)a above and include the status of those activities that have been or remain to be completed in item (2)b above.

d.

The programs and commitments described in the Subsequent License Renewal Final Safety Analysis Report Supplement shall continue in effect during the subsequent period of extended operation, to the extent set forth therein, unless modified in accordance with the process set forth in 10 CFR 50.59.

J.

This subsequent renewed license is effective as of the date of issuance and shall expire at midnight on December 20, 2053.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Jeremy Groom, Acting Director Office of Nuclear Reactor Regulation Attachments:

1.

BFN Unit 1 - Technical Specifications - Appendices A and B Date of Issuance: December 11, 2025