ML25219A011

Subsequent Renewed License - DPR-68

ML25219A011
Person / Time
Site:	Browns Ferry
Issue date:	12/11/2025
From:	Jeremy Groom NRC/NRR/DNRL/NLRP
To:	Tennessee Valley Authority
References
Download: ML25219A011 (0)

Text

TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-296 BROWNS FERRY NUCLEAR PLANT, UNIT 3 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-68 1.

The Nuclear Regulatory Commission (NRC or the Commission), having previously made the findings set forth in Facility Operating License No. DPR-68 issued July 2, 1976, as superseded by Renewed Facility Operating License No. DPR-68 issued May 4, 2006, has now found that:

A.

The application to subsequently renew Facility Operating License No. DPR-68 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 3 (BFN or the facility) has been substantially completed in conformity with Construction Permit No.

CPPR-48 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-68 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-68, 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-68 (issued July 2, 1976),

which was superseded by Renewed Facility Operating License No. DPR-68 issued May 4, 2006, is hereby superseded by Subsequent Renewed Facility Operating License No. DPR-68, which is issued to TVA to read as follows:

A.

This subsequent renewed license applies to the Browns Ferry Nuclear Plant, Unit 3, 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 65), 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-68 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. 319, are hereby incorporated in the 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 212 to Facility Operating License DPR-68, the first performance is due at the end of the first surveillance interval that begins at implementation of the Amendment 212. For SRs that existed prior to Amendment 212, 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 212.

(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-68 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 May 1, August 14, November 5 and 14, December 3, 4, 11, 22, 23, 29, and 30, 1997; January 23, March 12, 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)

Deleted.

(5)

Classroom and simulator training on all power uprate related changes that affect operator performance will be conducted prior to operating at uprated conditions. Simulator changes that are consistent with power uprate conditions will be made and simulator fidelity will be validated in accordance with ANSI/ANS 3.5-1985. Training and the plant simulator will be modified, as necessary, to incorporate changes identified during startup testing. This amendment is effective immediately.

(6)(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, Revision 4, 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. 265, as amended by changes approved by License Amendment Nos. 271 and 281.

(7)

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 Subsequent Renewed License No. DPR-68 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 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 Subsequent Renewed License No. DPR-68 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 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 Subsequent Renewed License No. DPR-68 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 associated with modifications and will be completed after all procedure updates, modifications, and training are complete.

(8)

Deleted.

(9)

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).

(10)

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 (11)

The licensee shall implement and maintain all Actions required by to NRC Order EA-06-137, issued June 20, 2006, except the Subsequent Renewed License No. DPR-68 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.

(12)

Upon implementation of Amendment No. 261, 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 measurement 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.

(d) For License Amendment 268, the licensee shall implement changes to BFN, Unit 3 TSs 5.6.5 and 3.3.1.1 within 60 days of approval. The remaining BFN, Unit 3, changes will be implemented upon completion of required supporting modification work and prior to entering Mode 3 (i.e., Hot Shutdown) from the spring 2014 refueling outage.

(13)

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.10) 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 No. 270.

Subsequent Renewed License No. DPR-68 (14)

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 3 steam dryer testing. This plan shall include:

a.

Criteria for comparison and evaluation of projected strain and acceleration with on-dryer instrument data.

b.

Acceptance limits developed for each on-dryer strain gauge.

c.

Tables of predicted dryer stresses at a power level of 3458 MWt, strain amplitudes and power spectral densities at strain gauge locations, and maximum stresses and locations.

The PAT plan shall provide correlations between measured strains and the corresponding maximum stresses. 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 and on-dryer instrumentation at a minimum of three power levels up to 3458 MWt. Based on a comparison of projected and measured strains and accelerations, BFN will assess whether the dryer acoustic and structural models have adequately captured the response significant to peak stress projections. If the measured strains and accelerations are not within the 3458 MWt acceptance limits, the new measured data will be used to re-perform the full structural re-analysis for the purposes of generating modified EPU acceptance limits.

a.

If the on-dryer instrumentation is unavailable, the BFN Unit 3 power ascension will be monitored using the available MSL strain gauges. The predicted dryer loads during the power ascension will be calculated with the Plant Based Load Evaluation (PBLE) Method 2 transfer function used in the steam dryer design analyses for EPU. The acceptance limits will ensure that the steam dryer stress margins remain above the final minimum alternating stress ratio (MASR) accepted in the EPU design analyses.

Subsequent Renewed License No. DPR-68 3.

BFN shall provide a summary of the data and evaluation of predicted and measured pressures, strains, and accelerations at a power level of 3458 MWt. These data will include the BFN-specific bias and uncertainty data and transfer function, revised peak stress table and any revised acceptance limits. The predicted pressures shall include those using both PBLE methods (that is, Method 1 using on-dryer data, and Method 2 using MSL data). It shall be provided to the NRC Project Manager upon completion of the evaluation. TVA shall not increase power above 3458 MWt until the NRC Project Manager notifies TVA that NRC accepts the evaluation or NRC questions regarding the evaluation have been addressed. If no questions are identified within 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> after the NRC receives the evaluation, power ascension may continue.

a.

If the on-dryer instrumentation is unavailable and the BFN-specific bias and uncertainty data and transfer function cannot be developed when BFN Unit 3 reaches a power level of 3458 MWt, the BFN Unit 3 power ascension above 3458 MWt will be monitored using the available MSL strain gauges.

The predicted dryer loads during the power ascension will be calculated with the PBLE Method 2 transfer function used in the steam dryer design analyses for EPU. The acceptance limits will ensure that the steam dryer stress margins remain above the final MASR accepted in the EPU design analyses.

(b) The following requirements are placed on operation of the facility during the initial power ascension from 3458 MWt to the approved EPU level (3952 MWt):

1.

At test increments that do not exceed 2.5 percent of 3458 MWt (approximately 86 MWt), TVA shall hold the facility at approximately steady state conditions and collect data from available MSL strain gauges and available on-dryer instrumentation. This data will be evaluated, including the comparison of measured dryer strains to acceptance limits and the comparison of predicted dryer loads based on MSL strain gauge data to acceptance limits. It will also be used to trend and project loads at the next test point and to EPU conditions to demonstrate margin for continued power ascension.

a.

If the on-dryer instrumentation becomes unavailable during power ascension above 3458 MWt, the BFN Unit 3 power ascension above 3458 MWt will be monitored using the available MSL strain gauges. The predicted dryer loads during the power ascension will be calculated with the BFN-specific PBLE Method 2 transfer function developed from the on-dryer instrumentation and MSL strain gauge data taken at the 3458 MWt hold point, the BFN-specific bias and uncertainty data, the revised peak stresses, and revised acceptance criteria developed in item (a)3 above. The acceptance limits Subsequent Renewed License No. DPR-68 will maintain the steam dryer stress margins above a MASR of 1.0.

2.

Following the data collection and evaluation at the plateaus at approximately 3630 MWt, 3803 MWt, and 3952 MWt, TVA shall provide a summary of the data and the evaluation performed in item (b)1 above to the NRC Project Manager. TVA shall not increase power above these power levels for up to 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 summary, 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 objection to continuation of power ascension.

3.

Should the measured strains on the dryer exceed the Level 1 acceptance limits, or alternatively if the dryer instrumentation is not available and the projected load on the dryer from the MSL strain gauge data exceeds the Level 1 acceptance limits, TVA shall return the facility to a power level at which the limits are 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. TVA shall not increase power for up to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to allow for NRC review and approval of the information.

a.

In the event that acoustic signals (in MSL strain gauge signals) are identified that challenge the dryer acceptance limits during power ascension above 3458 MWt, TVA shall evaluate dryer loads, and stresses, including the effect of +/-10 percent frequency shift, and re-establish the acceptance limits and determine whether there is margin for continued power ascension.

b.

During power ascension above 3458 MWt, if an engineering evaluation for the steam dryer is required because a Level 1 acceptance limit is exceeded, TVA shall perform the structural analysis using the Steam Dryer Report, Appendix A methods to address frequency uncertainties up to +/- 10 percent and assure that peak responses that fall within this uncertainty band are addressed.

4.

a.

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 BFN-specific bias and uncertainties and transfer function, to the NRC. The BFN-specific bias and uncertainties summary shall include both PBLE Method 1 and Method 2. This report shall be submitted to the NRC within 90 days of the completion of EPU power ascension testing for BFN Unit 3. Should the results of this stress analysis indicate the allowable stress in any part of the dryer is exceeded, TVA Subsequent Renewed License No. DPR-68 shall reduce power to a level at which the allowable stress is met, evaluate the 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 dryer integrity at EPU conditions, shall be provided to the NRC for review and approval prior to return to EPU conditions.

b.

Within 30 days after completion of the core flow sweep test at EPU conditions to determine any compounding effect due to alignment of Vane Passing Frequency and Safety Relief Valve resonance frequencies, TVA shall provide the core flow sweep test results for NRC review.

5.

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 3.

The vibration summary report shall include the information in items 5.a through 5.c, as follows:

a.

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.

b.

An evaluation of the measured vibration data collected in item 5.a above compared against acceptance limits.

c.

Vibration values and associated acceptance limits at approximately 3630 MWt, 3803 MWt, and 3952 MWt using the data collected in item 5.a, above.

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

1.

Level 1 and Level 2 acceptance limits for on-dryer strain gauges and for projected dryer loads from MSL strain gauge data to be used up to 3952 MWt.

2.

Specific hold points and their duration during EPU power ascension.

3.

Activities to be accomplished during 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.

Subsequent Renewed License No. DPR-68 6.

Methods to be used to trend plant parameters.

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 TVA application and all supplements to the application in support of the EPU LAR pertaining to the steam dryer before 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.

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) During the first two scheduled refueling outages after reaching full EPU conditions, TVA shall conduct a visual inspection of all accessible, susceptible locations of the steam dryer in accordance with Boiling Water Reactor Vessels and Internals Project (BWRVIP)-139A (Steam Dryer Inspection and Flaw Evaluation Guidelines) and General Electric) GE inspection guidelines (SIL 644, BWR Steam Dryer Integrity).

(f) 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 NRC within 90 days following startup from each of the first two respective refueling outages.

Subsequent Renewed License No. DPR-68 (g) 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.

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

(15)

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.

(16)

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.

(17)

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.

(18)

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.

(19)

Maximum Extended Load Line Limit Analysis Plus (MELLLA+)

Implementation Prior to the first implementation of MELLLA+, TVA shall perform reload safety analyses using codes that have been corrected for the errors described in TVA letter CNL-19-125, dated December 19, 2019.

Subsequent Renewed License No. DPR-68 (20)

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.

(21)

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 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. 300.

(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 50.90, 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, Subsequent Renewed License No. DPR-68 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 July 2, 2016, 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.

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 July 2, 2036.

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-68 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.

H.

This subsequent renewed license is effective as of the date of issuance and shall expire at midnight on July 2, 2056.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

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

1.

BFN Unit 3 - Technical Specifications - Appendices A and B Date of Issuance: December 11, 2025