CNL-22-100, to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3,in Support of Atrium 11 Fuel Use at Browns Ferry (TS-535)

to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3,in Support of Atrium 11 Fuel Use at Browns Ferry (TS-535)

ML22343A092
Person / Time
Site:	Browns Ferry
Issue date:	12/09/2022
From:	Jim Barstow Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML22343A091	List: CNL-22-100, to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3,in Support of Atrium 11 Fuel Use at Browns Ferry (TS-535)
References
EPID L-2021-LLA-0132, CNL-22-100
Download: ML22343A092 (1)

Text

Proprietary Information - Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosures 1 and 4 1101 Market Street, Chattanooga, Tennessee 37402 CNL-22-100 December 9, 2022 10 CFR 50.90 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:

Supplement 7 to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3, in Support of ATRIUM 11 Fuel Use at Browns Ferry (TS-535)

(EPID L-2021-LLA-0132)

References:

1. TVA letter to NRC, CNL-21-053, Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3, in Support of ATRIUM 11 Fuel Use at Browns Ferry (TS-535), dated July 23, 2021 (ML21204A128 and ML21204A129)

2. TVA letter to NRC, CNL-22-076, Supplement 4 to Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3, in Support of ATRIUM 11 Fuel Use at Browns Ferry (TS-535) (EPID L-2021-LLA-0132), dated July 28, 2022 (ML22209A238)

In Reference 1, Tennessee Valley Authority (TVA) submitted a request for a Technical Specification (TS) amendment for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3.

The proposed license amendments, in part, revise TS 5.6.5.b, Core Operating Limits Report (COLR), to allow application of Advanced Framatome Methodologies for determining core operating limits in support of loading Framatome fuel type ATRIUMTM 1 11.

1 ATRIUM 11 is a trademark or registered trademarks of Framatome, Inc., its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

Proprietary Information - Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosures 1 and 4

Proprietary Information - Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosures 1 and 4 U.S. Nuclear Regulatory Commission CNL-22-100 Page 2 December 9, 2022 In Attachment 4 to Reference 1, in part, TVA committed to submit the BFN Unit 2 Cycle 23 Reload Analysis Report (i.e., ANP-4017P) to the Nuclear Regulatory Commission (NRC), for information only, within 15 days following TVA approval of the report. This report was approved by TVA on November 29, 2022, and is included in Enclosure 1.

Additionally, in Reference 2, TVA submitted the BFN Unit 2 Cycle 23 Fuel Rod Design Report to the NRC for information. It was later determined that Tables 3-2 and 3-3 of the Reference 2 enclosure contained some values that had minor rounding errors. Therefore, Framatome Inc. (Framatome) revised the Fuel Rod Design Report to correct these errors.

The issue was entered into the TVA corrective action program. Enclosure 4 of this submittal provides the revised report.

Enclosures 1 and 4 contain information considered proprietary to Framatome. As owner of the proprietary information, Framatome has executed affidavits contained in Enclosures 3 and 6 which identify the information as proprietary, is customarily held in confidence, and should be withheld from public disclosure in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390. Enclosures 2 and 5 provide the non-proprietary versions of Enclosures 1 and 4, respectively. Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Framatome affidavits should reference the corresponding reports and should be addressed to Alan Meginnis, Framatome, Manager, Product Licensing, 2101 Horn Rapids Road, Richland, WA 99354.

This letter does not change the no significant hazards consideration or the environmental considerations contained in the referenced letter. Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and the non-proprietary enclosures to the Alabama Department of Public Health.

There are no new regulatory commitments associated with this submittal. Please address any questions regarding this submittal to Stuart L. Rymer, Director (Acting), Nuclear Regulatory Affairs, at slrymer@tva.gov.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 9th day of December 2022.

Respectfully, Digitally signed by Rearden, Pamela S Date: 2022.12.09 08:50:47 -05'00' James Barstow Vice President, Nuclear Regulatory Affairs & Support Services Enclosures cc: See Page 3 Proprietary Information - Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosures 1 and 4

Proprietary Information - Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosures 1 and 4 U.S. Nuclear Regulatory Commission CNL-22-100 Page 3 December 9, 2022

Enclosures:

1. ANP-4017P, Browns Ferry Unit 2 Cycle 23 Reload Analysis Report, Revision 0, Framatome, November 2022 (Proprietary)

2. ANP-4017NP, Browns Ferry Unit 2 Cycle 23 Reload Analysis Report, Revision 0, Framatome, November 2022 (Non-Proprietary)

3. Framatome Affidavit for Enclosure 1

4. ANP-4001P, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23, Revision 1, Framatome, August 2022 (Proprietary)

5. ANP-4001NP, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23, Revision 1, Framatome, August 2022 (Non-Proprietary)

6. Framatome Affidavit for Enclosure 4 cc: (Enclosures):

NRC Regional Administrator - Region II NRC Project Manager - Browns Ferry Nuclear Plant NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health (w/o Enclosures 1 and 4)

Proprietary Information - Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosures 1 and 4

Proprietary Information - Withhold Under 10 CFR § 2.390 Enclosure 1 ANP-4017P, Browns Ferry Unit 2 Cycle 23 Reload Analysis Report, Revision 0, Framatome, November 2022 (Proprietary)

CNL-22-100 Proprietary Information - Withhold Under 10 CFR § 2.390

Enclosure 2 ANP-4017NP, Browns Ferry Unit 2 Cycle 23 Reload Analysis Report, Revision 0, Framatome, November 2022 (Non-Proprietary)

CNL-22-100

Browns Ferry Unit 2 Cycle 23 ANP-4017NP Revision 0 Reload Analysis November 2022

© 2022 Framatome Inc.

0414-12-F04 (Rev. 004, 04/27/2020)

ANP-4017NP Revision 0 Copyright © 2022 Framatome Inc.

All Rights Reserved ATRIUM is a trademark or registered trademark of Framatome or its affiliates, in the USA or other countries.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page i Nature of Changes Section(s) or Item Page(s) Description and Justification

1. All This is the initial release.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page ii Contents INTRODUCTION ............................................................................................... 1-1 DISPOSITION OF EVENTS AND PLANT MODELING SENSITIVITIES ................................................................................................. 2-1 Disposition of Events for ATRIUM 11 Fuel Introduction .......................... 2-1 Plant Specific Modeling Sensitivities..................................................... 2-12 Inadvertent Startup of the HPCI Pump Modeling Considerations...................................................................................... 2-25 MECHANICAL DESIGN ANALYSIS.................................................................. 3-1 THERMAL-HYDRAULIC DESIGN ANALYSIS .................................................. 4-1 Thermal-Hydraulic Design and Compatibility .......................................... 4-1 Safety Limit MCPR Analysis ................................................................... 4-1 Core Hydrodynamic Stability................................................................... 4-2 Voiding in the Channel Bypass Region................................................... 4-2 ANTICIPATED OPERATIONAL OCCURRENCES ........................................... 5-1 System Transients .................................................................................. 5-2 Load Rejection No Bypass (LRNB) .............................................. 5-4 Turbine Trip No Bypass (TTNB)................................................... 5-4 Feedwater Controller Failure (FWCF) .......................................... 5-5 Loss of Feedwater Heating .......................................................... 5-6 Control Rod Withdrawal Error ...................................................... 5-6 Inadvertent HPCI Pump Start....................................................... 5-7 Two Loop Pump Seizure.............................................................. 5-8 Slow Flow Runup Analysis...................................................................... 5-8 Equipment Out-of-Service Scenarios...................................................... 5-9 TBVOOS ...................................................................................... 5-9 FHOOS ...................................................................................... 5-10 PLUOOS .................................................................................... 5-10 Combined TBVOOS and FHOOS .............................................. 5-11 Combined TBVOOS and PLUOOS ............................................ 5-11 Combined FHOOS and PLUOOS .............................................. 5-11 Combined TBVOOS, FHOOS, and PLUOOS ............................ 5-11 Reduced Feedwater Temperature at Startup............................. 5-11 Recirculation Pump Out-of-Service ............................................ 5-12 Licensing Compliance ........................................................................... 5-13 Axial Exposure Ratio.................................................................. 5-13

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page iii Licensing Power Shape ............................................................. 5-13 POSTULATED ACCIDENTS ............................................................................. 6-1 Loss-of-Coolant-Accident (LOCA) .......................................................... 6-1 Control Rod Drop Accident (CRDA)........................................................ 6-1 Fuel and Equipment Handling Accident .................................................. 6-2 Fuel Loading Error (Infrequent Event) .................................................... 6-2 Mislocated Fuel Bundle................................................................ 6-3 Misoriented Fuel Bundle .............................................................. 6-3 SPECIAL ANALYSES ....................................................................................... 7-1 ASME Overpressurization Analysis ........................................................ 7-1 ATWS Event Evaluation.......................................................................... 7-2 ATWS Overpressurization Analysis ............................................. 7-2 Long-Term Evaluation.................................................................. 7-3 Standby Liquid Control System............................................................... 7-3 Fuel Criticality ......................................................................................... 7-3 OPERATING LIMITS AND COLR INPUT.......................................................... 8-1 MCPR Limits........................................................................................... 8-1 LHGR Limits ........................................................................................... 8-2 MAPLHGR Limits.................................................................................... 8-3 REFERENCES .................................................................................................. 9-1

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page iv Tables Table 1.1 EOD and EOOS Operating Conditions ...................................................... 1-2 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry ............................................................................................. 2-3 Table 2.2 Disposition of Operating Flexibility and EOOS Options on Limiting Events ..................................................................................................... 2-11 Table 2.3 Plant Parameter Sensitivity Results for MCPR ..................................... 2-15 Table 2.4 Plant Parameter Sensitivity Results for Transient Nodal Power .............. 2-19 Table 2.5 Plant Parameter Sensitivity Results for Overpressurization ................... 2-23 Table 2.6 Lower Plenum Mixing Fractions............................................................... 2-27 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses .................................................................................................... 4-4 Table 4.2 Results Summary for Safety Limit MCPR Analyses .................................. 4-5 Table 4.3 BSP Endpoints for Nominal Feedwater Temperature ................................ 4-6 Table 4.4 BSP Endpoints for Reduced Feedwater Temperature............................... 4-6 Table 4.5 Nominal Feedwater Temperature Boundary Points ................................... 4-7 Table 4.6 Reduced Feedwater Temperature Boundary Points.................................. 4-8 Table 4.7 ABSP Setpoints for the Scram Region ...................................................... 4-9 Table 5.1 Exposure Basis for Transient Analysis .................................................... 5-15 Table 5.2 Scram Speed Insertion Times ................................................................. 5-16 Table 5.3 Base Case Limiting Transient Event NSS Insertion Time........................ 5-17 Table 5.4 Base Case Limiting Transient Event TSSS Insertion Time...................... 5-18 Table 5.5 Loss of Feedwater Heating Transient Analysis Results........................... 5-19 Table 5.6 Control Rod Withdrawal Error CPR Results .......................................... 5-20 Table 5.7 RBM Operability Requirements ............................................................... 5-20 Table 5.8 Flow-Dependent MCPR Results .............................................................. 5-21 Table 5.9 TLO and SLO Pump Seizure Results ...................................................... 5-21 Table 5.10 ATRIUM 11 LHGRFACp Transient Results ............................................. 5-22 Table 5.11 ATRIUM 10XM LHGRFACp Transient Results ....................................... 5-23 Table 5.12 Licensing Basis Core Average Axial Power Profile................................. 5-24 Table 7.1 ASME Overpressurization Analysis Results .............................................. 7-4 Table 7.2 ATWS Overpressurization Analysis Results .............................................. 7-5

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page v Table 8.1 TLO MCPRp Limits for OSS Insertion Times BOC to NEOC ..................... 8-4 Table 8.2 TLO MCPRp Limits for NSS Insertion Times BOC to NEOC ..................... 8-5 Table 8.3 TLO MCPRp Limits for TSSS Insertion Times BOC to NEOC ................... 8-8 Table 8.4 TLO MCPRp Limits for OSS Insertion Times NEOC to End of Coast ...... 8-11 Table 8.5 TLO MCPRp Limits for NSS Insertion Times NEOC to End of Coast ...... 8-12 Table 8.6 TLO MCPRp Limits for TSSS Insertion Times NEOC to End of Coast .... 8-15 Table 8.7 MCPRf Limits ........................................................................................... 8-18 Table 8.8 Steady-State LHGR Limits ...................................................................... 8-18 Table 8.9 LHGRFACp Multipliers ............................................................................. 8-19 Table 8.10 LHGRFACf Multipliers.............................................................................. 8-20 Table 8.11 MAPLHGR Limits .................................................................................... 8-20

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page vi Figures Figure 1.1 Browns Ferry Power / Flow Map................................................................ 1-3 Figure 5.1 Limiting LRNB at 100P / 105F - TSSS Key Parameters ......................... 5-25 Figure 5.2 Limiting LRNB at 100P / 105F - TSSS Sensed Water Level................... 5-26 Figure 5.3 Limiting LRNB at 100P / 105F - TSSS Vessel Pressures ....................... 5-27 Figure 5.4 Limiting TTNB at 100P / 105F - TSSS Key Parameters ......................... 5-28 Figure 5.5 Limiting TTNB at 100P / 105F - TSSS Sensed Water Level ................... 5-29 Figure 5.6 Limiting TTNB at 100P / 105F - TSSS Vessel Pressures ...................... 5-30 Figure 5.7 Limiting FWCF at 100P / 105F - TSSS Key Parameters ........................ 5-31 Figure 5.8 Limiting FWCF at 100P / 105F - TSSS Sensed Water Level .................. 5-32 Figure 5.9 Limiting FWCF at 100P / 105F - TSSS Vessel Pressures ...................... 5-33 Figure 5.10 Limiting IHPS at 100P / 105F - TSSS Key Parameters .......................... 5-34 Figure 5.11 Limiting IHPS at 100P / 105F - TSSS Sensed Water Level .................... 5-35 Figure 5.12 Limiting IHPS at 100P / 105F - TSSS Vessel Pressures ....................... 5-36 Figure 7.1 ASME-MSIV Overpressurization Event at 102P / 105F - Key Parameters ................................................................................................ 7-6 Figure 7.2 ASME-MSIV Overpressurization Event at 102P / 105F - Sensed Water Level ............................................................................................... 7-7 Figure 7.3 ASME-MSIV Overpressurization Event at 102P / 105F - Vessel Pressures .................................................................................................. 7-8 Figure 7.4 ASME-MSIV Overpressurization Event at 102P / 105F - Safety /

Relief Valve Flow Rates ............................................................................ 7-9 Figure 7.5 ATWS-MSIV Overpressurization Event at 100P / 85F - Key Parameters .............................................................................................. 7-10 Figure 7.6 ATWS-MSIV Overpressurization Event at 100P / 85F - Sensed Water Level ............................................................................................. 7-11 Figure 7.7 ATWS-MSIV Overpressurization Event at 100P / 85F - Vessel Pressures ................................................................................................ 7-12 Figure 7.8 ATWS-MSIV Overpressurization Event at 100P / 85F - Safety /

Relief Valve Flow Rates .......................................................................... 7-13 Figure 8.1 [

] ................................................ 8-21 Figure 8.2 [

]................................................................ 8-22

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page vii Nomenclature ABSP automated backup stability ADS automatic depressurization system AER axial exposure ratio AOO anticipated operational occurrence AOT abnormal operational transient APRM average power range monitor ARO all control rods out ASME American Society of Mechanical Engineers AST alternate source term ATWS anticipated transient without scram ATWS-RPT anticipated transient without scram recirculation pump trip BFN Browns Ferry Nuclear Plant BOC beginning-of-cycle BPWS banked position withdrawal sequence BSP backup stability protection BWR boiling water reactor CAD containment atmospheric dilution CDA confirmation density algorithm CFD computational fluid dynamics CFR Code of Federal Regulations COLR core operating limits report CPR critical power ratio CRDA control rod drop accident CRWE control rod withdrawal error EFPD effective full-power days EFPH effective full-power hours EFPY effective full-power years EM evaluation model EOC end-of-cycle EOCLB end-of-cycle licensing basis EOC-RPT end-of-cycle recirculation pump trip EOC-RPT-OOS end-of-cycle recirculation pump trip out-of-service EOD extended operating domain EOFP end of full power EOOS equipment out-of-service EPU extended power uprate defined as 120 % original licensed thermal power FFTR final feedwater temperature reduction FHOOS feedwater heaters out-of-service FoM figure of merit FSAR final safety analysis report FW feedwater FWCF feedwater controller failure

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page viii Nomenclature (Continued)

GSF generic shape function HPCI high pressure coolant injection ICF increased core flow IHPS inadvertent HPCI pump start LFWH loss of feedwater heating LHGR linear heat generation rate LHGRFACf flow-dependent linear heat generation rate multipliers LHGRFACp power-dependent linear heat generation rate multipliers LOCA loss-of-coolant accident LPRM local power range monitor LRNB generator load rejection with no bypass MAPFAC maximum average planar multipliers MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MCPRf flow-dependent minimum critical power ratio MCPRp power-dependent minimum critical power ratio MELLLA maximum extended load line limit analysis MELLLA+ maximum extended load line limit analysis plus MSIV main steam isolation valve MSRV main steam relief valve MSRVOOS main steam relief valve out-of-service NCL natural circulation line NEOC near end-of-cycle NSS nominal scram speed NRC Nuclear Regulatory Commission, U.S.

OPRM oscillation power range monitor OSS optimum scram speed Pbypass power below which direct scram on TSV / TCV closure is bypassed PCT peak cladding temperature PLU power load unbalance PLUOOS power load unbalance out-of-service PRFO pressure regulator failure open RBM (control) rod block monitor RCPOOS recirculation pump out-of-service RDF rated drive flow RHR residual heat removal RPT recirculation pump trip RTP rated thermal power

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page ix Nomenclature (Continued)

SE safety evaluation SLC standby liquid control SLCS standby liquid control system SLMCPR safety limit minimum critical power ratio SLO single-loop operation SRV safety/relief valve STP simulated thermal power TBV turbine bypass valve TBVIS turbine bypass valves in service TBVOOS turbine bypass valves out-of-service TCV turbine control valve TIP traversing incore probe TIPOOS traversing incore probe out-of-service TLO two-loop operation TSSS technical specifications scram speed TSV turbine stop valve TTNB turbine trip with no bypass TVA Tennessee Valley Authority CPR change in critical power ratio MCPR change in minimum critical power ratio

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 1-1 INTRODUCTION Reload licensing analyses results generated by Framatome Inc. (Framatome) are presented in support of Browns Ferry Unit 2 Cycle 23. The analyses reported in this document were performed using methodologies currently under review (Reference 1) by the U. S. Nuclear Regulatory Commission (NRC) for application to Browns Ferry Nuclear Plant (BFN). The technical limitations and conditions associated with the application of the approved methodologies have been satisfied by these analyses.

The Cycle 23 core consists of a total of 764 fuel assemblies including 332 fresh ATRIUM 11 assemblies and 432 irradiated ATRIUM 10XM assemblies. The licensing analyses support the core design presented in Reference 2 and the use of the maximum extended load line limit analysis plus (MELLLA+) operating domain.

The Cycle 23 reload licensing analyses were performed for potentially limiting events and analyses identified in Section 2.0. The results of analyses are used to establish the Technical Specifications / core operating limits report (COLR) limits and ensure design and licensing criteria are met. The design and safety analyses are based on both operational assumptions and plant parameters provided by the utility. The results of the reload licensing analyses support operation for the power / flow map presented in Figure 1.1 and also support operation with the equipment out-of-service (EOOS) scenarios presented in Table 1.1.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 1-2 Table 1.1 EOD and EOOS Operating Conditions Extended Operating Domain (EOD) Conditions Increased core flow (ICF)

Maximum extended load line limit analysis plus (MELLLA+)

Combined final feedwater temperature reduction (FFTR) / coastdown Equipment Out-of-Service (EOOS) Conditions*

Turbine bypass valves out-of-service (TBVOOS)

Feedwater heaters out-of-service (FHOOS)

Power load unbalance out-of-service (PLUOOS)

Combined TBVOOS and FHOOS Combined TBVOOS and PLUOOS Combined FHOOS and PLUOOS Combined TBVOOS, FHOOS, and PLUOOS Recirculation pump out-of-service (RCPOOS),

• Base case and each EOOS condition are supported in combination with 1 main steam relief valve out-of-service (MSRVOOS), end-of-cycle recirculation pump trip out-of-service (EOC-RPT-OOS), up to 18 traversing incore probe (TIP) channels out-of-service (TIPOOS) (per operating requirements defined in Section 4.2), and up to 50 % of the local power range monitor (LPRM) out-of-service.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

RCPOOS is the EOOS implying single loop operation (SLO). Operation in single loop is only supported up to a maximum core flow of 50 % of rated, a maximum power level of 43.75 % of rated, and an active recirculation drive flow of 17.73 Mlb/hr.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 1-3 Figure 1.1 Browns Ferry Power / Flow Map

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-1 DISPOSITION OF EVENTS AND PLANT MODELING SENSITIVITIES Disp osition of Events for A T RI U M 1 1 Fuel I ntroduction A disposition of events to identify the limiting events which need to be analyzed to support operation at the Browns Ferry Nuclear Plant with extended power uprate (EPU) MELLLA+

conditions was performed for the introduction of ATRIUM 11 fuel. Events and analyses identified as potentially limiting were either evaluated generically for the introduction of ATRIUM 11 fuel or are performed on a cycle-specific basis.

The first step in the disposition of events is to identify the licensing basis of the plant. Included in the licensing basis are descriptions of the postulated events/analyses and the associated criteria. Fuel-related system design criteria must be met, ensuring regulatory compliance and safe operation. The BFN licensing basis is contained in the Final Safety Analysis Report (FSAR), the Technical Specifications, COLR, and other reload analysis reports.

Framatome reviewed all the fuel-related design criteria, events, and analyses identified in the licensing basis. In many cases, when the operating limits are established to ensure acceptable consequences of an abnormal operational transient (AOT) or accident, the fuel-related aspects of the system design criteria are met. All the fuel-related events were reviewed and dispositioned into one of the following categories:

• N o further analysis req uired. This classification may result from one of the following:

The consequences of the event are bound by consequences of a different event.

The consequences of the event are benign, i.e., the event causes no significant change in margins to the operating limits.

The event is not affected by the introduction of a new fuel design and/or the current analysis of record remains applicable.

• A ddress event each reload. The consequences of the event are potentially limiting and need to be addressed each reload.

• A ddress for initial reload. This classification may result from one of the following:

The analysis is performed using conservative bounding assumptions and inputs such that the initial reload results will remain applicable for future reloads of the same fuel design.

Results from the first reload will be used to quantitatively demonstrate that the results remain applicable for future reloads of the same fuel design because the consequences are benign or bound by those of another event.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-2 The impacts of the EOOS scenarios presented in Table 1.1 were also considered in the disposition of events.

A summary of the disposition of events is presented in Tables 2.1 and 2.2. Table 2.1 presents a list of the events and analyses, the corresponding FSAR section, the disposition status, and any applicable comments. Table 2.2 presents a summary of the disposition of events for the EOOS scenarios. Note that operation in the ICF and MELLLA+ regions of the power/flow map are included in the disposition results presented in Table 2.1.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-3 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry FSAR Section Event /Analysis Disposition Status Comments 3.2 Fuel mechanical Address initial reload Cycle specific analysis (results and design analyses generally do not change from cycle-to-cycle unless a design feature is modified).

Refer to Reference 3 for the analysis, acceptance criteria, methodology and evaluation model.

Demonstrate design criteria are met.

3.6 Nuclear design Address each reload Cycle specific analysis.

Refer to Reference 2 for the analysis, acceptance criteria, methodology and evaluation model.

Demonstrate design criteria are met.

3.7 Thermal and Address each reload Plant specific and cycle specific analysis.

hydraulic design Demonstrate design criteria are met. Fuel hydraulic design and compatibility results are provided in the Thermal-Hydraulic Design report. Refer to Reference 4 for the analysis, acceptance criteria, methodology, and evaluation model.

Other cycle specific criteria are presented in this report, i.e., thermal operating limits.

3.8 Standby liquid Address each reload Cycle specific analysis.

control system Analysis performed each reload to verify (SLCS) adequate SLCS shutdown capacity.

See Section 7.3.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-4 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry (Continued)

FSAR Section Event /Analysis Disposition Status Comments 4.2 Reactor vessel and No further analyses FSAR analysis.

appurtenances required The vessel fluence irradiation is primarily mechanical design dependent upon the effective full power years (EFPY), power distribution, power level, and fuel management scheme. The neutron spectrum of ATRIUM 11 is sufficiently similar to the spectrum applied in the current licensing basis evaluation of the vessel irradiation limits addressed in FSAR 4.2.5. Therefore, no additional vessel irradiation limit analyses are required for the introduction of ATRIUM 11 fuel (Section 8.4 of Reference 11).

4.4 Nuclear system Address each reload Cycle specific analysis (overpresurization),

pressure relief plant specific analysis (LOCA).

system Analysis of limiting overpressurization events required each reload.

Evaluations of the automatic depressurization system (ADS) capability are addressed as part of the loss-of-coolant accident (LOCA) analyses (Reference 5).

5.2 Primary No further analyses FSAR analysis.

containment required Except for the containment atmospheric system dilution (CAD) evaluation, the primary containment characteristics following a postulated LOCA are not fuel related. The CAD system criteria are met for ATRIUM 11 (Reference 6).

5.3 Secondary No further analyses The secondary containment basis is Containment required independent of fuel design.

System

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-5 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry (Continued)

FSAR Section Event /Analysis Disposition Status Comments 6.0 Emergency core Address each reload Plant specific analysis and cycle specific cooling systems analysis.

LOCA is a potentially limiting accident.

Refer to Reference 5 for the analysis, acceptance criteria, methodology, and evaluation model. Limiting break characteristics and exposure study results are identified for ATRIUM 11 reload fuel.

Confirmation of bounding limiting power history and gadolinia LHGR is performed for follow-on reloads to confirm the applicability of Reference 5.

7.5 Neutron monitoring Address each reload Plant specific and cycle specific analysis.

system Rod block monitor (RBM) setpoints evaluated for the control rod withdrawal error (CRWE) event. Cycle specific Best-estimate Enhanced Option III (BEO-III) with the Confirmation Density Algorithm (CDA) stability analysis is performed to confirm the CDA setpoints provide adequate protection and define the cycle-specific backup stability protection (BSP) regions.

7.19 Anticipated Address each reload Cycle specific analysis.

transient without Analyses are performed to demonstrate scram that the peak vessel pressure for the (ATWS) limiting ATWS event is less than 120% of design pressure.

Peak cladding temperature (PCT) and local cladding oxidation are bound by LOCA.

Standby liquid control system shutdown capacity is verified each cycle (see Section 7.3)

Long term ATWS analyses remain applicable for Framatome fuel (Section 7.2.2).

ATWS with core instability is addressed in Reference 7 for the introduction of ATRIUM 11 fuel. For ATRIUM 10XM fuel, the acceptability has been assessed as satisfactory in Reference 8, Section 3.9.3.3.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-6 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry (Continued)

FSAR Section Event /Analysis Disposition Status Comments 8.10 Station blackout No further analyses FSAR analysis.

required The licensing basis analysis remains applicable to Framatome fuel since station blackout is solely driven by decay heat. All other criteria are not fuel dependent.

Framatome fuel is designed to perform in a manner similar to and analogous with fuel of current and previous designs.

10.2 New fuel storage Not applicable Fuel assemblies will not be stored in the new fuel storage vault (Technical Specification 4.3.1.2).

10.3 Spent fuel storage Address each reload Plant specific analysis.

Refer to Reference 9 for the analysis, acceptance criteria, methodology, and evaluation model.

Evaluated for spent fuel storage racks.

Confirm applicability each reload (Section 7.4).

10.11 Fire protection Address initial reload Plant specific analysis.

systems NFPA 805 fire protection acceptance criteria are met for ATRIUM 11 fuel.

14.5.2.1 Generator trip (TCV No further analyses Bound by the generator trip with turbine fast closure) required bypass valve failure (LRNB).

14.5.2.2 Generator trip (TCV Address each reload Cycle specific analysis.

fast closure) with This event is a potentially limiting AOT.

turbine bypass valve failure 14.5.2.2.4 LRNB with EOC- Address each reload Cycle specific analysis.

RPT-OOS This event is a potentially limiting AOT.

14.5.2.3 Loss of condenser No further analyses FSAR analysis.

vacuum required Transient is equivalent to the turbine trip with bypass operable; therefore, the event is bound by the turbine trip with turbine bypass valve failure.

14.5.2.4 Turbine trip (TSV No further analyses Bound by the turbine trip with turbine closure) required bypass valve failure.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-7 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry (Continued)

FSAR Section Event /Analysis Disposition Status Comments 14.5.2.5 Turbine bypass Address initial reload Cycle specific analysis, for initial reload.

valves failure Generally bound by the generator trip with following turbine turbine bypass valve failure. Comparison trip (TTNB), high is provided in Section 5.

power 14.5.2.6 Turbine bypass No further analyses FSAR analysis.

valves failure required Generally bound by the generator trip with following turbine turbine bypass valve failure.

trip (TTNB), low power 14.5.2.7 Main steam No further analyses FSAR analysis.

isolation valve required The MSIV closure event is bound by the (MSIV) closure LRNB event for thermal operating limits.

However, the MSIV closure is evaluated for overpressure limits (Sections 7.1 and 7.2).

14.5.2.8 Pressure regulator No further analyses FSAR analysis.

failure (downscale) required Eliminated as an AOT by the installation of a digital fault-tolerant main turbine electro-hydraulic control system.

14.5.3.1 Loss of feedwater Address each reload Cycle specific analysis.

heater (LFWH)

Generally bound by the feedwater controller failure (FWCF) event.

Addressed each cycle to demonstrate that it remains bound by the other events.

14.5.3.2 Shutdown cooling No further analyses FSAR analysis.

(RHR) malfunction required Benign event.

- decreasing temperature 14.5.3.3 Inadvertent HPCI Address each reload See Section 5.1.6.

pump start For minimum critical power ratio (MCPR),

(IHPS) the event was determined to be non-limiting, bound by the FWCF event, for the initial reload.

However, for thermal-mechanical linear heat generation rate (LHGR) margin, the event is potentially limiting and analyzed on a cycle-specific basis.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-8 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry (Continued)

FSAR Section Event /Analysis Disposition Status Comments 14.5.4.1 Continuous rod Address each reload Cycle specific analysis.

withdrawal during This event is a potentially limiting AOT.

power range operation 14.5.4.2 Continuous rod No further analyses FSAR analysis.

withdrawal during required Benign event.

reactor startup 14.5.4.3 Control rod removal No further analyses FSAR analysis.

error during required This event is not credible.

refueling 14.5.4.4 Fuel assembly No further analyses FSAR analysis.

insertion error required An unplanned criticality during refueling during refueling due to a single fuel assembly insertion error is not credible.

Mislocated or Address each reload Cycle specific analysis.

misoriented fuel assembly 14.5.5.1 Pressure regulator Address each reload FSAR analysis and cycle specific analysis.

failure open The PRFO - maximum steam demand is a (PRFO) potentially limiting ATWS overpressurization event, which is addressed each reload.

Relative to AOT thermal operating limits, the PRFO is a benign event.

14.5.5.2 Inadvertent No further analysis FSAR analysis.

opening of a MSRV required Benign event.

14.5.5.3 Loss of feedwater No further analysis FSAR analysis.

flow required Benign event.

14.5.5.4 Loss of auxiliary No further analyses FSAR analysis.

power required Benign event.

14.5.6.1 Recirculation flow No further analysis FSAR analysis.

control failure - required Consequences of this event are bound by decreasing flow the pump seizure event.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-9 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry (Continued)

FSAR Section Event /Analysis Disposition Status Comments 14.5.6.2 Trip of one No further analyses FSAR analysis.

recirculation pump required Consequences of this event are benign and bound by the TTNB event.

14.5.6.3 Trip of two No further analyses FSAR analysis.

recirculation pumps required Consequences of this event are benign and bound by the TTNB event.

14.5.6.4 Recirculation pump Address each reload Cycle specific analysis.

seizure See Section 5.1.7 for two-loop operation (TLO) pump seizure and Section 5.3.9 for SLO pump seizure.

For both TLO and SLO pump seizure, the events are confirmed to be non-limiting for the initial cycle.

SLO is not allowed in the MELLLA+

operating domain.

14.5.7.1 Recirculation flow Address each reload Cycle specific analysis.

control failure -

Consequences of the slow flow run-up increasing flow event determine the flow dependent MCPR and LHGR operating limits and are evaluated each reload.

14.5.7.2 Startup of idle No further analysis FSAR analysis.

recirculation loop required Benign event.

14.5.8.1 Feedwater Address each reload Cycle specific analysis.

controller failure This event is a potentially limiting AOT.

(FWCF) - maximum demand 14.5.8.2 Feedwater Address each reload Cycle specific analysis.

controller failure This event is a potentially limiting AOT.

(FWCF) - maximum demand with EOC-RPT-OOS 14.5.8.3 Feedwater Address each reload Cycle specific analysis.

controller failure This event is a potentially limiting AOT.

(FWCF) - maximum demand with TBVOOS

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-10 Table 2.1 Disposition of Events Summary for ATRIUM 11 Fuel Introduction at Browns Ferry (Continued)

FSAR Section Event /Analysis Disposition Status Comments 14.5.9 Loss of habitability No further analyses FSAR analysis.

of the control room required This is postulated as a special event to demonstrate the ability to safely shutdown the reactor from outside the control room.

Not impacted by addition of ATRIUM 11 fuel design.

14.6.2 Control rod drop Address each reload Cycle specific analysis.

accident (CRDA)

Consequences of the CRDA are evaluated to confirm that the acceptance criteria are satisfied.

14.6.3 Loss-of-coolant Address each reload Plant specific analysis and cycle specific accident (LOCA) parameter confirmation.

The break spectrum analysis is addressed for the initial ATRIUM 11 reload, Reference 5. Consequences of the LOCA are evaluated to determine appropriate fuel-specific maximum average planar linear heat generation rate (MAPLHGR) limits which are independent of cycle-specific assembly designs. Limiting power history, gad LHGR confirmation, and MAPLHGR checks are performed for follow-on reloads.

14.6.4 Refueling accident Address each reload Plant specific analysis.

Consequences of the refueling accident are evaluated to confirm that the acceptance criteria are satisfied.

14.6.5 Main steam line No further analysis FSAR analysis and Reference 5.

break accident required The consequences of a large steam line break are far from limiting with respect to 10 CFR 50.46 acceptance criteria.

Radiological dose consequences are performed utilizing alternate source term (AST) in accordance with 10 CFR 50.67.

The consequences of the event are not a function of fuel type since no fuel failures are calculated to occur. The dose is a function of the radionuclide inventory in the coolant itself prior to the event.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-11 Table 2.2 Disposition of Operating Flexibility and EOOS Options on Limiting Events Affected Limiting Option Events/Analyses Comments One MSRV OOS ASME Overpressurization This scenario is included as part of the base case condition for the events/analyses FWCF identified.

LRNB TTNB ATWS RCPOOS

• LOCA The impact of SLO on LOCA is addressed in Safety limit minimum critical Section 8.

power ratio (SLMCPR) The SLO SLMCPR is addressed each reload.

FWCF The RCPOOS events are addressed each LRNB reload.

TTNB BEO-III is evaluated for single-loop Pump Seizure conditions for the transition reload.

BEO-III FFTR / FHOOS* FWCF This scenario is included in each reload for BEO-III / Backup Stability each of these events/analyses.

Protection (BSP)

TBVOOS FWCF The FWCF event with TBVOOS is evaluated each reload.

EOC-RPT OOS FWCF This scenario is included in each reload for each of these events/analyses.

LRNB TTNB PLUOOS LRNB The LRNB event with PLUOOS is evaluated each reload.

Traversing in-core probe SLMCPR TIPOOS is included in the SLMCPR (TIP) OOS analysis.

• Note that feedwater heaters out-of-service and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-12 P lant Sp ecific Modeling Sensitivities As part of the initial application of the AURORA-B AOO methodology to a plant, justification must be provided to ensure that conservative plant parameters are being used. This requirement is defined in Limitation and Conditions 7 and 11 of the Reference 10 safety evaluation. In particular, these limitations and conditions state:

7 . As discussed in Section 3 . 6 of this SE, licensees should provide j ustification for the k ey plant parameters and initial conditions selected for performing sensitivity analyses on an event-specific b asis. L icensees should further j ustify that the input values ultimately chosen for these k ey plant parameters and initial conditions w ill result in a conservative prediction of FoMs w hen performing calculations according to the AU RO RA-B EM describ ed in ANP-1 0 3 0 0 P.

1 1 . AREVA w ill provide j ustification for the uncertainties used for the highly rank ed plant-specific PIRT parameters C1 2 , R0 1 , R0 2 , and SL 0 2 on a plant-specific b asis, as describ ed in Tab le 3 . 2 of this SE.

In order to comply with these requirements, a set of sensitivity studies was performed.

Separate sensitivity studies were performed for each of the three figures of merit that were required to license Browns Ferry Unit 2 Cycle 23: change in minimum critical power ratio (MCPR) (Table 2.3), transient nodal power (Table 2.4), and overpressure (Table 2.5). These sensitivity studies address the key parameters required for licensing with the exception of C12 which is described below. In addition to these sensitivity studies, licensing calculations will also look at a wide range of core exposures and flow rates to ensure that the conservative statepoints have been analyzed.

Uncertainties associated with PIRT parameters R01, R02, and SL02 were evaluated for the initial transition. [

]

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 2-27 Table 2.6 Lower Plenum Mixing Fractions

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 3-1 MECHANICAL DESIGN ANALYSIS The mechanical design analyses for ATRIUM 10XM and ATRIUM 11 fuel assemblies are presented in the applicable mechanical design reports (References 3, 12, 13, and 14). The maximum exposure limits for the ATRIUM 10XM and ATRIUM 11 fuel designs are:

54.0 GWd/MTU average assembly exposure (ATRIUM 10XM) 57.0 GWd/MTU average assembly exposure (ATRIUM 11) 62.0 GWd/MTU rod average exposure (full-length fuel rods)

The maximum calculated rod oxide thickness for ATRIUM 11 fuel is presented in Tables 3-2 and 3-3 of Reference 13. The maximum calculated rod oxide thickness for ATRIUM 10XM fuel is presented in Tables 3-2 and 3-3 of Reference 14. The calculated oxide thickness complies with the approved limit provided in Reference 15.

The ATRIUM 10XM and ATRIUM 11 LHGR limits are presented in Section 8.0. The fuel cycle design analyses (Reference 2) have verified that the ATRIUM 10XM and ATRIUM 11 fuel assemblies remain within licensed burnup limits.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-1 THERMAL-HYDRAULIC DESIGN ANALYSIS T hermal- H ydraulic Design and C omp atib ility The results of thermal-hydraulic characterization and compatibility analyses are presented in the thermal-hydraulic design report (Reference 4). The analysis results demonstrate that the thermal-hydraulic design and compatibility criteria are satisfied for the Browns Ferry Unit 2 Cycle 23 transition core consisting of ATRIUM 10XM and ATRIUM 11 fuel assemblies.

Safety L imit MC P R A nalysis The SLMCPR is defined as the minimum value of the CPR ensuring less than 0.1 % of the fuel rods are expected to experience boiling transition during normal operation or an AOT. The SLMCPR for all fuel in the core was determined using the methodology described in Reference 16. The analysis was performed with a power distribution that conservatively represents expected reactor operation throughout the cycle.

The Browns Ferry Unit 2 Cycle 23 SLMCPR analysis used the ACE/ATRIUM 10XM critical power correlation, described in Reference 17, for the ATRIUM 10XM fuel. The ACE/ATRIUM 11 critical power correlation, described in Reference 18, was applied to the ATRIUM 11 fuel assemblies.

In the Framatome methodology, the effects of channel bow on the critical power performance are accounted for in the SLMCPR analysis. Reference 16 discusses the application of a realistic channel bow model. For this cycle, the channel bow model uncertainty has been augmented for those channels experiencing fluence gradients outside the bounds of the measurement database.

The fuel- and plant-related uncertainties used in the SLMCPR analysis are presented in Table 4.1. The radial power uncertainty used in the analysis includes the effects of up to 18 TIP channels out-of-service, up to 50 % of the LPRM out-of-service, and a 2,500 effective full-power hours (EFPH) LPRM calibration interval.

Analysis results support a TLO SLMCPR of 1.08 and a SLO SLMCPR of 1.09. Analysis results, including the SLMCPR and the percentage of rods expected to experience boiling transition, are summarized in Table 4.2.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-2 C ore H ydrodynamic Stab ility Browns Ferry Unit 2 will implement a plant specific application of the Best-estimate Enhanced Option III (BEO-III) with the Confirmation Density Algorithm (CDA) analysis methodology as described in Reference 19. The CDA enabled through the oscillation power range monitor (OPRM) system and the BSP solution described in Reference 19 will be the stability licensing basis for Browns Ferry. Cycle-specific analyses have been performed with RAMONA5-FA modeling recirculation pump trips from limiting MELLLA+, maximum extended load line limit analysis (MELLLA) with FHOOS and SLO statepoints. The LPRM traces for all statistical cases were analyzed with the CDA consistent with the Reference 19 methodology. The minimum required TLO and SLO stability operating limits are bounded by the MCPR limits provided in Section 8.1. All cases with a channel decay ratio greater than 1.0 within the 95/95 population were confirmed to meet the requirements of the Reference 19 methodology. The 95/95 statistical Tmin calculation was confirmed to be greater than 1.2 seconds for Browns Ferry Unit 2 Cycle 23.

The BSP solution may be used by the plant in the event that the OPRM system is declared inoperable. Reference 20 describes two BSP options based on selected elements from three distinct constituents: BSP Manual Regions, BSP Boundary, and Automated BSP (ABSP) setpoints.

The Manual BSP region boundaries were calculated for Browns Ferry Unit 2 Cycle 23 using STAIF (Reference 21) for nominal and reduced feedwater temperature operation (both FFTR and FHOOS). The endpoints of the regions are defined in Table 4.3 and Table 4.4 for nominal and reduced feedwater temperature, respectively. The Manual BSP region boundary endpoints are connected using the Generic Shape Function (GSF) and are provided with Table 4.5 and Table 4.6 for nominal and reduced feedwater temperature, respectively. The BSP Boundary for nominal and reduced feedwater temperature is defined by the MELLLA boundary line, per Reference 22. The ABSP Average Power Range Monitor (APRM) Simulated Thermal Power (STP) setpoints associated with the ABSP Scram Region are listed in Table 4.7. These ABSP setpoints are applicable to nominal and reduced feedwater temperature operation.

V oiding in the C hannel B yp ass Region To demonstrate compliance with the NRCs 5 % maximum bypass voiding around the LPRM requirement (see Section 5.1.1.5.1 of the Reference 23 Safety Evaluation), the bypass void

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-3 level has been evaluated throughout the cycle. The maximum bypass void value at the LPRM D level and at the axial elevation equivalent to the top of the TIP tube have been confirmed to remain below this limit for the Cycle 23 design.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-4 Table 4.1 Fuel- and Plant-Related Uncertainties for Safety Limit MCPR Analyses Parameter Uncertainty Fuel-Related Uncertainties Plant-Related Uncertainties Feedwater flow rate 1.8%

Feedwater temperature 0.8%

Core pressure 0.7%

Total core flow rate TLO 2.5%

SLO 6.0%

[ ]

[

]

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-5 Table 4.2 Results Summary for Safety Limit MCPR Analyses Minimum Percentage of Supported Rods in Boiling SLMCPR Transition TLO - 1.08 0.0994 SLO - 1.09 0.0693

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-6 Table 4.3 BSP Endpoints for Nominal Feedwater Temperature Power Flow Endpoint (%) (%) Definition Scram Region (Region I)

A1 75.9 52.7 Boundary Intercept on MELLLA+ Line Scram Region (Region I)

B1 35.5 29.0 Boundary Intercept on natural circulation line (NCL)

Controlled Entry Region A2 66.1 52.0 (Region II) Boundary Intercept on MELLLA Line Controlled Entry Region B2 25.5 29.0 (Region II) Boundary Intercept on NCL Table 4.4 BSP Endpoints for Reduced Feedwater Temperature Power Flow Endpoint (%) (%) Definition Scram Region (Region I)

A1 64.9 50.5 Boundary Intercept on MELLLA Line Scram Region (Region I)

B1 29.4 29.0 Boundary Intercept on NCL Controlled Entry Region A2 68.3 54.9 (Region II) Boundary Intercept on MELLLA Line Controlled Entry Region B2 24.5 29.0 (Region II) Boundary Intercept on NCL

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-7 Table 4.5 Nominal Feedwater Temperature Boundary Points Scram Region Exit Region Flow Power Flow Power

(% rated) (% rated) (% rated) (% rated) 52.70 75.90 52.00 66.10 51.52 71.76 50.85 61.62 50.33 67.98 49.70 57.57 49.15 64.52 48.55 53.92 47.96 61.35 47.40 50.63 46.78 58.45 46.25 47.64 45.59 55.79 45.10 44.94 44.41 53.36 43.95 42.50 43.22 51.13 42.80 40.28 42.04 49.08 41.65 38.27 40.85 47.20 40.50 36.45 39.67 45.49 39.35 34.79 38.48 43.92 38.20 33.29 37.30 42.48 37.05 31.94 36.11 41.17 35.90 30.70 34.93 39.98 34.75 29.59 33.74 38.89 33.60 28.59 32.56 37.90 32.45 27.68 31.37 37.02 31.30 26.87 30.19 36.22 30.15 26.15 29.00 35.50 29.00 25.50

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-8 Table 4.6 Reduced Feedwater Temperature Boundary Points Scram Region Exit Region Flow Power Flow Power

(% rated) (% rated) (% rated) (% rated) 50.50 64.90 54.90 68.30 49.43 61.22 53.61 63.33 48.35 57.86 52.31 58.87 47.28 54.79 51.02 54.86 46.20 51.99 49.72 51.26 45.13 49.43 48.43 48.01 44.05 47.09 47.13 45.09 42.98 44.95 45.84 42.46 41.90 43.00 44.54 40.08 40.83 41.20 43.25 37.93 39.75 39.56 41.95 35.99 38.68 38.07 40.66 34.23 37.60 36.70 39.36 32.65 36.53 35.45 38.07 31.21 35.45 34.31 36.77 29.92 34.38 33.27 35.48 28.76 33.30 32.33 34.18 27.71 32.23 31.48 32.89 26.77 31.15 30.71 31.59 25.92 30.08 30.02 30.30 25.17 29.00 29.40 29.00 24.50

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 4-9 Table 4.7 ABSP Setpoints for the Scram Region Parameter Symbol Setting Value (Unit) Comments Slope of ABSP APRM Slope for Trip mTRIP 2.00 (% RTP / % RDF) flow-biased trip linear segment.

ABSP APRM flow-biased trip setpoint power intercept.

Constant Power PBSP-TRIP 35.0 (% RTP) Constant Power Line for Trip Line for Trip from zero Drive Flow to Flow Breakpoint value.

ABSP APRM flow-biased trip Constant Flow setpoint drive flow intercept.

WBSP-TRIP 49 (% RDF)

Line for Trip Constant Flow Line for Trip.

(see Note 1)

Flow Breakpoint WBSP-BREAK 30.0 (% RDF) Flow Breakpoint value Note 1: WBSP-TRIP can be set to 49.0 % RDF or any higher value up to the intersection of the ABSP sloped line with the APRM flow-biased STP scram line.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-1 ANTICIPATED OPERATIONAL OCCURRENCES This section describes the analyses performed to determine the power- and flow-dependent MCPR operating limits (MCPRp and MCPRf) for base case operation.

The AURORA-B methodology (Reference 10) is used with the Framatome THERMEX methodology (Reference 24) for the generation of thermal limits. AURORA-B is a comprehensive evaluation model developed for predicting the dynamic response of boiling water reactors (BWRs) during transient, postulated accident, and beyond design-basis accident scenarios. The evaluation model (EM) contains a multi-physics code system with flexibility to incorporate all the necessary elements for analysis of the full spectrum of BWR events that are postulated to affect the nuclear steam supply system of the BWR plant. Deterministic analysis principles are applied to satisfy plant operational and Technical Specification requirements through the use of conservative initial conditions and boundary conditions.

The foundation of AURORA-B AOO is built upon three computer codes, S-RELAP5, MB2-K, and RODEX4. Working together as a system, they make up the multi-physics evaluation model that provides the necessary systems, components, geometries, processes, etc. to assure adequate predictions of the relevant BWR event characteristics for its intended applications.

The three codes making up the foundation of the code system are:

• S-RELAP5 - This code provides the transient thermal-hydraulic, thermal conduction, control systems, and special process capabilities (i.e. valves, jet-pumps, steam separator, critical power correlations, etc.) necessary to simulate a BWR plant.

• MB2-K - This code uses advanced nodal expansion methods to solve the three-dimensional, two-group, neutron kinetics equations. The MB2-K code is consistent with the MICROBURN-B2 steady state core simulator. MB2-K receives a significant portion of its input from the steady state core simulator.

• RODEX4 - A subset of routines from this code are used to evaluate the transient thermal-mechanical fuel rod (including fuel/clad gap) properties as a function of temperature, rod internal pressure, etc. The fuel rod properties are used by S-RELAP5 when solving the transient thermal conduction equations in lieu of standard S-RELAP5 material property tables.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-2 The AURORA-B AOO methodology (Reference 10) includes an evaluation of the impact of code uncertainties on Figures of Merit (FoM) (e.g. MCPR, peak pressure) [

] that has wide acceptance in the nuclear industry.

The ACE/ATRIUM 10XM critical power correlation (Reference 17) is used to evaluate the thermal margin for the ATRIUM 10XM fuel. The ACE/ATRIUM 11 critical power correlation (Reference 18) is used in the thermal margin evaluations for the ATRIUM 11 fuel.

System T ransients The reactor plant parameters for the system transient analyses were provided by the utility.

Analyses have been performed to determine MCPRp limits protecting operation throughout the power / flow domain depicted in Figure 1.1.

At Browns Ferry, direct scram on turbine stop valve (TSV) position and turbine control valve (TCV) fast closure are bypassed at power levels less than 26 % of rated (Pbypass). Below Pbypass scram occurs when either the high pressure or high neutron flux scram setpoint is reached. MCPR limits are monitored at power levels greater than or equal to 23 % of rated, which is the lowest power analyzed for this report, consistent with Reference 25.

The limiting exposure for rated power pressurization transients is typically at end of full power (EOFP) when the control rods are fully withdrawn. To provide additional margin to the operating limits earlier in the cycle, analyses were also performed to establish operating limits at a near end-of-cycle (NEOC) core average exposure. Analyses were performed at cycle exposures prior to NEOC to ensure the operating limits provide the necessary protection. The end-of-cycle licensing basis (EOCLB) analysis was performed at EOFP + 15 effective full power days (EFPD). Analyses were also performed to support extended cycle operation with FFTR and power coastdown. The licensing basis exposures used to develop the neutronics inputs to the transient analyses are presented in Table 5.1.

All pressurization transients assumed one of the lowest setpoint MSRV is inoperable. The basis supports operation with 1 MSRV out-of-service.

Reductions in feedwater temperature of less than 15 F from the nominal feedwater temperature and variations of +/- 10 psi in dome pressure are considered base case operation not an EOOS

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-3 condition. Although the base case operating condition assumes a maximum reduction of 15 F from the nominal feedwater temperature, the Browns Ferry Operating License does not allow operation at 100 % power in the MELLLA+ domain with final feedwater temperature less than 384.5 F. Analyses were performed to determine the limiting conditions in the allowable ranges.

FFTR is used to extend rated power operation by decreasing the feedwater temperature. The amount of feedwater temperature reduction is a function of power with the maximum decrease of 70 F (55 F + 15 F bias) at rated power. Analyses were performed to support combined FFTR / Coastdown operation to the core average exposure provided in Table 5.1. The analyses were performed with the limiting feedwater and dome pressure conditions in the allowable ranges. Operation with FFTR is not allowed in the MELLLA+ operating domain.

System pressurization transient results are sensitive to scram speed assumptions. To take advantage of average scram speeds faster than those associated with the Technical Specifications requirements, scram speed-dependent MCPRp limits are provided. The analytically adjusted timing for optimum scram speed (OSS) insertion times, nominal scram speed (NSS) insertion times, and the Technical Specifications scram speed (TSSS) insertion times used in the analyses are presented in Table 5.2 compared to the surveillance testing timing. The OSS and NSS MCPRp limits can only be applied if the scram speed test results meet the required insertion times. System transient analyses were performed to establish MCPRp limits for OSS, NSS, and TSSS insertion times.

The Technical Specifications (Reference 25) allow for operation with up to 13 slow and 1 stuck control rod. One additional control rod is assumed to fail to scram. The OSS, NSS, and TSSS analyses were performed to conservatively account for the effect of the slow and stuck rods on scram reactivity. For transient events below 26% power (below Pbypass) without direct scram, the results are relatively insensitive to scram speed, and only TSSS analyses are performed.

Thermal limits are typically based on the worst MCPR from the highest core flow and lowest core flow for a given core power. Therefore, transient analyses are performed for a wide range of core flows for a given power level. This range of transient statepoints support desired thermal limit development for plant operation. If less restrictive limits are required, then more than a single power dependent limit could be developed utilizing results from intermediate flows for a given power. See Section 8 for thermal limit development discussion.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-4 Tables 5.3 and 5.4 present the base case limiting transient event and results as a function of power used to generate the base case operating limits for NSS and TSSS insertion times, respectively.

Load Rej ection No Bypass (LRNB)

Load rejection causes a fast closure of the TCV. The TCV closure creates a pressure compression wave traveling through the steam lines into the vessel causing a rapid pressurization. The increase in pressure causes a decrease in core voids which in turn causes a rapid increase in power. Fast closure of the TCV also causes a reactor scram and recirculation pump trip (RPT). Turbine bypass system operation, which also mitigates the consequences of the event, is not credited. The excursion of the core power due to void collapse is terminated primarily by the reactor scram and revoiding of the core.

LRNB analyses assume the power load unbalance (PLU) is inoperable for power levels less than 50 % of rated. The LRNB sequence of events is different than the standard event when the PLU is inoperable. Instead of a fast closure, the TCV close in servo mode and there is no direct scram on TCV closure. The power and pressure excursion continues until the high pressure scram occurs.

LRNB analyses were performed for a range of power / flow conditions to support generation of the thermal limits. Responses of various reactor and plant parameters during the LRNB event initiated at 100 % of rated power and 105 % of rated core flow with TSSS insertion times are shown in Figure 5.1 - Figure 5.3.

Turbine Trip No Bypass (TTNB)

A turbine trip event can be initiated as a result of several different signals. The initiating signal causes the TSV to close in order to prevent damage to the turbine. The TSV closure creates a pressure compression wave traveling through the steam lines into the vessel causing a rapid pressurization. The increase in pressure causes a decrease in core voids which in turn causes a rapid increase in power. Closure of the TSV also causes a reactor scram and an RPT which helps mitigate the pressurization effects. Turbine bypass system operation, which also mitigates the consequences of the event, is not credited. The excursion of the core power due to void collapse is terminated primarily by the reactor scram and revoiding of the core.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-5 In addition to closing the TSV, a signal is also sent to close the TCV in fast mode. The consequences of a fast closure of the TCV are very similar to those resulting from a TSV closure.

The main difference is the time required to close the valves. While the TCV full stroke closure time is greater than the TSV (0.150 second compared to 0.100 second), the initial position of the TCV is dependent on the initial steam flow. At rated power and lower, the initial position of the TCV is such that the closure time is less than the TSV. However, the TCV closure characteristics are nonlinear such that the resulting core pressurization and MCPR may not always bound those of the slower TSV closure.

TTNB analyses were performed for a range of power / flow conditions to support generation of the thermal limits. Responses of various reactor and plant parameters during the TTNB event initiated at 100 % of rated power and 105 % of rated core flow with TSSS insertion times are shown in Figure 5.4 - Figure 5.6.

Feedwater Controller Failure (FW CF)

The increase in feedwater flow due to a failure of the feedwater control system to maximum demand results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at maximum demand, the water level continues to rise and eventually reaches the high water level trip setpoint. The initial water level is conservatively assumed to be at the low level normal operating range to delay the high-level trip and maximize the core inlet subcooling resulting from the FWCF. The high water level trip causes the TSV to close in order to prevent damage to the turbine from excessive liquid inventory in the steam line. Valve closure creates a pressure compression wave traveling back to the core causing void collapse and a subsequent rapid power excursion. The closure of the TSV also initiates a reactor scram and an RPT. In addition to the TSV closure, the TCV also close in the fast closure mode. Because of the partially closed initial position of the control valves, they will typically close faster than the stop valves and control the pressurization portion of the event. However, TCV closure characteristics are nonlinear such that the resulting core pressurization and MCPR results may not always bound those of the slower TSV closure at rated power (steam flow increases above rated before fast TCV closure). The limiting of TCV or TSV closure, for the initial operating conditions, was used in the FWCF analyses based on sensitivity analyses. The turbine bypass valves (TBV) are assumed operable and provide some pressure relief. The core power

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-6 excursion is mitigated in part by pressure relief; however, the primary mechanisms for termination of the event are reactor scram and revoiding of the core.

FWCF analyses were performed for a range of power / flow conditions to support generation of the thermal limits. Analyses performed at power levels greater than or equal to 65 % assume a maximum feedwater runout of 22.79 Mlbm/hr. For power levels equal to above Pbypass (26 %

power) up to 65 %, analyses assumed a maximum feedwater runout of 19.82 Mlbm/hr. For power levels below Pbypass, a maximum feedwater runout of 16.68 Mlbm/hr was assumed. A discussion of this input is provided in Comment 24 of Reference 26.

Figure 5.7 - Figure 5.9 show the responses of various reactor and plant parameters during the FWCF event initiated at 100 % of rated power and 105 % of rated core flow with TSSS insertion times.

Loss of Feedwater Heating The loss of feedwater heating (LFWH) event analysis supports an assumed 100 F decrease in the feedwater temperature. The result is an increase in core inlet subcooling which reduces voids thereby increasing core power and shifting the axial power distribution towards the bottom of the core. As a result of the axial power shift and increased core power, voids begin to build up in the bottom region of the core acting as negative feedback to the increased subcooling effect.

The negative feedback moderates the core power increase. Although there is a substantial increase in core thermal power during the event, the increase in steam flow is much less because a large part of the added power is used to overcome the increase in inlet subcooling.

The increase in steam flow is accommodated by the pressure control system via the TCV or turbine bypass valves (TBV), so no pressurization occurs. A cycle-specific analysis was performed in accordance with the Reference 27 methodology to determine the change in MCPR for the event. The LFWH results are presented in Table 5.5.

Control Rod W ithdrawal Error The control rod withdrawal error (CRWE) transient is an inadvertent reactor operator initiated withdrawal of a control rod. This withdrawal increases local power and core thermal power which lowers the core MCPR. The CRWE transient is typically terminated by control rod blocks initiated by the rod block monitor (RBM). The CRWE event was analyzed assuming no Xenon and allowing credible instrumentation out-of-service in the RBM system. The analysis further

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-7 assumes the plant could be operating in either an A or B sequence control rod pattern. The rated power CRWE results are shown in Table 5.6 for the analytical unfiltered RBM high power setpoint values of 107 % to 117 %. Analysis results indicate standard filtered RBM setpoint reductions are supported. Analyses demonstrate the 1 % strain and centerline melt criteria are met. LHGR limits and associated multipliers are presented in Section 8.2. Recommended operability requirements supporting unblocked CRWE operation are shown in Table 5.7 based on the SLMCPR values presented in Section 4.2.

Inadvertent HPCI Pump Start The inadvertent startup of the HPCI system results in the injection of cold water to the reactor vessel from the HPCI pump through the feedwater sparger. Injection of this subcooled water increases the subcooling at the inlet to the core and results in an increase in the core power.

The feedwater control system will attempt to control the water level in the reactor by reducing the feedwater flow. As long as the mass of steam leaving the reactor through the steam lines is more than the mass of HPCI water being injected, the water level will be controlled and a new steady-state condition will be established. In the scenario when HPCI flow becomes more than the steam flow, water level can increase until the HPCI pump trip is reached, thereby shutting of HPCI flow. In this case, a L8 trip is avoided.

The HPCI flow in the Browns Ferry units is only injected into one of the two feedwater lines and thus through the feedwater sparger on only one side of the reactor vessel, resulting in an asymmetric flow distribution of the injected HPCI flow. This asymmetric injection of the HPCI flow may cause an asymmetric core inlet enthalpy distribution and a larger enthalpy decrease for part of the core. [

]

The IHPS analyses were performed for a range of power and flow conditions to support generation of the base case operating limits for realistic and maximum allowable average scram insertion times. Figure 5.10 - Figure 5.12 present the responses of various reactor and plant parameters for IHPS transient initiated at rated core power.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-8 Two Loop Pump Seiz ure FSAR Section 14.5.6.4 addresses the one recirculation pump seizure event in two-loop operation. The event assumes an instantaneous seizure of one of the recirculation pump motor shafts. Flow through the affected loop is rapidly reduced due to the large hydraulic resistance introduced by the stopped rotor. This causes the core thermal power to decrease and reactor water level to swell. The sudden decrease in core coolant flow while the reactor is at power results in a degradation of core heat transfer to cause fuel damage.

The two-loop pump seizure analysis was performed for rated power and increased core flow to support generation of the base case operating limits for realistic and maximum allowable average scram insertion times. The results for the TLO pump seizure event are provided in Table 5.9. A comparison to the limiting MCPRs provided in Table 5.3 demonstrate TLO pump seizure is a non-limiting transient.

Slow Flow Runup A nalysis Flow-dependent MCPR limits (MCPRf) and LHGR multipliers (LHGRFACf) are established to support operation at off-rated core flow conditions. Limits are based on the CPR and heat flux changes experienced by the fuel during slow flow excursions. The slow flow excursion event assumes recirculation flow control system failure such that core flow increases slowly to the maximum flow physically attainable by the equipment (107 % of rated core flow). An uncontrolled increase in flow creates the potential for a significant increase in core power and heat flux. A conservatively steep flow runup path was used in the analysis. Evaluations were performed to support operation in all the EOOS scenarios.

A steady-state hydraulic model, using bounding statepoint assumptions, is used to calculate the change in CPR during a two-loop flow runup to the maximum flow rate. The MCPRf limit is set so an increase in core power, resulting from the maximum increase in core flow, assures the TLO SLMCPR is not violated. Calculations were performed over a range of initial flow rates to determine the corresponding MCPR values causing the limiting assembly to be at the SLMCPR for the high flow condition at the end of the flow excursion.

Analysis results are presented in Table 5.8. MCPRf limits providing the required protection are presented in Table 8.7. MCPRf limits are applicable for all exposures.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-9 Flow runup analyses were performed to determine LHGRFACf multipliers. The analysis assumes recirculation flow increases slowly along the limiting rod line to the maximum flow physically attainable by the equipment. A series of flow excursion analyses were performed at several exposures throughout the cycle starting from different initial power / flow conditions.

Xenon is assumed to remain constant during the event. LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a flow runup. LHGRFACf multipliers are presented in Table 8.10.

The maximum flow during a flow excursion in SLO is much less than the maximum flow during TLO. Therefore, the MCPRf limits and LHGRFACf multipliers for TLO are applicable for SLO.

Eq uip ment Out- of- Service Scenarios The EOOS scenarios supported are shown in Table 1.1. As noted in Table 1.1, base case and each EOOS condition is supported in combination with 1 MSRVOOS, EOC-RPT-OOS, up to 18 TIP channels out-of-service (per operating requirements defined in Section 4.2), and up to 50 % of the LPRM out-of-service.

When EOC-RPT is inoperable, no credit is assumed for RPT on TSV position or TCV fast closure. The function of the EOC-RPT feature is to reduce the severity of the core power excursion caused by the pressurization transient. The RPT accomplishes this by helping revoid the core thereby reducing the magnitude of the reactivity insertion resulting from the pressurization transient. Failure of the RPT feature can result in higher operating limits.

Analyses were performed for LRNB and FWCF events assuming EOC-RPT-OOS.

The analyses presented in this section also include these EOOS conditions protected by the base case limits. No further discussion for these EOOS conditions is presented in this section.

Base thermal limits presented in Section 8.0 are applicable with or without function of the EOC-RPT.

Tables 5.10 and 5.11 presents the limiting LHGRFACp transient analysis results for each EOOS scenario used to generate the operating limits for all scram insertion times.

TBVOOS The effect of operation with TBVOOS is a reduction in the system pressure relief capacity which makes the pressurization events more severe. While the base case LRNB and TTNB events are

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-10 analyzed assuming the TBV out-of-service, operation with TBVOOS has an adverse effect on the FWCF event. Analyses of the FWCF event with TBVOOS were performed to establish the TBVOOS operating limits.

FHOOS The FHOOS scenario assumes a feedwater temperature reduction of 70 F (55 F + 15 F bias) at rated power and steam flow. The effect of reduced feedwater temperature is an increase in core inlet subcooling changing the axial power shape and core void fraction. Additionally, steam flow for a given power level decreases because more power is required to increase coolant enthalpy to saturated conditions. Generally, LRNB and TTNB events are less severe with FHOOS conditions due to the decrease in steam flow relative to nominal conditions. FWCF events with FHOOS conditions are generally worse due to a larger change in inlet subcooling and core power prior to the pressurization phase of the event.

Separate FHOOS limits are not needed for operation beyond the EOCLB exposure since a feedwater (FW) temperature reduction is included to attain the additional cycle extension to the FFTR / coastdown exposure, i.e., FFTR is equivalent to FHOOS since both are based on the same feedwater temperature reduction.

PLUOOS The PLU device in normal operation is assumed to not function below 50 % power. PLUOOS is assumed to mean the PLU device does not function for any power level and does not initiate fast TCV closure. The following PLUOOS scenario was assumed for the load reject event.

• Initially, the TCV remain in pressure / speed control mode. There is no direct scram or EOC-RPT on valve motion.

• Loss of load results in increasing turbine speed. Depending on initial power, a turbine overspeed condition may be reached to initiate a turbine trip resulting in scram and EOC-RPT.

• Without a turbine trip signal, scram occurs on either high flux or high dome pressure to terminate the event.

Analyses were performed for LRNB events assuming PLUOOS.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-11 Combined TBVOOS and FHOOS FWCF analyses with both TBVOOS and FHOOS were performed. Operating limits for this combined EOOS scenario were established using these FWCF results and results previously discussed. Separate TBVOOS and FHOOS combined limits are not needed for operation beyond the EOCLB exposure since a FW temperature reduction is included to attain the additional cycle extension to the FFTR / coastdown exposure.

Combined TBVOOS and PLUOOS Limits were established to support operation with both TBVOOS and PLUOOS. No additional analyses are required to construct MCPRp operating limits for TBVOOS and PLUOOS since TBVOOS and PLUOOS are independent EOOS conditions (TBVOOS only impacts FWCF events; PLUOOS only impacts LRNB events).

Combined FHOOS and PLUOOS LRNB analyses with both FHOOS and PLUOOS were performed. Operating limits for this combined EOOS scenario were established using these LRNB results and results previously discussed. Separate FHOOS and PLUOOS combined limits are not needed for operation beyond the EOCLB exposure since a FW temperature reduction is included to attain the additional cycle extension to the FFTR / coastdown exposure.

Combined TBVOOS, FHOOS, and PLUOOS Limits were established to support operation with TBVOOS, FHOOS, and PLUOOS. No additional analyses are required to construct MCPRp operating limits for TBVOOS, FHOOS, and PLUOOS since TBVOOS and PLUOOS are independent EOOS conditions (TBVOOS only impacts FWCF events; PLUOOS only impacts LRNB events). Separate TBVOOS, FHOOS, and PLUOOS combined limits are not needed for operation beyond the EOCLB exposure since a FW temperature reduction is included to attain the additional cycle extension to the FFTR / coastdown exposure.

Reduced Feedwater Temperature at Startup During reactor startup, it is beneficial to reduce feedwater temperature to avoid excessive wear on reactor equipment. The desired feedwater temperature is less than the temperature assumed in the FHOOS licensing analyses performed each cycle. Therefore, previously defined

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-12 EOOS scenarios are not adequate to cover operation during startup with the desired reduction in feedwater temperature.

Analyses were performed to support all cycle exposures with or without EOC-RPT-OOS.

Analyses were also performed to support all cycle exposures with TBVOOS in combination with or without EOC-RPT-OOS. The analyses consider both NSS (above Pbypass cases) and TSSS. In addition, these analyses inherently cover all remaining non-PLUOOS equipment out-of-service scenarios defined in Table 1.1. Two separate startup feedwater temperatures are evaluated as provided in Item 6.6.1 of Reference 26. Limits for startup feedwater temperatures are included in Tables 8.1 - 8.6 and Table 8.9.

The reduced feedwater temperatures are not applicable above 50 % of rated power. The startup feedwater temperatures cannot be less than the values defined in Item 6.6.1 of Reference 26. If this requirement is met, reactor startup is restricted to the 85 % rod line or less.

Recirculation Pump Out-of-Service Recirculation pump out-of-service (RCPOOS) is the EOOS implying single loop operation. The pump seizure event assumes the reactor is operating with one recirculation pump inactive and an instantaneous seizure of the pump motor shaft of the active recirculation pump occurs. Flow through the active loop is rapidly reduced due to the large hydraulic resistance introduced by the stopped rotor causing core thermal power to decrease and reactor water level to swell. The sudden decrease in core coolant flow while the reactor is at power results in a degradation of core heat transfer which could result in fuel damage. The high water level setpoint is not reached; therefore, no reactor scram occurs.

Analysis assumptions have been constructed to seek a balance between operating flexibility and margin to thermal limits. Maximum core power is restricted to 43.75 % of rated and core flow is restricted to 50 % of rated; active recirculation drive flow is assumed 17.73 Mlb/hr. The results for the SLO pump seizure event are provided in Table 5.9. A comparison to the limiting MCPRs provided in Table 5.3 demonstrate SLO pump seizure is a non-limiting transient.

For RCPOOS, the TLO transient CPRs and LHGRFAC multipliers remain applicable.

Therefore, when developing the thermal limits, the only impacts on the LHGR and maximum average planar linear heat generation rate (MAPLHGR) limits is the application of a MAPLHGR

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-13 multiplier discussed in Section 8.3. The same situation is true for the EOOS scenarios. The TLO EOOS LHGRFAC multipliers remain applicable.

L icensing C omp liance In order to ensure that plant operation stays within the licensing evaluations that have been performed, two sets of criteria are defined. The first is defined as an exposure ratio and is intended to ensure that the licensing bases for the limiting NEOC limits will continue to be met.

The second is the licensing basis power shape. These criteria are intended to ensure that end of cycle limits also remain valid.

Axial Exposure Ratio The intent of the axial exposure ratio (AER) monitoring is to provide a guideline for reactor operation to assure that the actual operation remains within the bounds of the licensing basis relative to plant transients sensitive to spectral shift operation for the NEOC cycle exposure.

Only the AER at the full power condition at the NEOC core average exposure of 31,103.3 MWd/MTU is relevant in meeting this criterion. The AER should be less than or equal to the target AER of 1.0904.

Licensing Power Shape The licensing axial power profile used by Framatome for the plant transient analyses bounds the projected end of full power axial power profile. The conservative licensing axial power profile generated at the EOCLB core average exposure of 33,983.2 MWd/MTU is given in Table 5.12.

Operation is considered to be in compliance when:

• The integrated normalized power generated in the bottom 7 nodes from the projected EOFP solution at the state conditions provided in Table 5.12 is greater than the integrated normalized power generated in the bottom 7 nodes in the licensing basis axial power profile, and the individual normalized power from the projected EOFP solution is greater than the corresponding normalized power from the licensing basis axial power profile for at least 6 of the 7 bottom nodes.

• The projected EOFP condition occurs at a core average exposure less than or equal to EOCLB.

If the criteria cannot be fully met, the licensing basis may nevertheless remain valid but further assessment will be required.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-14 The licensing basis power profile in Table 5.12 was calculated using the MICROBURN-B2 code.

Compliance analyses must also be performed using MICROBURN-B2 or POWERPLEX-XD.

Note the power profile comparison should be done without incorporating instrument updates to the axial profile because the updated power is not used in the core monitoring system to accumulate assembly burnups.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-15 Table 5.1 Exposure Basis for Transient Analysis Core Average Exposure (MWd/MTU) Comments 15,103.3 Beginning of cycle Break point for exposure-dependent 31,103.3 MCPRp limits (NEOC)

Design basis rod patterns to 33,983.2 EOFP + 15 EFPD (EOCLB)

Maximum licensing core exposure 35,678.9 including FFTR / Coastdown 34,027.3 Cycle 22 EOC (short window) 34,610.0 Cycle 22 EOC (nominal window) 34,970.6 Cycle 22 EOC (long window)

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-16 Table 5.2 Scram Speed Insertion Times Control Surveillance Timing*

Rod Position TSSS NSS OSS (notch) (seconds) (seconds) (seconds) 48

--- 0.000 0.000 (full out) 48 --- 0.200 0.200 46 0.45 0.380 0.350 36 1.08 0.960 0.930 26 1.84 1.590 1.560 6 3.36 2.900 2.800 0

(full in)

• Transient analyses explicitly account for the Technical Specifications operation allowance for up to 13 slow and 1 stuck control rod by assuming 15 control blades fail to scram.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-17 Table 5.3 Base Case Limiting Transient Event NSS Insertion Time ATRIUM 10XM Limiting ATRIUM 11 Limiting Power MCPR Event MCPR Event 100 0.45 TTNB 0.47 FWCF 90 0.58 FWCF 0.54 FWCF 77.6 0.70 FWCF 0.84 FWCF 65 0.88 FWCF 0.82 FWCF 50 1.02 FWCF 1.04 FWCF 26 1.41 FWCF 1.35 FWCF 26 at > 50%F below Pbypass 1.62 FWCF 1.50 FWCF 26 at 50%F below Pbypass 1.37 FWCF 1.30 FWCF 23 at > 50%F below Pbypass 1.72 FWCF 1.66 FWCF 23 at 50%F below Pbypass 1.45 FWCF 1.44 FWCF

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-18 Table 5.4 Base Case Limiting Transient Event TSSS Insertion Time ATRIUM 10XM Limiting ATRIUM 11 Limiting Power MCPR Event MCPR Event 100 0.49 LRNB 0.52 FWCF 90 0.59 FWCF 0.60 FWCF 77.6 0.80 FWCF 0.82 FWCF 65 0.98 FWCF 0.96 FWCF 50 1.15 FWCF 1.16 FWCF 26 1.48 FWCF 1.50 FWCF 26 at > 50%F below Pbypass 1.62 FWCF 2.63 FWCF 26 at 50%F below Pbypass 1.37 FWCF 2.43 FWCF 23 at > 50%F below Pbypass 1.72 FWCF 2.79 FWCF 23 at 50%F below Pbypass 1.45 FWCF 2.57 FWCF

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-19 Table 5.5 Loss of Feedwater Heating Transient Analysis Results Power

(% rated) CPR 100 0.14 90 0.15 80 0.16 70 0.17 60 0.18 50 0.20 40 0.23 30 0.27 23 0.33

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-20 Table 5.6 Control Rod Withdrawal Error CPR Results Analytical RBM Setpoint (without filter) CRWE

(%) CPR MCPR*

107 0.27 1.35 111 0.31 1.39 114 0.34 1.42 117 0.40 1.48 Table 5.7 RBM Operability Requirements Thermal Power Applicable

(% rated) MCPR 2.02 TLO 27 % and < 90 %

2.04 SLO 90 % 1.52 TLO

• For rated power and a 1.08 SLMCPR.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-21 Table 5.8 Flow-Dependent MCPR Results Core Flow ATRIUM 10XM ATRIUM 11

(% rated) MCPR MCPR 30 1.58 1.55 40 1.48 1.42 50 1.45 1.37 60 1.42 1.32 70 1.34 1.28 80 1.26 1.23 90 1.21 1.18 100 1.15 1.13 107 1.08 1.08 Table 5.9 TLO and SLO Pump Seizure Results State point Power / Flow ATRIUM 10XM ATRIUM 11

(% rated) CPR CPR 100 / 105 TLO 0.24 0.23 43.75 / 50 SLO 0.83 0.70

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-22 Table 5.10 ATRIUM 11 LHGRFACp Transient Results*

PLUOOS TBVOOS Power Base and and

(% rated) Case FHOOS PLUOOS FHOOS TBVOOS FHOOS 100.0 0.95 0.96 1.00 1.00 0.94 0.93 90.0 0.92 0.92 1.00 1.00 0.92 0.89 77.6 0.89 0.90 1.00 1.00 0.89 0.83 65.0 0.84 0.87 --- --- 0.78 0.74 60.0 0.80 0.85 --- --- 0.79 0.72 55.0 0.87 0.83 --- --- 0.75 0.70 50.0 0.74 0.78 0.92 0.99 0.72 0.68 40.0 0.64 0.72 --- --- 0.67 0.63 26.0 0.54 0.59 --- --- 0.58 0.55 26.0 at > 50 % F below Pbypass 0.39 0.37 --- --- 0.37 0.35 23.0 at > 50 % F below Pbypass 0.37 0.35 --- --- 0.34 0.31 26.0 at 50 % F below Pbypass 0.45 0.42 --- --- 0.58 0.54 23.0 at 50 % F below Pbypass 0.46 0.44 --- --- 0.53 0.50 Power SFHOOS1 SFHOOS1 SFHOOS2 SFHOOS2

(% rated) TBVIS TBVOOS TBVIS TBVOOS 50.0 0.67 0.60 0.66 0.59 40.0 0.62 0.55 0.60 0.55 26.0 0.50 0.47 0.50 0.46 26.0 at > 50 % F below Pbypass 0.36 0.34 0.36 0.34 23.0 at > 50 % F below Pbypass 0.34 0.30 0.34 0.30 26.0 at 50 % F below Pbypass 0.40 0.45 0.40 0.45 23.0 at 50 % F below Pbypass 0.39 0.42 0.39 0.42

• Results support operation with or without EOC-RPT-OOS and are presented for all cycle exposures and scram insertion times.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-23 Table 5.11 ATRIUM 10XM LHGRFACp Transient Results*

PLUOOS TBVOOS Power Base and and

(% rated) Case FHOOS PLUOOS FHOOS TBVOOS FHOOS 100.0 0.99 1.00 1.00 1.00 0.99 1.00 90.0 0.95 1.00 1.00 1.00 0.95 0.97 77.6 0.93 0.98 1.00 1.00 0.97 0.90 65.0 0.88 0.94 --- --- 0.87 0.82 60.0 0.86 0.92 --- --- 0.87 0.81 55.0 0.94 0.91 --- --- 0.84 0.80 50.0 0.80 0.86 1.00 1.00 0.81 0.76 40.0 0.71 0.81 --- --- 0.75 0.71 26.0 0.60 0.68 --- --- 0.66 0.63 26.0 at > 50 % F below Pbypass 0.44 0.41 --- --- 0.43 0.40 23.0 at > 50 % F below Pbypass 0.42 0.40 --- --- 0.38 0.36 26.0 at 50 % F below Pbypass 0.50 0.48 --- --- 0.66 0.62 23.0 at 50 % F below Pbypass 0.51 0.49 --- --- 0.61 0.54 Power SFHOOS1 SFHOOS1 SFHOOS2 SFHOOS2

(% rated) TBVIS TBVOOS TBVIS TBVOOS 50.0 0.74 0.68 0.75 0.67 40.0 0.69 0.63 0.69 0.63 26.0 0.58 0.54 0.58 0.53 26.0 at > 50 % F below Pbypass 0.41 0.39 0.41 0.38 23.0 at > 50 % F below Pbypass 0.39 0.34 0.38 0.34 26.0 at 50 % F below Pbypass 0.45 0.50 0.44 0.50 23.0 at 50 % F below Pbypass 0.44 0.49 0.43 0.49

• Results support operation with or without EOC-RPT-OOS and are presented for all cycle exposures and scram insertion times.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-25 Figure 5.1 Limiting LRNB at 100P / 105F - TSSS Key Parameters

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-26 Figure 5.2 Limiting LRNB at 100P / 105F - TSSS Sensed W ater Level

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-27 Figure 5.3 Limiting LRNB at 100P / 105F - TSSS Vessel Pressures

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-28 Figure 5.4 Limiting TTNB at 100P / 105F - TSSS Key Parameters

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-29 Figure 5.5 Limiting TTNB at 100P / 105F - TSSS Sensed W ater Level

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-30 Figure 5.6 Limiting TTNB at 100P / 105F - TSSS Vessel Pressures

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-31 Figure 5.7 Limiting FW CF at 100P / 105F - TSSS Key Parameters

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-32 Figure 5.8 Limiting FW CF at 100P / 105F - TSSS Sensed W ater Level

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-33 Figure 5.9 Limiting FW CF at 100P / 105F - TSSS Vessel Pressures

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-34 Figure 5.10 Limiting IHPS at 100P / 105F - TSSS Key Parameters

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-35 Figure 5.11 Limiting IHPS at 100P / 105F - TSSS Sensed Water Level

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 5-36 Figure 5.12 Limiting IHPS at 100P / 105F - TSSS Vessel Pressures

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 6-1 POSTULATED ACCIDENTS L oss- of- C oolant- A ccident ( L OC A )

The results of the ATRIUM 10XM LOCA analysis are presented in References 28 and 29 as supplemented by Reference 30. The ATRIUM 10XM peak cladding temperature (PCT) is 2,052 F. The peak local metal water reaction is 2.06% and the maximum core wide metal-water reaction (for hydrogen generation) for a full ATRIUM 10XM core is < 1.0 %.

The results of the ATRIUM 11 LOCA analysis are presented in Reference 5, as supplemented by Reference 31. The ATRIUM 11 PCT is 1,898 F. The peak local metal water reaction is 8.27% and the maximum core wide metal-water reaction (for hydrogen generation) for a full ATRIUM 11 core is < 0.73%.

Analyses and results support the EOD and EOOS conditions listed in Table 1.1. Note TBVOOS, EOC-RPT-OOS, PLUOOS, and TIPOOS/LPRM out-of-service have no direct influence on the LOCA events.

C ontrol Rod Drop A ccident ( C RDA )

Plant startup utilizes a bank position withdrawal sequence (BPWS) including rod worth minimization strategies. The CRDA evaluation was performed for both A and B sequence startups consistent with the withdrawal sequences specified by TVA. The approved Framatome AURORA-B CRDA methodology is used (Reference 32). The applicability of this methodology for the Browns Ferry plants is demonstrated in Reference 33.

The analysis utilized the RG 1.236 criteria (Reference 34) consistent with the approach previously demonstrated in Reference 33. The CRDA analysis results demonstrate that the core coolability is maintained with total fuel enthalpy remaining below 230 cal/g and no fuel melting.

The radiological consequences are shown to be bounded by the Browns Ferry CRDA AST analysis. The number of effective rod failures (adjusted for revised release fractions) were confirmed to be less than the Browns Ferry licensing limit of 850 which is based upon the current AST dose analysis.

The following table identifies the limiting rod drop with the actual number of rod failures and the number of rod failures scaled up to account for revised release fractions of DG-1327 (Reference 35).

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 6-2 ATRIUM 10XM ATRIUM 11 Max Max FSAR FSAR Fraction of Prompt Bundles Actual Actual Total Fuel Dose- Dose- Allowed Sequence Enthalpy with Rod Rod Enthalpy Melt Equivalent Equivalent Rod Increase Failures Failures Failures (cal / g) Failures Failures Failures (cal / g)

MOC_ B4312 113.0 132.9 no 0 0 0 0 0 0 MOC_ B4321 Fuel and Eq uip ment H andling A ccident The fuel handling accident radiological analysis implementing the AST as approved in Reference 36 was performed with consideration of ATRIUM-10 core source terms. The ATRIUM 10XM and ATRIUM 11 source terms have been dispositioned relative to those in the AST analysis of record and found to support the same conclusions. Fuel assembly and reactor core isotopic inventories used as input to design basis radiological accident analyses are applicable to all three units (Reference 36). The number of failed fuel rods for the ATRIUM-10 fuel as previously provided to TVA in Reference 37 for use in the AST analysis is unchanged. The number of failed fuel rods for the ATRIUM 10XM fuel is 163 which remains bounded by the analysis of record. Framatome has also performed an analysis to demonstrate the number of failed rods due to a fuel handling accident involving ATRIUM 11 fuel does not exceed 194 (Reference 38). These results are consistent with the number of failed rods supported by the current Browns Ferry AST analysis. No other aspect of utilizing the ATRIUM 10XM and ATRIUM 11 fuel affects the current analysis; therefore, the AST fuel handling accident analysis remains applicable.

Fuel L oading Error ( I nfreq uent Event)

There are two types of fuel loading errors possible in a BWR - the mislocation of a fuel assembly in a core position prescribed to be loaded with another fuel assembly and the misorientation of a fuel assembly with respect to the control blade. As described in Reference 39, the fuel loading error is characterized as an infrequent event. The acceptance criterion is the offsite dose consequences due to the event shall not exceed a small fraction of the 10 CFR 50.67 limits.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 6-3 Mislocated Fuel Bundle Framatome has performed a fuel mislocation error analysis for Browns Ferry Unit 2 Cycle 23.

The analysis considered the impact of a mislocated assembly against potential fuel rod failure mechanisms due to increased LHGR and reduced CPR. Based on the analyses the offsite dose criteria (a small fraction of 10 CFR 50.67) is conservatively satisfied. A dose consequence evaluation is not necessary since no rod approaches the fuel centerline melt or 1 % strain limits and less than 0.1 % of the fuel rods are expected to experience boiling transition.

Misoriented Fuel Bundle Framatome has performed a fuel assembly misorientation analysis for Browns Ferry Unit 2 Cycle 23 (monitored with the ACE critical power correlation). The analysis was performed assuming the limiting assembly was loaded in the worst orientation (rotated 180 ), and depleted through the cycle without operator interaction. The analysis demonstrates the small fraction of 10 CFR 50.67 offsite dose criteria is conservatively satisfied. A dose consequence evaluation is not necessary since no rod approaches the fuel centerline melt or 1 % strain limits and less than 0.1 % of the fuel rods are expected to experience boiling transition.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-1 SPECIAL ANALYSES A SME Overp ressuriz ation A nalysis This section describes the maximum overpressurization analyses performed to demonstrate compliance with the ASME Boiler and Pressure Vessel Code. The analysis shows the safety / relief valves have sufficient capacity and performance to prevent the reactor vessel pressure from reaching the safety limit of 110 % of the design pressure.

MSIV closure, TSV closure, and TCV closure (without bypass) analyses were performed with the AURORA-B methodology (Reference 10) for 102 % power and both 85 % and 105 % flow at the highest cycle exposure. The MSIV closure event is similar to the other steam line valve closure events in that the valve closure results in a rapid pressurization of the core. The increase in pressure causes a decrease in voids which in turn causes a rapid increase in power.

The turbine bypass valves do not impact the system response and are not modeled in the analysis. [

] The following assumptions were made in the analysis.

• The most critical active component (direct scram on valve position or motion) was assumed to fail. However, scram on high neutron flux and high dome pressure is available.

• The SRV opening setpoints used in the analysis are set to the Technical Specification values increased by 3%, plus an additional 5 psi. To support operation with 1 MSRVOOS, the plant configuration analyzed assumed one of the lowest setpoint MSRV was inoperable.

• TSSS insertion times were used.

• The initial dome pressure was set at the maximum allowed by the Technical Specifications plus an additional 5 psi bias, 1,070 psia (1,055 psig).

• A fast MSIV closure time of 3.0 seconds was used.

• The analytical limit ATWS-RPT setpoint and function were assumed.

Results of the limiting valve closure, MSIV closure overpressurization analyses are presented in Table 7.1. Various reactor plant parameters during the limiting MSIV closure event are presented in Figure 7.1 - Figure 7.4. The maximum pressure of 1354 psig occurs in the lower

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-2 plenum. The maximum dome pressure for the same event is 1321 psig. The results demonstrate the maximum vessel pressure limit of 1,375 psig and dome pressure limit of 1,325 psig are not exceeded for any analyses.

A T W S Event Evaluation ATW S Overpressuriz ation Analysis This section describes analyses performed to demonstrate the peak vessel pressure for the limiting anticipated transient without scram (ATWS) event is less than the ASME Service Level C limit of 120 % of the design pressure (1,500 psig). Overpressurization analyses were performed at 100 % power at both 85 % and 105 % flow over the cycle exposure range for both the MSIV closure event and the PRFO event. The PRFO event assumes a step increase in pressure demand such that the pressure control system opens the TCV and TBV. Steam flow demand is assumed to increase to 125 % demand (equivalent to 131.3 % of rated steam flow) allowing a maximum TCV flow of 105 % and a maximum bypass system flow of 21.3 %. The system pressure decreases until the low pressure setpoint is reached resulting in the closure of the MSIV. The subsequent pressurization wave collapses core voids, thereby increasing core power.

[

]

The following assumptions were made in the analyses.

• The analytical limit ATWS-RPT setpoint and function were assumed.

• The SRV opening setpoints used in the analysis are set to the Technical Specification values increased by 3%, plus an additional 5 psi. To support operation with 1 MSRVOOS, the plant configuration analyzed assumed one of the lowest setpoint MSRV was inoperable.

• All scram functions were disabled.

• The initial dome pressure was set to the nominal pressure of 1,050 psia.

• A nominal MSIV closure time of 4.0 seconds was used for both events.

Analyses results are presented in Table 7.2. The response of various reactor plant parameters during the limiting MSIV closure event are shown in Figure 7.5 - Figure 7.8. The maximum

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-3 lower plenum pressure is 1498 psig and the maximum dome pressure is 1480 psig. The results demonstrate the ATWS maximum vessel pressure limit of 1,500 psig is not exceeded.

Long-Term Evaluation Fuel design differences may impact the power and pressure excursion experienced during the ATWS event. This in turn may impact the amount of steam discharged to the suppression pool and containment.

An evaluation for ATRIUM 11 fuel was presented in Section 7.3 of Reference 11. This evaluation concluded the introduction of the ATRIUM 11 fuel design does not significantly impact the long term ATWS response (suppression pool temperature and containment pressure) and the current analysis remains applicable. This conclusion is applicable for the Browns Ferry Unit 2 Cycle 23 core design.

Standb y L iq uid C ontrol System In the event the control rod scram function becomes incapable of rendering the core in a shutdown state, the standby liquid control (SLC) system is required to be capable of bringing the reactor from full power to a cold shutdown condition at any time in the core life. The Browns Ferry Unit 2 SLC system is required to be able to inject 720 ppm natural boron equivalent at 70 F into the reactor coolant. Framatome has performed an analysis demonstrating the SLC system meets the required shutdown capability for the cycle. The analysis was performed at a coolant temperature of 366 F with a boron concentration equivalent to 720 ppm at 68 F*. The temperature of 366 F corresponds to the low pressure permissive for the RHR shutdown cooling suction valves and represents the maximum reactivity condition with soluble boron in the coolant. The analysis shows the core to be subcritical throughout the cycle by at least 1.79 % k/k based on the short Cycle 22 EOC.

Fuel C riticality The spent fuel pool criticality analysis for ATRIUM 11 fuel is presented in Reference 9. The ATRIUM 11 fuel assemblies identified for the cycle meet the spent fuel storage requirements.

ATRIUM 11 fuel assemblies will not be stored in the new fuel storage vault.

• TVA Browns Ferry SLC licensing basis documents indicate a minimum of 720 ppm boron at a temperature of 70 F. The Framatome cold analysis basis of 68 F represents a negligible difference and the results are adequate to protect the 70 F licensing basis for the plant.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-4 Table 7.1 ASME Overpressurization Analysis Results Maximum Vessel Pressure Maximum Lower-Plenum Dome Pressure Event (psig) (psig)

MSIV closure 1,354 1,321 (102P / 105F)

Pressure 1,375 1,325 Limit

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-5 Table 7.2 ATWS Overpressurization Analysis Results Maximum Vessel Pressure Maximum Lower-Plenum Dome Pressure Event (psig) (psig)

MSIV closure 1,498 1,480 (100P / 85F)

Pressure 1,500 1,500 Limit

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-6 Figure 7.1 ASME- MSIV Overpressuriz ation Event at 102P / 105F - Key Parameters

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-7 Figure 7.2 ASME- MSIV Overpressuriz ation Event at 102P / 105F - Sensed W ater Level

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-8 Figure 7.3 ASME- MSIV Overpressuriz ation Event at 102P / 105F - Vessel Pressures

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-9 Figure 7.4 ASME- MSIV Overpressuriz ation Event at 102P / 105F - Safety / Relief Valve Flow Rates

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-10 Figure 7.5 ATW S- MSIV Overpressuriz ation Event at 100P / 85F - Key Parameters

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-11 Figure 7.6 ATW S- MSIV Overpressuriz ation Event at 100P / 85F - Sensed W ater Level

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-12 Figure 7.7 ATW S- MSIV Overpressuriz ation Event at 100P / 85F - Vessel Pressures

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 7-13 Figure 7.8 ATW S- MSIV Overpressuriz ation Event at 100P / 85F - Safety / Relief Valve Flow Rates

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-1 OPERATING LIMITS AND COLR INPUT MC P R L imits Determination of MCPR limits are based on analyses of the limiting AOTs. For Browns Ferry Unit 2 Cycle 23, [

]

MCPR operating limits are established such that less than 0.1 % of the fuel rods in the core are expected to experience boiling transition during an AOT initiated from rated or off-rated conditions and are based on a TLO SLMCPR of 1.08 and SLO SLMCPR of 1.09. Exposure-dependent MCPR limits were established to support operation from BOC to NEOC and NEOC to End of combined FFTR / Coastdown (Coast). MCPR limits are established to support base case operation and the EOOS scenarios presented in Table 1.1.

MCPRp limits above the bypass power level implement a flow dependency as a MCPR margin improvement when operating at lower core flows associated with the MELLLA+ operating domain. At a given power level, MCPRp limits are defined for specific ranges of core flow.

TLO MCPRp limits are presented for base case operation and the EOOS conditions in Table 8.1 - Table 8.6. Limits are presented for OSS, NSS, and TSSS insertion times for the exposure ranges considered. Tables 8.1 through 8.3 present the MCPRp limits for the BOC to NEOC exposure range. Tables 8.4 through 8.6 present the MCPRp limits for the NEOC to End of Coast exposure range. To develop MCPRp limits for SLO, thermal limits are a combination of

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-2 the corresponding TLO limit set and a 0.01 adder, which accounts for the difference in TLO and SLO SLMCPR.

MCPRf limits protect against fuel failures during a postulated slow flow excursion. The MCPRf limits presented in Table 8.7 are applicable for all cycle exposures and EOOS conditions identified in Table 1.1.

L H G R L imits The steady-state LHGR limits are presented in Table 8.8. Power- and flow-dependent multipliers (LHGRFACp and LHGRFACf) are applied directly to the LHGR limits to protect against fuel melting and overstraining of the cladding during an AOT.

The LHGRFACp multipliers are determined using the RODEX4 thermal-mechanical methodology (Reference 40) using the AURORA-B transient simulations. For the LHGRFACp evaluations [

]

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-3 LHGRFACp multipliers were established to support operation at all cycle exposures for all scram insertion times and for the EOOS conditions identified in Table 1.1 with and without TBVOOS.

LHGRFACp limits are presented in Table 8.9.

LHGRFACf multipliers are established to provide protection against fuel centerline melt and overstraining of the cladding during a postulated slow flow excursion. LHGRFACf limits are presented in Table 8.10. LHGRFACf multipliers are applicable for all cycle exposures and EOOS conditions identified in Table 1.1.

MA P L H G R L imits The ATRIUM 10XM TLO MAPLHGR limits are discussed in Table 8.11. For SLO, a multiplier of 0.85 must be applied to the TLO MAPLHGR limits. Power- and flow-dependent maximum average planar multipliers (MAPFAC) set-downs are not required; therefore, MAPFAC = 1.0.

The ATRIUM 11 TLO MAPLHGR limits are discussed in Table 8.11. For SLO, a multiplier of 0.85 must be applied to the TLO MAPLHGR limits. Power- and flow-dependent MAPFAC set-downs are not required; therefore, MAPFAC = 1.0.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-4 Table 8.1 TLO MCPRp Limits for OSS Insertion Times BOC to NEOC*

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.54 1.51 90.0 1.66 1.60 77.6 1.71 1.75 65.0 1.91 1.93 Base case 50.0 2.05 2.14 operation 26.0 2.45 2.39

> 50%F 50%F > 50%F 50%F 26.0 2.76 2.51 2.63 2.43 23.0 2.86 2.59 2.79 2.57 100.0 1.57 1.62 90.0 1.66 1.65 77.6 1.77 2.04 65.0 1.99 2.22 FHOOS 50.0 2.10 2.22 26.0 2.55 2.74

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.89 2.43 23.0 2.87 2.59 2.90 2.57

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-5 Table 8.2 TLO MCPRp Limits for NSS Insertion Times BOC to NEOC*

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.54 1.53 90.0 1.67 1.61 77.6 1.79 1.75 65.0 1.97 1.93 Base case 50.0 2.13 2.14 operation 26.0 2.52 2.45

> 50%F 50%F > 50%F 50%F 26.0 2.76 2.51 2.63 2.43 23.0 2.86 2.59 2.79 2.57 100.0 1.62 1.68 90.0 1.76 1.82 77.6 1.85 2.23 65.0 1.99 2.38 TBVOOS 50.0 2.16 2.38 26.0 2.52 2.58

> 50%F 50%F > 50%F 50%F 26.0 3.09 2.88 3.05 2.75 23.0 3.33 3.03 3.29 2.98 100.0 1.60 1.66 90.0 1.70 1.68 77.6 1.86 2.04 65.0 2.09 2.22 FHOOS 50.0 2.21 2.22 26.0 2.61 2.75

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.89 2.43 23.0 2.87 2.59 2.90 2.57 100.0 1.54 1.53 90.0 1.67 1.61 77.6 1.79 1.75 65.0 1.97 1.93 PLUOOS 50.0 2.13 2.14 26.0 2.52 2.45

> 50%F 50%F > 50%F 50%F 26.0 2.76 2.51 2.63 2.43 23.0 2.86 2.59 2.79 2.57

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-6 Table 8.2 TLO MCPRp Limits for NSS Insertion Times BOC to NEOC* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.65 1.96 90.0 1.76 1.96 77.6 1.91 2.26 TBVOOS 65.0 2.09 2.38 and 50.0 2.22 2.50 FHOOS 26.0 2.67 3.13

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 3.16 2.75 23.0 3.47 3.03 3.40 2.98 100.0 1.62 1.68 90.0 1.76 1.82 77.6 1.85 2.23 TBVOOS 65.0 1.99 2.38 and 50.0 2.16 2.38 PLUOOS 26.0 2.52 2.58

> 50%F 50%F > 50%F 50%F 26.0 3.09 2.88 3.05 2.75 23.0 3.33 3.03 3.29 2.98 100.0 1.60 1.66 90.0 1.70 1.68 77.6 1.86 2.04 FHOOS 65.0 2.09 2.22 and 50.0 2.21 2.22 PLUOOS 26.0 2.61 2.75

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.89 2.43 23.0 2.87 2.59 2.90 2.57 100.0 1.65 1.96 90.0 1.76 1.96 77.6 1.91 2.26 TBVOOS, 65.0 2.09 2.29 FHOOS, 50.0 2.22 2.50 and 26.0 2.67 3.13 PLUOOS

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 3.16 2.75 23.0 3.47 3.03 3.40 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-7 Table 8.2 TLO MCPRp Limits for NSS Insertion Times BOC to NEOC* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 2.21 2.22 50.0 2.21 2.22 Startup 26.0 2.84 2.75 FHOOS 1

> 50%F 50%F > 50%F 50%F TBVIS 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.22 2.50 50.0 2.22 2.50 Startup 26.0 2.84 3.13 FHOOS 1

> 50%F 50%F > 50%F 50%F TBVOOS 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98 100.0 2.21 2.22 50.0 2.21 2.22 Startup 26.0 2.84 2.75 FHOOS 2

> 50%F 50%F > 50%F 50%F TBVIS 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.22 2.50 50.0 2.22 2.50 Startup 26.0 2.84 3.13 FHOOS 2

> 50%F 50%F > 50%F 50%F TBVOOS 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Turbine bypass valves in-service (TBVIS) limits are applicable for all EOOS scenarios presented in Table 1.1 except those that include TBVOOS. TBVOOS limits are applicable for all EOOS scenarios presented in Table 1.1. Startup FHOOS 1 and Startup FHOOS 2 temperatures are presented as FW Set 1 and FW Set 2, respectively, in item 6.6.1 of Reference 26. Note feedwater heaters out-of-service conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-8 Table 8.3 TLO MCPRp Limits for TSSS Insertion Times BOC to NEOC*

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.58 1.61 90.0 1.68 1.70 77.6 1.89 1.84 65.0 2.07 2.00 Base case 50.0 2.26 2.26 operation 26.0 2.59 2.60

> 50%F 50%F > 50%F 50%F 26.0 2.76 2.51 2.63 2.43 23.0 2.86 2.59 2.79 2.57 100.0 1.74 2.10 90.0 1.80 2.22 77.6 1.98 2.65 65.0 2.09 2.65 TBVOOS 50.0 2.32 2.65 26.0 2.66 2.81

> 50%F 50%F > 50%F 50%F 26.0 3.09 2.88 3.05 2.75 23.0 3.33 3.03 3.29 2.98 100.0 1.62 1.87 90.0 1.77 1.87 77.6 1.95 2.32 65.0 2.17 2.37 FHOOS 50.0 2.35 2.38 26.0 2.64 2.80

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.89 2.43 23.0 2.87 2.59 2.90 2.57 100.0 1.58 1.61 90.0 1.68 1.70 77.6 1.89 1.84 65.0 2.07 2.00 PLUOOS 50.0 2.26 2.26 26.0 2.59 2.60

> 50%F 50%F > 50%F 50%F 26.0 2.76 2.51 2.63 2.43 23.0 2.86 2.59 2.79 2.57

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-9 Table 8.3 TLO MCPRp Limits for TSSS Insertion Times BOC to NEOC* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.75 2.27 90.0 1.81 2.27 77.6 1.98 2.86 TBVOOS 65.0 2.19 2.86 and 50.0 2.40 2.86 FHOOS 26.0 2.74 3.35

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 3.35 2.75 23.0 3.47 3.03 3.40 2.98 100.0 1.74 2.10 90.0 1.80 2.22 77.6 1.98 2.65 TBVOOS 65.0 2.09 2.65 and 50.0 2.32 2.65 PLUOOS 26.0 2.66 2.81

> 50%F 50%F > 50%F 50%F 26.0 3.09 2.88 3.05 2.75 23.0 3.33 3.03 3.29 2.98 100.0 1.62 1.87 90.0 1.77 1.87 77.6 1.95 2.32 FHOOS 65.0 2.17 2.37 and 50.0 2.35 2.38 PLUOOS 26.0 2.64 2.80

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.89 2.43 23.0 2.87 2.59 2.90 2.57 100.0 1.75 2.27 90.0 1.81 2.27 77.6 1.98 2.86 TBVOOS, 65.0 2.19 2.86 FHOOS, 50.0 2.40 2.86 and 26.0 2.74 3.35 PLUOOS

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 3.35 2.75 23.0 3.47 3.03 3.40 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-10 Table 8.3 TLO MCPRp Limits for TSSS Insertion Times BOC to NEOC* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 2.35 2.38 50.0 2.35 2.38 Startup 26.0 2.88 2.80 FHOOS 1

> 50%F 50%F > 50%F 50%F TBVIS 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.40 2.86 50.0 2.40 2.86 Startup 26.0 2.92 3.35 FHOOS 1

> 50%F 50%F > 50%F 50%F TBVOOS 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98 100.0 2.35 2.38 50.0 2.35 2.38 Startup 26.0 2.88 2.80 FHOOS 2

> 50%F 50%F > 50%F 50%F TBVIS 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.40 2.86 50.0 2.40 2.86 Startup 26.0 2.92 3.35 FHOOS 2

> 50%F 50%F > 50%F 50%F TBVOOS 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

TBVIS limits are applicable for all EOOS scenarios presented in Table 1.1 except those that include TBVOOS. TBVOOS limits are applicable for all EOOS scenarios presented in Table 1.1. Startup FHOOS 1 and Startup FHOOS 2 temperatures are presented as FW Set 1 and FW Set 2, respectively, in item 6.6.1 of Reference 26. Note feedwater heaters out-of-service conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-11 Table 8.4 TLO MCPRp Limits for OSS Insertion Times NEOC to End of Coast*

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.57 1.58 90.0 1.66 1.64 77.6 1.77 1.86 65.0 1.99 1.99 Base case 50.0 2.10 2.10 operation 26.0 2.55 2.33

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.48 2.43 23.0 2.87 2.59 2.57 2.57 100.0 1.57 1.58 90.0 1.66 1.64 77.6 1.77 1.86 65.0 1.99 1.99 FHOOS 50.0 2.10 2.10 26.0 2.55 2.33

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.48 2.43 23.0 2.87 2.59 2.57 2.57

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-12 Table 8.5 TLO MCPRp Limits for NSS Insertion Times NEOC to End of Coast*

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp

> 90%F 90%F > 90%F 90%F 100.0 1.60 1.60 1.63 1.52 90.0 1.70 1.70 1.74 1.56 77.6 1.86 1.86 1.92 1.72 Base case 65.0 2.09 2.09 2.00 1.79 operation 50.0 2.21 2.21 2.10 2.10 26.0 2.61 2.61 2.48 2.28

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.48 2.43 23.0 2.87 2.59 2.57 2.57 100.0 1.65 1.87 90.0 1.76 2.07 77.6 1.91 2.15 65.0 2.09 2.35 TBVOOS 50.0 2.22 2.35 26.0 2.67 2.51

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98

> 90%F 90%F > 90%F 90%F 100.0 1.60 1.60 1.63 1.52 90.0 1.70 1.70 1.74 1.56 77.6 1.86 1.86 1.92 1.72 FHOOS 65.0 2.09 2.09 2.00 1.79 50.0 2.21 2.21 2.10 2.10 26.0 2.61 2.61 2.48 2.28

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.48 2.43 23.0 2.87 2.59 2.57 2.57 100.0 1.60 1.63 90.0 1.70 1.74 77.6 1.86 1.92 65.0 2.09 2.00 PLUOOS 50.0 2.21 2.10 26.0 2.61 2.48

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.48 2.43 23.0 2.87 2.59 2.57 2.57

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-13 Table 8.5 TLO MCPRp Limits for NSS Insertion Times NEOC to End of Coast* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.65 1.87 90.0 1.76 2.07 77.6 1.91 2.15 TBVOOS 65.0 2.09 2.35 and 50.0 2.22 2.35 FHOOS 26.0 2.67 2.51

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98 100.0 1.65 1.87 90.0 1.76 2.07 77.6 1.91 2.15 TBVOOS 65.0 2.09 2.35 and 50.0 2.22 2.35 PLUOOS 26.0 2.67 2.51

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98 100.0 1.60 1.63 90.0 1.70 1.74 77.6 1.86 1.92 FHOOS 65.0 2.09 2.00 and 50.0 2.21 2.10 PLUOOS 26.0 2.61 2.48

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.48 2.43 23.0 2.87 2.59 2.57 2.57 100.0 1.65 1.87 90.0 1.76 2.07 TBVOOS, 77.6 1.91 2.15 FHOOS, 65.0 2.09 2.35 and 50.0 2.22 2.35 PLUOOS 26.0 2.67 2.51

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-14 Table 8.5 TLO MCPRp Limits for NSS Insertion Times NEOC to End of Coast* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 2.21 2.22 Startup 50.0 2.21 2.22 FHOOS 1 26.0 2.84 2.75 TBVIS > 50%F 50%F > 50%F 50%F 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.22 2.50 Startup 50.0 2.22 2.50 FHOOS 1 26.0 2.84 3.13 TBVOOS > 50%F 50%F > 50%F 50%F 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98 100.0 2.21 2.22 Startup 50.0 2.21 2.22 FHOOS 2 26.0 2.84 2.75 TBVIS > 50%F 50%F > 50%F 50%F 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.22 2.50 Startup 50.0 2.22 2.50 FHOOS 2 26.0 2.84 3.13 TBVOOS > 50%F 50%F > 50%F 50%F 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

TBVIS limits are applicable for all EOOS scenarios presented in Table 1.1 except those that include TBVOOS. TBVOOS limits are applicable for all EOOS scenarios presented in Table 1.1. Startup FHOOS 1 and Startup FHOOS 2 temperatures are presented as FW Set 1 and FW Set 2, respectively, in item 6.6.1 of Reference 26. Note feedwater heaters out-of-service conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-15 Table 8.6 TLO MCPRp Limits for TSSS Insertion Times NEOC to End of Coast*

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp

> 90%F 90%F > 90%F 90%F 100.0 1.62 1.62 1.63 1.56 90.0 1.77 1.77 1.79 1.57 77.6 1.95 1.95 1.92 1.74 Base case 65.0 2.17 2.17 2.04 1.82 operation 50.0 2.35 2.35 2.10 2.10 26.0 2.64 2.64 2.50 2.28

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.50 2.43 23.0 2.87 2.59 2.57 2.57 100.0 1.75 1.87 90.0 1.81 2.19 77.6 1.98 2.27 65.0 2.19 2.35 TBVOOS 50.0 2.40 2.35 26.0 2.74 2.55

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98

> 90%F 90%F > 90%F 90%F 100.0 1.62 1.62 1.63 1.56 90.0 1.77 1.77 1.79 1.57 77.6 1.95 1.95 1.92 1.74 65.0 2.17 2.17 2.04 1.82 FHOOS 50.0 2.35 2.35 2.10 2.10 26.0 2.64 2.64 2.50 2.28

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.50 2.43 23.0 2.87 2.59 2.57 2.57 100.0 1.62 1.63 90.0 1.77 1.79 77.6 1.95 1.92 65.0 2.17 2.04 PLUOOS 50.0 2.35 2.10 26.0 2.64 2.50

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.50 2.43 23.0 2.87 2.59 2.57 2.57

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-16 Table 8.6 TLO MCPRp Limits for TSSS Insertion Times NEOC to End of Coast* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 1.75 1.87 90.0 1.81 2.19 77.6 1.98 2.27 TBVOOS 65.0 2.19 2.35 and 50.0 2.40 2.35 FHOOS 26.0 2.74 2.55

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98 100.0 1.75 1.87 90.0 1.81 2.19 77.6 1.98 2.27 TBVOOS 65.0 2.19 2.35 and 50.0 2.40 2.35 PLUOOS 26.0 2.74 2.55

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98 100.0 1.62 1.63 90.0 1.77 1.79 77.6 1.95 1.92 FHOOS 65.0 2.17 2.04 and 50.0 2.35 2.10 PLUOOS 26.0 2.64 2.50

> 50%F 50%F > 50%F 50%F 26.0 2.87 2.51 2.50 2.43 23.0 2.87 2.59 2.57 2.57 100.0 1.75 1.87 90.0 1.81 2.19 TBVOOS, 77.6 1.98 2.27 FHOOS, 65.0 2.19 2.35 and 50.0 2.40 2.35 PLUOOS 26.0 2.74 2.55

> 50%F 50%F > 50%F 50%F 26.0 3.25 2.88 2.97 2.75 23.0 3.47 3.03 3.03 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

Note feedwater heaters out-of-service / FFTR and single-loop operation conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-17 Table 8.6 TLO MCPRp Limits for TSSS Insertion Times NEOC to End of Coast* (Continued)

Operating Power ATRIUM 10XM ATRIUM 11 Condition (% of rated) MCPRp MCPRp 100.0 2.35 2.38 Startup 50.0 2.35 2.38 FHOOS 1 26.0 2.88 2.80 TBVIS > 50%F 50%F > 50%F 50%F 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.40 2.86 Startup 50.0 2.40 2.86 FHOOS 1 26.0 2.92 3.35 TBVOOS > 50%F 50%F > 50%F 50%F 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98 100.0 2.35 2.38 Startup 50.0 2.35 2.38 FHOOS 2 26.0 2.88 2.80 TBVIS > 50%F 50%F > 50%F 50%F 26.0 3.13 2.51 3.05 2.43 23.0 3.13 2.59 3.05 2.57 100.0 2.40 2.86 Startup 50.0 2.40 2.86 FHOOS 2 26.0 2.92 3.35 TBVOOS > 50%F 50%F > 50%F 50%F 26.0 3.51 2.88 3.45 2.75 23.0 3.74 3.03 3.65 2.98

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. For single-loop operation, MCPRp limits will be 0.01 higher which accounts for the difference in TLO and SLO SLMCPR. Note that operation in SLO is only supported up to 43.75% core power, 50% core flow, and an active recirculation drive flow of 17.73 Mlb/hr.

TBVIS limits are applicable for all EOOS scenarios presented in Table 1.1 except those that include TBVOOS. TBVOOS limits are applicable for all EOOS scenarios presented in Table 1.1. Startup FHOOS 1 and Startup FHOOS 2 temperatures are presented as FW Set 1 and FW Set 2, respectively, in item 6.6.1 of Reference 26. Note feedwater heaters out-of-service conditions are not allowed when operating in the MELLLA+ operating domain.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-18 Table 8.7 MCPRf Limits Core Flow ATRIUM 10XM ATRIUM 11

(% of rated) MCPRf Limit MCPRf Limit 30.0 1.64 1.70 84.0 1.40 1.45 107.0 1.40 1.45 Table 8.8 Steady-State LHGR Limits ATRIUM 10XM ATRIUM 11 Peak Steady-State Steady-State Pellet Exposure LHGR Limit LHGR Limit (GWd/MTU) (kW/ft) (kW/ft) 0.0 14.1 13.6 18.9 14.1 ---

21.0 --- 13.6 53.0 --- 10.2 74.4 7.4 ---

80.0 --- 3.5

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-19 Table 8.9 LHGRFACp Multipliers*

Base case operation (TBVIS) TBVOOS LHGRFACp LHGRFACp EOOS Power Condition (% rated) ATRIUM 11 ATRIUM 10XM ATRIUM 11 ATRIUM 10XM 100.0 0.95 0.99 0.93 0.99 26.0 0.54 0.60 0.54 0.60 Nominal 26.0 at > 50 % F 0.37 0.41 0.35 0.40 operation and FHOOS § 23.0 at > 50 % F 0.35 0.40 0.31 0.36 26.0 at 50 % F 0.42 0.48 0.42 0.48 23.0 at 50 % F 0.42 0.48 0.42 0.48 100.0 0.95 0.99 0.93 0.99 26.0 0.50 0.58 0.45 0.54 Startup 26.0 at > 50 % F 0.36 0.41 0.34 0.39 FHOOS 1§ 23.0 at > 50 % F 0.34 0.39 0.30 0.34 26.0 at 50 % F 0.40 0.45 0.40 0.45 23.0 at 50 % F 0.39 0.44 0.39 0.44 100.0 0.95 0.99 0.93 0.99 26.0 0.50 0.58 0.44 0.53 Startup 26.0 at > 50 % F 0.36 0.41 0.34 0.38 FHOOS 2§ 23.0 at > 50 % F 0.34 0.38 0.30 0.34 26.0 at 50 % F 0.40 0.44 0.40 0.44 23.0 at 50 % F 0.39 0.43 0.39 0.43

• Limits support operation with or without EOC-RPT-OOS and any combination of 1 MSRVOOS, up to 18 TIP channels OOS, and up to 50 % of the LPRM out-of-service. Base case supports single-loop operation.

Limits are applicable for all the EOOS scenarios presented in Table 1.1 except those including TBVOOS.

Limits are applicable for all the EOOS scenarios presented in Table 1.1 including those with TBVOOS.

§ Nominal operation and FHOOS represents the feedwater temperatures shown in Figure 2.2 of Reference 26. Startup FHOOS 1 and Startup FHOOS 2 temperatures are presented as FW Set 1 and FW Set 2, respectively, in Item 6.6.1 of Reference 26.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 8-20 Table 8.10 LHGRFACf Multipliers ATRIUM 10XM ATRIUM 11 Core Flow LHGRFACf LHGRFACf

(% of rated) Multiplier Multiplier 0.0 0.60 0.60 30.0 0.60 0.60 77.5 1.00 ---

79.3 --- 1.00 107.0 1.00 1.00 Table 8.11 MAPLHGR Limits Average Planar ATRIUM 10XM ATRIUM 11 Exposure MAPLHGR Limit MAPLHGR Limit (GWd/MTU) (kW/ft) (kW/ft) 0.0 13.0 11.5 15.0 13.0 ---

20.0 --- 11.5 60.0 --- 9.0 67.0 7.6 ---

69.0 --- 7.2

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 9-1 REFERENCES

1. Letter, TVA to NRC, CNL-21-053, Request for License Amendment Regarding Application of Advanced Framatome Methodologies, and Adoption of TSTF-564 Revision 2 for Browns Ferry Nuclear Plant Units 1, 2, and 3, in Support of ATRIUM 11 Fuel Use at Browns Ferry (TS-535), dated July 23, 2021 (ML21204A128 and ML21204A129)

2. ANP-3996P Revision 0, Browns Ferry Unit 2 Cycle 23 Fuel Cycle Design Report, Framatome Inc., April 2022.

3. ANP-3860P Revision 1, Mechanical Design Report for Browns Ferry ATRIUM 11 Fuel Assemblies, Framatome Inc., June 2022.

4. ANP-3859P Revision 1, Browns Ferry Thermal-Hydraulic Design Report for ATRIUM 11 Fuel Assemblies, Framatome Inc., June 2022.

5. ANP-3905P Revision 2, Browns Ferry Units 1, 2, and 3 LOCA Analysis for ATRIUM 11 Fuel, Framatome Inc., June 2022.

6. FS1-0055816 Revision 1.0, Evaluation of ATRIUM 11 Fuel and the Containment Atmospheric Dilution System, Framatome Inc., April 2021.

7. ANP-3906P Revision 0, Browns Ferry ATWS-I Evaluation for ATRIUM 11 Fuel, Framatome Inc., April 2021.

8. Letter, F. Saba (USNRC) to J. Barstow (TVA), Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendment Nos. 310, 333, and 293 Regarding Maximum Extended Load Line Limit Analysis Plus (EPID L-2018-LLA-0048). ADAMS Accession Number ML19210C308.

9. ANP-3910P Revision 3, Browns Ferry Nuclear Plan Units 1, 2, and 3 Spent Fuel Storage Pool Criticality Safety Analysis for ATRIUM 11 Fuel, Framatome Inc., April 2022.

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12. ANP-3150P Revision 4, Mechanical Design Report for Browns Ferry ATRIUM' 10XM Fuel Assemblies, AREVA Inc., November 2017.

13. ANP-4001P Revision 1, ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23, Framatome Inc., August 2022.

14. ANP-3997P Revision 0, ATRIUM 10XM Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23, Framatome Inc., May 2022.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 9-2

15. Letter from Farideh E. Saba (NRC) to Joseph W. Shea (TVA), Browns Ferry Nuclear Plants, Units 1, 2, and 3 - Issuance of Amendments Regarding Technical Specifications (TS) Changes TS-478 Addition of Analytical Methodologies to TS 5.6.5 and Revision of TS 2.1.1.2 for Unit 2 (TAC NOS. MF0878 and MF0879), ML14108A334, July 31, 2014.

16. ANP-10307PA Revision 0, AREVA MCPR Safety Limit Methodology for Boiling Water Reactors, AREVA NP Inc., June 2011.

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(ADAMS Accession Number ML13324A099)

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23. NEDO-33006-A Revision 3, General Electric Boiling Water Reactor Maximum Extended Load Line Limit Analysis Plus, General Electric Hitachi Nuclear Energy America, LLC, June 2009. (ADAMS Accession Number ML091800530)

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25. Technical Specification Requirements for Browns Ferry Nuclear Plant Unit 2, Tennessee Valley Authority, as amended.

26. ANP-3986P Revision 0, Browns Ferry Unit 2 Cycle 23 Plant Parameters Document, Framatome Inc., February 2022.

27. ANF-1358(P)(A) Revision 3, The Loss of Feedwater Heating Transient in Boiling Water Reactors, Framatome ANP, Inc., September 2005.

28. ANP-3546P Revision 0, Browns Ferry Units 1, 2, and 3 LOCA Break Spectrum Analysis for ATRIUM 10XM Fuel (EPU MELLLA+), AREVA Inc., March 2017.

29. ANP-3547P Revision 2, Browns Ferry Units 1, 2, and 3 LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM 10XM Fuel (EPU MELLLA+), Framatome Inc.,

January 2020.

Framatome Inc. ANP-4017NP Revision 0 Browns Ferry Unit 2 Cycle 23 Reload Analysis Page 9-3

30. FS1-0044279 Revision 4.0, 10 CFR 50.46 PCT Error Report for Browns Ferry Units 1, 2, and 3 with EPU/MELLLA+ Conditions, Framatome Inc., November 19, 2021.

31. FS1-0063865 Revision 1.0, 10 CFR 50.46 PCT Error Report for Browns Ferry Units 1, 2, and 3 ATRIUM 11 Fuel, Framatome Inc., August 2022.

32. ANP-10333P-A Revision 0, AURORA-B: An Evaluation Model for Boiling Water Reactors; Application of Control Rod Drop Accident (CRDA), Framatome, March 2018 (as supplemented by Section 6.4 of ANP-3908P Revision 4, Applicability of Framatome BWR Methods to Browns Ferry with ATRIUM 11 Fuel, Framatome Inc., June 2022).

33. ANP-3874P Revision 3,Browns Ferry ATRIUM 11 Control Rod Drop Accident Analysis with the AURORA-B CRDA Methodology, Framatome Inc., June 2022.

34. RG 1.236, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, June 2020 (NRC ADAMS ML20055F490).

35. DG-1327, Pressurized-Water Reactor Control Rod Ejection and Boiling-Water Reactor Control Rod Drop Accidents, July 2019 (NRC ADAMS ML18302A106).

36. Letter, EA Brown (NRC) to KW Singer (TVA), Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Issuance of Amendments Regarding Full-Scope Implementation of Alternative Source Term (TAC Nos. MB5733, MB5734, MB5735, MC0156, MC0157 and MC0158)

(TS-405), September 27, 2004.

37. Letter, TA Galioto (FANP) to JF Lemons (TVA), Fuel Handling Accident Assumptions for Browns Ferry, TAG:02:012, January 23, 2002.

38. ANP-3307P Revision 1, AREVA Support of TVAs ECP for Implementation of ATRIUM 11 Into Browns Ferry, AREVA Inc., November 2014.

39. XN-NF-80-19(P)(A) Volume 4 Revision 1, Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads, Exxon Nuclear Company, June 1986.

40. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., February 2008.

Enclosure 3 Framatome Affidavit for Enclosure 1 CNL-22-100

AFFIDAVIT

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for Framatome Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in the report ANP-4017P, Revision 0, Browns Ferry Unit 2 Cycle 23 Reload Analysis, dated November 2022 and referred to herein as Document. Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6. The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatomes research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e) The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

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

Executed on: November 22, 2022 MEGINNIS Alan Digitally signed by MEGINNIS Alan Date: 2022.11.22 11:55:23 -08'00' Alan B. Meginnis

Proprietary Information - Withhold Under 10 CFR § 2.390 Enclosure 4 ANP-4001P, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23, Revision 1, Framatome, August 2022 (Proprietary)

CNL-22-100 Proprietary Information - Withhold Under 10 CFR § 2.390

Enclosure 5 ANP-4001NP, Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23, Revision 1, Framatome, August 2022 (Non-Proprietary)

CNL-22-100

Controlled Document ATRIUM 11 Fuel Rod Thermal- ANP-4001NP Revision 1 Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report August 2022 (c) 2022 Framatome Inc.

0414-12-F04 (Rev. 004, 04/27/2020)

Controlled Document ANP-4001NP Revision 1 Copyright © 2022 Framatome Inc.

All Rights Reserved FUELGUARD and ATRIUM are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries.

0414-12-F04 (Rev. 004, 04/27/2020)

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page i Nature of Changes Section(s)

Item or Page(s) Description and Justification 1 Pages 3-13 Incorrect values were reported in Table 3-2 and Table 3-3

& 3-14 of revision 0. [

].

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0

SUMMARY

AND CONCLUSIONS .................................................................... 2-1 3.0 FUEL ROD DESIGN EVALUATION .................................................................. 3-1 3.1 Fuel Rod Design ..................................................................................... 3-1 3.2 RODEX4 and Statistical Methodology Summary .................................... 3-2 3.3 Summary of Fuel Rod Design Evaluation ............................................... 3-4 3.3.1 Internal Hydriding ......................................................................... 3-6 3.3.2 Cladding Collapse ........................................................................ 3-6 3.3.3 Overheating of Fuel Pellets.......................................................... 3-6 3.3.4 Stress and Strain Limits ............................................................... 3-7 3.3.5 Fuel Densification and Swelling ................................................... 3-8 3.3.6 Fatigue ......................................................................................... 3-8 3.3.7 Oxidation, Hydriding, and Crud Buildup ....................................... 3-9 3.3.8 Rod Internal Pressure ................................................................ 3-10 3.3.9 Plenum Spring Design (Fuel Assembly Handling)...................... 3-10

4.0 REFERENCES

.................................................................................................. 4-1

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page iii List of Tables Table 2-1 Summary of Fuel Rod Design Evaluation Results (MELLLA+) ................... 2-2 Table 3-1 Key Fuel Rod Design Parameters, ATRIUM 11 for BFE2-23 .................... 3-11 Table 3-2 RODEX4 Fuel Rod Results Equilibrium CycleMELLLA+ ....................... 3-13 Table 3-3 RODEX4 Fuel Rod Results for ATRIUM 11 BFE2-23 Cycle MELLLA+................................................................................................. 3-14 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses ........................... 3-15 List of Figures Figure 2-1 LHGR Limit (Normal Operation).................................................................. 2-3

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page iv Nomenclature Acronym Definition 3GFG 3rd generation FUELGUARD AOO anticipated operational occurrences ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel BOL beginning of life BWR boiling water reactor CRWE control rod withdrawal error CUF cumulative usage factor EOL end of life FDL fuel design limit ID inside diameter LAR License Amendment Request LHGR linear heat generation rate LTP lower tie plate MWd/kgU megawatt days per kilogram of initial uranium MELLLA+ maximum extended load line limit analysis plus NRC Nuclear Regulatory Commission, U. S.

OD outside diameter PCI pellet-to-cladding-interaction PLFR part length fuel rod ppm parts per million SRA stress relieved annealed S-N stress amplitude versus number of cycles UTL upper tolerance limit

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 1-1

1.0 INTRODUCTION

Results of the fuel rod thermal-mechanical analyses are presented to demonstrate that the applicable design criteria are satisfied. The analyses are for the Framatome Inc. ATRIUM 11 fuel that will be inserted for operation in Browns Ferry Unit 2 Cycle 23 as reload batch BFE2-23.

These analyses assume the use of chromia additive in the enriched and natural urania portions of the fuel and assume operation in the Maximum Extended Load Line Limit Plus (MELLLA+)

operation domain. Both the design criteria and the analysis methodology have been approved by the U. S. NRC (NRC).

The analysis results are evaluated according to the generic fuel rod thermal and mechanical design criteria contained in ANF-89-98(P)(A) Revision 1 and Supplement 1 (Reference 1) along with design criteria provided in the RODEX4 fuel rod thermal-mechanical topical report (Reference 2)*. Approved methodology for the inclusion of chromia additive in the fuel pellets is also used (Reference 3).

The RODEX4 fuel rod thermal-mechanical analysis code is used to analyze the fuel rod for fuel centerline temperature, cladding strain, rod internal pressure, cladding collapse, cladding fatigue and external oxidation. The code and application methodology are described in the RODEX4 topical report (Reference 2). The cladding steady-state stress and plenum spring design methodology are summarized in Reference 1.

The following sections describe the fuel rod design, design criteria and methodology with reference to the source topical reports. Results from the analyses are summarized for comparison to the design criteria.

• (N.B., the cladding external oxidation limit from that topical report of [ ] was reduced to

[ ] when the RODEX4 methodology was approved for application to the Browns Ferry units (Reference 4)).

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 2-1 2.0

SUMMARY

AND CONCLUSIONS Key results are compared against each design criterion in:

- Table 2-1 for MELLLA+ operating domain Results are presented for the limiting cases. Additional RODEX4 results are given in Section 3.0.

The analyses support a maximum fuel rod discharge exposure of 62 MWd/kgU.

Fuel rod criteria applicable to the design are summarized in Section 3.0. Analyses show the criteria are satisfied when the fuel is operated at or below the LHGR (linear heat generation rate) limit (Fuel Design Limit - FDL) presented in Figure 2-1.

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 2-2 Table 2-1 Summary of Fuel Rod Design Evaluation Results (MELLLA+)

Criteria Description Criteria Result, Margin or Comment Section*

3.2 Fuel Rod Criteria 3.2.1 Internal hydriding [

]

(3.1.1) Cladding collapse [ ]

(3.1.2) Overheating of fuel No fuel melting [ ]

pellets margin to fuel melt > 0, °C 3.2.5 Stress and strain limits (3.1.1) Pellet-cladding [ ]

(3.1.2) interaction 3.2.5.2 Cladding steady-state [

stresses

]

3.3 Fuel System Criteria (3.1.1) Fatigue [ ]

(3.1.1) Oxidation, hydriding, [ ]

and crud buildup (3.1.1) Rod internal pressure [ ]

(3.1.2) 3.3.9 Fuel rod plenum spring Plenum spring to [

(fuel handling)

]

• Numbers in the column refer to paragraph sections in the generic design criteria document, ANF 98(P)(A) Revision 1 and Supplement 1 (Reference 1). A number in parentheses is the paragraph section in the RODEX4 fuel rod topical report (Reference 2).

Margin is defined as (limit - result).

The cladding external oxidation limit is restricted to [ ] by Reference 4.

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 2-3

[

]

Figure 2-1 LHGR Limit (Normal Operation)

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-1 3.0 FUEL ROD DESIGN EVALUATION Summaries of the design criteria and methodology are provided in this section along with analysis results in comparison to criteria. Both the fuel rod criteria and fuel system criteria as directly related to the fuel rod analyses are covered.

The fuel rod analyses cover normal operating conditions and AOOs (anticipated operational occurrences). The fuel centerline temperature analysis (overheating of fuel) and cladding strain analysis take into account slow transients at rated operating conditions.

Other fuel rod-related topics on overheating of cladding, cladding rupture, fuel rod mechanical fracturing, rod bow, axial irradiation growth, cladding embrittlement, violent expulsion of fuel and fuel ballooning are evaluated as part of the respective fuel assembly structural analysis, thermal hydraulic analyses, or LOCA analyses and are reported elsewhere. The evaluation of fast transients and transients at off-rated conditions also are reported separately from this report.

3.1 Fuel Rod Design The ATRIUM 11 fuel rod is conventional in design configuration and very similar to past designs such as the ATRIUM 10XM, ATRIUM-10 and ATRIUM-9 fuel rods.

The fuel rods are made with Zircaloy-2 cladding [

] plenum spring on the upper end of the fuel column assists in maintaining a compact fuel column during shipment and initial reactor operation.

There are two Part-length Fuel Rod (PLFR) designs incorporated in the fuel assembly. The longer is [ ] long, while the shorter is [ ] long. [

].

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-2

[

].

As on previous ATRIUM fuel designs that incorporated the 3rd generation FUELGUARD (3GFG)

Lower Tie Plate (LTP), the PLFRs have a [

].

Table 3-1 lists the main parameters for the fuel rod and components.

3.2 RODEX4 and Statistical Methodology Summary RODEX4 evaluates the thermal-mechanical response of the fuel rod surrounded by coolant.

The fuel rod model considers the fuel column, gap region, cladding, gas plena and the fill gas and released fission gases. The fuel rod is divided into axial and radial regions with conditions computed for each region. The operational conditions are controlled by the [

].

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-3 The heat conduction in the fuel and clad is [

].

Mechanical processes include [

].

As part of the methodology, fuel rod power histories are generated [

].

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-4 Since RODEX4 is a best-estimate code, uncertainties are taken into account by a [

]. Uncertainties taken into account in the analysis are summarized as:

• Power measurement and operational uncertainties - [

].

• Manufacturing uncertainties - [

].

• Model uncertainties - [

].

[

].

3.3 Summary of Fuel Rod Design Evaluation Results from the analyses are listed in Table 3-2 and Table 3-3. Summaries of the methods and codes used in the evaluation are provided in the following paragraphs. The design criteria are also listed along with references to the sections of the design criteria topical reports (References 1 and 2).

The fuel rod thermal and mechanical design criteria are summarized as follows.

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-5

• Internal Hydriding. The fabrication limit [

] to preclude cladding failure caused by internal sources of hydrogen (Section 3.2.1 of Reference 1).

• Cladding Collapse. Clad creep collapse shall be prevented. [

] (Section 3.1.1 of Reference 2).

• Overheating of Fuel Pellets. The fuel pellet centerline temperature during anticipated transients shall remain below the melting temperature (Section 3.1.2 of Reference 2).

• Stress and Strain Limits. [

] during normal operation and during anticipated transients (Sections 3.1.1 and 3.1.2 of Reference 2).

Fuel rod cladding steady-state stresses are restricted to satisfy limits derived from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code (Section 3.2.5.2 of Reference 1).

• Cladding Fatigue. The fatigue cumulative usage factor for clad stresses during normal operation and design cyclic maneuvers shall be below [ ] (Section 3.1.1 of Reference 2).

• Cladding Oxidation, Hydriding and Crud Buildup. Section 3.1.1 of Reference 2 limits the maximum cladding oxidation to less than [ ] to prevent clad corrosion failure. The oxidation limit is further reduced to [

] (Reference 4).

• Rod Internal Pressure. The rod internal pressure is limited [

] to ensure that significant outward clad creep does not occur and unfavorable hydride reorientation on cooldown does not occur (Section 3.1.1 of Reference 2).

• Plenum Spring Design (Fuel Handling). The rod plenum spring must maintain a force against the fuel column stack [ ] (Section 3.3.9 of Reference 1).

Cladding collapse, overheating of fuel, cladding transient strain, cladding cyclic fatigue, cladding oxidation, and rod pressure are evaluated [ ]. Cladding stress and the plenum spring are evaluated [ ].

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-6 3.3.1 Internal Hydriding The absorption of hydrogen by the cladding can result in cladding failure due to reduced ductility and formation of hydride platelets. Careful moisture control during fuel fabrication reduces the potential for hydrogen absorption on the inside of the cladding. The fabrication limit [

] is verified by quality control inspection during fuel manufacturing.

3.3.2 Cladding Collapse Creep collapse of the cladding and the subsequent potential for fuel failure is avoided in the design by limiting the gap formation due to fuel densification subsequent to pellet-clad contact.

The size of the axial gaps which may form due to densification following first pellet-clad contact shall be less than [ ].

The evaluation is performed using the RODEX4 code and methodology. RODEX4 takes into account the [

].

Table 3-2 lists the results for an equilibrium cycle operating in the MELLLA+ operating domain.

Table 3-3 lists the results for the ATRIUM 11 BFE2-23 cycle operating in the MELLLA+

operating domain.

3.3.3 Overheating of Fuel Pellets Fuel failure from the overheating of the fuel pellets is not allowed. The centerline temperature of the fuel pellets must remain below melting during normal operation and AOOs. The melting point of the fuel includes adjustments for [ ]. Framatome establishes an LHGR limit to protect against fuel centerline melting during steady-state operation and during AOOs.

Fuel centerline temperature is evaluated using the RODEX4 code and methodology for both normal operating conditions and AOOs.

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-7 Table 3-2 lists the results for an equilibrium cycle operating in the MELLLA+ operating domain.

Table 3-3 lists the results for the ATRIUM 11 BFE2-23 cycle operating in the MELLLA+

operating domain.

3.3.4 Stress and Strain Limits 3.3.4.1 Pellet/Cladding Interaction Cladding strain caused by transient-induced deformations of the cladding is calculated using the RODEX4 code and methodology. [

]. The strain limit is 1%.

Table 3-2 lists the results for an equilibrium cycle operating in the MELLLA+ operating domain.

Table 3-3 lists the results for the ATRIUM 11 BFE2-23 cycle operating in the MELLLA+

operating domain.

3.3.4.2 Cladding Stress Cladding stresses are calculated using solid mechanics elasticity solutions and finite element methods. The stresses are conservatively calculated for the individual loadings and are categorized as follows:

Category Membrane Bending Primary [

]

Secondary [

]

Stresses are calculated at the cladding outer and inner diameter in the three principal directions for both beginning of life (BOL) and end of life (EOL) conditions. At EOL, the stresses due to mechanical bow and contact stress are decreased due to irradiation relaxation. The separate

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-8 stress components are then combined, and the stress intensities for each category are compared to their respective limits.

The cladding-to-end cap weld stresses are evaluated for loadings from differential pressure, differential thermal expansion, rod weight, and plenum spring force.

The design limits are derived from the ASME (American Society of Mechanical Engineers)

Boiler and Pressure Vessel (B&PV) Code Section III (Reference 5) and the minimum specified material properties.

Table 3-4 lists the results in comparison to the limits for Beginning-of-Life (BOL) Hot conditions and End-of-Life (EOL) at both Hot and Cold conditions.

3.3.5 Fuel Densification and Swelling Fuel densification and swelling are limited by the design criteria for fuel temperature, cladding strain, cladding collapse, and rod internal pressure criteria. Although there are no explicit criteria for fuel densification and swelling, the effect of these phenomena are included in the RODEX4 code and methodology.

3.3.6 Fatigue Fuel rod cladding fatigue is calculated using the RODEX4 code and methodology. [

]. The CUF (cumulative usage factor) is summed for each of the axial regions of the fuel rod using Miners rule. The axial region with the highest CUF is used in the subsequent [

]. The maximum CUF for the cladding must remain below [ ] to satisfy the design criterion.

Table 3-2 lists the results for an equilibrium cycle operating in the MELLLA+ operating domain.

Table 3-3 lists the results for the ATRIUM 11 BFE2-23 cycle operating in the MELLLA+

operating domain.

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-9 3.3.7 Oxidation, Hydriding, and Crud Buildup Cladding external oxidation is calculated using the RODEX4 code and methodology. The corrosion model includes an enhancement factor that is derived from poolside measurement data to obtain a fit of the expected oxide thickness. An uncertainty value for the model enhancement factor also is determined from the data. The model uncertainty is included as part of the [ ].

[

].

In the event abnormal crud is observed at a plant, a specific analysis is required to address the higher crud level. An abnormal level of crud is defined by a formation that increases the calculated fuel average temperature by 25°C above the design basis calculation. The formation of crud is not calculated within RODEX4. Instead, an upper bound of expected crud based on plant observations is input by the use of the crud heat transfer coefficient. The corrosion model also takes into consideration the effect of the higher thermal resistance from the crud on the corrosion rate. A higher corrosion rate is therefore included as part of the abnormal crud evaluation. A similar specific analysis is required if an abnormal corrosion layer is observed instead of crud.

In the case of the Browns Ferry units, no additional crud is taken into account in the calculations because an abnormal crud or corrosion layer (beyond the design basis) has not been observed at the Browns Ferry units.

[

].

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-10 Currently, [

].

The oxide limit is evaluated such that greater than [

].

Table 3-2 lists the results for an equilibrium cycle operating in the MELLLA+ operating domain.

Table 3-3 lists the results for the ATRIUM 11 BFE2-23 cycle operating in the MELLLA+

operating domain.

3.3.8 Rod Internal Pressure Fuel rod internal pressure is calculated using the RODEX4 code and methodology. The maximum rod pressure is calculated under steady-state conditions and also takes into account slow transients. Rod internal pressure is limited to [

]. The expected upper bound of rod pressure [

] is calculated for comparison to the limit.

Table 3-2 lists the results for an equilibrium cycle operating in the MELLLA+ operating domain.

Table 3-3 lists the results for the ATRIUM 11 BFE2-23 cycle operating in the MELLLA+

operating domain.

3.3.9 Plenum Spring Design (Fuel Assembly Handling)

The plenum spring must maintain a force against the fuel column to prevent [

]. This is accomplished by designing and verifying the spring force in relation to the fuel column weight. The plenum spring is designed such that the [

].

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-11 Table 3-1 Key Fuel Rod Design Parameters, ATRIUM 11 for BFE2-23

[

]

Active fuel length, inch Full length rod 150.00

• The theoretical density of enriched and naturally enriched UO2-Cr pellets is 10.94 g/cm3 while that for UO2-Gd2O3 pellets is 10.96 g/cm3.

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-12 Table 3-1 Key Fuel Rod Design Parameters, ATRIUM 11 for BFE2-23 (contd)

[

]

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-13 Table 3-2 RODEX4 Fuel Rod Results Equilibrium CycleMELLLA+*

[

]

• Note that the results are provided up to fuel assembly discharge.

Margin is defined as (limit - result).

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-14 Table 3-3 RODEX4 Fuel Rod Results for ATRIUM 11 BFE2-23 Cycle MELLLA+*

[

]

• Note that the results are provided up to fuel assembly discharge.

Margin is defined as (limit - result).

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 3-15 Table 3-4 Cladding and Cladding-End Cap Steady-State Stresses Result Description, Stress Category Criteria BOL BOL EOL Cold Hot Hot Cladding stress Pm (primary membrane stress) [ ]

Pm + Pb (primary membrane + [ ]

bending)

P + Q (primary + secondary) [ ]

Cladding-End Cap stress Pm + Pb [ ]

Controlled Document Framatome Inc. ANP-4001NP Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23 Licensing Report Page 4-1

4.0 REFERENCES

1. ANF-89-98(P)(A) Revision 1 and Supplement 1, Generic Mechanical Design Criteria for BWR Fuel Designs, Advanced Nuclear Fuels Corporation, May 1995.

2. BAW-10247PA Revision 0, Realistic Thermal-Mechanical Fuel Rod Methodology for Boiling Water Reactors, AREVA NP Inc., February 2008.

3. ANP-10340P-A Revision 0. Incorporation of Chromia-Doped Fuel Properties in AREVA Approved Methods, Framatome Inc., May 2018.

4. Letter from Farideh E. Saba (NRC) to Joseph W. Shea (TVA), Browns Ferry Nuclear Plants, Units 1, 2, and 3 - Issuance of Amendments Regarding Technical Specifications (TS) Changes TS-478 Addition of Analytical Methodologies to TS 5.6.5 and Revision of TS 2.1.1.2 for Unit 2 (TAC NOS. MF0878 and MF0879), ML14108A334, July 31, 2014.

5. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, 1977.

6. ODonnell, W.J., and B. F. Langer, Fatigue Design Basis for Zircaloy Components, Nuclear Science and Engineering, Vol. 20, 1964.

Enclosure 6 Framatome Affidavit for Enclosure 4 CNL-22-100

AFFIDAVIT

1. My name is Alan B. Meginnis. I am Manager, Product Licensing, for Framatome Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Framatome to determine whether certain Framatome information is proprietary. I am familiar with the policies established by Framatome to ensure the proper application of these criteria.

3. I am familiar with the Framatome information contained in the report ANP-4001P, Revision 1 ATRIUM 11 Fuel Rod Thermal-Mechanical Evaluation for Browns Ferry Unit 2 Cycle 23, dated August 2022 and referred to herein as Document. Information contained in this Document has been classified by Framatome as proprietary in accordance with the policies established by Framatome for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Framatome and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) Trade secrets and commercial or financial information.

6. The following criteria are customarily applied by Framatome to determine whether information should be classified as proprietary:

(a) The information reveals details of Framatomes research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Framatome.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Framatome in product optimization or marketability.

(e) The information is vital to a competitive advantage held by Framatome, would be helpful to competitors to Framatome, and would likely cause substantial harm to the competitive position of Framatome.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(d) and 6(e) above.

7. In accordance with Framatomes policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside Framatome only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. Framatome policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

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

Executed on: August 25, 2022 MEGINNIS Alan Digitally signed by MEGINNIS Alan Date: 2022.08.25 15:40:57 -07'00' Alan B. Meginnis