Text

Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision
	ML16302A441
Person / Time
Site:	Browns Ferry
Issue date:	10/28/2016
From:	James Shea; Hitachi-GE Nuclear Energy, Ltd, Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16302A440	List: CNL-16-169, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision;
References
	CNL-16-169, NEDC-33860P NEDO-33860, Rev 1
	Download: ML16302A441 (588)
	v • d • e

[[:#Wiki_filter:Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosure 1 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-16-169 October 28, 2016 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:

Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision

References:

1. Letter from TVA to NRC, CNL-15-169, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU), dated September 21, 2015 (ML15282A152)

By the Reference 1 letter, Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for the Extended Power Uprate (EPU) of Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3. The proposed LAR modifies the renewed operating licenses to increase the maximum authorized core thermal power level from the current licensed thermal power of 3458 megawatts to 3952 megawatts. to this letter provides Revision 1 to the Power Uprate Safety Analysis Report (PUSAR) (NEDC-33860P). This revision updates the PUSAR to incorporate the information submitted, as marked-up PUSAR pages, in previous BFN EPU LAR Supplements. GE-Hitachi Nuclear Energy Americas LLC (GEH) and the Electric Power Research Institute (EPRI) consider portions of the information provided in Enclosure 1 to this letter to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390, Public inspections, exemptions, requests for withholding. An affidavit for withholding information, executed by GEH, is provided in Enclosure 3 to this letter. An affidavit for withholding information, executed by EPRI, is provided in Enclosure 4 to this letter. Enclosure 2 to this letter is a non-proprietary version of the document provided in Enclosure 1. Therefore, on behalf of GEH and EPRI, TVA requests that Enclosure 2 to this letter be withheld from public disclosure in accordance with the GEH and EPRI affidavits and the provisions of 10 CFR 2.390. Enclosures 1 and 2 to this letter replace and supersede Attachments 6 and 7, respectively, of the BFN EPU LAR (Reference 1).

U.S. Nuclear Regulatory Commission CNL-16-169 Page 2 October 28, 2016 TVA has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in the Reference 1 letter. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the supplemental information in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed license amendment. Additionally, in accordance with 10 CFR 50.91 (b)(1 ), TVA is sending a copy of this letter, without the proprietary information, to the Alabama State Department of Public Health. There are no new regulatory commitments associated with this submittal. If there are any questions or if additional information is needed, please contact Edward D. Schrull at (423) 751-3850. I declare under penalty of perjury that the foregoing is true and correct. Executed on the 28th day of October 2016. J. W . Shea Vice President, Nuclear Licensing

Enclosures:

1. NEDC-33860P, Revision 1, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate - (Proprietary version)

2. NED0-33860, Revision 1, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate - (Non-proprietary version)

3. Affidavit - GE-Hitachi Nuclear Energy Americas LLC

4. Affidavit - Electric Power Research Institute cc:

NRC Regional Administrator - Reg ion II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health (w/o Enclosure 1)

Withhold from Public Disclosure Under 10 CFR 2.390 ENCLOSURE 1 NEDC-33860P, Revision 1, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate (Proprietary version)

ENCLOSURE 2 NEDO-33860, Revision 1, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate (Non-proprietary version)

NEDO-33860 Revision 1 October 2016 Non-Proprietary Information - Class I (Public) SAFETY ANALYSIS REPORT FOR BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 EXTENDED POWER UPRATE Copyright 2016 GE - Hitachi Nuclear Energy Americas LLC All Rights Reserved

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) INFORMATION NOTICE This is a non-proprietary version of the document NEDC-33860P Revision 1, which has the proprietary information removed. Portions of the document that have been removed are indicated by a set of open and closed double square brackets as shown here (( )). IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purposes of supporting the Tennessee Valley Authority (TVA) license amendment request for an extended power uprate at Browns Ferry Nuclear Plants 1, 2, and 3 in proceedings before the U.S. Nuclear Regulatory Commission. The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilities, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document. No use of or right to copy any of this information contained in this document, other than by the NRC and its contractors in support of GEHs application, is authorized except by contract with GEH, as noted above. The information provided in this document is part of and dependent upon a larger set of knowledge, technology, and intellectual property rights pertaining to the design of standardized, nuclear powered, electric generating facilities. Without access and a GEH grant of rights to that larger set of knowledge, technology, and intellectual property rights, this document is not practically or rightfully usable by others, except by the NRC or through contractual agreements with TVA, as set forth in the previous paragraph. ii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Revision Summary Revision Description No. 0 Initial issue. 1 Revised to incorporate all changed pages. iii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) TABLE OF CONTENTS Page EXECUTIVE

SUMMARY

...................................................................................................... xxvii 1 . INTRODUCTION .................................................................................................................. 1-1 1.1 Report Approach ........................................................................................................ 1-1 1.1.1   Generic Assessments ..................................................................................... 1-1 1.1.2   Plant-Specific Evaluation............................................................................... 1-2 1.2 Purpose and Approach ............................................................................................... 1-3 1.2.1   Uprate Analysis Basis .................................................................................... 1-3 1.2.2   Computer Codes............................................................................................. 1-3 1.2.3   Approach ........................................................................................................ 1-4 1.3 EPU Plant Operating Conditions ............................................................................... 1-6 1.3.1   Reactor Heat Balance ..................................................................................... 1-6 1.3.2   Reactor Performance Improvement Features................................................. 1-6 1.4 Summary and Conclusions ........................................................................................ 1-6 2 . SAFETY EVALUATION ...................................................................................................... 2-1 2.1 Materials and Chemical Engineering ......................................................................... 2-1 2.1.1   Reactor Vessel Material Surveillance Program ............................................. 2-1 2.1.2   Pressure-Temperature Limits and Upper-Shelf Energy ................................. 2-3 2.1.3   Reactor Internal and Core Support Materials ................................................ 2-7 2.1.4   Reactor Coolant Pressure Boundary Materials ............................................ 2-10 2.1.5   Protective Coating Systems (Paints) - Organic Materials............................ 2-14 2.1.6   Flow-Accelerated Corrosion ........................................................................ 2-18 2.1.7   Reactor Water Cleanup System ................................................................... 2-22 2.2 Mechanical and Civil Engineering........................................................................... 2-51 2.2.1   Pipe Rupture Locations and Associated Dynamic Effects .......................... 2-51 2.2.2   Pressure-Retaining Components and Component Supports ........................ 2-55 2.2.3   Reactor Pressure Vessel Internals and Core Supports ................................. 2-76 2.2.4   Safety-Related Valves and Pumps ............................................................... 2-88 2.2.5   Seismic and Dynamic Qualification of Mechanical and Electrical Equipment .................................................................................................... 2-97 2.3 Electrical Engineering ............................................................................................ 2-131 iv

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.3.1 Environmental Qualification of Electrical Equipment .............................. 2-131 2.3.2 Offsite Power System ................................................................................ 2-135 2.3.3 AC Onsite Power System........................................................................... 2-140 2.3.4 DC Onsite Power System........................................................................... 2-142 2.3.5 Station Blackout ......................................................................................... 2-145 2.4 Instrumentation and Controls................................................................................. 2-169 2.4.1 Reactor Protection, Safety Features Actuation, and Control Systems ....... 2-169 2.5 Plant Systems ......................................................................................................... 2-188 2.5.1 Internal Hazards ......................................................................................... 2-188 2.5.2 Fission Product Control ............................................................................. 2-209 2.5.3 Component Cooling and Decay Heat Removal ......................................... 2-216 2.5.4 Balance-of-Plant Systems .......................................................................... 2-231 2.5.5 Waste Management Systems ..................................................................... 2-241 2.5.6 Additional Considerations ......................................................................... 2-250 2.5.7 Additional Review Areas (Plant Systems) ................................................. 2-253 2.6 Containment Review Considerations ..................................................................... 2-274 2.6.1 Primary Containment Functional Design................................................... 2-274 2.6.2 Subcompartment Analyses......................................................................... 2-288 2.6.3 Mass and Energy Release .......................................................................... 2-293 2.6.4 Combustible Gas Control in Containment ................................................. 2-296 2.6.5 Containment Heat Removal ....................................................................... 2-299 2.6.6 Secondary Containment Functional Design............................................... 2-322 2.7 Habitability, Filtration, and Ventilation ................................................................. 2-382 2.7.1 Control Room Habitability System ............................................................ 2-382 2.7.2 Engineered Safety Feature Atmosphere Cleanup ...................................... 2-384 2.7.3 Control Room Area Ventilation System .................................................... 2-386 2.7.4 Spent Fuel Pool Area Ventilation System ................................................. 2-388 2.7.5 Reactor, Turbine, and Radwaste Building Ventilation Systems ................ 2-389 2.7.6 Engineered Safety Feature Ventilation System ......................................... 2-391 2.8 Reactor Systems ..................................................................................................... 2-395 2.8.1 Fuel System Design ................................................................................... 2-395 2.8.2 Nuclear Design........................................................................................... 2-396 v

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.8.3 Thermal and Hydraulic Design .................................................................. 2-398 2.8.4 Emergency Systems ................................................................................... 2-399 2.8.5 Accident and Transient Analyses............................................................... 2-417 2.8.6 Fuel Storage ............................................................................................... 2-441 2.9 Source Terms and Radiological Consequences Analyses ..................................... 2-470 2.9.1 Source Terms for Radwaste Systems Analyses ......................................... 2-470 2.9.2 Radiological Consequences Analyses Using Alternative Source Terms ... 2-473 2.10 Health Physics ........................................................................................................ 2-490 2.10.1 Occupational and Public Radiation Doses ................................................. 2-490 2.11 Human Performance .............................................................................................. 2-512 2.11.1 Human Factors ........................................................................................... 2-512 2.12 Power Ascension and Testing Plan ........................................................................ 2-519 2.12.1 Approach to EPU Power Level and Test Plan ........................................... 2-519 2.13 Risk Evaluation ...................................................................................................... 2-522 2.13.1 Risk Evaluation of EPU ............................................................................. 2-522 3 . REFERENCES ....................................................................................................................... 3-1 vi

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) TABLES Table 1-1 Computer Codes Used For EPU Table 1-2 OLTP, CLTP, and EPU Plant Operating Conditions Table 2.1-1a Browns Ferry Unit 1 USE EMA - 60 Year Life (38 EFPY) Table 2.1-1b Browns Ferry Unit 2 USE EMA Year Life (48 EFPY) Table 2.1-1c Browns Ferry Unit 3 USE EMA Year Life (54 EFPY) Table 2.1-2a Browns Ferry Unit 1 Adjusted Reference Temperatures 60-Year License (38 EFPY) Table 2.1-2b Browns Ferry Unit 2 Adjusted Reference Temperatures 60-Year License (48 EFPY) Table 2.1-2c Browns Ferry Unit 3 Adjusted Reference Temperatures 60-Year License (54 EFPY) Table 2.1-3 Effects of Irradiation on Browns Ferry RPV Circumferential Weld Properties Table 2.1-4a Browns Ferry Unit 1 Comparison of Key Parameters Influencing FAC Wear Rate Table 2.1-4b Browns Ferry Unit 2 Comparison of Key Parameters Influencing FAC Wear Rate Table 2.1-4c Browns Ferry Unit 3 Comparison of Key Parameters Influencing FAC Wear Rate Table 2.1-5a Browns Ferry Unit 1 Components with Highest Predicted Wear Rate for Each Wear Rate Analysis Run Definition CHECWORKSTM SFA-Predicted Thickness vs. Measured Thickness Table 2.1-5b Browns Ferry Unit 2 Components with Highest Predicted Wear Rate for Each Wear Rate Analysis Run Definition CHECWORKSTM SFA-Predicted Thickness vs. Measured Thickness Table 2.1-5c Browns Ferry Unit 3 Components with Highest Predicted Wear Rate for Each Wear Rate Analysis Run Definition CHECWORKSTM SFA-Predicted Thickness vs. Measured Thickness Table 2.1-6 Comparison of RWCU System Operating Conditions Table 2.1-7 Comparisons of Chemistry Parameters for CLTP and EPU Cases Table 2.1-8 Selection Process Criteria for Components in the FAC Program Table 2.1-9a RCPB Piping and Safe End Materials of Construction Table 2.1-9b Summary of RCPB Welds per Generic Letter 88-01/BWRVIP-75-A Table 2.2-1 High Energy Line Break Outside Containment: Liquid Line Breaks Table 2.2-2 Reactor Coolant Pressure Boundary Structural Evaluation Table 2.2-3a Main Steam Pipe Stresses Due to EPU Conditions Table 2.2-3b Feedwater Pipe Stresses Due to EPU Conditions Table 2.2-3c Feedwater Pipe Stresses Due to Feedwater Transient Table 2.2-3d Feedwater and Condensate Pipe Stresses Due to Feedwater Transient Table 2.2-4a Main Steam System Piping (Outside Containment) vii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-4b Feedwater Piping Table 2.2-4c Condensate Piping Table 2.2-4d Extraction Steam Piping Table 2.2-4e FW Heater Drains and Vents Piping Table 2.2-4f Moisture Separator Vents and Drains Piping Table 2.2-5 BOP Piping System Evaluation Table 2.2-6 CUFs and Sp+q Values of Limiting Components Table 2.2-7 RIPDs for Normal Conditions Table 2.2-8 RIPDs for Upset Conditions Table 2.2-9 RIPDs for Faulted Conditions Table 2.2-10 Governing Stress Results for RPV Internal Components Table 2.2-11 Systems with Pumps and Valves in the IST Program Table 2.2-12 EPU Effects to Browns Ferry Program Valves Table 2.3-1 Summary of EPU Effect on EQ DBA Environmental Parameters Table 2.3-2 Evaluation of Pressure Qualification of EQ Components in the Drywell Table 2.3-3 Evaluation of Radiation Qualification of EQ Components Table 2.3-4 Normal Maximum and Total Radiation Requirements for Rooms at Browns Ferry Table 2.3-5 RWCU LOCA/HELB Temperature Evaluation Outside Containment Table 2.3-6 Offsite Electrical Equipment Ratings and Margins Table 2.3-7 Electrical Distribution System Load Changes Table 2.3-8a Key Inputs for Browns Ferry Station Blackout Table 2.3-8b Browns Ferry Station Blackout Sequence of Events Table 2.4-1 Technical Specification Setpoint Information Table 2.4-2 Changes to Instrumentation and Controls Table 2.5-1 NFPA 805 Fire Event Key Inputs Table 2.5-2 NFPA 805 Case 4 (EPU) Fire Event Evaluation Results Table 2.5-3 NFPA 805 Case 4 (EPU) Sequence of Events Table 2.5-4 SGTS Iodine Removal Capacity Parameters Table 2.5-5 Basis for Classification of No Significant Effect Table 2.6-1 Browns Ferry Containment Performance Results Table 2.6-2a Containment Response Key Analysis Input Values Table 2.6-2b Non-Accident Unit Containment Response Key Analysis Input Values viii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.6-3 Browns Ferry Peak Suppression Pool Temperatures for Postulated ATWS, Station Blackout, and NFPA 805 Events Table 2.6-4 ECCS Pump EPU NPSH Summary Table 2.6-4a ECCS Pump EPU NPSH Summary - Supplemental Evaluation Table 2.6-5 Input Comparisons for Containment Short-Term Analysis between CLTP and EPU Table 2.6-6 Input Comparisons Between CLTP and EPU for Limiting Long Term SP Temperature Analysis Table 2.6-6a Heat Sink Input Comparisons Between CLTP and EPU for Limiting Long Term SP Temperature Analysis Table 2.7-1 EPU Effect on Ventilation Systems Table 2.8-1 Browns Ferry Key Inputs for EPU ATWS Analysis Table 2.8-2 Browns Ferry Containment Results for ATWS Analysis Table 2.8-3 MSIVC Sequence of Events Table 2.8-4 PRFO Sequence of Events Table 2.8-5 LOOP Sequence of Events Table 2.8-6 IORV Sequence of Events Table 2.9-1 Total Activity Levels Table 2.9-2 Activity Concentrations of Principal Radionuclides in Fluid Streams for EPU Table 2.9-3 Comparison of Normal Operation (CLTP) and EPU Activation and Fission Products Table 2.9-4 Comparison of Design Basis to EPU Noble Gas Radionuclide Source Terms Table 2.9-5 Comparison of Design Basis to EPU Radiation Sources Table 2.9-6 LOCA Radiological Consequences Table 2.9-7 FHA Radiological Consequences Table 2.9-8 CRDA Radiological Consequences Table 2.9-9 MSLB Pre-Incident Iodine Spike Radiological Consequences Table 2.9-10 MSLB Equilibrium Iodine Concentration Radiological Consequences Table 2.9-11 Post-LOCA Vital Areas Requiring Continuous Occupancies Table 2.9-12 Post-LOCA Mission Doses Table 2.9-13 Post-Accident Sampling Mission Dose Summary CLTP versus EPU Table 2.10-1 Browns Ferry Average Occupational Dose for CLTP and EPU Table 2.10-1a Current and Anticipated Measured Radiation Dose in Selected Areas of the Reactor Building for 60 Year Normal Operation Table 2.10-1b Current and Anticipated Measured Radiation Dose in Selected Areas of the Drywell ix

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.10-1c Current and Anticipated Measured Radiation Dose in Selected Areas of the Turbine Building Table 2.10-2 Design Basis and Reported Annual Dose to Members of the Public x

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) FIGURES Figure 1-1 Power/Flow Operating Map for EPU Figure 1-2 EPU Heat Balance - Nominal Figure 1-3 EPU Heat Balance - Overpressure Protection Analysis Figure 2.3-1 Worst Case Drywell Temperature Profile Figure 2.3-2 Worst Case Secondary Containment EQ Temperature Profile Figure 2.5-1 NFPA 805 Case 4 (EPU) Fire Event Suppression Pool Temperature Figure 2.5-2a Browns Ferry Unit 1 Generator Reactive Capability Curve Figure 2.5-2b Browns Ferry Units 2 and 3 Generator Reactive Capability Curve Figure 2.6-1 EPU Suppression Pool Temperature Response to RSLB DBA-LOCA (CIC) Figure 2.6-1a EPU SP Temperature Response of Non-Accident Unit Shutdown - CST Available Figure 2.6-1b EPU SP Temperature Response of Non-Accident Unit Shutdown - CST Not Available Figure 2.6-1c Illustration of Browns Ferry 4 kV Distribution System to 480V RMOV Board Level Figure 2.6-2 EPU DW and WW Temperature Response to RSLB DBA-LOCA (SPC) Figure 2.6-3 EPU Short-Term RSLB DBA-LOCA Containment Pressure Response (Reference Condition: Initial DW Temperature=150ºF) Figure 2.6-4 EPU Short-Term RSLB DBA-LOCA Containment Temperature Response (Reference Condition: Initial DW Temperature =150ºF) Figure 2.6-5 EPU Short-Term RSLB DBA-LOCA Containment Pressure Response (Bounding Condition: Initial DW Temperature =130ºF) Figure 2.6-6 EPU Short-Term RSLB DBA-LOCA Containment Temperature Response (Bounding Condition: Initial DW Temperature =130ºF) Figure 2.6-7 EPU Short-Term RSLB DBA-LOCA Containment Pressure Response (Design Condition: Initial DW Temperature =70ºF) Figure 2.6-8 EPU Short-Term RSLB DBA-LOCA Containment Temperature Response (Design Condition: Initial DW Temperature =70ºF) Figure 2.6-9 EPU Long-Term Small Steam Line Break LOCA Drywell Temperature Response Figure 2.6-10 EPU Long-Term Small Steam Break LOCA Suppression Pool Temperature Response - 0.01 ft2 Break with HPCI Available Figure 2.6-11a Large Break-LOCA Short Term RHR NPSH versus Time Figure 2.6-11b DBA-LOCA Long Term RHR NPSH versus Time Figure 2.6-12a DBA-LOCA Short Term CS NPSH versus Time Figure 2.6-12b DBA-LOCA Long Term CS NPSH versus Time xi

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Figure 2.6-13a Small Break LOCA RHR NPSH versus Time Figure 2.6-13b Small Break LOCA CS NPSH versus Time Figure 2.6-14a Loss of RHR SDC - RHR NPSH versus Time Figure 2.6-14b Loss of RHR SDC - CS NPSH versus Time Figure 2.6-15a SORV with RPV Isolation Event - RHR NPSH versus Time Figure 2.6-15b SORV with RPV Isolation Event - CS NPSH versus Time Figure 2.6-16 Fire Event - RHR NPSH versus Time Figure 2.6-17 SBO Event - RHR NPSH versus Time Figure 2.6-18a ATWS - RHR NPSH versus Time Figure 2.6-18b ATWS - RHR NPSH versus Time Figure 2.6-19a Shutdown of the Non-Accident Unit - RHR NPSH versus Time Figure 2.6-19b Shutdown of the Non-Accident Unit - CS NPSH versus Time Figure 2.8-1 EPU MELLLA BOC MSIVC Figure 2.8-2 EPU MELLLA BOC MSIVC Figure 2.8-3 EPU MELLLA BOC MSIVC Figure 2.8-4 EPU MELLLA BOC PRFO Figure 2.8-5 EPU MELLLA BOC PRFO Figure 2.8-6 EPU MELLLA BOC PRFO Figure 2.8-7 EPU MELLLA EOC MSIVC Figure 2.8-8 EPU MELLLA EOC MSIVC Figure 2.8-9 EPU MELLLA EOC MSIVC Figure 2.8-10 EPU MELLLA EOC PRFO Figure 2.8-11 EPU MELLLA EOC PRFO Figure 2.8-12 EPU MELLLA EOC PRFO Figure 2.8-13 EPU MELLLA EOC LOOP Figure 2.8-14 EPU MELLLA EOC LOOP Figure 2.8-15 EPU MELLLA EOC LOOP Figure 2.8-16 EPU MELLLA EOC IORV Figure 2.8-17 EPU MELLLA EOC IORV Figure 2.8-18 EPU MELLLA EOC IORV xii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) ACRONYMS AND ABBREVIATIONS Term Definition AC Alternating Current ADHR Auxiliary Decay Heat Removal ADS Automatic Depressurization System AEC Atomic Energy Commission AHC Access Hole Cover AL Analytical Limit ALARA As Low As Is Reasonably Achievable ANS American Nuclear Society ANSI American National Standards Institute AOI Abnormal Operator Instruction AOO Anticipated Operational Occurrence (moderate frequency transient event) AOV Air-Operated Valve AP Annulus Pressurization APRM Average Power Range Monitor AREVA Areva Incorporated ARI Alternate Rod Insertion ART Adjusted Reference Temperature ARTS APRM/RBM/Technical Specifications ASME American Society of Mechanical Engineers ASDC Alternate Shutdown Cooling AST Alternate Source Term ASTM American Society for Testing and Materials ATWS Anticipated Transient Without Scram AV Allowable Value AVZ Above Vessel Zero B-10 Boron 10 xiii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition BHP Brake Horsepower BFN Browns Ferry Nuclear Plant BIIT Boron Injection Initiation Temperature BOC Beginning of Cycle BOP Balance-of-Plant Browns Ferry Browns Ferry Nuclear Plant (all units) BTU British Thermal Unit B&W Babcock and Wilcox BWR Boiling Water Reactor BWROG BWR Owners Group BWRVIP BWR Vessel and Internals Project CAD Containment Atmospheric Dilution CAP Containment Accident Pressure CARV Cross Around Relief Valve CCW Component Cooling Water CDF Core Damage Frequency CF Chemistry Factor CFD Condensate Filter Demineralizer CFR Code of Federal Regulations CIC Coolant Injection Cooling CLTP Current Licensed Thermal Power CLTR Constant Pressure Power Uprate Licensing Topical Report CO Condensation Oscillation COLR Core Operating Limits Report CPPU Constant Pressure Power Uprate CR Control Room CRAVS Control Room Area Ventilation System xiv

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition CRD Control Rod Drive CRDA Control Rod Drop Accident CRDH Control Rod Drive Housing CREVS Control Room Emergency Ventilation System CRGT Control Rod Guide Tube CRHZ Control Room Habitability Zone CS Core Spray CSBW Cold Shutdown Boron Weight CSC Containment Spray Cooling CSS Containment Spray System CST Condensate Storage Tank CUF Cumulative Usage Factor CWS Circulating Water System DBA Design Basis Accident DBA-LOCA Design Basis Loss-of-Coolant Accident DBE Design Basis Event DC Direct Current DFWCS Digital Feedwater Control System DHRP Decay Heat Removal Pressure DLO Dual (Recirculation) Loop Operation DOR Division of Operating Reactors DP Differential Pressure (psid) DW Drywell EAB Exclusion Area Boundary ECCS Emergency Core Cooling System ECCS-LOCA Emergency Core Cooling System - Loss-of-Coolant Accident EDG Emergency Diesel Generator xv

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition ECP Electrochemical Potential EECW Emergency Equipment Cooling Water EFDS Equipment and Floor Drainage System EFPY Effective Full Power Years EHC Electro-Hydraulic-Control ELTR1 Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate Licensing Topical Report ELTR2 Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate Licensing Topical Report EMA Equivalent Margin Analysis EOC End of Cycle EOI Emergency Operating Instruction EOL End of Life EOC-RPT End of Cycle-Recirculation Pump Trip EPRI Electric Power Research Institute EPU Extended Power Uprate EQ Environmental Qualification ESF Engineered Safety Feature ESFVS Engineered Safety Feature Ventilation System ESW Emergency Service Water FAC Flow Accelerated Corrosion FCF Flow Correction Factor FCV Flow Control Valve FFWTR Final Feedwater Temperature Reduction FHA Fuel Handling Accident FIV Flow Induced Vibration FPC Fuel Pool Cooling xvi

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition FPCCS Fuel Pool Cooling and Cleanup System FPP Fire Protection Program FSS Fire Safe Shutdown FSTF Full Scale Test Facility ft. Feet FUSAR Fuel Uprate Safety Analysis Report FW Feedwater FWCF Feedwater Controller Failure FWH Feedwater Heater FWHOOS Feedwater Heater Out-of-Service FWLB Feedwater Line Break GDC General Design Criteria GE General Electric GEH GE-Hitachi Nuclear Energy Americas LLC GL Generic Letter GSU Generator Step Up HCTL Heat Capacity Temperature Limit HCVS Hardened Containment Vent System HDR Header HDWP High Drywell Pressure HELB High Energy Line Break HEPA High Efficiency Particulate Adsorber HP High Pressure hp Horse Power HPCI High Pressure Coolant Injection HSBW Hot Shutdown Boron Weight HTR Heater xvii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition HVAC Heating Ventilating and Air Conditioning HWC Hydrogen Water Chemistry HWL High Water Level HWWV Hardened Wetwell Vent HX Heat Exchanger I&C Instrumentation and Control IASCC Irradiation-Assisted Stress Corrosion Cracking IBA Intermediate Break Accident ICF Increased Core Flow ICGT In Core Guide Tubes ID Identification IE Inspection and Enforcement ICHGT In Core Housing and Guide Tube ICS Integrated Computer System IEEE Institute of Electrical and Electronics Engineers IGSCC Intergranular Stress Corrosion Cracking IORV Inadvertent Opening of a Relief Valve IPB Isolated Phase Bus IPEEE Individual Plant Examination of External Events IRM Intermediate Range Monitor ISFSI Independent Spent Fuel Storage Installation ISP Integrated Surveillance Program ISI In-Service Inspection IST In-Service Testing JIT Just-in-Time JOG Joint Owners Group JP Jet Pump xviii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition JPSL Jet Pump Sensing Line kV Kilovolt LAR License Amendment Request LC Level Control LDI Liquid Drop Impingement LDR Load Definition Report LDS Leak Detection System LERF Large Early Release Frequency LFWH Loss of Feedwater Heater LOCA Loss-of-Coolant Accident LOFW Loss of Feedwater LOOP Loss of Offsite Power LP Low Pressure LPCI Low Pressure Coolant Injection LPRM Local Power Range Monitor LPSP Low Power Setpoint LPZ Low Population Zone LRPVP Low Reactor Pressure Vessel Pressure LSSS Limiting Safety System Setting LTP Long-Term Program LTR Licensing Topical Report LWL Low Water Level LWMS Liquid Waste Management System MCES Main Condenser Evacuation System MCR Main Control Room MCPR Minimum Critical Power Ratio MELB Moderate Energy Line Break xix

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition MELLL Maximum Extended Load Line Limit MELLLA Maximum Extended Load Line Limit Analysis MEQ Mechanical Equipment Qualification MeV Million Electron Volts MDRIR Minimum Debris Retention Injection Rate MEB Mechanical Electrical Branch Mlb or Mlbm Millions of Pounds MOV Motor Operated Valve MPS Minimum Recirculation Pump Speed MS Main Steam MSIV Main Steam Isolation Valve MSIVC Main Steam Isolation Valve Closure MSL Main Steam Line MSLB Main Steam Line Break MSLBA Main Steam Line Break Accident MSO Multiple Spurious Operation MSRV Main Steam Relief Valve MSRVDL Main Steam Relief Valve Discharge Line M&T Measurement and Test MVA Million Volt Amps MWe Megawatts-Electric MWt Megawatt-Thermal N/A Not Applicable NDE Non-Destructive Examination NEI Nuclear Energy Institute NERC North American Electric Reliability Council NFPA National Fire Protection Association xx

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition N-16 Nitrogen 16 NMCA Noble Metal Chemical Addition NobleChemTM Noble metal chemicals are added to coat metal surfaces as catalysts for HWC allowing IGSCC mitigation at lower hydrogen injection rates. NPSH Net Positive Suction Head NPSHa Net Positive Suction Head Available NPSHR Net Positive Suction Head - Required NQAM Nuclear Quality Assurance Manual NRC Nuclear Regulatory Commission NSI Next Scheduled Inspection NSSS Nuclear Steam Supply System NTSP Nominal Trip Set Point NUMAC Nuclear Measurement and Control NUMARC Nuclear Management and Resources Council NUREG Nuclear Regulatory Commission Technical Report Designation OC Outside Primary Containment ODCM Offsite Dose Calculation Manual OE Operating Experience OFS Orificed Fuel Support OI Operator Instruction OLMCPR Operating Limit Minimum Critical Power Ratio OLNC On-Line NobleChem - Process to inject NobleChemTM with the plant on-line. OLTP Original Licensed Thermal Power OOS Out-of-Service OSD Original Steam Dryer OSL Site Environmental Dosimeter Stations xxi

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition P Differential Pressure - psi PASS Post-Accident Sampling Station PC Primary Containment PCS Pressure Control System PCT Peak Clad Temperature PF Power Factor PLC Programmable Logic Controller PLUOOS Power Load Unbalance OOS PPT Peak Pool Temperature PRA Probabilistic Risk Assessment PRFO Pressure Regulator Failure Open PRNM Power Range Neutron Monitoring psi Pounds per Square Inch psia Pounds per Square Inch - Absolute psid Pounds per Square Inch - Differential psig Pounds per Square Inch - Gauge PSP Pressure Suppression Pressure P-T or P/T Pressure-Temperature PUAR Plant Unique Analysis Report PULD Plant Unique Load Definition QA/QC Quality Assurance / Quality Control RBCCW Reactor Building Closed Cooling Water RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RCW Raw Cooling Water xxii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition RDLB Recirculation Discharge Line Break RFP Reactor Feedwater Pump RFPT Reactor Feedwater Pump Turbine RFW Reactor Feedwater RG Regulatory Guide RHR Residual Heat Removal RHRSW Residual Heat Removal Service Water RIPD Reactor Internal Pressure Difference RMOV Reactor Motor Operated Valve RPT Recirculation Pump Trip RPTOOS Recirculation Pump Trip Out of Service RPV Reactor Pressure Vessel RRS Reactor Recirculation System RSD Replacement Steam Dryer RSLB Recirculation Suction Line Break RSW Raw Service Water RTNDT Reference Temperature of the Nil-Ductility Transition RTP Rated Thermal Power RWCU Reactor Water Cleanup RWE Rod Withdrawal Error RWM Rod Worth Minimizer RX Reactor SAF Single Active Failure SAMG Severe Accident Management Guideline SBA Small Break Accident SBO Station Blackout SC Safety Communication xxiii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition SCBA Self-Contained-Breathing-Apparatus SCW Stator Cooling Water SDC Shutdown Cooling SER Safety Evaluation Report SFA Steam/Feedwater Application SFIE Steam Flow Induced Error SFP Spent Fuel Pool SFPAVS Spent Fuel Pool Area Ventilation System SGTS Standby Gas Treatment System SHB Shroud Head Bolts SIF Stress Intensification Factor SIL Services Information Letter SIS System Impact Study SJAE Steam Jet Air Ejectors SL Service Level SLC Standby Liquid Control SLCS Standby Liquid Control System SLO Single-loop Operation SNM Susceptible Non-Modeled SORV Stuck Open Relief Valve SP Suppression Pool SPC Suppression Pool Cooling SPDS Safety Parameter Display System SQN Sequoyah Nuclear Plant SR Surveillance Requirement SRLR Supplemental Reload Licensing Report SRM Source Range Monitor xxiv

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition SRP Standard Review Plan SRSS Square Root of the Sum of the Squares SRV Safety Relief Valve SRVDL Safety Relief Valve Discharge Line SSC Systems, Structures and Components SSP Supplemental Surveillance Capsule Program SSW Sacrificial Shield Wall STM Steam STP Simulated Thermal Power SW Service Water SWMS Solid Waste Management System TAF Top of Active Fuel TBS Turbine Bypass System TBVOOS Turbine Bypass Valves OOS TCV Turbine Control Valve TEDE Total Effective Dose Equivalent TFSP Turbine First-Stage Pressure T-G Turbine-Generator TID Total Integrated Dose TIP Traversing In-core Probe TLAA Time Limiting Aging Analysis TRM Technical Requirement Manual TS Technical Specification TSC Technical Support Center TSV Turbine Stop Valve TSVC Turbine Stop Valve Closure TT Turbine Trip xxv

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Term Definition TTNBP Turbine Trip with no Steam Bypass Failure TVA Tennessee Valley Authority TW The TW sequence is a severe accident sequence that is the result of an anticipated transient followed by a total loss of decay heat removal. UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink UPS Uninterruptible Power Supply US United States USAS USA Standard USE Upper Shelf Energy USI Unresolved Safety Issue USNRC United States Nuclear Regulatory Commission USST Unit Station Service Transformer VFD Variable Flow Drive VPF Vane Passing Frequency VSL Vessel VWO Valve Wide-Open WB Whole Body WBN Watts Bar Nuclear Plant Wd Drive Flow WRA Wear Rate Analysis WW Wetwell xxvi

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) EXECUTIVE

SUMMARY

This Power Uprate Safety Analysis Report (PUSAR) summarizes the results of safety evaluations performed that justify uprating the licensed thermal power at Browns Ferry Nuclear Plant (Browns Ferry) Units 1, 2, and 3. The requested licensed power level is an increase to 3952 MWt from the current licensed reactor thermal power of 3458 MWt. The PUSAR is presented in a format consistent with the template safety evaluation report (SER) contained in Section 3.2 of the US NRC, Office of Nuclear Reactor Regulation, Review Standard for Extended Power Uprates, RS-001, December 2003. The Regulatory Evaluations from the template SER have been modified to reflect the licensing basis of Browns Ferry. GE-Hitachi Nuclear Energy Americas LLC (GEH) has previously developed and implemented a number of extended power uprates (EPUs) using the Nuclear Regulatory Commission (NRC) approved Licensing Topical Reports (LTRs), Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate, NEDC-32424P-A, February 1999 (ELTR1) and Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate, NEDC-32523P-A, February 2000 (ELTR2). Based on extended power uprate (EPU) experience, GEH has developed an approach to uprate reactor power that maintains the current plant maximum normal operating reactor dome pressure. This approach is referred to as Constant Pressure Power Uprate and was approved by the NRC in the Licensing Topical Report (LTR) NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, hereafter referred to as the CLTR. Some topics in the CLTR are directly fuel dependent because the fuel type affects the resulting evaluation or the consequences of transients or accidents. Because Browns Ferry Units 1, 2, and 3 will only contain Areva Incorporated, (AREVA) ATRIUM-10 and ATRIUM 10XM fuel types at the time of EPU implementation on the respective Browns Ferry units, the requested Browns Ferry EPU does not reference the CLTR as the basis for areas involving fuel-dependent topics, consistent with the NRCs Conditions and Limitations on the use of the CLTR. The fuel-dependent evaluations were performed by TVA or AREVA using NRC approved codes and methods. Due to the proprietary nature of the AREVA analyses, the fuel-dependent analyses in support of the requested EPU are contained in License Amendment Request (LAR) (Fuel Uprate Safety Analysis Report) to the Browns Ferry EPU LAR. The safety evaluation sections in this report provide appropriate cross references to LAR Attachment 8 for fuel-related topics. For evaluations independent of fuel type, this report provides a systematic application of the CLTR approach to systems, structures, components and evaluations, including the performance of plant-specific engineering assessments and confirmation of the applicability of the CLTR generic assessment required to support an EPU. It is not the intent of this report to explicitly address all the details of the analyses and evaluations described herein. For example, only previously NRC-approved or industry-accepted methods were used for the analyses of accidents and transients, as referred to in the CLTR, xxvii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) ELTR1, or ELTR2. Therefore, the safety analysis methods have been previously addressed, and thus, are not explicitly addressed in this report. Also, event and analysis descriptions that are already provided in other licensing reports or the Updated Final Safety Analysis Report (UFSAR) are not repeated within this report. This report, in conjunction with other Attachments to the EPU LAR, summarizes the significant evaluations needed to support a licensing amendment to allow for uprated power operation at Browns Ferry. Uprating the power level of nuclear power plants can be done safely within plant-specific limits and is a cost-effective way to increase installed electrical generating capacity. Many light water reactors have already been uprated worldwide, including many Boiling Water Reactor (BWR) plants. An increase in the electrical output of a BWR plant is accomplished primarily by generating and supplying higher steam flow to the turbine-generator. Browns Ferry, as currently licensed, has an as-designed equipment and system capability to accommodate steam flow rates above the current rating. Also, the plant has sufficient design margins to allow the plant to be safely uprated significantly beyond its current licensed power level. A higher steam flow is achieved by increasing the reactor power along specified control rod and core flow lines. A limited number of operating parameters are changed, some setpoints are adjusted and instruments are recalibrated. Plant procedures are revised and power ascension testing is performed to confirm the results of the safety analyses. Detailed evaluations of the reactor, engineered safety features, power conversion, emergency power, support systems, and design basis accidents were performed. This report, in conjunction with the fuel dependent evaluations contained in LAR Attachment 8, demonstrate that Browns Ferry can safely operate at the requested EPU level. However, non-safety power generation modifications will be implemented in order to obtain the electrical power output associated with the uprate power. Until these modifications are completed, the non-safety related, balance of plant equipment may limit the electrical power output, which in turn may limit the operating thermal power level to less than the rated thermal power level. These modifications have been evaluated and they do not constitute a material alteration to the plant. The evaluations and reviews were conducted in accordance with the CLTR and the criteria in ELTR1, ELTR2, or the TVA and AREVA codes and methods using NRC-approved or industry-accepted analysis methods. The results of these evaluations and reviews presented in this report are as follows: All fuel independent safety aspects of Browns Ferry that are affected by the increase in thermal power were evaluated (fuel dependent safety aspects are evaluated in LAR Attachment 8); No reliance on containment accident pressure is required to ensure adequate emergency core cooling system pump net positive suction head during accidents, abnormal operational transients or special (ATWS, SBO, Fire) events; Evaluations were performed using NRC-approved or industry-accepted analysis methods; xxviii

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Systems and components affected by EPU were reviewed to ensure there is no significant challenge to any safety system; No changes, which require compliance with more recent industry codes and standards, are being requested; No new design functions that require modifications are necessary for safety related systems, and any modification to non-safety related and/or power generation equipment will be implemented per 10 CFR 50.59; and The UFSAR will be updated for the EPU related changes, after EPU is implemented, per the requirements in 10 CFR 50.71(e). xxix

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public)

1. INTRODUCTION 1.1 Report Approach This Power Uprate Safety Analysis Report (PUSAR) summarizes the results of safety evaluations performed to justify uprating the licensed thermal power at Tennessee Valley Authority (TVA) Browns Ferry Nuclear Plant (Browns Ferry) Units 1, 2, and 3. The requested license power level is an increase to 3,952 MWt from the current licensed reactor thermal power (CLTP) of 3,458 MWt.

GE-Hitachi Nuclear Energy Americas LLC (GEH) has previously developed and implemented EPU at several nuclear power plants. Based on EPU experience, GEH has developed an approach to uprating reactor power that maintains the current plant maximum normal operating reactor dome pressure. This approach is referred to as Constant Pressure Power Uprate (CPPU) and is contained in the LTR NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, (Reference 1) hereafter referred to as the CLTR. The NRC approved the CLTR in the staff SER contained in Reference 1 for BWR plants containing General Electric (GE) fuel types through GE14 and using GEH accident analysis methods. Because Browns Ferry uses non-GE fuel, the CLTR is not applicable for fuel design dependent evaluations and transients. Analyses and evaluations performed in support of the generic dispositions in the CLTR are not applicable. Fuel dependent subjects are addressed in the complementary Fuel Uprate Safety Analysis Report (FUSAR) included as Attachment 8 to the License Amendment Request (LAR) for power uprate. This evaluation justifies an EPU to 3,952 MWt, with no increase in reactor operating pressure, which corresponds to 120% of the original licensed thermal power (OLTP) for Browns Ferry. This report is presented in a format consistent with the template contained in Section 3.2 of the United States Nuclear Regulatory Commission (USNRC), Office of Nuclear Reactor Regulation, Review Standard for Extended Power Uprates, RS-001, December 2003 (Reference 2). The Regulatory Evaluations from the template have been modified to reflect the licensing basis of Browns Ferry. 1.1.1 Generic Assessments Many of the component, system, and performance evaluations contained within this report have been generically evaluated in the CLTR and the Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate, (Reference 3) hereafter referred to as ELTR2, and found to be acceptable by the NRC. The plant-specific applicability of these generic assessments is identified and confirmed in the applicable sections of this report. Generic assessments are those safety evaluations that can be dispositioned for a group or all BWR plants by: A bounding analysis for the limiting conditions, Demonstrating that there is a negligible effect due to EPU, or 1-1

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Demonstrating that the required plant cycle-specific reload analyses are sufficient and appropriate for establishing the EPU licensing basis. Bounding analyses may be based on either: (1) a demonstration that assessments provided in previous EPU LTRs that included a pressure increase (References 3 and 4) are bounding; or (2) on specific generic studies provided in the CLTR. For these bounding analyses, the current EPU experience is provided in the CLTR, ELTR1, and ELTR2, along with the basis and results of the assessment. For those EPU assessments having a negligible effect, the current EPU experience plus a phenomenological discussion of the basis for the assessment is provided in the CLTR. Assessments that are dependent on the fuel design were performed by others and are included in the complementary FUSAR and associated fuel-related reports in Attachments 8 through 38 of the EPU License Amendment Request (LAR). Some of the safety evaluations affected by EPU are fuel cycle (reload) dependent. Reload dependent evaluations require that the reload fuel design, core loading pattern, and operational plan be established so that analyses can be performed to establish core operating limits. The reload analysis demonstrates that the core design for EPU meets the applicable NRC evaluation criteria and limits. Because of the lead-time required for the NRC review of this power uprate submittal, the Browns Ferry reload core design for the initial fuel cycle at uprated power are not established at the time of this submittal. As discussed in Section 2.8.2, the EPU has a relatively small effect on core operating and safety limits. Therefore, the reload fuel design and core loading pattern dependent plant evaluations for EPU operations are performed with the reload analysis as part of the standard reload licensing process. Because Browns Ferry Units 1, 2, and 3 will only contain AREVA ATRIUM-10 and ATRIUM 10XM fuel types at the time of EPU implementation on the respective Browns Ferry units, the requested Browns Ferry EPU does not reference the CLTR as the basis for areas involving fuel-dependent topics, consistent with the NRCs Conditions and Limitations on the use of the CLTR. The fuel-dependent evaluations were performed by TVA or AREVA using NRC approved codes and methods. Due to the proprietary nature of the AREVA analyses, the fuel-dependent analyses in support of the requested EPU are contained in FUSAR Attachment 8 to the Browns Ferry EPU licensing amendment request (LAR). No plant can implement a power uprate unless the appropriate reload core analysis is performed and all criteria and limits are satisfied. Otherwise, the plant would be in an unanalyzed condition. Based on current requirements, the reload analysis results are documented in the Supplemental Reload Licensing Report (SRLR), and the applicable core operating limits are documented in the plant-specific Core Operating Limits Report (COLR). 1.1.2 Plant-Specific Evaluation Plant-specific evaluations are assessments of the principal evaluations that are not addressed by the generic assessments described in Section 1.1.1. The relative effect of EPU on the plant-specific evaluations and the methods used for their performance are provided. Where applicable, the assessment methodology is referenced. If a specific computer code is used, the name of this 1-2

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) computer code is provided in the section. Table 1-1 provides a summary of the computer codes used. The plant-specific evaluations performed and reported in this document use plant-specific values to model the actual plant systems, transient response, and operating conditions. These plant-specific analyses are considered reload independent and are performed using a conservative core representative of Browns Ferry design for operation at 120% of OLTP for a cycle length of 24 months. 1.2 Purpose and Approach An increase in electrical output of a BWR is accomplished primarily by generation and supply of higher steam flow to the turbine generator (T-G). Most BWRs, as originally licensed, have an as-designed equipment and system capability to accommodate steam flow rates at least 5% above the original rating. In addition, continuing improvements in the analytical techniques (computer codes) based on several decades of BWR safety technology, plant performance feedback, operating experience (OE), and improved fuel and core designs have resulted in significant increases in the design and operating margins between the calculated safety analyses results and the current plant licensing limits. The available margins in calculated results, combined with the as-designed excess equipment, system, and component capabilities (1) have allowed many BWRs to increase their thermal power ratings by 5% without any Nuclear Steam Supply System (NSSS) hardware modification, and (2) provide for power increases up to 20% with some non-safety hardware modifications. These power increases involve no significant increase in the hazards presented by the plants as approved by the NRC in the original license. The method for achieving higher power is to extend the power/flow map (Figure 1-1) along the Maximum Extended Load Line Limit Analysis (MELLLA) line. However, there is no increase in the maximum normal operating reactor vessel dome pressure or the maximum licensed core flow over their CLTP values. EPU operation does not involve increasing the maximum normal operating reactor vessel dome pressure, because the plant, after modifications to non-safety power generation equipment, has sufficient pressure control and turbine flow capabilities to control the inlet pressure conditions at the turbine. 1.2.1 Uprate Analysis Basis Browns Ferry is currently licensed at the 100% CLTP level of 3,458 MWt. The EPU rated thermal power (RTP) level included in this evaluation is 120% of the OLTP. Plant-specific EPU parameters are listed in Table 1-2. The EPU safety analyses are based on a power level of 1.02 times the EPU power level unless the two percent power factor (PF) is already accounted for in the analysis methods consistent with the methodology described in Reference 5, or the 2% does not apply (e.g., Anticipated Transient Without Scram (ATWS) and SBO events). 1.2.2 Computer Codes NRC-approved or industry-accepted computer codes and calculational techniques are used to demonstrate compliance with the applicable regulatory acceptance criteria. The codes used in the analyses for this report are provided in Table 1-1. Computer codes for the fuel dependent 1-3

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) analyses are specified in the FUSAR, Attachment 8 to the LAR. The application of these codes to the EPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The limitations on use of these codes and methods as defined in the NRC staff position letter reprinted in ELTR1 and the NRC SER for ELTR2 were followed for this EPU analysis. Any exceptions to the use of the code or conditions of the applicable SERs are noted in Table 1-1. The application of the computer codes in Table 1-1 is consistent with the current Browns Ferry licensing basis except where noted. 1.2.3 Approach The planned approach to achieving the higher power level consists of the change to the Browns Ferry licensing and design basis to increase the licensed power level to 3,952 MWt, consistent with the approach outlined in the CLTR, except as specifically noted, and with the approach outlined in ELTR1 for fuel-dependent evaluations. Consistent with the CLTR, the following plant-specific exclusions are exercised: No increase in maximum normal operating reactor dome pressure No increase to maximum licensed core flow No increase to currently licensed MELLLA upper boundary No change to source term methodology No new fuel product line introduction No change to fuel cycle length No additions to currently licensed operational enhancements The plant-specific evaluations are based on a review of plant design and operating data, as applicable, to confirm excess design capabilities; and, if necessary, identify required modifications associated with EPU. All changes to the plant-licensing basis have been identified. For specified topics, generic analyses and evaluations in the CLTR, or ELTR1 and ELTR2 as applicable, demonstrate plant operability and safety. The dispositions in the CLTR are based on a 20% increase of OLTP, which is equal to the requested power uprate for Browns Ferry. For this increase in power, the conclusions of system/component acceptability stated in the CLTR and ELTR2 are bounding and have been confirmed for Browns Ferry. The scope and depth of the evaluation results provided herein are established based on the approach in the CLTR and ELTR2 and unique features of the plant. The results of the following evaluations are presented: Reactor Core and Fuel Performance: Assessments that are dependent on the fuel design were performed by others and are included in the complementary FUSAR included as Attachment 8 to the LAR for power uprate. Reactor Coolant System (RCS) and Connected Systems: Evaluations of the NSSS components and systems have been performed at EPU conditions. These evaluations confirm the acceptability of the effects of the higher power and the associated change in 1-4

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) process variables (i.e., increased steam and feedwater (FW) flows). Safety-related equipment performance is the primary focus, but key aspects of reactor operational capability are also included. Engineered Safety Feature Systems: The effects of EPU power operation on the Containment, emergency core cooling system (ECCS), Standby Gas Treatment System (SGTS) and other ESFs have been evaluated for key events. The evaluations include the containment responses during limiting Anticipated Operational Occurrences (AOOs) and special events, ECCS-LOCA, and safety relief valve (SRV) containment dynamic loads. Control and Instrumentation: The control and instrumentation signal ranges and ALs for setpoints have been evaluated to establish the effects of the changes in various process parameters such as power, neutron flux, steam flow and FW flow. As required, evaluations have been performed to determine the need for any Technical Specification (TS) allowable value (AV) changes for various functions (e.g., main steam line (MSL) high flow isolation setpoints). Electrical Power and Auxiliary Systems: Evaluations have been performed to establish the operational capability of the plant electrical power and distribution systems and auxiliary systems to ensure that they are capable of supporting safe plant operation at the EPU power level. Power Conversion Systems: Evaluations have been performed to establish the operational capability of various non-safety Balance-of-Plant (BOP) systems and components to ensure that they are capable of delivering the increased power output, and/or the modifications necessary to obtain full EPU power. Radwaste Systems and Radiation Sources: The liquid and gaseous waste management systems (GWMSs) have been evaluated at limiting conditions for EPU to show that applicable release limits continue to be met during operation at higher power. The radiological consequences have been evaluated for EPU to show that applicable regulations have been met for the EPU power conditions. This evaluation includes the effect of higher power level on source terms, on-site doses and off-site doses, during normal operation. Reactor Safety Performance Evaluations: Assessments that are dependent on the fuel design were performed by others and are included in the complementary FUSAR included as Attachment 8 to the LAR for power uprate. Additional Aspects of EPU: High-energy line break (HELB) and environmental qualification (EQ) evaluations have been performed at bounding conditions for EPU to show the continued operability of plant equipment under EPU conditions. The effects of EPU on the Browns Ferry Probabilistic Risk Assessment (PRA) have been analyzed to demonstrate that there are no new vulnerabilities to severe accidents. 1-5

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 1.3 EPU Plant Operating Conditions 1.3.1 Reactor Heat Balance The operating pressure, the total core flow, and the coolant thermodynamic state characterize the thermal hydraulic performance of a BWR reactor core. The EPU values of these parameters are used to establish the steady state operating conditions and serve as initial and boundary conditions for the required safety analyses. The EPU values for these parameters are determined by performing heat (energy) balance calculations for the reactor system at EPU conditions. The reactor heat balance relates the thermal-hydraulic parameters to the plant steam and FW flow conditions for the selected core thermal power level and operating pressure. Operational parameters from actual plant operation are considered (e.g., steam line pressure drop) when determining the expected EPU conditions. The thermal-hydraulic parameters define the conditions for evaluating the operation of the plant at EPU conditions. The thermal-hydraulic parameters obtained for the EPU conditions also define the steady state operating conditions for equipment evaluations. Heat balances at appropriately selected conditions define the initial and boundary conditions for plant safety analyses. Figure 1-2 shows the EPU heat balance at 100% of EPU RTP and 100% rated core flow. Figure 1-3 shows the EPU heat balance at 102% of EPU RTP and 100% core flow with dome pressure at 1,070 psia. Table 1-2 provides a summary of the reactor thermal-hydraulic parameters for the OLTP, CLTP and EPU conditions. At EPU conditions, the maximum nominal operating reactor vessel dome pressure is maintained at the current value, which minimizes the need for plant and licensing changes. With the increased steam flow and associated non-safety BOP modifications, the current dome pressure provides sufficient operating turbine inlet pressure to assure good pressure control characteristics. 1.3.2 Reactor Performance Improvement Features The reactor performance improvement features and the equipment allowed to be out-of-service (OOS) are listed in Table 1-2. When limiting, the input parameters related to the performance improvement features or the equipment OOS have been considered in the safety analyses for EPU, and as applicable, will be included in the reload core analyses. The use of these performance improvement features and allowing for equipment OOS are allowed during EPU operation. Where appropriate, the evaluations that are dependent upon cycle length are performed for EPU assuming a 24-month fuel cycle length. 1.4 Summary and Conclusions This evaluation has covered an EPU to 120% of OLTP. The strategy for achieving higher power is to extend the MELLLA power/flow map region along the upper boundary extension. The Browns Ferry licensing bases have been reviewed to demonstrate how this uprate can be accommodated without a significant increase in the probability or consequences of an accident previously evaluated, without creating the possibility of a new or different kind of accident from 1-6

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) any accident previously evaluated, and without exceeding any existing regulatory limits or design allowable limits applicable to the plant which might cause a reduction in a margin of safety. The EPU described herein involves no significant hazard consideration. 1-7

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 1-1 Computer Codes Used For EPU Computer Version NRC Task Comments Code* or Approved Revision Nominal Reactor ISCOR 09 Y(2) NEDE-24011P Rev. 0 SER Heat Balance Reactor Pressure TGBLA 06 Y Vessel (RPV) DORTG 01 N (8) (9) Fluence Reactor Internal ISCOR 09 Y(2) NEDE-24011P Rev. 0 SER Pressure LAMB 07 (3) NEDE-20566-P-A Differences TRACG 02 Y NEDE-32176P Rev. 0 (RIPDs) NEDC-32177P Rev. 1 NRC TAC No. M90270 Reactor Vessel ANSYS 6.1 N (1) Integrity - Stress FatiguePro 3.01 N (1) and Fatigue Evaluation RPV Flow-Induced ANSYS 6 N (1) Vibration SAP4G07 07 N NEDO-10909 (1) Reactor BILBO 04V N/A NEDE-23504, February 1977 (1) Recirculation System Reactor Coolant Pressure Boundary TPIPE Various N (7) Piping Piping Components SAP4G07 07 N GE NEDO-10909 (1) Flow Induced Vibration Anticipated ODYN 10 Y NEDE-24154P-A Supplement. 1, Vol. 4 Transient Without STEMP 04 (5) Scram PANACEA 11 Y(4) NEDE-30130-P-A ISCOR 09 Y(2) NEDE-24011P Rev. 0 SER Containment SHEX 06 Y (6) System Response M3CPT 05 Y NEDO-10320, Apr. 1971 (NUREG-0661) LAMB 08 (3) NEDE-20566-P-A September 1986 Annulus ISCOR 09 Y(2) NEDE-24011P Rev. 0 SER Pressurization (AP) Station Blackout SHEX 06 Y (6) Fission Product ORIGEN2 2.1 N Isotope Generation and Depletion Code Inventory MS Piping TPIPE 16 N Structural Analysis Program (7) Analysis 1-8

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 1-1 Computer Codes Used For EPU (continued) Computer Version NRC Task Comments Code* or Approved Revision Plant Life Flow CHECWORKSTM 4.0 N Industry supported software to assist the utility Accelerated SFA industry in planning and implementing inspection Corrosion programs to prevent FAC failures.

The application of these codes to the EPU analyses complies with the limitations, restrictions, and conditions specified in the approving NRC SER where applicable for each code. The application of the codes also complies with the SERs for the EPU programs.

(1) Not a safety analysis code that requires NRC approval. The code application is reviewed and approved by GEH for Level-2 application and is part of GEHs standard design process. Also, the application of this code has been used in previous power uprate submittals. (2) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011P Revision 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, Reactor Core and Fuel Performance and LOCA applications is consistent with the approved models and methods. (3) The LAMB code is approved for use in ECCS-LOCA applications (NEDE-20566-P-A and NEDO-20566A), but no approving SER exists for the use of LAMB in the evaluation of reactor internal pressure differences or containment system response. The use of LAMB for these applications is consistent with the model description of NEDE-20566-P-A (Reference 6). (4) The physics code PANACEA provides inputs to the transient code ODYN. The improvements to PANACEA that were documented in NEDE-30130-P-A were incorporated into ODYN by way of Amendment 11 of GESTAR II (NEDE-24011-P-A). The use of TGBLA Version 06 and PANACEA Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S. A. Richards (NRC) to G. A. Watford (GE)

Subject:

Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods, (TAC No. MA6481), November 10, 1999. (5) The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool heatup. The use of STEMP was noted in NEDE-24222, Assessment of BWR Mitigation of ATWS, Volume I and II (NUREG-0460 Alternate No. 3) December 1, 1979. The code has been used in ATWS applications since that time. It has also recently been accepted in the NRC review of NEDC-32868P, GE14 Compliance with Amendment 22 of NEDE-24011-P-A (GESTAR). There is no formal NRC review and approval of STEMP. (6) The application of the methodology in the SHEX code to the containment response is approved by the NRC in the letter to G. L. Sozzi (GE) from A. Thadani (NRC), Use of the SHEX Computer Program and 1-9

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis, July 13, 1993 (Reference 7). (7) TPIPE is a linear elastic analysis of piping program used by TVA for analysis of the Main Steam (MS) and FW piping. TPIPE is not a safety analysis code that requires NRC approval. TVA validation and verification of the TPIPE program and related approval data is stored in TVA System ID 262127. The TPIPE program is described in the Browns Ferry UFSAR, Appendix C, Section C.3.7, and has been benchmarked against the NRC program EPIPE in accordance with the Standard Review Plan, NUREG-0800, Section 3.9.1.II and NUREG/CR-1677. TPIPE is TVAs program used for pipe analysis for all three units at Browns Ferry. (8) The use of DORTG was approved by the NRC through the letter from H. N. Berkow (USNRC) to G. B. Stramback (GE), Final Safety Evaluation Regarding Removal of Methodology Limitations for NEDC-32983P-A, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations (TAC No. MC3788), November 17, 2005. (9) Letter, S.A. Richards (USNRC) to G. A. Watford (GE), Amendment 26 to GE Licensing Topical Report NEDE-24011-P-A, GESTAR II - Implementing Improved GE Steady-State Methods (TAC No. MA6481), November 10, 1999. 1-10

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 1-2 OLTP, CLTP, and EPU Plant Operating Conditions Parameter OLTP Value5 CLTP Value1 EPU Value4 Thermal Power (MWt) 3,293 3,458 3,952 Vessel Steam Flow (Mlb/hr) 2 13.370 14.153 16.440 Full Power Core Flow Range Mlb/hr 76.9 to 107.6 83.0 to 107.6 101.5 to 107.6

        % Rated                                    75 to 105              81 to 105      99 to 105 Maximum Nominal Dome Pressure (psia)                   1,005                 1,050          1,050 Maximum Nominal Dome Temperature (F)                  547.0                 550.5          550.5 Pressure at Upstream Side of Turbine Stop 960                  1,000           983 Valve (TSV) (psia)

Full Power FW Flow (Mlb/hr) 13.330 14.103 16.390 Temperature (F) 377.0 381.7 394.5 Core Inlet Enthalpy (Btu/lb) 3 521.6 524.7 523.2 Reactor Recirculation System (RRS) Outlet 575ºF 575ºF 575ºF Design Temperature RRS Outlet Maximum Temperature 546ºF 550.5ºF 6 550.5ºF RRS Inlet Design Temperature 575ºF 575ºF 575ºF RRS Inlet Maximum Temperature 546ºF 550.5ºF 550.5ºF FW Nozzle Design Temperature 575ºF 575ºF 575ºF FW Nozzle Maximum Temperature 7 573ºF 573ºF 573ºF Main Steam (MS) Nozzle Design 575ºF 575ºF 575ºF Temperature MS Nozzle Maximum Temperature 546ºF 550.5ºF 550.5ºF Core Spray (CS) Nozzle Design Temperature 575ºF 575ºF 575ºF CS Nozzle Maximum Temperature 8 546ºF 550.5ºF 550.5ºF Notes:

1. Based on current reactor heat balance.

2. At normal FW heating.

3. At 100% core flow conditions.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public)

4. Performance improvement features and/or equipment Out-of-Service (OOS) that are included in EPU evaluations:

a. MELLLA

b. Increased Core Flow (ICF)

c. Single-loop Operation (SLO)

d. Final Feedwater Temperature Reduction (FFWTR), 55°F Temperature Reduction

e. APRM/RBM/Technical Specifications (ARTS)

f. 3% Main Steam Relief Valve (MSRV) Setpoint Tolerance

g. One MSRV OOS

h. Turbine Bypass Valves OOS (TBVOOS)

i. End-of-Cycle Recirculation Pump Trip (EOC RPT) OOS (RPTOOS)

j. Feedwater Heaters Out-of-Service (FWHOOS), 55°F Temperature Reduction

k. 24 Month Fuel Cycle

l. Power Load Unbalance OOS (PLUOOS)

5. All nozzle maximum pressures are the same as the maximum normal dome pressure and the design pressure, 1,250 psig, remains unchanged from OLTP to EPU.

6. Maximum nominal dome temperature is 550.5ºF.

7. FW OLTP, CLTP and EPU maximum temperature values are based on loss of feedwater pumps for 102%

rated thermal power conditions. During these times, nozzles will be filled with steam.

8. CS has no flow under normal operating conditions and the maximum temperature values correspond to the maximum nominal dome temperatures above.

1-12

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Figure 1-1 Power/Flow Operating Map for EPU 1-13

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Legend

  # = Flow, lbm/hr                     1050 H = Enthalpy, Btu/lbm                  P F = Temperature, F M = Moisture, %                                           Main Steam Flow          16.440E+06 #

P = Pressure, psia 1190.4 H
0.48 M
Carryunder = 0.35% 983 P
3952 Main Feed Flow MWt Wd= 100.0 % 16.524E+06 # 16.390E+06 #

524.1 H 370.5 H 370.2 H 530.1 F Total 394.8 F 394.5 F Core Flow 102.5E+06 h= 1.0 H # 1.333E+05 # 410.9 H 523.2 432.2 F H Cleanup Demineralizer System 5.000E+04 # Control Rod Drive 1.333E+05 # 48.0 H Feed Flow 523.1 H 77.0 F 529.3 F

Conditions at upstream side of TSV Core Thermal Power 3952.0 Pump Heating 10.6 Cleanup Losses -4.4 Other System Losses -1.1 Turbine Cycle Use 3957.1 MWt Figure 1-2 EPU Heat Balance - Nominal

(@ 100% Power and 100% Core Flow) 1-14

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Legend

  # = Flow, lbm/hr                      1070 H = Enthalpy, Btu/lbm                   P F = Temperature, F M = Moisture, %                                           Main Steam Flow          16.831E+06 #

P = Pressure, psia 1189.6 H
0.52 M
Carryunder = 0.35% 999 P
4031 Main Feed Flow MWt Wd= 99.5 % 16.914E+06 # 16.781E+06 #

526.1 H 372.7 H 372.4 H 531.7 F Total 396.9 F 396.6 F Core Flow 102.0E+06 h= 1.0 H # 1.333E+05 # 412.9 H 526.9 434.0 F H Cleanup Demineralizer System 5.000E+04 # Control Rod Drive 1.333E+05 # 48.0 H Feed Flow 525.1 H 77.0 F 530.9 F

Conditions at upstream side of TSV Core Thermal Power 4031.0 Pump Heating 10.4 Cleanup Losses -4.4 Other System Losses -1.1 Turbine Cycle Use 4035.9 MWt Figure 1-3 EPU Heat Balance - Overpressure Protection Analysis

(@ 102% Power and 100% Core Flow) 1-15

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public)

2. SAFETY EVALUATION 2.1 Materials and Chemical Engineering 2.1.1 Reactor Vessel Material Surveillance Program Regulatory Evaluation The reactor vessel material surveillance program provides a means for determining and monitoring the fracture toughness of the reactor vessel beltline materials to support analyses for ensuring the structural integrity of the ferritic components of the reactor vessel.

The NRCs acceptance criteria are based on (1) General Design Criterion (GDC)-14, insofar as it requires that the reactor coolant pressure boundary (RCPB) be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; (2) GDC-31, insofar as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a non-brittle manner and the probability of a rapidly propagating fracture is minimized; (3) 10 CFR Part 50, Appendix H, which provides for monitoring changes in the fracture toughness properties of materials in the reactor vessel beltline region; and (4) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR Part 50, Appendix H. Specific NRC review criteria are contained in Standard Review Plan (SRP) Section 5.3.1 and other guidance provided in Matrix 1 of RS-001. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the Atomic Energy Commission (AEC) published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed GDC published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General 2-1

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-9 and 33. Final GDC-31 is applicable to Browns Ferry as described in Browns Ferry Nuclear Plant (BFN), Unit 1- Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-484), dated December 18, 2013 (Reference 8), Browns Ferry Nuclear Plant (BFN), Unit 2 - Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-491), dated June 19, 2014 (Reference 9), and Browns Ferry Nuclear Plant, Unit 3 - Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-494), dated January 27, 2015 (Reference 10). The Reactor Vessel Material Surveillance Program is described in Browns Ferry UFSAR Section 4.2, Reactor Vessel and Appurtenances Mechanical Design, and the Bases to TS 3.4.9, RCS Pressure and Temperature (P/T) Limits. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Reactor Vessel Material Surveillance Program is documented in NUREG-1843, Section 3.0.3.2.19. Technical Evaluation The RPV fracture toughness evaluation process is described in Section 2.1.2. RPV embrittlement is caused by neutron exposure of the wall adjacent to the core including the regions above and below the core that experience fluence greater than or equal to 1 x 1017 n/cm2. This region is defined as the beltline region. Operation at EPU conditions results in a higher neutron flux, which increases the integrated fluence over the period of plant life. The surveillance program consists of three capsules for each unit. No capsules have been removed from the Browns Ferry Unit 1 vessel. Therefore, three capsules remain in the vessel, and have been there since plant startup. One capsule containing Charpy specimens was removed from the Browns Ferry Unit 2 vessel after 8.2 effective full power years (EFPY) of operation (end of Fuel Cycle 7), tested, reconstituted, and placed into the vessel during the Unit 2 Cycle 8 refueling outage. A second capsule was removed after 22.9 EFPY of operation (end of Fuel Cycle 16), tested, and analyzed. The remaining one of the three original capsules has been in the reactor vessel since plant startup. The first Browns Ferry Unit 3 capsule was removed from the vessel during the Fuel Cycle 8 outage, but was not tested. Browns Ferry Units 1, 2, and 3 are part of the BWR Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) currently administered by Electric Power Research Institute (EPRI) and will comply with the withdrawal schedule specified for representative or surrogate surveillance capsules that now represent each unit. Therefore, the 10 CFR 50 Appendix H surveillance capsule schedule for the ISP governs. Implementation of EPU has no adverse effect on the BWRVIP withdrawal schedule. 2-2

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The maximum normal operating dome pressure for EPU is unchanged from that for CLTP thermal power operation. Therefore, the hydrostatic and leakage test pressures are acceptable for EPU. Operation with EPU does not have an adverse effect on the reactor vessel fracture toughness because the Unit 1, 2, and 3 vessels remain in compliance with the regulatory requirements as demonstrated in Section 2.1.2. Conclusion TVA has evaluated the effects of the proposed EPU on the reactor vessel surveillance withdrawal schedule and has addressed changes in neutron fluence and their effects on the schedule. The evaluation indicates that the material surveillance program will continue to meet the requirements of 10 CFR Part 50, Appendix H, and 10 CFR 50.60, and will ensure continued compliance with draft GDCs-9 and 33, and final GDC-31 in this respect following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the reactor vessel material surveillance program. 2.1.2 Pressure-Temperature Limits and Upper-Shelf Energy Regulatory Evaluation Pressure and Temperature (P-T) limits are established to ensure the structural integrity of the ferritic components of the RCPB during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests. The NRCs acceptance criteria for P-T limits are based on (1) GDC-14, insofar as it requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; (2) GDC-31, insofar as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a non-brittle manner and the probability of a rapidly propagating fracture is minimized; (3) 10 CFR Part 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB; and (4) 10 CFR 50.60, which requires compliance with the requirements of 10 CFR Part 50, Appendix G. Specific NRC review criteria are contained in SRP Section 5.3.2 and other guidance provided in Matrix 1 of RS-001. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this 2-3

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-9. Final GDC-31 is applicable to Browns Ferry as described in Browns Ferry Nuclear Plant (BFN), Unit 1- Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-484), dated December 18, 2013 (Reference 8), Browns Ferry Nuclear Plant (BFN), Unit 2 - Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-491), dated June 19, 2014 (Reference 9), and Browns Ferry Nuclear Plant, Unit 3 - Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-494), dated January 27, 2015 (Reference 10). The Pressure-Temperature Limits and Upper Shelf Energy is described in Browns Ferry UFSAR Section 4.2, Reactor Vessel and Appurtenances Mechanical Design, and the Bases to TS 3.4.9, RCS Pressure and Temperature (P/T) Limits. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluations associated with Pressure-Temperature Limits and Upper-Shelf Energy are documented in NUREG-1843, Sections 4.2.1 and 4.2.5. RCS Pressure and Temperature (P/T) Limits The Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits have been developed for EPU conditions and have been submitted to the NRC for approval as follows:

a. The Browns Ferry Unit 1 change was submitted to the NRC on December 18, 2013 and approved in License Amendment No. 287 on February 2, 2015.

b. The Browns Ferry Unit 2 change was submitted to the NRC on June 19, 2014 and approved in License Amendment No. 314 on June 2, 2015.

c. The Browns Ferry Unit 3 change was submitted to the NRC on January 27, 2015 and approved in License Amendment No. 278 on January 7, 2016.

2-4

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 3.2.1 of the CLTR addresses the effect of EPU on Pressure-Temperature (P-T) Limits and Upper-Shelf Energy (USE). The results of this evaluation are described below. As explicitly stated in Section 3.2.1 of the CLTR, EPU may result in a higher operating neutron flux at the vessel wall, consequently increasing the integrated flux over time (neutron fluence). The neutron fluence is recalculated using the NRC-approved GEH neutron fluence methodology (Reference 12). This method is consistent with Regulatory Guide (RG) 1.190 (Reference 13) and utilizes a more representative fluence than previous methods. Browns Ferry meets all CLTR dispositions. AREVA fuel will be used at Browns Ferry when EPU is implemented; however, the basis for the RPV flux is the GEH analysis using GE14 fuel. AREVA independently evaluates the bounding nature of the GEH results for the peak flux values for RPV inner diameter, and internals (shroud diameter, top guide, core plate) in FUSAR Section 2.1.2. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Fracture Toughness Plant Specific Disposition The revised fluence is used to evaluate the vessel against the requirements of 10 CFR 50, Appendix G. The results of these evaluations indicate that: (a) The reduction in USE, using Equivalent Margin methods, demonstrates that there is an equivalent margin of safety against fracture for RPV materials such that it will remain qualified with respect to 10 CFR 50 Appendix G criterion for the design life of the vessel. The maximum decrease in USE for the beltline plate materials is 16% ((( ))) for Unit 2 at 48 EFPY. The maximum decrease in USE for the beltline weld materials is 33.5% ((( ))) for Unit 1 at 38 EFPY. These values are provided in Tables 2.1-1a through 2.1-1c. (b) The beltline material Reference Temperature of Nil-Ductility Transition (RTNDT) remains below 200°F. The N-16 water level instrumentation nozzle is included in the evaluation. (c) The Technical Specification P-T curves were revised to incorporate the methodology of the GEH P-T curve LTR (Reference 14) and the ISP Browns Ferry Unit 2 second surveillance capsule results. The fracture toughness evaluation included the effects of the N-16 water level 2-5

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) instrumentation nozzle that occurs within the beltline region. The hydro test pressure for EPU is the minimum nominal operating pressure. (d) The end of life (EOL) shift is increased, and consequently, results in an increase in the Adjusted Reference Temperature (ART), which is the initial RTNDT plus the shift. These values are provided in Tables 2.1-2a through 2.1-2c. (e) The EOL beltline circumferential weld material mean RTNDT remains bounded by the requirements of Generic Letter (GL) 98-05 (Reference 15), BWRVIP-05 (References 16 and 17), and BWRVIP 74-A (Reference 18). This comparison is provided in Table 2.1-3. (f) GEH P-T limit curves include an adjustment for the column of water in a full RPV. The Browns Ferry EPU is a constant pressure power uprate, which, by definition, does not change the pressure from that considered for CLTP. The pressure head for Browns Ferry for a full vessel is 31.6 psig. (g) ISP plate and weld materials have been considered in development of the beltline ART as defined in BWRVIP-135. In accordance with the guidance from BWRVIP-135 and the methodology provided in RG 1.99 Revision 2 (Reference 19), the surveillance materials are considered in the development of the P-T limit curves for Units 1 and 2, but are not considered in the development of the P-T limit curves for Unit 3. (h) The generic pressure test P-T limit curve is based on dimensions cited in NEDC-33178P-A, Revision 1 (Reference 14). GEH P-T limit curves are considered acceptable for plant-specific application when it is demonstrated that the plant-specific dimensions are bounded by the generic dimensions, as is the case for Browns Ferry Units 1, 2, and 3. (i) Ferritic piping within the RCPB has not been replaced since plant start-up. Therefore, Browns Ferry meets all CLTR dispositions for fracture toughness. Conclusion TVA has evaluated the effects of the proposed EPU on the P-T limits for the plant and addressed changes in neutron fluence and their effects on the P-T limits. Revised P-T curves have been approved by the NRC per 10 CFR 50.90. As such, TVA concludes that the changes in neutron fluence and their effects on the P-T limits have been adequately addressed. TVA further concludes it has demonstrated the validity of the proposed P-T limits for operation under the proposed EPU conditions. Based on this, TVA concludes the P-T limits will continue to meet the requirements of 10 CFR 50, Appendix G, and 10 CFR 50.60 and will enable Browns Ferry to continue to comply with the current licensing basis following implementation of the proposed EPU. Therefore, TVA finds the proposed EPU acceptable with respect to the proposed P-T limits. 2-6

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.1.3 Reactor Internal and Core Support Materials Regulatory Evaluation The reactor internals and core supports include structures, systems, and components (SSCs) that perform safety functions or whose failure could affect safety functions performed by other SSCs. These safety functions include reactivity monitoring and control, core cooling, and fission product confinement (within both the fuel cladding and the RCS). The NRCs acceptance criteria for reactor internal and core support materials are based on GDC-1 and 10 CFR 50.55a for material specifications, controls on welding, and inspection of reactor internals and core supports. Specific NRC review criteria are contained in SRP Section 4.5.2 and Boiling Water Reactor Vessel and Internals Project (BWRVIP) -26. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-1. The reactor internals and core supports are described in Browns Ferry UFSAR Section 3.3, Reactor Vessel Internals Mechanical Design. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for 2-7

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Reactor Internal and Core Support Materials is documented in NUREG-1843, Section 2.3.1.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 10.7 of the CLTR addresses the effect of EPU on reactor internal and core support materials. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Irradiated Assisted Stress Corrosion Cracking Plant Specific Disposition As explicitly stated in Section 10.7 of the CLTR, the increase in irradiation of the core internal components influences Irradiation-Assisted Stress Corrosion Cracking (IASCC). The longevity of most equipment is not affected by EPU. ((

                         )) A plant-specific analysis of IASCC is required for EPU.

The reactor internal and core support materials evaluation included the materials specifications and mechanical properties, welds, weld controls, Non-destructive examination (NDE) procedures, corrosion resistance, and susceptibility to degradation. This evaluation of the reactor internals and core supports includes Structures, Systems, and Components (SSCs) that perform safety functions or whose failure could affect safety functions performed by other SSCs. None of these requirements, specifications, or controls is changed as a result of the EPU; therefore, these continue to be acceptable. Browns Ferry has a procedurally controlled program for the augmented NDE of selected RPV internal components in order to ensure their continued structural integrity. The inspection techniques utilized are primarily for the detection and characterization of service-induced, surface-connected planar discontinuities, such as Intergranular Stress Corrosion Cracking (IGSCC) and IASCC, in welds and in the adjacent base material. Browns Ferry belongs to the BWR Vessel and Internals Project (BWRVIP) organization and implementation of the procedurally controlled program is consistent with the BWRVIP issued documents. The inspection strategies recommended by the BWRVIP consider the effects of fluence on applicable components and are based on component configuration and field experience. The inspection program is modified for the inspection of the core plate bolts in accordance with Deviation 2-8

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Disposition No. DD-2011-01 (Reference 20). The inspection program is enhanced for additional inspections of the core plate beyond what is required by the BWRVIP. Components selected for inspection include those that are identified as susceptible to in-service degradation and those where augmented examination is conducted for verification of structural integrity. These components have been identified through the review of NRC Inspection and Enforcement Bulletins (IEBs), BWRVIP documents, and recommendations provided by General Electric Services Information Letters (GE SILs). The inspection program provides performance frequency for NDE and associated acceptance criteria. Components inspected include the following: Core Spray (CS) piping Core plate Core spray spargers Core shroud and core shroud support Jet pumps and associated components Top guide Lower plenum Vessel ID attachment welds Instrumentation penetrations Steam dryer drain channel welds FW spargers In-core flux monitoring guide tubes Control rod guide tubes Inspected components are considered as being potentially susceptible to IASCC if the end-of-life fluence is in excess of 5 x 1020 n/cm2 (E> 1 MeV). Three components have been identified as being potentially susceptible to IASCC, based upon the projected 54 EFPY fluence for Unit 1: (1) Top Guide, 2.06 x 1022 n/cm2 (E> 1 MeV); (2) Shroud, 5.34 x 1021 n/cm2 (E> 1 MeV); and (3) Core Plate, 7.33 x 1020 n/cm2 (E> 1 MeV). Three components have been identified as being potentially susceptible to IASCC, based upon the projected 52 EFPY fluence for Units 2 and 3: (1) Top Guide, 1.98 x 1022 n/cm2; (2) Shroud, 5.15 x 1021 n/cm2; and (3) Core Plate, 7.07 x 1020 n/cm2. The BWRVIP inspection recommendations that provide the scope, sample size, inspection method, and frequency of examination used to manage the effects of IASCC are as follows: Top Guide (BWRVIP-26-A and BWRVIP-183) (References 21 and 22) Shroud (BWRVIP-76) (Reference 23) 2-9

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Core Plate (BWRVIP-25) (Reference 24) Continued implementation of the current procedure program assures the prompt identification of any degradation of reactor vessel internal components experienced during EPU operating conditions. To mitigate the potential for IGSCC and IASCC, Browns Ferry utilizes hydrogen water chemistry and noble metals applications. Reactor vessel water chemistry conditions are also maintained consistent with the EPRI and established industry guidelines. The service life of most equipment is not affected by EPU. The peak fluence increase experienced by the reactor internals does not represent a significant increase in the potential for IASCC. The current inspection strategy for the reactor internal components is expected to be adequate to manage any potential effects of EPU. No relevant indications have been observed during in the most recent grid beam inspection of Browns Ferry Units 1, 2, or 3. Analysis of the core plate bolts was conducted as part of the Time Limiting Aging Analysis (TLAA) for the Browns Ferry license renewal, per Reference 25. Therefore, Browns Ferry meets all CLTR dispositions for IASCC. Conclusion TVA has evaluated the effects of the proposed EPU on the integrity of reactor internal and core support materials. The evaluation indicates that the reactor internal and core support materials will continue to be acceptable and will continue to meet the requirements of draft GDC-1 and 10 CFR 50.55a. Therefore, the proposed EPU is acceptable with respect to reactor internal and core support materials. 2.1.4 Reactor Coolant Pressure Boundary Materials Regulatory Evaluation The RCPB defines the boundary of systems and components containing the high-pressure fluids produced in the reactor. The NRCs acceptance criteria for RCPB materials are based on (1) 10 CFR 50.55a and GDC-1, insofar as they require that SSCs important to safety be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (3) GDC-14, insofar as it requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; (4) GDC-31, insofar as it requires that the RCPB be designed with margin sufficient to assure that, under specified conditions, it will behave in a non-brittle manner and the probability of a rapidly propagating fracture is minimized; and (5) 10 CFR Part 50, Appendix G, which specifies fracture toughness requirements for ferritic components of the RCPB. 2-10

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Specific NRC review criteria are contained in SRP Section 5.2.3 and other guidance provided in Matrix 1 of RS-001. Additional review guidance for primary water stress-corrosion cracking of dissimilar metal welds and associated inspection programs is contained in GL 97-01, Information Notice 00-17, Bulletins 01-01, 02-01, and 02-02. Additional review guidance for thermal embrittlement of cast austenitic stainless steel components is contained in a letter from C. Grimes, NRC, to D. Walters, Nuclear Energy Institute, dated May 19, 2000. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-1, 2, and 9. Final GDC-31 is applicable to Browns Ferry as described in Browns Ferry Nuclear Plant (BFN), Unit 1- Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-484), dated December 18, 2013 (Reference 8), Browns Ferry Nuclear Plant (BFN), Unit 2 - Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-491), dated June 19, 2014 (Reference 9), and Browns Ferry Nuclear Plant, Unit 3 - Application to Modify Technical Specification 3.4.9, RCS Pressure and Temperature (P/T) Limits (BFN TS-494), dated January 27, 2015 (Reference 10). The Reactor Coolant Pressure Boundary Materials is described in Browns Ferry UFSAR Sections 4.2, Reactor Vessel and Appurtenances Mechanical Design, and 4.3, Reactor Recirculation System. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials 2-11

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the reactor coolant pressure boundary is documented in NUREG-1843, Sections 2.3.1 and 4.3. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section10.7 of the CLTR addresses the effect of EPU on RCPB materials. The temperature and flow increase experienced by the RCPB does not represent a significant increase in the potential for IGSCC. Other degradation mechanisms are addressed in other sections of this report. Fracture toughness of the vessel components is addressed in Section 2.1.2. Flow-Accelerated Corrosion (FAC) for the plant is addressed in Section 2.1.6. The structural evaluation of the RCPB piping is addressed in Section 2.2.2.2.1. Flow Induced Vibration (FIV) for the safety-related piping components is addressed in Section 2.2.2.1.3. Therefore, the current inspection strategy for the RCPB is adequate to manage any potential effects of EPU. The Browns Ferry In-service Inspection (ISI) program for reactor coolant pressure boundary piping is in accordance with American Society of Mechanical Engineers (ASME) Section XI coupled with the augmented program for reactor coolant piping based on Generic Letter 88-01 (Reference 26), NUREG-0313 (Reference 27) and BWRVIP-75-A (Reference 28). The inspection techniques and NDE procedures utilized for ultrasonic examinations are qualified to the requirements of Appendix VIII of ASME Section XI (as implemented by the EPRI Performance Demonstration Initiative Program) for the detection and characterization of service-induced, surface-connected planar discontinuities, such as IGSCC. Continued implementation of the current program assures the prompt identification of any degradation of RCPB components experienced during EPU operating conditions. The augmented inspection program is designed to detect potential degradation from IGSCC. For IGSCC to occur, three conditions must be present: (1) a susceptible material; (2) the presence of residual or applied tensile stress (such as from welding); and (3) aggressive environment. Operation at EPU conditions results in an insignificant change to temperature and flow conditions for portions of the RCPB piping and does not affect the other susceptibility factors associated with IGSCC. This is consistent with the conclusions presented in Section 3.6.1 of ELTR2. The design of the RCPB piping and safe ends has been modified to reduce the amount of installed IGSCC susceptibility material. Table 2.1-9a lists the materials used in the Browns Ferry RCPB piping. 2-12

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The RCPB weldments have been categorized and inspected in accordance with NUREG-0313 (Reference 27) and BWRVIP-75-A (Reference 28). Table 2.1-9b depicts the number of welds by category in each unit. The two Category G welds per unit are physically located inside containment penetrations, which prohibits direct physical examinations of the welds. Approval of alternative in-service inspection methods has been obtained and is being followed with respect to the Category G welds. The nuclear industry has established that initiation and growth of IGSCC in stainless steel piping welds results from the combination of weld residual stress, an oxidizing environment, and a susceptible material. As described above, TVA has employed the use of IGSCC-resistant replacement material, applied weld stress improvement, and reduced the oxidizing environment with HWC. Operation at a higher power level will result in a slightly higher oxygen generation rate due to radiolysis of water; however, coolant chemistry will continue to be strictly controlled and maintained within specified limits. Implementation of EPU will not adversely affect the causative factors for IGSCC and, as such, the current established inspection and mitigation programs are adequate to support implementation of EPU. Several IGSCC mitigation processes have been applied to Browns Ferry to reduce the RCPB components susceptibility to IGSCC. Browns Ferry was designed, fabricated, and constructed with IGSCC addressed in most welds by one of three methods: (1) corrosion resistant materials; (2) solution heat treatment; or (3) clad with resistant materials. For the weldments where these three processes were not used, stress improvement processes were applied to reduce IGSCC susceptibility. Stress improvement processes and original construction processes used for IGSCC resistance are not affected by EPU. Also, Browns Ferry has implemented hydrogen water chemistry with noble metals, which reduces the potential for IGSCC of RCPB components. In the reactor core, the bulk dissolved oxygen concentration depends on the amount of oxygen generated through radiolysis and the amount consumed in the recombination reaction with reactor water dissolved hydrogen. The rate of radiolytic generation is directly dependent on reactor power (neutron and gamma flux levels) and the reactor water dissolved hydrogen concentration is directly dependent on the rate of hydrogen injection to feedwater. As the rate of radiolytic generation of oxygen increases with higher EPU power levels, the hydrogen injection rate to feedwater will be increased proportionate to the increased feedwater flow rate. As such, the EPU feedwater hydrogen concentration will be the same as the CLTP feedwater hydrogen concentration. The EPU predicted hydrogen-to-feedwater injection rate increases by less than three scfm from CLTP, and the EPU predicted oxygen-to-offgas injection rate increases by less than 1.5 scfm from CLTP. These increases are well within the capacity of the existing Browns Ferry HWC systems. Monitoring of HWC system parameters will be performed under the existing site chemistry programs to ensure required injection rates for IGSCC mitigation at EPU conditions. 2-13

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) In addition to HWC, all Browns Ferry units have implemented the NobleChem' (NMCA) process, with Browns Ferry Units 1 and 3 currently applying the annual On-line NobleChem' (OLNC) injection process and Browns Ferry Unit 2 planning to transition from the Classic NMCA process to OLNC in 2015. The NobleChem' processes are used in conjunction with HWC injection to feedwater to achieve IGSCC mitigation of reactor piping and internals at lower feedwater hydrogen addition rates than would be required for mitigation strategies that employ HWC only (e.g., Moderate HWC). Implementation of these programs at the Browns Ferry units post-EPU will continue to be performed in accordance with the recommendations of the applicable EPRI BWRVIP guidelines and experience reports (References 29 through 32). The primary parameters monitored for IGSCC mitigation at Browns Ferry are catalyst loading and Electrochemical Potential (ECP). The H2:O2 molar ratio (from the radiolysis/ECP model), hydrogen injection rate and reactor water oxygen concentration are secondary parameters monitored for IGSCC mitigation. These monitoring methods, currently employed or soon to be employed at all Browns Ferry units, will remain effective at EPU conditions. Conclusion TVA has evaluated the effects of the proposed EPU on the integrity of RCPB materials. The evaluation indicates that the RCPB materials will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of draft GDCs-1, 2, and 9, final GDC-31, 10 CFR Part 50, Appendix G, and 10 CFR 50.55a. Therefore, the proposed EPU is acceptable with respect to RCPB materials. 2.1.5 Protective Coating Systems (Paints) - Organic Materials Regulatory Evaluation Protective coating systems (paints) provide a means for protecting the surfaces of facilities and equipment from corrosion and contamination from radionuclides and also provide wear protection during plant operation and maintenance activities. The NRCs acceptance criteria for protective coating systems are based on (1) 10 CFR Part 50, Appendix B, which states quality assurance requirements for the design, fabrication, and construction of safety-related SSCs and (2) Regulatory Guide 1.54, Revision 1, for guidance on application and performance monitoring of coatings in nuclear power plants. Specific NRC review criteria are contained in SRP Section 6.1.2. Browns Ferry Current Licensing Basis The Browns Ferry current licensing basis regarding coatings is described in TVA letter to the NRC 120-day response to GL 98-04, dated November 10, 1998, Browns Ferry Nuclear Plant (BFN), Sequoyah Nuclear Plant (SQN), and Watts Bar Nuclear Plant (WBN), 120-Day Response Generic Letter (GL) 98-04, Potential for Degradation of the ECCS and the Containment Spray System (CSS) After a Loss-of-Coolant Accident (LOCA) Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment, Dated July 14, 1998. (Reference 33) 2-14

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Technical Evaluation The TVA protective coating program (Reference 34) 1) lists the coating systems approved for use in TVA nuclear plants; 2) separates them into service levels (SLs) designated SL-I, SL-II, SL-III; 3) provides the temperature and radiation qualification for SL-I; 4) defines the suitable application (i.e., dry, high humidity, immersion); and 5) defines proper surface preparation and application techniques. All SL-I coatings used at Browns Ferry are qualified for Design Basis Accident (DBA) conditions of temperature, pressure, radiation and chemical effects which bound worst case conditions at EPU. SL-I coating is required in the primary containment. Regulatory requirements such as 10 CFR 50 Appendix B, Regulatory Guide 1.54, GL 98-04, NUREG-1801, Information Notices, IE circulars, industry standards American Society for Testing and Materials (ASTM) D5144, ASTM D3843, ANSI 5.12, EPRI 1003102, and the TVA Nuclear Quality Assurance Manual (NQAM) are promulgated through Reference 34. The Browns Ferry Service Level I coatings are subject to the requirements of Regulatory Guide 1.54 - 1973 (Reference 35), American National Standard Institute (ANSI) N101.2 - 1972 (Reference 36) and ANSI N101.4 - 1972 (Reference 37). The qualification testing for Service Level I coatings used for new applications or repair/replacement activities inside containment meets the applicable requirements contained in the standards and regulatory commitments listed above. At EPU, the accumulated gamma dose for the DBA-LOCA is 1.5E8 Rad and is bounded by the SL-I coating qualification level of 1.0E9 Rad. At EPU, the peak drywell pressure and temperature for all LOCA events (See Table 2.6-1) are 50.9 psig (peak value for DBA-LOCA) and 336.9°F (steam line break LOCA), which are bounded by the SL-I coating qualification level of 70 psig pressure and 340°F temperature. The chemical constituency of the primary containment post-LOCA does not change as a result of EPU. The Service Level I coatings approved for use at Browns Ferry and applied inside containment are listed in the table below: 2-15

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Qualification Qualification Qualification Coating System Temperature for Application Dose (rads) (ºF) Immersion Unit 1 torus above water Valspar 78(1) 9 line tie in band 340 1 x 10 Yes AKA Vygard 78 Unit 2 and 3 torus D6/Amercoat 90N 340 1 x 109 No Drywell liner 9 Amerlock 400 NT 340 1 x 10 No Drywell structural steel UT-15 340 1 x 109 Yes Torus underwater repair 9 Kolor-Poxy 340 1 x 10 Unit 1 torus immersion Yes 6548/7107 zone Bio-Dur 561 340 1 x 109 Yes Torus underwater repair (1) 9 Plasite 340 1 x 10 Yes Unit 1 vapor space Note:

1. Coating system has been discontinued and is no longer applied. However, the coating is still resident inside the primary containment.

The Service Level I protective coating systems used inside the containment were evaluated for their continued suitability for and stability under DBA-LOCA and HELB conditions, considering radiation, temperature, pressure, and chemical effects at EPU conditions. The Harsh Environmental Data drawings and supporting calculations currently include the effect of life extension and EPU. Browns Ferry inspects the containment coating in accordance with plant procedures each refueling outage looking for failed or damaged coating. Coating conditions monitored by this program include checking for cracking, blistering, flaking, scaling, peeling, rust through, tiger striping, discoloration, embrittlement or mechanical damage. Any failed or damaged coating is remediated in accordance with plant procedures. The condition assessments and resulting repair, replacement, or removal activities ensure that the amount of coatings subject to detachment from the substrate during a LOCA is minimized to ensure post-accident operability of the ECCS suction strainers. The inspection of the coating in the immersion zone of the torus is coordinated with the desludging of the torus which is frequency based. Inspection of the immersion zone has typically occurred every second or third outage. Uncontrolled coatings are also identified and tracked to ensure the amount of uncontrolled coating which could contribute to ECCS strainer blockage is maintained below the established limit (157 ft2) used in the design of the replacement ECCS suction strainers (Reference 38) installed in the Browns Ferry units. EPU does not change this limit. In accordance with 2-16

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Reference 34 requirements, Browns Ferry inspects SL-I coatings during each refueling outage. This inspection periodicity does not change with EPU. Based on the conservative analysis summarized above, Browns Ferry has determined that reasonable assurance exists that when properly applied and maintained, the SL-I systems used in the primary containment will not detach under normal or accident conditions with the plant operating at EPU conditions. Conclusion TVA has evaluated the effects of the proposed EPU on the protective coatings. The evaluation indicates that the protective coatings will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of 10 CFR Part 50, Appendix B. Therefore, the proposed EPU is acceptable with respect to protective coatings. 2-17

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.1.6 Flow-Accelerated Corrosion Regulatory Evaluation FAC is a corrosion mechanism occurring in carbon steel components exposed to flowing single- or two-phase water. Components made from stainless steel are immune to FAC, and FAC is significantly reduced in components containing small amounts of chromium or molybdenum. The rates of material loss due to FAC depend on velocity of flow, fluid temperature, steam quality, oxygen content, and pH. During plant operation, control of these parameters is limited and the optimum conditions for minimizing FAC effects, in most cases, cannot be achieved. Loss of material by FAC will, therefore, occur. Browns Ferrys FAC program is based on NUREG-1344, GL 89-08, and the guidelines in Electric Power Research Institute (EPRI) Report NSAC-202L-R4. It consists of predicting loss of material using the CHECWORKS' computer code, and visual inspection and volumetric examination of the affected components. The NRCs acceptance criteria are based on the structural evaluation of the minimum acceptable wall thickness for the components undergoing degradation by FAC. Browns Ferry Current Licensing Basis The Browns Ferry program for addressing Flow Accelerated Corrosion is described in a TVA letter, dated July 19, 1989, Response to Generic Letter 89 Erosion/Corrosion-Induced Pipe Wall Thinning (Reference 39). This response provided information regarding administrative controls, procedures, and engineering activities associated with this program. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the FAC program is documented in NUREG-1843, Section 3.0.3.2.9. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 10.7 of the CLTR addresses the effect of EPU on FAC. Browns Ferry meets all CLTR dispositions. The results of this evaluation are described below. Browns Ferry Topic CLTR Disposition Result Meets CLTR Flow Accelerated Corrosion Plant Specific Disposition 2-18

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The CLTR states that the increase in steam and FW flow rate as a result of EPU influence FAC. In order to monitor and control FAC, Browns Ferry maintains an effective FAC program. The EPU implementation at Browns Ferry will change a number of system water and steam flow rates, temperatures, and enthalpies, in turn changing dissolved oxygen concentration. All these factors affect FAC susceptibility status and FAC wear rates. As a result of EPU operating conditions, some lines will experience accelerated rates of FAC, while others will have reduced rates. It should be noted that no lines that were previously non-susceptible to FAC (as defined by the EPU heat balance) will become susceptible due to EPU operating conditions. ((

                                             )) The FAC program will not significantly change for EPU.

The FAC program at Browns Ferry is based on: NRC I&E Bulletin 87-01, Thinning Pipe Walls in Nuclear Power Plants (Reference 40) Generic Letter 89-08, Erosion/Corrosion-Induced Pipe Wall Thinning (Reference 39) EPRI NSAC-202L-R4, Recommendations for an Effective Flow Accelerated Corrosion Program, Revision 4, November 2014 (Reference 41) With regard to EPRI NSAC-202L-R4, the choice of the method for detecting and evaluating the effect of FAC on a component is dependent on the type of component and its history. Results of the evaluation reveal if the component will remain above minimum allowable wall thickness throughout the next operating cycle and what the predicted minimum wall thickness will be at the end of the operating cycle. Additionally, the evaluation shows the remaining service life of the component (based on the calculated minimum allowable wall thickness) and the Next Scheduled Inspection (NSI) outage. The NSI is an outage prior to the time that the component reaches minimum allowable wall thickness. Component wall thickness is analyzed using minimum wall thickness according to the Browns Ferry design methodology and acceptable only if it meets all the design requirements of Browns Ferry. The Browns Ferry FAC program monitors all FAC susceptible piping - both small bore and large bore - to ensure the structural integrity and functionality are maintained. FAC susceptible piping can be divided into two categories: lines that meet the requirements to be modeled using CHECWORKS' Steam/Feedwater Application (SFA), and those that do not. For those that meet the requirements, Browns Ferry uses CHECWORKS' SFA, in conjunction with volumetric examination to predict FAC wear rates and remaining service life for components in single phase and two phase systems. The FAC susceptible lines that do not meet the minimum requirements for modeling and analysis by CHECWORKS' SFA are referred to as Susceptible Non-Modeled (SNM). This group is comprised of lines with unknown or widely varying operating conditions that prevent the 2-19

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) development of accurate predictive models. It includes bypass lines, recirculation lines, vent lines, high level dumps, and socket welded piping. Some small bore piping and piping susceptible to wall thinning mechanisms other than FAC are also included in this group. Selection of this piping for inspection is typically the result of industry experience, Browns Ferry experience, or engineering judgment. One of the most import aspects of the Browns Ferry FAC program is the proper selection of locations for FAC inspection and subsequent replacement of degraded piping. This is accomplished using the following (detailed in Table 2.1-8): CHECWORKS' SFA predictive wear analysis Susceptibility ranking of SNM piping Operating Experience (OE) Browns Ferry-specific experience Trending of historical inspection data Sound engineering judgment combining all of the above The proposed EPU may affect the following aspects of the Browns Ferry FAC program. FAC System Susceptibility Evaluation - This may include the addition of new lines in the FAC program based on changes in operating conditions as indicated in the heat balance. Wear rates - changes in operating conditions will result in some components wearing at an accelerated rate, while others will wear at a slower rate. Selection of component inspection and replacement locations and subsequent evaluation of inspection results (trending) - there could be a short-term increase in the number of inspections performed. These are evaluated as follows: FAC System Susceptibility Evaluation Browns Ferry performed a system susceptibility screening based on the revised EPU heat balance and determined that no additional lines were required to be added to the FAC program. Wear Rates - CHECWORKS' SFA Model Update for EPU The proposed EPU will result in changes to several variables that may directly influence FAC wear rates. The variables include operating temperature, steam quality, velocity and oxygen content. To account for these changes, Browns Ferry updated the affected parameters in the CHECWORKS' SFA model based on the EPU heat balance. Tables 2.1-4a, 4b and 4c contains a listing of the CHECWORKS' SFA run definitions (i.e., compilations of lines with similar operating conditions, water chemistry and usage 2-20

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) for analysis). A comparison of CLTP and EPU wear rate predictions identified changes for each unit. For Browns Ferry Unit 1 (Table 2.1-4a), 12 wear rate analysis runs were done: there was decrease of 30.2% to an increase of 16.74%. Of the run definitions for Unit 1, six had a decrease in the predicted wear rate while the remaining six run definitions exhibited an increase. For Browns Ferry Unit 2 (Table 2.1-4b), 13 wear rate analysis runs were done: there was decrease of 29.86% to an increase of 19.35%. Of the run definitions for Unit 2, six had a decrease in the predicted wear rate while the remaining seven run definitions exhibited an increase. For Browns Ferry Unit 3 (Table 2.1-4c), 13 wear rate analysis runs were done: there was decrease of 29.87% to an increase of 16.26%. Of the run definitions for Unit 3, seven had a decrease in the predicted wear rate while the remaining six run definitions exhibited an increase. Based on a review of the changes in operating conditions, Browns Ferry found the resulting predicted wear rates to be consistent with EPU conditions. Selection of Inspection and Replacement Locations The current approach to select locations for FAC inspection does not change as a result of the EPU. However, there will be an increase in the number of FAC inspections performed on both CHECWORKS' SFA-modeled and SNM piping over the next several refueling outages to ensure the effect of extended power uprate is understood. Inspections will be selected considering the changes in predicted wear rates, actual component thicknesses, operating time since last examination and design margin. This approach will ensure that FAC susceptible components are inspected or replaced prior to reaching code minimum wall thickness. Based on the EPU evaluation, no significant effect on the component replacement schedule is anticipated in the near term. The continued implementation of the existing Browns Ferry FAC program, updated appropriately to include EPU system parameters, will ensure that any required changes to the component inspection and replacement schedules are made prior to EPU implementation. This data will be used to further calibrate the CHECWORKS' SFA model and susceptibility for SNM piping. Benchmarking CHECWORKS' SFA Predicted Component Thickness Tables 2.1-5a, 5b and 5c presents a comparison of CHECWORKS'-predicted thicknesses to measured thicknesses for a sample component from each of the Wear Rate Analysis (WRA) run definitions. The selection process includes components with the highest predicted wear rates prior to EPU for each unit. The measured thicknesses were determined by ultrasonic testing non-destructive examination performed during the refueling outage as noted in the tables. 2-21

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The tables show that, with the exception of two cases (one in Browns Ferry Unit 2 and the other in Browns Ferry Unit 3), the measured thickness from the inspection was greater than the predicted thickness, indicating that CHECWORKS' SFA predictions are typically conservative. Other than FAC, Browns Ferry also inspects certain components for degradation caused by Liquid Droplet Impingement (LDI). Indications that LDI may be present are valve leak-bys, or conditions (open valves, leaks) that cause the velocity of the two-phased mixture to increase dramatically. The FAC program also inspects for cavitation per system engineering requests. The Browns Ferry FAC program adequately manages the effects on FAC due to EPU. Therefore, Browns Ferry meets all CLTR dispositions for FAC. Conclusion TVA has evaluated the effect of the proposed EPU on the FAC analysis for the plant and has addressed changes in the plant operating conditions on the FAC analysis. The evaluation indicates that the updated analyses will predict the loss of material by FAC and will ensure timely repair or replacement of degraded components following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to FAC. 2.1.7 Reactor Water Cleanup System Regulatory Evaluation The Reactor Water Cleanup (RWCU) system provides a means for maintaining reactor water quality by filtration and ion exchange and a path for removal of reactor coolant when necessary. Portions of the RWCU system comprise the RCPB. The NRCs acceptance criteria for the RWCU system are based on (1) GDC-14, insofar as it requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; (2) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (3) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate confinement. Specific NRC review criteria are contained in SRP Section 5.4.8. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the 2-22

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-9, 34, 51, and 70. The Reactor Water Cleanup System is described in Browns Ferry UFSAR Section 4.9, Reactor Water Cleanup System. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Reactor Water Cleanup System is documented in NUREG-1843, Section 2.3.3.21. Management of aging effects on the Reactor Water Cleanup System is documented in NUREG-1843, Section 3.0.3.2.15. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 3.11 of the CLTR addresses the effect of EPU on the reactor water cleanup system. The results of this evaluation are described below. The RWCU system is a normally operating system with no safety-related functions other than RCPB and containment isolation. This system is designed to remove solid and dissolved impurities from recirculated reactor coolant, thereby reducing the concentration of radioactive and corrosive species in the reactor coolant. The evaluation of the system performance of the Browns Ferry RWCU system under EPU conditions is presented below. The effects of EPU on the RWCU containment isolation function and valves are included in the containment isolation assessment in Sections 2.2.4 and 2.6.1.3. Tables 2.1-6 and 2.1-7 contain the magnitude of changes in RWCU system operating conditions and a summary of the chemistry values. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: 2-23

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR System Performance Plant Specific Disposition Addressed in Containment Isolation Plant Specific Section 2.6.1.3 As explicitly stated in Section 3.11 of the CLTR, the RWCU system may be slightly affected by the increase in FW flow due to the power uprate. RWCU system operation at the EPU RTP level slightly decreases the temperature within the RWCU system (from 530.5°F to 529.3°F). This system currently operates at flow rates consistent with the original design flow. The operating flow rates are not being changed for EPU. Table 2.1-6 provides the magnitude of changes in RWCU system operating conditions (e.g., a decrease in operating inlet temperature). RWCU system flow is usually selected to be approximately 1% of FW system flow based on operational history. For the Browns Ferry EPU, the RWCU system was analyzed for flow at 133,300 lbm/hr. This flow rate is approximately 0.81% of EPU rated FW flow. The evaluation of RWCU performance for the Browns Ferry EPU considered water chemistry, heat exchanger performance, pump performance, flow control valve capability and filter / demineralizer performance. All aspects of performance were found to be within the design of the RWCU system at the analyzed flow for EPU conditions. The RWCU system analysis concludes that:

1. An increase in filter / demineralizer backwash frequency occurs, but this is within the capacity of the radwaste system.

2. The changes in operating system conditions result from a decrease in inlet temperature and a negligible increase in FW system operating pressure.

3. The RWCU system filter / demineralizer control valves will operate in the slightly more open position because of the negligible increase to the RWCU system discharge pressure.

4. No changes to instrumentation are required, and setpoint changes are not required due to the system process parameter changes.

Previous operating experience has shown that the increased FW flow results in increases in three key reactor coolant chemistry parameters. Table 2.1-7 provides a summary of the chemistry values and the evaluation results for each are presented below. These values use the maximum values for actual plant rolling averages for all three plants: Sulfates concentration - The current maximum average level of sulfates is 1.29 ppb for all three units. The expected reactor water sulfate level for EPU, considering the FW flow increase, is 1.50 ppb. This level is well below the administrative goal of 2.0 ppb and the action level of 5.0 ppb for sulfates. 2-24

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Chlorides concentration - The current maximum average level of chlorides is 0.33 ppb for all three units. The expected reactor water chloride level for EPU, considering the FW flow increase, is 0.38 ppb. This level is well below the administrative goal of 1.0 ppb and the action level of 5.0 ppb for chlorides. Reactor water conductivity - The calculated reactor water conductivity increases from 0.121 S/cm to 0.132 S/cm because of the increase in FW flow. This expected level is below the administrative goal for conductivity of 0.14 S/cm and the action level of 0.30 S/cm. The effects of EPU on the RWCU system functional capability have been reviewed, and the system can perform adequately at EPU RTP with the CLTP RWCU system flow. As can be seen from Table 2.1-6, the changes in RWCU system operating conditions from CLTP to EPU are small. The Browns Ferry RWCU system has sufficient capacity to respond to the EPU conditions and maintain the chemistry parameters within administrative goals. Therefore, Browns Ferry meets all CLTR dispositions for system performance. Conclusion TVA has evaluated the effects of the proposed EPU on the RWCU system. The evaluation indicates that the RWCU system will continue to be acceptable following implementation of the proposed EPU and will continue to meet the requirements of the draft GDCs-9, 34, 51, and 70. Therefore, the proposed EPU is acceptable with respect to the RWCU system. 2-25

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-1a Browns Ferry Unit 1 USE EMA - 60 Year Life (38 EFPY) Equivalent Margin Analysis (EMA) Plant Applicability Verification Form for Browns Ferry Unit 1 60 Years (38 EFPY) BWR/3-6 PLATE Surveillance Plate USE:

                                                  %Cu   =    N/A 1st Capsule Fluence   =    N/A        n/cm 2 1st Capsule Measured % Decrease     =    N/A               (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease      =    N/A               (RG 1.99, Rev. 2, Figure 2)

Ratio of Measured to Predicted % Decrease = N/A (RG 1.99, Rev. 2, Position 2.2) Limiting Beltline Plate USE (Heat C2884-2):

                                                  %Cu   =          0.12 38 EFPY 1/4T Fluence    =     1.09E+18 n/cm 2 RG 1.99 Predicted % Decrease    =          13.0        (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = N/A (RG 1.99, Rev. 2, Position 2.2) 13.0% < (( )) Therefore, vessel plates are bounded by Equivalent Margin Analysis 2-26

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 1 60 Years (38 EFPY) BWR/2-6 WELD Surveillance Weld USE 406L44

                                                     %Cu   =         0.29 1st Capsule Fluence  =    1.00E+18 n/cm2 2nd Capsule Fluence   =    1.83E+18 n/cm2 3rd Capsule Fluence  =    1.77E+18 n/cm2 4th Capsule Fluence  =    2.89E+18 n/cm2 5th Capsule Fluence  =    3.97E+17 n/cm2 6th Capsule Fluence  =    4.93E+17 n/cm2 1st Capsule Measured % Decrease    =         32.0      (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease   =         26.0      (RG 1.99, Rev. 2, Figure 2) 2nd Capsule Measured % Decrease     =         33.0      (Charpy Curves) 2nd Capsule RG 1.99 Predicted % Decrease   =         29.5      (RG 1.99, Rev. 2, Figure 2) 3rd Capsule Measured % Decrease    =         36.5      (Charpy Curves) 3rd Capsule RG 1.99 Predicted % Decrease   =         29.5      (RG 1.99, Rev. 2, Figure 2) 4th Capsule Measured % Decrease    =         42.5      (Charpy Curves) 4th Capsule RG 1.99 Predicted % Decrease   =         32.5      (RG 1.99, Rev. 2, Figure 2) 5th Capsule Measured % Decrease    =         20.5      (Charpy Curves) 5th Capsule RG 1.99 Predicted % Decrease   =         21.0      (RG 1.99, Rev. 2, Figure 2) 6th Capsule Measured % Decrease    =         21.0      (Charpy Curves) 6th Capsule RG 1.99 Predicted % Decrease   =         22.0      (RG 1.99, Rev. 2, Figure 2)

Ratio of Measured to Predicted % Decrease (4th Capsule) = 1.3 (RG 1.99, Rev. 2, Position 2.2) Limiting Beltline Weld USE (406L44):

                                                     %Cu =           0.29 38 EFPY 1/4T Fluence =      8.86E+17 n/cm2 RG 1.99 Predicted % Decrease =             25      (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = 33.5 (RG 1.99, Rev. 2, Position 2.2) 33.5% < (( )) Therefore, vessel welds are bounded by Equivalent Margin Analysis 2-27

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-1b Browns Ferry Unit 2 USE EMA Year Life (48 EFPY) Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 2 60 Years (48 EFPY) BWR/3-6 PLATE Surveillance Plate USE1 A0981-1

                                                  %Cu   =     0.14 1st Capsule Fluence   =    2.40E+17 n/cm 2 2nd Capsule Fluence   =    6.44E+17 n/cm 2 1st Capsule Measured % Decrease     =          6            (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease     =        9.5            (RG 1.99, Rev. 2, Figure 2) 2nd Capsule Measured % Decrease     =        -3.6     [1]   (Charpy Curves) 2nd Capsule RG 1.99 Predicted % Decrease      =         12            (RG 1.99, Rev. 2, Figure 2)

Ratio of Measured to Predicted % Decrease = <1 (RG 1.99, Rev. 2, Position 2.2) Limiting Beltline Plate USE (Heat C2467-1):

                                                  %Cu   =            0.16 48 EFPY 1/4T Fluence    =     1.34E+18 n/cm 2 RG 1.99 Predicted % Decrease    =            16.0       (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = N/A (RG 1.99, Rev. 2, Position 2.2) 16.0% < (( )) Therefore, vessel plates are bounded by Equivalent Margin Analysis Note [1]: The 2nd capsule measured results demonstrated an increase in USE. 2-28

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 2 60 Years (48 EFPY) BWR/2-6 WELD Surveillance Weld USE BF2 ESW

                                                       %Cu =       0.20 1st Capsule Fluence =     2.40E+17 n/cm2 2nd Capsule Fluence =      6.44E+17 n/cm2 1st Capsule Measured % Decrease    =           5.9     (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease     =          14.1     (RG 1.99, Rev. 2, Figure 2) 2nd Capsule Measured % Decrease     =           3.4     (Charpy Curves) 2nd Capsule RG 1.99 Predicted % Decrease     =          17.8     (RG 1.99, Rev. 2, Figure 2)

Ratio of Measured to Predicted % Decrease = <1 (RG 1.99, Rev. 2, Position 2.2) Limiting Beltline Weld USE (ESW):

                                                       %Cu =            0.24 48 EFPY 1/4T Fluence =      9.14E+17 n/cm2 RG 1.99 Predicted % Decrease =            22.0     (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = N/A (RG 1.99, Rev. 2, Position 2.2) 22.0% < (( )) Therefore, vessel welds are bounded by Equivalent Margin Analysis 2-29

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-1c Browns Ferry Unit 3 USE EMA Year Life (54 EFPY) Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 3 60 Years (54 EFPY) BWR/3-6 PLATE Surveillance Plate USE:

                                                %Cu =         N/A 1st Capsule Fluence =        N/A    n/cm2 1st Capsule Measured % Decrease =          N/A          (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease =          N/A          (RG 1.99, Rev. 2, Figure 2)

Ratio of Measured to Predicted % Decrease = N/A (RG 1.99, Rev. 2, Position 2.2) Limiting Beltline Plate USE (Heat C3222-2):

                                                %Cu =           0.15 54 EFPY 1/4T Fluence =      1.25E+18 n/cm2 RG 1.99 Predicted % Decrease =           15.0       (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = N/A (RG 1.99, Rev. 2, Position 2.2) 15.0% < (( )) Therefore, vessel plates are bounded by Equivalent Margin Analysis 2-30

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Equivalent Margin Analysis Plant Applicability Verification Form for Browns Ferry Unit 3 60 Years (54 EFPY) BWR/2-6 WELD Surveillance Weld USE:

                                                     %Cu =         N/A 1st Capsule Fluence =        N/A    n/cm2 1st Capsule Measured % Decrease =          N/A          (Charpy Curves) 1st Capsule RG 1.99 Predicted % Decrease =         N/A          (RG 1.99, Rev. 2, Figure 2)

Limiting Beltline Weld USE (ESW):

                                                     %Cu =           0.24 54 EFPY 1/4T Fluence =      1.05E+18 n/cm2 RG 1.99 Predicted % Decrease =           23.0       (RG 1.99, Rev. 2, Figure 2)

Adjusted % Decrease = N/A (RG 1.99, Rev. 2, Position 2.2) 23.0% < (( )) Therefore, vessel welds are bounded by Equivalent Margin Analysis 2-31

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-2a Browns Ferry Unit 1 Adjusted Reference Temperatures 60-Year License (38 EFPY) Lower-Intermediate Plates Thickness in inches = 6.125 38 EFPY Peak I.D. fluence = 1.58E+18 n/cm2 38 EFPY Peak 1/4 T fluence = 1.09E+18 n/cm2 Lower Plates & Lower to Lower-Intermediate Girth Weld Thickness in inches= 6.125 38 EFPY Peak I.D. fluence = 1.28E+18 n/cm2 38 EFPY Peak 1/4 T fluence = 8.86E+17 n/cm2 Axial Welds Thickness in inches= 6.125 38 EFPY Peak I.D. fluence = 1.58E+18 n/cm2 38 EFPY Peak 1/4 T fluence = 1.09E+18 n/cm2 Water Level Instrumentation Nozzle Thickness in inches= 6.125 38 EFPY Peak I.D. fluence = 4.77E+17 n/cm2 38 EFPY Peak 1/4 T fluence = 3.30E+17 n/cm2 38 EFPY 38 EFPY 38 EFPY Adjusted Initial 1/4 T 1/4T 1/4T 1/4T COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF CF RTNDT Fluence RTNDT I Margin Shift ART

                                                                                                          °F       n/cm2       °F                                °F      °F       °F PLATES:

Lower Shell 6-127-1 A0999-1 0.14 0.60 100 -20 8.86E+17 39 0 17 34 73 53 6-127-2 B5864-1 0.15 0.44 101 -20 8.86E+17 40 0 17 34 74 54 6-127-4 A1009-1 0.14 0.50 96 -10 8.86E+17 38 0 17 34 72 62 Lower-Intermediate Shell 6-139-19 C2884-2 0.12 0.53 82 14 1.09E+18 36 0 17 34 70 84 6-139-20 C2868-2 0.09 0.48 58 30 1.09E+18 25 0 13 25 50 80 6-139-21 C2753-1 0.08 0.50 51 2 1.09E+18 22 0 11 22 44 46 WELDS: Axial Welds ESW -- 0.24 0.37 141 23.1 1.09E+18 61 13 28 62 123 146 Lower to Lower-Intermediate Girth Weld WF154 406L44 0.27 0.60 184 20 8.86E+17 72 10 28 59 132 152 BEST ESTIMATE CHEMISTRIES: None NOZZLES: N16 Water Level Instrumentation Forging Inconel (1) 0.12 0.53 82 14 3.30E+17 19 0 9 19 38 52 Weld Inconel INTEGRATED SURVEILLANCE PROGRAM: Plate (2) A0981-1 Weld (3, 4) SSP-406144 0.29 0.69 205 (( )) 23.1 8.86E+17 110 10 28 59 170 193 Notes: [1] The material properties used are those for the bounding adjacent shell plate from the lower-intermediate shell. The fluence considered is applicable at the nozzle location. [2] The representative plate material is not the same heat number as the target plate; therefore the RG 1.99 Chemistry Factor (CF) is used. This information is not applicable to development of the P-T curves and is provided for information only. [3] The initial RTNDT is obtained from the limiting plant-specific plate and weld. [4] The representative weld material is the same heat as the target weld; therefore, these results are considered in development of the P-T curves. Surveillance data from six (6) capsules are available. Scatter of this data exceeds credibility criteria. The fitted CF of (( )), based on the surveillance data, is higher than the RG 1.99 table CF of 204.95°F; therefore, the full margin term is applied. The CF is adjusted per RG 1.99 to be: (184°F/205°F) * (( )). 2-32

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-2b Browns Ferry Unit 2 Adjusted Reference Temperatures 60-Year License (48 EFPY) Lower-Intermediate Plates Thickness = 6.125 inches 48 EFPY Peak I.D. fluence = 1.93E+18 n/cm2 48 EFPY Peak 1/4 T fluence = 1.34E+18 n/cm2 Lower Plates & Lower to Lower-Intermediate Girth Weld Thickness = 6.125 inches 48 EFPY Peak I.D. fluence = 1.56E+18 n/cm2 48 EFPY Peak 1/4 T fluence = 1.08E+18 n/cm2 Axial Welds Thickness = 6.125 inches 48 EFPY Peak I.D. fluence = 1.32E+18 n/cm2 48 EFPY Peak 1/4 T fluence = 9.14E+17 n/cm2 Water Level Instrumentation Nozzle Thickness = 6.125 inches 48 EFPY Peak I.D. fluence = 5.84E+17 n/cm2 48 EFPY Peak 1/4 T fluence = 4.04E+17 n/cm2 Adjusted Initial 1/4 T 48 EFPY I 48 EFPY 48 EFPY 1/4 T 1/4 T 1/4 T COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF CF RTNDT Fluence RTNDT Margin Shift ART

                                                                                                                °F       n/cm2     °F                                °F        °F       °F PLATES:

Lower Shell 6-127-14 C2467-2 0.16 0.52 112 -20 1.08E+18 48 0 17 34 82 62 6-127-15 C2463-1 0.17 0.48 117 -20 1.08E+18 51 0 17 34 85 65 6-127-17 C2460-2 0.13 0.51 88 0 1.08E+18 38 0 17 34 72 72 Lower-Intermediate Shell 6-127-6 A0981-1 0.14 0.55 98 -10 1.34E+18 47 0 17 34 81 71 6-127-16 C2467-1 0.16 0.52 112 -10 1.34E+18 53 0 17 34 87 77 6-127-20 C2849-1 0.11 0.50 73 -10 1.34E+18 35 0 17 34 69 59 WELDS: Axial Welds ESW -- 0.24 0.37 141 23.1 9.14E+17 56 13 28 62 118 141 Lower to Lower-Intermediate Girth Weld D55733 0.09 0.65 117 -40 1.08E+18 51 0 25 51 101 61 BEST ESTIMATE CHEMISTRIES: None NOZZLES: N16 Water Level Instrumentation Forging Inconel [1] 0.16 0.52 112 -10 4.04E+17 29 0 15 29 58 48 Weld Inconel INTEGRATED SURVEILLANCE PROGRAM: Plate [2, 3] A0981-1 0.14 0.55 (( )) -10 1.34E+18 68 0 8.5 17 85 75 Weld [4, 5] BF2 ESW 0.20 0.33 (( )) 23.1 9.14E+17 114 13 14 38 152 175 Notes: [1] The material properties used are those for the bounding adjacent shell plate from the lower-intermediate shell. The fluence considered is applicable at the nozzle location. [2] The representative plate material is not the same heat as the target plate; therefore, the Position 1.1 RG 1.99 CF is used. This information is, however, applicable to development of the P-T curves because the same heat of material exists in the Unit 2 vessel lower-intermediate shell. The surveillance data is credible; therefore is reduced as permitted by RG 1.99. [3] The initial RTNDT is obtained from the plant-specific material in the lower-intermediate shell. [4] The initial RTNDT is obtained from the plant-specific material in the axial welds. [5] The representative weld material is considered to be the same heat as the target weld; therefore, these results are considered in the development of the P-T curves. As this material is considered to be the same heat as the Unit 2 vessel, the adjusted CF is calculated per RG 1.99 to be (140.55°F/120.25°F) * (( )). The surveillance data is credible; therefore is reduced as permitted by RG 1.99. 2-33

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-2c Browns Ferry Unit 3 Adjusted Reference Temperatures 60-Year License (54 EFPY) Lower-Intermediate Plates Thickness = 6.125 inches 54 EFPY Peak I.D. fluence = 2.23E+18 n/cm 2 54 EFPY Peak 1/4 T fluence = 1.54E+18 n/cm 2 Lower Plates & Lower to Lower-Intermediate Girth Weld Thickness = 6.125 inches 54 EFPY Peak I.D. fluence = 1.80E+18 n/cm 2 54 EFPY Peak 1/4 T fluence = 1.25E+18 n/cm 2 Axial Welds Thickness = 6.125 inches 54 EFPY Peak I.D. fluence = 1.52E+18 n/cm 2 54 EFPY Peak 1/4 T fluence = 1.05E+18 n/cm 2 Water Level Instrumentation Nozzle Thickness = 6.125 inches 54 EFPY Peak I.D. fluence = 6.75E+17 n/cm 2 54 EFPY Peak 1/4 T fluence = 4.67E+17 n/cm 2 1/4T 1/4T 1/4T Initial 1/4 T 54 EFPY I 54 EFPY 54 EFPY COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF RTNDT Fluence RTNDT Margin Shift ART

                                                                                                     °F      n/cm^2      °F                                 °F        °F        °F PLATES:

Lower Shell 6-145-4 C3222-2 0.15 0.52 106 10 1.25E+18 49 0 17 34 83 93 6-145-7 C3213-1 0.13 0.58 90 -20 1.25E+18 42 0 17 34 76 56 6-145-12 C3217-2 0.14 0.66 101.5 -4 1.25E+18 47 0 17 34 81 77 Lower-Intermediate Shell 6-145-1 C3201-2 0.13 0.60 91 -20 1.54E+18 46 0 17 34 80 60 6-145-2 C3188-2 0.10 0.48 65 -20 1.54E+18 33 0 17 33 66 46 6-145-6 B7267-1 0.13 0.51 88 -20 1.54E+18 45 0 17 34 79 59 WELDS: Axial Welds ESW -- 0.24 0.37 141 23.1 1.05E+18 60 13 28 62 122 145 Lower to Lower-Intermediate Girth Weld D55733 0.09 0.66 117 -40 1.25E+18 54 0 27 54 108 68 BEST ESTIMATE CHEMISTRIES: None NOZZLES: N16 Water Level Instrumentation Forging Inconel [1] 0.13 0.6 91 -20 4.67E+17 26 0 13 26 51 31 Weld Inconel INTEGRATED SURVEILLANCE PROGRAM: Plate [2] A0981-1 Weld [3] BF2 ESW Notes: [1] The material properties used are those for the bounding adjacent shell plate from the lower-intermediate shell. The fluence considered is applicable at the nozzle location. [2] The representative plate material is not the same heat number as the target plate; therefore the RG 1.99 CF is used. This information is not applicable to development of the P-T curves and is provided for information only. [3] The initial RTNDT is obtained from the limiting plant-specific plate and weld. [4] The representative weld material is not the same heat as the target weld; therefore, these results are not considered in development of the P-T curves. 2-34

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-3 Effects of Irradiation on Browns Ferry RPV Circumferential Weld Properties Parameter B&W Unit 1 Unit 2 Unit 3 64 EFPY[1] 38 EFPY 48 EFPY 54 EFPY Cu% 0.31 0.27 0.09 0.09 Ni% 0.59 0.60 0.65 0.66 CF 196.7 184 117 117 Fluence at 0.19 0.128 0.156 0.18 clad/weld interface (1019 n/cm2) RTNDT w/o 109.4 86 60 64 margin (F) (Note 2) RTNDT(U) 20 20 -40 -40 (F) Mean RTNDT 129.4 106 20 24 (F) P(F/E) 4.83 x 10-4 (Note 4) (Note 4) (Note 4) NRC (Note 3) P(F/E) --- --- --- --- BWRVIP (Note 3) Notes: [1] Data for the Babcock and Wilcox (B&W) group of plants was obtained from BWRVIP-05 and its SER. [2] RTNDT = CF

f (0.28 - 0.10 log f) as defined in RG 1.99.

[3] P(F/E) means Probability of a failure event. [4] Although a conditional failure probability has not been calculated, the fact that the Browns Ferry values at the end of license are less than the 64 EFPY value provided by the NRC leads to the conclusion that the Browns Ferry RPV conditional failure probability is bounded by the NRC analysis, consistent with the requirements defined in GL 98-05. 2-35

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-4a Browns Ferry Unit 1 Comparison of Key Parameters Influencing FAC Wear Rate CLTP vs. EPU Temperature Oxygen Percent Velocity (ft/sec) Quality Change in CHECWORKSTM (oF) (ppb) Predicted Wear Rate Change % % % Wear Rate Analysis Run CLTP EPU (oF) CLTP EPU Change CLTP EPU Change CLTP EPU Change Due to Power Definition Name 105% 120% 105% 120% 105% 120% 105% 120% Uprate (See NOTE) FWH 4 - FWH 3 242.3 248.8 6.5 13.841 16.104 16.35 52.7 52.7 0.00 0.0 0.0 N/A 16.27 FWH 3 - RFPs 301.5 309.3 7.8 14.269 16.625 16.51 52.7 52.7 0.00 0.0 0.0 N/A -10.04 RFPs - FWH 2 301.5 309.3 7.8 16.094 18.751 16.51 52.7 52.7 0.00 0.0 0.0 N/A -10.03 FWH 2 - FWH 1 334.4 343.5 9.1 13.261 15.47 16.66 52.7 52.7 0.00 0.0 0.0 N/A 7.04 FWH 1 - Rx 381.3 391.6 10.3 16.538 19.328 16.87 52.7 52.7 0.00 0.00 0.00 N/A -9.67 EXT STM #1 390.7 403.0 12.3 19.751 21.991 11.34 6.0 7.5 24.1 0.892 0.888 -0.45 7.35 MSEP - FCVs 389.2 401.5 12.3 2.069 2.495 20.59 5.9 7.3 24.1 0.0 0.0 N/A -29.95 FCVs - FWH 2 389.2 401.5 12.3 6.175 7.447 20.60 5.9 7.3 24.1 0.0 0.0 N/A -30.2 FWH 1 - FWH 2 344.3 356.9 12.6 4.06 4.994 23.00 851.9 1009.1 18.46 0.0 0.0 N/A 16.74 FWH 2 - FWH 3 313.0 323.8 10.8 7.815 9.461 21.06 410.0 481.7 17.49 0.0 0.0 N/A -10.24 FWH 3 - FWH 4 252.4 262.2 9.8 9.718 11.724 20.64 244.2 285.0 16.69 0.0 0.0 N/A 4.54 FWH 4 - FL TNK 196.5 204.8 8.3 3.871 4.64 19.87 80.2 92.4 15.19 0.0 0.0 N/A 10.21 2-36

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-4b Browns Ferry Unit 2 Comparison of Key Parameters Influencing FAC Wear Rate CLTP vs. EPU Temperature Oxygen Percent Velocity (ft/sec) Quality Change in (oF) (ppb) CHECWORKSTM Predicted Wear Rate Analysis Change Wear Rate CLTP EPU CLTP EPU Change CLTP EPU Change CLTP EPU Change Run Definition Name (oF) Due to Power 105% 120% 105% 120% 105% 120% 105% 120% Uprate (See NOTE) 301.5 309.3 7.8 14.269 16.625 16.51 60.2 60.2 0.00 0 0 N/A -10.06 CON-FWH 3 to RFPs 242.3 248.8 6.5 13.841 16.104 16.35 60.2 60.2 0.00 0 0 N/A 16.27 CON-FWH 4 to FWH 3 390.7 403 12.3 19.75 21.987 11.33 5.8 7.3 24.76 0.892 0.888 -0.45 9.59 EX-Extraction #1 345.7 356.6 10.9 25.169 31.583 25.48 2.7 3.3 22.42 0.967 0.965 -0.21 19.35 EX-Extraction #2 344.3 356.9 12.6 4.06 4.994 23.00 821.9 973.6 18.46 0 0 N/A 14.95 HDV-FWH 1 to FWH 2 313 323.8 10.8 7.815 9.461 21.06 395.6 464.8 17.49 0 0 N/A -10.34 HDV-FWH 2 to FWH 3 252.4 262.2 9.8 9.718 11.723 20.63 235.6 275.0 16.69 0 0 N/A 4.46 HDV-FWH 3 to FWH 4 HDV-FWH 4 to FL 77.4 89.1 15.19 0 0 N/A 196.6 204.9 8.3 3.871 4.639 19.84 11.05 TNK 389.2 401.5 12.3 2.069 2.496 20.64 5.7 7.1 24.71 0 0 N/A -29.61 HDV-MSEP to FCVs HDV-MSP FCV to 5.7 7.1 24.71 0 0 N/A 389.2 401.5 12.3 6.176 7.447 20.58 -29.86 FWH 2 381.9 391.6 10.3 16.938 19.796 16.87 60.2 60.2 0.00 0 0 N/A -9.66 RFW-FWH 1 to Rx 334.4 343.5 9.1 13.261 15.47 16.66 60.2 60.2 0.00 0 0 N/A 7.03 RFW-FWH 2 to FWH 1 301.5 309.3 7.8 16.094 18.751 16.51 60.2 60.2 0.00 0 0 N/A -10.05 RFW-RFPs to FWH 2 2-37

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-4c Browns Ferry Unit 3 Comparison of Key Parameters Influencing FAC Wear Rate CLTP vs. EPU Temperature Oxygen Percent Velocity (ft/sec) Quality Change in CHECWORKSTM (oF) (ppb) Predicted Wear Rate Change % % % Wear Rate Analysis Run CLTP EPU (oF) CLTP EPU Change CLTP EPU Change CLTP EPU Change Due to Power Definition Name 105% 120% 105% 120% 105% 120% 105% 120% Uprate (See NOTE) 58.5 0.00 0.0 0.0 N/A FWH 4 - FWH 3 242.3 248.8 6.5 13.841 16.104 16.35 58.5 16.26 58.5 58.5 0.00 0.0 0.0 N/A FWH 3 - RFPs 301.5 309.3 7.8 14.269 16.625 16.51 -10.03 58.5 58.5 0.00 0.0 0.0 N/A RFPs - FWH 2 301.5 309.3 7.8 16.094 18.751 16.51 -10.04 58.5 58.5 0.00 0.0 0.0 N/A FWH 2 - FWH 1 334.4 343.5 9.1 13.261 15.47 16.66 7.03 58.5 0.00 N/A FWH 1 - Rx 381.3 391.6 10.3 16.938 19.796 16.87 58.5 0.0 0.0 -9.68 EXT STM #1 390.7 403 12.3 19.75 21.987 11.33 5.7 7.1 24.73 0.892 0.888 -0.45 9.65 5.6 7.0 24.73 0.0 0.0 N/A MSEP - FCVs 389.2 401.5 12.3 2.069 2.496 20.64 -29.62 5.6 7.0 24.73 0.0 0.0 N/A FCVs - FWH 2 389.2 401.5 12.3 6.176 7.447 20.58 -29.87 809.0 958.3 18.46 0.0 0.0 N/A FWH 1 - FWH 2 344.3 356.9 12.6 4.06 4.994 23.00 14.88 FWH A2 - FWH 389.4 457.5 17.49 0.0 0.0 N/A 313 323.8 10.8 7.815 9.461 21.06 -10.38 A3 FWH B&C2 - 389.4 457.5 17.49 0.0 0.0 N/A 313 323.8 10.8 7.815 9.461 21.06 -10.39 B&C3 231.9 270.6 16.69 0.0 0.0 N/A FWH 3 - FWH 4 252.4 262.2 9.8 9.718 11.723 20.63 4.42 0.0 0.0 N/A FWH 4 - FL TNK 196.6 204.9 8.3 3.871 4.639 19.84 76.2 87.7 15.19 11.41 2-38

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Note: The rate of Flow Accelerated Corrosion (FAC) is a complex process that is interdependent on many variables including temperature, velocity, oxygen concentration, and steam quality. Each variable affects the overall wear rate for a component differently. The algorithm in the CHECWORKSTM code has the ability to determine the overall effect on wear rate based on changes in each variable. The rate of FAC is related to the interaction of the parameters; thus, the primary reason for the predicted decrease under EPU conditions is associated with the change in operating temperature. The influence of temperature is represented by a bell curve. Flow accelerated corrosion rates increase as temperature increases up to approximately 300°F and then decrease as the temperature continues to increase beyond 300oF. The slopes of the bell curve are quite steep, which results in a relatively large decrease in wear rate based on a relatively small increase in temperature. The influence of velocity on the rate of flow accelerated corrosion is fairly linear. The slope of the velocity curve is relatively flat indicating that larger changes in velocity will have a lesser effect on rate of FAC degradation verses temperature. Evaluation of the entries in Tables 2-1.4a - c with negative changes in the predicted FAC wear rate indicates that the increase in temperature resulted in a larger overall reduction in the predicted wear rate than the corresponding increase from velocity. This results in a net reduction in the predicted wear rate. 2-39

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-5a Browns Ferry Unit 1 Components with Highest Predicted Wear Rate for Each Wear Rate Analysis Run Definition CHECWORKSTM SFA-Predicted Thickness vs. Measured Thickness Ratio of Measured Predicted Measured CHECWORKS' Wear Nominal Nominal Thickness Thickness Thickness to Rate Analysis Run Component Pipe Size Thickness Tmeas Tpred Predicted Time of Definition Component Name Type (inches) (inches) (inches) (inches) Thickness Inspection REFUEL 6 - FWH 4 - FWH 3 1CON11A-2RT 180o Return 18 0.438 0.452 0.308 1.47 Restart FWH 3 - RFPs 1CON12A-13E Elbow 18 0.438 0.421 0.363 1.16 REFUEL 7 RFPs - FWH 2 1RFW1A - BFN Orifice 18 0.861 0.771 0.602 1.28 REFUEL 10 REFUEL 6 - FWH 2 - FWH 1 1RFW2A2-1N Reducer 24.5 3.281 3.310 3.125 1.06 Restart FWH 1 - Rx 1RFW3B - 10FN Orifice 24 1.219 1.149 0.812 1.41 REFUEL 10 EXT STM #1 1EX11 - 15T Tee 30 0.375 0.440 0.257 1.71 REFUEL 10 MSEP - FCVs 1HDV9MSB2 - 10N Exit Nozzle 4 0.237 0.538 0.144 3.75 REFUEL 8 FCVs - FWH 2 1HDV10MSA2 - 29O Orifice 6 0.280 0.251 0.092 2.73 REFUEL 8 FWH 1 - FWH 2 1HDV1A1 - 3R Reducer 8 0.322 0.277 0.045 6.19 REFUEL 8 REFUEL 6 - FWH 2 - FWH 3 1HDV2A2 - 45R Reducer 10 0.365 0.369 0.253 1.46 Restart REFUEL 6 - FWH 3 - FWH 4 1HDV3A3 - 9R Reducer 10 0.365 0.331 0.255 1.30 Restart FWH 4 - FL TNK 1HDV4B4 - 16R Reducer 18 0.375 0.339 0.229 1.61 REFUEL 9 2-40

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-5b Browns Ferry Unit 2 Components with Highest Predicted Wear Rate for Each Wear Rate Analysis Run Definition CHECWORKSTM SFA-Predicted Thickness vs. Measured Thickness Ratio of Measured Predicted Measured CHECWORKS' Wear Nominal Nominal Thickness Thickness Thickness to Rate Analysis Run Component Pipe Size Thickness Tmeas Tpred Predicted Time of Definition Component Name Type (inches) (inches) (inches) (inches) Thickness Inspection CON-FWH 3 to RFPs 2CON12A -7T Tee 18 0.438 0.618 0.312 1.98 REFUEL 10 0 CON-FWH 4 to FWH 3 2CON11A - 2R 180 Return 18 0.438 0.496 0.287 1.73 REFUEL 12 EX-Extraction #1 2EX11 - 15T Tee 30 0.375 0.343 0.205 1.67 REFUEL 12 EX-Extraction #2 2EX32A - 20N Reducer 12 0.500 0.330 0.477 0.69 REFUEL 12 HDV-FWH 1 to FWH 2 2HDV1A1- 3R Reducer 8 0.322 0.310 0.107 2.91 REFUEL 11 HDV-FWH 2 to FWH 3 2HDV4A2 - 12R Reducer 10 0.365 0.310 0.207 1.50 REFUEL 14 HDV-FWH 3 to FWH 4 2HDV6A3 - 9R Reducer 10 0.365 0.334 0.269 1.24 REFUEL 10 HDV-FWH 4 to FL TNK 2HDV8A4 - 16A Reducer 18 0.375 0.423 0.217 1.95 REFUEL 14 HDV-MSEP to FCVs 2HDV10MA2 - 30T Tee 6 0.280 0.306 0.086 3.58 REFUEL 15 HDV-MSP FCV to FWH 2 2HDV2MSA1 - 19R Reducer 6 0.280 0.269 0.086 3.12 REFUEL 10 RFW-FWH 1 to Rx 2RFW6B - 10FN Orifice 24 1.219 1.154 0.950 1.21 REFUEL 11 RFW-FWH 2 to FWH 1 2RFW4C2 - 7N Exit Nozzle 20 1.031 0.952 0.857 1.11 REFUEL 11 RFW-RFPs to FWH 2 2RFW1A - 8FN Orifice 18 0.938 0.778 0.675 1.15 REFUEL 11 2-41

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-5c Browns Ferry Unit 3 Components with Highest Predicted Wear Rate for Each Wear Rate Analysis Run Definition CHECWORKSTM SFA-Predicted Thickness vs. Measured Thickness Ratio of Measured Predicted Measured CHECWORKS' Wear Nominal Nominal Thickness Thickness Thickness Rate Analysis Run Component Pipe Size Thickness Tmeas Tpred to Predicted Time of Definition Component Name Type (inches) (inches) (inches) (inches) Thickness Inspection FWH 4 - FWH 3 3CON11A -2RT 1800 Return 18 0.438 0.501 0.305 1.64 REFUEL 12 FWH 3 - RFPs 3CON12A -10E Elbow 18 0.438 0.551 0.348 1.58 REFUEL 14 RFPs - FWH 2 3RFW1C - 4E Elbow 18 0.938 0.973 0.635 1.53 REFUEL 15 FWH 2 - FWH 1 3RFW2A2 - 1N Reducer 24.5 3.281 3.340 2.736 1.22 REFUEL 10 FWH 1 - Rx 3RFW3A - 16FN Orifice 24 1.219 1.136 0.799 1.42 REFUEL 11 EXT STM #1 3EX11 - 14T Tee 30 0.375 0.396 0.316 1.26 REFUEL 10 MSEP - FCVs 3HDV7MSB1 - 17N Exit Nozzle 5.75 0.962 0.504 0.828 0.61 REFUEL 10 FCVs - FWH 2 3HDV8MSA1 - 34R Reducer 6 0.280 0.242 0.107 2.26 REFUEL 10 FWH 1 - FWH 2 3HDV1A1 - 3R Reducer 8 0.322 0.305 0.104 2.92 REFUEL 11 FWH A2 - FWH A3 3HDV2A2 - 44R Reducer 10 0.365 0.310 0.228 1.36 REFUEL 10 FWH B&C2 - B&C3 3HDV2B2 - 22T Tee 10 0.365 0.405 0.294 1.38 REFUEL 9 FWH 3 - FWH 4 3HDV3A3 - 9R Reducer 10 0.365 0.315 0.252 1.25 REFUEL 11 FWH 4 - FL TNK 3HDV4A4 - 16R Reducer 18 0.375 0.407 0.238 1.71 REFUEL 13 2-42

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-6 Comparison of RWCU System Operating Conditions Parameter Units CLTP EPU Thermal Power MWt 3,458 3,952 RWCU System Inlet Temperature °F 530.5 529.3 RWCU System Inlet Pressure psia 1,050 1,050 (RPV dome pressure, neglecting head) RWCU System Outlet Temperature °F 433.5 432.2 RWCU System Flow lbm/hr 133,300 133,300 Table 2.1-7 Comparisons of Chemistry Parameters for CLTP and EPU Cases CLTP EPU Item Parameter Units Values Values Maximum average sulfate 1 ppb 1.29 1.50 concentration Maximum average chloride 2 ppb 0.33 0.38 concentration Maximum average reactor water 3 S/cm 0.121 0.132 conductivity 2-43

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-8 Selection Process Criteria for Components in the FAC Program Selection Process Criteria Description The Browns Ferry FAC program selects components based on the results of the model's output (i.e., wear rate and remaining life). Components CHECWORKSTM Model are selected from both lines that have not been inspected and from lines that have inspected components. The Browns Ferry FAC program selects inspection components based on the susceptibility (highest trended wear rate and shortest remaining service life) of the non-modeled piping. A large amount of FAC Susceptible Non-Modeled susceptible piping cannot be modeled because of a lack of operating (SNM) parameter data. This includes almost all of the small-bore piping. This also includes FW heater shells. Lines that are deemed highly susceptible and could have detrimental consequences if failure occurred are slated for inspection. The Browns Ferry FAC program selects inspection components based on Operating Experiences (OEs) from the industry that are applicable to Industry Operating Experience Browns Ferry. Periodically, OEs are reviewed for Browns Ferry applicability. If the event is applicable, suitable components are selected to address the issue. The Browns Ferry FAC program selects inspection components based on site-specific events. Site Operating Experience also encompasses site specific information obtained from other site groups and other TVA sites. Periodically, the corrective action program is reviewed to discover Site Operating Experience if any situations had occurred that would be applicable to the program, (Internal) (i.e., valve leak-bys, steam leaks, abnormal valve usage (open when should be closed)). The thermal performance report is also reviewed periodically to identify any applicable leaking valves whose piping may need to be inspected. Inspection components are also selected based on requests from system engineers or from design changes. The Browns Ferry FAC program selects inspection components based on the NSI number. The components NSI is based on the wear rate and the Inspection Trending / minimum allowable wall thickness. Components with a remaining life Re-Inspections less than the time to the upcoming outage +1 full operating cycle are inspected. Carbon steel components downstream of FAC-resistant material have Replacement Transition / shown to have higher wear rates. A sample of carbon steel components Entrance Effects downstream of known resistant materials are considered for inspection. The Browns Ferry FAC program also selects inspection components based on engineering judgment using the criteria above. Engineering Engineering Judgment judgment is used when selecting inspection locations through industry and Browns Ferry operating experience. 2-44

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-9a RCPB Piping and Safe End Materials of Construction Nozzle Unit 1 Unit 2 Unit 3 Designation / System N1x: Recirculation SA-182 F316NG SA-376 GR 316 SA-376 GR 316 Suction N2x: Recirculation SA-182 F316NG SA-182 F316NG SA-182 F316NG Discharge N3x: Main Steam SA-105 SA-105 SA-105 N4x: Reactor SA-105 SA-105 SA-105 Feedwater N5x: Core Spray SA-182 F316NG SA-182 F316NG SA-182 F316NG N6x: Rx Vessel Head Spray A508 CL 2 A508 CL 2 A508 CL 2 (Not Used) N7: Rx Vessel A508 CL 2 A508 CL 2 A508 CL 2 Head Vent N8x: Jet Pump SA-182 F316NG SA-182 F316NG SA-182 F316NG Instrumentation N9x: Control Rod Drive Return - SA336 F8 SA336 F8M SA336 F8M Capped (Not Used) N10: SLC and Core Plate SA336 F8 SA336 F8M SA336 F8M Differential N11x: Rx Vessel Level SA336 F8 SA336 F8M SA336 F8M Instrumentation N12x: Rx Vessel Level SA336 F8 SA336 F8M SA336 F8M Instrumentation N15: Rx Vessel SA-105 SA-105 SA-105 Drain 2-45

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-9a RCPB Piping and Safe End Materials of Construction (continued) Nozzle Unit 1 Unit 2 Unit 3 Designation / System N16x: Rx Vessel Level SA336 F8 SA336 F8M SA336 F8M Instrumentation SB-166 SB-166 SB-166 CRD Drive Nozzles A333 Gr 1 A333 Gr 1 A333 Gr 1 A182 F316 A182 F316 A182 F316 A358 TP304 SA376 TP316NG A358 TP304 SA403 WP316NG SA403 WP 316NG A376 TP304 SA376 TP316 Reactor Recirculation SA182 F316NG SA403 WP316NG SA182 F316 System (Note 1) SA182 F316L A182 F304 SA182 F316L Cast A351 CF8 Cast A351 CF8 Cast A351 CF8 Cast A351 CF8M Cast A351 CF8M Cast A351 CF8M A155 GR KC70 CL 1 A155 GR KC70 CL 1 A155 GR KC70 CL 1 A516 A516 A516 A106 GR B Main Steam System A106 GR B A106 GR B A234 GR B A234 GR B A234 GR B A420 GR WPLI A105 GR II A105 GR II A350 GR LFI A155 GR KC70 CL 1 A155 GR KC70 CL 1 A106 GR B A155 GR KC70 CL 1 A234 GFR WPB A234 GFR WPB Reactor Feedwater A106 GR B A333 GR 6 A333 GR 6 System A105 GR II A420 GR WPL1 A420 GR WPL1 A234 GR B A105 GR II A105 GR II A234 GR B A234 GR B 2-46

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-9a RCPB Piping and Safe End Materials of Construction (continued) Nozzle Unit 1 Unit 2 Unit 3 Designation / System SA333 GR 6 SA333 GR 6 SA333 GR 6 A358 GR 304 A358 GR 304 Core Spray System SA350 GR LF2 SA350 GR LF2 SA350 GR LF2 Cast A351 CF8 SA420 GR WPL6 SA420 GR WPL6 Cast A351 CF8 Cast A351 CF8 Rx Vessel Head Spray A508 CL 2 A508 CL 2 A508 CL 2 (Not Used) SA105 GR II SA105 GR II SA105 GR II A/SA 106 Grade B A312 TP304 A/SA 106 Grade C A106 GR B A508 CL 2 A508 CL 2 A106 GR B Rx Vessel Head Vent A/SA 105 SA105 GR II A508 CL 2 SA105 GR II A182 F304 SA105 GR II SA216 WCB SA403 WP304 A/SA 234 GR WPB A234 SA376 TP304 SA376 TP316 A312 SA375 TP316 Reactor Vessel Level SA312 TP304 A376 TP304 SA312 TP316 Instrumentation SA312 TP316 A376 TP316 SA376 TP316 SA182 F316 A182 F316 SA182 F316 SA182 F316L A312 TP304 A312 TP304 A312 TP304 A312 TP316 A312 TP316 A312 TP316 SLC and Core Plate A376 TP304 A376 TP304 A376 TP304 Differential Pressure A376 TP316 A376 TP316 A376 TP316 A 182 GR F304 A 182 GR F304 A 182 GR F304 A 182 GR F316 A 182 GR F316 A 182 GR F316 2-47

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-9a RCPB Piping and Safe End Materials of Construction (continued) Nozzle Unit 1 Unit 2 Unit 3 Designation / System SA376 TP316 A376 TP304 A376 TP304 Jet Pump SA312 TP316 A376 TP316 A376 TP316 Instrumentation SA182 F316 A 182 GR F304 A 182 GR F304 SA182 F316L A 182 GR F316 A 182 GR F316 A/SA 376 TP304 A 376 TP304 SA376 TP316 SA-105 SA105 SA182 F316L Reactor Vessel Drain A/SA 182 F304 A182 F304 SA182 F304 A/SA 182 F316 A182 F316 SA351 CF8M SA351 CF8M SA351 CF8M CRD Return Line SA 182 F316L SA 182 F316L SA 182 F316L (Capped/Not Used) Note:

1. The Unit 1 Reactor Recirculation System (RRS) piping (pump suction and discharge piping, the ring header, the riser piping, and the inlet and outlet safe ends) has been replaced with ASME SA376 Type 316 NG stainless material, which is resistant to IGSCC. The recirculation suction and inlet safe ends are an improved crevice-free design. The replacement piping utilized an improved design which eliminated several piping welds. Additionally, the use of EPRI welding techniques (such as machine welding where practical and reduced energy input) and the application of a Mechanical Stress Improvement Process (MSIP) were utilized to reduce the potential for IGSCC.

As a result of these efforts, all the Unit 1 RRS welds are Category A welds in accordance with NUREG-0313, Revision 2 classifications. The Unit 2 RRS pump suction and discharge piping and the bottom portion of the ten system risers are fabricated with ASTM A358 Type 304 stainless steel. The top portion of each riser including the riser elbow is ASME SA403 WP 316 NG. The recirculation inlet safe ends are ASME SA376 Type 316 NG. The recirculation inlet safe ends are an improved crevice-free design. To mitigate weld residual stresses in the Unit 2 RRS, the application of a MSIP or Induction Heat Stress Improvement (IHSI) were utilized on the accessible welds to reduce the potential for IGSCC. The Unit 3 RRS consists of suction and discharge piping fabricated with ASTM A358 Type 304 stainless steel. All piping downstream of the 28 inch recirculation discharge piping, including the cross, ring header, and the risers are ASME SA403 WP 316 NG stainless steel. The recirculation inlet safe ends are ASME SA376 Type 316 NG. The 2-48

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) recirculation inlet safe ends are an improved crevice-free design. The replacement piping utilized an improved design which eliminated several piping welds. To mitigate weld residual stresses in the Unit 3 RRS, the application of a MSIP or IHSI were utilized on the accessible welds to reduce the potential for IGSCC. 2-49

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.1-9b Summary of RCPB Welds per Generic Letter 88-01/BWRVIP-75-A IGSCC Weld Category Unit 1 Unit 2 Unit 3 A 136 48 70 B 0 0 0 C 10 112 79 D 5 9 2 E 0 16 10 F 0 0 0 G 2 2 2 2-50

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.2 Mechanical and Civil Engineering 2.2.1 Pipe Rupture Locations and Associated Dynamic Effects Regulatory Evaluation SSCs important to safety could be affected by the pipe-whip dynamic effects of a pipe rupture. The NRCs acceptance criteria are based on GDC-4, which requires SSCs important to safety to be designed to accommodate the dynamic effects of a postulated pipe rupture. Specific NRC review criteria are contained in SRP Section 3.6.2. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-40. Browns Ferry Pipe Rupture Locations and Associated Dynamic Effects are described in Browns Ferry UFSAR Section 12.2, Principal Structures and Foundations, and Appendix M, Report on Pipe Failures Outside Containment in the Browns Ferry Nuclear Plant. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 10.1 of the CLTR addresses the effect of EPU on High Energy Line Breaks (HELBs). The results of this evaluation are described below. 2-51

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Inside containment, the high-energy piping systems potentially affected by EPU are: Main steam Main steam drains Reactor Core Isolation Cooling (RCIC) steam line High Pressure Coolant Injection (HPCI) steam line FW Main steam safety relief valve piping (between the MSL and each SRV) Reactor pressure vessel head vent line Main steam drains, RCIC steam, and HPCI steam line flow rates, pressures, and temperatures are unchanged from CLTP to EPU operating conditions. Therefore, the MS attached piping did not have additional break locations resulting from EPU operating conditions. Outside containment, high-energy piping systems include: Main steam Feedwater HPCI (normally pressurized steam supply to turbine drive) RCIC (normally pressurized steam supply to turbine drive) RWCU Of these, the only systems affected by EPU are main steam and feedwater. While main steam pressures and temperatures do not increase with EPU, the piping stress analysis of record was performed for EPU conditions to account for the increase in flows. A review was performed of piping stresses that increased due to EPU and postulated pipe break locations. The review was in accordance with the requirements of the current licensing basis methodology. No changes to the implementation of the existing criteria for defining pipe break and crack locations and configurations are being made for EPU. No new break or crack locations are required to be postulated as a result of the increased piping stresses associated with EPU. No changes to the implementation of the existing criteria for special features, such as augmented in-service inspection (ISI) programs or the use of special protective devices, such as pipe-whip restraints are being made for EPU. For EPU, high energy line breaks (HELBs) are evaluated for their effects on equipment qualification. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: 2-52

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR Steam Lines Generic Disposition Meets CLTR Liquid Lines Plant Specific Disposition 2.2.1.1 Steam Line Breaks The CLTR states that there is no effect on steam line breaks because steam conditions at postulated break locations are unchanged. Therefore, EPU has no effect on the mass and energy releases from a HELB in a steam line. A review of the heat balances produced for the Browns Ferry EPU confirmed that ((

))

The evaluation of the steam line HELB events in the Browns Ferry licensing basis meet all CLTR dispositions except as discussed below for the plant-specific MS Line intermediate break. At Browns Ferry, the intermediate size steam line break is defined in the current licensing basis as the largest main steam line break (MSLB) that is not isolated by the MS line high flow sensors, which are set to limit flow below the analytical limit of 144% of rated MS flow. Because rated MS flow increases for EPU, the mass flow rate for the intermediate steam line break also increases. The CLTP design mass and energy releases for the intermediate size MSLB were evaluated and revised to account for the increase in the percent rated steam flow at EPU. This evaluation included the effects identified within the 10 CFR Part 21 Potentially Reportable Condition Notification: Error in Main Steam Line High Flow Calculational Methodology (Reference 42). Therefore, the intermediate size MSLB remains bounded by the doubled-ended break in the main steam valve vault. The mass and energy releases for the doubled-ended break in the main steam valve vault at EPU are unchanged from the CLTP analyses. 2.2.1.2 Liquid Line Breaks As stated in Section 10.1 of the CLTR, EPU may increase subcooling in the reactor vessel, which may lead to increased break flow rates for liquid line breaks. For Browns Ferry, the increase in vessel subcooling could affect the RWCU line break analysis. In addition, operation at EPU conditions requires an increase in the FW flow, which results in an increase in FW system pressure and a small increase in RWCU discharge pressure at the FW inlet. This increase in FW system pressure may lead to increased break flow rates for FW line breaks. For Browns Ferry, mass and energy releases for HELBs were re-evaluated at EPU conditions. The plant-specific evaluation of liquid line breaks included the RWCU and FW systems as well as the effect of increased RWCU and FW operating pressure on pipe whip and jet impingement. 2-53

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The results of the Browns Ferry evaluation of liquid line breaks outside containment are provided in Table 2.2-1. 2.2.1.2.1 RWCU Line Breaks Operation at EPU involves an increase in the steam and feedwater flows, which results in a small increase in downcomer subcooling. This condition results in a small increase in the CLTP RWCU System mass flow rates. New mass and energy releases for RWCU line breaks have been analyzed for EPU conditions which include the effects of subcooling and the small increase in RWCU discharge pressure. Structural effects of increased peak pressures were reviewed and found to be acceptable. The effects of increased peak calculated room temperatures in the Reactor Building resulting from the RWCU line breaks are addressed in the EPU EQ analysis. See Section 2.3.1 for EQ results. 2.2.1.2.2 Feedwater System Line Break The FW System process conditions are changed for EPU to support the increased feedwater flows. The base RELAP5 feedwater system break models for CLTP were revised to account for EPU changes in the FW System process conditions including the changes to system temperatures and to the increased discharge head for the FW pumps. The EPU evaluation concludes that the associated changes in FW system line break mass and energy release will not challenge the bases for the current HELB analysis because the effects of the FW line break in the main steam valve vault are bounded by the effects of the postulated MSLB. Also, for the portion of the smaller RWCU piping attached to the FW piping in the main steam valve vault, mass and energy releases from breaks in the smaller RWCU piping are bounded by the FW line break mass and energy releases. 2.2.1.2.3 Pipe Whip and Jet Impingement Pipe whip and jet impingement loads resulting from high energy pipe breaks are a function of system pressure, temperature, and size, as well as proximity to relatively constant pressure sources connected to the line, and the effect of friction or line area restrictions between the break and the constant pressure source. Inside containment, the only high-energy piping that experiences an increase in operating pressure due to EPU is in the FW and RRS systems. Outside containment, the only high-energy piping experiencing an increase in operating pressure due to EPU is the FW and RWCU system. Increased FW fluid conditions associated with EPU will not affect the current HELB analysis in the main steam valve vault. The increase in FW fluid conditions are bounded by the MSLB, which remains unchanged from CLTP conditions. Pipe whip and jet impingement loads resulting from high energy pipe breaks are dependent on system pressure. The feedwater system operating pressure for the design of pipe whip and jet impingement is based on the feedwater system design pressure of 1,250 psig, which is higher than the EPU pressure of 1,200 psig. Therefore, EPU will have a negligible effect on FW pipe whip and jet impingement and will still be bounded by current design analyses. 2-54

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The RRS will have a slight increase in operating pressure for EPU conditions. This slight increase in operating pressure remains bounded by the existing margin in the current analysis. Therefore, EPU will not affect the RRS pipe whip and jet impingement loads. The potential effect of increased FW, RRS and RWCU operating pressures at the existing HELB break locations relative to the subsequent effects of pipe whip (targets) and jet impingement loads were evaluated. The resulting EPU pipe whip (targets) and jet impingement loads are bounded by the current licensing basis pipe whip and jet impingement loads. The adequacies of pipe stress and pipe support loads relative to pipe whip and jet impingement loads are evaluated in Section 2.2.2. Therefore, Browns Ferry meets all CLTR dispositions for liquid line breaks. Conclusion TVA has evaluated the effects of the proposed EPU on rupture locations and associated dynamic effects. The evaluation indicates that SSCs important to safety will continue to meet the requirements of the current licensing basis with respect to the dynamic effects of a postulated pipe rupture following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the dynamic effects associated with the postulated rupture of piping. 2.2.2 Pressure-Retaining Components and Component Supports Regulatory Evaluation The NRCs acceptance criteria are based on (1) 10 CFR 50.55a and GDC-1, insofar as they require that SSCs important to safety be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-2, insofar as it requires that SSCs important to safety be designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions; (3) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (4) GDC-14, insofar as it requires that the Reactor Coolant Pressure Boundary (RCPB) be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; and (5) GDC-15, insofar as it requires that the RCS be designed with margin sufficient to ensure that the design conditions of the RCPB are not exceeded during any condition of normal operation. Specific NRC review criteria are contained in SRP Sections 3.9.1, 3.9.2, 3.9.3, and 5.2.1.1 and other guidance provided in Matrix 2 of RS-001. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a 2-55

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-1, 2, 9, 33, 34, 40, and 42. The Pressure Retaining Components and Component Supports are described in Browns Ferry UFSAR Section 3.3, Reactor Vessel Internals Mechanical Design, and Chapter 12, Structures and Shielding. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). Systems and system components, programs used to manage aging effects, and time limited aging analyses are documented in NUREG-1843, Sections 2.3, 3.1, 3.5, and 4.3. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 3.2.2, 3.4 and 3.5 of the CLTR addresses the effect of EPU on Reactor Vessel Structural Evaluation, Flow-Induced Vibration and Piping Evaluation, respectively. The results of this evaluation are described below. 2.2.2.1 Flow-Induced Vibration (FIV) The FIV evaluation addresses the influence of an increase in flow during EPU on RCPB piping and RCPB piping components. 2-56

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Key applicable structures include the Reactor Recirculation System (RRS) piping and suspension, the Main Steam (MS) system piping and suspension, the FW system piping and suspension and the branch lines attached to the MS system piping or FW system piping. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Structural Evaluation of Recirculation Piping Generic Disposition Meets CLTR Structural Evaluation of Main Steam and FW Piping Generic Disposition Meets CLTR Safety-Related Thermowells and Probes Plant Specific Disposition Structural Evaluation of Core Flow Dependent RPV Meets CLTR Plant Specific Internals Disposition 2.2.2.1.1 Structural Evaluation of Recirculation Piping The CLTR states that there is no significant increase in the recirculation flow rate at EPU conditions. The recirculation system drive flow remains unchanged at 17.7 Mlb/hr per loop at EPU conditions, resulting in no increase from CLTP to EPU operation. Consequently, the FIV levels of the RRS components are expected to remain essentially the same. Because RRS flow rates for EPU are essentially the same as previously experienced, no further evaluation or testing of the FIV levels of the RRS piping, branch piping (e.g., attached residual heat removal piping), or its suspension system is required. The FIV effect on RRS piping inside containment at Browns Ferry meets all CLTR dispositions because the nominal reactor dome pressure remains the same and the RRS maximum drive flow does not increase. 2.2.2.1.2 Structural Evaluation of Main Steam and Feedwater Piping The CLTR states that MS and FW flow rates increase due to the power uprate. As a result of the increased flow rates and flow velocities, the MS and FW piping experience increased vibration levels, approximately proportional to the square of the flow velocities. Thus, for Browns Ferry, vibration levels may increase by up to 58.5% of OLTP. The ASME Code (NB-3622.3) and nuclear regulatory guidelines require some vibration test data be taken and evaluated for these high energy piping systems during initial operation at EPU conditions. Vibration data for the MS and FW piping inside containment will be acquired using remote 2-57

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) sensors, such as accelerometers and displacement probes as appropriate. A piping vibration startup test program, which meets the ASME code and regulatory requirements, will be performed. Therefore, the assessment of the structural evaluation of MS and FW piping meets all CLTR dispositions. The EPU increase in MS and FW flow rates only affects the structural analysis of the MS and FW piping. The structural analyses of all other systems inside and outside of containment are not affected by the increase in MS and FW flow rates. FIV testing of the MS and FW piping system will be performed during EPU power ascension. Additional information related to the MS piping is provided in Section 2.5.4.1.1. The piping vibration monitoring program for Browns Ferry EPU is described in LAR Attachment 45. LAR Attachment 45 also provides projections of piping vibration levels at the EPU power level for piping systems both inside and outside containment. 2.2.2.1.3 Safety-Related Thermowells and Probes As explicitly stated in Section 3.4 of the CLTR, MS and FW flow rates increase due to the power uprate. The CLTR requires a plant-specific evaluation of safety-related thermowells and probes in the MS and FW piping systems at EPU conditions. Browns Ferry Units 1, 2 and 3 have no safety-related thermowells in the MS lines, but do have safety-related thermowells installed on the MSRV discharge pipes in MS system. There are no safety-related sample probes installed in the MS and FW systems at Browns Ferry. The MS system flow increased from 3.54 Mlb/hr per line at CLTP to 4.11 Mlb/hr per line resulting in an increase of 16.2% during EPU operation. The FW system flow increased from 7.05 Mlb/hr per line at CLTP to 8.20 Mlb/hr per line resulting in an increase of 16.2% during EPU operation. The RRS thermowell evaluation was performed with a bounding ICF RRS flow rate of 18.81 Mlb/hr under EPU condition. The safety-related thermowells and probes in the MS, FW and RRS piping systems were evaluated and found to be adequate for EPU. The methodology used to evaluate the FIV effects on piping components under EPU is described in Section 3.4.1 of the CLTR. This evaluation utilizes computer program SAP4G07 to develop dynamic finite element models of the MS/FW/RRS thermowells and sample probes. Three-dimensional beam elements with six degrees of freedom are used to model the thermowell and sample probe and their sockolet/pipe weld. At the sockolet/pipe weld to the outer pipe wall, all six degrees of freedom are fixed. The masses of the thermowells or sample probes and their sockolet/pipe weld are lumped at the nodal points, which include both the structural mass and fluid mass displaced by the thermowells and sample probes. These added masses are used to account for the effects of fluid on the thermowells and sample probes vibration responses. This evaluation of the piping components follows the FIV analysis guideline, as outlined in ASME code N-1300 (Reference 43). The resonance separation rule: fn/fs<0.7 or fn/fs >1.3 as established in Reference 43 (N-1324.1(d)) is used to determine if there exists an adequate 2-58

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) separation between the vortex shedding frequencies and the natural frequencies of the piping components. When one of the natural frequencies is close to the vortex shedding frequency, the equation from Table N-1324.2(a)-1 of Reference 43 is used to calculate structural response of the piping components under lock-in condition. For off lock-in (no resonance) condition, the structural response was calculated using the standard methods (Reference 43, N-1324.2). The safety-related thermowells and sample probes in the recirculation piping system were evaluated to be adequate for the RRS flow associated with ICF at EPU conditions. The evaluation in accordance with ASME code, Section III, Division 1, Appendices, N-1300 (Reference 43) concludes that the safety-related thermowells and sample probes in MS, FW, and RRS systems at Browns Ferry Units 1, 2, and 3, remain structurally adequate to withstand the FIV effects under EPU conditions. The maximum vibratory stress is calculated by using the square root of the sum of the squares (SRSS) of the responses in lift and drag directions with Stress Concentration Factor (SCF) of 2.0. The results of the analyses are presented below: Criteria Maximum Stress Item Component Allowable (psi) fn/fs <0.7 under EPU (psi) OR fn/fs >1.3 MSRV Discharge 1 6,075(1) 13,600 fn/fs<0.7 Line Thermowell 2 FW Thermowell 1,329 13,600 fn/fs>1.3 3 RRS Thermowell 3,020 13,600 fn/fs>1.3 4 RRS Sample Probe 142 13,600 fn/fs>>1.3 Note: (1) Because the structural fundamental frequency of MS thermowell is lower than the vortex shedding frequency, the structural response of the thermowell at the lock-in condition is conservatively assumed and calculated by the equation from table N-1324.2(a)-1, Reference 43. This FIV stress bounds the current MS thermowell under EPU, which is expected to be less than 6,075 psi at the off lock-in condition. Therefore, Browns Ferry meets all CLTR dispositions for safety-related thermowells and probes. 2.2.2.2 Piping Evaluation 2.2.2.2.1 Reactor Coolant Pressure Boundary Piping (Non-FIV) Evaluation The RCPB system evaluation consists of a number of safety-related piping subsystems that move fluid through the reactor and other safety systems. The code of record for Browns Ferry safety-related piping, with the exception of the primary containment torus attached piping, is USA 2-59

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Standard (USAS) B31.1.0 - 1967 (Reference 44). Because Reference 44 is incomplete with respect to plant operating conditions and code equations, the later ASME Section III code has been used in the development of load combinations and allowable stress criteria. Section III of the 1971 ASME Boiler and Pressure Code, including the Summer 1973 addenda, Subsection NC is used as guidance. However, analysis parameters, such as material allowable stresses, stress intensification factor (SIF) coefficient of thermal expansion, and elastic modulus are in accordance with USAS B31.1.0 - 1967. The Browns Ferry EPU piping evaluations are performed to these same codes of record without exception. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Structural Evaluation for Unaffected Safety-Related Meets CLTR Generic Piping Disposition Structural Evaluation for Affected Safety-Related Meets CLTR Plant Specific Piping Disposition 2.2.2.2.1.1 Structural Evaluation for Unaffected Safety-Related Piping As stated in Section 3.5.1 of the CLTR, the flow, pressure, temperature, and mechanical loading for most of the RCPB piping systems do not increase for EPU. Consequently, there is no change in stress and fatigue evaluations. ((

                                      )) and therefore, Browns Ferry meets all CLTR dispositions for the structural evaluation for unaffected safety-related piping. Table 2.2-2 provides the justification for confirming the generic disposition for the above piping systems and segments.

Section 2.8.4.2 demonstrates that the RCPB piping remains below the ASME pressure limit during the most severe pressurization transient. 2-60

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Pipe Whip and Jet Impingement Pipe whip and jet impingement loads resulting from high energy pipe breaks are a function of system pressure, temperature, and size, as well as proximity to relatively constant pressure sources connected to the line, and the effect of friction or line area restrictions between the break and the constant pressure source. The resulting EPU pipe whip and jet impingement loads are bounded by the current licensing basis pipe whip and jet impingement loads. Additionally, pipe stress calculations were revised to reflect EPU operating conditions for MS and FW. There are no increased pipe stress levels above the thresholds required for postulating HELBs, except at locations already evaluated for breaks. As a result, EPU conditions do not result in new HELB locations, nor affect existing HELB evaluations of pipe whip restraints and jet targets. 2.2.2.2.1.2 Structural Evaluation for Affected Safety-Related Piping As stated in Section 3.5.1 of the CLTR, the FW and MS piping and associated branch piping up to the first anchor or support will experience an increase in the flow, pressure, and/or temperature, resulting in an increase in operating stress and fatigue. For all systems, the maximum stress levels were reviewed based on specific increases in temperature, pressure, and flow rate. EPU also increases the operating pipe support loads due to the above effects as well as increased fluid transient turbine stop valve closure (TSVC) loads that result from the increased steam flow rates. The analyses of record for MS piping were revised to evaluate TSVC loads for all pipe nodes. These evaluations determined that the interface loads on snubbers, struts, guides, and flange connections at EPU conditions are within the design limits (capacities) of these components, and any required modifications have been completed. The analyses of record for FW piping inside containment were revised to account for the pressures and temperatures of EPU operating conditions. Therefore, design loads and stresses remain bounding for EPU. These evaluations determined that the interface loads on snubbers, struts, guides, and flange connections at EPU conditions are within the design limits (capacities) of these components and any required modifications have been completed. For RCPB MS piping outside containment between the containment penetration and the outboard main steam isolation valve (MSIV), the TSVC fluid transient was evaluated in the revised piping stress analyses of record for EPU conditions. The MS analysis resulted in MS piping outside containment meeting all Code criteria. There are no required pipe support modifications for MS piping outside containment due to EPU. There are no pipe supports on the RCPB MS piping outside containment, between the containment penetration anchor and the outboard MSIV. The FW system has been evaluated and found to meet the appropriate code criteria for EPU conditions, based on the design margins between calculated stresses and code limits in the current design. All piping stresses are below the code allowable values of the Browns Ferry analysis of record. The MS piping was analyzed for TSVC loads. Stresses in the MS and attached piping are below the code allowable values. 2-61

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The pipe supports of the systems affected by EPU loading increases are reviewed to determine if there is sufficient margin to code acceptance criteria to accommodate the increased loadings. RCPB FW piping inside containment support loads are acceptable for EPU. The MS analysis resulted in MS piping inside containment meeting all Code criteria. EPU pipe stresses and support loads for RCPB FW piping outside containment are acceptable and meet all Code criteria. The MS analysis resulted in MS piping outside containment meeting all Code criteria. Main Steam and Associated Piping System Evaluation For Browns Ferry, an increase in flow and mechanical loads was evaluated on a plant-specific basis consistent with the methods specified in Appendix K of ELTR1. Plant-specific evaluations are required to demonstrate that the calculated stresses are less than the code allowable limits in accordance with the requirements of the applicable code of record in the existing design basis stress report. The MS and associated branch piping inside containment and RCPB piping outside containment was evaluated to the USAS B31.1.0 - 1967 stress criteria (Reference 44), including the effects of EPU on piping stresses, piping support loads including the associated building structure, penetrations, piping interfaces with the RPV nozzles, flanges, and valves. Allowable stress values for MS piping inside containment and associated branch lines were taken from USAS B31.1.0 - 1967 (Reference 44). SRP Section 3.6.2, MEB 3-1 criteria is not a licensing commitment for Browns Ferry, but all pipe ruptures are postulated in accordance with current licensing basis. Because the MS piping pressures and temperatures are not significantly affected by EPU, there is no effect on the analyses for these parameters. Seismic inertia loads, seismic building displacement loads, and MSRV discharge loads are not affected by EPU; thus, there is no effect on the analyses for these load cases. The increase in MS flow results in increased fluid transient loads from a TSVC transient. The TSVC loads bound the MSIV closure loads because the MSIV closure time is significantly longer than the TSV closure time. The analyses of record were revised for MS to include the TSVC transient. The TSVC transient loading will increase due to the increase in the MS flow rate under EPU. Detailed and conservative modeling of this transient was performed to ensure that components, pipe stress, and support loads do not exceed their allowable code limits. Pipe Stresses A review of the increase in flow associated with EPU indicates that piping stress changes do not result in stress limits being exceeded for the MS system and attached branch piping or for RPV nozzles and containment penetrations. The revised design analyses have sufficient margin between calculated stresses and USAS B31.1.0 allowable limits (see Table 2.2-3a) to justify operation at EPU conditions. The pressure and temperature of the MS piping are unchanged for EPU. 2-62

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Similarly, the branch pipelines (Safety Relief Valve Discharge Line (SRVDL), RCIC, HPCI, RPV head vent, and main steam drains including the MSIV drain) connected to the MS headers were evaluated to determine the effect of the increased MS flow on the lines. This evaluation concluded that there is no adverse effect on the existing MS branch line qualifications due to the increased MS flows resulting from EPU. As with the MS piping, the pressures and temperatures for these branch pipelines do not change as a result of EPU. A review was performed of postulated pipe break locations. The review was conducted in accordance with the requirements of the current licensing basis methodology. As a result of this review, no new postulated break locations were identified. Based on existing margins available for the MS piping, it was concluded that EPU does not result in reactions in excess of the current design capacity. The pipe stress analyses of record for RCPB MS piping outside containment were revised to evaluate the increased TSVC fluid transient loading with EPU. The revised analysis for the MS system outside containment demonstrates that the design has sufficient margin between calculated stresses and the allowable limits in the code of record, USAS B31.1.0 - 1967 (Reference 44) to justify operation at EPU conditions. Pipe Supports A review of the change in flow associated with EPU indicates that piping load changes do not result in load limits being exceeded for the MS piping system; therefore, the pipe supports for the MS piping system are adequate at EPU conditions. The current design analyses were updated for conditions representative of EPU operation in the MS piping system as applicable. No inside containment pipe support modifications are required for EPU. Main Steam Isolation Valves The MSIVs are part of the RCPB, and perform the safety function of steam line isolation during certain abnormal events and accidents. The MSIVs must be able to close within a specified time range at all design and operating conditions. They are designed to satisfy leakage limits set forth in the plant TSs. These design requirements are not adversely affected by increased EPU flow, thus the original design remains adequate for EPU conditions. The MSIVs have been evaluated, as discussed in Section 4.7 of ELTR2, Supplement 1. The evaluation covers both the effects of the changes to the structural capability of the MSIV to meet pressure boundary requirements, and the potential effects of EPU-related changes to the safety functions of the MSIVs. The generic evaluation from ELTR2 is based on: (1) a 20% thermal power increase; (2) an increased operating dome pressure to 1,095 psia; (3) a reactor temperature increase to 556°F; and (4) steam and FW flow increases of about 24%. Table 1-2 provides the maximum nominal dome pressure and temperature as well as the changes in steam and FW flows. From these parameters, it can be determined that the evaluation from ELTR2 is applicable to Browns Ferry. The MSIV has design features that ensure that MSIV closure time is maintained within the stroke time limits. The closing time of the MSIVs is controlled by the design of the hydraulic control valves and the function of the hydraulic damper. Therefore, the MSIV performance is 2-63

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) bounded by conclusions of the evaluation in Section 4.7 of ELTR2, and the Browns Ferry MSIVs are acceptable for EPU operation. Feedwater System Evaluation The pressure changes are insignificant for EPU and are bounded by those used in the analysis of record. The calculations of record were revised to reflect EPU operating temperatures (Table 2.2-5). The current licensing basis for the reactor FW system inside containment complies with the Browns Ferry code of record stress criteria (Reference 44) for the effect of thermal expansion displacement on the piping snubbers, hangers, and struts. Piping interfaces with RPV nozzles, penetrations, flanges, and valves also remain valid per current licensing basis. Note that the FW flow change of approximately 16.2% does not affect the reactor FW piping system loads and stresses because fluid transient and fatigue loads are not a part of the design basis in the original stress calculation. SRP Section 3.6.2, MEB 3-1 criteria is not a licensing commitment for Browns Ferry, but all pipe ruptures are postulated in accordance with the current licensing basis. This discussion also applies to the FW system and associated branch piping outside containment. The FW piping design at EPU conditions was evaluated for compliance with the analysis code of record stress criteria (Reference 44). Because the FW system piping operating temperatures increase slightly due to EPU, the effect of these parameters on the existing analyses was evaluated. Seismic inertia loads, and seismic building displacement loads are not affected by EPU; thus, there is no effect on the analyses for these load cases. Other external loading conditions are not changed by EPU. For the FW piping inside and outside containment, there is no FW system fluid transient analysis in the original or current design basis analysis so the increase in FW system flow has no effect on the current analysis. Although fluid transient analyses are not required as part of the piping design basis, an analysis was performed that assessed the effect of transients on the FW piping at EPU conditions. The bounding transient considered was for a simultaneous trip of all three FW pump turbines that results in transient loading on piping as the FW pumps coast down. The results of this analysis showed that the FW piping is acceptable for FW fluid transients that occur at EPU conditions and that the FW piping design has sufficient margin between calculated stresses/loads and the allowable limits in the code of record (Table 2.2-3d). Pipe Stresses For FW piping inside containment, a review of the changes in operating pressure, temperature and flow associated with EPU indicates that piping stress changes do not result in stress limits being exceeded for the reactor FW piping system, for RPV nozzles, and at postulated pipe break locations (see Table 2.2-3b). The current Browns Ferry design analyses were revised for conditions representative of EPU operating modes in the FW piping system. 2-64

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) This discussion also pertains to the RCPB portion of the FW piping outside containment. A review of the increase in flow, operating pressure, and temperature associated with EPU indicates that piping load changes do not result in load limits being exceeded for the FW piping system and attached branch piping. A review was also performed of postulated pipe break locations in accordance with the current licensing basis methodology. As a result of this review, no new postulated break locations were identified. The analysis for the FW system outside containment demonstrates that the design has sufficient margin between calculated stresses and the allowable limits in the code of record (Reference 44). Pipe Supports A review of the changes in operating pressure, temperature and flow associated with EPU indicates that piping load changes and thermal expansion displacements do not result in load limits being exceeded for the FW piping system; therefore, the pipe supports for the FW piping system are adequate at EPU conditions. The current design analyses were revised for conditions representative of EPU operating modes in the FW piping system. Seismic inertia loads and seismic building displacement loads are not affected by EPU; thus, there is no effect on the analyses for these load cases. The FW system piping outside containment was evaluated for the effects of EPU temperature increase on the piping design analyses. It was concluded that EPU does not have an adverse effect on FW pipe support design, and all loads were within limits. Other Piping Evaluation As previously noted, the nominal operating pressure and temperature of the reactor are not changed by EPU. Aside from MS and FW, no other system connected to the RCPB experiences a material increase in flow rate at EPU conditions. Only minor changes to fluid conditions are experienced by these systems due to higher steam flow from the reactor and the subsequent change in fluid conditions within the reactor. Additionally, piping dynamic loads due to MSRV discharge at EPU conditions are bounded by those used in the existing analyses. These systems were previously evaluated for compliance with the code of record (Reference 44) stress criteria as required. Because none of these piping systems connected to the RCPB experience any significant change in operating conditions due to EPU, they are all acceptable as currently designed. Therefore, Browns Ferry meets all CLTR dispositions for RCPB piping. 2.2.2.2.2 Balance-of-Plant (BOP) Piping Systems The BOP piping systems evaluation consists of a number of piping subsystems that move fluid through systems outside the RCPB piping. Browns Ferry meets all CLTR dispositions. The topics considered in this section are: 2-65

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Structural Evaluation for Unaffected Safety-Related Meets CLTR Generic Piping Disposition Structural Evaluation for Unaffected Non-Safety Meets CLTR Generic Related Piping Disposition Structural Evaluation for Affected Safety-Related Meets CLTR Plant Specific Piping Disposition Structural Evaluation for Affected Non-Safety Meets CLTR Plant Specific Related Piping Disposition 2.2.2.2.2.1 Structural Evaluation for Unaffected BOP Piping As stated in Section 3.5.2 of the CLTR, the flow, pressure, temperature, and mechanical loading for some BOP piping systems do not increase for EPU. Consequently, there is no change in stress and fatigue evaluations and these BOP piping systems meet all CLTR dispositions. The following piping for BOP and NSSS outside containment only were confirmed to be unaffected by EPU conditions because either the flow, temperature, pressure, or other mechanical loads do not change in the system for EPU or the change is insignificant and has no effect on the piping system design: Auxiliary Steam Piping Circulating Water Piping Condensate Storage and Supply Piping Condenser Air Removal Piping CRD Piping Drywell (DW) Chilled Water Piping Fuel Pool Cooling (FPC) and Cleanup Piping Liquid Radwaste Piping Off Gas Piping Plant Chilled Water Piping Raw Service Water (RSW) Piping Raw Cooling Water (RCW) Piping Post-Accident Sampling Piping 2-66

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Process Sampling Piping RWCU Piping Standby Liquid Control (SLC) Piping (Outside Containment) Emergency Equipment Cooling Water (EECW) Piping Reactor Building Cooling Water (RBCCW) Piping Main Steam Relief Valve Drain Line (MSRVDL) Piping Beyond the First Anchor to the Quenchers RCIC Piping Outside Containment HPCI Piping Outside Containment 2.2.2.2.2.2 Structural Evaluation for Affected BOP Piping As stated in Section 3.5.2 of the CLTR, the FW and MSL piping including the associated branch piping will experience an increase in the flow and/or temperature resulting in an increase in stress. The Browns Ferry piping systems determined to be affected by EPU operation include: RHR Piping CS Piping Outside Containment MS Piping (Outside Containment) Extraction Steam Piping FW Piping (Outside Containment) Condensate Piping Moisture Separator Drains Piping FW Heater Vents and Drains Piping Cross Around Relief Valve (CARV) Discharge Piping Condensate Demineralizer Piping For those systems with analyses, the maximum stress level analysis results were reviewed based on specific increases in temperature, pressure and flow rate (see Tables 2.2-4a through 2.2-4f and Table 2.2-5). The code of record for Browns Ferry safety-related piping, with the exception of the primary containment torus attached piping, is USAS B31.1.0 - 1967 (Reference 44). Because Reference 44 is incomplete with respect to plant operating conditions and code equations, the later ASME Section III code has been used in the development of load combinations and allowable stress criteria. Section III of the 1971 ASME Boiler and Pressure Code, including the Summer 1973 addenda, Subsection NC is used as guidance. However, analysis parameters, such as material allowable stresses, SIF coefficient of thermal expansion, and elastic modulus are in 2-67

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) accordance with USAS B31.1.0 - 1967. The Browns Ferry code of record for primary containment torus attached piping is ASME Boiler and Pressure Code Section III, Division 1, Subsection NC, through the Summer 1977 addenda. The piping systems affected by EPU have been evaluated in accordance with these codes of record criteria for the EPU conditions based on the design margins between actual stresses and code limits in the original design. All piping stresses have been found to be below the code allowable limits of the present code of record. A review was performed of postulated high energy pipe break locations in accordance with the requirements of the current licensing basis methodology. As a result of this review, no new postulated break locations were identified. Details regarding analyses pertaining to the dynamic effects of high-energy piping failures outside containment are provided in Section 2.2.1 and the environmental effects of piping failures outside containment are discussed in Section 2.5.1.3. Pipe failures in high energy non-nuclear safety and field routed piping not rigorously analyzed are postulated at all adverse locations with regards to systems, structures, and components that are important to safety. Although condensate piping is high energy per Browns Ferry licensing basis definitions, a break in the condensate system does not affect structures, systems, and components that are important to safety. Therefore, the condensate system has not been evaluated for postulated pipe failures. Browns Ferry does not evaluate stratification in the piping evaluations of record and does not monitor for stratification at CLTP conditions. This disposition remains unchanged for EPU. Main Steam and Associated Piping System Evaluation The MS piping system outside containment was evaluated for compliance with all codes and standards that are captured under Browns Ferry criteria, including the effects of EPU on piping stresses, equipment nozzles, pipe break postulation, flanges and valves. Temperatures and pressures in the MS piping, including attached MS branch piping and turbine bypass piping, will not increase with EPU. Because MS piping pressures and temperatures do not increase with EPU, there was no effect on the analyses due to these parameters. The increase in MS flow results in increased transient forces from the TSV closure. The TSVC transient load is the only load increase for the MS piping and supports at EPU conditions. For MS piping outside containment, a new pipe stress analysis was performed to evaluate the increased TSVC fluid transient loading with EPU. Detailed TSVC fluid transient forcing functions were developed and the piping stress analysis was evaluated at EPU conditions to determine loads at EPU due to the TSVC transient. The EPU MS analysis resulted in MS piping outside containment meeting all code of record criteria (Reference 44), and the pipe stress results are shown in Table 2.2-4a. With the implementation of support modifications as described in EPU LAR Attachment 47, the revised analysis for the MS system outside containment demonstrates that the design has sufficient margin between calculated stresses and the allowable limits in the Browns Ferry code of record to justify operation at EPU conditions. 2-68

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Pipe Stresses Reactor dome pressure and temperature remain unchanged for EPU. MS piping pressure and temperature at the TSV decrease slightly with increased friction losses at EPU. The results of the EPU stress analysis for MS outside containment demonstrate that the piping design has sufficient margin between calculated stresses and code allowable limits (see Table 2.2-4a) to justify operation at EPU conditions and that no modifications are necessary as a result of EPU. Pipe Supports Based on the MS pipe stress analysis, the pipe support loads, after the implementation of pipe support modifications discussed in LAR Attachment 47, remain within design load limits to Code allowables. Therefore, the MS and associated piping meets all CLTR dispositions. Feedwater System Evaluation Operation at EPU conditions increases stresses on piping and piping system components due to slightly higher operating temperatures. Higher FW operating pressures result from the higher head loss associated with a higher FW flow rate. The increase in FW system operating pressure at EPU remains bounded by the FW system design pressure (1,250 psig) used in the current licensing basis stress calculations. The FW piping systems outside containment have been evaluated in accordance with the plant code of record criteria (Reference 44) for the EPU conditions based on the design margins between actual stresses and applicable code limits. All piping is below the code allowable of the present code of record (Reference 44). No new postulated pipe break locations were identified. Pipe Stresses Because the FW system piping pressures and temperatures increase slightly due to EPU, the effects of these parameters on the existing analyses were evaluated. Existing FW piping analyses are performed to design pressures and temperatures which remain bounding relative to EPU conditions. Seismic inertia loads, and seismic building displacement loads are not affected by EPU; thus, there is no effect on the analyses for these load cases. Other external loading conditions also are not changed by EPU. For the FW piping outside containment, there is no FW system fluid transient analysis in the existing design basis analysis, so the increase in FW system flow has no effect on the current analysis. The FW temperature and pressure changes are insignificant relative to piping design for EPU. Also, fluid conditions in the FW piping design analyses bound the FW operating conditions at EPU. The flow change does not affect the FW piping design system because fluid transient loading is not a design load in the original or current stress analyses. Therefore, the current licensing basis for the FW system complies with the code of record for the effect of thermal expansion displacement on the piping snubbers, hangers, and struts. Piping interfaces with penetrations, flanges, and valves also remain valid per the current licensing basis. 2-69

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Although fluid transient analyses are not required as part of the FW piping design basis, an analysis was performed that assessed the effect of transients on the FW piping at EPU conditions. The bounding transient considered was for a simultaneous trip of all three FW pump turbines that results in transient loading on piping as the FW pumps coast down. The results of this analysis showed that the BOP FW piping is acceptable for FW fluid transients that occur at EPU conditions and that the design has sufficient margin between calculated stresses/loads and the allowable limits in the code of record (Table 2.2-3d). A review was also performed of postulated high energy pipe break locations in accordance with the requirements of the current licensing basis methodology. As a result of this review, no new postulated break locations were identified. Based on existing margins available for the FW piping, it was concluded that EPU does not have an adverse effect on the FW piping design. Pipe Supports The FW system piping outside containment was evaluated for the effects of EPU operating pressure and temperature increase on the piping design analyses. There is no fluid transient analysis in the current FW design basis for Browns Ferry. Because the existing analyses bound the EPU conditions, it was concluded that EPU does not have an adverse effect on FW pipe support loads and design. Therefore, the FW system meets all CLTR dispositions. Other Piping Evaluation Torus Attached Piping The DBA-LOCA hydrodynamic loads, including the pool swell loads, condensation oscillation (CO) loads and chugging loads are not changed for Browns Ferry EPU. For EPU conditions, the DBA-LOCA containment response loads were evaluated and found to be unchanged by EPU (see Section 2.6.1.2.1) and thus, there are no resulting effects on the containment/torus attached piping and valves. The suppression pool temperature response for large and small break LOCA and other events is evaluated in Section 2.6 and is reported in Table 2.6-1. The peak suppression pool (SP) temperatures for these events at EPU are bounded by the current design analyses of the torus attached piping where the piping was analyzed at a conservatively high peak SP temperature of 187.3°F. With the implementation of support modifications to reinforce an existing pad at an ECCS ring header branch connection as described in EPU LAR 7, the bounding analysis for the ECCS ring header at a bounding peak SP temperature of 187.3°F demonstrates that the design has sufficient margin between calculated stresses and allowable limits in the Browns Ferry code of record to justify operation at EPU conditions. Other external loading conditions for the torus attached piping (e.g., seismic loads) are not affected by EPU. Additionally, piping dynamic loads due to MSRV discharge at EPU conditions (Section 2.6.1.2.2) are bounded by those used in the current Browns Ferry analyses of record. 2-70

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The containment hydrodynamic load evaluation in Section 2.6.1.2.3 states that the SBA event thermal loads at EPU conditions (based on a SP temperature of 146°F) do not bound the thermal loads used in the Reference 45 load definition report (based on a SP temperature of 136°F), which provided thermal load input for subsequent use in the Reference 46 structural analysis of torus attached piping. Review of the Browns Ferry design calculations for the torus attached piping shows that the current design calculations conservatively used the Reference 45 thermal loads from the intermediate break accident (IBA) event (based on a SP temperature of 158°F) for load combinations that required hydrodynamic loading in the SBA and IBA analysis. The Reference 45 IBA event thermal load bounds both the EPU SBA and IBA thermal loads. Other SBA/IBA containment hydrodynamic loads/load combinations at EPU are either unchanged or bounded by the loads/load combinations used in the current design analyses of the torus attached piping. For fire events (classified by TVA for piping design as an emergency condition service level) the EPU peak SP temperature (Section 2.5.1.4) is bounded by the current design analyses of the torus attached piping where the piping was analyzed at a conservative peak SP temperature of 223°F for fire events. The load conditions for the torus attached piping are either unchanged for EPU or bounded by the loads used in the Browns Ferry current analysis of record. With the implementation of pipe support modifications to reinforce an existing pad at an ECCS ring header branch connection as discussed in LAR Attachment 47, the torus attached piping meets all code of record criteria at EPU conditions. Other BOP Piping Systems The piping and pipe supports of the other BOP systems affected by EPU loading increases were reviewed to determine if there is sufficient capacity margin to accommodate the increased loadings. This review shows that existing piping design analyses are performed to pressures and temperatures which bound EPU conditions for some systems. For others, the design analyses have sufficient margin between the calculated and Code allowable stress limits to accommodate the small increases in pressure or temperature with EPU. The evaluations (see Tables 2.2-4b through 2.2-4f and 2.2-5) demonstrate that for all systems, design margins for piping, supports and equipment nozzles are either unaffected by EPU or are adequate to accommodate the increased loads and movements resulting from EPU. For BOP systems that do not require a detailed analysis, pipe routing and flexibility are considered to remain acceptable for EPU. These are non-safety-related BOP systems for which no piping or support analyses are documented. These are generally cold systems (< 200°F) where thermal stresses and displacements are not significant. For these systems, pipe routing and flexibility are considered to remain acceptable, as the pipe routing is not being changed with EPU, and any increase in operating temperature range with EPU is not significant. 2-71

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.2.2.3 Reactor Vessel The RPV structure and support components form a pressure boundary to contain reactor coolant and moderator, and form a boundary against leakage of radioactive materials into the drywell. The RPV also provides structural support for the reactor core and internals. The fatigue of plant-specific components is monitored for license renewal (Reference 11) and those components are listed in the table below. These components have been reviewed (( 2-72

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2-73

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public)

))

The high and low pressure seal leak detection nozzles have been reviewed and found acceptable for 60-year EPU conditions. The effect of EPU was evaluated to ensure that the reactor vessel components continue to comply with the existing structural requirements of the ASME Boiler and Pressure Vessel Code. For the OLTP components under consideration, the ASME Boiler and Pressure Vessel Code, Section III, 1965 Code with Addenda to and including Summer 1965 (Units 1 and 2) and the ASME Boiler and Pressure Vessel Code, Section III, 1965 Code with Addenda to and including Summer 1966 (Unit 3) are applicable. These were used as the governing code and are considered the Code of Construction. However, if a components design has been modified, the governing code for that component was the code used in the stress analysis of the modified component. The following components that ((

                                                                )) were modified since the original construction are:

FW Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976 (Units 1, 2 and 3). Recirculation Inlet Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1980 Edition with Addenda to and including Winter 1981 (Units 1, 2 and 3). Recirculation Outlet Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1980 Edition with Addenda to and including Winter 1981 (Units 1, 2 and 3). Core Spray Nozzle: This component was modified and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976 (Units 1, 2 and 3). CRD Hydraulic System Return Nozzle Cap: This component was newly installed and the governing Code for the modification is the ASME Boiler and Pressure Vessel Code, Section III, 1974 Edition with Addenda to and including Summer 1976 (Units 1, 2 and 3). 2-74

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Jet Pump Instrumentation Seal Safe End: This component was modified and the governing Codes for the modification are the ASME Boiler and Pressure Vessel Code, Section III, 1980 Edition with Addenda to and including Winter 1981 (Units 1 and 2) and the ASME Boiler and Pressure Vessel Code, Section III, 1986 Edition (Unit 3). New stresses are determined by scaling the original stresses based on the EPU conditions (( )). The analyses were performed for the design, the normal and upset, and the emergency and faulted conditions. If there is an increase in jet reaction, pipe restraint or fuel lift loads, the changes are considered in the analysis of the components affected for normal, upset, emergency and faulted conditions. In all evaluations, all six stress components were considered; no simplified single stress methodology was employed. Design Conditions Because there are no changes in the design conditions due to EPU, the design stresses are unchanged and the Code requirements are met. Normal and Upset Conditions The reactor coolant temperature and flows (except: FW flow, recirculation flow, and main steam flow) at EPU conditions are only slightly changed from those at current rated conditions. Evaluations were performed at conditions that bound the change in operating conditions. The evaluation type is mainly reconciliation of the stresses and usage factors to reflect EPU conditions. A primary plus secondary stress analysis was performed showing EPU stresses still meet the requirements of the ASME Code, Section III. Lastly, the fatigue usage was evaluated for the limiting location of components with a (( )) The Browns Ferry fatigue analysis results for the limiting components are provided in Table 2.2-6. The plant-specific evaluations were performed with environmental fatigue using NUREG/CR-6909 (Reference 47) for the fatigue life correction factor to the ASME fatigue analyses. This was done to support Browns Ferry license renewal (Reference 11). Emergency and Faulted Conditions The stresses due to emergency and faulted conditions are based on loads such as peak dome pressure, which are unchanged for EPU. Therefore, Code requirements are met for all RPV components under emergency and faulted conditions. Therefore, reactor vessel meets all EPU dispositions. Conclusion TVA has evaluated the structural integrity of pressure-retaining components and their supports and has addressed the effects of the proposed EPU on these components and supports. The evaluation indicates that pressure-retaining components and their supports will continue to meet the requirements of 10 CFR 50.55a, draft GDCs-1, 2, 9, 33, 34, 40, and 42 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the structural integrity of the pressure-retaining components and their supports. 2-75

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.2.3 Reactor Pressure Vessel Internals and Core Supports Regulatory Evaluation Reactor pressure vessel internals consist of all the structural and mechanical elements inside the reactor vessel, including core support structures. The NRCs acceptance criteria are based on (1) 10 CFR 50.55a and GDC-1, insofar as they require that SSCs important to safety be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-2, insofar as it requires that SSCs important to safety be designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions; (3) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; and (4) GDC-10, insofar as it requires that the reactor core be designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. Specific NRC review criteria are contained in SRP Sections 3.9.1, 3.9.2, 3.9.3, and 3.9.5 and other guidance provided in Matrix 2 of RS-001. Browns Ferry Current Licensing Basis: The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-1, 2, 40 and 42. Final GDC-10 is applicable to Browns Ferry as 2-76

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) described in Browns Ferry Nuclear Plant, Unit 1 - Issuance of Amendments Regarding the Transition to Areva Fuel, dated July 3, 2012. (Reference 48) The Reactor Pressure Vessel Internals and Core Supports are described in Browns Ferry UFSAR Sections 3.3, Reactor Vessel Internals Mechanical Design, and 4.2, Reactor Vessel and Appurtenances Mechanical Design. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The reactor internals and core support structural components evaluation for license renewal are discussed in NUREG-1843, Section 3.1. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Sections 3.3 and 3.4 of the CLTR address the effect of EPU on Reactor Vessel and Reactor Internals, respectively. The results of this evaluation are described below. 2.2.3.1 FIV Influence on Reactor Internal Components The FIV evaluation of the RPV internals addresses the influence of an increase in flow during EPU. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Structural Evaluation of Core Flow Dependent RPV Meets CLTR Generic Internals Disposition Meets CLTR Structural Evaluation of Other RPV Internals Plant Specific Disposition 2.2.3.1.1 Structural Evaluation of Core Flow Dependent RPV Internals As stated in Section 3.4.2 of the CLTR, EPU causes an increase in reactor coolant quality and an increase in FW, steam, and recirculation pump drive flow. (( )) The core flow dependent RPV internal components (in-core guide tube and control rod guide tube) are confirmed to be consistent with the generic dispositions provided in the CLTR ((

))

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.2.3.1.2 Structural Evaluation of Other RPV Internals As stated in Section 3.4.2 of the CLTR, EPU causes an increase in reactor coolant quality and an increase in FW, steam, and recirculation pump drive flow. The required RPV internals vibration assessment of the other RPV internals is described in the CLTR. EPU operation increases the steam production in the core, resulting in an increase in the core pressure drop. ((

                                                                      )) The increase in power may increase the vibration level of reactor internals. Analyses were performed to evaluate the effects of FIV on the reactor internals at EPU conditions. This evaluation used a bounding reactor power of 102% of 3,952 MWt and 105% of rated core flow. ((
                                              )) For components requiring an evaluation but not instrumented in Browns Ferry Unit 1, ((
         )) The expected vibration levels for EPU were estimated by extrapolating the measured vibration data in Browns Ferry or similar plants and based on GEH BWR operating experience.

These expected vibration levels were then compared with the established vibration acceptance criteria. The following components were evaluated: a) Control Rod Guide Tube (CRGT) b) In Core Guide Tubes (ICGT) c) Feedwater Spargers d) Jet Pumps (JP) e) Jet Pump Sensing Lines (JPSL) f) Shroud g) Shroud Head and Separator Assembly h) Core Spray Piping Line and Sparger i) Fuel Assembly j) Guide Rod k) Shroud Head Bolts l) Top Guide m) Head Spare Instrument Nozzle n) Top Head Instrument Nozzle o) Top Head Vent Nozzle 2-78

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) p) Steam Line Nozzle q) Water Level Instrument Nozzle The results of the vibration evaluation show that continuous operation at a reactor power of 102% of 3,952 MWt and 105% of rated core flow does not result in any detrimental effects on the critical or safety-related reactor internal components shown above. Flow induced vibration of critical reactor internal components at EPU is predicted based on the available startup test data at ((

                                                                     )) Vibration amplitudes are also adjusted by a ((
                                    )) The extrapolated vibration amplitude response under EPU conditions is compared with the acceptance criterion in the percent criteria for each mode. The percentages of the criteria for all modes are cumulative as total percent criteria. ((
                        )) The summary of the evaluation methods and results for the following components are:

Control Rod Guide Tubes (CRGT) and In-Core Guide Tubes (ICGT) The vibration of the CRGT and ICGT in the lower plenum is a function of core flow. Because the maximum core flow under EPU remains unchanged from the OLTP or CLTP, the flow-induced vibrations of these components are not affected under EPU. Hence, there will be no increase in FIV stresses due to EPU. Maximum stresses during OLTP are well within the acceptance criteria and will remain about the same at EPU conditions. FW Sparger The FW sparger in Browns Ferry is of the improved triple thermal sleeve design. ((

                            )) Therefore, the Browns Ferry FW sparger is acceptable under EPU conditions.

Jet Pumps Results from strain gage measurements ((

))

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Jet Pump Sensing Lines Resonance of the recirculation pump Vane Passing Frequency (VPF) with the natural frequency of the JPSL is the cause of the JPSL stress. ((

                                       )) The jet pump sensing lines remain acceptable for EPU conditions.

Shroud For the shroud, the measured vibrations were extrapolated to the EPU conditions. Maximum stresses are less than 2,400 psi at OLTP and will remain well within acceptance criteria at EPU conditions. The calculated maximum stress is about 38% of the acceptance criteria or 3,800 psi at EPU conditions. Shroud Head and Separator Assembly For the shroud head and separator, ((

))

Core Spray Piping and Sparger ((

                                                                              )) Therefore, the FIV stress on core spray piping due to vortex shedding at EPU conditions is minimal.

During EPU, the components in the core region and components such as the core spray line are primarily affected by the core flow. Components in the annulus region such as the jet pump are primarily affected by the recirculation pump drive flow and core flow. For EPU conditions at Browns Ferry, there is no change in the maximum licensed core flow as compared to the CLTP condition, resulting in negligible changes in FIV on the components in the annular and core regions. The core spray sparger is subjected to a very low flow velocity with no resonance. This indicates that FIV adequacy is assured during EPU operation. Fuel Assembly The EPU FIV effect on fuel assembly is contained in Attachments 24 and 26 (proprietary) and 25 and 27 (non-proprietary) of the EPU LAR. 2-80

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Guide Rods The guide rod is subjected to cross flow, and the procedure and criteria as established in ASME Code Section III N-1300 (Reference 43) is used. ((

                                                    )) Therefore, the guide rod at Browns Ferry is acceptable under FIV for EPU conditions.

Shroud Head Bolts The shroud head bolt is subjected to cross flow, and the procedure and criteria as established in ASME Code, Section III, N-1300 (Reference 43) is used. ((

))

Therefore, the shroud head bolt at Browns Ferry is acceptable under FIV for EPU conditions. Top Guide The core flow does not change under EPU, and the flow velocity around the top guide is lower than (( )) Therefore, the increase in FIV loads due to EPU conditions is minimal and will not adversely affect the structural adequacy of the top guide. RPV Head Spare Instrument Nozzle ((

                                                                                      )) Thus, the stress due to FIV at EPU conditions is negligible.

RPV Top Head Instrument Nozzle ((

))

Thus, the stress in the nozzle due to FIV at EPU conditions is negligible. RPV Top Head Vent Nozzle ((

             )) Therefore, the top head vent nozzle will be structurally adequate from a vibration viewpoint at EPU conditions.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Steam Line Nozzle ((

                               )) Thus, the stress in the nozzle due to FIV at EPU conditions is negligible.

Water Level Instrument Nozzle The water level instrumentation nozzle connected to the steam leg is in the stagnant steam region, and is not affected by EPU, and the instrumentation nozzle connected to the reference leg of the narrow range reactor water level experiences similar changes as the lower part of the shroud guide rods or core spray pipe. The water level instrument nozzle with high fundamental natural frequency under low flow velocities assures that the FIV effects would be negligibly small. The calculations for EPU conditions indicate that vibrations of all safety-related reactor internal components are within the GEH acceptance criteria. The analysis is conservative for the following reason: The GEH criterion of 10,000 psi peak stress intensity is more conservative than the ASME allowable peak stress intensity of 13,600 psi for service cycles 1011. Conservatively, the peak responses of the applicable modes are absolute summed. The maximum vibration amplitude in each mode is used in the absolute sum process, whereas in reality the peak vibration amplitudes are unlikely to occur at the same time. Therefore, it is concluded that the flow-induced vibrations for all evaluated components remain within the acceptance limits, and the FIV effects on reactor internal components meets all CLTR dispositions. Steam Dryer During EPU, the components in the upper zone of the reactor, such as the steam dryer, are mostly affected by the increased steam flow. As a result, the steam dryer can be significantly affected by EPU conditions. The steam dryer is a non-safety-related components. Recent uprate experience indicates that FIV at EPU conditions may lead to high cycle fatigue failure of some dryer components. Failure of a dryer component does not represent a safety concern, but can result in a large economic effect. Quantitative analyses of the Browns Ferry steam dryers have been performed. The results showed that modifications to enhance structural integrity of the steam dryers would be needed for EPU conditions. Rather than modify the existing dryers, TVA has made a decision to replace the steam dryers. Attachment 40 of the Browns Ferry EPU license amendment request provides the analyses of the replacement steam dryers. 2-82

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.2.3.2 Reactor Internals The RPV internals consist of the core support structure components and non-core support structure components. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Reactor Internal Pressure Differences Plant Specific Disposition Meets CLTR Reactor Internals Structural Evaluation Plant Specific Disposition Meets CLTR Steam Dryer Separator Performance Plant Specific Disposition 2.2.3.2.1 Reactor Internal Pressure Differences As stated in Section 3.3.1 of the CLTR, EPU results in higher pressure differences across the RPV internals due to higher core exit steam flow. The increase in core average power alone would result in higher core loads and Reactor Internals Pressure Differences (RIPDs) due to the higher core exit steam quality. The RIPDs are calculated for Normal (steady-state operation), Upset, and Faulted conditions for all major reactor internal components. For minor components (jet pump sensing lines, dryer/separator guide rods, and in-core guide tube braces), the pressure drops during Normal, Upset, and Faulted conditions are minimal and represent insignificant portions of the RIPDs because of the small surface area. They are not affected by EPU and are not evaluated for EPU. Tables 2.2-7 through 2.2-9 compare the RIPDs across the major reactor internal components during current and EPU operation in the Normal, Upset, and Faulted conditions, respectively. The EPU reactor internal pressure difference (RIPD) calculations that are sensitive to fuel type are performed with a full core of GE13 fuel that includes the debris filter lower tie plate option. The RIPDs for GE13 fuel were demonstrated to be bounding for GE14 fuel as part of the GE14 new fuel introduction program. This is due to the higher flow resistance and resultant higher pressure drop of the GE13 fuel bundle. The RIPDs for GE13 fuel were also found to be bounding for both ATRIUM-10 and ATRIUM 10XM fuels except for the channel wall differential pressure (DP). The fuel channel RIPDs are discussed in Section 2.2.3.2.2. The EPU RIPDs are therefore applicable to GE13, GE14, ATRIUM-10, and ATRIUM 10XM fuel types. The acoustic and flow-induced loads following a postulated recirculation line break were also evaluated using TRACG models (see Table 1-1). The methodology for determining the Browns Ferry acoustic and flow-induced loads at EPU rated thermal power is unchanged from that used 2-83

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) for current rated thermal power and is unaffected by the issue identified in GEH Safety Communication 12-20 (Reference 50). The acoustic and flow-induced loads associated with the extension of the MELLLA and ICF domain to include EPU operation are bounded by the acoustic and flow-induced loads associated with reduced feedwater temperature operation at the minimum pump speed point on the MELLLA line (Point C of Figure 1-1). 2.2.3.2.2 Reactor Internals Structural Evaluation (Non-FIV) As stated in Section 3.3.2 of the CLTR, the typical loads considered in the EPU structural evaluation of the internals include: dead weight, RIPDs, seismic loads, thermal loads, flow loads, and acoustic and flow-induced loads due to a recirculation line break, consistent with the design basis. ((

))

The RPV internals consist of the core support structure components and non-core support structure components. The RPV internals are not ASME Code components. However, the requirements of the ASME Code are used as guidelines in their design/analysis. The evaluations/stress reconciliation in support of EPU was performed consistent with the design basis analysis of the components. The reactor internal components evaluated are: Core Support Components Shroud Shroud Support Core Plate Top Guide Control Rod Drive Housing (CRDH) Control Rod Guide Tube (CRGT) Orificed Fuel Support (OFS) Fuel Channel Non-Core Support Components FW Sparger Jet Pumps Core Spray Line and Sparger Access Hole Cover (AHC) Shroud Head and Steam Separator Assembly In-Core Housing and Guide Tube (ICHGT) Vessel Head Cooling Spray Nozzle 2-84

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Jet Pump Instrument Penetration Seal Differential Pressure and Standby Liquid Control Line The original configurations of the internal components are considered in the EPU evaluation unless a component has undergone permanent structural modifications, in which case, the modified configuration is used as the basis for the evaluation. The effects of the thermal-hydraulic changes due to EPU on the reactor internals were evaluated. All applicable loads and load combinations were considered consistent with the existing design basis analysis. These loads include the RIPDs (Section 2.2.3.2.1), dead weight, seismic loads, acoustic and flow induced loads, scram and thermal loads. EPU loads are compared to those used in the existing design basis analysis. If the EPU loads are bounded by the design basis loads for the RPV internals, the existing design basis qualification is valid for EPU. In such cases, no further evaluations are required or performed. For RPV internals exhibiting increases in loads, the method of analysis is to linearly scale the critical/governing stresses based on increases in loads as applicable (or the incremental stresses are calculated), and compare the resulting stresses against the allowable stress limits, consistent with the design basis. Conservative assessment is the initial approach; however, if required, excessive conservatism is removed from the existing assessment and/or the design basis analysis, as appropriate, and if justifiable. Table 2.2-10 presents the governing stresses for the various reactor internal components as affected by EPU. All stresses are within the design basis ASME Code allowable limits, and the RPV internal components are demonstrated to be structurally qualified for operation at EPU conditions. The following reactor vessel internals are evaluated for the effects of changes in loads due to EPU: a) Shroud: The only shroud load affected by EPU is RIPD. Seismic and flow induced loads remain unaffected by EPU. Acoustic loads are bounded by the design basis values. The RIPDs show increases with respect to the Normal/Upset condition; whereas, for the Faulted condition, only the shroud head RIPD increases. Because buckling is the limiting stress condition for the shroud, the tensile (positive) stresses produced by pressure are conservatively neglected in the shroud buckling analysis, consistent with the current design basis analysis. Therefore, for EPU, there is no change to the previous results, and the stresses remain unchanged and within Code allowable values. Therefore, the shroud is structurally qualified for EPU. b) Shroud Support: The only shroud support load affected by EPU is RIPD. Seismic loads and flow induced loads remain unchanged by EPU. Acoustic loads are bounded by the design basis values. The RIPDs show increases for the Normal/Upset service conditions; for the Faulted condition, only the shroud head RIPD increases. The effect of the change in EPU Normal/Upset condition RIPDs was assessed with respect to the design basis analysis 2-85

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) and the resultant stress was found to be within the allowable value. Therefore, the shroud support remains structurally qualified for EPU. c) Core Plate: The only core plate load affected by EPU is RIPD. All other applicable loads (including dead weight and seismic) remain unaffected. Only the Normal/Upset condition RIPDs increase for EPU. The existing basis beam buckling loads and sliding were reconciled based on the changes in the pressure loads. The reconciled beam buckling and sliding results remain within allowable limits. Normal condition RIPD is bounded by the core plate plug design basis value. The minimum core plate plug design basis 14 EFPY design life based on IGSCC is still bounding for EPU conditions. Therefore, the core plate (including core plate plugs) remains structurally qualified for EPU. d) Top Guide: The only top guide load affected by EPU is RIPD. All other applicable loads remain unaffected. The top guide EPU RIPDs are less than the previously qualified CLTP values for all service conditions. Therefore, the top guide remains unaffected by and structurally qualified for EPU. e) Control Rod Drive Housing (CRDH): The CRDH (internal to the vessel) is subjected to the following primary loads: weight (guide tube + fuel), pressure, scram loads, seismic and the flow loads in the lower plenum. As a result of the EPU, the CRDH design pressure loads (vessel pressure), design scram loads, seismic loads and the flow in the lower plenum remain unaffected. The temperature change in the lower plenum is insignificant based on the results of the recirculation system analysis. Thus, there is no significant effect on the thermal stress conditions of the CRDH. Therefore, the CRDH remains unaffected and structurally qualified for EPU. f) Control Rod Guide Tube (CRGT): Only the EPU Normal/Upset RIPDs increase relative to the previously qualified value. All other loads remain unaffected by EPU. Adequate CRGT lift margin exists in the EPU condition. The temperature and flow remain unchanged for EPU. The contribution of the increased RIPDs is small and the resulting stress for EPU is well within the stress allowable values. Therefore, the CRGT remains structurally qualified for EPU. g) Orificed Fuel Support (OFS): The only OFS load affected by EPU is RIPD, which increases for the Normal/Upset conditions only. All other loads (dead weight and seismic) remain unaffected. The contribution of the increased RIPDs is small, and the resulting stress for EPU conditions is well within the stress allowable values. Therefore, the OFS remains structurally qualified for EPU. h) Fuel Channel: The fuel channel RIPDs are within the design limits of the fuel for all service conditions. See LAR Attachments 24 and 26 for the fuel channel evaluations. i) Feedwater Sparger: The only change as a result of EPU is the change in the feedwater flow and the temperature. All other applicable loads remain unaffected. Flow related loading is a minimal contributor to the primary stress in the feedwater sparger. The effect of increase in feedwater temperature due to EPU is bounded by the CLTP evaluation. The 2-86

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) change in the maximum flow, as documented in the reactor heat balance (Table 1-2), has an insignificant effect on the primary stress integrity of the component. Therefore, the feedwater sparger remains structurally qualified for EPU. j) Jet Pumps: The predominant loads for jet pumps are: seismic loads, hydraulic flow loads, acoustic and flow induced loads. The change in the jet pump drive flow due to EPU is insignificant. The change in the flow temperature due to EPU is insignificant. Acoustic loads are bounded by the design basis values. The load conditions pertaining to the jet pump riser brace repair (Unit 3) remain unaffected by EPU. The existing repair design basis remains valid. The repair inspection interval is every other refueling outage, and the next scheduled inspection is during the Unit 3 Refueling Outage 18 (U3R18) in the Spring of 2018. Therefore, the jet pumps remain structurally qualified for EPU. k) Core Spray Line and Sparger: The core spray system flow load and pressure remain unaffected. Because vessel pressure is unchanged due to EPU, the Faulted condition annulus downcomer load also remains unaffected. Seismic loads are unaffected by EPU. The thermal condition (< 2°F temperature difference) in the annulus remains practically unchanged. Therefore, the core spray line and the sparger remain structurally qualified for the EPU condition. Unrelated to EPU, the Unit 3 core spray line T-box and downcomer have been modified. However, because the applicable loads for the core spray system remain unaffected by EPU, the Unit 3 core spray line T-box and downcomer remain qualified in the as modified condition. l) Access Hole Cover (AHC): The AHC experiences the same pressures as the shroud support plate and these RIPDs increase for Normal/Upset EPU conditions only. There is no significant change to the temperature and seismic loads due to EPU. The AHC location specific acoustic load is considered in the assessment. The effect of the EPU RIPDs and acoustic load is evaluated and reconciled with respect to the AHC design basis analysis. The design basis analysis remains valid for EPU conditions. Therefore, the AHC remains structurally qualified for EPU. m) Shroud Head and Steam Separator Assembly: The only shroud head load affected by EPU is RIPD, which increases for all (Normal, Upset, and Faulted) conditions. Seismic and other dynamic loads are not affected by EPU. While the limiting factor for the assembly is the Shroud Head Bolt (SHB), the stress on the SHBs due to the increased shroud head RIPDs remains within the allowable for EPU. Thus, the shroud head and steam separator assembly remains qualified for EPU. n) In-Core Housing and Guide Tube (ICHGT): There is no change in the dead weight and seismic loads due to EPU. The temperature (< 2°F) and flow in the lower plenum remain essentially unchanged for EPU. The existing design basis remains acceptable. Thus, the ICHGT remains structurally qualified for EPU. o) Vessel Head Cooling Spray Nozzle: The vessel head cooling spray nozzle is subject to dome pressure, seismic, and temperature effects. For EPU, there is no change in the nominal 2-87

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) dome pressure or temperature; seismic load also remains unchanged. Thus, the vessel head cooling spray nozzle remains structurally qualified for EPU. The vessel head cooling spray nozzle was capped and is no longer operational. The structural qualification of the vessel head cooling spray nozzle remains valid because there is no change to the loads when the nozzle is not in use. p) Jet Pump Instrument Penetration Seal: The jet pump instrument penetration seal is not affected by EPU conditions because the vessel pressure and temperature remains essentially unchanged. EPU has no effect on seismic loading, which remains unchanged. Therefore, the jet pump instrument penetration seal remains structurally qualified for EPU conditions. q) Differential Pressure and Standby Liquid Control Line: The core flow and the temperature are essentially unchanged for EPU conditions. Also, EPU has no effect on the existing seismic response of the differential pressure and standby liquid control line system. Therefore, the differential pressure and standby liquid control line remains structurally qualified for EPU. 2.2.3.2.3 Steam Dryer/Separator Performance For Browns Ferry, the EPU performance of the steam dryer/separator was evaluated to ensure that the quality of steam leaving the reactor pressure vessel continues to meet existing operational criteria at EPU conditions. EPU results in an increase in saturated steam generated in the reactor core. For constant core flow, this in turn results in an increase in the separator inlet quality and dryer face velocity and a decrease in the water level inside the dryer skirt. These factors, in addition to the core radial power distribution, affect the steam dryer/separator performance. The results of the evaluation demonstrated that Browns Ferry, using a representative equilibrium core design, has acceptable steam dryer/separator performance (i.e., moisture carryover 0.1 wt. %) at EPU conditions. Moisture carryover measurements are to be performed as part of the power ascension test plan as described in LAR Attachment 46. Conclusion TVA has evaluated the effects of the proposed EPU on the reactor internals and core supports. The evaluation indicates that the reactor internals and core supports will continue to meet the requirements of 10 CFR 50.55a, final GDC-10 and draft GDCs-1, 2, 40 and 42 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the design of the reactor internal and core supports. 2.2.4 Safety-Related Valves and Pumps Regulatory Evaluation The NRCs acceptance criteria are based on (1) GDC-1, insofar as it requires that SSCs important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-37, GDC-40, GDC-43, and GDC-46, insofar as they require that the ECCS, the containment heat 2-88

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) removal system, the containment atmospheric cleanup systems, and the cooling water system, respectively, be designed to permit appropriate periodic testing to ensure the leak-tight integrity and performance of their active components; (3) GDC-54, insofar as it requires that piping systems penetrating containment be designed with the capability to periodically test the operability of the isolation valves to determine if valve leakage is within acceptable limits; and (4) 10 CFR 50.55a(f), insofar as it requires that pumps and valves subject to that section must meet the inservice testing program requirements identified in that section. Specific NRC review criteria are contained in SRP Sections 3.9.3 and 3.9.6 and other guidance provided in Matrix 2 of RS-001. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-1, 38, 46, 47, 48, 57, 59, 60, 61, 63, 64, and 65. There is no draft GDC directly associated with final GDC-46. The inservice testing of safety-related valves and pumps is described in Browns Ferry UFSAR Section 6.6, Inspection and Testing. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation 2-89

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Report (SER), NUREG-1843, dated April 2006 (Reference 11). The safety-related valves and pumps are addressed within NUREG-1843 under the systems that contain them. Technical Evaluation 2.2.4.1 Background In-Service Testing of Safety-Related Pumps and Valves The In-service Testing (IST) of safety-related pumps and valves is addressed and documented in the Browns Ferry In-service Testing Program contained in Attachment 3, Part C In-service Testing of Pumps and Valves of TVA procedure NPG-SPP-09.1; ASME Code and Augmented Programs. The Browns Ferry pump and valve in-service testing program, hereafter referred to as the IST program, meets the requirements of 10 CFR 50.55a(f). The Browns Ferry Technical Specifications, Section 5.5.6, IST Program, states that this program provides controls for in-service testing of ASME Class 1, 2, and 3 pumps and valves and that the program shall include testing frequencies as specified in ASME OM Code, 2004 Edition through 2006 Addenda. Containment Leakage Rate Testing Program Containment leakage rate testing is addressed in UFSAR Section 5.2.5, and Browns Ferry TS Section 5.5.12. The Browns Ferry Containment Leakage Rate Testing Program implements testing requirements in accordance with 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, and guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak Test Program, dated September 1995 (Reference 51). Tests that measure containment and isolation valve leak rates (Type A, B and C tests) are performed using the Technical Specification value for Pa. The containment design pressure is 56 psig. From the containment analysis at EPU conditions, the peak containment pressure (Pa) is 49.1 psig. For Unit 2 and Unit 3 the containment peak pressure for EPU is lower than the CLTP Technical Specification Pa of 50.6 psig. For Unit 1 the Pa increases from the CLTP pressure of 48.5 psig to an EPU value of 49.1 psig. Therefore the leak rate testing requirements for containment and applicable isolation valves are affected by the proposed EPU. No components affected by the tests are transitioning from non-safety to safety as a result of EPU. Existing programmatic controls for the classification and maintenance of components and test values associated with Appendix J tests are sufficient to support EPU. Pumps in the IST Program The scope of the IST program is derived from the ASME OM Code, Subsection ISTB, In-service Testing of Pumps in Light-Water Reactor Nuclear Power Plants. ASME Code Class boundaries and component safety functions are not affected by EPU and no system parameter changes are being introduced that will require revisions to the programs supporting 10 CFR 50.55a(f) requirements. Therefore, existing programmatic controls for the classification and maintenance of testing requirements associated with safety-related pumps are consistent with 2-90

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) the existing IST program and are sufficient to support EPU. Table 2.2-11 lists the systems with pumps in the IST program. Valves in the IST Program The scope of the IST program is derived from the ASME OM Code, Subsection ISTC, In-service Testing of Valves in Light-Water Reactor Nuclear Power Plants. ASME Code Class boundaries and component safety functions are not affected by EPU and no system parameter changes are being introduced that will require revisions to the programs supporting 10 CFR 50.55a(f) requirements. Therefore, existing programmatic controls for the classification and maintenance of testing requirements associated with safety-related valves are consistent with the existing IST program and are sufficient to support EPU. Table 2.2-11 lists the systems with valves in the IST program. Motor Operated Valve Program The Browns Ferry Motor Operated Valve (MOV) Program implements the recommendations and requirements made in Generic Letter 89-10, Safety-Related Motor Operated Valve Testing and Surveillance (Reference 52). The scope of the program also includes the requirements of Generic Letter 96-05, Verification of Design-Basis Capability of Safety-Related Motor Operated Valves (Reference 53). Existing programmatic controls for periodic verification requirements associated with safety-related valves are consistent with the existing GL 89-10 and GL 96-05 programs and are sufficient to support EPU. The existing Browns Ferry calculations for GL 89-10 MOVs were reviewed and the review shows that the maximum ambient temperatures used in existing MOV calculations bound the maximum ambient temperatures for EPU with the exception of one Unit 3 motor operated valve, 3-FCV-75-53. (See Table 2.2-12). This temperature increase has no effect on the affected valve capability or margin. Other parameters such as valve differential pressure/line pressure, motor terminal voltage, pressure locking and thermal binding, and valve stroke time effect were evaluated and were found to be either unaffected by EPU or the EPU effect was bounded by the parameters used in the existing calculations of record. The peak containment pressure following a LOCA increases slightly due to EPU (less than 1.4 psig from the peak pressures used in the existing MOV calculations). MOVs that are required to operate during a LOCA were evaluated for the changes in peak containment pressure and were found to maintain positive thrust/torque margin against the thrust/torque required for the valves to perform their open or close function. No valves under the Browns Ferry GL 89-10 and GL 96-05 program require modification to support EPU implementation. Operation at EPU conditions does not affect the capability of the GL 89-10 MOVs to perform their design basis functions. Table 2.2-11 indicates the systems that contain GL 89-10 MOVs. Air-Operated Valve Program The TVA Air Operated Valve program (NETP-114) was evaluated for compliance with the Joint Owners Group (JOG) air operated valve testing requirements and will continue to provide assurance that AOVs will be appropriately monitored and maintained during plant operations 2-91

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) under EPU conditions. Currently, no Browns Ferry air operated valves are classified as Category 1. Category 2 and 3 valves do not require design verification. No AOVs change to Category 1 as a result of EPU. Generic Letter 95-07 GL 95-07, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," August 17, 1995 (Reference 54) addresses the phenomena of pressure locking and thermal binding of safety-related power-operated gate valves. Pressure locking and thermal binding had been previously evaluated for all Browns Ferry safety-related gate valves. There were no air-operated or hydraulic-operated valves that were susceptible to pressure locking or thermal binding. The evaluation identified a number of MOVs that were susceptible to these conditions, which resulted in valve disc modifications and/or valve replacements. A review of the station commitments and modifications related to GL 95-07 indicates that EPU will not cause additional safety-related gate valves, formerly excluded by screening criteria, to be susceptible to pressure locking or thermal binding, and EPU will not affect the susceptibility of valves already modified to prevent these problems. Therefore, EPU has no effect on the potential for pressure locking or thermal binding of safety-related power-operated gate valves. Table 2.2-11 indicates the systems that contain GL 95-07 valves. Lessons Learned The Browns Ferry IST program, Containment Leak Rate program, MOV program and air operated valve (AOV) program utilize the Browns Ferry Corrective Action Program to evaluate and resolve non-conforming conditions identified during program performance. The purpose of the Browns Ferry Corrective Action Program is to stimulate and manage continuous improvement of station and organizational performance through identification, evaluation, correction and prevention of reoccurrence of unwanted and/or unexpected conditions, deviations, events, or issues that have the potential for affecting the safe, reliable, and efficient operation of Browns Ferry. Included in the program is recognition of any lessons learned or improvement opportunities identified from an assessment of missed opportunities to avoid the event/condition/issue. NPG SPP-22.300, Revision 2, is the administrative procedure that implements the requirements of the corrective action program that complies with 10 CFR 50, Appendix B, Criterion 16. 2.2.4.2 Description of Analyses and Evaluations This section addresses the effect of EPU on the performance requirements of Browns Ferry safety-related components in the IST, Motor Operated Valve, and Air Operated Valve programs. For the majority of the valves analyzed for EPU effects, there are minor effects on normal operating and DBA ambient temperatures. However, because EPU does not result in a significant change to the temperature assumptions used in the MOV calculations, the operation of the affected valves is not affected. 2-92

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Because the drywell and wetwell normal and accident pressures are not significantly changed by EPU, the associated containment valves were not affected. The calculations of design basis parameters for MOVs within the scope of GL 89-10 and GL 96-05 were reviewed to determine the effect of EPU on the valves. In most cases, the values of existing parameters bound the values expected under EPU conditions. In a few cases, there are slight increases above the current value. In those cases, the effect of the slight increase on the MOV has been evaluated in accordance with the requirements of the station MOV program and found to be within the design parameters. Systems Not Significantly Affected by EPU: The following systems contain pumps and/or power operated valves but are not significantly affected by EPU: Service Air, Instrument Air, Pneumatic Nitrogen, Hydrogen Supply, Carbon Dioxide supply, Containment Systems, Floor Drains, Sanitary Drains, Radioactive Drains, Sewer, Torus Drain, Miscellaneous Drains, Amertap, Suppression Pool Cleanup, Fuel Oil systems, Lube Oil systems, Process Steam and Aux Boilers, Monitoring and Sampling systems, Laundry, Sewage Treatment, Showers, Water Supply Systems, and Water Quality Systems. With no changes or effects to these systems due to EPU, the Browns Ferry program valves in these systems are not affected by EPU. Nuclear Steam Supply Systems: The Nuclear Steam Supply System (NSSS) systems include Core Spray (Section 2.8.5.6.2), High Pressure Coolant Injection (Section 2.8.5.6.2), Reactor Core Isolation Cooling (Section 2.8.4.3), Residual Heat Removal (Section 2.8.4.4), Reactor Water Cleanup (Section 2.1.7) and Standby Liquid Control (Section 2.8.4.5). Evaluations show that EPU has no effect on system operating pressures, flow rates, and pump head performance for Core Spray, High Pressure Coolant Injection, Reactor Core Isolation Cooling and the Reactor Water Cleanup systems. Changes in the Standby Liquid Control system are addressed in Section 2.8.4.5. ASME Code Class boundaries and component safety functions within these systems are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Changes in containment response affect certain program valves of the NSSS systems. The affected valves are evaluated in Section 2.2.4.3. The individual NSSS systems are evaluated below. Reactor Water Cleanup System The Reactor Water Cleanup system is not changed as a result of EPU and is addressed in Section 2.1.7. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. 2-93

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Standby Liquid Control System The Standby Liquid Control system changes due to EPU are addressed in Section 2.8.4.5. For EPU, the maximum pump discharge pressure occurring during the limiting ATWS event is calculated at 1,201 psig. The program valves in the system were found to be acceptable for EPU. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. System discharge increases slightly for EPU but due to the fact that IST test pressures do not require changes, the IST program is unaffected. Reactor Core Isolation Cooling No changes are being made to the RCIC system as a result of EPU. This system is addressed in Section 2.8.4.3. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Residual Heat Removal System No Residual Heat Removal system changes are being considered due to EPU. This system is addressed in Section 2.8.4.4. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. High Pressure Coolant Injection System No changes are being made to the HPCI system as a result of EPU. This system is addressed in Section 2.8.5.6.2. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Core Spray System No changes are being made to the Core Spray system as a result of EPU. This system is addressed in Section 2.8.5.6.2. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Balance of Plant Systems: Changes in system flow rates for EPU affect certain components in the balance of plant systems. The affected components are evaluated in Section 2.2.4.3. 2-94

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Reactor Building Closed Loop Cooling System The Reactor Building Closed Loop Cooling system changes due to EPU are addressed in Section 2.5.3.3.1. No modifications are being made as a result of EPU. The Browns Ferry program valves are not affected by EPU. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Feedwater / Feedwater Pump Recirculation Systems The FW system changes due to EPU are addressed in Section 2.5.4.4. The temperature, flow, and operating pressure will increase; however, the current design conditions bound the conditions at EPU. The Browns Ferry MOVs affected by EPU are evaluated in Section 2.2.4.3. The ability of the IST program check valves to perform their safety functions is not affected by EPU. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Main Steam System The proposed EPU results in MS flow increase; however, the current component design bounds the conditions at EPU regarding pressure and temperature. The Main Steam Safety Relief Valves are addressed in Section 2.8.4.2. The MSRV setpoints remain the same. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Reactor Recirculation System The Reactor Recirculation system is addressed in Section 2.8.4.6. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Fuel Pool Cooling System The Fuel Pool Cooling system is addressed in Section 2.5.3.1. Fuel Pool Cooling requirements are unaffected by EPU and additional heat limits are administratively controlled. 2-95

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Control Rod Drive Hydraulic System The Control Rod Drive Hydraulic system is not changed due to EPU and the system is addressed in Section 2.8.4.1. Although there is a slight pressure increase for operation at EPU, it is within the design capability of the system components and no temperature changes occur for EPU. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Raw Cooling Water System The RCW system has a small increase in temperature over current operation for EPU but remains within design limits. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. Emergency Equipment Cooling Water System For the EECW system, the heat load increases are insignificant and flow demand, pump duty, and system pressure will not significantly change. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. RHR Service Water System The RHR shutdown cooling mode initiating pressure and temperature do not change with EPU. Therefore, there is no increase in the maximum Residual Heat Removal Service Water (RHRSW) system heat load when the RHR heat exchangers operate in the shutdown cooling mode during normal reactor shutdown. ASME Code Class boundaries and component safety functions within this system are not affected by EPU. Additionally, EPU does not introduce any system parameter changes that will require IST program revisions. 2.2.4.3 Individual Component Evaluations Valves Affected by Changes in Containment Response Certain motor operated valves in the program are affected by changes in the containment response, see Section 2.6.1.1. These valves were evaluated and one was found to be affected by EPU conditions. The valve and its effect is presented in Table 2.2-12. 2-96

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Conclusion TVA has evaluated the effects of the proposed EPU on safety-related valves. The evaluation addressed the effects of the proposed EPU on its MOV programs related to GL 89-10 and GL 95-07. The evaluation indicates that safety-related valves will continue to meet the requirements of 10 CFR 50.55a(f) and draft GDCs-1, 38, 46, 47, 48, 57, 59, 60, 61, 63, 64, and 65 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to safety-related valves. 2.2.5 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment Regulatory Evaluation Mechanical and electrical equipment covered by this section includes equipment associated with systems that are essential to emergency reactor shutdown, containment isolation, reactor core cooling, and containment and reactor heat removal. Equipment associated with systems essential to preventing significant releases of radioactive materials to the environment are also covered by this section. The NRCs acceptance criteria are based on (1) GDC-1, insofar as it requires that SSCs important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed; (2) GDC-30, insofar as it requires that components that are part of the RCPB be designed, fabricated, erected, and tested to the highest quality standards practical; (3) GDC-2, insofar as it requires that SSCs important to safety be designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions; (4) 10 CFR Part 100, Appendix A, which sets forth the principal seismic and geologic considerations for the evaluation of the suitability of plant design bases established in consideration of the seismic and geologic characteristics of the plant site; (5) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents; (6) GDC-14, insofar as it requires that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of rapidly propagating fracture; and (7) 10 CFR Part 50, Appendix B, which sets quality assurance requirements for safety-related equipment. Specific NRC review criteria are contained in SRP Section 3.10. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis 2-97

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-1, 2, 9, 33, 34, 40 and 42. The Seismic and Dynamic Qualification of Mechanical and Electrical Equipment is described in Browns Ferry UFSAR Section 12.2, Principal Structures and Foundations. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Sections 10.1 and 10.3 of the CLTR address the effect of EPU on the seismic and dynamic qualification of mechanical and electrical equipment. The EPU dynamic forces (pipe whip and jet impingement loads) are bounded by the current licensing basis pipe whip and jet impingement loads (see Section 2.2.2). The primary input motions due to the safe shutdown earthquake are not affected by EPU and therefore, there are no consequences to the existing seismic analyses. No quality standards related to the design, fabrication, erection, and testing of the RCPB or SSCs important to safety are relaxed or removed as a result of the EPU and no changes have been made to the plant design bases established in consideration of the seismic and geologic characteristics of the plant site. For protective (mechanical) devices located in an area designated as a harsh environment and which perform safety-related functions, the change in that environment resulting from a Design Basis Event (DBE) during EPU operation imposes no adverse effects on the performance of the mechanical equipment. The incremental increases in the environmental conditions due to radiation described for EPU operation do not result in reaching the environmental threshold levels where noticeable degradation may occur. The internal process and external environmental changes associated with EPU will have no detrimental effect on the mechanical equipment's ability to perform safety-related functions in accordance with the original design basis of the plant. 2-98

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Based on the above, mechanical equipment in Browns Ferry Units 1, 2 and 3, performing safety-related functions in a harsh environment do not experience detrimental effects when operating at 3,952 MWt core thermal power level. The incremental changes in the environmental conditions, mainly radiation, due to EPU operation do not affect the ability of the mechanical equipment to perform their intended safety functions. The increase in radiation levels experienced by equipment during normal operation and accident conditions is expected to be proportional to the increase in power level. There is only a very small effect on pressure and temperature conditions due to the constant pressure assumption. Operation at EPU conditions increases the temperatures in areas near the FW lines due to the increased feedwater system operating temperature. However, the increases are expected to have little or no effect on the mechanical equipment materials. The Browns Ferry design and licensing basis does not require a formal Mechanical Equipment Qualification (MEQ) program such as the EQ program for electrical equipment. Browns Ferry uses other existing programs to evaluate the qualification of mechanical components. The key elements are design control, procurement evaluations, testing/preventative maintenance and equipment monitoring. The design control program ensures that mechanical components are specified and procured for the environment in which they are intended to function. Periodic maintenance and testing are performed in accordance with plant and industry operating experience and vendor recommendations to ensure continued functionality. The mechanical design of equipment/components (e.g., pumps, heat exchangers) in certain systems is affected by operation at EPU due to slightly increased temperatures (< 10%), and in some cases, flows ( 15%). However, experience has shown that the uprated operating conditions do not significantly affect the cumulative usage fatigue factors of mechanical components. EPU effects on fluid induced loads due to postulated RRS pipe breaks (Section 2.6.1.2.1) inside containment remain bounded by the current analysis. The dynamic loading on the safety-related components (RHR containment spray headers and LPCI protection) are not affected by EPU conditions. The margin in the current analysis bounds the slight increase in operating pressure in the RRS due to EPU. The RCPB systems affected by EPU were evaluated within the piping assessments in Section 2.2.2.2.1. The piping systems affected by EPU remain bounded by the current piping analysis. Therefore, the piping and piping supports for the affected systems are adequately designed for EPU conditions. Conclusion TVA has evaluated the effects of the proposed EPU on the qualification of mechanical and electrical equipment and addressed the effects of the proposed EPU on this equipment. The evaluation indicates that the equipment will continue to meet the requirements of 10 CFR Part 100, Appendix A; 10 CFR Part 50, Appendix B; and draft GDCs-1, 2, 9, 33, 34, 40 and 42 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the qualification of the mechanical and electrical equipment. 2-99

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-1 High Energy Line Break Outside Containment: Liquid Line Breaks System or Break EPU Effect on CLTP Mass Flow Release Rate Location MS System, RCIC Unchanged System, HPCI System MS Line Intermediate The total mass release increased by approximately 11%. Break RWCU System Mass flow rate increased by approximately 4.4% for RWCU breaks. FW System (Double The total mass release increased by approximately 12.5%. Ended Break) 2-100

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-2 Reactor Coolant Pressure Boundary Structural Evaluation System Temperature Pressure Flow Rate Mechanical Loading CLTP EPU CLTP EPU CLTP EPU ((

))

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3a Main Steam Pipe Stresses Due to EPU Conditions Maximum Stress Summary: Unit 1 Line A CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U 150 Note 1 13,867 18,000 Service Level B Eq. 9E 150 Note 1 17,110 27,000 Service Level C Eq. 9E 150 Note 1 13,867 22,500 Service Level D Eq. 9E 150 Note 1 17,110 30,000 Sustained + Thermal Eq. 9U+10 5 Note 1 27,040 52,500 Maximum Stress Summary: Unit 1 Line B CLTP EPU EPU Service Level Equation Node Stress Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U S47 Note 1 16,113 18,000 Service Level B Eq. 9E S47 Note 1 25,478 27,000 Service Level C Eq. 9E S47 Note 1 16,113 22,500 Service Level D Eq. 9E S47 Note 1 25,478 30,000 Sustained + Thermal Eq. 9U+10 H2 Note 1 37,312 45,000 Maximum Stress Summary: Unit 1 Line C CLTP EPU EPU Service Level Equation Node Stress Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U J22 Note 1 15,123 18,000 Service Level B Eq. 9E J22 Note 1 25,263 27,000 Service Level C Eq. 9E J22 Note 1 15,123 22,500 Service Level D Eq. 9E J22 Note 1 25,263 30,000 Sustained + Thermal Eq. 9U+10 252 Note 1 29,237 52,500 2-102

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3a Main Steam Pipe Stresses Due to EPU Conditions (continued) Maximum Stress Summary: Unit 1 Line D CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U J01 Note 1 15,819 18,000 Service Level B Eq. 9E J01 Note 1 26,001 27,000 Service Level C Eq. 9E J01 Note 1 16,102 22,500 Service Level D Eq. 9E J01 Note 1 26,284 30,000 Sustained + Thermal Eq. 9U+10 32 Note 1 24,767 52,500 Maximum Stress Summary: Unit 2 Line A CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U 150 Note 1 17,926 21,000 Service Level B Eq. 9E 150 Note 1 20,921 31,500 Service Level C Eq. 9E 150 Note 1 19,919 26,250 Service Level D Eq. 9E 150 Note 1 22,914 35,000 Sustained + Thermal Eq. 9U+10 5 Note 1 33,656 52,500 Maximum Stress Summary: Unit 2 Line B CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U H16 Note 1 14,199 18,000 Service Level B Eq. 9E H16 Note 1 17,756 27,000 Service Level C Eq. 9E H16 Note 1 15,816 22,500 Service Level D Eq. 9E H16 Note 1 18,429 30,000 Sustained + Thermal Eq. 9U+10 H3 Note 1 29,397 45,000 2-103

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3a Main Steam Pipe Stresses Due to EPU Conditions (continued) Maximum Stress Summary: Unit 2 Line C CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U RC-1 Note 1 16,010 18,000 Service Level B Eq. 9E RC-1 Note 1 16,286 27,000 Service Level C Eq. 9E RC-1 Note 1 17,041 22,500 Service Level D Eq. 9E RC-1 Note 1 17,317 30,000 Sustained + Thermal Eq. 9U+10 252 Note 1 34,254 52,500 Maximum Stress Summary: Unit 2 Line D CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U 32 Note 1 18,910 21,000 Service Level B Eq. 9E 55 Note 1 23,219 31,500 Service Level C Eq. 9E 55 Note 1 19,941 26,250 Service Level D Eq. 9E 55 Note 1 24,250 35,000 Sustained + Thermal Eq. 9U+10 32 Note 1 30,971 52,500 Maximum Stress Summary: Unit 3 Line A CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U 150 Note 1 12,384 21,000 Service Level B Eq. 9E 150 Note 1 15,110 31,500 Service Level C Eq. 9E 150 Note 1 13,703 26,250 Service Level D Eq. 9E 150 Note 1 16,429 35,000 Sustained + Thermal Eq. 9U+10 39, 40 Note 1 28,977 52,500 2-104

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3a Main Steam Pipe Stresses Due to EPU Conditions (continued) Maximum Stress Summary: Unit 3 Line B CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U H16 Note 1 16,413 18,000 Service Level B Eq. 9E H16 Note 1 17,467 27,000 Service Level C Eq. 9E H16 Note 1 18,070 22,500 Service Level D Eq. 9E H16 Note 1 19,124 30,000 Sustained + Thermal Eq. 9U+10 CENTR Note 1 37,120 45,000 Maximum Stress Summary: Unit 3 Line C CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U RC-1 Note 1 16,010 18,000 Service Level B Eq. 9E RC-1 Note 1 16,286 27,000 Service Level C Eq. 9E RC-1 Note 1 18,558 22,500 Service Level D Eq. 9E RC-1 Note 1 18,835 30,000 Sustained + Thermal Eq. 9U+10 252 Note 1 34,254 52,500 Maximum Stress Summary: Unit 3 Line D CLTP EPU EPU Stress Service Level Equation Node Stress Allowable (psi) (psi) (psi) Service Level A Eq. 9U 55 Note 1 18,943 21,000 Service Level B Eq. 9E 55 Note 1 23,290 31,500 Service Level C Eq. 9E 55 Note 1 21,492 26,250 Service Level D Eq. 9E 55 Note 1 25,839 35,000 Sustained + Thermal Eq. 9U+10 36, 37 Note 1 37,284 52,500 Note:

1. The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 timeframe. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3b Feedwater Pipe Stresses Due to EPU Conditions Maximum Stress Summary: Unit 1 Line A B31.1 Description Node EPU Allowable (psi) Stress Ratio Equation Joint Stress (psi) 11 Primary + Secondary 55B 35,053 37,500 0.935 (Normal) (=SA+Sh) 9U+10 Primary + Secondary 55BD 42,346 45,000 0.941 (Upset) (=1.2*(SA+Sh)) Note: The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 timeframe. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions. Maximum Stress Summary: Unit 1 Line B B31.1 Description Node EPU Allowable (psi) Stress Ratio Equation Joint Stress (psi) 11 Primary + Secondary CENTR 33,226 37,500 0.886 (Normal) (=SA+Sh) 9U+10 Primary + Secondary 47 38,370 45,000 0.853 (Upset) (=1.2*(SA+Sh)) Note: The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 timeframe. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions. 2-106

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3b Feedwater Pipe Stresses Due to EPU Conditions (continued) Maximum Stress Summary: Unit 2 Line A B31.1 Description Node EPU Allowable (psi) Stress Ratio Equation Joint Stress (psi) 11 Primary + Secondary SB20 29,807 30,000 0.994 (Normal) (=SA+Sh) 9U+10 Primary + Secondary SB20 35,924 36,000 0.998 (Upset) (=1.2*(SA+Sh)) Note: The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 timeframe. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions. Maximum Stress Summary: Unit 2 Line B B31.1 Description Node EPU Allowable (psi) Stress Ratio Equation Joint Stress (psi) 11 Primary + Secondary 47 34,181 37,500 0.911 (Normal) (=SA+Sh) 9U+10 Primary + Secondary 47 39,324 45,000 0.874 (Upset) (=1.2*(SA+Sh)) Note: The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 timeframe. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions. 2-107

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3b Feedwater Pipe Stresses Due to EPU Conditions (continued) Maximum Stress Summary: Unit 3 Line A B31.1 Description Node EPU Allowable (psi) Stress Ratio Equation Joint Stress (psi) 11 Primary + Secondary 47, 48 37,463 37,500 0.999 (Normal) (=SA+Sh) 9U+10 Primary + Secondary 48 40,800 45,000 0.907 (Upset) (=1.2*(SA+Sh)) Note: The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 timeframe. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions. Maximum Stress Summary: Unit 3 Line B B31.1 Description Node EPU Allowable (psi) Stress Ratio Equation Joint Stress (psi) 11 Primary + Secondary CENTR 31,736 37,500 0.846 (Normal) (=SA+Sh) 9U+10 Primary + Secondary 47 36,186 45,000 0.804 (Upset) (=1.2*(SA+Sh)) Note: The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 timeframe. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions. 2-108

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-3c Feedwater Pipe Stresses Due to Feedwater Transient Feedwater Piping Inside Containment (Unit 1 Node 55B) Allowable 27,000 psi Max EPU Eqn. 9E Stress 25,751 psi Existing EPU Eqn. 9E Stress Ratio 0.954 Max EPU Eqn. 9E Stress with RFPT Load 26,563 psi EPU Eqn. 9E Stress Ratio with RFPT Load 0.984 Note: An evaluation was performed which evaluated the effects of a simultaneous three reactor feedwater pump turbine trip. This considered case is bounding and conservative. The evaluation was performed for a representative piping segment for FW piping inside containment and is applicable to Units 1, 2, and 3. The results are for the bounding node with the highest Eqn. 9U/9E/9F/9U+10 stress ratio which was Node 55B of Unit 1 for Eqn. 9E. The FW transient only affects those equation stresses. Table 2.2-3d Feedwater and Condensate Pipe Stresses Due to Feedwater Transient Feedwater Piping Condensate Piping Outside Containment (Node 200.1) (Node 65) Allowable 22,500 psi 22,500 psi Max EPU Stress (1) 17,133 psi 12,217 psi Existing EPU Max Stress Ratio (1) 0.761 0.543 Max EPU Stress with RFPT Load 17,811 psi 12,274 psi EPU Stress Ratio with RFPT Load 0.792 0.545 Note:

1. Stress corresponding to the maximum stress ratio of all Code equations. For feedwater piping outside containment and condensate piping, the maximum EPU stress ratio corresponds to Eqn. 10. Although Eqn. 10 is for thermal loadings, increasing the maximum stress ratio and confirming it is less than 1.0 is conservative.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-4a Main Steam System Piping (Outside Containment) Maximum Stress Interactions for Main Steam Piping Outside Containment CLTP EPU Allowable Interaction Service Level Stress Stress (2) Node (psi) Ratio (psi) (psi) Eqn. 8 Sustained Note 1 6,631 D90 15,000 0.442 Eqn. 9U Occasional (Upset) Note 1 17,563 L30A 18,000 0.976 Eqn. 10 Thermal Expansion Note 1 16,688 B15 22,500 0.742 Sustained + Thermal Eqn. 11 Note 1 21,821 B15 37,500 0.582 Expansion Note:

1. The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 time frame. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required. Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions.

2. EPU stress from MS piping analysis; based on all four loops between the containment penetration anchor and the HP turbine inlet nozzles. Applies to Units 1, 2, and 3.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-4b Feedwater Piping Maximum Stress for Feedwater Piping Outside Containment (1) CLTP EPU Allowable Interaction Criteria Per ANSI B31.1 Node Stress Stress (psi) Ratio (psi) (psi) Thermal Eqn. 10 65 Note 2 17,133 22,500 0.761 Expansion Notes:

1. Only the Equation 10 stresses increase due to the thermal increases associated with EPU.

Applies to Units 1, 2, and 3.

2. The original EPU stress calculations for the BOP and safety-related (RCPB) piping for Browns Ferry units were completed in the 2002-2003 time frame. Since completion, piping systems in both safety-related (RCPB) and BOP systems have been modified, using the EPU values. A significant effort to reconstitute CLTP stress values with the current analysis would be required.

Given that the existing calculations (with EPU values) have determined that no code of record stress allowables have been exceeded indicates that the piping is acceptable for EPU conditions. Maximum pipe stress increase from CLTP analysis**: Temperature Expansion 5.4%

Pressure 0%
Fluid Transients N/A
Maximum pipe support loading increase (due to 5.3%
thermal expansion loading)**:
The maximum increase in FW temperature range from CLTP to EPU is 5.4%. Pipe stresses remain within code allowables. Maximum increase in FW temperature for piping with rigid supports is 5.3%. There is no fluid transient loading in the current FW piping design basis.

 ** Bounding value of Units 1, 2, and 3.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-4c Condensate Piping Maximum pipe stress increase from CLTP analysis**: Temperature Expansion 4.7% Pressure 16.7%

Fluid Transients N/A
Maximum pipe support loading increase (due to 4.7%

thermal expansion loading)**:

Condensate piping design pressure bounds EPU operating pressures. Pipe stress remains within code allowables. There is no fluid transient loading in the current condensate piping design basis.
- Bounding value of Units 1, 2, and 3.

Table 2.2-4d Extraction Steam Piping Maximum pipe stress increase from CLTP analysis**: Temperature Expansion 6.5% Pressure 13.3%

Fluid Transients N/A
Maximum pipe support loading increase (due to 4.6%
thermal expansion loading)**:
Extraction steam piping design pressure remains bounding for EPU. The maximum increase in support loading is for piping with rigid supports. There is no fluid transient loading in the current extraction steam piping design basis.
- Bounding value of Units 1, 2, and 3.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-4e FW Heater Drains and Vents Piping Maximum pipe stress increase from CLTP analysis**: Temperature Expansion 9.4% Pressure 0%

Fluid Transients N/A
Maximum pipe support loading increase (due to 0%
thermal expansion loading)**:
FW heater drains and vents piping design pressure remains bounding for EPU.

Pipe stresses remain within code allowables. There is no fluid transient loading in the current FW heater drains and vents piping design basis. There are no rigid supports for this piping, but increased movement on spring supports is 0.1 or less.

- Bounding value of Units 1, 2, and 3 Table 2.2-4f Moisture Separator Vents and Drains Piping Maximum pipe stress increase from CLTP analysis**:

Temperature Expansion 4.24% Pressure 0%

Fluid Transients N/A
Maximum pipe support loading increase (due to 4.24%

thermal expansion loading)**:

Moisture separator vents and drains piping design pressure remains bounding for EPU. There is no fluid transient loading in the current moisture separator vents and drains piping design basis.
- Bounding value of Units 1, 2, and 3.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-5 BOP Piping System Evaluation Temperature (°F) Pressure (psig) Flowrate (Mlb/hr) System Mechanical Loading CLTP EPU CLTP EPU CLTP EPU Condensate Piping (3rd stage heater to RFP) (( FW Piping (from RFP to RPV) MS Piping (max conditions at RPV) Extraction Steam Piping (High Pressure (HP) turbine exhaust to 1st stage heater) FW Heater Drains (1st stage heater to 2nd stage heater)

))

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-6 CUFs and Sp+q Values of Limiting Components P + Q Stress[1] (ksi) CUF EPU with Allowable CLTP CLTP Environmental CLTP EPU EPU 40 / 60 years Component (ASME 40 years 60 years Fatigue Uen Allowable (3,527 MWt)[2] (4,031 MWt) (4,031 MWt) Code Limit) (3,527 MWt) [2] (3,527 MWt) 60 years (4,031 MWt) Feedwater Nozzle Unit 1: 37.3 57.7 69.9 0.984[3] 0.458/0.248[6] 1.0 Unit 2: 37.3 37.7 69.9 0.984 [3] Note 4 Note 5 0.444/0.240 [6] 1.0 Unit 3: 37.3 37.7 69.9 0.984 [3] 0.450/0.243 [6] 1.0 Recirculation Inlet Nozzle Unit 1: 77.1/47.0 [7] 88.73/54.10 [7,17] 58.40 0.425[8] 0.212/0.981[9] 1.0 [7] [7,17] [8] [9] Unit 2: 77.1/47.0 88.73/54.10 58.40 0.425 Note 4 Note 5 0.282/0.981 1.0 [7] [7,17] [8] [9] Unit 3: 77.1/47.0 88.73/54.10 58.40 0.425 0.279/0.981 1.0 2-115

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-6 CUFs and Sp+q Values of Limiting Components (continued) P + Q Stress[1] (ksi) CUF EPU with Allowable CLTP Environmental CLTP EPU CLTP EPU 40/ 60 years Component (ASME 60 years Fatigue Uen Allowable (3,527 MWt)[2] (4,031 MWt) (3,527 MWt) [2] (4,031 MWt) Code Limit) (3,527 MWt) 60 years (4,031 MWt) Recirculation Outlet Nozzle Unit 1: 73.20 75.53 80.10 0.779[10] 0.034/0.015[11] 1.0 Unit 2: 73.20 75.53 80.10 0.779[10] Note 4 Note 5 0.103/0.015[11] 1.0 Unit 3: 73.20 75.53 80.10 0.779[10] 0.102/0.015[11] 1.0 Core Spray Nozzle Unit 1: 42.50 44.37 52.46 0.073[10] 0.112/0.237[9] 1.0 Unit 2: 42.50 44.37 52.46 0.073[10] Note 4 Note 5 0.345/0.237 [9] 1.0 Unit 3: 42.50 44.37 52.46 0.073[10] 0.340/0.237 [9] 1.0 CRD Hydraulic System Return Nozzle Unit 1: 68.0 70.99 80.0 0.363[10] 0.394/0.287 [12] 1.0 [10] [12] Unit 2: 68.0 70.99 80.0 0.363 Note 4 0.363/0.287 NA 1.0 [10] [12] Unit 3: 68.0 70.99 80.0 0.363 0.363/0.287 1.0 2-116

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-6 CUFs and Sp+q Values of Limiting Components (continued) P + Q Stress[1] (ksi) CUF EPU with Allowable CLTP Environmental CLTP EPU CLTP EPU 40/ 60 years Component (ASME 60 years Fatigue Uen Allowable (3,527 MWt)[2] (4,031 MWt) (3,527 MWt) [2] (4,031 MWt) Code Limit) (3,527 MWt) 60 years (4,031 MWt) 2-inch Instrumentation Nozzle Unit 1: 80.20/0.03 [13] 83.73/0.048 [13] 69.9/1.0 [13] 0.06[10] 1.0 Unit 2: 80.20/0.03 [13] 83.73/0.048 [13] 69.9/1.0 [13] 0.06[10] NA Note 14 NA 1.0 Unit 3: 80.20/0.03 [13] 83.73/0.048 [13] 69.9/1.0 [13] 0.06[10] 1.0 Support Skirt Unit 1: 115.9/NA[7] 90.053/51.273 [7,17] 80.10/80.10 0.904 0.114/0.129 [12] 1.0 [7] [7,17] [12] Unit 2: 115.9/NA 90.053/51.273 80.10/80.10 0.904 Note 4 0.090/0.129 NA 1.0 [7] [7,17] [12] Unit 3: 115.9/NA 90.053/51.273 80.10/80.10 0.904 0.090/0.129 1.0 2-117

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-6 CUFs and Sp+q Values of Limiting Components (continued) P + Q Stress[1] (ksi) CUF EPU with Allowable CLTP Environmental CLTP EPU CLTP EPU 40/ 60 years Component (ASME 60 years Fatigue Uen Allowable (3,527 MWt)[2] (4,031 MWt) (3,527 MWt) [2] (4,031 MWt) Code Limit) (3,527 MWt) 60 years (4,031 MWt) Refueling Containment Skirt Unit 1: 86.70 87.87 88.00 0.328 0.283/0.304 [12] 1.0 [12] Unit 2: 86.70 87.87 88.00 0.328 Note 4 0.348/0.304 NA 1.0 [12] Unit 3: 86.70 87.87 88.00 0.328 0.348/0.304 1.0 Shroud Support Unit 1: 136.8/0.170 142.82/0.263 [13] 69.90/1.0 0.170 1.0 Unit 2: [13] 142.82/0.263 [13] [13] 0.170 NA Note 14 1.0 Unit 3: 136.8/0.170 142.82/0.263 [13] 69.90/1.0 0.170 1.0 NA [13] [13] 136.8/0.170 69.90/1.0 [13] [13] 2-118

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-6 CUFs and Sp+q Values of Limiting Components (continued) P + Q Stress[1] (ksi) CUF EPU with Allowable CLTP Environmental CLTP EPU CLTP EPU 40/ 60 years Component (ASME 60 years Fatigue Uen Allowable (3,527 MWt)[2] (4,031 MWt) (3,527 MWt) [2] (4,031 MWt) Code Limit) (3,527 MWt) 60 years (4,031 MWt) Main Closure Studs Unit 1: 103.3 103.3 110.1 0.762 1.0 Unit 2: 103.3 103.3 110.1 0.762 NA Note 15 NA 1.0 Unit 3: 103.3 103.3 110.1 0.762 1.0 Vessel Shell Unit 1: 39.00 40.72 80.00 0.032 0.003 1.0 Unit 2: 39.00 40.72 80.00 0.032 Note 4 Note 5 0.010 1.0 Unit 3: 39.00 40.72 80.00 0.032 0.010 1.0 2-119

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-6 CUFs and Sp+q Values of Limiting Components (continued) P + Q Stress[1] (ksi) CUF EPU with Allowable CLTP Environmental CLTP EPU CLTP EPU 40/ 60 years Component (ASME 60 years Fatigue Uen Allowable (3,527 MWt)[2] (4,031 MWt) (3,527 MWt) [2] (4,031 MWt) Code Limit) (3,527 MWt) 60 years (4,031 MWt) CRD Penetration Unit 1: 73.00/ 76.21/0.006 [13] 70.00/1.0 0.005 1.0 Unit 2: 0.005[13] 119.02/0.110 [13] [13] 0.093 NA Note 14 NA 1.0 Unit 3: 114.0/ 119.02/0.110 [13] 70.00/1.0 0.093 1.0 0.093[13] [13] 114.0/ 70.00/1.0 0.093[13] [13] Stabilizer Bracket Unit 1: 51.20 51.89 80.00 0.170 1.0 Unit 2: 51.20 51.89 80.00 0.170 NA Note 14 NA 1.0 Unit 3: 51.20 51.89 80.00 0.170 1.0 2-120

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-6 CUFs and Sp+q Values of Limiting Components (continued) P + Q Stress[1] (ksi) CUF EPU with Allowable CLTP Environmental CLTP EPU CLTP EPU 40/ 60 years Component (ASME 60 years Fatigue Uen Allowable (3,527 MWt)[2] (4,031 MWt) (3,527 MWt) [2] (4,031 MWt) Code Limit) (3,527 MWt) 60 years (4,031 MWt) Jet Pump Instrumentation Seal Unit 1: 38.35 41.18 51.75 Unit 2: 38.35 41.18 51.75 Note 16 NA Note 16 NA Note 16 Unit 3: 44.10 47.35 51.75 2-121

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Notes for Table 2.2-6:

1. Stress value for the limiting location of the component.

2. CLTP was conservatively evaluated at 102% of the stretch power uprate level (3,458

1.02 = 3,527 MWt).

3. These values are for the nozzle blend radius, which is the bounding location.

4. For 60-year license, EPU values are used.

5. Instead of EPU CUF, the environmentally assisted fatigue usage factors for 60 years are provided.

6. The first value is for the nozzle blend radius and the second value is for the safe end. These are the limiting locations evaluated for the FW nozzle.

7. Thermal bending included/Thermal bending removed.

8. The value is for the safe end which is the limiting location.

9. The first value is for the nozzle blend radius and the second value is for the safe end.

10. The value is for the limiting location of the nozzle.

11. The first value is for the nozzle blend radius and the second value is for the nozzle body cladding.

12. Value reported for 40-year license and 60-year license (40-year value / 60-year value).

13. P + Q value / Elastic-plastic CUF value

14. Fatigue evaluation is not performed as the generic disposition is applied.

15. Fatigue evaluation is not performed as the component specific disposition is applied. The fatigue value CUF < 1.0.

16. Fatigue evaluation is exempted by ASME Code Section III, NB-3222.4(d) 1980 Edition with Addenda to and including Winter 1981 (Units 1 and 2) and 1986 Edition (Unit 3).

17. P + Q without thermal bending is less than the ASME allowable. Therefore it meets ASME requirements. Note that the simplified elastic-plastic analysis is performed for fatigue evaluations according to ASME requirements.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-7 RIPDs for Normal Conditions Parameter CLTP 1 EPU 2 (psid) (psid) Shroud Support Ring and Lower Shroud 31.06 32.89 Core Plate and Guide Tube 22.84 24.40 Upper Shroud 8.23 8.55 Shroud Head 8.42 9.43 Shroud Head to Reactor Water Level (Irreversible 3) 10.8 12.24 Shroud Head to Reactor Water Level (Elevation 3) 1.07 0.94 Fuel Channel Wall (Core Average Power Bundle) 9.1 10.4 Fuel Channel Wall (Maximum Power Bundle) 11.67 13.31 Top Guide 0.61 0.61 Steam Dryer (OSD / RSD) 4 0.33 0.42 / 0.42 Notes:

1. At 105% rated core flow with GE13 fuel.

2. At 105% rated core flow with GE13 fuel. The GE13 results are bounding for operation with GE14, ATRIUM-10, and ATRIUM 10XM.

3. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud (at the midpoint between the top of fuel and the shroud dome) and the exit of the separators.

4. OSD = Original Steam Dryer. RSD = Replacement Steam Dryer.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-8 RIPDs for Upset Conditions Parameter CLTP 1 EPU 2 (psid) (psid) Shroud Support Ring and Lower Shroud 33.46 35.29 Core Plate and Guide Tube 25.24 26.80 Upper Shroud 12.34 12.82 Shroud Head 12.63 14.14 Shroud Head to Reactor Water Level (Irreversible 3) 16.20 18.36 Shroud Head to Reactor Water Level (Elevation 3) 1.61 1.41 Fuel Channel Wall (Core Average Power Bundle) 12.0 13.3 Fuel Channel Wall (Maximum Power Bundle) 14.57 16.21 Top Guide 1.10 0.92 Steam Dryer (OSD / RSD) 4 0.50 0.62 / 0.62 Notes:

1. At 105% rated core flow with GE13 fuel.

2. At 105% rated core flow with GE13 fuel. The GE13 results are bounding for operation with GE14, ATRIUM-10, and ATRIUM 10XM.

3. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud (at the midpoint between the top of fuel and the shroud dome) and the exit of the separators.

4. OSD = Original Steam Dryer. RSD = Replacement Steam Dryer.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-9 RIPDs for Faulted Conditions Parameter CLTP 1 EPU 2 (psid) (psid) Shroud Support Ring and Lower Shroud 52.0 51.0 Core Plate and Guide Tube 30.0 28.5 Upper Shroud 30.0 29.0 Shroud Head 30.0 29.5 Shroud Head to Reactor Water Level (Irreversible 3) 32.0 32.0 Shroud Head to Reactor Water Level (Elevation 3) 2.1 1.4 Fuel Channel Wall (Core Average Power Bundle) 12.9 14.0 Fuel Channel Wall (Maximum Power Bundle) 14.6 15.5 Top Guide 2.8 1.1 Steam Dryer (OSD / RSD) 4 5 7.7 / 8.0 7.7 / 8.0 Notes:

1. At 105% rated core flow with GE13 fuel. The GE13 results are bounding for operation with GE14, ATRIUM-10, and ATRIUM 10XM.

2. Evaluations at these points considered both normal and reduced FW temperatures.

The reduced FW temperature of 55ºF was used for EPU. The GE13 RIPD results are bounding for operation with GE14 fuel.

3. Irreversible loss is the loss across the separators; the elevation loss or reversible head loss is the loss between the inside shroud (at the midpoint between the top of fuel and the shroud dome) and the exit of the separators.

4. OSD = Original Steam Dryer. RSD = Replacement Steam Dryer.

5. Steam dryer loads are limiting at the cavitation interlock condition. The faulted condition steam dryer load is therefore unaffected by EPU implementation.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-10 Governing Stress Results for RPV Internal Components Component Category/ Stress/Load CLTP EPU Allowable Location Service Category Basis Value Condition Value 1 Shroud Normal/Upset Buckling (psi) 2,176 2,176 7,720 Faulted Buckling (psi) 10,804 10,804 15,440 2 Shroud Support Design Stress (psi) 24,500 30,062 34,950 Operating Faulted Stress (psi) 66,000 66,000 69,900 3 Core Plate Normal/Upset Buckling/Sliding 25.24 26.80 28.0 (including core Delta P (psid) plate plugs) Emerg./Fault. Buckling/Sliding 32.0 32.0 42.0 Delta P (psid) 4 Top Guide Normal/Upset Longest Beam 23,754 23,754 25,388 Stress (psi) Emergency Longest Beam 33,674 33,674 38,081 Stress (psi) Faulted Longest Beam 33,674 33,674 50,775 Stress (psi) 5 Control Rod Drive Qualitative assessment. Not affected and remains qualified for Housing EPU. 6 Control Rod Guide Upset Buckling (p/pC) 0.24 0.255 0.40 Tube Emergency Buckling (p/pC) 0.304 0.304 0.60 Faulted Buckling (p/pC) 0.304 0.304 0.80 7 Orificed Fuel Upset Stress (psi) 12,413 12,527 15,580 Support Faulted Stress (psi) 23,505 23,505 35,440 8 Fuel Channels Qualified per Proprietary Fuel Design Basis. 2-126

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-10 Governing Stress Results for RPV Internal Components (continued) Component Category/ Stress/Load CLTP EPU Allowable Location Service Category Basis Value Condition Value 9 Feedwater Normal/Upset (Pm + Pb + Q - 70,800 70,910 76,500 Sparger (Slotted Ring) Therm. Bending) (psi) Normal/Upset (Pm + Pb) (psi) 5,190 6,990 21,450 (Header Pipe/Tee) Emergency (Pm + Pb) (psi) 6,020 7,820 28,600 (Header Pipe/Tee) Faulted (Pm + Pb) (psi) 33,690 35,490 42,900 (Header Pipe/Tee) 10 Jet Pump Qualitative assessment. Not affected and remains qualified for (including riser EPU. brace attachment repair - Unit-3) 11 Core Spray Line Qualitative assessment. Not affected and remains qualified for and Sparger EPU. (includes T-Box and downcomer repairs - Unit-3) 12 Access Hole Cover Normal/Upset (Pm + Pb) (psi) 6,756 7,093 34,950 Emergency/ Qualitative assessment. Remains qualified for EPU. Faulted 13 Shroud Head and Normal/Upset (Pm + Pb) (psi) 33,993 34,489 34,950 Steam Separator Emergency (Pm + Pb) (psi) 31,348 34,671 52,425 Assembly Faulted (Pm + Pb) (psi) 41,432 41,758 69,900 14 In-Core Housing Qualitative assessment. Not affected and remains qualified for and Guide Tube EPU. 2-127

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-10 Governing Stress Results for RPV Internal Components (continued) Component Category/ Stress/Load CLTP EPU Allowable Location Service Category Basis Value Condition Value 15 Vessel Head Qualitative assessment. Not affected and remains qualified for Cooling Spray EPU. Nozzle 16 Jet Pump Qualitative assessment. Not affected and remains qualified for Instrument EPU. Penetration Seal 17 Differential Qualitative assessment. Not affected and remains qualified for Pressure and EPU. Standby Liquid Control Line 2-128

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-11 Systems with Pumps and Valves in the IST Program System IST IST GL 89-10 GL 95-07 System System GL 96-05 Affected Number Pumps Valves Valves by EPU Valves Main Steam 001 NA X X NA X Feedwater 003 NA X NA NA X Heater Drains and Vents 006 NA X NA NA NA Boiler Drains and Vents and 010 NA X NA NA NA Blowdown Auxiliary Boiler System 012 NA X NA NA NA RHR Service Water 023 X X X NA NA Raw Water Cooling 024 NA X NA NA NA Reactor Water Sampling 043 NA X NA NA NA Raw Water Chemical Treatment 050 NA X NA NA NA System Standby Liquid Control 063 X X NA NA Primary Containment 064 NA X NA NA X Emergency Equipment Cooling 067 NA X NA NA NA Water Reactor Recirculation 068 NA X X NA NA Reactor Water Cleanup 069 NA X X NA X 2-129

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.2-11 Systems with Pumps and Valves in the IST Program (continued) System IST IST GL 89-10 GL 95-07 System System GL 96-05 Affected Number Pumps Valves Valves by EPU Valves Reactor Building Closed 070 NA X X NA NA Cooling Water RCIC 071 X X X X NA HPCI 073 X X X X NA RHR 074 X X X X NA CS 075 X X X X NA Radwaste 077 NA X NA NA NA Fuel Pool Cooling 078 NA X NA NA NA Control Rod Drive 085 NA X NA NA NA Note: Cells with NA indicate that the system has no components in the respective program. Table 2.2-12 EPU Effects to Browns Ferry Program Valves Maximum Ambient Valve ID Valve Function Differential Temp EPU Effect Pressure Change Change, psi No effect on Core Spray Inboard Injection 3-FCV-75-53 -- + 5°F valve capability Valve or margin. 2-130

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.3 Electrical Engineering 2.3.1 Environmental Qualification of Electrical Equipment Regulatory Evaluation Environmental qualification (EQ) of electrical equipment involves demonstrating that the equipment is capable of performing its safety function under significant environmental stresses that could result from DBAs. The NRCs acceptance criteria for EQ of electrical equipment are based on 10 CFR 50.49, which sets forth requirements for the qualification of electrical equipment important to safety that is located in a harsh environment. Specific NRC review criteria are contained in SRP Section 3.11. Browns Ferry Current Licensing Basis The Browns Ferry program for environmental qualification of electrical equipment is described in Browns Ferry UFSAR Section 7.1.6, Environmental Qualification of Electrical Equipment. In addition to the evaluations described in the Browns Ferry UFSAR, Browns Ferrys environmental qualification of electrical equipment was evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The environmental qualification of electrical equipment for license renewal is discussed in NUREG-1843, Sections 2.6.1.4 and 4.4. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 10.3.1 of the CLTR addresses the effect of EPU on the Environmental Qualification of Electrical Equipment. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Electrical Equipment Plant Specific Disposition The CLTR states that the increase in power level increases the radiation levels experienced by equipment during normal operation and accident conditions. Because of the constant pressure assumption, there is only a very small effect on pressure and temperature conditions. 2-131

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) All electrical equipment in the EQ program was evaluated by developing a list of components that are identified as being in the electrical EQ program. For areas affected by EPU operating conditions, associated safety-related electrical equipment was reviewed consistent with the requirements of NUREG-0588 Category II, Division of Operating Reactors (DOR) Guidelines of IE Bulletin No.79-01B or 10 CFR 50.49 (or NUREG-0588 Category I) and NRC Regulatory Guide 1.89 to ensure the existing qualification for the normal and accident conditions expected in the areas where the devices are located remains adequate. The review focused on the effect of environmental changes due to EPU. The DOR Guidelines of IE Bulletin No. 79-01B and 10 CFR 50.49 acceptance criteria were used in making this determination. Margin evaluation complies with the recommendations of IEEE 323-1971 and IEEE 323-1974 or is qualitatively justified based on separate data that establish material capabilities. The EQ Program equipment qualification basis was evaluated using the expected changes to existing normal and accident radiation doses when operating at the EPU increased reactor power level. Table 2.3-1 summarizes the changes to the EQ environmental parameters due to EPU and changes due to remodeling Standby Gas Treatment Flow in all Reactor Buildings. The normal and post-LOCA radiation dose value changes are based on an EPU scaling factor and applied to plant areas inside and outside primary containment. The EPU scaling factor is for gamma and beta contributors applied over the post-accident time intervals up to 100 days. Once the effect on post-EPU radiation dose value was determined, all equipment was evaluated and found to remain qualified for post-EPU parameters with respect to radiation. This includes equipment with sufficient life to demonstrate radiation qualification through the end of plant life (60 years) or with designated qualified life of less than 60 years. Limited life components are addressed within the Browns Ferry EQ program as warranted. EQ file updates will be completed as required prior to EPU implementation per TVA procedure NPG-SPP-09.2, Equipment Environmental Qualification (EQ) Program. Post-EPU EQ compliance is not contingent upon any plant modifications, replacement of equipment, or other compensatory measure. No new EQ electrical components are being added to the 10 CFR 50.49 program due to EPU. Inside Primary Containment EQ for safety-related electrical equipment located inside containment is based on steam line break (SLB) and/or DBA-LOCA conditions along with the temperature, pressure, humidity, and radiation consequences, and includes the normal operating environments expected to exist during plant operation. These changes will occur over the 100-day profile, necessitating the evaluation of the new requirements based on the series of individual component type tests available to the EQ Program. The higher Drywell EPU accident temperature profile has been determined to be acceptable because there is sufficient positive margin in the CLTP Accident Degradation Evaluation calculations to ensure the equipment would still function as required at the higher temperatures. The current Drywell EQ CLTP temperature profile and revised EPU EQ 2-132

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) temperature profile are shown in Figure 2.3-1. These profiles were developed by combining the bounding curves for both the SLB and LOCA DBAs. These profiles are also the worst case EQ enveloping profiles for all plant EQ locations in the primary and secondary containments. The EPU profile exceeds the existing Drywell profile over the entire 100 days. The EPU temperature peak occurs at the same time as the CLTP peak, it is slower to degrade until the RPV depressurizes, and the 100 day final temperature is slightly higher than the CLTP ending temperature. The peak accident pressure will increase from 50.64 psig (65.04 psia) to 50.9 psig (65.3 psia). All components were re-evaluated with respect to the EPU peak pressure of 50.9 psig and are documented in Table 2.3-2, which demonstrates that the component qualification limits bound the postulated EPU accident pressure with sufficient margin. The inside containment normal pressure and normal and accident humidity conditions will not change as a result of EPU. The maximum normal and maximum abnormal temperatures will increase by 0.12°F, which will be rounded to 1°F. This small temperature increase will have a minor effect on thermal qualified life and will be addressed as part of normal EQ maintenance replacement. There are no EQ components in the Wetwell (Torus); therefore, an accident curve is not provided for the Wetwell. The current radiation levels under normal plant conditions were conservatively evaluated to increase in proportion to the increase in reactor thermal power. The total integrated dose (TID) levels generally will increase by < 16% above CLTP levels inside primary containment, see Table 2.3-4 for comparisons. The total integrated doses (normal plus accident) for EPU conditions were evaluated and determined not to exceed the radiation doses for most of the equipment located inside primary containment. Only some cables and solenoids have a decreased qualified life due to EPU implementation. Table 2.3-3 provides a comparison of each type of component qualification dose with the EPU EQ TID for relevant plant locations. It was determined that some Drywell cables do not have sufficient radiation qualification to meet or exceed the EPU total integrated dose (normal + accident) for 60 years of operation. These cables, as a result of the EPU dose values, will require maintenance replacement once their accumulated normal dose equals the qualification dose minus the accident dose. The normal EQ maintenance activity will replace those cables which have a limited qualified life due to radiation aging. The increased radiation doses will result in a reduction of the radiation life for some solenoids located inside primary containment. However, the qualified life based on thermal aging is shorter than that of the radiation life for these solenoids. Therefore, the component qualified life will not be reduced due to the increased radiation doses at EPU. Normal EQ maintenance activities will replace those components which have a limited qualified life due to thermal aging. Humidity during LOCA events inside primary containment typically reaches saturation (100% RH and condensing) early in the event progression and remains saturated for most if not all of the analyzed period. Because of this characteristic, humidity is typically not graphed. The Browns Ferry EQ program assumes saturated conditions for the duration of the LOCA/SLB event; therefore EPU has no effect on the EQ qualification to humidity. 2-133

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Outside Primary Containment The HELB analysis for the secondary containment evaluates numerous break locations and sizes occurring in the HPCI, RCIC, RWCU and MS/FW systems. The effect of operating at EPU conditions was evaluated based on the mass and energy releases and the analytical bases used in Browns Ferry HELB analyses. It was determined that the RWCU HELB will change peak temperatures in some EQ rooms. However, the increased temperatures at EPU conditions remain bounded by the temperatures used for the equipment qualification. See Table 2.3-5 for the effected EQ room descriptions. Table 2.3-5 provides a listing of the EQ components affected by the increase in RWCU HELB peak accident temperatures and their qualification limit. Figure 2.3-2 depicts the bounding HELB temperature EPU and CLTP profiles, which is for the Steam Tunnel. The Steam Tunnel HELB temperature profile is not affected by EPU. Unlike the Drywell, the EQ Program does not assume saturated conditions for the duration for the HELB events outside primary containment. The only HELB event which will change due to EPU is the RWCU line break. For reactor building areas where the relative humidity reaches 100% following an RWCU line break, the period for 100% relative humidity conditions will increase to 4 hours. This increased period at 100% relative humidity does not exceed the post-accident conditions for which the affected safety-related electrical equipment is qualified. The Steam Tunnel (EQ Room 7) bulk temperature at EPU conditions will increase by 0.37°F and is conservatively rounded up to a 1°F increase for the EPU evaluation. This temperature increase has a minor effect on thermal qualified life and is addressed as part of normal EQ maintenance replacement. The pressure and humidity conditions in the Steam Tunnel do not change due to EPU. The pressures, ambient temperatures, and humidity conditions for all other EQ locations outside primary containment remain unchanged by EPU. The current radiation levels under normal plant conditions were conservatively evaluated to increase in proportion to the increase in reactor thermal power at EPU; see Table 2.3-4 for a listing of the CLTP and EPU values. The TID levels will generally increase by less than 20% above CLTP levels outside primary containment. A few areas will increase greater than 20% due to EPU dose rate increases. The qualification basis for the EQ program equipment was evaluated based on the revised EPU normal and accident radiation dose values, except where EPU component (location)-specific dose values are applied. See Table 2.3-5 for identification of each type of component, along with the EPU EQ TID for relevant plant locations and qualification dose. EQ equipment was evaluated and most of the EQ equipment was found to remain fully qualified for post-EPU parameters with respect to radiation. Some of the components, which were determined to not have sufficient radiation qualification to meet or exceed the EPU total integrated dose (normal + accident) for 60 years of operation, will require replacement. The normal EQ maintenance activity will replace those components which have a limited qualified life due to thermal aging, mechanical aging or radiation aging. Accident temperature, pressure, and humidity environments used for qualification of equipment outside primary containment result from an MSLB, or other HELBs, whichever is limiting for each plant area. The HELB pressure profiles for CLTP conditions were determined to be 2-134

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) bounding for EPU conditions. The peak HELB temperatures at EPU rated thermal power are bounded by the values used for equipment qualification at CLTP conditions. Conclusion TVA has evaluated the effects of the proposed EPU on the environmental conditions for the qualification of electrical equipment. The evaluation indicates that the electrical equipment will continue to meet the relevant requirements of 10 CFR 50.49 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the EQ of electrical equipment. 2.3.2 Offsite Power System Regulatory Evaluation The offsite power system includes two or more physically independent circuits capable of operating independently of the onsite standby power sources. The NRCs acceptance criteria for offsite power systems are based on GDC-17. Specific NRC review criteria are contained in SRP Sections 8.1 and 8.2, Appendix A to SRP Section 8.2, and Branch Technical Positions PSB-1 and ICSB-11. Browns Ferry Current Licensing Basis The Browns Ferry offsite power system is described in Browns Ferry UFSAR Section 8.3, Transmission System. Final GDC-17 is applicable to Browns Ferry as described in UFSAR Section 8.3. Browns Ferrys Offsite Power System was evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The offsite power system is discussed in the NUREG-1843, Sections 2.5.1 and 3.6. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.1 of the CLTR addresses the effect of EPU on the alternating current (AC) Power System. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: 2-135

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR AC Power (Degraded Voltage) Plant Specific Disposition Meets CLTR AC Power (Normal Operation) Plant Specific Disposition 2.3.2.1 AC Power (Degraded Voltage) As explicitly stated in Section 6.1 of the CLTR, the increase in thermal power translates to an increased electrical output from the station. Changes in electrical requirements to support normal plant operation are not safety-related. The increased power from the main generator will have no adverse effect on the transmission systems ability to supply loads required for safe shutdown. The Browns Ferry Unit 1, Unit 2, and Unit 3 main generators are each connected to a set of three single phase main generator step-up transformers. The 500 kV and 161 kV switchyards consist of the buswork, disconnect switches, circuit breakers, and the associated control and protection systems. Browns Ferry has two sources of offsite power from the 500 kV and 161 kV transmission network. The offsite power circuits from the transmission network to the safety-related Division I (4.16 kV shutdown boards A and B) and Division II (4.16 kV shutdown boards C and D) for Units 1 and 2 are as follows: From the 500 kV switchyard through Unit Station Service Transformer (USST) 1B to a 4.16 kV unit board. That unit board feeds 4.16 kV shutdown bus 1 or 2, which then feed two of the Unit 1 and 2 4.16 kV shutdown boards (A and B or C and D). From the 500 kV switchyard through USST 2B to a 4.16 kV unit board. That unit board feeds 4.16 kV shutdown bus 1 or 2, which then feed two of the Unit 1 and 2 4.16 kV shutdown boards (A and B or C and D). From the 161 kV transmission network, through Common Station Service Transformer (CSST) A to Start bus 1A or 1B, to a 4.16 kV unit board, to 4.16 kV shutdown bus 1 or 2, which then feeds two of the Unit 1 and 2 4.16 kV shutdown boards (A and B or C and D). The offsite power circuits from the transmission network to the safety-related Division I (4.16 kV shutdown boards 3EA and 3EB) and Division II (4.16 kV shutdown boards 3EC and 3ED) for Unit 3 are as follows: From the 500 kV switchyard through USST 3B to 4.16 kV unit board 3A and/or 3B. Each unit board feeds two of the Unit 3 4.16 kV shutdown boards (3EA and 3EB or 3EC and 3ED). 2-136

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) From the 161 kV transmission network, through CSST A to Start bus 1A or 1B, then to a 4.16 kV unit board. That unit board feeds two of the Unit 3 4.16 kV shutdown boards (3EA and 3EB or 3EC and 3ED). From the 161 kV transmission network, through CSST B to Start bus 1A or 1B, then to a 4.16 kV unit board. That unit board feeds two of the Unit 3 4.16 kV shutdown boards (3EA and 3EB or 3EC and 3ED). Multiple Browns Ferry units can utilize the 161 kV offsite power circuit simultaneously. However, once a load from one Browns Ferry unit is connected to a 161 kV offsite power circuit (via the Start buses), Browns Ferry operating procedures require disabling the automatic transfer of selected 4.16 kV unit boards and/or 4.16 kV common boards on the other Browns Ferry units to the 161 kV circuits. This is to prevent overloading of the CSSTs. Upon a loss of the normal 500 kV offsite circuit, the emergency diesel generators would supply the associated safety-related loads in both divisions needed to mitigate the immediate consequences of an accident or analyzed operational transient. Therefore, the 161 kV circuit CSSTs can still be credited as a qualified alternate offsite circuit for multiple units. However, access to the 161 kV circuit will require a delayed manual transfer when operators can manually control the loads on the 4.16 kV Start buses to support long term post-accident or transient recovery and shutdown. This description of the Browns Ferry power distribution system is unchanged by EPU. The protective relaying schemes are designed to protect the equipment from electrical faults. Electrical ratings and margins associated with major components of the offsite power system are given in Table 2.3-6. The review of the protective relaying for the main generator determined that no changes are required for operation at EPU. The transmission system stability study modeled electrical ratings and margins associated with major components of the offsite power system. The transmission system stability study documents the load flow analysis of the off-site power supply and provides input to site calculations. A transmission system stability study has been performed, considering the increase in electrical output, to demonstrate conformance to 10 CFR 50 Appendix A, General Design Criteria (GDC) 17. Details of this study are provided in LAR Attachment 43 Transmission System Stability Evaluation to the EPU license amendment request. The analysis shows the limiting pre-event outages for the 2015 (CLTP) and 2019 (post-EPU) peak cases. The pre-event outages include the loss of TVAs largest generating unit, loss of a Browns Ferry unit, and loss of each of the transmission lines within the TVA grid. The loss of 500 kV-161 kV transformer banks bounds the effect to the system of loss of any large loads on the grid. Browns Ferry offsite power is adequate to operate loads required for safe shutdown and will preclude the inadvertent separation from the offsite supply. Therefore, the offsite power at degraded voltage meets all CLTR dispositions. The grid stability analysis evaluated the effect of the EPU on the off-site power transmission system (i.e., the grid) and the ability to meet the minimum required voltage levels to the on-site power system of each Browns Ferry Unit. 2-137

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The effect of EPU on the on-site power distribution system, including the degraded voltage protection system, was reviewed separately in plant calculations using the Browns Ferry analytical electrical system computer mode. Degraded voltage protection is provided for the 4 kV safety-related shutdown boards. Each board has three upper degraded voltage relays and three lower degraded voltage relays that sense each of the three phase-to-phase voltages on the shutdown board potential transformer secondaries. A third set of three relays detect loss of power. If two of the three upper degraded voltage relays sense a shutdown board voltage above the setpoint (4400 V) for more than five seconds, the time delay relays will actuate and give annunciation. The annunciation will alert the operators to reduce board voltage. If two of the three loss of voltage relays sense a shutdown board voltage below the setpoint (2870 V) for more than 1.5 seconds, the diesel generator starts. If the condition exists for 5 seconds, the relays will initiate load shedding and 4 kV shutdown board power isolation for diesel generator breaker closure. If two of the three relays lower degraded voltage relays sense a shutdown board voltage below the setpoint (3920V) for about 0.3 seconds, the time delay relays will begin timing. Due to inaccuracy of the relay, dropout may occur at any voltage between 3940 V and 3899 V. The analysis was performed at the degraded voltage lower boundary of 3900 V to ensure all connected safety-related loads and boards remain within their rated operating voltage ranges. The relays reset when voltage recovers greater than 3962 V. Due to inaccuracy of the relay, reset may occur at any voltage between 3941 V and 3983 V. In the analysis, to ensure the relays reset, the voltage must recover to at least 3983 V. If a degraded voltage exists for approximately 4 seconds, the diesel generator will start. If the relays are actuated and the voltage recovers within 5.95 seconds, the relays will reset and the board will not transfer to the diesels. If a degraded voltage exists for greater than 5.95 seconds, the relays will initiate load shedding and 4 kV shutdown board power isolation for diesel generator breaker closure. To ensure margin, the analysis uses 5.6 seconds for resetting the degraded voltage relays. The results of the analysis indicate that no changes are required to the degraded voltage setpoints as a result of EPU. 2.3.2.2 AC Power (Normal Operation) TVA owns both the transmission system and Browns Ferry Units 1, 2, and 3. The Off-site Power System is part of the transmission system. The operation and maintenance of the Off-site Power System is under the control of the Transmission Power Systems organization under TVA. The operation and maintenance of the On-site Power systems are under the control of Browns Ferry. The On-site power system ends at the high side outputs of the main power transformers, the common station service transformers, and the cooling tower transformers. As explicitly stated in Section 6.1 of the CLTR, the increase in thermal power translates to an increased electrical output from the station. For the off-site power supply, other than the main 2-138

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) generator, the equipment is adequate for operation with the uprated electrical output. Changes in electrical requirements to support normal plant operation are not safety-related. The existing off-site electrical equipment was determined to not need reinforcements for operation with the uprated electrical output and increased electrical loading. The increased power from the generator will have no adverse effect on the transmission systems ability to supply loads required for safe shutdown. The review of the transmission system stability study concluded the following: A transmission system stability study has been performed, considering the increase in electrical output, to demonstrate conformance to GDC 17 (10 CFR 50 Appendix A). Details of this study are provided in the Attachment 43 Transmission System Stability Evaluation to the EPU license amendment request. Browns Ferry offsite power voltages resulting from loss of TVAs largest generating unit, loss of a Browns Ferry unit, and loss of each of the transmission lines within the TVA grid, are adequate to operate loads required for safe shutdown and will preclude the inadvertent separation from the offsite supply. The transmission system stability study determined that there was no significant effect to the ability of the grid to supply sufficient shutdown power after the power uprate. Reactive load capabilities are stated in Attachment 43 Transmission System Stability Evaluation to the EPU license amendment request. In addition, the Interconnect System Impact Study being performed in accordance with the TVA Large Generator Interconnect Procedure will address reactive load in detail. The Generator Step-up (GSU) transformer rating is 1,500 MVA. The main generator ratings are 1,330 MVA for Unit 1 and 1,332 MVA for Unit 2 and Unit 3. The 500 kV switchyard components (i.e., bus, breakers, switches, transformers, and lines) are adequate for increased generator output associated with EPU. The Unit 1, Unit 2, and Unit 3 GSU transformers have been replaced and upgraded to support the increase in generator output. The maximum rating of the rewound generator is 1,330 MVA for Unit 1 and 1,332 MVA for Units 2 and 3, which is less than the generator step-up transformers rating of 1,500 MVA @ 65°C. The amount of power the generator sends through the GSU is equal to the generator output minus the house loads (that are tapped off the iso-phase bus through the USST before going through the GSU) and the transformer losses. As a result, under normal operations the transformers have substantial margin. The Unit 1, Unit 2 and Unit 3 isolated phase bus (IPB) duct work, cooling coils and fans have been modified to increase the continuous current rating to provide for operation at EPU output. The transmission system stability study did not identify any required upgrades for the 500 kV switchyard components associated with operation at the EPU electrical output. The model used in the study included components in the switchyard, GSU, and the main generator. The components performance are unaffected by operation at EPU conditions during normal operation and meets all CLTR dispositions. 2-139

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Conclusion TVA has evaluated the effects of the proposed EPU on the offsite power system. The evaluation indicates that the offsite power system will continue to meet the requirements of the final GDC-17 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the offsite power system. 2.3.3 AC Onsite Power System Regulatory Evaluation The alternating current (AC) onsite power system includes those standby power sources, distribution systems, and auxiliary supporting systems provided to supply power to safety-related equipment. The NRCs acceptance criteria for the AC onsite power system are based on GDC-17, insofar as it requires the system to have the capacity and capability to perform its intended functions during anticipated operational occurrences and accident conditions. Specific NRC review criteria are contained in SRP Sections 8.1 and 8.3.1. Browns Ferry Current Licensing Basis The Browns Ferry onsite AC power system is described in Browns Ferry UFSAR Section 8.4, Normal Auxiliary Power System and Section 8.5, Standby AC Power Supply and Distribution. Final GDC-17 is applicable to Browns Ferry as described in UFSAR Section 8.3. Browns Ferrys Onsite AC Power System was evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The onsite AC power system is discussed in NUREG-1843, Sections 2.5.1 and 3.6. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.1 of the CLTR addresses the effect of EPU on the AC Onsite Power System. The Browns Ferry AC on-site power distribution system consists of transformers, buses, and switchgear. AC power to the distribution system is provided from the transmission system, Transmission Switchyard, and from onsite Diesel Generators. The AC onsite power system consists of equipment and systems required to provide AC power to safety-related and non-safety-related loads as long as offsite power is available. This includes 500 kV transformers, 161 kV transformers, 22 kV transformers, 4.16 kV transformers, 4.16 kV switchgears, 480 V transformers, 480 V load centers and motor control centers, 208/120V 2-140

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) distribution panels, and Uninterruptible Power Supply (UPS) systems. The AC onsite power system provides standby power for safety-related unit functions from onsite standby diesel generators. The AC onsite power distribution system loads were reviewed under both normal and abnormal operating scenarios. In both cases, loads are computed based on equipment nameplate data or brake horsepower (BHP), with conservative demand factors applied. These loads are used as inputs for the computation of load, voltage drop and short circuit current values which were modeled in a commercially available electrical analysis software package. The significant changes in electrical load demand are associated with increasing the size of the condensate pumps and condensate booster pumps to restore hydraulic margin. The Browns Ferry review covered the AC power components with respect to their functional performance as affected by various configurations and loading conditions including full operation and unit trip with LOCA. The Browns Ferry review focused on the additional electric load that would result from the proposed EPU. Sufficient margin is available so that no electrical distribution system modifications are required. There are no changes to the emergency diesel generator loads or load sequencing for EPU. Therefore the fuel oil requirements do not change and the existing supply is adequate. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR AC Power (Degraded Voltage) Plant Specific Disposition Meets CLTR AC Power (Normal Operation) Plant Specific Disposition 2.3.3.1 AC Onsite Power (Degraded Voltage) As explicitly stated in Section 6.1 of the CLTR, the increase in thermal power translates to an increased electrical output from the station. Operation at the EPU power level is achieved in both normal and emergency conditions by operating equipment at or below the nameplate ratings. The EPU load flow and voltage drop analyses conservatively assume the transient electrical loading conditions that could exist upon a trip of either a condensate or condensate booster pump (CBP). Table 2.3-7 provides a summary of the loading changes to the onsite power analysis model due to EPU operation. The electrical system can tolerate the CBP overload condition for the short duration of the transient. Browns Ferry operation at EPU RTP includes upgraded condensate pumps and CBPs that deliver higher head, which improves operating margins. These modifications are complete except for the Unit 3 CBPs, which will be upgraded prior to Unit 3 EPU implementation. Larger reactor 2-141

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) recirculation pump motors have been installed to provide the ability to reliably achieve ICF conditions (105% RCF) at both CLTP and EPU RTP. The larger condensate pump motors (1,250 hp), CBP motors (3,000 hp) and reactor recirculation pump motors (8,657 hp) do not change the conclusions of the current Browns Ferry degraded voltage analysis. The analysis encompasses the safety-related 4.16 kV buses and is independent of voltage profiles for the balance of plant buses. Therefore, AC onsite power at degraded voltage meets all CLTR dispositions. 2.3.3.2 AC Onsite Power (Normal Operation) The existing protective relay settings are adequate; coordination is maintained between the pump motor breakers and the 4.16 kV and 480 V switchgear main feeder breakers. The existing protective relay settings for pump motors are based on the motors nameplate rating. The proposed loading of the buses with the larger condensate pump motors and CBP motors was evaluated and determined to be acceptable with retained margin. Detailed design of the replacement condensate pumps and condensate booster pumps addressed the revised relay settings to maintain coordination and ensure adequate cable sizing. The analytical electrical system computer model developed for Browns Ferry updated the main power transformer size to reflect the recent change of main power transformers and the proposed changes to the main generators and condensate pumps. Load flow, voltage drop and short circuit current evaluations were performed to verify the adequacy of the AC on-site power system for the proposed changes. Analyzed EPU BHP loads as discussed above are within the electrical distribution equipment capabilities (i.e., unit station service transformers, common station service transformers, cooling tower transformers and buses). The running and starting voltages for motors are within the acceptable values. Therefore, AC onsite power during normal operation meets all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on the AC onsite power system including the effects of the proposed EPU on the systems functional design. The evaluation indicates that the AC onsite power system will continue to meet the requirements of final GDC-17 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the AC onsite power system. 2.3.4 DC Onsite Power System Regulatory Evaluation The direct current (DC) onsite power system includes the DC power sources and their distribution and auxiliary supporting systems that are provided to supply motive or control power to safety-related equipment. 2-142

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The NRCs acceptance criteria for the DC onsite power system are based on GDC-17, insofar as it requires the system to have the capacity and capability to perform its intended functions during anticipated operational occurrences and accident conditions. Specific NRC review criteria are contained in SRP Sections 8.1 and 8.3.2. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-24 and 39. The Browns Ferry onsite DC power system is described in Browns Ferry UFSAR Section 8.8, Auxiliary DC Power Supply and Distribution. Browns Ferrys onsite DC power system was evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The onsite DC power system was determined to be within the scope of license renewal and the components subject to aging management review are evaluated on a plant wide basis as commodities. The onsite DC power supplies are described in NUREG-1843, Section 2.5.1. The electrical commodity groups are described in NUREG-1843, Section 2.5.1, and aging management for electrical commodities is described in NUREG-1843, Section 3.6. 2-143

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.2 of the CLTR addresses the effect of EPU on DC Power. The results of this evaluation are described below. The Browns Ferry direct current (DC) power distribution system provides control and motive power for various systems/components within the plant. The results of the battery sizing calculation for the LOCA/LOOP analysis scenario show that the existing batteries have adequate voltage at the end of the duty cycle. It also shows all required DC devices are within their design voltage range. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR DC Power Requirements Plant Specific Disposition As stated in Section 6.2 of the CLTR, DC loads are not significantly increased as a result of power uprate. The DC power system provides DC power to instrumentation, controls and motive force power required for equipment required to operate during accident conditions. Equipment includes safety-related switchgear, DC motor operated valves, HPCI turbine auxiliary oil pumps, HPCI gland seal condenser condensate pumps, RCIC gland seal vacuum tank condensate pumps, RCIC gland seal vacuum pumps, emergency lighting, and 120V inverters. These DC loads are not affected by EPU and there are no Class 1E DC Power load changes required for EPU implementation. Therefore this analysis concludes that the DC power system is adequate to support the EPU power increase. Conclusion TVA has evaluated the effects of the proposed EPU on the onsite DC power system and has accounted for the effects of the proposed EPU on the systems functional design. The evaluation indicates that the DC onsite power system will continue to meet the requirements of draft GDCs-24 and 39 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the DC onsite power system. 2-144

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.3.5 Station Blackout Regulatory Evaluation Station blackout (SBO) refers to a complete loss of AC electric power to the essential and nonessential switchgear buses in a nuclear power plant. SBO involves the Loss of Offsite Power (LOOP) concurrent with a turbine trip and failure of the onsite emergency AC power system. SBO does not include the loss of available AC power to buses fed by station batteries through inverters or the loss of power from "alternate AC sources". The NRCs acceptance criteria for SBO are based on 10 CFR 50.63. Specific NRC review criteria are contained in SRP Section 8.1 and other guidance provided in Matrix 3 of RS-001. Browns Ferry Current Licensing Basis The licensing basis for station blackout is described in Browns Ferry UFSAR Section 8.10, Station Blackout. Station blackout coping equipment was evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). Station blackout is discussed in NUREG-1843, Section 2.1.3. The station blackout coping equipment was determined to be within the scope of license renewal and the components subject to aging management review are evaluated on a plant wide basis as commodities. The electrical commodity groups are described in NUREG-1843, Section 2.5.1, and aging management for electrical commodities is described in NUREG-1843, Section 3.6. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 9.3.2 of the CLTR addresses the effect of EPU on Station Blackout. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topic addressed in this evaluation is: Browns Ferry Topic CLTR Disposition Result Meets CLTR Station Blackout Plant Specific Disposition The CLTR states that the plant responses to and coping abilities for an SBO event are affected slightly by operation at the power uprate level, due to the increase in the decay heat. 2-145

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) SBO was re-evaluated using the guidelines of Nuclear Management and Resources Council (NUMARC) 87-00, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors (Reference 55), and Regulatory Guide 1.155, Station Blackout (Reference 56). ((

))

The major characteristics that affect the ability to cope with a SBO event are identified in NUMARC 87-00 Revision 1 as:

1. Condensate inventory for decay heat removal

2. Class 1E battery capacity

3. Compressed gas capacity

4. Effects of loss of ventilation

5. Containment isolation By satisfying the criteria used in assessing the above characteristics, the plant is able to show satisfactory response to an SBO event.

NUMARC 87-00 Revision 1 (Section 7) provides two methods for conducting the assessment. The second method, the Alternate AC Approach, is used in the Browns Ferry SBO assessment. This method uses equipment that is capable of being electrically isolated from the preferred off-site and emergency on-site AC power sources. Alternate AC approach would entail a short period of time in an AC Independent state (up to one hour) while operators initiate power from the backup source. Browns Ferry SBO assessment assumes only one unit in station blackout with the other two units available to supply Alternate AC to the blacked-out unit. Use of Alternate AC power is limited to providing the required cooling systems to certain areas (control room, control bay, and electrical board rooms). The Alternate AC Approach is the method for calculating the coping period where the plant uses equipment that is capable of being electrically isolated from the preferred off-site and emergency on-site AC power sources. The four-hour coping duration criteria for Alternate AC Approach plants applies to Browns Ferry. Thus, Browns Ferry must meet the SBO requirements for at least four hours. Condensate Inventory for Decay Heat Removal Analyses have shown that the Browns Ferry condensate inventory is adequate to meet the SBO coping requirement for EPU conditions. The current CST inventory reserve (135,000 gallons) for RCIC and HPCI use ensures that adequate water volume is available to remove decay heat, depressurize the reactor and maintain reactor vessel level above the top of active fuel (approximately 114,000 gallons required at EPU conditions) during the coping period. 2-146

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Class 1E Battery Capacity There are no changes to the systems and equipment used to respond to a SBO and no change in the required coping time for operation at EPU conditions. The capacity of the existing Unit 1, Unit 2, and Unit 3 Class 1E batteries have adequate capacity and voltage to support the loads required to cope with a SBO event for a period of four hours. The calculated short circuit current is within the allowable limit of all protective devices to support a SBO event for a period of four hours. There are no changes to the design capacities of the Class 1E batteries required for EPU implementation. Evaluation of the Browns Ferry Class 1E Battery Capacity has shown that Browns Ferry has adequate battery capacity to support decay heat removal during a SBO for the required coping duration. The battery capacity analysis of record is conservative in that it includes an assumption in the model that various HPCI System loads, which are relatively large, operate for long periods during the SBO mitigation sequence. The CLTP mitigation sequence includes a single and relatively short HPCI cycle, and the resulting HPCI loads are bounded by the analysis. EPU does not significantly increase the HPCI loading, and similar to CLTP, only one relatively short HPCI cycle (approximately 7 minutes) is predicted by the EPU containment analysis analytical model (SHEX) model for SBO mitigation. Similarly, the number of required RCIC cycles in the CLTP and EPU mitigation sequence as predicted by the model is well below the RCIC initiations assumed in the analysis of record. Given the above, the battery capacity remains adequate to support operation of the required coping equipment operation after EPU. Compressed Gas Capacity The EPU SBO evaluation has shown that the Browns Ferry air operated safety relief valves (MSRVs) required for decay heat removal have sufficient compressed gas capacity for the required automatic and manual operation during the SBO event for EPU conditions. Simulation of SBO at EPU conditions, using the GEH SHEX code (See Table 1-1), indicates 74 total MSRV cycles are required as compared to the compressed gas inventory capable of thousands of cycles. Sufficient capacity remains to perform emergency RPV depressurization in case it is required. Therefore, adequate compressed gas capacity exists to support the MSRV actuations because the maximum number of MSRV valve operations is less than the capacity of the pneumatic supply. Timing of the operator action to cross-tie drywell control air to the containment atmospheric dilution (CAD) system is discussed in Section 2.11.1.2.1. Effects of Loss of Ventilation The effect of loss of ventilation in dominant areas of concern containing equipment necessary to achieve and maintain safe shutdown during a station blackout is evaluated for SBO. Areas containing equipment necessary to cope with an SBO event were evaluated for the effect of loss of ventilation due to an SBO. The evaluation shows that equipment operability is maintained because the SBO environment is milder than the existing design and qualification bases. 2-147

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) These areas for Browns Ferry included:

1. Control Building Rooms

2. Reactor Building Shutdown Board Rooms/Electrical Board Rooms

3. Drywell

4. RCIC Room

5. HPCI Room

6. Main Steam Tunnel

7. Torus Room The EPU SBO evaluation used the Alternate AC power source approach (Section 2.3) and the methodology of Section 2.7, Effects of the loss of ventilation methodology, of NUMARC 87-00, Revision 1 (Reference 55). The evaluation shows that equipment operability is maintained because the SBO environment is milder than the existing design and qualification bases, as summarized below:

The drywell evaluation using SHEX determined that the maximum EPU temperature compared to the maximum CLTP temperature does not change and is bounded by the existing design and qualification bases. Outside the drywell, the SBO loss-of-ventilation evaluation for the Control Building Rooms, Reactor Building Shutdown Board Rooms/Electrical Board Rooms, RCIC Room, HPCI Room, Main Steam Tunnel, Reactor Building General Floor Area, and Tours Room determined that, compared to CLTP, equipment operability is maintained because the SBO environment is milder than the existing design and qualification bases. Containment Isolation Containment isolation capability is not adversely affected by the SBO event for EPU as the SBO environment conditions do not change significantly after EPU and containment isolation is not adversely affected by the SBO for EPU. SBO Containment Response Analysis Key inputs for the Browns Ferry EPU SBO evaluation are contained in Table 2.3-8a. The SBO sequence of events is provided in Table 2.3-8b. The plant response to and coping capabilities for an SBO event are affected slightly by operation at EPU due to the increase in initial power level and decay heat. There are no changes to the systems and equipment used to respond to an SBO event, nor is the required coping time of four hours changed. The SBO event calculations for CLTP and EPU conditions are performed using the NRC-approved SHEX computer program and nominal ANSI/ANS 5.1-1979 decay heat source term at 100% equilibrium power for containment long-term pressure and temperature analysis. The energy contribution from metal-water reaction in the core is not modeled as the core does not uncover during the event and metal-water reaction would not occur (Reference 7). 2-148

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The battery capacity remains adequate to support RCIC and HPCI operation after EPU. Adequate compressed gas capacity exists to support all SBO mitigation equipment requirements. The SBO evaluation at EPU conditions shows a need for an additional 13% over CLTP of Condensate Storage Tank water for RCIC and HPCI use to ensure that adequate water volume is available to remove decay heat, depressurize the reactor, and maintain reactor vessel level above the top of active fuel. This increases the total Condensate Storage Tank volume required to approximately 114,000 gallons, which is well within the current Condensate Storage Tank inventory reserve of < 135,000 gallons. The key parameters for the SBO calculations for containment response at CLTP, EPU conditions, and the design limits are provided in the following table. Key Containment Parameters Comparison Parameter Units CLTP EPU Design Limit Peak Drywell Pressure psia 41.4 43.4 < 70.0 Peak SP (Torus)

                            ºF                194.1               203.7          < 281.0 Temperature Peak Drywell
                            °F                 276                 276           < 281.0 Temperature The containment response comparison is based on a scenario that provides conservative containment parameters designed to result in a more severe containment response.

Based on the above evaluations, Browns Ferry continues to meet the requirements of 10 CFR 50.63 after the EPU. Therefore, SBO meets all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on the plants ability to cope with and recover from an SBO event for the period of time established in the plants licensing basis. The evaluation indicates that the plant will continue to meet the requirements of 10 CFR 50.63 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to SBO. 2-149

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-1 Summary of EPU Effect on EQ DBA Environmental Parameters Effect of EPU Environmental Parameter Inside Primary Outside Primary Containment Containment Temperature, 1°F increase for all Drywell There will be no change to EQ locations, Maximum Normal Elevations. except for the Steam Tunnel, which will and Maximum increase by 1°F. Abnormal1 Temperature, Peak temperature increases to HELB- There will be no changes in the Accident1 336.9°F MS/FW, RCIC and HPCI peaks. Between 5 to 20°F increase in RWCU peaks in Rooms 6B, 8, 9A, 9D, 12, 13, 16, 19 and

20. There will be a decrease in EQ Room

18. (See Table 2.3-4 for Room descriptions).

LOCA - All but a few EQ locations will fractionally increase in temperature. The Torus Room and NE Pump Room will increase by 6.3°F and 1.2°F, respectively. Pressure Peak pressure increases to There will be no change. 65.3 psia. Humidity No change RWCU HELB time at 100% Relative Humidity will increase to four hours. Containment Spray No change Not applicable. Submergence No change There will be no change. Radiation TID will increase in all areas. The TID will increase in all areas except the Stack, RWCU Backwash Receiving Tank Room and Cleanup Demineralizer Valve Room, which will decrease. (Note: there are no EQ components in the Stack), Note:

1. Current component qualification testing bounds the temperature increases.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-2 Evaluation of Pressure Qualification of EQ Components in the Drywell Qualification Margin1 Component Pressure (psig) (%) Cable - Rockbestos Company [Type PXJ / PXMJ] 117.2 130 Cable - Eaton Cable [Type MS] 72 41 Cable - Brand Rex [Type PXJ / PXMJ] 113 122 Cable - Okonite Company [Type PX / PXJ / PXMJ] 112 120 Cable - Rockbestos Company [Type PXJ / PXMJ] 107 110 Cable - Rockbestos Company [Type MS] 117.2 130 Cable - Rockbestos Company [Coaxial Cable] 133 161 Cable - Okonite Company [Type PXJ / PXMJ] 112 120 Conduit Seal - CON/AX Corp. ECSA 75 47 Connector - CON/AX Coaxial Connector 80.4 57 Temperature Element - Weed SP611-1A-A-3-C-2-75-4D4- 75 47 2, 1B1D/612D-1A-C-6-C-17-00 Special Measure Transmitter - TEC VFMS2273A, 504A, 82 61 160-2 and 2273-C2 Limit Switch - N/AMCO EA-740 (QTR-180) 127 149 MOV - Limitorque AC/IPC SMB-000, 00, 0, 2, and SB-3 105 106 EPA - Conax 7504-10001-03, -04, -05 and 7FO2-10000-01 74.5/80/55 46/57/8 EPA - GE 100 Series (TVA ID Nos. EA and EF) 104/103 104/102 EPA - GE Canister 238X600RH 63 23 Solenoid - ASCO NP Series, 206 Series 113.3 122 Solenoid - Automatic Valve C-5497 65 27 Solenoid - Target Rock 1/2 SMS-S-02-5 68.2 33 Splices - Raychem WCSF-N Series 66 29 Splices - Raychem NPkV, NPKC, NPKP, NPKS, NMCK, 132/66 159/29 NCBK, NESK Raychem WCSF-N Series, NPkV, NPKC, NPKP, NPKS, 120 135 NMCK, NCBK, NESK Note:

1. Margin is calculated based on gauge pressure relative to the EPU peak Drywell pressure of 50.9 psig (65.3 psia).

2-151

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-3 Evaluation of Radiation Qualification of EQ Components Rooms EPU TID [rads] Qualification Component (Note 1) (Note 2) Dose [rads] Cable - American Insulated Wire [Type PXJ / PXMJ] OC 1.11E+07 1.70E+07 Cable - American Insulated Wire [Type PXJ / PXMJ OC 1.70E+07 9.01E+07 Cable - Anaconda Power and Control [Type PXMJ] OC 1.86E+07 1.80E+08 3 Cable - Rockbestos Company [Type PXJ / PXMJ] PC/OC 2.15E+08 1.80E+08 Cable - Continental (Anaconda) [Type MS] OC 1.11E+07 2.22E+07 Cable - Anaconda Cable Co. [Type PJJ] OC 1.85E+07 2.40E+07 Cable - Brand-Rex Company [Type PN / PNJ] OC 1.62E+07 2.40E+07 Cable - Brand Rex Company [Type MS] OC 1.85E+07 1.89E+08 3 Cable - Eaton Cable [Type MS] PC/OC 2.29E+08 1.80E+08 Cable - Essex International, Inc. [Type CPJ / CPJJ] OC 1.62E+07 1.70E+07 Cable - Essex Cable [Type PXMJ] OC 1.70E+07 1.70E+07 Cable - Essex Group, Inc. [Type PXJ / PXMJ] OC 1.70E+07 1.85E+08 Cable - General Cable Corporation [Type PNJ] OC 1.12E+07 2.40E+07 Cable - General Cable Corporation [Type CPSJ] OC 1.12E+07 1.70E+07 Cable - General Cable Corporation [Type CPJ] OC 1.12E+07 1.70E+07 Cable - Okonite Company [Type PXJ] OC 1.85E+07 1.81E+08 Cable - Phelps Dodge Cable [Type CPJ] OC 1.62E+07 1.70E+07 3 Cable - Triangle / PWC Inc. [Type PN / PNJ] OC 4.32E+08 2.40E+07 Cable - Triangle / PWC Inc. [Type CPJ / CPJJ] OC 1.12E+07 1.70E+07 Cable - Triangle / PWC Inc. [Type CPSJ] OC 1.12E+07 1.70E+07 Cable - Rockbestos KXL-780, Firewall III [Type MS] OC 1.11E+07 2.00E+08 Cable - Rome Cable [Type CPJ / CPJJ] OC 1.62E+07 1.70E+07 Cable - Rome (Cyprus) [Type PJJ] OC 5.19E+07 6.75E+07 Cable - Simplex Wire and Cable Co.[Type CPJ] OC 1.12E+07 1.7E+07 Cable - Sumitomo Electric Industries, Ltd. [Type CPJJ] OC 1.62E+07 2.23E+07 Cable - Tamaqua - Products Cable Corp. [Type PNJ] OC 1.70E+07 2.4E+07 Cable - Times Wire and Cable [Type MS] OC 1.11E+07 6.76E+07 Cable - Brand Rex [Type PXJ / PXMJ] PC/OC 2.29E+083 1.8E+08 2-152

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-3 Evaluation of Radiation Qualification of EQ Components (continued) Rooms EPU TID [rads] Qualification Component (Note 1) (Note 2) Dose [rads] Cable - Simplex Wire & Cable Co.[ Type CPSJ] OC 1.11E+07 1.7E+07 Cable - Triangle / PWC Inc. [Type PJJ] OC 1.62E+07 5.85E+07 Cable - Okonite Company [Type PX / PXJ / PXMJ] PC/OC 2.29E+083 1.81E+08 Cable - Rockbestos Company [Type SIS] OC 1.58E+08 1.8E+08 Cable - Okonite Company [Type EPSJ] OC 1.57E+08 1.8E+08 Cable - Rockbestos Company [Type PXJ / PXMJ] PC/OC 2.04E+083 1.80E+08 Cable - Rockbestos Company [Type MS] PC/OC 2.29E+083 1.80E+08 3 Cable - Rockbestos Company [Type Coaxial Cable] PC/OC 2.29E+08 1.84E+08 Cable - Okonite Company [Type PXJ / PXMJ] PC/OC 1.35E+08 1.81E+08 Cable - Anaconda Cable [Type MS] OC 1.11E+07 1.80E+08 Cable - ITT Surprenant Cable [Type MS] OC 7.44E+06 1.80E+08 Cable - American Insulated Wire [Type PXMJ] OC 1.70E+07 7.5E+07 Cable - Essex Cable [Type PXMJ] OC 1.70E+07 7.81E+07 Conduit Seal - CON/AX Corp. ECSA PC/OC 1.82E+08 2.25E+08 Conduit Seal - Rosemount Inc. 353C OC 1.62E+07 1.11E+08 Conduit Seal - EGS Quick Disconnect OC 1.85E+07 2.00E+08 Connector - CON/AX Coaxial Connector PC/OC 2.08E+08 2.25E+08 Handswitch - Cutler-Hammer 1025OT OC 5.19E+07 8.00E+07 Handswitch - General Electric CR2940 OC 1.62E+07 1.80E+07 Flow Switch - SOR 103AS OC 9.30E+06 3.30E+07 Flow Switch - SOR 141 Series OC 9.30E+06 3.00E+07 Flow/Level Switch - Fluid Components Inc. FR72-45A, OC 5.19E+07 5.81E+07 FR72-4HTR-DLL, FR72-1R Level Switch - Magnetrol 291 Series OC 3.18E+06 6.76E+06 Level Switch - Magnetrol 402 Series OC 1.34E+06 1.98E+08 Pressure Switch - SOR 5N/6N/12N OC 8.48E+06 3.00E+07 Pressure Switch - SOR Test Report 9058-102 OC 1.12E+07 3.30E+07 Temperature Element - Weed SP611-1A-A-3-C-2-75-4D4- PC/OC 2.29E+08 2.73E+08 2, 1B1D/612D-1A-C-6-C-17-00 2-153

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-3 Evaluation of Radiation Qualification of EQ Components (continued) Rooms EPU TID [rads] Qualification Component (Note 1) (Note 2) Dose [rads] Temperature Switch - EGS (Fenwal) 01-170230-090, OC 9.12E+06 5.05E+07 01-170020-090 Temperature Switch - SOR 201, 205 OC 1.43E+07 3.30E+07 Temperature Switch - SOR Test Report 9058-102 OC 9.30E+06 3.45E+07 Special Measure Transmitter - TEC VFMS2273A, 504A, PC 1.63E+08 2.22E+08 160-2 and 2273-C2 Limit Switch - N/AMCO EA-740 0C 1.70E+07 2.04E+08 Limit Switch - N/AMCO EA-180 OC 1.12E+07 2.04E+08 Limit Switch - Honeywell/Microswitch OC 8.95E+06 1.1E+07 OP-AR/OPD-AR/OPD-AR-30 Limit Switch - N/AMCO EA-740 (QTR-180) PC/OC 2.01E+08 2.04E+08 Motors - GE 5K6348XC23A and 5K6336XC198A OC 1.02E+06 1.46E+07 Motor - Reliance TEFC-XT Type P, Random Wound OC 9.30E+06 2.2E+08 Motors Motor - Reliance 4160VAC OC 9.30E+06 2.0 E+08 MOV - Limitorque AC/IPC SMB-000, 00, 0, 2, and SB-3 PC 1.97E+08 1.93E+08 2.27E+08 (MOVs components have various qualification doses and 2.33E+083 2.11E+08 TIDs. Listed in order are Fiberite, Phenolic, Motor 1.88E+083 2.04E+08 Insulation and Wiring and Splices) 1.80E+08 MOV - Limitorque AC/OPC SMB-000- Thru SMB-5T OC 1.85E+07 2.04E+08 MOV - Limitorque DC/OPC SMB-000, 00, 0, 2, 3, 4T and OC 1.05E+07(motors)3 1.0E+7 SB-0 (MOVs motor and all other components have a 1.17E+07 (motors) separate qualification doses and TIDs.) 2.11 E+08 EPA - Conax 7504-10001-03, -04, -05 and 7FO2-10000-01 PC/OC 1.0E+08 1.0E+08 EPA - GE 100 Series (Penetration Seals and Pigtails have PC/OC 4.94E+07 5.0E+07 separate qualification doses and TIDs) 7.70E+07 1.0E+08 EPA - GE Canister 238X600RH PC/OC 7.70E+07 8.3E+07 HVAC - Ellis-Watts ACH275.LC39 OC 6.95E+05 1.36E+07 8.37E+05 3.12E+07 Damper Motor Actuators - OC 2.87E+06 3.3E+06 Raymond Controls Sure 24-10-4 2-154

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-3 Evaluation of Radiation Qualification of EQ Components (continued) Rooms EPU TID [rads] Qualification Component (Note 1) (Note 2) Dose [rads] Damper Motor Actuators - Raymond Controls Sure 25 OC 5.25E+06 6.10E+06 4CW Solenoid - VALCOR V526-529-2 OC 1.85E+07 1.69E+08 Solenoid - ASCO NP Series, 206 Series PC/OC 1.92E+08 2.01E+08 Solenoid - Automatic Valve C-5497 PC/OC 1.36E+083 2.55E+07 Solenoid - Target Rock 81NN, 92Z OC 1.70E+07 3.35E+07 3 Solenoid - Target Rock 1/2 SMS-S-02-5 PC 1.45E+08 1.0E+07 Solenoid - AVCO Scram Solenoid Pilot Valve 0C 9.21E+04 2.49E+05 Splices - Raychem WCSF-N Series PC/OC 2.29E+083 2.0E+08 Splices - Raychem NPkV, NPKC, NPKP, NPKS, NMCK, PC/OC 2.11E+08 2.20E+08 NCBK, NESK Splices - Raychem NMCK8/NHVT OC 1.11E+07 5.0E+07 Raychem WCSF-N Series, NPkV, NPKC, NPKP, NPKS, PC/OC 2.04E+083 1.951E+08 NMCK, NCBK, NESK Raychem Nuclear High Voltage (5/8 kV) Splices OC 1.12E+07 2.15E+08 Terminal Block - General Electric CR151A, CR151B, EB-5, OC 1.57E+08 2.27E+08 EB-25 Transformer - BBC VPE OC 6.76E+05 8.7E+05 Pressure Transmitter - Rosemount 1153 B OC 1.05E+07 2.62E+07 Pressure Transmitter - Rosemount 1153D/1154/115 SERIES OC 1.79E+07 5.19E+07 Pressure Transmitter - Gould PD3200-100 Series, 400, OC 1.07E+06 5.55E+07 PDH3200-030 Pressure Transmitter - Weed DTN2010 OC 8.49E+06 1.10E+07 Notes:

1) PC indicates primary containment. OC indicates Outside Primary Containment.

2) The EPU TID is the sum of the normal dose, accident dose and no margin because the Browns Ferry radiation parameters were calculated using methods of Appendix D to NUREG 0588 Revision 1, per Section 1.4 of NUREG 0588 Revision 1, a 10% margin is not required.

3) The component has a limited life of less than 60 years due to EPU radiation and will be replaced periodically as part of the normal EQ maintenance program.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-4 Normal Maximum and Total Radiation Requirements for Rooms at Browns Ferry DBA Dose1 Total Integrated Dose1 Area Normal Operating Dose (Gamma + Beta) (Normal + Accident) LOCA + 60 yr + 40 Years 60 years2 60 yr +EPU LOCA 40 Years 60 years Room Description EPU EPU (RADS) (RADS) (RADS) (RADS) (RADS) (RADS) (RADS) (RADS) 0 Drywell El. 6.2E+07 9.3E+07 1.31E+08 2.47E+09 2.83E+09 2.54E+09 2.57E+09 2.97E+09 549.92 to 585.0 0 Drywell El. 6.2E+07 9.3E+07 1.31E+08 2.46E+09 2.83E+09 2.53E+09 2.56E+09 2.97E+09 585.0 to 617.0 0 Drywell El. 6.2E+07 9.3E+07 1.31E+08 2.44E+09 2.80E+09 2.51E+09 2.54E+09 2.94E+09 617.0 to 639.0 0 Drywell El. 6.2E+07 9.3E+07 1.31E+08 2.44E+09 2.80E+09 2.51E+09 2.54E+09 2.94E+09 639.0 and above 00 Wetwell 2.34E+06 3.51E+06 4.01E+06 2.43E+09 2.80E+09 2.44E+09 2.44E+09 2.81E+09 1 RX. Bldg. EL. 519.0 1.4E+04 2.1E+04 2.52E+04 1.67E+06 2.21E+06 1.69E+06 1.70E+06 2.24E+06 HPCI Room 2 RX. Bldg. EL. 519.0 1.8E+05 2.7E+05 3.15E+05 8.07E+06 8.91E+06 8.25E+06 8.34E+06 9.23E+06 Southwest Pump Room 3 RX. Bldg. EL. 519.0 1.5E+04 2.25E+04 2.70E+04 8.07E+06 8.91E+06 8.09E+06 8.10E+06 8.94E+06 Northwest Pump Room 4 RX. Bldg. EL. 519.0 3.9E+04 5.85E+04 6.93E+04 8.07E+06 8.91E+06 8.11E+06 8.13E+06 8.98E+06 Northeast Pump Room 2-156

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-4 Normal Maximum and Total Radiation Requirements for Rooms at Browns Ferry (continued) DBA Dose1 Total Integrated Dose1 Area Normal Operating Dose (Gamma + Beta) (Normal + Accident) LOCA + 60 yr + Room Description 40 Years 60 years2 60 yr +EPU LOCA 40 Years 60 years EPU EPU 5 RX. Bldg. EL. 519.0 2.1E+05 3.15+05 3.78E+05 8.07E+06 8.91E+06 8.28E+06 8.39E+06 9.29E+06 Southeast Pump Room 6A-D RX. Bldg. EL. 519.0 7.01E+05 1.06E+06 1.26E+6 1.33E+7 1.50E+07 1.40E+07 1.44E+07 1.63E+07 Torus Room 7 RX. Bldg. EL. 565.0 8.1E+06 1.22E+07 1.21E+07 5.37E+06 6.35E+06 1.35E+07 1.76E+07 1.85E+07 Main Steam Tunnel 8 RX. Bldg. EL. 565.0 7.0E+05 1.05E+06 1.26E+06 8.77E+06 9.79E+06 9.47E+06 9.82E+06 1.11E+07 General Floor Area 9A RX. Bldg. EL. 593.0 4.2E+05 6.30E+05 7.58E+05 5.97E+06 6.68E+06 6.39E+06 6.60E+06 7.44E+06 RHR Heat Exchanger Rooms 9B RX. Bldg. EL. 593.0 General Area 4.21E+05 6.32E+05 1.80E+06 5.97E+06 6.68E+06 6.40E+06 6.61E+06 7.48E+06 (Southwest Quadrant) 9C RX. Bldg. EL. 593.0 General Area 4.2E+05 6.30E+05 7.58E+05 5.97E+06 6.68E+06 6.39E+06 6.60E+06 7.44E+06 (Northwest Quadrant) 2-157

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-4 Normal Maximum and Total Radiation Requirements for Rooms at Browns Ferry (continued) DBA Dose1 Total Integrated Dose1 Area Normal Operating Dose (Gamma + Beta) (Normal + Accident) LOCA + 60 yr + Room Description 40 Years 60 years2 60 yr +EPU LOCA 40 Years 60 years EPU EPU 5 RX. Bldg. EL. 519.0 2.1E+05 3.15+05 3.78E+05 8.07E+06 8.91E+06 8.28E+06 8.39E+06 9.29E+06 Southeast Pump Room 9D RX. Bldg. EL. 593.0 General Area 4.2E+05 6.30E+05 7.58E+05 5.97E+06 6.68E+06 6.39E+06 6.60E+06 7.47E+06 (Northeast Quadrant) 9E RX. Bldg. EL. 593.0 General Area 4.21E+05 6.32E+05 1.80E+06 5.97E+06 6.68E+06 6.40E+06 6.61E+06 8.48E+06 (Southeast Quadrant) 10 RX. Bldg. EL. 593.0 9.2E+05 1.38E+06 1.58E+06 2.89E+05 6.60E+05 1.21E+06 1.67E+06 2.24E+06 RWCU Pump Rooms 11 RX. Bldg. EL. 593.0 Nonregenerative 6.1E+06 9.15E+06 1.04E+07 7.9E+05 1.22E+06 6.89E+06 9.94E+06 1.17E+07 Heat Exchanger Room 12 RX. Bldg. EL. 621.25 4.2E+05 6.3E+05 7.56E+05 3.47E+06 3.99E+06 3.89E+06 4.10E+06 4.75E+06 General Floor Area 2-158

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-4 Normal Maximum and Total Radiation Requirements for Rooms at Browns Ferry (continued) DBA Dose1 Total Integrated Dose1 Area Normal Operating Dose (Gamma + Beta) (Normal + Accident) LOCA + 60 yr + Room Description 40 Years 60 years2 60 yr +EPU LOCA 40 Years 60 years EPU EPU 13 RX. Bldg. EL. 639.0 2.6E+04 3.9E+04 4.73E+04 2.89E+05 6.60E+05 3.15E+05 3.28E+05 7.08E+05 General Floor (South) Area 14 RX. Bldg. EL. 639.0 7.0E+04 1.05E+05 1.26E+05 2.89E+05 6.60E+05 3.59E+06 3.94E+05 7.86E+05 General Floor (North) Area 15 RX. Bldg. EL. 664.0 1.1E+04 1.65E+04 2.23E+04 2.89E+05 6.60E+05 3.00E+06 3.06E+05 6.83E+05 Refueling Floor 16 RX. Bldg. EL. 593 RWCU Backwash 3.01E+08 4.52E+08 4.31E+08 2.89E+05 6.60E+05 3.02E+08 4.53E+08 4.32E+08 Receiving Tank Room 17A and B RX. Bldg. EL. 639.0 RWCU Note 3 Note 3 Note 3 2.89E+05 6.60E+05 2.89E+05 2.89E+05 6.60E+05 Demineralizers A and B 18 RX. Bldg. EL. 621.25 Cleanup 3.0E+08 4.50E+08 4.50E+08 2.89E+05 6.60E+05 3.01E+08 4.51E+08 4.51E+08 Demineralizer Valve Room 2-159

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-4 Normal Maximum and Total Radiation Requirements for Rooms at Browns Ferry (continued) DBA Dose1 Total Integrated Dose1 Area Normal Operating Dose (Gamma + Beta) (Normal + Accident) LOCA + 60 yr + Room Description 40 Years 60 years2 60 yr +EPU LOCA 40 Years 60 years EPU EPU RX. Bldg. EL. 565.0 19 7.4E+04 1.11E+05 1.26E+05 1.63E+07 1.84E+07 1.64E+07 1.65E+07 1.86E+07 Drywell Access Area RX. Bldg. EL. 565.0 20 1.2E+07 1.80E+07 2.02E+07 4.8E+05 8.70E+05 1.25E+07 1.85E+07 2.11E+07 Traversing In-core Probe (TIP) Room RX. Bldg. EL. 565.0 22 1.0E+03 1.50E+03 1.50E+03 1.43E+08 1.57E+08 1.44E+08 1.44E+08 1.58E+08 SGTS Building General Spaces 23 Stack Area 8.8E+03 1.32E+04 1.65E+04 8.45E+05 7.14E+05 8.54E+05 8.58E+05 7.31E+05 Notes: General Note - Unit 2 values used as representative for Units 1, 2, and 3. All three Drywells have the same source term and the Drywell doses are the same due to being identical in terms of operating power, Reactor Pressure Vessels and sacrificial shield concrete density. The measured normal dose rates in the Unit 2 and 3 portions of the Reactor Building and Standby Gas Treatment Building are the same and Unit 1 was confirmed to be less than Units 2 and 3. The Reactor Building gamma accident doses are the same for all three Units and the Beta accident doses are slightly higher on Unit 1 for some areas, but insignificant to the normal plus accident gamma doses.

1. The CLTP doses do not reflect the changes made to the EPU doses as a result of flow changes for SGTS.

2. 60 year normal dose calculated by multiplying the 40 year dose by 1.5.

3. Normal dose not calculated because there is no class 1E equipment located in this room.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-5 RWCU LOCA/HELB Temperature Evaluation Outside Containment Room Qualification Peak EPU Accident Room Description Component Number2 Limit (°F)1 Temperature (°F) MOV -DC Limitorque Outside Primary 6B RX. Bldg. EL. 519.0 Torus Room 340°F 135°F Containment 8 RX. Bldg. EL. 565.0 General Floor Area Limit switches ( N/AMCO, Honeywell) 308°F 175°F 8 RX. Bldg. EL. 565.0 General Floor Area Penetrations (Conax, GE) 340°F 175°F 8 RX. Bldg. EL. 565.0 General Floor Area Level Switches (SOR. FCI) 203°F 180°F 8 RX. Bldg. EL. 565.0 General Floor Area Connector (Conax) 445°F 180°F RX. Bldg. EL. 593.0 General Area MOV - AC Limitorque Outside Primary 9D 250°F 180°F (Northwest Quadrant) Containment RX. Bldg. EL. 593.0 General Area 9D Raychem Splices 358°F 180°F (Northwest Quadrant) RX. Bldg. EL. 593.0 General Area 9D Transmitters (Rosemount, Weed, Gould) 180°F 180°F (Northwest Quadrant) RX. Bldg. EL. 593.0 General Area 439°F 9D Solenoids ( Valcor, ASCO, AVCO, Target Rock) 535°F (Northwest Quadrant) (includes heat rise) RX. Bldg. EL. 593.0 General Area 9D Temperature Elements (Weed, Fenwal, SOR) 350°F 180°F (Northwest Quadrant) RX. Bldg. EL. 593.0 General Area 9D Conduit Seals (Conax, Rosemount, EGS) 375°F 180°F (Northwest Quadrant) RX. Bldg. EL. 593.0 General Area 9D Hand Switches (Cutler-Hammer, GE) 330°F 180°F (Northwest Quadrant) RX. Bldg. EL. 593.0 General Area 9D Pressure Switches (SOR) 227°F 180°F (Northwest Quadrant) RX. Bldg. EL. 593.0 General Area 9D Terminal Block (GE) 350°F 180°F (Northwest Quadrant) RX. Bldg. EL. 593.0 General Area 9D HVAC (Ellis-Watts) 215°F 180°F (Northwest Quadrant) 2-161

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-5 RWCU LOCA/HELB Temperature Evaluation Outside Containment (continued) Room Qualification Peak EPU Accident Room Description Component Number2 Limit (°F)1 Temperature (°F) RX. Bldg. EL. 593.0 General Area 9D Flow Switch (SOR) 325°F 180°F (Northwest Quadrant) RX. Bldg. EL. 621.25 General Floor Area 380°F 12 Transformer (Brown Boveri) 447°F (includes heat rise) 16 RX. Bldg. EL. 593 RWCU Backwash Cable -Various Vendors 215°F 205°F3 Receiving Tank Room [Unit 2] ( for ~60 seconds) Notes:

1. Lowest qualification peak LOCA/HELB temperature for the component type.

2. Worst case EQ Location (Room)

3. DOR cable is worst case limited to 205°F long term peak temperature but may exceed 205°F for approximately 10 minutes as long it is under a maximum temperature of 300°F (TVA Letter to the NRC Dated 05/11/1989 (Docket no. 50-250)).

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-6 Offsite Electrical Equipment Ratings and Margins CLTP EPU Component CLTP Component Margin EPU Duty Margin Rating Duty (%) (%) Main Generator 1,330/0.95 U1 1,280/0.90 U1 3.8 U1 1,330/0.95 U1 0.0 (MVA Capability 1,332/0.93 U2 1,280/0.90 U2 3.9 U2 1,332/0.93 U2 0.0 /power factor) 1,332/0.93 U3 1,280/0.90 U3 3.9 U3 1,332/0.93 U3 0.0 Isolated Phase Bus 36,740 U1 35,359 U1 3.8 U1 36,740 U1 0.0 Maximum Continuous 36,796 U2 (Note 1) 35,359 U2 3.9 U2 36,796 U2 0.0 Current (Amps) 36,796 U3 35,359 U3 3.9 U3 36,796 U3 0.0 Main Generator Step- 1,500 U1 1,150 U1 23.3 U1 1,280 U1 14.7 Up Transformers 1,500 U2 1,150 U2 23.3 U2 1,282 U2 14.5 (MVA) 1,500 U3 1,150 U3 23.3 U3 1,282 U3 14.5 (Note 2) (Note 3) Unit Station Service 24/32/40 MVA 27.34 U1 31.65 Transformers (MVA) OA/FA/FOA 27.61 U2 30.98

                             @55°C                                          22.94 U3    42.65 (Note 4)

Unit Station Service 24/32 MVA OA/FA 19.91 U1 37.78 Transformers (MVA) @55°C 18.20 U2 43.13 20.41 U3 36.22 (Note 4) Common Station 21.9/29.2/36.5 36.15 0.96 Service Transformers MVA OA/FA/FOA 36.15 0.96 (A and B) (MVA) @55°C (Note 5) Notes:

1. The Unit 2 isophase bus forced cooling system has the capability to remove the heat generated from 36,796 amps, which is based on an original rating of 36,740 amps with 45,000 scfm of cooling flow, and testing which demonstrated an increased cooling capability based on a cooling flow of 55,000 scfm with two fans operating.

2. Based on maximum historical transformer loading data.

3. Based on maximum generator output minus auxiliary loads of 50MVA.

4. Based on normal loading.

5. Based on maximum shutdown loading.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-7 Electrical Distribution System Load Changes Max. Analyzed Required BHP Motor Description Nameplate hp BHP EPU EPU Condensate Pumps (U1) 1,250 1,025 (1) 1,212.5 (2) Condensate Pumps (U2) 1,250 1,025 (1) 1,212.5 (2) Condensate Pumps (U3) 1,250 1,025 (1) 1,212.5 (2) Condensate Booster Pumps (U1) 3,000 2,470.5 (1) 3,720 (2) Condensate Booster Pumps (U2) 3,000 2,470.5 (1) 3,720 (2) Condensate Booster Pumps (U3) 3,000 2,470.5 (1) 3,720 (2) Reactor Recirculation Pumps (U1) 8,657 8,657 (3) 8,657 (3) Reactor Recirculation Pumps (U2) 8,657 8,657 (3) 8,657 (3) Reactor Recirculation Pumps (U3) 8,657 8,657 (3) 8,657 (3) Notes:

1. Normal operation at EPU RTP with three condensate and three condensate booster pumps in service.

2. Maximum transient load assuming either a trip of one condensate pump or one condensate booster pump.

3. Electrical analyses at EPU conditions assume reactor recirculation pump motors operate at nameplate hp.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-8a Key Inputs for Browns Ferry Station Blackout Parameter(1) Value Initial Reactor Power 3,952 MWt Initial Reactor Pressure 1,055 psia Decay Heat ANS/ANSI 5.1 1979 standard consistent with recommendations of GEH SIL 636 Initial Suppression Pool 95°F Temperature Initial Suppression Pool Volume 122,940 ft3 Low Water Level (LWL) Initial Wetwell Pressure 14.4 psia Initial Drywell Temperature 150°F Initial Drywell Pressure 15.5 psia Initial Drywell free airspace 171,000 ft3 volume Initial Wetwell free airspace 135,000 ft3 volume Initial wetwell (WW) airspace 95°F temperature CST Water Temperature 130°F CST Inventory 135,000 gallons available Initial Drywell Relative 20% Humidity Initial Wetwell Relative 100% Humidity RHR Heat exchanger K factor 265 BTU/Sec-°F (per heat exchanger) RHR pump flow rate (per pump) 6500 GPM RHR service water flow rate to 4000 gpm RHR heat exchangers RHR service water temperature 95°F Leakage rate from primary 2% of containment air mass per day containment Containment heat sinks modeled Yes (1) RPV volume, related masses, and wetwell to drywell vacuum breakers are provided in Table 2.6-2a. (2) Containment heat sinks are modeled in the SBO evaluation using the EPU values shown in Table 2.6-6a. 2-165

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.3-8b Browns Ferry Station Blackout Sequence of Events Browns Ferry Station Blackout Sequence of Events for EPU Time (sec) Description

       ~0          Loss of Offsite Power Reactor scram MSIV start to close Loss of Feedwater RCIC available to maintain reactor water level HPCI available to maintain reactor water level
     ~4.0          MSIV closed
       ~5          FW flow stops

~10 to 256 MSRVs open (relief mode) 267 Begin HPCI Injection (high drywell pressure) 641 End HPCI Injection (Level 8) 1,200 Begin manual MSRV operation for RPV pressure control 2,683 Begin RCIC Injection (Level 2) 7,320 Manual MSRV cooldown complete. 11,194 End RCIC Injection (Level 8) 14,400 Offsite power restored (end of coping period) Containment cooling initiated with 2 RHR pumps/2 RHR heat exchangers in SP cooling mode (RHR flow of 6500gpm/pump; RHR heat exchanger K-factor of 265 BTU/Sec-°F) 2-166

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Figure 2.3-1 Worst Case Drywell Temperature Profile 2-167

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Figure 2.3-2 Worst Case Secondary Containment EQ Temperature Profile (Note the bounding CLTP and EPU temperature profiles are the same) 2-168

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.4 Instrumentation and Controls 2.4.1 Reactor Protection, Safety Features Actuation, and Control Systems Regulatory Evaluation Instrumentation and control systems are provided (1) to control plant processes having a significant effect on plant safety, (2) to initiate the reactivity control system (including control rods), (3) to initiate the engineered safety features (ESF) systems and essential auxiliary supporting systems, and (4) for use to achieve and maintain a safe shutdown condition of the plant. Diverse instrumentation and control systems and equipment are provided for the express purpose of protecting against potential common-mode failures of instrumentation and control protection systems. The NRCs acceptance criteria related to the quality of design of protection and control systems are based on 10 CFR 50.55a(a)(1), 10 CFR 50.55a(h), and GDCs-1, 4, 13, 19, 20, 21, 22, 23, and 24. Specific NRC review criteria are contained in SRP Sections 7.0, 7.2, 7.3, 7.4, 7.7, and 7.8. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-1, 12, 13, 14, 15, 19, 20, 22, 23, 25, 26, 40, and 42. Final GDC-19 is applicable to Browns Ferry. 2-169

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry instrumentation and control systems are described in Browns Ferry UFSAR Section 7, Control and Instrumentation. Browns Ferrys instrumentation and control systems were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The instrumentation and control systems were determined to be within the scope of license renewal and the components subject to aging management review are evaluated on a plant wide basis as commodities. The electrical commodity groups are described in NUREG-1843, Section 2.5, and aging management for electrical commodities is described in NUREG-1843, Section 3.6. Technical Evaluation The setpoint calculation methodology, safety limit-related Limiting Safety System Setting (LSSS) determination, and instrument setpoint controls are discussed in this section. NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 5 of the CLTR addresses the effect of EPU on Reactor Protection, Safety Features Actuation, and Control Systems. The results of this evaluation are described below. 2.4.1.1 Nuclear Steam Supply System Monitoring and Control Instrumentation As stated in Section 5.1 of the CLTR, the instruments and controls used to monitor and directly interact with or control reactor parameters are usually within the NSSS. Changes in process variables and their effects on instrument performance and setpoints were evaluated for EPU operation to determine any related changes. Process variable changes are implemented through changes in normal plant operating procedures. TSs address instrument AVs and/or setpoints for those NSSS sensed variables that initiate protective actions. The effects of EPU on TS instrument functions are addressed in Section 2.4.1.3. The EPU affects the performance of the Neutron Monitoring System. These performance effects are associated with the Average Power Range Monitors (APRMs), Local Power Range Monitors (LPRMs), Intermediate Range Monitors (IRMs), and Source Range Monitors (SRMs). 2-170

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition Browns Ferry Result Meets CLTR APRM, IRM, SRM Generic Disposition Meets CLTR Local Power Range Monitors Generic Disposition Meets CLTR Rod Block Monitor Generic Disposition Meets CLTR Rod Worth Minimizer Generic Disposition 2.4.1.1.1 Average Power Range, Intermediate Range and Source Range Monitors The CLTR states that at rated power, the increase in power level increases the average flux in the core and at the in-core detectors. The Average Power Range Monitor (APRM power signals are calibrated to read 100% at the new licensed power (i.e., EPU RTP). The Intermediate Range Monitors (IRM)s provide full overlap with the APRMs. The APRM, IRM, and source range monitor (SRM) systems installed at Browns Ferry are in accordance with the requirements established by the GEH design specifications. The specifications provide confirmation that the APRM, IRM, and SRM systems meet all CLTR dispositions. 2.4.1.1.2 Local Power Range Monitors The CLTR states that at rated power, the increase in power level increases the flux at the LPRMs. The average flux experienced by the detectors increases due to the average power increase in the core. The maximum flux experienced by an Local Power Range Monitor (LPRM) remains approximately the same because the peak bundle power does not increase. Due to the increase in neutron flux experienced by the LPRMs and traversing incore probes (TIPs), the neutronic life of the LPRM detectors may be reduced and radiation levels of the TIPs may be increased. LPRMs are designed as replaceable components. The LPRM accuracy at the increased flux is within specified limits, and LPRM lifetime is an operational consideration that is handled by routine replacement. TIPs are stored in shielded rooms. A small increase in radiation levels is accommodated by the radiation protection program for normal plant operation. 2-171

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Reliability of LPRM instrumentation and accurate prediction of in-bundle pin powers typically requires operation with bypass voids lower than 5% at nominal conditions. LAR Attachment 34 concludes that bypass voiding does not exceed 5% for any LPRMs. The LPRMs installed at Browns Ferry are in accordance with the requirements established by the GEH design specifications. The specifications provide confirmation that the LPRMs meet all CLTR dispositions. 2.4.1.1.3 Rod Block Monitor The CLTR states that the increase in power level at the same APRM reference level results in increased flux at the LPRMs that are used as inputs to the rod block monitor (RBM). The RBM instrumentation is referenced to an APRM channel. Because the APRM has been rescaled, there is only a small effect on the RBM performance due to the LPRM performance at the higher average local flux. The change in performance does not have a significant effect on the overall RBM performance. The RBMs installed at Browns Ferry are in accordance with the requirements established by the GEH design specifications. The specifications provide confirmation that the RBMs meet all CLTR dispositions. 2.4.1.1.4 Rod Worth Minimizer The assessment of the RWM is provided in FUSAR Section 2.4.1.1.4. 2.4.1.2 BOP Monitoring and Control As stated in Section 5.2 of the CLTR, operation of the plant at EPU conditions has minimal effect on the BOP system instrumentation and control devices. Based on EPU operating conditions for the power conversion and auxiliary systems, most process control valves and instrumentation have sufficient range/adjustment capability for use at the EPU conditions. However, some (non-safety) modifications may be needed to the power conversion systems to obtain EPU RTP. Browns Ferry meets all CLTR dispositions. The topics considered in this section are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Pressure Control System (PCS) Generic Disposition Meets CLTR Turbine Steam Bypass System (Normal Operation) Generic Disposition Meets CLTR Turbine Steam Bypass System (Safety Analysis) Generic Disposition 2-172

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR FW Control System (Normal Operation) Generic Disposition Meets CLTR FW Control System (Safety Analysis) Generic Disposition Meets CLTR Leak Detection System (LDS) Generic Disposition 2.4.1.2.1 Pressure Control System The CLTR states that the increase in power level increases the steam flow to the turbine. The PCS is a normal operating system to provide fast and stable responses to system disturbances related to steam pressure and flow changes to control reactor pressure within its normal operating range. This system does not perform a safety function. Pressure control operational testing is included in the EPU implementation plan as described in Section 2.12 to ensure adequate turbine control valve pressure control and flow margin is available. The PCS at Browns Ferry meets all CLTR dispositions. 2.4.1.2.2 Turbine Steam Bypass System The CLTR states that the bypass system capacity, in terms of mass flow, is not changed for EPU. As a result, the increase in power level and resulting increase in steam flow to the turbine effectively reduces the bypass system capacity in terms of percent steam flow. The turbine bypass system is not essential for turbine operation and is not credited in any limiting events analyses as discussed in Section 2.8.5. The Turbine Steam Bypass System is a normal operating system that is used to bypass excessive steam flow. This system is non-safety-related. The flow capacity of the bypass system, 3.5 Mlbm/hr, is not changed. ((

                                                     )) The AOO events are discussed further in Section 2.8.5.

The Turbine Steam Bypass System at Browns Ferry meets all CLTR dispositions. 2.4.1.2.3 Feedwater Control System The CLTR states that the increase in power results in an increase in FW flow. The FW Control System is a normally operating system to control and maintain the reactor vessel water level. EPU results in an increase in FW flow. FW control operational testing is 2-173

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) included in the EPU implementation plan as described in Section 2.12 to ensure that the FW response is acceptable. Failure of this system is evaluated in the reload analysis for each reload core with the FW controller failure-maximum demand event. An Loss of Feedwater (LOFW) event can be caused by downscale failure of the controls. The LOFW is discussed in Section 2.8. The FW Control System at Browns Ferry meets all CLTR dispositions ((

))

2.4.1.2.4 Leak Detection System The CLTR states that the only effect on the LDS due to EPU is a slight increase in the FW system temperature and increase in the steam flow. ((

))

Main Steam Tunnel in the Reactor Building and in the Turbine Building: The increased FW temperature results in a negligible increase (< 1°F) in the MS tunnel temperature. The LDS temperature and differential temperature setpoints remain unchanged for EPU. As a result, the MS tunnel temperature setpoint is conservative because it slightly increases leak detection sensitivity and is not changed. Drywell: The normal operating drywell area temperature experiences a negligible change for EPU conditions; therefore, the DW LDS is not affected. RWCU: There is no significant change to the RWCU system temperature and pressure and no change to the RWCU system flow; therefore, the RWCU LDS is not affected. RCIC: There is no increase in the system temperature, pressure, or flow; therefore, the RCIC LDS is not affected. RHR Shutdown Cooling Mode: There is no increase in the RHR Shutdown Cooling Mode temperature or pressure; therefore, the RHR system LDS is not affected. HPCI: There is no increase in the system temperature, pressure, or flow; therefore, the HPCI LDS is not affected. The flow-based LDS is not affected by EPU, with the exception of MSL high flow. MSL high flow is discussed in Section 2.4.1.3. The LDS at Browns Ferry meets all CLTR dispositions ((

))

2.4.1.3 Technical Specification Instrument Setpoints As stated in Section 5.3 of the CLTR, Allowable Values (AVs) and/or Nominal Trip Setpoints (setpoints) are those sensed variables which initiate protective actions and are generally associated with the safety analysis. AVs are highly dependent on the results of the safety 2-174

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) analysis. The safety analysis generally establishes the ALs. The AVs and other instrument setpoints include consideration of measurement uncertainties and are derived from the ALs. The settings are selected with sufficient margin to minimize inadvertent initiation of the protective action, while assuring that adequate operating margin is maintained between the system settings and the actual limits. There is typically substantial margin in the safety analysis process that should be considered in establishing the setpoint process used to establish the Technical Specification AVs and other setpoints. Increases in the core thermal power and steam flow affect some instrument setpoints. These setpoints are adjusted to maintain comparable differences between system settings and actual limits, and reviewed to ensure that adequate operational flexibility and necessary safety functions are maintained at the EPU RTP level. Where the power increase results in new instruments being employed, an appropriate setpoint calculation is performed and TS and/or Technical Requirement Manual (TRM) changes are implemented, as required. ((

))

((

))

Browns Ferry has elected not to use the simplified methodology and has applied the existing GE methodology, Reference 57, for the APRM and RBM setpoint functions and the existing TVA methodology, Reference 58, for all other setpoint functions to the Technical Specification instrument setpoints. The TVA setpoint methodology (Reference 58) has been accepted by the NRC for setpoint calculations as referenced in the NRC safety evaluation report for Browns Ferry technical specification change, TS-453 (Reference 59). The GE setpoint methodology (Reference 57) used in the setpoint calculations for the neutron monitoring system functions affected by EPU (e.g., APRM and RBM) was also used during the initial licensing and subsequent NRC approval for installation of the power range neutron monitoring (PRNM) system for each respective Browns Ferry unit. All Technical Specification instruments were evaluated for effects from EPU. This evaluation included a review of environmental (i.e., radiation and temperature) effects, process (i.e., measured parameter) effects and analytical (i.e., AL and margins) effects on the subject instruments. Table 2.4-1 summarizes the current and EPU ALs for Browns Ferry. The setpoint calculation methodology for the Browns Ferry EPU is not per the generic disposition of the CLTR because TVA has elected not to use the simplified methodology stated in the CLTR. The setpoint value for each topic addressed in this section meet all CLTR dispositions. The topics considered in this section are: 2-175

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Full setpoint calculation Main Steam Line High Flow Isolation - Setpoint Generic performed with Calculation Methodology Reference 58 methodology. Main Steam Line High Flow Isolation - Setpoint Meets CLTR Plant Specific Value Disposition Full setpoint calculation Turbine First-Stage Pressure Scram and performed with Recirculation Pump Trip Bypass - Setpoint Generic Reference 58 Calculation Methodology methodology due to HP turbine replacement. AL revised using guidelines of Turbine First-Stage Pressure Scram and Plant Specific Section F.4.2.3 of Recirculation Pump Trip Bypass - Setpoint Value Reference 4 (ELTR1). Full setpoint calculation APRM Flow Biased Scram - Setpoint Calculation Generic performed with Methodology Reference 57 methodology. Meets CLTR APRM Flow Biased Scram - Setpoint Value Plant Specific Disposition Full setpoint calculation Rod Worth Minimizer Low Power Setpoint - Generic performed with Setpoint Calculation Methodology Reference 58 methodology. Rod Worth Minimizer Low Power Setpoint - Meets CLTR Plant Specific Setpoint Value Disposition Meets CLTR Rod Block Monitor Generic Disposition 2-176

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Full setpoint calculation APRM Setdown in Startup Mode - Setpoint Generic performed with Calculation Methodology Reference 57 methodology. Meets CLTR APRM Setdown in Startup Mode - Setpoint Value Plant Specific Disposition 2.4.1.3.1 Main Steam Line High Flow Isolation The CLTR states that the effect on the Main Steam Line High Flow Isolation due to EPU is increased reactor power level and steam flow. The MSL high flow isolation setpoint is used to initiate the isolation of the Group 1 primary containment isolation valves. The only safety analysis event that credits this trip is the MSLB accident. For this accident, there are diverse trips from high area temperature and high area differential temperature in the main steam tunnel. For Browns Ferry, there is sufficient margin to choke flow, so the AL for EPU is unchanged from the current percent of rated steam flow (144% rated steam flow) in each MSL. No new instrumentation is required (the existing instrumentation has the required upper range limit and calibration span the instrument loops need to accommodate the new setpoint). A new setpoint was calculated using the Reference 60 methodology and an AV change is required to change the differential pressure at the allowable steam flow. The MSL AL to choke flow margin calculation was performed in accordance with the methodology specified in GEH Services Information Letter (SIL) No. 438 Revision 2 (Reference 61) and communicated to utilities in GEH 10 CFR Part 21 Safety Communication (SC) 12-18 Revision 2 (Reference 62). The Browns Ferry plant-specific EPU MSL flow element choke flow to MSL high flow AL margin evaluation incorporates the resolution of the GEH 10 CFR Part 21 issue (Reference 42). Therefore, the Main Steam Line High Flow Isolation setpoint meets all CLTR plant-specific dispositions. 2.4.1.3.2 Turbine First-Stage Pressure Scram and Recirculation Pump Trip Bypass The CLTR states that the effect on the turbine first stage pressure (TFSP) Scram and RPT Bypass Permissive due to EPU is increased reactor power level and a potential change to TFSP. EPU results in an increased power level, and the HP turbine modifications result in a change to the relationship of TFSP to reactor power level. The TFSP setpoint is used to reduce scrams and RPTs at low power levels where the turbine bypass system (TBS) is effective for TTs and generator load rejections. In the safety analysis, this trip bypass only applies to events at low power levels that result in a turbine trip (TT) or load rejection. Based on the guidelines in 2-177

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Section F.4.2.3 of Reference 4 (ELTR1), the TFSP Scram and RPT Bypass Permissive AL in percent RTP is reduced from 30% at CLTP to 26% at EPU. ((

))

((

                          )) Therefore, a new setpoint is calculated using the TVA methodology per Reference 60, and ((                                                                     )) The AV (in psig) for Browns Ferry is revised prior to EPU implementation.

To assure that the new value is appropriate, an EPU plant ascension startup test or normal plant surveillance is performed to validate that the actual plant interlock is cleared consistent with the safety analysis. EPU startup testing is described in Section 2.12. Therefore, the TFSP Scram and RPT Bypass Permissive meet all CLTR plant specific dispositions. 2.4.1.3.3 APRM Flow Biased Simulated Thermal Power - High Scram This function is referred to in the Browns Ferry TSs as the APRM Flow-Biased Simulated Thermal Power (STP) - High function. The CLTR states that the effect on the APRM Flow Biased Scram due to EPU is increased reactor power level. APRM Simulated Thermal Power - High function provides protection against transients where Thermal Power increases slowly and protects the fuel cladding integrity by ensuring that the minimum critical power ratio (MCPR) safety limit is not exceeded. This operating limit for the operating domain is established to provide a pre-emptive scram and to prevent a gross violation of the licensed domain. The Browns Ferry AL for this function is being revised based on the methodology outlined in the CLTR. Therefore, a new setpoint was calculated using the TVA methodology per Reference 58, and (( )) The AV (in %RTP) for Browns Ferry will be revised prior to EPU implementation. The clamped AL will retain its value in percent power. Therefore, APRM Flow-Biased Scram at Browns Ferry meets all of the CLTR dispositions. 2.4.1.3.4 Rod Worth Minimizer Low Power Setpoint The AL in terms of percent RTP does not change, and ((

                             )) Browns Ferry Technical Specifications do not define an AV for this setpoint function.

The CLTR states that the effect on the RWM Low Power Setpoint (LPSP) due to EPU is increased reactor power level and increased FW flow. For the RWM LPSP instrument function at Browns Ferry, the measurement parameter is main steam and feedwater flow. The RWM LPSP is used to bypass the rod pattern constraints established for the Control Rod Drop Accident (CRDA) at greater than a pre-established low power level. 2-178

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Therefore, the RWM LPSP calculation methodology meets all CLTR dispositions. The LPSP AL is maintained at the same value in terms of percent power (10% RTP) and the EPU has been evaluated on this basis. Below this setpoint, only banked position mode withdrawals or insertions are allowed. Therefore, the RWM LPSP meets all CLTR plant specific dispositions. The LPSP measurement parameter is main steam flow and feedwater flow. The existing steam flow and feed flow measurement transmitters have sufficient range for EPU rated steam flow and feed flow. A new setpoint in terms of rated steam flow and feedwater flow was calculated using the Reference 58 methodology. 2.4.1.3.5 Rod Block Monitor The generic disposition of the Rod Block Monitor in the CLTR states that the effect on the Rod Block Monitor due to EPU is increased reactor power level. Consistent with the generic disposition discussed above, the severity of a rod withdrawal error (RWE) during power operation event is dependent upon the RBM rod block setpoint. This setpoint is only applicable to the control rod withdrawal error. ((

))

2.4.1.3.6 APRM Setdown in Startup Mode This function is referred to in the Browns Ferry TSs as the APRM Neutron Flux - High, Setdown function. The CLTR states that the effect on the APRM Setdown in Startup Mode due to EPU is a reduced TS safety limit for reduced pressure or low core flow conditions. No specific safety analyses take direct credit for this function. It indirectly ensures that reactor power does not exceed 23% RTP before the Mode Switch is placed in "RUN." The APRM setdown in the startup mode provides margin to the safety limit. The value for the TS safety limit for reduced pressure or low core flow condition is established to satisfy the fuel thermal limits monitoring requirements. The Browns Ferry AL for this function will change due to the EPU based on the methodology outlined in the CLTR. Therefore, a new setpoint was calculated using the TVA methodology per Reference 58, and the TS applicable condition for fuel thermal limits monitoring requirements in % RTP has been changed. The AV (in % RTP) for Browns Ferry will be revised prior to EPU 2-179

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) implementation. Therefore, APRM Setdown in Startup Mode at Browns Ferry meets the CLTR disposition. 2.4.1.3.7 Main Steam Line Low Pressure Isolation in the Run Mode The PCS (see Section 2.4.1.2.1) pressure setpoint does not change in a power uprate. However, the steam line pressure near the turbine, where this sensor is located, is expected to change. The margin assessment confirmed that the remaining margin at EPU conditions will not impose any new constraints on the performance of surveillances with the existing setpoint. The margin assessment is performed as an operational screening check to ascertain the potential of normal turbine surveillances (individual stop and control valve full stroke closure at power) to cause pressure drops that could actuate the trip instrumentation. GE SIL 130 provides the criterion applied. SIL 130 provided a basis for reducing the margin down to a minimum of 100 psi to avoid spurious steam line isolations in the event of plant scrams. The margin for the projected uprate conditions is 125.3 psid. The MSL low pressure isolation AL, AV and nominal trip setpoint (NTSP) are unchanged for EPU. 2.4.1.4 Changes to Instrumentation and Controls In the CLTR SER, the staff requested that the plant-specific submittal address all EPU-related changes to instrumentation, such as scaling changes, changes to upgrade obsolescent instruments, and changes to the control philosophy. Table 2.4-2 provides this information. The instrument modifications described in Table 2.4-2 that have not been completed to date will be completed prior to EPU operation. Conclusion TVA has evaluated the effects of the proposed EPU on the functional design of the reactor trip system, safe shutdown system, and control systems. The evaluation indicates that Browns Ferry will continue to meet the requirements of 10 CFR 50.55a(a)(1), 10 CFR 50.55a(h), final GDC-19, and draft GDCs-1, 12, 13, 14, 15, 19, 20, 22, 23, 25, 26, 40, and 42. Therefore, the proposed EPU is acceptable with respect to instrumentation and controls. 2-180

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.4-1 Technical Specification Setpoint Information Values Parameter Current EPU APRM Calibration Basis (MWt) 3458 3952 APRM High Flux Scram AL (% RTP) 125.4 No Change2 APRM Simulated Thermal Power - High DLO AL1 (% RTP) 0.66W + 68.0 3 0.55W + 67.5 3 SLO AL1 (% RTP) 0.66(W-W) + 68.0 3,4 0.55(W -W) + 67.5 3,4 Clamp AL1 (% RTP) 122 No Change2 APRM Rod Block AL DLO Flow Biased(1) (% RTP) 0.66W + 64 0.55W + 63.5 SLO Flow Biased(1) (% RTP) 0.66(Wd-W) + 64 0.55(Wd -W) + 63.5 Clamp AL (% RTP) 118 No Change APRM Neutron Flux - High Setdown (% RTP) Scram AL 25 23 Rod Block (APRM Upscale (Startup) AL (% 15 13 RTP)) Rod Block Monitor ALs Low Power Setpoint (Enable) (% RTP) 30 No Change Intermediate Power Setpoint (% RTP) 65 No Change High Power Setpoint (% RTP) 85 No Change Low Trip Setpoint (% Reference Level) 121.0 No Change 5 Intermediate Trip Setpoint (% Reference Level) 116.0 No Change 5 High Trip Setpoint (% Reference Level) 111.0 No Change 5 2-181

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.4-1 Technical Specification Setpoint Information (continued) Values Parameter Current EPU RWM LPSP AL (% RTP) 10 No Change6 Main Steam Line High Flow Isolation (% rated 144 144 steam flow) AL Turbine First-Stage Pressure Scram and Recirculation Pump Trip Bypass (% RTP) AL 30 26 Reactor Vessel Water Level - Low, Level 3 Scram (inches Above Vessel Zero (AVZ)) AL 518 No Change7 Main Steam Line Low Pressure Isolation (in RUN 825 No Change8 Mode) Allowable Value Trip Setpoint, psig Notes:

1. No credit is taken in any safety analysis for the flow referenced setpoints.

2. The EPU APRM Neutron Flux - High Scram, APRM Simulated Thermal Power - High Clamp and APRM Neutron Flux - High Setdown remain the same in terms of percent rated thermal power.

3. W is the Recirculation Drive Flow in percent of Rated flow. W is the difference between the dual loop operation (DLO) and SLO drive flow at the same core flow. The current value of W is 10% and is not changed.

4. The ALs for SLO operation are unchanged in terms of MWt.

5. The cycle-specific reload analysis is used to determine any change in the rod block trip setpoint. The RBM trip setpoints listed are based on an OLMCPR of 1.25. The trip setpoints corresponding to other OLMCPR values also would remain the same for EPU.

6. The EPU RWM LPSP remains the same in terms of percent rated thermal power.

7. The AL, AV and NTSP are not changed for EPU for this setpoint function. EPU satisfies the issue with Steam Flow Induced Error (SFIE; also called Bernoulli error) in the case that the steam dryer skirt becomes uncovered for a loss of feedwater flow transient, per the related safety communication SC04-14 (Reference 63).

8. The MSL Low Pressure Isolation (in RUN Mode) Actual Trip Setpoint, is 843 psig and is unchanged for EPU.

2-182

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.4-2 Changes to Instrumentation and Controls Parameter EPU Change5 Implementation Status3 U1 U2 U3 MSL High Flow Rescale loops and revise setpoints for AL of 144% N N N rated steam flow (PDT/PDIS-1-13A-D,-25A-D,- 36A-D,-50A-D)1 MSL Flow Rescale loops (FT-1-13,-25,-36,-50)2 Y Y N Turbine 1st Stage Pressure (Scram Recalibrate for AL of 26.25% RTP N N N Bypass Permissive) (PT-1-81A/B,-91A/B) Turbine 1st Stage Pressure Rescale loop (PT-1-81)2 Y Y N Turbine Exhaust Intermediate Pressure Rescale loop (PT-1-100) Y Y N APRM Flow Biased STP Scram Recalibrate (APRM-92-1,-2,-3,-4) N N N APRM Flow Biased STP Rod Block Recalibrate N N N APRM Scram Setdown Recalibrate N N N APRM Rod Block Setdown Recalibrate N N N RBM Power Dependent Setpoints Recalibrate per cycle-specific reload analysis N N N RWM LPSP No change for EPU - - - Digital FW Controls Software Update RWM alarm/enable setpoints and DFWCS N N N parameters4 RFW Line A/B Flow Rescale loops (FT-3-78A,-78B)2 Y Y N RFW Pump A/B/C Suction Pressure Replace pressure gauge with new range Y Y N (PI-2-121,-122,-123) RFW Pump A/B/C Low Suction Recalibrate for new switch setpoints Y Y N Pressure Alarm/Trip (PS-2-121A/B,-122A/B,-123A/B) RFPT LP Steam Inlet Flow Rescale loops (FT-1-117, -120) Y Y N RFPT Low Condenser Vacuum Trip Added pressure switches and revise trip logic to Y Y Y 2-out-of-3 (PS-3-200A/B/C,-201A/B/C,-202A/B/C) RFP Discharge Flow Rescale loops (FT-3-20,-13,-6)2 Y Y N RFP Seal Injection DP Replace DP controller and revise setpoints Y Y N (PDC-3-80,-82,-84,-86,-88,-90) RRS Jet Pump Head Recalibrate loop (PDT/PDI-3-51) Y Y Y RCW Flow from RRS VFD HX A/B Rescale loops (FIT-24-182,-187) Y Y Y RRS Pump A/B Winding Temperature Revise high alarm setpoints (TA-68-58,-84) Y Y Y RRS VFD A/B Protective Relays Revise protective relay setting (MMR, DFR) Y Y Y 2-183

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.4-2 Changes to Instrumentation and Controls (continued) Parameter EPU Change5 Implementation Status3 U1 U2 U3 RRS Pump Motor Controls Software Revise upper power runback steam flow setpoint N N N (RRMST:UPWRSPT) SPE Bypass Line Flow Rescale flow indicator loop (FT/FI-2-42) Y NA NA SJAE A/B Trip and Standby Auto Start Modified logic to remove trip on low condenser Y Y Y vacuum and eliminate auto start of standby SJAE (PS-2-5B,-8B) SJAE A/B Condensate Pressure Revise pressure switch setpoints to prevent Y Y N inadvertent SJAE isolation (PS-2-34,-40) SJAE Steam Supply Stage I, II, III Revise low steam supply pressure isolation switch Y Y Y Pressure setpoints (PS-1-150, -152, -166, -167) Condensate Pump Discharge Header Replace flow transmitter, controller, recorder and Y Y N (HDR) Flow rescale loops (FT/FC-2-29 and XR-2-26)8 Condensate Pump Motor Current Replace motor ammeters in Main Control Room Y Y Y (MCR) (EI-2-26,-21,-15) and at 4 kV unit board (EI-2-26/8, -21/7, -15/5)7 with increased range Condensate Pump Breakers Revise breaker relay trip settings (U-8, U-7, U-5)9 Y Y Y Condensate Pump Motor Revise high temperature alarm setpoints Y Y Y Stator/Bearing Temperature (TE-2-25A-H and J, -20A-H and J, -14A-H and J) Condensate Booster Pump Motor Replace ammeters in MCR (EI-2-56,-62,-68) and Y Y N Current at 4 kV unit board (II-2-56,-62,-68) with increased range Condensate Booster Pump Breakers Revise breaker relay trip settings (U-9, U-8, U-6) Y Y N Condensate to RFW Pump A/B/C Replace pressure indicator with increased range Y - - Pressure (PI-2-81,-93,-106) Condenser A/B/C CCW Outlet Flow Rescale and add outlet flow signals to ICS Y Y Y (FIT-27-156,-157,-158,-159,-160,-161) Condenser A/B/C CCW Temperature Add inlet/outlet temperature signals to ICS Y Y Y (TE-27-33B,-36B-41B,-44B,-49B,-52B, -57B,

                                      -60B,-65B,-68B,-73B,-76B)

Condenser A/B/C CCW Pressure Replace transmitters and rescale inputs to ICS Y Y Y (PT-2-1,-5,-8) 2-184

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.4-2 Changes to Instrumentation and Controls (continued) Parameter EPU Change5 Implementation Status3 U1 U2 U3 Condenser A/B/C Low Vacuum Replace vacuum switches with pressure N N N Turbine Trip, Bypass Trip, and Alarm transmitters that provide input to the electro-hydraulic control (EHC) system (PS-47-72A/B,- 73A/B,-74A/B,-75A/B/C, -125A/B/C) Condenser A/B/C Low Vacuum - Revise turbine control software settings to support N N N Digital EHC System Software Update replacement of low condenser vacuum pressure switches with transmitters No. 1, 2 Extraction HDR Pressure Rescale loops (PT-5-3,-36) Y Y Y No. 2, 4 Extraction Steam Pressure Replace pressure gauges with increased range Y Y Y (from LP Turbine A/B/C to Heater (PI-5-33B,-34,-36,-46,-47,-49,-59,-60,-62) (HTR) A4/B4/C4) FWH - A1/B1/C1 Outlet Temperature Rescale ICS inputs (TE-3-44,-37,-30) Y Y N FWH - A1/B1/C1 Shell Pressure Rescale loops (PT/PI-5-6,-10,-14) Y Y N FWH - A2/B2/C2 Shell Pressure Rescale loops (PT/PI-5-18,-22,-26) Y Y N FWH - A1/B1/C1, A2/B2/C2, Replace obsolete transmitters Y Y Y A3/B3/C3 Level (LT-6-1A/B,-4A/B, -7A/B,-19A/B,-22A/B,- 25A/B,-37A/B,-40A/B,-43A/B) FWH - A1/B1/C1, A2/B2/C2, Rescale transmitter loops Y Y N A3/B3/C3 Level (LT-6-1A/B,-4A/B, -7A/B,-19A/B,-22A/B,- 25A/B,-37A/B,-40A/B,-43A/B) FWH - A1/B1/C1, A2/B2/C2, Replace obsolete FW HTR Level Indicating Y NA NA A3/B3/C3 Level Controllers (LIC-6-1,-4,-7,-19,-22,-25,-37,-40,-43) FWH - A1/B1/C1, A2/B2/C2, Replace obsolete level switches and recalibrate Y NA NA A3/B3/C3 Level (LS-6-1A/B,-4A/B, -7A/B,-19A/B,-22A/B,- 25A/B,-37A/B,-40A/B,-43A/B) Moisture Separator Level Control (LC) Remove trip function of switches (FIS-6-56A/B, Y Y Y Reservoir Drain Flow -57A/B,-58A/B) and retain as flow indicators with increased span (FI-6-56A/B,-57A/B,-58A/B)6 Condensate Demineralizer vessel Replace transmitters and rescale loops Y Y Y (VSL) Flow (FIT-2-208A thru H and J), add 10th VSL flow channel (FIT-2-208K) Condensate Demineralizer VSL DP Replace transmitters, rescale loops (PDIT-2-205A Y Y Y thru H and J), add 10th VSL DP channel (PDIT-2-205K) and revise Programmable Logic Controller (PLC) high DP setpoint (PDSH 130SP) 2-185

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.4-2 Changes to Instrumentation and Controls (continued) Parameter EPU Change5 Implementation Status3 U1 U2 U2 Condensate Demineralizer VSL Resin Replace transmitters and rescale loops Y Y Y Trap DP (PDIT-2-207A thru H and J) add 10th VSL flow channel (PDIT-2-207K) Condensate Demineralizer Inlet/Outlet Add pressure gauges for 10th demineralizer Y Y Y Pressure (PI-2-205AK, -BK) Condensate Injection Flow Add new flow indicator (FI-2-7043) Y Y Y Generator Hydrogen Pressure Replace obsolete low and high pressure N N N control/alarm switches and revise setpoints (PS-35-18A,-19,-18B) - revise control setpoints (PCV-35-5A,-5B,-39) Generator Protection Revise main generator over-excitation relay N N N setpoint (J1K) Generator Stator Cooling Water (SCW) Rescale transmitter (FIT-35-65) and revise low Y N Y Flow alarm setpoint (FA-35-65) Generator SCW Inlet Flow Revise low flow turbine runback/trip setpoints Y N Y (FS-35-65A/B/C) Generator SCW Inlet Pressure Revise high/low alarm setpoints (PA-35-90A/B) Y N Y Generator SCW DP Revise high/low/low low alarm setpoints Y N Y (PDA-35-91A/B/C) Generator SCW Outlet Temperature Revise high temperature turbine runback/trip and Y N Y alarm setpoints (TS-35-71A/B/C, TIS-35-72) Generator SCW Cooler Discharge Revise controller/control valve setpoints Y N Y Pressure (PC/PCV-35-55) Isophase Bus Duct Phase A/B/C Revise high temperature alarm setpoints Y Y Y Temperature at Generator (TS-262-3A1,-3A2,-3B1,-3B2,-3C1,-3C2) Isophase Bus Duct Phase A/B/C Revise high temperature alarm setpoints Y Y Y Temperature at Main Bus (TS-262-3A3,-3B3,-3C3) Digital EHC System Software Revise turbine control software settings for N N N electrical overspeed setpoint, intermediate pressure, power load unbalance, turbine first stage pressure, and MWe control Offgas Condenser Cooling Water Replace obsolete temperature sensor and rescale Y Y Y indicator (TE/TI-2-256) 2-186

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.4-2 Changes to Instrumentation and Controls (continued) Parameter EPU Change5 Implementation Status3 U1 U2 U3 Hydrogen Water Chemistry Replace and rescale HWC oxygen to condensate Y Y N flow indicator (FI-4-9) Hydrogen Water Chemistry PLC Install updated software in HWC PLC (PLC-4-40) Y Y N for H2 and O2 injection rates at EPU MCR Recorders Replace obsolete recorders - reactor vessel Y Y Y level/total FW flow (XR-3-53), main steam flow (FR-46-5) MCR Recorders Rescale recorders - reactor vessel level/total FW Y Y N flow (XR-3-53), main steam flow (FR-46-5) Notes:

1. Requires change to the differential pressure setpoint value.

2. Includes associated software updates to FWCS and RFPT Woodward Governor controls.

3. Implementation Status:

Y (Yes) N (No) - The modification will be installed prior to implementing EPU on the respective Browns Ferry Unit.

4. Software updates to FWCS.

5. All loops rescaled to EPU values include corresponding rescaling of Integrated Computer System (ICS) inputs where applicable.

6. For Unit 1, the flow indicating switches were removed and replaced with flow indicating transmitters (FIT-6-56A/B,-57A/B,-58A/B) and rescaled for EPU.

7. For Unit 1, the component number for Condensate Pump 1C is EI-2-15/6.

8. Unit 1 controller is FIC-2-29 and Unit 3 recorder is FR-2-29.

9. Unit 1 Pump 1C Breaker relay is U-6.

2-187

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.5 Plant Systems 2.5.1 Internal Hazards 2.5.1.1 Flooding 2.5.1.1.1 Flood Protection Regulatory Evaluation TVA conducted a review in the area of flood protection to ensure that SSCs important to safety are protected from flooding. The NRCs acceptance criteria for flood protection are based on GDC-2. Specific NRC review criteria are contained in SRP Section 3.4.1. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-2. Browns Ferry internal flooding hazards are described in Browns Ferry UFSAR Section 10.16.4.6 Evaluation for Flooding due to Failure of Low Energy Piping Systems Outside Primary Containment. Browns Ferrys internal flooding hazards were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry 2-188

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the flood protection barriers is documented in NUREG-1843, Section 2.4. During plant license renewal evaluations, tanks, and pipes which were not already in scope pursuant to 10 CFR 54.4(a)(1) or (a)(3) were evaluated to ensure they were not "non-safety equipment whose failure could affect a safety function" (Criterion (a)(2)). Components that met the inclusion criteria were evaluated within the system that contained them. Additionally, civil features whose function was to control, abate, or minimize the effects of flooding were identified and evaluated within the structure that contained them. Technical Evaluation 2.5.1.1.1.1 High Energy Line Break NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 10.1 of the CLTR addresses the effect of EPU on flooding. The results of this evaluation are described below. High Energy Line Breaks (HELBs) are evaluated for their effects on equipment qualification. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Liquid Lines Plant Specific Disposition As stated in Section 10.1 of the CLTR, EPU may increase subcooling in the reactor vessel, which may lead to increased mass and energy release rates for liquid line breaks for RWCU only. Components and/or equipment required for safe shutdown of the reactor were evaluated for the effect of flooding from breaks and cracks in high-energy lines. The evaluations verified that the plant can be safely shut down, assuming a concurrent single active failure in systems necessary to mitigate the consequences of the postulated component failure. Systems that are affected by EPU are FW and RWCU. The CLTP mass and energy releases for feedwater line breaks are affected by EPU implementation due to the changes in the feedwater system including increased feedwater flow rate and modifications to the condensate, condensate booster and feedwater pumps. At EPU, the RWCU system will operate at a lower enthalpy. Plant flooding due to internal piping failures in these systems was evaluated for changes due to EPU. Plant flooding is conservatively evaluated based on the most limiting event, which is the FW line break. In this event, the entire hotwell volume is being released in the main steam valve vault and main steam tunnel, and then drains to the Reactor Building. Because no changes are made to the existing hotwell inventory, draining systems, and flood barriers, the flood levels in the Reactor Building 2-189

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) due to a FW break are unchanged. RWCU line break flood level increase is due to the RWCU operating at a lower enthalpy, which will result in an increase in the critical crack flow by 4.41%. The change in the RWCU flood level remains bounded by the FW line break flood level. The remaining systems evaluated are not affected by EPU and remain bounded by the current flooding analyses. Internal flooding due to postulated failures in piping systems is not affected by EPU. Therefore, Browns Ferry meets all CLTR dispositions for liquid lines. 2.5.1.1.1.2 Moderate Energy Line Break NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 10.2 of the CLTR addresses the effect of EPU on flooding. The results of this evaluation are described below. The EPU effect on Moderate Energy Line Break (MELB) spray and subcompartment temperature is addressed in Section 2.5.1.3.2. This section discusses the EPU effect on flooding levels. Browns Ferry addresses the concern of moderate energy line breaks through various initiatives including: Probabilistic Risk Assessment - Internal Flooding Analysis Seismic Interaction Piping and Other Components Based on the Unresolved Safety Issues (USI) A-17, System Interactions in Nuclear Power Plants and USI A-46, Seismic Qualifications of Equipment in Operating Plants. While Browns Ferry was not originally licensed using the Standard Review Plan Section 3.6.1, the licensing and design basis includes specific conditions for MELBs for effects on safety-related equipment. MELBs are evaluated for their effects on equipment used for safe shutdown. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation for MELB are: Topic CLTR Disposition Browns Ferry Result Meets CLTR Flooding Generic Disposition The CLTR states that EPU results in no change in the inventory contained in moderate energy lines. The flow rates and/or the system inventories of analyzed moderate energy piping systems do not increase for EPU. System design limits (design pressure) used as input to the MELB flooding analyses are not changed by EPU. EPU does not affect the ability of the plant to cope with 2-190

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) effects of spray from MELBs. EPU does not introduce new MELB locations and does not introduce or move safety-related equipment. EPU will not affect the normal operating water levels and pressure of the suppression pool (torus), the flood seals between the Reactor and Turbine Building, flood barriers (curbs) of ECCS rooms, and flood level detection equipment in the lower Reactor Building elevation. The Intake Pumping Station and location of safety-related equipment within the structure will not change for EPU. Sources of flooding and protection measures in the Circulating Water System (CWS) are not affected by EPU. The CWS is located in the Turbine Building, which is sealed from the Reactor Building to an elevation 572.5 feet, and the time for operator action remains valid, as EPU does not increase the design pressure for the CWS. Therefore, MELB internal flooding meets all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on internal flooding hazards. The evaluation indicates that SSCs important to safety will continue to be protected from flooding and will continue to meet the requirements of draft GDC-2 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to flood protection. 2.5.1.1.2 Equipment and Floor Drains Regulatory Evaluation The function of the Equipment and Floor Drainage System (EFDS) is to assure that waste liquids, valve and pump leakoffs, and tank drains are directed to the proper area for processing or disposal. The EFDS is designed to handle the volume of leakage expected, prevent a backflow of water that might result from maximum flood levels to areas of the plant containing safety-related equipment, and protect against the potential for inadvertent transfer of contaminated fluids to an uncontaminated drainage system. The NRCs acceptance criteria for the EFDS are based on GDCs-2 and 4 insofar as they require the EFDS to be designed to withstand the effects of earthquakes and to be compatible with the environmental conditions (flooding) associated with normal operation, maintenance, testing, and postulated accidents (pipe failures and tank ruptures). Specific NRC review criteria are contained in SRP Section 9.3.3. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, 2-191

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-2. The equipment and floor drains are described in Browns Ferry UFSAR Section 10.16, Equipment and Floor Drainage Systems. Browns Ferrys equipment and floor drain systems were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). During plant license renewal evaluations, tanks and pipes which were not already in scope pursuant to 10 CFR 54.4(a)(1) or (a)(3) were evaluated to ensure they were not "non-safety equipment whose failure could affect a safety function" (Criterion (a)(2)). Components that met the inclusion criteria were evaluated within the system that contained them. Additionally, civil features whose function was to control, abate, or minimize the effects of flooding were identified and evaluated within the structure that contained them. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 8.1 of the CLTR addresses the effect of EPU on the Equipment and Floor Drain system. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Waste Volumes Plant Specific Disposition 2-192

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The CLTR states that power uprate does not affect the floor drain collector subsystem and the waste collector subsystem operation or equipment performance. The floor drain collector subsystem and the waste collector (equipment drain) subsystem both receive periodic inputs from a variety of sources. Neither subsystem is expected to experience a large increase in the total volume of liquid and solid waste due to operation at the EPU condition. The design of the Browns Ferry equipment and floor drains inside and outside of containment has been evaluated to ensure any EPU-related liquid radwaste increases can be processed. Browns Ferry has sufficient capacity to handle added liquid increases expected (i.e., it can collect and process the drain fluids). Therefore, EPU does not affect system operation or equipment performance and meets all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on the EFDS. The evaluation indicates that the EFDS has sufficient capacity to (1) handle any additional expected leakage resulting from the plant changes, (2) does not affect the backflow of water to areas with safety-related equipment. The EFDS will continue to meet the requirements of draft GDC-2 following implementation of the proposed EPU. Therefore the proposed EPU is acceptable with respect to the EFDS. 2.5.1.1.3 Circulating Water System Regulatory Evaluation The Circulating Water System (CWS) provides a continuous supply of cooling water to the main condenser to remove the heat rejected by the turbine cycle and auxiliary systems. The NRCs acceptance criteria for the CWS are based on GDC-4 for the effects of flooding of safety-related areas due to leakage from the CWS and the effects of malfunction or failure of a component or piping of the CWS on the functional performance capabilities of safety-related SSCs. Specific NRC review criteria are contained in SRP Section 10.4.5. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with 2-193

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A. No draft GDCs directly apply to the Circulating Water System. The Circulating Water System is described in Browns Ferry UFSAR Section 11.6, Condenser Circulating Water System. The Browns Ferry circulating water system was evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Circulating Water System is documented in NUREG-1843, Section 2.3.4.6. Management of aging effects on the Circulating Water System is documented in NUREG-1843, Section 3.3. Technical Evaluation The circulating water system is not being modified for EPU operation. The performance of the system was evaluated for EPU based on the original design capacity of the CWS and the cooling tower system over the actual range of circulating water inlet temperatures, and confirms that the circulating water system and heat sink are adequate for EPU operation. The evaluation of the CWS at EPU power indicates sufficient system capacity to ensure that the plant maintains adequate condenser backpressure. However, condenser backpressure limitations may require load reductions at the upper range of the anticipated circulating water inlet temperatures. Conclusion There are no EPU related modifications to the CWS. Performance was analyzed with respect to EPU power levels. Condenser backpressure limitations may require load reductions at the upper range of the anticipated circulating water inlet temperatures. The effect of EPU on the flooding analyses is addressed in Section 2.5.1.1.1. 2.5.1.2 Missile Protection 2.5.1.2.1 Internally Generated Missiles Regulatory Evaluation TVAs review concerns missiles that could result from in-plant component overspeed failures and high-pressure system ruptures. 2-194

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The NRCs acceptance criteria for the protection of SSCs important to safety against the effects of internally generated missiles that may result from equipment failures are based on GDC-4. Specific NRC review criteria are contained in SRP Sections 3.5.1.1 and 3.5.1.2. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-40. The missile protection for internally generated missiles is described in Browns Ferry UFSAR Sections 5.2.4.6, Missile and Pipe Whip Prevention, and 11.2.2, Power Generation Design Basis. Browns Ferrys systems and components were evaluated for license renewal. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The equipment and components credited with mitigating the effect of missiles are documented in NUREG-1843, Section 2.4, and the programs credited with managing that equipment aging are documented in NUREG-1843, Section 3.5. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the 2-195

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) effects of EPUs. Section 7.1 of the CLTR addresses the effect of EPU on the turbine generator. The results of this evaluation regarding turbine missiles are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Turbine-Generator Missile Avoidance Plant Specific Disposition As explicitly stated in Section 7.1 of the CLTR, the increase in steam flow can change the previous missile avoidance and protection analysis. The Browns Ferry design inherently provides missile protection for the safety-related SSCs, and important-to-safety non-safety related SSCs, by orienting the main and RFP turbines perpendicular to the control bay, Reactor Building and other structures containing safety-related and important-to-safety systems and components. This configuration ensures, in the unlikely event of a turbine failure, any missiles escaping the turbine shell are ejected away from the control bay, Reactor Building and other structures containing safety-related and important-to-safety systems and components. All three units have favorably oriented turbines as defined by RG 1.115 (Reference 64). Unit 1 has replaced the LP turbine rotors with monoblock integral rotors. The new rotors are not susceptible to low speed rotor failure. A specific missile generation study is not required for the Unit 1 turbine. Integral (monoblock) rotors are not considered a source for potential missile generation for EPU for Unit 1. For Browns Ferry Units 2 and 3, specific calculations have been performed to determine the probability of a turbine missile. The worst case probability based on inspection frequency and turbine valve testing is 3.3 x10-5/year. When this turbine missile probability is applied using the NRC approved methodology (Reference 64) for calculating turbine missile damage probability, the resultant probability is 3.3 x 10-8/year, which is below the acceptance criteria of 1 x 10-7/year. The probability of a missile as the result of a runaway overspeed event is acceptable for EPU. See Section 2.5.1.2.2 for additional information on the main turbine. Transients which affect the feedwater pumps and turbines will be limited by the protective features of the feedwater control system. Therefore, there is no adverse effect associated with transients on the feedwater system. Because the extended power uprate is at a constant pressure, there is no increase in the operating pressure of other auxiliary systems located in the Reactor Building for either the normally operating systems or the standby ECCS required to mitigate the consequences of abnormal transients or accidents. There is no change in the potential for generation of missiles or the energy of analyzed missiles in either safety-related systems or non-safety related systems in the proximity of safety-related SSCs. Therefore, the missile analyses remain valid. 2-196

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The spent fuel pool system is located in the reinforced concrete Reactor Building. There is no large normally operating rotating equipment adjacent to the spent fuel pool. Dynamic effects and missiles that might result from plant equipment failures in the vicinity of the spent fuel pool have not changed with respect the plant's current design. The review criterion specified in Matrix 5 of RS-001 is applicable to EPUs that result in substantially higher system pressures or changes in existing system configuration. Pressure does increase in the condensate and feedwater systems. However, the areas of increased pressure are not in the vicinity of SSCs important to safety as defined by RG 1.115 Appendix A. The Browns Ferry EPU does not create any condition resulting in an increase in probability of the generation of internal missiles. In addition, the Browns Ferry EPU does not entail any changes in equipment configurations that could change the effect of internally generated missiles on important-to-safety equipment. Therefore, internally generated missiles meet all CLTR dispositions. Conclusion TVA has evaluated changes in system pressures, configurations, and equipment rotational speeds necessary to support the proposed EPU. The evaluation indicates that SSCs important to safety will continue to be protected from the effects of internally generated missiles in accordance with draft GDC-40. Therefore, the proposed EPU is acceptable with respect to the protection of SSCs important to safety from internally generated missiles. 2.5.1.2.2 Turbine Generator Regulatory Evaluation The turbine control system, steam inlet stop and control valves, low pressure turbine steam intercept and inlet control valves control the speed of the turbine under normal and abnormal conditions, and are thus related to the overall safe operation of the plant. The NRCs acceptance criteria for the turbine generator are based on GDC-4, and relates to protection of SSCs important to safety from the effects of turbine missiles by providing a turbine overspeed protection system (with suitable redundancy) to minimize the probability of generating turbine missiles. Specific NRC review criteria are contained in SRP Section 10.2. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis 2-197

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-40. The turbine generator is described in Browns Ferry UFSAR Sections 11.2, Turbine-Generator, and 7.11, Pressure Regulator and Turbine-Generator Control. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the turbine generator is documented in NUREG-1843, Section 2.3.4. Management of aging effects on the turbine generator is documented in NUREG-1843, Section 3.4. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 7.1 of the CLTR addresses the effect of EPU on the turbine-generator. The results of this evaluation are described below. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Turbine-Generator Performance Plant Specific Disposition The turbine-generator converts the thermal energy in the steam into electrical energy. The increase in thermal energy and steam flow from the reactor is translated to an increased electrical output from the station by the turbine-generator. The increase in steam flow can also change the previous missile avoidance and protection analysis (See Section 2.5.1.2.1). 2-198

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The turbine-generator is required for normal plant operation and is not safety-related. Experience with previous power uprate applications indicates that turbine and generator modifications (e.g., turbine rotating element modification) are required to support power uprate. These modifications are required to support normal operation and are non-safety related. The turbine-generator overspeed protection systems were evaluated to ensure that adequate protection is provided for EPU conditions. The turbine and generator were originally designed with a maximum flow-passing capability and generator output in excess of rated conditions to ensure that the original rated steam-passing capability and generator output were achieved. This excess design capacity ensured that the turbine and generator meet rated conditions for continuous operating capability with allowances for variations in flow coefficients from expected values, manufacturing tolerances, and other variables that may adversely affect the flow-passing capability of the units. The difference in the steam-passing capability between the design condition and the rated condition is called the flow margin. At CLTP and at a reactor dome pressure of 1,050 psia, the main turbines operate with a current rated throttle steam flow of 14.153 Mlbm/hr at a throttle pressure of 1,000 psia. The generators are rated at 1,280 MVA at a power factor of 0.9. At EPU RTP and at a reactor dome pressure of 1,050 psia, the main turbines will operate with a rated throttle steam flow of 16.44 Mlbm/hr at a throttle pressure of 983 psia. The original Browns Ferry main generators were rewound in anticipation of uprating the power. The reactive capability curves are shown in Figures 2.5-2a (Unit 1) and 2.5-2b (Units 2 and 3). The current main generators are rated as follows for EPU: Unit 1: 1,330 MVA at a 0.95 power factor Units 2 and 3: 1,332 MVA at a 0.93 power factor The existing HP turbine for each Browns Ferry unit is not capable of passing the required EPU steam flow rate and will be replaced prior to EPU. The new HP turbine section has been designed with an effective throttle flow margin of 5 percent above the required EPU throttle flow. The design point of the new HP turbine included the flow margin in order to ensure that the HP turbine will pass the rated throttle flow, as well as to allow for reactor pressure control. Therefore, the Valves Wide Open (VWO) condition refers to the turbine supply steam flow with additional margin over rated condition when adjusted for the lower inlet pressure associated with higher flow. For operation at EPU, the high pressure turbine has been re-designed with replacement diaphragms, buckets, and a new rotor, for at least the minimum target throttle flow margin, to increase the flow passing capability. The expected environmental changes, such as diurnal heating and cooling effects changing cycle efficiency, periodically require management of reactor power to remain within the generator rating. The required variations in reactor power do not approach the magnitude of changes periodically required for surveillance testing and rod pattern alignments and other occasional events requiring de-rating, such as equipment out-of-service for maintenance. 2-199

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) As part of the EPU on Unit 1, the original shrunk-on Low Pressure (LP) rotors were replaced with rotors of monoblock (integral) design. The HP rotors will also be replaced with rotors of monoblock design. Per CLTR Section 7.1, The only safety related evaluation is the plant specific turbine-generator missile avoidance and protection analysis. The entrapped energy following a turbine trip or load rejection increases slightly for CPPU. Relative to the turbine generator missile protection analysis, many power plants have replaced high pressure and low pressure shrunk-on rotors with an integral rotor without shrunk-on wheels. These integral rotors are not considered a source for potential missile generation for CPPU for the slight increase in entrapped energy; therefore, a plant specific analysis is not required. As part of the EPU on Units 2 and 3, modifications and inspections have been performed to the shrunk-on wheels for the LP turbine rotors to reduce the probability of LP turbine rotor blade failure and ejection. The HP turbine rotors will also be replaced with rotors of monoblock design. A specific missile generation study was performed. See Section 2.5.1.2.1 for the turbine missile evaluation. The turbine overspeed calculation compares the entrapped steam energy contained within the turbine and the associated piping, after the stop valves trip, and the sensitivity of the rotor train for the potential overspeed capability. The entrapped energy increases slightly for EPU conditions. Appendix A of the CLTR states that although the power uprate slightly increases the energy trapped in the turbine following a load rejection, the turbine overspeed would remain within design limits. The turbine overspeed scenario considered is the emergency case where the EHC controls and the control and intercept valves fail to respond to the initial overspeed due to a load rejection event. For this scenario, the unit rapidly accelerates to the overspeed trip setpoint, thereby trip-closing the main and intermediate stop valves. The operating condition analyzed was the maximum power, valves wide open case, with low backpressure. This approach accounts for the two basic contributors to peak overspeed due to a load rejection event: 1) the energy due to entrapped (or entrained) steam within the steam path and inlet piping downstream of the main and intermediate steam valves; and 2) what is termed "valve lag overspeed," which takes into account the energy contributed by new steam entering the machine during the response time of the control and trip systems, and during the actual closing time of these valves. The overspeed trip setpoint is established such that the resulting peak speed will not exceed the 120% emergency overspeed limit due to overshoot. This ensures that the turbine is protected in an overspeed event. The turbine and turbine control system design changes for EPU have not yet been installed and the specific control setpoints have not been established. The setpoints will be adjusted to ensure that the turbine will not exceed 120% of rated speed due to overshoot. Equipment important to safety associated with the plant is protected from main turbine missiles by physical barriers and favorable alignment. Additionally, the independent spent fuel storage installation has been evaluated and determined acceptable with regard to plant generated main turbine missiles using the EPU turbine failure probability analyses as input. The effect of EPU is 2-200

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) offset by ensuring that the turbine speed will not exceed 120% of rated during an overspeed event. The main generator excitation systems, that existed on all three units, were GE Alterrex Excitation systems. The energy for excitation of the generator field is derived from the turbine using a self-excited shaft driven alternator. The output of the alternator is rectified by a set of water cooled stationary diode bridges mounted in the doghouse of the generator. The resultant DC voltage is supplied to the generator field via brushes and collector rings. In 2011 and 2012, TVA replaced the voltage regulators in the Alterrex Excitation systems with new ABB Unitrol 5000 voltage regulators. The Interconnection System Impact Study (SIS), performed for Browns Ferry at EPU conditions, determined that main generator stability issues exist for a 3-phase fault on one of several Browns Ferry transmission lines coincident with certain transmission lines already out of service (N-1-1 event.). The present excitation system cannot raise the field voltage fast enough and high enough to prevent the main generator from becoming unstable. To address the transient stability issue, Browns Ferry will install a new shunt-fed, static excitation system on each of the three units. The new excitation system will provide the field current directly to the brushes and collector rings eliminating the need for the rotating Alterrex exciter and the stationary diodes. Additional computer simulations of the N-1-1 event for the SIS, with this new excitation system modeled, have shown that the generators remain stable. As the new excitation system requires a higher DC field voltage, the insulating material in the rotors will need to be evaluated as acceptable or replaced. If the rotors must be changed (e.g., insulating material replaced), the design change process will ensure that the modifications will not have an adverse effect on GDC-17 compliance, generator stability, short circuit analysis acceptance criteria compliance, and reliability and availability of the offsite transmission system. When conditions exist, that make generator instabilities possible during an N-1-1 event, administrative controls will be implemented to limit unit output whenever certain transmission lines are out of service. Once the excitation systems are replaced on all three units, the administrative controls will no longer be necessary. Use of administrative controls to prevent transient instabilities, prior to the excitation systems being installed, is allowed by North American Electric Reliability Council (NERC) standards. Conclusion TVA has evaluated the effects of the proposed EPU on the turbine generator. The evaluation indicates that the turbine generator will continue to provide adequate turbine overspeed protection to minimize the probability of generating turbine missiles and will continue to meet the requirements of draft GDC-40 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the turbine generator. 2-201

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.5.1.3 Pipe Failures Regulatory Evaluation A review of the plant design was conducted regarding protection from piping failures outside containment to ensure that (1) such failures would not cause the loss of needed functions of safety-related systems and (2) the plant could be safely shut down in the event of such failures. The NRCs acceptance criteria for pipe failures are based on GDC-4, which requires, in part, that SSCs important to safety be designed to accommodate the dynamic effects of postulated pipe ruptures, including the effects of pipe whipping and discharging fluids. Specific NRC review criteria are contained in SRP Section 3.6.1. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-40. Piping failures outside containment are described in Browns Ferry UFSAR Appendix M, Report on Pipe Failures Outside Containment in the Browns Ferry Nuclear Plant. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Sections 9.2.1, 10.1, and 10.2 of the CLTR address the effects of EPU on Piping Failures. The results of this evaluation are described below. 2-202

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.5.1.3.1 High Energy Piping Outside Containment Where EPU resulted in increased piping stresses in high energy piping outside containment, the increased stresses were evaluated against existing line break criteria to identify any potential new break locations. The results of that evaluation (see Section 2.2.1) determined that there are no new high energy line break locations outside containment due to operation at EPU conditions. Pipe break criteria were evaluated based on the requirements of Appendix M of the UFSAR, which is based on current licensing basis requirements. The combinations of stresses were evaluated to meet the requirement of pipe break criteria. Based on these criteria, no new postulated pipe break locations were identified. Existing high-energy line break locations outside containment that are affected by EPU are identified in Section 2.2.1 with the effects summarized in Table 2.2-1. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Steam Lines Generic Disposition Meets CLTR Liquid Lines Plant Specific Disposition 2.5.1.3.1.1 Steam Lines The effect of EPU on HELB mass and energy release rates for steam lines outside containment is documented in Section 2.2.1.1. Section 2.2.1.1 concludes that the generic CLTR disposition for high-energy line breaks in steam lines is applicable and that EPU has no effect on HELB mass and energy release rates for steam lines outside containment. The Browns Ferry design basis for steam line breaks also includes a plant-specific MS Line intermediate break for which the effects of EPU are documented in Section 2.2.1.1. The CLTR states that there is no effect on steam line breaks because steam conditions at the postulated break conditions are unchanged. EPU has no effect on the steam pressure or enthalpy at the postulated break locations. Therefore, EPU has no effect on the mass and energy releases from a HELB in a steam line. Therefore, the Browns Ferry steam lines meet all CLTR dispositions. 2.5.1.3.1.2 Liquid Lines The effect of EPU on HELB mass and energy release rates for liquid lines outside containment is documented in Section 2.2.1.2. The evaluations document energy release rates for the RWCU and FW systems. Section 2.2.1.2.1 documents the plant specific HELB evaluation of RWCU line breaks, and Section 2.2.1.2.2 documents the plant specific HELB evaluation for FW line 2-203

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) breaks. The effects of EPU operation on the feedwater line break (FWLB) pipe whip, jet impingement, jet reaction and flooding analyses are addressed in Sections 2.2.1.2 and 2.5.1.1. The CLTR states that EPU may increase subcooling in the reactor vessel, which may lead to increased break flow rates for liquid line breaks. EPU conditions may result in an increase in the mass and energy release for liquid line breaks. Therefore, liquid line breaks are evaluated for EPU, and the evaluations include EPU effects on subcompartment pressures and temperatures, pipe whip and jet impingement, and flooding. The ability of the plant to cope with the flooding effects from HELBs outside containment that are affected by EPU is evaluated in Section 2.5.1.1. RWCU mass and energy release rates and their effect on environmental conditions (compartment pressures and temperatures) were re-analyzed for both CLTP and EPU (for Units 2 and 3). Unit 1 was evaluated for EPU only. The CLTP mass and energy release rates (Units 2 and 3) for FW line breaks are negligibly affected by EPU. However, the effects of a FW system line break on main steam valve vault peak pressures and temperatures will continue to be bounded by a main steam line break in the main steam valve vault. Therefore, the Browns Ferry Liquid Lines meet all CLTR dispositions. 2.5.1.3.2 Moderate Energy Piping Outside Containment As stated in Section 2.5.1.1, system design limits (design pressure) used as input to the MELB flooding analyses and are not changed by EPU. Because the Browns Ferry MELB mass releases and environmental conditions (pressures and temperatures) are not affected by the EPU, there is no adverse effect on post-MELB control room habitability or on access to areas important to safe control of post-accident operations. Browns Ferry Topic CLTR Disposition Result Meets CLTR Flooding Generic Disposition The CLTR states that EPU results in no change in the inventory contained in moderate energy lines. Therefore, flooding meets all CLTR dispositions. 2.5.1.3.3 Environmental Conditions All EQ equipment outside primary containment remains above the flood levels resulting from postulated pipe breaks and therefore are not subject to submergence. Conclusion TVA has evaluated the changes that are necessary for the proposed EPU. The evaluation indicates that SSCs important to safety will continue to be protected from the dynamic effects of 2-204

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) postulated piping failures in fluid systems outside containment and will continue to meet the requirements of draft GDC-40 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to protection against postulated piping failures in fluid systems outside containment. 2.5.1.4 Fire Protection Regulatory Evaluation The purpose of the Fire Protection Program (FPP) is to provide assurance, through a defense-in-depth design, that a fire will not prevent the performance of necessary safe plant shutdown functions and will not significantly increase the risk of radioactive releases to the environment. The NRCs acceptance criteria for the FPP are based on (1) 10 CFR 50.48 and associated Appendix R to 10 CFR Part 50, insofar as they require the development of an FPP to ensure, among other things, the capability to safely shut down the plant; (2) GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; (3) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions. Specific NRC review criteria are contained in SRP Section 9.5.1.1, as supplemented by the guidance provided in Attachment 1 to Matrix 5 of Section 2.1 of RS-001. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. 2-205

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-4. Final GDC-3 is applicable to Browns Ferry as described in the Browns Ferry Fire Protection Report, Volume 1, Revision 20. Fire Protection is described in Browns Ferry UFSAR Section 10.11, Fire Protection Systems and the Fire Protection Report. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The fire protection systems are documented in NUREG-1843, Section 2.3.3.6. Fire barrier materials are addressed as a commodity group, while walls, floors, doors, and structural steel are evaluated within the building that contains them. Components credited with achieving safe shutdown following a fire are evaluated within the system that contains them. Management of aging effects on the fire protection systems is documented in NUREG-1843, Section 3.3. By letter dated October 28, 2015 (Reference 65), the NRC issued license amendments approving the transition of Browns Ferrys licensing basis to the National Fire Protection Association (NFPA) 805 standard in accordance with 10 CFR 50.48(c). Technical Evaluation 2.5.1.4.1 Fire Protection Program NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.7 of the CLTR addresses the effect of EPU on the fire protection program. The results of this evaluation are described below. As explicitly stated in Section 6.7 of the CLTR, ((

))

Therefore, the reactor and containment responses and operator actions will be evaluated on a plant-specific basis for EPU. This section addresses the effect of EPU on the fire protection program, fire suppression and detection systems, and reactor and containment system responses to postulated fire events. Once the NFPA 805 (Reference 66) fire protection transition is implemented, Browns Ferry will meet all CLTR dispositions. The topics addressed in this evaluation are: 2-206

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR Fire Suppression and Detection Systems Plant Specific Disposition Meets CLTR Operator Response Time Plant Specific Disposition Meets CLTR Peak Cladding Temperature Plant Specific Disposition Meets CLTR Vessel Water Level Plant Specific Disposition Meets CLTR Suppression Pool Temperature Plant Specific Disposition The higher decay heat associated with EPU results in higher heat input into the suppression pool which, without mitigation, will result in higher suppression pool temperatures. The higher decay heat may also result in lower vessel water levels or higher Peak Cladding Temperatures (PCTs), depending on the plant-specific analysis basis. As a result of these effects, fire suppression and detection systems, operator response time, peak clad temperature (PCT), and suppression pool temperature need to be addressed. Tennessee Valley Authority (TVA) implemented the Nuclear Energy Institute methodology NEI 04-02, Revision 2, Guidance for Implementing a Risk-informed, Performance-based Fire Protection Program Under 10 CFR 50.48(c) (NEI 04-02) (Reference 67), to transition Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3 from its current fire protection licensing basis to the new requirements as outlined in NFPA 805. The BFN NFPA 805 Fire Safe Shutdown Analysis consists of a deterministic analysis and a performance based analysis. The deterministic analysis (NFPA 805 Section 4.2.3) identifies and evaluates one success path for each fire area to meet the nuclear safety performance criteria of Section 1.5. Section 2.5.1.4.2, Fire Event, addresses the deterministic analysis. For instances where the nuclear safety performance criteria are not met, a performance based analysis (NFPA 805 Section 4.2.4) is performed to demonstrate that risk is acceptable and that defense in depth and safety margin are maintained. The performance based analysis is addressed in LAR Attachment 44. Safe Shutdown Systems, equipment, and compensatory measures will be sufficient to support EPU. EPU is found to not affect the elements of the fire protection program related to: (1) fire suppression and detection systems, (2) fire zones/areas, (3) fire barriers, and (4) fire protection responsibilities of plant personnel. Administrative controls, associated with fire protection in the 2-207

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Technical Specifications, the Fire Protection Report, and the Nuclear QA Plan, will be adequate for EPU conditions. EPU modifications will be assessed and assured not to adversely affect the ability to achieve and maintain the fuel in a safe and stable condition in the event of a fire. Original NFPA 805 analyses were performed at EPU conditions and therefore operator action times cannot be compared to CLTP conditions. To ensure that PCT remains less than the acceptance criterion in the most limiting scenario, one LPCI pump must be manually aligned for injection within 20 minutes. The EPU requires no new operator actions for fire safe shutdown of the plant and there are no actions required inside the primary containment. The reactor and containment responses to the postulated fire events at EPU conditions are described in Section 2.5.1.4.2. The results show that for the limiting thermal-hydraulic cases, peak fuel cladding temperature, vessel water level, and suppression pool temperature meet the acceptance criteria and there is sufficient time for the operators to perform the necessary actions to meet the NFPA 805 requirement to achieve and maintain the fuel in a safe and stable condition in the event of a fire. Therefore, once the NFPA 805 fire protection transition is implemented, Browns Ferry will meet all CLTR dispositions. 2.5.1.4.2 Fire Event The limiting NFPA 805 fire events were analyzed under EPU conditions. The fuel heat-up analysis was performed using the NRC accepted AREVA LOCA methodology (RELAX/HUXY). The containment analysis was performed using the GEH SHEX model. These analyses determined the effect of EPU on fuel cladding integrity, reactor vessel integrity, and containment integrity as a result of the fire event. The two bounding cases described below are identified as Case 1 and Case 4. See Tables 2.5-1, 2.5-2, and 2.5-3 for the inputs and results of the fire event analyses. Case 1: The bounding safe shutdown case for PCT has Multiple Spurious Operation (MSO) of 11 of the 13 MSRVs which depressurize the reactor, and one RHR pump aligned in the LPCI/ASDC mode at 20 minutes. The analysis shows that the calculated PCT of 1,330°F is acceptable from a deterministic perspective (< 1,500°F) (See FUSAR Section 2.5.1.4). Case 4 (See Figure 2.5-1): The bounding safe shutdown case for peak suppression pool temperature has reactor depressurization beginning at 25 minutes using three MSRVs. As the reactor is depressurized, condensate pumps replenish reactor inventory until hotwell inventory is depleted. After condensate is secured, one RHR pump is aligned into LPCI/ASDC mode. One RHRSW pump is initiated at two hours. Peak SP temperature reaches 207.7°F and this meets the containment integrity acceptance criteria of < 281°F and the torus attached piping limit of <223°F (See Section 2.2.2.2.2.2). Analyses show that containment accident pressure credit is not required to ensure adequate pump net positive suction head (NPSH) to mitigate a fire event (see Section 2.6.5.2 and LAR Attachment 39). 2-208

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The results of Case 4, and the evaluations in Section 2.6.5.2, FUSAR Section 2.5.1.4, and LAR 9, demonstrate that the peak fuel cladding temperature, vessel water level, and suppression pool temperature meet the acceptance criteria and the time available for the operators to perform the necessary actions is sufficient. Therefore, EPU has no adverse effect on the ability of the systems and personnel to mitigate the effects of a fire event and satisfies the requirement of achieving and maintaining the fuel in a safe and stable condition in the event of a fire. Conclusion TVA has evaluated fire-related safe shutdown requirements and has accounted for the effects of the increased decay heat on the ability of the required systems to achieve and maintain safe shutdown conditions. The evaluation indicates that the FPP will continue to meet the requirements of 10 CFR 50.48, final GDC-3, and draft GDC-4 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to fire protection. 2.5.2 Fission Product Control 2.5.2.1 Fission Product Control Systems and Structures Regulatory Evaluation The NRCs acceptance criteria are based on GDC-41, insofar as it requires that the containment atmosphere cleanup system be provided to reduce the concentration of fission products released to the environment following postulated accidents. Specific NRC review criteria are contained in SRP Section 6.5.3. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General 2-209

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-70. The Standby Gas Treatment System is described in Browns Ferry UFSAR Section 5.3.3.7, Standby Gas Treatment System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Standby Gas Treatment System is documented in NUREG-1843, Section 2.3.2.2. Management of aging effects on the Standby Gas Treatment System is documented in NUREG-1843, Section 3.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 4.5 of the CLTR addresses the effect of EPU on the SGTS. The assumptions regarding leakage and exhaust paths from the primary and secondary containments and other sources are as described in Alternative Source Term (AST) methodology for Browns Ferry (Reference 68). Browns Ferry meets all CLTR dispositions. Therefore, a plant-specific evaluation is not required. The topic addressed in this evaluation is: Browns Ferry Topic CLTR Disposition Result Meets CLTR Flow Capacity Generic Disposition Meets CLTR Iodine Removal Capability Generic Disposition The CLTR states that the core inventory of iodine and subsequent loading on the SGTS filters or charcoal adsorbers are affected by EPU. The SGTS is designed to maintain secondary containment at a negative pressure and to provide an elevated release path for the removal of fission products potentially present during abnormal conditions. By preventing the ground level release of airborne particulates and halogens, the SGTS limits off-site dose following a postulated DBA. The flow capacity of the SGTS and its ability to maintain a negative pressure in the secondary containment are discussed in Section 2.6.6. 2-210

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) At Browns Ferry, neither the SGTS component design nor the filter materials are being altered due to the EPU. The total (radioactive plus stable) post-LOCA iodine loading on the charcoal adsorbers increases proportionally with the increase in core iodine inventory, which increases with core thermal power. However, and as accepted by the CLTR, sufficient charcoal mass is present so that the post-LOCA iodine loading on the charcoal remains does not increase decay heating such that operation is challenged or there is a threat of charcoal ignition. Browns Ferry is not committed to RG-1.52 with respect to iodine loading onto SGTS charcoal. As is stated in Reference 1, ((

))

Two bounding analyses have been performed in the CLTR to evaluate decay heating in the SGTS for: 1) plants that implement AST in accordance with RG 1.183 (Reference 68), and

2) plants committed to RG 1.3 (Reference 71) for fission product transport. From Reference 1,

((

                                                                                       )) The parameters and their bounding values, with a comparison to the Browns Ferry specific values, are shown in Table 2.5-4.

As seen in Table 2.5-4, the Browns Ferry SGTS design is bounded with respect to the applicable parameters. ((

                                                                                         )) No credit is taken for charcoal adsorption for any DBA. Credit is taken for high efficiency particulate adsorber (HEPA) filter removal of 90% of the particulate activity in the DBA-LOCA analysis (Reference 72).

((

)).

Browns Ferry credits 2 (out of 3) trains operating during the accident period, therefore, the 2-211

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) values presented are accepted as the maximum heating and iodine loading for one of the two trains. The Browns Ferry SGTS utilizes a low flow cooling system to assure no desorption of radionuclides in the case of increased decay heating. While decay heat from fission products accumulated within the system filters and charcoal adsorbers increases with the increase in thermal power, the low flow cooling sub-system of the SGTS will still continue to protect the system from desorption should there be a loss of a system fan. The parameters used in the CLTR bounding analysis for AST application are confirmed to bound the Browns Ferry plant-specific values. Therefore the Browns Ferry SGTS design and operation under EPU conditions is consistent with the overall CLTR disposition for the SGTS (that the ability of the SGTS to remove fission products is not adversely affected by EPU) and satisfies applicable regulatory guidance. Conclusion TVA has evaluated the effects of the proposed EPU on the Standby Gas Treatment System. The evaluation indicates that the system will continue to provide adequate fission product removal in post-accident environments following implementation of the proposed EPU and will continue to meet the requirements of draft GDC-70. Therefore, the proposed EPU is acceptable with respect to the fission product control systems and structures. 2.5.2.2 Main Condenser Evacuation System Regulatory Evaluation The Main Condenser Evacuation System (MCES) generally consists of two subsystems: (1) the "hogging" or startup system that initially establishes main condenser vacuum and (2) the system that maintains condenser vacuum once it has been established. The NRCs acceptance criteria for the MCES are based on (1) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (2) GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences and postulated accidents. Specific NRC review criteria are contained in SRP Section 10.4.2. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis 2-212

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-17 and 70. The Main Condenser Evacuation System is described in Browns Ferry UFSAR Section 11.4, Main Condenser Gas Removal and Turbine Sealing System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Main Condenser Evacuation System is documented in NUREG-1843, Section 2.3.4. Management of aging effects on the Main Condenser Evacuation System is documented in NUREG-1843, Section 3.4.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 7.2 of the CLTR addresses the effect of EPU on the Condenser and Steam Jet Air Ejectors (SJAE). The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Condenser and SJAE Plant Specific Disposition The CLTR states that the increase in steam flow increases the heat removal requirement for the condenser. The additional power level increases the non-condensable gases generated by the reactor. 2-213

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The main condenser hogging (mechanical vacuum pump) and the SJAE functions are required for normal plant operation and are not safety-related. The design of the condenser air removal system is not adversely affected by EPU and no modification to the system is required. The following aspects of the condenser air removal system were evaluated for this determination: Non-condensable gas flow capacity of the SJAE system; Capability of the SJAEs to operate satisfactorily with available dilution / motive steam flow; and Mechanical vacuum (hogging) pump capability to remove required non-condensable gases from the condenser at EPU start-up conditions The capacity of the SJAEs is adequate because they were originally designed for operation at flows greater than those required at EPU conditions. Therefore, the main condenser evacuation system design bases for Browns Ferry are unchanged for EPU. Conclusion There are no EPU related changes to the MCES and the MCES will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment following implementation of the proposed EPU. The MCES will continue to meet the requirements of draft GDCs-17 and 70. Therefore, the proposed EPU is acceptable with respect to the MCES. 2.5.2.3 Turbine Gland Sealing System Regulatory Evaluation The turbine gland sealing system is provided to control the release of radioactive material from steam in the turbine to the environment. The NRCs acceptance criteria for the turbine gland sealing system are based on (1) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; and (2) GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences and postulated accidents. Specific NRC review criteria are contained in SRP Section 10.4.3. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, 2-214

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-17 and 70. The Turbine Gland Sealing System is described in Browns Ferry UFSAR Section 11.4, Main Condenser Gas Removal and Turbine Sealing System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Turbine Gland Sealing System is documented in NUREG-1843, Section 2.3.4. Management of aging effects on the Turbine Gland Sealing System is documented in NUREG-1843, Section 3.4.2. Technical Evaluation Each turbine sealing system includes a steam seal regulator with the necessary valves to maintain a constant positive pressure in the steam seal supply header and a single steam-packing exhauster condenser equipped with two full-capacity blowers to prevent steam leakage at the turbine shaft seals. The turbine sealing system prevents the leakage of steam into the Turbine Building and also prevents the leakage of air into the main condenser. During normal power operations, a pressure regulator valve and two seal steam header unloader valves maintain the seal steam header pressure at approximately 4 psig. To regulate the seal steam header pressure, the unloader valves divert excess seal steam to the main condenser. For EPU, larger unloader valves (8 to 10) and associated piping are being installed to provide additional capability to maintain the seal steam header pressure at approximately 4 psig. EPU conditions will not affect the capability of the turbine sealing system to contain activated nitrogen and limit exposure to radiation. Conclusion TVA has evaluated the effects of the proposed EPU on the turbine gland sealing system. This evaluation indicated that the turbine gland sealing system will continue to meet the performance 2-215

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) requirements following modification of the non-safety related seal steam header unloader valves and associated piping. After the modifications are implemented, the turbine gland sealing system will continue to maintain its ability to control and provide monitoring for releases of radioactive materials to the environment consistent with draft GDCs-17 and 70. Therefore, the proposed EPU is acceptable with respect to the turbine gland sealing system. 2.5.2.4 Main Steam Isolation Valve Leakage Control System Regulatory Evaluation Redundant quick-acting isolation valves are provided on each main steam line. The leakage control system is designed to reduce the amount of direct, untreated leakage from the Main Steam Isolation Valves (MSIVs) when isolation of the primary system and containment is required. The NRCs acceptance criteria for the MSIV leakage control system are based on GDC-54, insofar as it requires that piping systems penetrating containment be provided with leakage detection and isolation capabilities. Specific NRC review criteria are contained in SRP Section 6.7. Browns Ferry Current Licensing Basis The Browns Ferry design does not include a Main Steam Isolation Valve Leakage Control System. Technical Evaluation Not applicable. Conclusion Not applicable. 2.5.3 Component Cooling and Decay Heat Removal 2.5.3.1 Spent Fuel Pool Cooling and Cleanup System Regulatory Evaluation The spent fuel pool provides wet storage of spent fuel assemblies. The safety function of the spent fuel pool cooling and cleanup system is to cool the spent fuel assemblies and keep the spent fuel assemblies covered with water during all storage conditions. The NRCs acceptance criteria for the spent fuel pool cooling and cleanup system are based on (1) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; (2) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided; and (3) GDC-61, insofar as it requires that fuel storage systems be designed with RHR capability reflecting the importance to safety of decay 2-216

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) heat removal, and measures to prevent a significant loss of fuel storage coolant inventory under accident conditions. Specific NRC review criteria are contained in SRP Section 9.1.3, as supplemented by the guidance provided in Attachment 2 to Matrix 5 of Section 2.1 of RS-001. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-4, 67, and 69. There is no draft GDC directly associated with final GDC-44. The Spent Fuel Pool Cooling and Cleanup System is described in Browns Ferry UFSAR Section 10.5, Fuel Pool Cooling and Cleanup System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Spent Fuel Pool Cooling and Cleanup System is documented in NUREG-1843, Section 2.3.3.26. Management of aging effects on the Spent Fuel Pool Cooling and Cleanup System is documented in NUREG-1843, Section 3.3.2. 2-217

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.3 of the CLTR addresses the effect of EPU on the fuel pool. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Fuel Pool Cooling (Normal Core Offload and Full Meets CLTR Plant Specific Core Offload) Disposition Meets CLTR Crud Activity and Corrosion Products Plant Specific Disposition Meets CLTR Radiation Levels Plant Specific Disposition Meets CLTR Fuel Racks Generic Disposition 2.5.3.1.1 Fuel Pool Cooling (Normal and Full Core Offload) As stated in Section 6.3.1 of the CLTR, for the same time after shutdown, the spent fuel pool heat load increases due to the decay heat generation as a result of EPU. The spent fuel cooling section of the Fuel Pool Cooling and Cleanup System (FPCCS) consists of two trains of pumps and heat exchangers and two trains of the non-safety Auxiliary Decay Heat Removal (ADHR) system. The RHR safety-related system supplemental fuel pool cooling mode may be used to augment the capacity of the FPCCS when the ADHR system is unavailable. The Browns Ferry Spent Fuel Pool (SFP) bulk water temperature must be maintained below the licensing limit of 150F. The temperature requirement assures operator comfort (an operational requirement), and provides ample margin against an inventory loss in the fuel pool due to evaporation or boiling. The limiting condition is a full core discharge with all remaining storage locations filled with used fuel from prior discharges. EPU does not affect the alignments, availability or safety-related designations of these systems. EPU did not change the trains of cooling used to evaluate the effects of core offload. EPU will increase the decay heat load 14.29% for fuel being offloaded from the reactor. This will result in a small overall increase in the heat load on the FPCCS during and after refueling outages because of the increase in decay heat. The decay heat for the EPU was calculated using 2-218

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) the formulation and uncertainty factors from ANSI/ANS-5.1-1979 with two-sigma uncertainty and a correction for miscellaneous actinides and activation products. The use of ANSI/ANS-5.1-1979 has been endorsed in NUREG-0800 Section 9.2.5, Revision 3 (Reference 73). The effect of this heat load on the SFP temperature was then evaluated for bounding full core offloads added to a bounding SFP heat load from previously offloaded batches. The evaluation of the full core offload credits one loop of the FPCCS and one loop of the ADHR system for directly removing the decay heat from the SFP. The result of this conservative evaluation shows that, using the single loop of FPCCS and ADHR alone, the SFP temperature can be maintained below 150°F. EPU does not affect the heat removal capability of the FPCCS, the ADHR system, or the supplemental fuel pool cooling mode of the RHR system. EPU results in slightly higher core decay heat loads during refueling. Each reload affects the decay heat generation in the SFP after a batch discharge of fuel from the reactor. The full core offload heat load in the SFP reaches a maximum immediately after the full core discharge. Plant procedures limit the rate of heat addition to the fuel pool based on calculated operational heat load limits and available heat removal systems. Operational considerations for these procedural limits include delaying initial fuel movements into the pool and/or limitations on the rate of transfer of the fuel to the pool. The SFP normal makeup source is from the Seismic Category II condensate storage system with a capacity of 100 gpm and is not affected by EPU and remains adequate for EPU conditions. Browns Ferry has two Seismic Category I emergency makeup sources, the RHR/RHR service water crosstie and the emergency equipment cooling water system; each has a makeup capability of at least 150 gpm. Existing plant instrumentation and procedures provide adequate indications and direction for monitoring and controlling SFP temperature and level during normal batch offloads and the unexpected case of the limiting full core offload. Symptom based operating procedures exist to provide mitigation strategies including placing additional cooling trains or systems in service, stopping fuel movement, and initiating make-up if necessary. The symptom based entry conditions and mitigation strategies for these procedures do not require changes for EPU. A normal batch offload (approximately 332 fuel bundles) is assumed for outage planning with the additional assumptions in either case (batch or full core) of only one of two trains of FPCCS and only one of two trains of ADHR available, 24-month fuel cycle, and ANSI/ANS 5.1-1979 + 2. 2.5.3.1.2 Crud Activity and Corrosion Products Section 6.3 of the CLTR requires a plant-specific evaluation for the fuel pool crud activity and corrosion products. As stated in Section 6.3.2 of the CLTR, crud activity and corrosion products associated with spent fuel can increase slightly due to power uprate. The amount of crud activity and pool quality are operational considerations and are unrelated to safety. An evaluation of the capability 2-219

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) of the FPCCS to maintain water clarity concludes that water clarity will not be affected by EPU. Therefore, the crud activity and corrosion products meet all CLTR dispositions. 2.5.3.1.3 Radiation Levels As stated in Section 6.3.3 of the CLTR, the normal radiation levels around the SFP may increase slightly, primarily during fuel handling operations. Radiation levels in those areas of the plant, which are directly affected by the reactor core and spent fuel, increase by the percentage increase in the average power density of the fuel bundles. Therefore, for an EPU increase of 14.29%, the radiation dose rates increase by 14.29%. The radiation level around the SFP is an operational consideration and is unrelated to safety. EPU will increase the core thermal power by up to 14.29% from 3,458 MWt to 3,952 MWt. The radiation levels in the spent fuel are therefore assumed to increase by 14.29% due to EPU. This increase is acceptable as compared to worst case area dose limits. The design of spent fuel pools is typically very conservative from the perspective of radiation exposure such that changes in the fuel inventory/bundle surface dose rate of 14.29% results in inconsequential changes in operating dose. The current Browns Ferry radiation procedures and radiation monitoring program would detect any changes in radiation levels and initiate appropriate actions. Therefore, the radiation levels around the SFP meet all CLTR dispositions. 2.5.3.1.4 Fuel Racks The fuel racks at Browns Ferry are generically addressed in the Section 6.3.4 of the CLTR. The increase in decay heat from EPU results in a higher heat load in the fuel pool during long-term storage. The fuel racks are designed for higher temperatures (212ºF) than the licensing limit of 150ºF. The fuel racks at Browns Ferry are confirmed to be consistent with the generic description provided in the CLTR because the fuel racks design temperature is greater than the licensing limit. Conclusion TVA has evaluated the spent fuel pool cooling and cleanup system and accounted for the effects of the proposed EPU on the spent fuel pool cooling function of the system. The evaluation concludes that the system will continue to provide sufficient cooling capability to cool the spent fuel pool following implementation of the proposed EPU and will continue to meet the requirements of draft GDCs-4, 67, and 69. Therefore, the proposed EPU is acceptable with respect to the spent fuel pool cooling and cleanup system. 2-220

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.5.3.2 Station Service Water Systems Regulatory Evaluation The station service water system provides essential cooling to safety-related equipment and may also provide cooling to non-safety-related auxiliary components that are used for normal plant operation. The NRCs acceptance criteria are based on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, including flow instabilities and loads (e.g., water hammer), maintenance, testing, and postulated accidents; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided. Specific NRC review criteria are contained in SRP Section 9.2.1, as supplemented by GL 89-13 and GL 96-06. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-4, 40 and 42. 2-221

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The Browns Ferry design includes three open loop cooling water systems. The Raw Cooling Water System supplies water to the Reactor and Turbine Buildings for cooling. The Plant Service Water System is described in Browns Ferry UFSAR Section 10.7, Raw Cooling Water System. The Residual Heat Removal Service Water System is provided to remove the heat rejected by the residual heat removal system during normal shutdown and accident operations. In addition this system provides a source of water for the Emergency Equipment Cooling Water System. The Residual Heat Removal Service Water System is described in Browns Ferry UFSAR Section 10.9, RHR Service Water System. The Emergency Equipment Cooling Water System is provided to remove the heat rejected by the equipment that must operate under accident conditions. The Emergency Equipment Cooling Water System is described in Browns Ferry UFSAR Section 10.10, Emergency Equipment Cooling Water System. Browns Ferrys current licensing basis regarding GL 89-13 is discussed in TVAs response to the NRC by letter dated March 16, 1990, Response to Generic Letter 89-13 Service Water Problems Affecting Safety-Related Equipment. Browns Ferrys current licensing basis regarding GL 96-06 is discussed in TVAs response to the NRC, Browns Ferry Revision 1-Response to Generic Letter 96 Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions, dated October 23, 1997. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Emergency Service Water System and Residual Heat Removal Service Water System is documented in NUREG-1843, Section 2.3.3.3. The license renewal evaluation associated with the Plant Service Water System is documented in NUREG-1843, Section 2.3.3.5. Management of aging effects on the Emergency Equipment Cooling Water System, Residual Heat Removal Service Water System, and Plant Service Water System is documented in NUREG-1843, Section 3.3.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.4 of the CLTR addresses the effect of EPU on Water System Performance. The results of this evaluation are described below. The EECW system includes pumps, valves, piping and instrumentation to provide cooling water from the Ultimate Heat Sink to safety-related plant equipment and backup cooling water to non-essential plant equipment. The EECW system is safety-related and is designed to operate during design basis events. The EECW System provides backup cooling flow to the RCW System. The EECW supply valves to the RCW System automatically isolate on low EECW header pressure to guarantee adequate flow to the essential components. 2-222

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The non-safety-related RCW system provides screened and chemically treated once through cooling water to various non-safety-related plant systems, components, and space coolers. The RCW system may also be operated during loss of power conditions only when standby diesel-generated power reserve margin is available. The RCW system includes pumps, valves, piping and instrumentation that provide cooling water to various non-safety-related systems and components, including the turbine-associated equipment heat exchangers and RBCCW heat exchangers. The non-safety-related RSW system supplies river water for yard-watering, cooling for plant equipment which the RCW system may not conveniently serve, and to function as a keep-fill system for the raw water Fire Protection System. The RHRSW system pumps and associated piping and valves are safety-related and provide cooling water from the Ultimate Heat Sink to the RHR heat exchangers during normal shutdown, flood conditions, and during post-accident conditions (LOCA). Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Water Systems Performance (Safety-Related) Plant Specific Disposition Meets CLTR Water Systems Performance (Normal Operation) Plant Specific Disposition Meets CLTR Suppression Pool Cooling (RHR Service Operation) Plant Specific Disposition 2.5.3.2.1 Water System Performance (Safety-Related) As explicitly stated in Section 6.4 of the CLTR, EPU results in a greater decay heat rate which increases the safety-related water systems cooling requirement during accident conditions. The performance of safety-related service water systems during and immediately following the most limiting design basis event, the LOCA, is not significantly affected by reactor power. For DBA-LOCA conditions, the RHRSW heat loads will increase slightly due to an increase in maximum suppression pool temperature from 172.1 to 179.0 for EPU. For normal shutdown, the maximum RHRSW heat loads will not increase for EPU because the associated pressure and temperature process conditions for normal shutdown cooling are not changing from CLTP to EPU. The safety-related portions of the RHRSW and Emergency Equipment Cooling Water Systems are designed to provide a reliable supply of cooling water during and following a DBA, design 2-223

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) basis flood, or loss of offsite power conditions, for the following essential equipment and systems: Services which have increased heat loads with EPU: RHR Heat Exchangers RBCCW Heat Exchangers* RHR Pumps Room Coolers CS Pump Room Coolers Services for which heat loads are not dependent on RTP: Emergency Diesel Generator (EDG) Heat Exchangers (Jacket Water, Air, and Lube Oil Coolers) Standby Coolant Supply System (Emergency RHRSW cross-connect to RHR system to provide reactor core or primary containment cooling if RHR is lost) Supplemental Cooling to SFP Makeup flow to the SFP* Unit 3 Electric Board Room air conditioning unit Unit 3 Control Bay Chillers Unit 3 Shutdown Board Room Chillers Control Air Compressors* Unit 1/2 Emergency Condensing Unit*

   *Denotes non-essential load The increase in heat load to the RHR Pump Room Coolers and CS Room Coolers is a result of a post-LOCA increase in room temperature in each area. This increase in room temperature will slightly increase the EECW discharge temperature but will not be significant as the room temperatures increase is negligible.

The increase in heat load to the RBCCW Heat Exchangers results in a negligible temperature increase. Control Air Compressors and RBCCW heat exchangers are normally serviced by non-safety-related RCW. EECW provides backup water in the case of a RCW failure. These loads isolate on low EECW header pressure to ensure flow to the essential EECW loads. The EECW system flow rates, and thereby flow velocities, remain unchanged due to EPU. The EECW and RHRSW systems were evaluated for changes due to EPU and are adequate as currently designed. Therefore, the EECW and RHRSW meet all CLTR dispositions. 2-224

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.5.3.2.2 Water System Performance (Normal Operation) As stated in Section 6.4 of the CLTR, EPU results in an increased heat load during normal operation. The increased non-safety related RCW system heat loads at EPU are due primarily to the increase in heat loads from the isolated phase bus duct heat exchanger(s), certain turbine building pump area coolers and condensate booster pump motor cooler(s). Plant modifications to rerate the main generator have been implemented to accommodate EPU. The main generator stator cooling water and hydrogen cooler heat loads for the uprated main generator (See PUSAR Section 2.5.1.2.2) are bounded by heat loads for the original generator rating. This is because the uprated generator reactive power output, the primary contributor to stator cooling and hydrogen cooler heat load, is constrained to be less than the reactive power output of the original generator rating. Additionally, the Isophase Bus Duct modifications increased the RCW flow. The RCW system is capable of providing the additional flow. With these increased heat loads, the RCW system discharge temperature increases approximately 0.1°F at EPU RTP. Therefore, the RCW system is expected to meet the requirements of the system with respect to heat loads and flow due to EPU because the RCW system temperature increase at EPU is negligible. There are no power dependent loads on the RSW system, and therefore there are no heat load increases due to EPU. Therefore, RCW and RSW performance during normal operation meets all CLTR dispositions. 2.5.3.2.3 Suppression Pool Cooling (RHR Service Water Operation) As stated in Section 6.4 of the CLTR, EPU results in a greater decay heat rate. The containment cooling analysis in Section 2.6.5 shows that the post-LOCA RHR heat load increases due in part to an increase in reactor decay heat. The post-LOCA containment and suppression pool responses have been calculated based on an energy balance between the post-LOCA heat loads and the heat removal capacity of the RHR and RHRSW. The containment cooling analysis and equipment review demonstrate that the suppression pool temperature can be maintained within acceptable limits in the post-accident condition at EPU based on the existing capability of the RHRSW system. The EPU post-accident containment system response results in an increase in the maximum Suppression Pool temperature from 172.1 to 179 . The containment cooling analysis results in a total heat load rejected to the RHRSW system due to post-accident suppression pool cooling of 74.2 MBtu/hr/in-service RHR Heat Exchanger. The maximum RHRSW fluid outlet temperature during suppression pool cooling from the RHR heat exchangers will increase to 133.4 , which remains below the 150 design temperature for the RHRSW discharge piping. The RHRSW system transfers heat to the ultimate heat sink (UHS), which is addressed in Section 2.5.3.4. Therefore, Suppression Pool Cooling meets all CLTR dispositions. 2-225

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Conclusion TVA has evaluated the effects of the proposed EPU on the station service water system including any increased heat loads on system performance that would result from the proposed EPU. The evaluation indicates that the station service water systems will continue to provide sufficient cooling for SSCs important to safety following implementation of the proposed EPU. Therefore, the station service water systems will continue to meet the requirements of draft GDCs-4, 40 and

42. Additionally, the Browns Ferry GL 89-13 Program (i.e., scope, maintenance, and testing) to manage and monitor raw water cooling systems and the Browns Ferry GL 96-06 Program to ensure equipment operability and containment integrity during design basis accident conditions, are not affected by the proposed EPU. Based on the above, the proposed EPU is acceptable with respect to the station service water systems.

2.5.3.3 Reactor Auxiliary Cooling Water Systems Regulatory Evaluation These systems include closed-loop auxiliary cooling water systems for reactor system components, reactor shutdown equipment, ventilation equipment, and components of the ECCS. The NRCs acceptance criteria for the reactor auxiliary cooling water system are based on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation including flow instabilities and attendant loads (i.e., water hammer), maintenance, testing, and postulated accidents; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided. Specific NRC review criteria are contained in SRP Section 9.2.2, as supplemented by GL 89-13 and GL 96-06. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding 2-226

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, with the exception of final GDC-44, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-4, 40 and 42. The Browns Ferry design includes one closed loop cooling water system. The Reactor Building Closed Cooling Water System is designed to remove heat from the reactor auxiliary systems equipment and their accessories. The Reactor Building Closed Cooling Water System is described in Browns Ferry UFSAR Section 10.6, Reactor Building Closed Cooling Water System. Browns Ferrys current licensing basis regarding GL 89-13 is discussed in TVAs response to the NRC by letter dated March 16, 1990, Response to Generic Letter 89-13 Service Water Problems Affecting Safety-Related Equipment. Browns Ferrys current licensing basis regarding GL 96-06 is discussed in TVAs responses to the NRC, Browns Ferry Revision 1-Response to Generic Letter 96 Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions, dated October 23, 1997 and Browns Ferry Nuclear Plant (BFN) Unit 1 - Generic Letter 96-06, Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions, dated May 12, 2004. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Reactor Building Closed Cooling Water System is documented in NUREG-1843, Section 2.3.3.22. Management of the effects of aging on the RBCCW system is documented in NUREG-1843, Section 3.3.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.4 of the CLTR addresses the effect of EPU on Water Systems. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: 2-227

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR Water Systems Performance (Non-Safety Related) Plant Specific Disposition The non-safety-related Reactor Auxiliary Cooling Water system includes the RBCCW system. The safety-related and normal operation water systems are evaluated in Section 2.5.3.2, Station Service Water Systems. Reactor Building Closed Cooling Water System EPU increases the heat loads on the RBCCW system, but due to an overly conservative analysis that was performed for CLTP, the computed heat load for EPU is decreased. The RBCCW heat loads are mainly dependent on the reactor vessel temperature and/or flow rates in the systems cooled by the RBCCW. The flow rates in the RBCCW system do not change due to EPU. The only component heat load increase at EPU conditions is an estimated 6.5% increase in Reactor Recirculation Pump and motor heat load. The remaining heat loads remain the same or decrease due to excessive conservatism in the CLTP heat load analysis. There are negligible changes to system operating temperatures and pressures as a result of EPU. There are no changes to RBCCW System operation. The RBCCW system contains sufficient redundancy in pumps and heat exchangers to ensure that adequate heat removal capability is available during normal operation. Therefore, RBCCW meets all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on the reactor auxiliary cooling water systems including any increased heat loads from the proposed EPU on system performance. The evaluation indicates that the reactor auxiliary cooling water systems will continue to provide sufficient cooling for SSCs important to safety following implementation of the proposed EPU. Therefore, the reactor auxiliary cooling water systems will continue to meet the requirements of draft GDCs-4, 40 and 42. Additionally, the Browns Ferry GL 89-13 Program (i.e., scope, maintenance, and testing) to manage and monitor raw water cooling systems and the Browns Ferry GL 96-06 Program to ensure equipment operability and containment integrity during design basis accident conditions, are not affected by the proposed EPU. Based on the above, the proposed EPU is acceptable with respect to the reactor auxiliary cooling water systems. 2.5.3.4 Ultimate Heat Sink Regulatory Evaluation The UHS is the source of cooling water provided to dissipate reactor decay heat and essential cooling system heat loads after a normal reactor shutdown or a shutdown following an accident. The NRCs acceptance criteria for the UHS are based on (1) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that 2-228

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) sharing will not significantly impair their ability to perform their safety functions; and (2) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided. Specific NRC review criteria are contained in SRP Section 9.2.5. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, with the exception of final GDC-44, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-4. There is no draft GDC directly associated with final GDC-44. The Wheeler Reservoir/Tennessee River serves as the ultimate heat sink for the plant. The ultimate heat sink temperature limit is described in Browns Ferry UFSAR Section 14.6.3.3.2.3, Long-Term Response. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.4 of the CLTR addresses the effect of EPU on the UHS. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: 2-229

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR Ultimate Heat Sink Plant Specific Disposition The Browns Ferry UHS is the Wheeler Reservoir/Tennessee River. The maximum allowable supply temperature from the UHS is 95 , which is governed by the limits established in TS 3.7.2. The UHS temperature limit is not affected by EPU. EPU will have no effect on the UHS as a source of cooling water for EECW and RHRSW systems, which dissipate reactor decay heat and essential cooling loads during normal or emergency reactor shutdowns. The Browns Ferry design includes an UHS which provides heat removal capability for safe reactor shutdown in the event of the site related natural phenomena and failures of man-made structures associated with the safety evaluation of the UHS. The UHS safety function is to provide sufficient cooling water to support 1 accident unit and 2 units in shutdown for at least 30 days. The UHS must remain capable of withstanding the following events without loss of safety function: the most severe single natural phenomena expected at the site, the site-related event (e.g., transportation accident, river diversion), reasonable combinations of less severe natural phenomena and/or site related event, or a single failure of a manmade structure. The EECW and RHRSW systems use the UHS to provide cooling water during accident and shutdown. They are capable of meeting their requirements at EPU with this heat sink. As explicitly stated in Section 6.4 of the CLTR, EPU results in increased heat load during normal operation and a greater decay heat rate, which increases the safety-related water systems cooling requirements during accident conditions. The RHR heat exchanger heat load increase, along with other smaller increases discussed in Section 2.5.3.2 must be accommodated by the UHS at EPU. The UHS is operated so that the present limits (e.g., UHS maximum temperature, minimum Wheeler Reservoir level, and minimum Tennessee River flow rate) are not changed or exceeded as a result of EPU. The UHS was evaluated for its capability to handle the increased EPU heat load. The evaluation demonstrates that UHS can maintain the cooling water supplied within the design basis minimum water level and minimum flow rate. EPU has no effect on the UHS design function. Therefore, UHS meets all CLTR dispositions. Conclusion The effects that the proposed EPU would have on the UHS safety function have been reviewed. The proposed EPU will not compromise the design basis safety function of the UHS. The UHS will continue to satisfy the requirements of the current licensing basis following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the UHS. 2-230

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.5.4 Balance-of-Plant Systems 2.5.4.1 Main Steam Regulatory Evaluation The main steam supply system (MS) transports steam from the NSSS to the power conversion system and various safety-related and non-safety-related auxiliaries. The NRCs acceptance criteria for the MS are based on (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects, including the effects of missiles, pipe whip, and jet impingement forces associated with pipe breaks; and (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions. Specific NRC review criteria are contained in SRP Section 10.3. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-4 and 40. Main steam piping is discussed in several UFSAR sections including Chapter 4, Reactor Coolant System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 2-231

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) (Reference 11). The license renewal evaluation associated with the main steam piping is documented in NUREG-1843, Section 2.3.4. Management of the effects of aging on the main steam piping is documented in NUREG-1843, Section 3.4.2. Technical Evaluation The heat balance for the EPU conditions is provided in Section 1.3. The heat balance shows the transport of steam to the power conversion equipment, the heat sink, and to steam driven components. Flow induced vibration and structural loading of the MS system piping and supports is addressed in Sections 2.2.2. Dynamic loading is discussed below. SRV dynamic loads are discussed in Sections 2.2.2 and 2.2.3. The function and capability of the MSIVs are discussed in Section 2.2.2. SRV setpoint tolerance and FIV effects are discussed below. 2.5.4.1.1 Structural Evaluation of Main Steam Piping NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 3.4.1 of the CLTR addresses the effect of EPU on flow induced vibration in the MSL. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Structural Evaluation of Main Steam Piping Generic Disposition The CLTR states that because the MS piping pressures and temperatures are not affected by EPU, there is no effect on the analyses for these parameters. Seismic inertia loads, seismic building displacement loads, and SRV discharge loads are not affected by EPU, thus, there is no effect on the analyses for these load cases. The increase in MS flow results in increased forces from the turbine stop valve closure transient. The turbine stop valve closure loads bound the MSIV closure loads because the MSIV closure time is significantly longer than the stop valve closure time. The capability of the MS piping to withstand dynamic loads at EPU conditions was evaluated. A summary of the results of the MS piping system evaluation that contains the increased loading associated with EPU conditions (i.e., temperature, pressure, and flow, including the effects of the MS flow induced transient loads at EPU conditions) along with a comparison to the code allowable limits is provided in Section 2.2.2. 2-232

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) SRV setpoint tolerance is independent of an EPU. Browns Ferry transient analyses conservatively bound the existing SRV setpoint tolerance ALs. Actual historical in-service surveillance of SRV setpoint performance test results are monitored separately for compliance to the TSs and In-Service Testing program. Browns Ferry has an ongoing evaluation program to resolve problems resulting in SRV surveillance testing exceeding the 3% tolerance. Increased MSL flow may affect vibration of the piping during normal operation. The vibration frequency, extent, and magnitude depend upon plant-specific parameters, valve locations, the valve design, and piping support arrangements. The effects of EPU on Flow-Induced Vibration (FIV) of the piping will be assessed by vibration testing during initial plant operation at the higher steam flow rates. This topic is addressed in Section 2.2.2.1.2. Attachment 45 to the EPU license amendment request contains details of the vibration monitoring program. FIV may increase incidents of SRV leakage. Browns Ferry currently has procedures and installed instrumentation in place to detect and take actions concerning SRV seat leakage. These procedures and installed instrumentation are considered acceptable to monitor for SRV seat leakage at EPU rated steam flow conditions. TVA has conducted drywell vibration studies directly related to SRV standpipes and branch connections and the effects of acoustic resonance. This has resulted in installation of acoustic vibration suppressors. This is to ensure that SRV vibration resulting from acoustic resonance is not expected at EPU operating conditions. Therefore, the structural evaluation of MS piping meets all CLTR dispositions. 2.5.4.1.2 Main Steam Line Flow Restrictors NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 3.7 of the CLTR addresses the effect of EPU on the MSL flow restrictors. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Structural Integrity Generic Disposition The CLTR states that at uprated power, the flow restrictors are required to pass a higher flow rate, which will result in an increased pressure drop. The increase in steam flow rate has no significant effect on flow restrictor erosion. There is no effect on the structural integrity of the MSL flow element (restrictor) due to the increased differential pressure because the restrictors were designed and analyzed for the choke flow condition. 2-233

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) After a postulated steam line break outside containment, the fluid flow in the broken steam line increases until it is limited by the MSL flow restrictor. ((

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The Browns Ferry restrictors were originally analyzed for these flow conditions and therefore the restrictors remain within the acceptable calculated differential pressure drop and choke flow limits under EPU conditions. Therefore, the flow restrictors meet all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on MS including the effects of changes in plant conditions on the design of MS. The evaluation indicates that the system will continue to meet the requirements of draft GDCs-4 and 40. Therefore, the proposed EPU is acceptable with respect to MS. 2.5.4.2 Main Condenser Regulatory Evaluation The main condenser system is designed to condense and deaerate the exhaust steam from the main turbine and provide a heat sink for the turbine bypass system. For BWRs without an MSIV leakage control system, the main condenser system may also serve an accident mitigation function to act as a holdup volume for the plate out of fission products leaking through the MSIVs following core damage. The NRCs acceptance criteria for the main condenser system are based on GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents. Specific NRC review criteria are contained in SRP Section 10.4.1. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a 2-234

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDC-70. The main condenser system is described in Browns Ferry UFSAR Section 11.3, Main Condenser System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the main condenser system is documented in NUREG-1843, Section 2.3.4. The management of the effects of aging on the main condenser system is documented in NUREG-1843, Section 3.4.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 7.2 of the CLTR addresses the effect of EPU on the Condenser and Steam Jet Air Ejectors. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Condenser and SJAE Plant Specific Disposition As stated in the CLTR, the increase in steam flow increases the heat removal requirement for the condenser. The additional power level increases the non-condensable gases generated by the reactor. The main condenser is designed to reject heat to the circulating water system and thereby maintain adequately low condenser pressure as recommended by the turbine vendor. Maintaining adequately low condenser pressure assures the efficient operation of the turbine-generator and minimizes wear on the turbine last stage blades. EPU operation increases the heat rejected to the condenser and, therefore, reduces the difference between the operating backpressure and the recommended maximum condenser backpressure. If condenser backpressures approach the main turbine backpressure limitation, then reactor thermal 2-235

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) power reduction would be required to reduce the heat rejected to the condenser and maintain condenser pressure within the turbine requirements. The main condenser is not being modified for EPU operation. The performance of the condenser was evaluated for EPU. This evaluation was based on a design duty over the actual range of circulating water inlet temperatures, and confirms that the condenser backpressure remains below the high alarm setpoint, and the turbine trip setpoint during normal operation. Condenser backpressure limitations may require load reductions at the upper range of the anticipated circulating water inlet temperatures. Main condenser storage capacity has been evaluated for hotwell retention time and found to be acceptable for EPU operation. The holdup time for the decay of short-lived radioisotopes (primarily N-16) remains a conservative decay time and is acceptable for EPU operation. The absolute value in lbm/hr of the steam bypassed to the main condenser during a load rejection event is not increased for EPU as discussed in FUSAR Section 2.5.4.2. Therefore, the Condenser and Steam Jet Air Ejectors for Browns Ferry meet all CLTR dispositions. Conclusion TVA has considered the effects of the proposed EPU with ATRIUM 10XM fuel on the main condenser system. It is concluded that the main condenser system will continue to maintain its ability to withstand the blowdown effects of the steam from the TBS and thereby continue to meet the current licensing basis with respect to controlling releases of radioactive effluents. Therefore, the proposed EPU with ATRIUM 10XM fuel is acceptable with respect to the main condenser system. 2.5.4.3 Turbine Bypass Regulatory Evaluation The TBS is designed to discharge a stated percentage of rated main steam flow directly to the main condenser system, bypassing the turbine. This steam bypass enables the plant to take step-load reductions up to the TBS capacity without the reactor or turbine tripping. The system is also used during startup and shutdown to control reactor pressure. For a BWR without an MSIV leakage control system, the TBS could also provide an accident mitigation function. The TBS, along with the main steam supply system and main condenser system, may be credited for mitigating the effects of MSIV leakage during a LOCA by the holdup and plate out of fission products. The NRCs acceptance criteria for the TBS are based on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents (including pipe breaks or malfunctions of the TBS), and (2) GDC-34, insofar as it requires that a RHR system be provided to transfer fission product decay heat and 2-236

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) other residual heat from the reactor core at a rate such that SAFDLs and the design conditions of the RCPB are not exceeded. Specific NRC review criteria are contained in SRP Section 10.4.4. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-40 and 42. There is no draft GDC directly associated with final GDC-34. The TBS is described in Browns Ferry UFSAR Section 11.5, Turbine Bypass System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The TBS is included in the discussion of the license renewal evaluation for the Main Steam System. That discussion can be found in NUREG-1843, Section 2.3.4. Management of aging effects on the Main Steam System is documented in NUREG-1843, Section 3.4.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the 2-237

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) effects of EPUs. Section 7.3 of the CLTR addresses the effect of EPU on the Turbine Bypass System. The results of this evaluation are described below. The Turbine Steam Bypass System provides a means of accommodating excess steam generated during normal plant maneuvers and transients. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Turbine Steam Bypass (Safety Analysis) Generic Disposition The CLTR states that the increase in steam flow reduces the relative capacity of the Turbine Steam Bypass System. See FUSAR Section 2.5.4.3 for the TBS safety analysis effect. The Turbine Steam Bypass system provides a means of accommodating excess steam generated during normal plant maneuvers and transients. The turbine bypass valves are rated for a total steam flow capacity of not less than 25% of the rated reactor steam flow, or 3.5 Mlbm/hr. Each of nine bypass valves is designed to pass a steam flow of 389,000 lbm/hr and does not change at EPU RTP. At EPU conditions, rated reactor steam flow is 16.44 Mlbm/hr, resulting in a bypass capacity of 21.3% of EPU rated steam flow. The bypass capacity at Browns Ferry remains adequate for normal operational flexibility at EPU RTP. The bypass capacity is used as an input to the reload analysis process for the evaluation of transient events that credit the Turbine Steam Bypass System. Therefore, the Browns Ferry steam bypass capacity used in the turbine steam bypass safety analysis meets all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on the TBS. The evaluation indicates that the same absolute value of steam flow bypass capacity will exist at EPU. The relative bypass capability with respect to rated steam flow at EPU conditions is reduced slightly. The TBS will continue to provide a means of accommodating excess steam generation during normal plant maneuvers and transients. Therefore, the proposed EPU is acceptable with respect to the TBS. 2.5.4.4 Condensate and Feedwater Regulatory Evaluation The condensate and feedwater system provides feedwater at a particular temperature, pressure, and flow rate to the reactor. The only part of the condensate and feedwater system classified as 2-238

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) safety-related is the feedwater piping from the NSSS up to and including the outermost containment isolation valve. The NRCs acceptance criteria for the condensate and feedwater system are based on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation including possible fluid flow instabilities (e.g., water hammer), maintenance, testing, and postulated accidents; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-44, insofar as it requires that a system with the capability to transfer heat loads from safety-related SSCs to a heat sink under both normal operating and accident conditions be provided, and that the system be provided with suitable isolation capabilities to assure the safety function can be accomplished with electric power available from only the onsite system or only the offsite system, assuming a single failure. Specific NRC review criteria are contained in SRP Section 10.4.7. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, with the exception of final GDC-44, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-4, 40 and 42. There is no draft GDC directly associated with final GDC-44. 2-239

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The condensate and feedwater system is described in Browns Ferry UFSAR Section 11.8, Condensate and Reactor Feedwater Systems. The condensate demineralizer system is described in Browns Ferry UFSAR Section 11.7, Condensate Filter-Demineralizer System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the condensate and feedwater system is documented in NUREG-1843, Section 2.3.4. The management of the effects of aging on the condensate and feedwater system is documented in NUREG-1843, Section 3.4.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 7.4 of the CLTR addresses the effect of EPU on the Condensate and FW Systems. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR FW and Condensate Systems Plant Specific Disposition The CLTR states that the increase in power level increases the FW requirements of the reactor. The FW and condensate systems are required for normal plant operation and are not safety-related. The FW and condensate systems do not perform a system level safety-related function, and are designed to provide a reliable supply of FW at the temperature, pressure, quality, and flow rate as required by the reactor. However, their performance has a major effect on plant availability and capability to operate at EPU conditions. Normal Operation System operating flows at EPU increase approximately 16% of rated flow at the CLTP. The condensate and FW systems will be modified to ensure acceptable performance with the new system operating conditions. See LAR Attachment 47 for modifications description. Transient Operation To account for FW demand transients, the FW system was evaluated to ensure that a minimum of 5% margin above the EPU FW flow was available. For system operation with all system pumps available, the predicted operating parameters were acceptable and within the component capabilities. 2-240

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The FW system post-feed pump trip capacity was evaluated to confirm that with the modifications to the FW and condensate system configurations, the capability to supply the transient flow requirements is maintained or increased. A transient analysis was performed to determine the reactor level response following a single FW pump trip. The results of the analysis in FUSAR Section 2.8.5.2.3.2 show that the system response is adequate during EPU conditions. Condensate Demineralizers The condensate filter demineralizers (CFD) are acceptable for EPU. The system experiences slightly higher loadings resulting in slightly reduced CFD run times. However, the reduced run times are acceptable (refer to Section 2.5.5 for the effects on the radwaste systems). Therefore, the FW and condensate systems meet all CLTR dispositions. Conclusion TVA has evaluated the effects of the proposed EPU on the condensate and feedwater system. The evaluation indicates that the condensate and feedwater systems will continue to meet their performance requirements following modifications to several non-safety-related components. Additionally, the modified condensate and feedwater pumps will provide a minimum of 5 percent margin above the EPU rated flow to account for feedwater transients. Therefore, the proposed EPU is acceptable with respect to the condensate and feedwater system. 2.5.5 Waste Management Systems 2.5.5.1 Gaseous Waste Management Systems Regulatory Evaluation The gaseous waste management systems involve the gaseous radwaste system, which deals with the management of radioactive gases collected in the offgas system or the waste gas storage and decay tanks. In addition, it involves the management of the condenser air removal system; the gland seal exhaust and the mechanical vacuum pump operation exhaust; and the building ventilation system exhausts. The NRCs acceptance criteria for gaseous waste management systems are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not exceed specified values; (2) GDC-3, insofar as it requires that (a) SSCs important to safety be designed and located to minimize the probability and effect of fires, (b) noncombustible and heat resistant materials be used, and (c) fire detection and fighting systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; (3) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; (4) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate confinement; and (5) 10 CFR Part 50 Appendix I, Sections II.B, II.C, and II.D, which set numerical guides for design objectives and limiting conditions for operation to meet the "as low as is reasonably achievable" (ALARA) criterion. 2-241

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Specific NRC review criteria are contained in SRP Section 11.3. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-69 and 70. Final GDC-3 is applicable to Browns Ferry as described in the Browns Ferry Fire Protection Report, Volume 1, Revision 20. The gaseous waste management system is described in Browns Ferry UFSAR Section 9.4, Gaseous Radwaste System and Section 9.5, Gaseous Radwaste System (Modified). Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 8.2 of the CLTR addresses the effect of EPU on Gaseous Waste Management. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: 2-242

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR Offsite Release Rate Generic Disposition Meets CLTR Recombiner Performance Generic Disposition 2.5.5.1.1 Offsite Release Rate The CLTR states that under EPU conditions, offgas system functions other than the recombiner and related components, (( )) The Browns Ferry site-specific CLTP design basis radiolytic gas production rate, 0.070 cfm/MWt, is greater than or equal to ((

)).

These are constant rates. As these rates are proportional to reactor power in each unit, the radiolytic gas flow rate is expected to increase in proportion to the change in power, approximately 20% under EPU conditions as compared to OLTP. Because the actual radiolytic gas flow rate at EPU conditions is within the design basis (radiolytic gas) flow rate at OLTP, the design basis production value is acceptable at EPU conditions. As such, the OLTP design basis is maintained at EPU conditions and an evaluation was conducted. This evaluation verified that all structures, systems and components of the offgas system were acceptable for EPU operation. The primary function of the gaseous waste management system is to process and control the release of gaseous radioactive effluents to the site environs so that the total radiation exposure of persons in offsite areas is within the guideline values of 10 CFR 50 Appendix I. The offgas system radiological release rate is administratively controlled to remain within existing site release rate limits and is a function of fuel cladding performance, main condenser air in-leakage, charcoal adsorber inlet dew point, and charcoal adsorber temperature. The Browns Ferry TS require administrative controls (i.e., Radioactive Effluent Controls Program) to limit radioactive gas releases to the environment. These controls require plant procedures for addressing fuel cladding failure or high activity in offgas. Such procedures are not affected by EPU. Further information regarding the production of noble gases at EPU conditions is found in Section 2.9.1.2. The gaseous waste management system (offgas system) design criteria ensure that it will meet the plant licensing basis for controlling gaseous waste such that the total radiation exposure of persons in offsite areas will be within the applicable guideline values of 10 CFR 20.1302 and 10 CFR 50 Appendix I. The plant gaseous waste licensing basis and the gaseous waste management system design criteria (for the offgas portion) that support the licensing basis are 2-243

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) unchanged by EPU. The gaseous waste management system will continue to satisfy this licensing basis under EPU operating conditions. The gaseous waste management system methods of treatment for radiological releases from the offgas system consist of holdup and filtration to reduce the gaseous radioactivity that could be potentially released to offsite areas. The capacity and capability of the offgas holdup and filtration system to adequately perform its design function are unchanged by EPU. The gaseous waste management systems are designed to meet the requirements of 10 CFR 20 and 10 CFR 50 Appendix I in accordance with the Offsite Dose Calculation Manual (ODCM), Reference 74. Browns Ferry compliance with the dose limits to the public of 10 CFR 20 and 10 CFR 50 Appendix I is described in Section 2.10.1.2.4 (Table 2.10-2). The offsite release rate at Browns Ferry meets all CLTR dispositions. 2.5.5.1.2 Recombiner Performance The CLTR states that under EPU conditions, core radiolysis increases linearly with reactor thermal power, thus increasing the heat load on the offgas recombiner and related components. The design features for precluding the possibility of an explosion include: (a) dilution to control the concentration of hydrogen; and (b) catalytic recombination to remove the combustible gas. The gaseous waste management system at Browns Ferry is consistent with GEH design specifications for radiolytic flow rate, and the Browns Ferry-specific value for radiolytic gas production rate is 0.045 cfm/MWt, which is well below the Browns Ferry site specific design value of 0.070 cfm/MWt (130ºF and 1 atm.). Therefore, the recombiner and condenser, as well as downstream system components, are designed to handle the increase in thermal power of the EPU. The gaseous waste management system component design requirements are determined by the quantity of radiolytic hydrogen and oxygen, which is expected to increase in proportion to the EPU power increase. The additional radiolytic hydrogen will also increase the catalytic recombiner temperature and offgas condenser heat load. These increases have been evaluated and it has been confirmed that sufficient margin remains in the Browns Ferry offgas system component design to ensure that the system will continue to satisfy the plant licensing basis. The recombiner performance at Browns Ferry meets all CLTR dispositions. Conclusion TVA has evaluated the gaseous waste management systems and the increase in fission product and amount of gaseous waste on the abilities of the system to control releases of radioactive materials and preclude the possibility of an explosion if the potential for explosive mixtures exists. The evaluation indicates that the gaseous waste management systems will continue to meet their design functions following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the gaseous waste management systems. 2-244

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.5.5.2 Liquid Waste Management Systems Regulatory Evaluation The NRCs acceptance criteria for the liquid waste management systems are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not exceed specified values; (2) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; (3) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate confinement; and (4) 10 CFR Part 50 Appendix I, Sections II.A and II.D, which set numerical guides for dose design objectives and limiting conditions for operation to meet the ALARA criterion. Specific NRC review criteria are contained in SRP Section 11.2. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-69 and 70. The liquid waste management system is described in Browns Ferry UFSAR Section 9.2, Liquid Radwaste System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 2-245

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) (Reference 11). The license renewal evaluation associated with the liquid waste management system is documented in NUREG-1843, Section 2.3.3.25. Management of aging effects on the liquid waste management system is documented in NUREG-1843, Section 3.3.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 8.1 of the CLTR addresses the effect of EPU on Liquid Waste Management. The results of this evaluation are described below. As stated in Section 8.1 of the CLTR, the Liquid Radwaste System collects, monitors, processes, stores and returns processed radioactive waste to the plant for reuse or for discharge. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Waste Volumes Plant Specific Disposition Meets CLTR Coolant Fission and Corrosion Product Levels Plant Specific Disposition 2.5.5.2.1 Waste Volumes The CLTR states that increased power levels and steam flow result in the generation of slightly higher levels of liquid radwaste (about 3.44% for Browns Ferry). The largest sources of liquid waste are from the backwash of condensate and RWCU filter-demineralizers. Other increases in the liquid waste management system (LWMS) loads are minimal. The effect of EPU on the LWMS is primarily a result of the increased load on condensate filter/demineralizers. Similarly, the RWCU filter-demineralizer requires more frequent backwashes due to slightly higher levels of activation and fission products. Because the RWCU flow rate will remain the same as CLTP, but an increase in contaminate concentration is projected, the RWCU system is projected to experience a slight increase in filter demineralizer backwash frequency. The current capacity of the LWMS can accommodate this small increase. Because the liquid volume does not increase appreciably for EPU, the current design and operation of the LWMS will accommodate the effects of EPU with no changes. The offsite concentration of liquid effluents at CLTP and EPU conditions meets 10 CFR 20 and the dose from liquid effluents meets 10 CFR 50 Appendix I, as shown in Table 2.10-2. The existing equipment and procedures that control releases to the environment will continue to ensure that releases remain within the applicable guideline values of 10 CFR 20.1302, 10 CFR 50 2-246

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Appendix I, and 40 CFR 190. Browns Ferry compliance with these dose limits to the public is described in Section 2.10.1.2.4 (Table 2.10-2). Therefore, the waste volumes meet all CLTR dispositions. 2.5.5.2.2 Coolant Fission and Corrosion Product Levels The CLTR states that increased power levels and steam flow result in the generation of slightly higher levels of coolant concentrations of fission and corrosion products. The coolant activation and corrosion products are slightly increased as a result of EPU as discussed in Section 8.4 of the CLTR. Per the AST submittal, a calculation of activated corrosion and fission products in the reactor coolant was performed in accordance with ANSI/ANS-18.1-1984, "Radioactive Source Term for Normal Operation of Light Water Reactors" (Reference 75). Input parameters that change as a result of EPU conditions include core power, weight of water in the reactor vessel, condensate demineralizer flow rate, and steam flow rate. The determination of activated corrosion products in the reactor coolant was performed the same way for all three units. The current design and operation of the LWMS will accommodate the effects of the EPU with no changes. The existing equipment and procedures that control releases to the environment will continue to ensure that releases remain within the applicable guideline values of 10 CFR 20.1302, 10 CFR 50 Appendix I, and 40 CFR 190. Conclusion TVA has evaluated the liquid waste management systems including the effects of the increase in fission product and amount of liquid waste on the ability of the liquid waste management systems to control releases of radioactive materials. The evaluation indicates that the liquid waste management systems will continue to meet their design functions following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the liquid waste management systems. 2.5.5.3 Solid Waste Management Systems Regulatory Evaluation The NRCs acceptance criteria for the solid waste management system are based on (1) 10 CFR 20.1302, insofar as it provides for demonstrating that annual average concentrations of radioactive materials released at the boundary of the unrestricted area do not exceed specified values; (2) GDC-60, insofar as it requires that the plant design include means to control the release of radioactive effluents; (3) GDC-63, insofar as it requires that systems be provided in waste handling areas to detect conditions that may result in excessive radiation levels; (4) GDC-64, insofar as it requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including 2-247

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) AOOs, and postulated accidents; and (5) 10 CFR Part 71, which states requirements for radioactive material packaging. Specific NRC review criteria are contained in SRP Section 11.4. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-17, 18, and 70. The solid waste management system is described in Browns Ferry UFSAR Section 9.3, Solid Radwaste System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry license renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the solid waste management system is documented in NUREG-1843, Section 2.3.3.25. Management of aging effects on the solid waste management system is documented in NUREG-1843, Section 3.3.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 8.1 of the CLTR addresses the effect of EPU on Solid Waste Management. The results of this evaluation are described below. 2-248

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The Solid Radwaste System collects, monitors, processes, and stores processed radioactive waste prior to offsite disposal. Browns Ferry meets all CLTR dispositions. The topics considered in this section are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Coolant Fission and Corrosion Product Levels Plant Specific Disposition Meets CLTR Waste Volumes Plant Specific Disposition 2.5.5.3.1 Coolant Fission and Corrosion Product Levels EPU does not change the types of solid radwaste which are generated, or add a new type of solid radwaste, as there are no new inputs being added to the radwaste system, and the radwaste system will not be modified as part of the EPU. The primary source of solid radwaste is in the form of spent resins. However, the resin is replaced based on pressure drop across the demineralizer and conductivity design criteria prior to exceeding radiological criteria. Therefore, any increase in the primary coolant activity will not significantly increase the activity of the spent resin. The existing equipment and procedures that control waste shipments and releases to the environment will continue to ensure that releases remain within the applicable regulatory guidance. 2.5.5.3.2 Waste Volumes The CLTR states that increased power levels and steam flow result in the generation of slightly higher levels of liquid and solid radwaste. The effect of EPU on the Solid Waste Management System (SWMS) is primarily a result of the increased load on condensate filter/demineralizers. The result is that the increase in solid radwaste volume is conservatively considered as up to 15%. Based on previous EPU experience from other plants, there is enough margin between the actual solid radwaste volume and design basis volume to accommodate this increase. The EPU projected usage of the Browns Ferry SWMS process capacity will be approximately 50% of the installed processing capacity. EPU does not generate a new type of waste or create a new waste stream. Therefore, the types of radwaste that require shipment are unchanged. Because the solid volume does not increase appreciably, the current design and operation of the SWMS will accommodate the effects of EPU with no changes, and the existing equipment and procedures that control waste shipments and releases to the environment will continue to ensure that releases remain within the applicable regulatory guidance. Therefore, the waste volumes meet all CLTR dispositions. 2-249

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Conclusion TVA has evaluated the effects of the increase in fission product and amount of solid waste on the ability of the solid waste management system to process the waste. The evaluation indicates that the solid waste management system will continue to meet its design functions following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to the solid waste management system. 2.5.6 Additional Considerations 2.5.6.1 Emergency Diesel Engine Fuel Oil Storage and Transfer System Regulatory Evaluation Nuclear power plants are required to have redundant onsite emergency power supplies of sufficient capacity to perform their safety functions (e.g., power diesel engine-driven generator sets), assuming a single failure. The NRCs acceptance criteria for the emergency diesel engine fuel oil storage and transfer system are based on (1) GDC-4, insofar as it requires that SSCs important to safety be protected against dynamic effects, including missiles, pipe whip, and jet impingement forces associated with pipe breaks; (2) GDC-5, insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions; and (3) GDC-17, insofar as it requires onsite power supplies to have sufficient independence and redundancy to perform their safety functions, assuming a single failure. Specific NRC review criteria are contained in SRP Section 9.5.4. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. 2-250

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-4 and 40. Final GDC-17 is applicable to Browns Ferry as described in UFSAR Section 8.3. The Diesel Engine Fuel Oil Storage and Transfer capability is described in Browns Ferry UFSAR Section 8.5, Standby AC Power Supply and Distribution. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the Diesel Engine Fuel Oil Storage and Transfer capability is documented in NUREG-1843, Section 2.3.3.2. Management of aging effects on the Diesel Engine Fuel Oil Storage and Transfer capability is documented in NUREG-1843, Section 3.3.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.8 of the CLTR addresses the effect of EPU on other systems not addressed in the CLTR. It concludes that systems not specifically addressed in the CLTR are not significantly affected by the power uprate. The Emergency Diesel Engine Fuel Oil Storage and Transfer System is not addressed in the CLTR, and this disposition applies to Browns Ferry. There are no changes to the EDG loads for EPU. EPU conditions are achieved by utilizing existing equipment operating at or below the nameplate rating and within the calculated BHP for the required pump motors. No increase in electrical equipment demand on the EDG's is expected as a result of EPU. Therefore, under emergency conditions, the electrical supply and distribution components are adequate. No increase in flow or pressure is required of any AC powered ECCS equipment. Therefore, the amount of power required to perform safety-related functions (pump and valve loads) is not increased with EPU, and the current emergency power system remains adequate. The systems have sufficient capacity to support all required loads to achieve and maintain safe shutdown conditions and to operate the ECCS equipment following postulated accidents and transients. Because the loads and mission times are not changed for EPU, no changes to the emergency diesel engine fuel oil storage and transfer system are necessary. Conclusion TVA has evaluated the required fuel oil for the emergency diesel generators and the effects of any increased electrical demand on fuel oil consumption. The evaluation indicates that the fuel 2-251

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) oil storage and transfer system will continue to provide an adequate amount of fuel oil to allow the diesel generators to meet the onsite power requirements of final GDC-17 and draft GDCs-4 and 40. Therefore, the proposed EPU is acceptable with respect to the fuel oil storage and transfer system. 2.5.6.2 Light Load Handling System (Related to Refueling) Regulatory Evaluation The light load handling system includes components and equipment used in handling new fuel at the receiving station and the loading of spent fuel into shipping casks. The NRCs acceptance criteria for the light load handling system are based on (1) GDC-61, insofar as it requires that systems that contain radioactivity be designed with appropriate confinement and with suitable shielding for radiation protection; and (2) GDC-62, insofar as it requires that criticality be prevented. Specific NRC review criteria are contained in SRP Section 9.1.4. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-66, 68, and 69. The light load handling system is described in Browns Ferry UFSAR Section 10.3, Spent Fuel Storage. 2-252

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the light load handling system is documented in NUREG-1843, Section 3.0.3.2.13. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.8 of the CLTR addresses the evaluation of the effect of the EPU on several plant systems that were not addressed elsewhere in that report. The Light Load Handling System (related to Fuel Handling and Storage System) is one of the systems so evaluated (see Table 2.5-5, Item 18). CLTR Section 6.8 is supported by ELTR1 (Reference 4), Section 5.12 and Appendix J, also previously approved by the NRC for use as guidelines for EPUs. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Fuel Handling and Storage System Generic Disposition The EPU has been found to not have any significant effect on the Fuel Handling and Storage System. The Fuel Handling and Storage System meets the CLTR disposition. Conclusion Implementing EPU does not require introducing any new fuel designs. Therefore, the fuel handling analysis is not affected by EPU. An evaluation of the light load handling system for the proposed EPU is not required. The proposed EPU is acceptable with respect to the light load handling system. 2.5.7 Additional Review Areas (Plant Systems) NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 6.8 of the CLTR addresses the evaluation of the effect of the EPU on several plant systems that were not addressed elsewhere in that report. The systems included in this evaluation are listed in Table 2.5-5. CLTR Section 6.8 is supported by ELTR1 (Reference 4), Section 5.12 and Appendix J, also previously approved by the NRC for use as guidelines for EPUs. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: 2-253

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Browns Ferry Topic CLTR Disposition Result Meets CLTR Other Systems Generic Disposition The EPU has been found to not have any significant effect on the systems listed in Table 2.5-5. The assessment of other systems meets the CLTR disposition. 2-254

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-1 NFPA 805 Fire Event Key Inputs Input Parameters Values Reactor Thermal Power 3,952 MWt RPV Dome Pressure 1,055 psia Decay Heat ANS 5.1-1979 without 2 uncertainty adder and with GEH SIL 636 recommendations Initial Suppression Pool Liquid Volume 122,940 ft3 (Note 1) Initial Suppression Pool and Wetwell Airspace 92°F (Note 2) Temperature Initial Wetwell Pressure 14.4 psia Initial Drywell Pressure 15.5 psia Initial Drywell Temperature 150°F Initial Wetwell Relative Humidity 100% Initial Drywell Relative Humidity 20% Drywell and Wetwell and Pool Heat Sinks Modeled Yes Drywell Heat Load Modeled Yes RHR Service Water Temperature 88°F (Note 2) RHR Heat Exchanger K Factor per Loop 290 Btu/sec-°F (Note 3) Number of RHR Loops Available 1 Number of RHR Pumps in One RHR Loop 1 ASDC RHR Flow Rate 7,500 gpm Condensate Available for Injection 90,000 gallons 2-255

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Notes:

1. Suppression pool volume corresponding to Browns Ferry Technical Specification low suppression pool water level with differential pressure control in service.

2. Nominal values based on Browns Ferry plant data over a seven year period from January 2008 through December 2014. Data analysis for this parameter shows that Browns Ferry operates at least 95% of time below this value.

3. RHR heat exchanger K factor based on RHR flow of 7,500 gpm, RHRSW flow of 4,500 gpm, RHRSW temperature of 88°F and conservative RHR heat exchanger fouling resistance. See LAR Attachment 39 for calculation of K factor.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-2 NFPA 805 Case 4 (EPU) Fire Event Evaluation Results Item Parameters Values 24.2 1 Peak DW Pressure (psia)

                                                    ~24,640 seconds 276.3 2     Peak DW Temperature (ºF)
                                                     ~1,500 seconds 24.6 3     Peak WW Airspace Pressure (psia)
                                                    ~24,640 seconds 209.0 4     Peak WW Airspace Temperature (ºF)
                                                    ~54,110 seconds 207.7 5     Peak Pool Temperature (ºF)
                                                    ~19,850 seconds 2-257

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-3 NFPA 805 Case 4 (EPU) Sequence of Events Approximate Elapsed Events Time Reactor scram occurs. Main Steam Isolation Valves (MSIVs) start to close. 0 seconds Feedwater pump is tripped. Drywell coolers are tripped. Condensate system continues to operate. MSIVs are fully closed. After isolation, MSRVs automatically 3.5 seconds start to open and close to maintain RPV pressure. Begin rapid depressurization using three MSRVs. RPV makeup 25 minutes is supplied by the condensate system. Condensate inventory available for injection is depleted.

  ~ 40 minutes         Operators secure condensate flow and initiate ASDC using 7,500 gpm of RHR flow in the LPCI mode.

2 hours RHR heat exchanger is placed into service. 72 hours Event is terminated. 2-258

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-4 SGTS Iodine Removal Capacity Parameters Browns Generic Ferry-Parameter Input Specific Criteria Value ((

))

Notes:

1. The value provided for this parameter is based on the configuration of the Browns Ferry SGTS.

Two trains of that system will be the minimum number that will operate in a DBA scenario. As such, the fuel iodine inventory value for a single Browns Ferry unit is split evenly between the two. Also the flowrate provided is for a single train.

2. Actual MSIV leakage is 100 scfh. It is not routed to the SGTS.

Results of the CLTR AST evaluation are applicable to Browns Ferry and show that the maximum charcoal loading, ((

                                                                          )) well below the 2.5 mg/gm maximum value in RG 1.52 (although Browns Ferry is not committed to that regulatory guide for iodine loading to the charcoal). The maximum component temperature is approximately 168ºF with normal flow conditions and 500ºF under conditions of a failed fan with minimum cooling flow, well below the 625ºF charcoal ignition temperature.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The parameters used in the CLTR analysis for AST application are confirmed to bound the Browns Ferry SGTS plant specific values. Therefore, the SGTS at Browns Ferry is confirmed to be consistent with the generic description provided in the CLTR. 2-260

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification (( 2-261

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-262

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-263

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-264

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-265

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-266

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-267

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-268

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification 2-269

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Table 2.5-5 Basis for Classification of No Significant Effect (continued) Item System Type Browns Ferry System Name Functional Description Basis for (Number) Classification

))

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Figure 2.5-1 NFPA 805 Case 4 (EPU) Fire Event Suppression Pool Temperature 2-271

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Figure 2.5-2a Browns Ferry Unit 1 Generator Reactive Capability Curve 2-272

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Figure 2.5-2b Browns Ferry Units 2 and 3 Generator Reactive Capability Curve 2-273

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.6 Containment Review Considerations 2.6.1 Primary Containment Functional Design Regulatory Evaluation The containment encloses the reactor system and is the final barrier against the release of significant amounts of radioactive fission products in the event of an accident. The NRCs acceptance criteria for the primary containment functional design are based on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and that such SSCs be protected against dynamic effects; (2) GDC-16, insofar as it requires that reactor containment be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment; (3) GDC-50, insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated temperature and pressure conditions resulting from any LOCA; (4) GDC-13, insofar as it requires that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation and for accident conditions, as appropriate, to assure adequate safety; and (5) GDC-64, insofar as it requires that means be provided to monitor the reactor containment atmosphere for radioactivity that may be released from normal operations and from postulated accidents. Specific NRC review criteria are contained in SRP Section 6.2.1.1.C. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. 2-274

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-10, 12, 17, 40, 42, and 49. The primary containment is described in Browns Ferry UFSAR Sections 5.2, Primary Containment System and 7.3, Primary Containment Isolation System. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the primary containment is documented in NUREG-1843, Sections 2.3.2.1.1 and 2.4.1.1. Management of aging effects on the primary containment is documented in NUREG-1843, Sections 3.2.2 and 3.5.2. Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 4.1 of the CLTR addresses the effect of EPU on Primary Containment Functional Design. The results of this evaluation are described below. The Browns Ferry UFSAR provides the containment responses to various postulated accidents that validate the design basis for the containment. EPU operation changes some of the conditions for the containment analyses. For example, the short-term DBA-LOCA containment response during the blowdown is governed by the blowdown flow rate. This blowdown flow rate is dependent on the reactor initial thermal-hydraulic conditions, such as vessel pressure and the mass and energy of the vessel fluid inventory, which change slightly with EPU. Also, the long-term heat-up of the suppression pool following a LOCA or a transient is governed by the ability of the RHR to remove decay heat. Because the decay heat depends on the initial reactor power level, the long-term containment response is affected by EPU. The containment response was reanalyzed to demonstrate the plant's capability to operate with a rated power increase to 3,952 MWt. The key plant parameters used to model and analyze the plant response at EPU are provided in Table 2.6-2a. The analyses of containment pressure and temperature responses, as described in Section 2.6.1.1, were performed at a power level of 102% of EPU RTP in accordance with ELTR1 using GEH codes and models. The M3CPT code was used to model the short-term containment pressure and temperature response. The modeling used in the M3CPT analyses is described in References 45 and 76. References 45 and 76 describe the basic containment analytical models used in GEH codes. Reference 6 describes the more detailed RPV model (LAMB) used for determining the vessel break flow in the containment analyses for EPU. 2-275

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The LAMB code models the recirculation loop as a separate pressure node. It also allows for inclusion of flashing in the pipe and vessel during the blowdown and flow choking at the jet pump nozzles when the conditions warrant. The use of the LAMB blowdown flow in M3CPT was identified in ELTR1 by reference to the LAMB code qualification in Reference 6. The SHEX code was used to model the long-term containment pressure and temperature response. The key models in SHEX are based on models described in References 45, 76 and 77. The GEH containment analysis methodologies have been applied to all BWR power uprate projects performed by GEH and accepted by the NRC. The Browns Ferry original long-term containment analyses did not credit passive heat sinks in the drywell, wetwell airspace, and suppression pool. This conservative assumption was identified to the NRC as Assumption 6 of Attachment 1 to the March 12, 1993 GE letter referenced in Reference 7. Long-term containment analyses performed for Browns Ferry EPU now includes credit for these passive heat sinks. This is herein identified as a change in methodology. ((

            )) (Assumption 8 of the same GE letter).

The effects of EPU on the containment dynamic loads due to a LOCA or MSRV discharge have also been evaluated as described in Section 2.6.1.2. The containment hydrodynamic loads have been defined generically for Mark I plants as part of the Mark I Containment Long-Term Program (LTP) (Reference 78) and approved by the NRC in Reference 79. The Browns Ferry plant-specific dynamic loads were defined in References 46 and 80, using the NRC approved methods of Reference 78. The evaluation of the LOCA containment dynamic loads is based primarily on the results of the short-term analysis described in Section 2.6.1.2. The MSRV discharge load evaluation is based on no changes in the MSRV opening setpoints for EPU. The metal-water reaction energy versus time relationship is calculated using the method described in USNRC Regulatory Guide 1.7 (Reference 81) as a normalized value (fraction of reactor thermal power). All of the energy from the metal-water reaction is assumed transferred to the reactor coolant in the first 120 seconds into the LOCA. The metal-water reaction energy represents a very small fraction of the total shutdown energy transferred to the coolant. Browns Ferry uses the Mark I containment design. Per the discussion in Section 4.1 of the NRC Safety Evaluation (SE) for the CLTR (Reference 1), benchmarking cases, originally stipulated in Reference 4 and Reference 7, using SHEX are not required for Mark I and Mark III containment analyses. To quote: The NRC has performed independent confirmatory analyses on extended uprates for both Mark I and Mark III containment designs and found the results consistent with SHEX results. Therefore, the confirmatory calculations with SHEX (benchmarking with current licensing basis assumptions - pre-uprate) for plant specific modeling are not required for extended power uprates for Mark I and Mark III containment designs. Therefore, following the 2-276

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) NRC safety evaluation of the CLTR (Reference 1), confirmatory benchmarking cases of SHEX are not required and were not performed for Browns Ferry EPU. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Topic CLTR Disposition Browns Ferry Result Meets CLTR Pool Temperature Response Plant Specific Disposition Meets CLTR Wetwell Pressure Plant Specific Disposition Meets CLTR Drywell Temperature Plant Specific Disposition Meets CLTR Drywell Pressure Plant Specific Disposition Meets CLTR Containment Dynamic Loads Plant Specific Disposition Meets CLTR Containment Isolation Plant Specific Disposition Meets CLTR Motor-Operated Valves Plant Specific Disposition Meets CLTR Hardened Wetwell Vent System Plant Specific Disposition Meets CLTR Equipment Operability Plant Specific Disposition 2.6.1.1 Containment Pressure and Temperature Response The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. As a result of this, the suppression pool temperature response, wetwell pressure, drywell temperature, and drywell pressure need to be addressed. Short-term and long-term containment analysis results are reported in the UFSAR. The short-term analysis is directed primarily at determining the drywell pressure response during the initial blowdown of the reactor vessel inventory to the containment following a large break inside the drywell. Short-term containment response analyses were performed for the limiting DBA-LOCA that assumes a double-ended guillotine break of a recirculation suction line (RSLB) to demonstrate that EPU does not result in exceeding the containment design limits. 2-277

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The long-term analysis is directed primarily at the suppression pool temperature response, considering the decay heat addition to the suppression pool. The RSLB DBA-LOCA, the double-ended guillotine break of a recirculation discharge line (RDLB) and small steam break LOCAs were reanalyzed for EPU. Peak values of the containment pressure and temperature responses to these LOCA events are given in Table 2.6-1. The effect of local suppression pool temperatures during MSRV discharges was addressed in accordance with the NUREG-0783 (Reference 82) criteria. Peak suppression pool temperatures resulting from the postulated ATWS, Station Blackout, and Fire events are given in Table 2.6-3. The effect of EPU on the events which yield the limiting containment pressure and temperature response is provided below. 2.6.1.1.1 Long-Term Suppression Pool Temperature Response 2.6.1.1.1.1 Bulk Pool Temperature The long-term bulk pool temperature response for EPU was evaluated for the limiting DBA-LOCA in Section 14.6.3.3 (Case C) of the UFSAR. This DBA-LOCA is an instantaneous guillotine break of the RSLB. For Browns Ferry EPU, RDLB LOCA and small break LOCAs were also analyzed at EPU conditions Per GE Safety Communication SC 06-01 (Reference 83), the potential was identified that a single failure that eliminated only the RHR heat exchanger could prove more limiting than the typically analyzed scenario of the single failure of an entire AC electrical power source. The Browns Ferry RHR system is configured with two loops of RHR, with each loop having its own separate injection point to the reactor pressure vessel, and with each loop having its own separate return to the suppression pool. Each loop is comprised of two RHR pumps with each pump having its own separate heat exchanger on its discharge. The current licensing basis analysis (Reference 84) for the short-term (first 10 minutes after the accident) evaluation of the RSLB assumed a Single Active Failure (SAF) where only two of the four RHR pumps were available. In order to address the issue identified in SC 06-01 (Reference 83), the RSLB EPU analysis assumes that all four RHR pumps are running in the short-term phase of the RSLB DBA-LOCA. This assumption will maximize the ECCS pump heat addition to the suppression pool and thereby maximize the suppression pool temperature. The RDLB analysis for CLTP conservatively assumed that all four RHR pumps are running in the short-term phase of the RSLB DBA-LOCA. The EPU analysis also conservatively assumes all four RHR pumps are running in the short-term phase of the RSLB DBA-LOCA. Therefore, the issue identified in SC 06-01(Reference 83) is addressed in the EPU analysis The acceptability of ECCS pump NPSH based on the containment analysis suppression pool temperature response is demonstrated in Section 2.6.5.2. RHR and core spray pumps can be throttled to decrease required NPSH from the required NPSH at pump run-out flow conditions, provided containment cooling requirements are satisfied. The analysis of the RSLB DBA-LOCA was performed at 102% of EPU RTP. The time-dependent SP temperature response is presented in Figure 2.6-1 and the calculated peak values 2-278

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) for LOCA bulk pool temperature for the CLTP and the EPU RTP case are compared in Table 2.6-1. The EPU analyses were performed using a decay heat table based on ANS/ANSI 5.1-1979 with 2-sigma adders with additional actinides and activation products per GE SIL 636 (Reference 85). No modifications were made to this standard. The containment system response to the accident is divided into two analysis phases. The first phase, hereafter referred to as the short-term phase covers the period up to 10 minutes after the accident initiation. During the short-term phase, no operator action is credited in the analysis. The second phase, hereafter referred to as the long-term phase covers the period after 10 minutes following the accident initiation. During the long-term phase, operator actions such as those to reduce electrical loading on the emergency diesel generators and to re-align portions of the ECCS from core cooling mode to containment cooling mode are credited. The RSLB DBA-LOCA analysis assumes that offsite power is lost concurrently with the accident initiation and that offsite power is not available during the accident analysis period. Separate RSLB analysis cases are run with initial conditions to either maximize or minimize the containment drywell and wetwell pressure response while maximizing the suppression pool temperature response in order to determine the sensitivity of the peak suppression pool temperature response to perturbed initial conditions. No containment leakage is assumed except for the RSLB cases with initial conditions to minimize the containment drywell and wetwell pressure response while maximizing the suppression pool temperature response, for which containment leakage (2% per day) and the leakage from MSIVs (150 scfh for all steam lines) are considered. In addition, the containment responses to various modes of containment cooling are evaluated. These three RHR cooling modes are (1) Coolant Injection Cooling (CIC), where RHR flow is cooled by the RHR heat exchanger before being discharged into the reactor vessel; (2) Containment Spray Cooling (CSC), where RHR flow is cooled by the RHR heat exchanger and then discharged to the containment via the DW spray and wetwell spray headers; and (3) Suppression Pool Cooling (SPC), where RHR flow is cooled by the RHR heat exchanger and then discharged back to the suppression pool. A complete LOOP is assumed to occur concurrent with the accident initiation. If a worst-case SAF such as failure of one emergency electrical power source (emergency diesel generator or loss of a 4 kV shutdown board) is assumed concurrent with the accident, then less than the full complement of low pressure ECCS pumps (four RHR pumps and four CS pumps) would be available during the short-term phase of the accident. However, if no SAF is assumed, then the full complement of ECCS pumps would be available. The initial condition of no SAF during the short-term phase is limiting for the determination of ECCS pump NPSH during the accident because of the Browns Ferry ECCS pump suction configuration where each ECCS pump does not have a dedicated ECCS suction strainer and piping suction directly from the suppression pool (torus). For each Browns Ferry unit, there are four ECCS suction strainers installed in the torus. The torus water volume then communicates to the ECCS pump suctions via a torus ring header located below the torus. This configuration result in higher ECCS piping head loss when there are multiple ECCS pumps running. In addition, a larger number of running ECCS pumps will lead to higher pump heat addition to the suppression pool. Conformance with GEH SC 06-01 2-279

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) (Reference 83) is made by assuming all low pressure ECCS pumps start during the short-term phase of the accident. All ECCS pumps are assumed to be available for the first 600 seconds after accident initiation. No RHRSW flow is assumed to the RHR heat exchangers and there is no heat removal from the RHR heat exchangers during the short-term phase. RPV liquid is discharged from the break into the drywell causing rapid vessel depressurization and a rapid increase in the drywell pressure and temperature. For the first 600 seconds following the accident, four RHR pumps in LPCI mode (with two RHR pumps injecting liquid into the intact recirculation loop and the other two RHR pumps into the broken recirculation loop) and four CS pumps are used to cool the core. For the RSLB DBA-LOCA, the RHR flow into the broken recirculation loop will be directed to the RPV and RHR flow will not go into runout flow because the RHR injection point is between the RPV and the closed reactor recirculation discharge valve (the reactor recirculation discharge valve in each reactor recirculation loop receives an automatic closure signal during a LOCA). HPCI is assumed available and will start on either high DW pressure or low RPV level. However, HPCI will isolate on low steam pressure. The ECCS injection of suppression pool water, along with the assumed addition of feedwater, produces a recovery of the reactor water level. This allows water heated by decay heat and vessel sensible energy to be discharged into the drywell, and subsequently into the suppression pool. If the accident were to occur on either Unit 1 or 2 and a worst-case SAF such as failure of one emergency electrical power source (emergency diesel generator or loss of a 4 kV shutdown board) is assumed concurrent with the accident, then less than the full complement of low pressure ECCS pumps (four RHR pumps and four CS pumps) would be available during the long-term phase of the accident. Assuming that one RHR pump is required for shutdown of the non-accident unit, only two RHR pumps and two RHR heat exchangers are assumed available for long-term containment cooling in the accident unit. After 600 seconds, operator actions are credited. One loop of CS with two CS pumps continues to be available for RPV water makeup. One loop of CS with two pumps is secured because two CS pumps can supply adequate long-term core cooling. One loop of RHR with two pumps is secured, and another loop of RHR with two pumps is switched to a RHR mode of containment cooling with its associated RHRSW flow activated for two heat exchangers. Three RHR cooling modes are investigated: (1) Coolant Injection Cooling (CIC) where RHR in LPCI mode with flow from the suppression pool is cooled by the RHR heat exchanger before being discharged into the reactor vessel; (2) CSC where RHR flow from the suppression pool is discharged as drywell and wetwell sprays; and (3) SPC where the RHR flow from the suppression pool is cooled by the RHR heat exchanger before being discharged back into the suppression pool. The heat exchanger K-value and RHR pump flow rate are presented in Table 2.6-2a. Initial conditions (initial DW pressure, initial wetwell pressure and initial DW temperature) were also perturbed to both maximize and minimize the peak containment pressure and thereby investigate the effect on peak suppression pool temperature. The resulting calculated peak bulk SP 2-280

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) temperature for RSLB DBA-LOCA at 10 minutes after the accident initiation is 152.8°F and the peak bulk SP temperature for RSLB DBA-LOCA is 179.0°F. The containment response during the first 10 minutes following the accident initiation for a RDLB-LOCA was calculated using Browns Ferry specific inputs to maximize suppression pool temperature and minimize containment pressure, similar to the RSLB DBA-LOCA analysis. The key parameter differences between the RDLB and the RSLB during the short-term phase of the accident are: (1) the break area (4.2 ft2 for the RSLB versus 1.94 ft2 for the RDLB) and, (2) the RHR flow rate and RHR injection path into the broken recirculation loop. For the RDLB, the RHR flow into the broken recirculation loop discharges directly to the drywell and the RHR flow into the broken loop is assumed at runout conditions (11,000 gpm per RHR pump for the RDLB versus 9,000 gpm per RHR pump for the RSLB). The resulting calculated peak bulk SP temperature for the RDLB at 10 minutes after the accident initiation is 152.0°F. The suppression pool temperature and corresponding wetwell pressure for the RDLB analyses are used in the evaluation of the available NPSH for the CS and the RHR pumps. The results of that evaluation are provided in Section 2.6.5.2. The suppression pool temperature response was also calculated for the spectrum of small steam line break LOCAs as evaluated for the drywell temperature response. The most limiting bulk suppression pool temperature response to the small steam line break LOCA was found to occur for the smallest break size evaluated, a 0.01 ft2 break, which produced a peak bulk suppression pool temperature of 182.7°F (See Section 2.6.5.1). The suppression pool temperature response was also calculated for the shutdown of a non-accident unit. The time-dependent SP temperature response is presented in Figure 2.6-1a. The peak bulk suppression pool temperature for this case is 185.1°F (See Section 2.6.5.1). Based on the analysis and limit values shown in Table 2.6-1, the peak bulk pool temperature for the LOCA events at EPU RTP is acceptable from a structural design standpoint. With calculated peak bulk suppression pool temperatures below the design limit, small break LOCAs and non-LOCA events with EPU are also acceptable from a structural design standpoint 2.6.1.1.1.2 Local Pool Temperature with MSRV Discharge The local pool temperature limit for MSRV discharge was specified in NUREG-0783 (Reference 82) because of concerns resulting from unstable condensation observed at high pool temperatures in plants without quenchers. Quencher devices such as the T-quenchers used in the Browns Ferry units mitigate these loads. The peak local suppression pool temperature at Browns Ferry has been evaluated for EPU, with the same scenario assumptions as evaluated in the original analysis of Reference 86, and meets the NUREG-0783 criteria. This evaluation demonstrated a minimum subcooling of approximately 20°F locally at the quencher. This meets the acceptance criteria included in NUREG-0783 and also ensures that the exiting quencher steam is condensed before posing a steam ingestion potential for any ECCS pump suction. Therefore, the peak local suppression pool temperature at Browns Ferry remains acceptable at EPU conditions. 2-281

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Containment pressure response and temperature response as a result of ((

                                  )) were evaluated and found to be acceptable. Therefore, the containment pressure and temperature response meets all CLTR dispositions.

2.6.1.1.2 Steam Bypass Capability Containment response is based on maintaining the pressure suppression function by limiting the leakage from the drywell to the suppression pool due to leakage between the drywell and wetwell airspace. In the event that excessive bypass leakage was to occur, over-pressurization of the primary containment could occur. The acceptance criterion for Mark I plants such as Browns Ferry, with regard to steam bypass leakage, is that the measured leakage is not greater than the leakage that would result from a one inch diameter opening. This maximum bypass leakage is confirmed by plant tests as directed in Browns Ferry Technical Specification Surveillance Requirement (SR) 3.6.1.1.2. The current steam bypass effective area capability, A/K , which was established from Browns Ferry analysis, is 0.18 ft2. This 0.18 ft2 effective area is approximately 54 times greater than the effective area of a one-inch opening (a one-inch plate orifice has an /K of ~0.0033 ft2). The steam bypass analysis was performed at an initial power level of 102% of EPU RTP. At EPU conditions, the steam bypass analyses were performed by assuming a spectrum of steam line breaks and by crediting containment heat sinks. Mechanistic energy and mass transfer between the suppression pool and airspace more realistically model the physical phenomenon for steam bypass conditions. In the current licensing basis analysis, operator action to initiate containment spray and thereby mitigate containment over pressurization is assumed to occur 10 minutes after the wetwell pressure reaches 35 psig. In order to address the possible interruption of containment cooling due to receipt of a LOCA signal caused by high drywell pressure concurrent with low RPV pressure, the EPU analysis conservatively assumed, for all break sizes except the smallest analyzed break size of 0.01 ft2, a 20 minute delay for the initiation of containment sprays after the wetwell pressure reaches 35 psig. For the smallest break size of 0.01 ft2, the RPV depressurization rate is sufficiently small that operators can inhibit the containment cooling interruption caused by high drywell pressure concurrent with low RPV pressure. The EPU analysis for the 0.01 ft2 break assumed the operators initiate containment spray 10 minutes after the wetwell pressure reaches 35 psig. The EPU evaluation shows that the peak containment pressure remains below the containment design pressure with no change in the current steam bypass effective area capability (0.18 ft2) with an initial DW temperature of 130ºF. EPU requires no change to the existing Browns Ferry TS SR 3.6.1.1.2 which is to detect flow paths between the drywell and wetwell whose total capacity is equal to or greater than the capacity of a one-inch diameter plate orifice (a one-inch plate orifice has an /K of ~0.0033 ft2). 2.6.1.2 Containment Dynamic Loads The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. As a result, containment dynamic loads are addressed in the following sections. 2-282

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.6.1.2.1 Loss-of-Coolant Accident Loads The LOCA containment dynamic loads analysis for EPU is primarily based on the short-term RSLB LOCA analyses and compliance with generic criteria developed through testing programs. The analyses were performed as described in Section 2.6.1.1 with break flows calculated using a more detailed RPV model (Reference 6). The NRC approved use of this model for the EPU containment evaluations in Reference 4. These analyses also provide calculated values for the controlling parameters for the dynamic loads throughout the blowdown. The key parameters are drywell and wetwell pressures, vent flow rates and suppression pool temperature. The LOCA dynamic loads considered in the EPU evaluations include pool swell, CO and chugging. For Mark I plants like Browns Ferry, the vent thrust loads were also evaluated. The results of the EPU pool swell evaluation confirmed that the current pool swell load definition remains bounding. The containment response conditions for EPU are within the range of test conditions used to define CO loads for the plant. The containment response conditions for EPU are within the conditions used to define the chugging loads. The vent thrust loads at EPU conditions were calculated to be less than the plant-specific values calculated during the Mark I containment LTP. The Mark I containment program Load Definition Report (LDR) Table 4.5.1-1 (Reference 78) defines the onset and duration times for chugging based on break size. For intermediate break sizes, chugging ends at 900 seconds after onset of chugging, 905 seconds after the break; for small break sizes, chugging ends at 900 seconds after onset of chugging, 1,200 seconds after the break. Discussion of the chugging duration time is provided in Sections 2.2, 2.3, and 4.4.1.1 of the LDR (Reference 78). For the load definition, chugging is assumed to end when reactor pressure is reduced to or below the drywell pressure, essentially stopping break flow and therefore vent steam flow. This vessel depressurization for the IBA and SBA events is due to initiation of the Automatic Depressurization System (ADS). The load definition of the LDR does not include any credit for operation of containment (drywell) sprays. However, Emergency Operating Procedures (EOPs) for Browns Ferry include direction to initiate drywell sprays when wetwell pressure exceeds 12.0 psig. Containment analyses performed for Browns Ferry EPU have shown that wetwell pressure will exceed this drywell spray initiation pressure of 12.0 psig by 600 seconds following initiation of the event if conditions for chugging are present. Initiation of drywell sprays will rapidly reduce drywell pressure and stop chugging. However, as reported in GEH SC 11-10 (Reference 87), for plants like Browns Ferry, where a LOCA signal is initiated on concurrent high DW pressure plus low RPV pressure, DW spray initiation could be delayed up to 1,200 seconds after initiation of an IBA or SBA LOCA. Therefore the chugging duration could be extended to a maximum of 1,200 seconds, which exceeds the duration times identified in Reference 78 for Mark I plants and Reference 46 for Browns Ferry. The effect of the chugging duration extension to 1,200 seconds was evaluated for Browns Ferry. From the Browns Ferry plant unique analysis report (PUAR (Reference 46)), the limiting fatigue usage factors for containment components are listed below: 2-283

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Component Allowable Fatigue DBA Usage SBA/IBA Usage Usage Factor Factor Drywell Structure General Shell and Cradle 1.0 0.051 0.096 Penetration X-204C 1.0 0.020 0.103 Containment Vent System Downcomer/Vent Header Intersection 1.0 0.559 0.610 Downcomer/Tie Bar Intersection 1.0 0.107 0.353 Torus Bellow Intersection 1.0 0.000 0.000 The highest fatigue usage factor from the above table for the SBA/IBA is 0.610. This 0.610 fatigue usage factor is the sum of the fatigue usage factor due to MSRV actuation (0.392) and the fatigue usage factor due to chugging (0.218). The number of chugging cycles is factored to consider a 900 second duration. To account for the 1,200 second chugging duration, the fatigue usage factor due to chugging can be linearly extrapolated: 0.218 * (1,200/900) = 0.291. Adding this chugging fatigue usage factor to the MSRV fatigue usage factor of 0.392 provides a revised fatigue usage factor of 0.683, which remains below the allowable fatigue usage factor of 1.0. All other components remain well below the allowable fatigue usage factor even if all fatigue usage is conservatively attributed only to chugging and the PUAR value identified in the above table is factored by 1,200/900. 2.6.1.2.2 Safety Relief Valve Loads The MSRV loads include MSRVDL loads, suppression pool boundary pressure loads, and drag loads on submerged structures. The MSRV opening setpoint pressure, the initial water leg in the MSRVDL, the MSRVDL geometry, and the suppression pool geometry influence these loads. The MSRV loads were evaluated for two different actuation phases: initial actuation and subsequent actuation. For the initial MSRV actuation following an event involving RPV pressurization, the only parameter change potentially introduced by EPU, which can affect the MSRV loads definition, is an increase in MSRV opening setpoint pressure. However, the changes proposed for EPU do not include an increase in the MSRV opening setpoint pressure. The load definition for subsequent MSRV actuations is not affected by EPU because the MSRVDL reflood height used for Browns Ferry is the maximum reflood height (Reference 46) that is not affected by the time between MSRV closing and MSRV reopening. The maximum reflood height is controlled by the MSRVDL geometry and the MSRVDL vacuum breaker capacity. Because all these parameters, including the MSRV setpoints, do not change, loads due to subsequent MSRV actuations are not affected by EPU. Therefore, EPU does not affect the MSRV loads or load definitions. 2-284

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) 2.6.1.2.3 LOCA Pressure and Temperature Loads The Reference 80 plant unique load definition report (PULD) provided LOCA-induced pressure and temperature results from DBA-LOCA (DBA), IBA, and SBA events as an input for subsequent use in the Reference 46 structural analysis. The IBA and SBA events were re-evaluated at 102% EPU RTP using initial conditions and assumptions consistent with the Reference 80 analysis. The results of the Browns Ferry EPU analysis show that all DW and WW pressure and temperatures at EPU conditions are bounded by the values of Reference 80 with the exception of the peak WW and SP temperature for the SBA. At EPU conditions, the SBA peak WW and SP temperature is 146°F, which does not bound the Reference 80 result of 136°F. ((

))

The evaluation of WW and SP piping that have the SBA temperature as a structural load combination input is contained in Section 2.2.2.2.2.2 (Other Piping Evaluation). Given the fact that current containment dynamic load evaluations remain bounding and applicable for plant operation at EPU conditions, and that the current MSRV load definition is still applicable, all CLTR dispositions are met. 2.6.1.3 Containment Isolation The CLTR states that the suppression pool temperature increases as a result of the higher decay heat associated with EPU. However, the system designs for containment isolation are not affected by EPU. The capabilities of isolation actuation devices to perform during normal operations and under post-accident conditions have been determined to be acceptable. Therefore, the Browns Ferry containment isolation capabilities are not adversely affected by the EPU and all CLTR dispositions are met. 2.6.1.4 Generic Letter 89-16 Hardened Wetwell Vent In response to GL 89-16, Browns Ferry installed a Hardened Wetwell Vent (HWWV) to mitigate the pressure increase during a TW severe accident sequence. The vent capacity is currently sized to prevent the containment pressure from exceeding the primary containment pressure limit with constant heat input equal to 1% of 3,458 MWt or 34.58 MWt. The Browns Ferry HWWV design for Unit 1, which is functionally the same for Units 2 and 3 with respect to the 1% vent capacity, was approved by Amendment No. 269 to Renewed License No. DPR-33 (Reference 89). At 2-285

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) EPU conditions, the existing vent capacity will be reduced to 0.88% of rated thermal power. This capacity will be restored to 1% of rated EPU power as discussed below. In response to EA-13-109, Issuance of Order to Modify Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions (Reference 90), Browns Ferry will modify the hardened containment vent system (HCVS) as described in the Browns Ferry letter to the NRC dated August 28, 2014 (Reference 91) and approved by NRC letter dated December 23, 2014 (Reference 92). The consideration of EPU conditions and the 1% vent capacity requirement is consistent with NEI 13-02, Industry Guidance for Compliance with Order EA-13-109, Revision 0 (Reference 93). NRC endorsed NEI 13-02 by JLD-ISG-2013-02, Revision 0, Compliance with Order EA-13-109, Order Modifying Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident Conditions, Interim Staff Guidance, November 14, 2013 (Reference 94). The Browns Ferry letter details the modification schedule for the HCVS for full compliance with Phase 1 of EA-13-109. The HCVS modifications for each Browns Ferry unit will be completed prior to implementing EPU at that unit. 2.6.1.5 Generic Letter 96-06 GL 96-06 identified potential problems with equipment operability and containment integrity during design-basis accident conditions as a result of: (1) water hammer and/or two-phase flow conditions in cooling water systems serving the containment air coolers; and (2) thermally induced over-pressurization of isolated piping sections in containment. The calculation that supports the Browns Ferry responses to GL 96-06 states that in the event of a DBA-LOCA or steam line break event with a coincident LOOP, the RBCCW pumps will be load shed at the start of the event. The first RBCCW pump will then be given a start signal 40 seconds after it was shed. Should the first pump fail to start, the second pump will be given a start signal three seconds later. Based on computed results, voiding in the RBCCW drywell atmosphere cooling coils will not occur for at least:

61 seconds on Units 2 and 3 for the DBA-LOCA
62 seconds on Units 2 and 3 for the steam line break
46 seconds on Unit 1 for the steam line break
49 seconds on Unit 1 for the DBA-LOCA As an RBCCW pump will have been restarted prior to voiding occurring, water hammer and/or two-phase flow in the RBCCW system are not a concern. The resultant time to boil is less in the Unit 1 analyses primarily due to two factors: (1) the drywell coolers were replaced in Unit 1 and the tubes in the Unit 1 replacement drywell coolers have a smaller inner diameter than the tubes used in the drywell coolers installed in Units 2 and 3, and 2) the modeling of the Unit 1 drywell cooler response was revised to more accurately represent the heat transfer by the drywell cooler tube fins.

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NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) A comparison of the drywell temperature profiles, during a DBA-LOCA for current licensing basis and EPU conditions, shows a minor difference in temperature (less than 2°F) and lasting only seconds. Based on the minor differences in the drywell temperature profiles and the margin between the RBCCW pump start (43 seconds) and calculated time to boil (61 seconds for Units 2 and 3 and 49 seconds for Unit 1), it is concluded that voiding will not occur under EPU conditions prior to the restart of an RBCCW pump. The maximum drywell temperature for the steam line break event at EPU conditions is bounded by the maximum drywell temperature used in the drywell cooler voiding analysis for the steam line break event at current licensing basis conditions. The drywell cooler voiding analysis for steam line breaks conservatively assumes that the drywell atmosphere instantaneously increases at the start of the event from the initial drywell temperature (150°F) to the peak drywell temperature (336°F). The drywell temperature is then assumed to remain at the peak temperature (336°F) for the duration of the event. Observation of Figure 2.6-9 shows that the drywell temperature response for steam line breaks during the first 100 seconds does not exceed 330°F. Therefore, the conclusion that an RBCCW pump will have been restarted prior to voiding occurring and that water hammer and two-phase flow in the RBCCW system are not a concern, is still valid. EPU is not altering the RBCCW system serving the drywell atmosphere cooling coils. The Browns Ferry response to GL 96-06 included four primary containment penetrations and process lines that were identified as being susceptible to thermal pressurization:

1) Demineralized water system,

2) Drywell floor drain sump discharge,

3) Drywell equipment drain sump discharge, and

4) Reactor water sampling system.

The demineralized water system is acceptable at EPU as controls are in place to ensure the header is drained prior to power operation each cycle. The current analysis of record calculation, using an assumed constant DW temperature of 336°F following a LOCA showed that 2.06 gallons of water would have to be drained from the system prior to power operation in order to prevent system over-pressurization following a LOCA. A slightly higher assumed EPU DW temperature of 336.9°F will increase the drain requirement to 2.07 gallons, which is negligible. The DW floor and DW equipment drain sump discharge lines are acceptable as a 0.06-inch (1/16-inch diameter) orifice has been drilled in each discharge check valve. This orifice ensures adequate leakage to prevent over-pressurization due to thermal expansion. The current analysis of record calculation, using an assumed constant DW temperature of 336°F following a LOCA, showed that an orifice diameter of 0.052-inches was sufficient to relieve the flow associated with thermal expansion in the discharge lines. A slightly higher assumed EPU DW temperature of 336.9°F will negligibly increase the flow requirements, and the existing orifices are adequately sized to pass this flow at EPU. 2-287

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The thermally induced over-pressurization of the reactor sampling system is calculated at CLTP (using a constant DW temperature of 336°F following a LOCA) to reach approximately 2,546 psig which will lift the inboard globe isolation valve disc off its seat and relieve the pressure back to the reactor vessel. The pressure associated with the seat lift is well within the design pressures of the reactor sampling equipment in the drywell subject to the thermal over-pressurization. The increase in drywell temperature to 336.9°F at EPU will result in a negligible reduction in margin for the prevention of over-pressurization in the reactor water sampling system. The pressure associated with the seat lift at EPU remains well within the design pressures of the reactor sampling equipment in the drywell subject to the thermal over-pressurization. The list of penetrations susceptible to thermal over-pressurization during design-basis accident conditions does not change at EPU conditions. A review of the EPU process conditions was performed for other systems (e.g., main steam, RWCU, RHR, and SLC) that could be potentially susceptible to thermal over-pressurization. The review concluded that EPU does not result in any changes to either the physical configuration or process conditions of the systems that would change the current Browns Ferry disposition of these systems as acceptable for thermal over-pressurization. EPU is not adding any new containment penetrations or performing any physical/procedural changes to the penetration configuration, or the process lines that pass through them; therefore, the Browns Ferry analysis of thermally induced over-pressurization of isolated piping sections in containment remains valid. Therefore, the existing Browns Ferry response to GL 96-06 remains valid for EPU and all CLTR dispositions are met. Conclusion TVA has evaluated the containment temperature and pressure transient and accounted for the increase of mass and energy resulting from the proposed EPU. The evaluation indicates that containment systems will continue to provide sufficient pressure and temperature mitigation capability to ensure that containment integrity is maintained. The evaluation further indicates that containment systems and instrumentation will continue to be adequate for monitoring containment parameters and release of radioactivity during normal and accident conditions and the containment and associated systems will continue to meet the requirements of draft GDCs-10, 12, 17, 40, 42, and 49 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to primary containment functional design. 2.6.2 Subcompartment Analyses Regulatory Evaluation A subcompartment is defined as any fully or partially enclosed volume within the primary containment that houses high-energy piping and would limit the flow of fluid to the main containment volume in the event of a postulated pipe rupture within the volume. 2-288

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) The NRCs acceptance criteria for subcompartment analyses are based on (1) GDC-4, insofar as it requires that SSCs important to safety be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, and that such SSCs be protected against dynamic effects; and (2) GDC-50, insofar as it requires that containment subcompartments be designed with sufficient margin to prevent fracture of the structure due to the calculated pressure differential conditions across the walls of the subcompartments. Specific NRC review criteria are contained in SRP Section 6.2.1.2. Browns Ferry Current Licensing Basis The General Design Criteria (GDC) listed in RS-001 are those currently specified in 10 CFR 50, Appendix A. The applicable Browns Ferry Nuclear Plant (Browns Ferry) principal design criteria predate these criteria. The Browns Ferry principal design criteria are listed in UFSAR Section 1.5, Principal Design Criteria. In 1967, the AEC published for public comment a revised set of proposed General Design Criteria (Federal Register 32FR10213, July 11, 1967). Although not explicitly licensed to the AEC proposed General Design Criteria published in 1967, the Tennessee Valley Authority (TVA) performed a comparative evaluation of the design basis of Browns Ferry with the AEC proposed General Design Criteria of 1967. The Browns Ferry UFSAR, Appendix A, Conformance to AEC Proposed General Design Criteria, contains this comparative evaluation. This evaluation discusses each of the groups of criteria set out in the July 1967 AEC release. For each group of criteria, there is a statement of TVAs understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria. Following a restatement of each of the proposed criteria is a table of references to locations in the Browns Ferry UFSAR where there is subject matter relating to the intent of that particular criteria. While Browns Ferry is not generally licensed to the final GDC or the 1967 AEC proposed General Design Criteria, a comparison of the final GDC to the applicable AEC proposed General Design Criteria can usually be made. For the final GDC listed in the Regulatory Evaluation above, the Browns Ferry comparative evaluation of the comparable 1967 AEC proposed General Design Criteria (referred to here as draft GDC) is contained in Browns Ferry UFSAR Appendix A: draft GDCs-40, 42, and 49. The primary containment is described in Browns Ferry UFSAR Sections 5.2, Primary Containment System and 12.2, Principal Structures and Foundations. Systems and system component materials of construction, operating history, and programs used to manage aging effects were evaluated for plant license renewal and documented in the Browns Ferry License Renewal Safety Evaluation Report (SER), NUREG-1843, dated April 2006 (Reference 11). The license renewal evaluation associated with the primary containment is documented in NUREG-1843, Sections 2.3.2.1.1 and 2.4.1.1. Management of aging effects on the primary containment is documented in NUREG-1843, Sections 3.2.2 and 3.5.2. 2-289

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) Technical Evaluation NEDC-33004P-A, Revision 4, Constant Pressure Power Uprate, Class III, July 2003 (also referred to as CLTR) was approved by the NRC as an acceptable method for evaluating the effects of EPUs. Section 10.1 of the CLTR addresses the effect of EPU on subcompartment analyses. The results of this evaluation are described below. Browns Ferry meets all CLTR dispositions. The topics addressed in this evaluation are: Browns Ferry Topic CLTR Disposition Result Meets CLTR Liquid Lines Plant Specific Disposition As stated in Section 10.1 of the CLTR, EPU may increase subcooling in the reactor vessel, which may lead to increased break flow rates for liquid line breaks. An annular structure is located inside the drywell around the RPV in order to provide thermal and radiation shielding, and is called the Sacrificial Shield Wall (SSW). The SSW is designed to withstand the differential pressure that would develop across the wall as a result of a high pressure pipe break within the annulus (i.e., between the RPV and the SSW). For Browns Ferry, pipes with nominal diameters of four inches or smaller are the only reactor coolant lines investigated, because the reactor vessel safe-end welds for these nozzles are located within the sacrificial shield area. The minimum wall thickness for the various piping systems occurs at the safe-end joint to the piping. All other sections from this joint back to the reactor vessel have thicker wall sections and, therefore, have lower stresses. The largest line which has the safe end located in the annulus is the 4-inch jet pump instrument line. The EPU assessment is an extension of the prior 5% power uprate assessment. The Moody slip critical flow model is used to generate critical mass flux values based on downcomer conditions corresponding to OLTP, Current Licensed Thermal Power (CLTP), EPU thermal power and the operating condition with maximum downcomer subcooling (minimum recirculation pump speed with feedwater temperature reduction). Shield wall differential pressure estimates are developed by scaling the OLTP analysis shield wall differential pressure based on the following assumptions:

1. Non-condensable gases that are initially in the annulus are effectively vented by the time of peak annulus to drywell vent differential pressure.

2. The break fluid flashes to a two phase homogeneous mixture with 100% water entrainment.

The mixture saturation pressure is equal to the annulus pressure, and the mixture enthalpy is equal to the break fluid enthalpy.

3. The drywell pressure is assumed to remain constant during the event and to be identical for all operating conditions. Evaluations were performed for a set of drywell pressures ranging from 14.4 psia to 17.0 psia to ensure conservative differential pressure estimates. Note that 2-290

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) the results of the shield wall differential pressure calculations demonstrate that the shield wall differential pressure estimates for a given condition vary by less than 1% for this range of initial drywell pressures.

4. Based on the low differential pressures, the annulus to drywell vent does not experience critical flow during the event. The annulus pressure is therefore equal to the sum of the vent differential pressure and the assumed drywell pressure.

5. Flow through the annulus to drywell vent is assumed to be isentropic.

Scaling is performed with a modified version of Darcys formula that accounts for second order effects of compressible vent flow. The following equation is used to estimate the shield wall annulus to drywell vent differential pressure: DP2 = DP1 * (Gc,22/Gc,12)* (1/2) (Y12/Y22) Where: DP2 is the SSW annulus to drywell vent differential pressure for the new operating condition. DP1 is the OLTP condition shield wall differential pressure (2.0 psid). Gc,2 is the Moody Slip critical mass flux based on the pressure and enthalpy of the fluid in the downcomer for the new operating condition. Gc,1 is the Moody Slip critical mass flux based on the pressure and enthalpy of the fluid in the downcomer for the OLTP operating condition. 1 is the density of the annulus mixture for the OLTP condition. 2 is the density of the annulus mixture for the new operating condition. Y1 is vapor expansion factor applicable to the OLTP condition. Y2 is vapor expansion factor applicable to the new operating condition. The vapor expansion factors (Y1 and Y2) are calculated as the ratio of the downstream density over the upstream density based on the assumption that the saturated two-phase mixture undergoes an isentropic expansion from the annulus pressure to the drywell pressure. The annulus pressure and annulus mixture enthalpy are used to determine the annulus mixture specific entropy. The downstream condition properties are established based on the assumption that the downstream fluid is a saturated two-phase mixture with a specific entropy equal to the upstream mixture specific entropy and a pressure equal to the assumed drywell pressure. The shield wall differential pressure is calculated by adding a conservative estimate of the static head of the two phase mixture in the annulus to the calculated annulus to drywell vent differential pressure estimate. The annulus pressure load on the biological shield wall due to a postulated break in a 4-inch jet pump instrument line nozzle is evaluated at EPU conditions. The CLTP annulus pressure load (2.4 psid), documented in UFSAR Section 12.2.2.6, remains bounding compared to the 102% EPU annulus pressure load of 2.3 psid for normal feedwater temperature operation. For 2-291

NEDO-33860 Revision 1 Non-Proprietary Information - Class I (Public) reduced feedwater temperature operation at 102% EPU power, the annulus pressure load is 2.5 psid. For the limiting minimum pump speed, reduced feedwater temperature operating condition, the annulus pressure load is 3.6 psid. The results of the EPU evaluation, which addresses the effects of both EPU and operation at the limiting off-rated condition along the MELLLA operating domain upper boundary (Minimum Recirculation Pump Speed (MPS) point with feedwater temperature reduction), indicate that the SSW pressure difference design limit is not exceeded. 102% CLTP 102% EPU 102% EPU 55.4% EPU(1) Design Parameter MELLLA Rated Flow Rated Flow MELLLA Line Limit NFWT NFWT RFWT RFWT (psid) Critical Mass 10,091.9 9,803.6 10,244.7 12,485.8 N/A Flux (lbm/sec-ft2) Maximum SSW 2.4 2.3 2.5 3.6 19 DP (psid) Note:

1. 102% of the power level at the intersection of the MELLLA line and the minimum pump speed line (1.02

54.3%).

Subcompartment Pressurization Evaluation As discussed earlier, the differential pressure loading on the SSW is not significantly affected by EPU. For normal feedwater temperature operation, EPU implementation will result in a small decrease in SSW differential pressures. Increased downcomer subcooling associated with reduced feedwater temperature operation at EPU results in a small increase in SSW differential pressures. Reduced feedwater temperature operation at the intersection of the MELLLA line and the minimum pump speed line, which is not affected by EPU implementation, produces the highest downcomer subcooling, and limiting SSW differential pressures. The peak SSW differential

v t e Letter Sequence Supplement
CAC:MF6741, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance, Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance, Withholding Request Acceptance Meeting, Meeting, Meeting Questions 25 February 2016 25 February 2016 18 February 2016 29 February 2016 15 March 2016 4 April 2016 25 March 2016 14 April 2016 29 March 2016 7 April 2016 14 April 2016 20 April 2016 13 June 2016 21 June 2016 21 July 2016 Answers 27 May 2016 3 August 2016 Results Approval Other: CNL-16-008, Update on Progress of Facility Modifications for MSIV Leak Rate, CNL-18-112, Extended Power Uprate - Flow Induced Vibration Summary Report, CNL-18-124, Extended Power Uprate - Replacement Steam Dryer Final Load Definition and Stress Reports, CNL-18-134, Extended Power Uprate - Replacement Steam Dryer Revised Analysis and Limit Curves Report, ML16085A301, ML16105A441, ML16168A026, ML16176A316
MONTHYEAR2016-04-07 [Table View]

CNL-16-169, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 35, Consolidated Power Uprate Safety Analysis Report Revision

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