ML15295A025

From kanterella
Jump to navigation Jump to search
Updated Final Safety Analysis Report, Amendment 26
ML15295A025
Person / Time
Site: Browns Ferry  Tennessee Valley Authority icon.png
Issue date: 10/05/2015
From: Bono S
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML15295A025 (16)
v • d • e


Text

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 October 5, 2015 10 CER 50.4 10 CFR 50.71 (e)(4)

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

Subject:

Browns Ferry Nuclear Plant - Updated Final Safety Analysis Report, Amendment 26 Pursuant to the requirements of Title 10 of the Code of Federal Regulations (10 CFR)

Part 50.71, "Maintenance of Records, Making of Reports," paragraph (e)(4), the Tennessee Valley Authority (TVA) is submitting Amendment 26 of the Updated Final Safety Analysis Report (UFSAR) for the Browns Ferry Nuclear Plant (BEN).

Amendment 26 of the BEN UFSAR is a 24-month update that reflects editorial corrections, changes, and analyses pursuant to 10 CFR 50.59, "Changes Tests, and Experiments," to the facility, which were in effect through July 1, 2015. All UFSAR pages issued as a result of this update are clearly delineated with a "BFN-26" in the page header and a revision bar is shown in the right margin next to the specific change.

The summaries of evaluations of these changes made under the provisions of 10 CFR 50.59, will be provided to the NRC in separate correspondence by November 30, 2015.

The Enclosure to this letter contains one (1) Compact Disk - Recordable (CD-R) of the BEN UFSAR, Amendment 26. TVA has also included a listing of the UFSAR files and their size in the enclosure. Please note that some files do not have embedded fonts; however, these files are from a scanned source and are thus consistent with Section 2.23 of the "Guidance for Electronic Submissions to the NRC, Revision 6.1."

TVA has determined that some of the informration contained on the CD-R is considered to be sensitive unclassified non-safeguards information. As such, the enclosed CD-R is labeled "Tennessee Valley Authority, Browns Ferry Nuclear Plant, Updated Final Safety Analysis,*.

Report, Amendment 26, Non-Public Version." This CD'R submittal of the entire BEN UFSAR {, LL supersedes all previously submitted hard copies and CD-R copies of the document.

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

U.S. Nuclear Regulatory Commission Page 2 October 5, 2015 Pursuant to the requirements of 10 CFR 54.37, "Additional Records and Recordkeeping Requirements," paragraph (b), TVA has determined that there are no newly identified systems, structures, and components that would have been subject to an aging management review or evaluation of time-limited aging analyses.

I certify that I am duly authorized by TVA, and that, to the best of my knowledge and belief, the information contained herein accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirements.

There are no new regulatory commitments contained in this letter. If you have any questions, please contact Jamie L. Paul, Nuclear Site Licensing Manager, at (256) 729-2636.

Enclosures:

CD-ROM Containing Browns Ferry Nuclear Plant Updated Final Safety Analyses Report, Amendment 26, Non-Public Version and a Listing of the UFSAR Files, Including File Size cc (w/ Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

U.S. Nuclear Regulatory Commission Page 3 October 5, 2015 JLP:MWO Enclosure bcc (wi Enclosure):

NRC Project Manager - Browns Ferry Nuclear Plant bcc (w/o Enclosure):

M. A. Balduzzi G. A. Boerschig S. M. Bono K. H. Bronson C. R. Church D. M. Czufin S. M. Douglas J. P. Grimes E. K. Henderson D. L. Hughes S. W. Hunnewell G. W. Maudlin J. L. Paul E. D. Schrull J. W. Shea P. B. Summers, Jr.

S. A. Vance B. A. Wetzel G. R. Williams P. R. Wilson CNL-1 5-199

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

Enclosure Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Updated Final Safety Analysis Report (UFSAR)

Amendment 26 This enclosure provides the following:

  • One CD-ROM Copy of Amendment 26 of the Browns Ferry Nuclear Plant UFSAR, Non-Public Version
  • Listing of the Browns Ferry Nuclear Plant UFSAR, Amendment 26 files, including file size

BROWNS FERRY NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR), AMENDMENT 26 CHAPTERS 1 THROUGH 14 AND APPENDICES A THROUGH 0 LIST OF FILES ON DISK (ALL LISTED FILES - NON-PUBLIC VERSION)

FILE NAME FILE SIZE (KB) 01 BEN FSAR Amendment 26 EPL 001 BEN FSAR Table of Contents CHAPTER 1 - INTRODUCTION AND

SUMMARY

002 1.1 Project Identification 54 003 1.2 Definitions 66 004 1.2 Figure - Definitions 78 005 1.3 Methods of Technical Presentation 41 006 1.3 Table - Methods of Technical Presentation 85 007 1.3 Figure - Methods of Technical Presentation 80 008 1.4 Class of BWR Systems Criteria Requirements for Safety Evaluation 37 009 1.4 Table - Class of BWR Systems Criteria. Requirements for Safety Evaluation 76 010 1.5 Principal Design Criteria 60 011 1.6 Plant Description 89 012 1.6 Figure - Plant Description 1136 013 1.7 Comparison of Principal Design Characteristics 35 014 1.7 Table - Comparison of Principal Design Characteristics 84 015 1.8 Summary of Radiation Effects 34 016 1.9 Plant Management 29 017 1.10 Quality Assurance Program 31 018 1.11 Identification Resolution of Construction Permit Concern - Summary 37 019 1.11 Table - Identification Resolution of Construction Permit Concern - Summary 59 020 1.12 General Conclusions 35 021 1-Table of Contents 56 CHAPTER 2 - SITE 022 2.1 Summary Description 34 023 2.2 Site Description 13 024 2.2 Table - Site Description 85 025 2.2 Figure - Site Description 362 I

FILE NAME FILE SIZE (KB) 026 2.3 Meteorology 64 027 2.3 Table - Meteorology 122 028 2,3 Figure - Meteorology 1695 029 2.4 Hydrology, Water Quality, and Aquatic Biology 58 030 2.4 Table - Hydrology, Water Quality, and Aquatic Biology 43 031 2.4 Figure - Hydrology, Water Quality, and Aquatic Biology 782 032 2.4A Maximum Possible Flood 69 033 2.4A Table - Maximum Possible Flood 449 034 2-4A-Table 1 Maximum Possible Flood 390 035 2.4A Figure - Maximum Possible Flood 5417 036 2.5 Geology and Seismology 77 037 2.5 Table - Geology and Seismology 49 038 2.5 Figure - Geology and Seismology 2093 039 2.6 Environmental Radiological Monitoring Program 38 040 2 Table of Contents 58 CHAPTER 3 - REACTOR 041 3,1 Summary Description 38 042 3,2 Fuel Mechanical Design 56 043 3,2 Table - Fuel Mechanical Design 27 044 3.2 Figure - Fuel Mechanical Design 387 045 3.3 Reactor Vessel Internals Mechanical Design 74 046 3,3 Table - Reactor Vessel Internals Mechanical Design 38 047 3.3 Figure - Reactor Vessel Internals 1020 048 3.4 Reactivity Control Mechanical Design 102 049 3.4 Figure - Reactivity Control Mechanical Design 1879 050 3,5 Control Rod Drive Housing Supports 46 051 3,5 Figure - Control Rod Drive 420 052 3,6 Nuclear Design 59 053 3.6 table - Nuclear Design 41 054 3,6 Figure - Nuclear Design 390 055 3.7 Thermal and Hydraulic Design 144 056 3.7 Figure - Thermal and Hydraulic Design 624 2

FILE NAME FILE SIZE (KB) 057 3.8 Standby Liquid Control 36 058 3.8 Figure - Standby Liquid Control 472 059 3 Table of Contents 43 CHAPTER 4 - REACTOR COOLANT SYSTEM 060 4.1 Summary Description 39 061 4.2 Reactor Vessel and Appurtenances Mechanical Design 76 062 4.2 Table - Reactor Pressure Vessel Materials 44 063 4.2 Figure - Reactor Vessel and Appurtenances Mechanical Design 1165 064 Sytem57.3 Ractr Reircuatin 065 .3 Sytem46 able- Racto Reircuatin 066 ecicultio

.3 Sytem1269 igue -Reator 067 4.4 Nuclear Steam Pressure Relief System 38 068 4.4 Table -Nuclear Steam Pressure Relief System 38 069 4.4 Figure - Nuclear Steam Pressure Relief System 1230 070 4.5 Main Steam Line Flow Restrictor 40 071 4.5 Figure - Main Steam Line Flow Restrictor 287 072 4.6 Main Steam Isolation Valves 58 073 4.6 Figure - Main Steam Isolation Valves 419 074 4.7 Reactor Core Isolation Cooling 48 075 4.7 Table -Reactor Core Isolation Cooling 41 076 Cve soltionCooing1956

.7 igur - eacor 077 .8 esiual eatRemval63 078 eatRemval44

.8 abl - esidal 079 .8 igue - esiualHeatRemval421 080 4.9 Reactor Water Cleanup 1 081 4.9 Table - Reactor Water Cleanup 4 082 4.9 Figure - Reactor Water Cleanup 20 083 4.10 Nuclear System Leakage Rate Limits 5 084 4.10 Figure - Nuclear System Leakage Rate44 085 4.11 Main Steam Lines, Feedwater Piping and Drains 39 086 4.11 Figure - Main Steam Lines, Feedwater Piping and Drains 67 087 4.12 Inservice Inspection and Testing 38 3

FILE NAME FILE SIZE (KB) 088 4- Tabie of Contents 46 CHAPTER 5 - CONTAINMENT 089 5.1 Summary Description 34

-090 5.2 Primary Containment 496 091 5.2 Table - Primary Containment 58 092 5.2 Figure - Primary Containment 4784

-093 5.3 Secondary Containment 93 094 5.3 Figure - Secondary Containment 1454 095 5 -Table of Contents 41 CHAPTER 6 - EMERGENCY CORE COOLING SYSTEM

-096 6.1 Safety Objective 33

-097 6.2 Safety Design Basis 36 098 6.3 Emergency Core Cooling 38

-099 6.3 Table - Emergency Core Cooling 43 100 6.3 Figure - Emergency Core Cooling 389 101 6.4 Description 39 102 6.4 Table - Description 24 103 6.4 Figure - Description 1890' 104 6.5 Safety Evaluation 62 105 6.5 Table - Safety Evaluation 18 1*06 6.5 Figure -Safety Evaluation 712 107 6.6 Inspection and Testing 41 "1086 Table of Contents 39 CHAPTER 7 - CONTROL AND INSTRUMENTATION 109 7.1 Summary Description 45

-110 7.1 Figure - Summary Description 420 1117.2 Reactor Protection System 78 112 7.2 Table - Reactor Protection System 42 113 7.2 Figure - Reactor Protection System 2278 114 7.3 Primary Containment Isolation 75 115 7.3 Table - Primary Containment Isolation 40

-116 7.3 Figure - Primary Containment Isolation 868 4

FILE NAME FILE SIZE (KB) 117 7.4 Emergency Core Cooling Control 87 118 7.4 Table - Emergency Core Cooling Control 46 119 7.4 Figure - Emergency Core Cooling Control 4788 120 7.5 Neutron Monitoring System 84 121 7.5 Table - Neutron Monitoring System 38 122 7.5 Figure- Neutron Monitoring System 1788 123 7.6 Refueling Interlocks 4 124 7.6 Table - Refueling Interlocks 37 125 7.6 Figure - Refueling Interlocks 450 126 7.7 Reactor Manual Control System 58 127 7.7 Table - Reactor Manual Control System 42 128 7.7 Figure - Reactor Manual Control System 506 129 7.7A_B Rod Sequence Control 32 130 7.8 Reactor Vessel Instrumentation 48 131 7.8 Table - Reactor Vessel Instrumentation 45 132 7.8 Figure - Reactor Vessel Instrumentation 1873 133 7.9 Recirculation Flow Control 57 134 7.9 Figure - Recirculation Flow Control 31 135 7.10 Feedwater Control System 54 136 7.10 Figure - Feedwater Control System 763 137 7.11 Pressure Regulator and Turbine 46 138 7.11 Figure - Pressure Regulator and Turbine 436 139 7.12 Process Radiation Monitoring 55 140 7.12 Table - Process Radiation Monitoring 36 141 7.12 Figure - Process Radiation Monitoring 2423 142 7.13 Area Radiation Monitoring 41 143 7.13 Table -Area Radiation Monitoring 50 144 7.13 Figure - Area Radiation Monitoring 389 145 7.14 Drywell Leak Detection Radiation 37 146 7.15 Health Physics Laboratory 35 147 7.16 Process Computer System 53 148 7.16 Table - Process Computer System 25 5

FILE NAME FILE SIZE (KB) 149 7.16 Figure - Process Computer System 388 150 7.17 Deleted 28 151 7.17 Table -Deleted 24 152 7.17 Figure - Deleted 391 153 7.18 Backup Control System 4 154 7.19 Anticipated Transient Without Scram 45 155 7.20 Instrument Setpoint Methodology 44 156 7.20 Table - Instrument Setpoint Methodology 50 157 7 Table of Contents 53 CHAPTER 8 - ELECTRICAL POWER SYSTEMS 158 8.1 Summary Description 41 159 8.2 Generators 36 160 8.3 Transmission System 16 161 8.3 Figure - Transmission System 4193 162 8.4 Normal Auxiliary Power 79 163 8.4 Table - Normal Auxiliary Power 5 164 8.4 Figure - Normal Auxiliary Power 1385 165 8.5 Standby AC Power Supply 47 166 8.5 Table - Standby AC Power Supply 112 167 8.5 Figure - Standby AC Power Supply 19510 168 8.6 250-V DC Power Supply 49 169 8.6 Table - 250-V DC Power Supply 37 170 8.6 Figure - 250-V DC Power Supply 1296 171 8.7 120-V AC Power Supply 42 172 8.7 Figure-I120-V AC Power Supply 978 173 8.8 Auxiliary DC Power Supply 48 174 8.8 Table -Auxiliary DC Power Supply 37 175 8.8 Figure - Auxiliary DC Power Supply 391 176 8.9 Safety Systems Independence Criteria 57 177 8.9 Table - Safety Systems Independence Criteria 25 178 8.9 Figure - Safety Systems Independence Criteria 389 179 8.10 Station Blackout 39 6

FILE NAME FILE SIZE (KB) 180 8 Table of Contents 20 CHAPTER 9 - RADIOACTIVE WASTE CONTROL SYSTEMS 181 9.1 Summary Description 3 182 9.2 Liquid Radwaste System 26 183 9.2 Table - Liquid Radwaste System 49 184 9.2 Figure - Liquid Radwaste System 3475 185 9.3 Solid Radwaste System 45 187Sytem25.4 Gseos Rawase 190 9.5 Figure - Gaseous Radwaste System - Modified 1432 191 9 Table of Contents 38 CHAPTER 10 - AUXILIARY SYSTEMS 192 10.1 Summary Description 34 193 10.2 New Fuel Storage 40 194 0.2Figue -NewFuelStoage437

-19510.3Spen Fue Stoage66 198 10.4 Table - Tools and Servicing Equipment 36 199 10.4 Figure - Tools and Servicing Equipment 439 200 10.5 Fuel Pool Cooling and Cleanup 52 201 10.5 Table - Fuel Pool Cooling and Cleanup 41 202 10.5 Figure - Fuel Pool Cooling and Cleanup 2086 203 10.6 Reactor Building Closed Cooling 41 204 10.6 Table - Reactor Building Closed Cooling 80 205 10.6 Figure -Reactor Building Closed Cooling 711 206 10.7 Raw Cooling Water 39 207 10.7 Figure -Raw Cooling Water 2671 208 10.8 Raw Service Water 37 209 10.9 RHR Service Water 19 7

FILE NAME FILE SIZE 210 10.9 Figure - RHR Service Water 1313 211 10.10 Emergency Equipment Cooiing Water 44 212 10.10 Figure -Emergency Equipment Cooling Water 1665 213 10.11 Fire Protection System 24 214 10.11 Figure- Fire Protection System 391 215 10.12 Heating Ventilating and Air Conditioning 60 216 10.12 Figure - Heating Ventilating and Air Conditioning 2122 217 10.13 Demineralized Water System 42 218 10.13 Figure - Demineralized Water System 390 219 10.14 Control and Service Air 46 220 10.14 Figure -Control and Service Air 1249 221 10.15 Potable Water and Sanitary 32 222 10.16 Equipment and Floor Drainage 45 223 10.17 Process Sampling Systems 41 224 10.17 Table - Process Sampling Systems 42 225 10.17 Figure - Process Sampling Systems 1204 226 10.18 Plant Communications System 49 227 10.18 Figure -Plant Communications System 726 228 10.19 Lighting System 36 229 10.20 Auxiliary Boiler System 4 230 10.21 Postaccident Sampling 43 231 10.21 Table - Postaccident Sampling 39 232 10.21 Figure - Postaccident Sampling 634 233 10.22 Auxiliary Decay Heat Removal 37 234 10.23 Hydrogen Water Chemistry 44 235 10 Table of Contents 47 CHAPTER 11 - POWER CONVERSION SYSTEMS 236 11.1 Summary Description 30 237 11.1 Figure - Summary Description 910 238 11.2 Turbine Generator 48 239 11.3 Main Condenser 44 240 11.4 Main Condenser Gas Removal 37 8

FILE NAME FILE SIZE 241 11.5 Turbine Bypass 36 242 11.6 Condenser Circulating Water 21 242a 11.6 Table -Condenser Circulating Water 21 243 11.6 Figure - Condenser Circulating Water 1512 244 11.7 Condensate Filter Demineralizer 41 245 11.7 Figure - Condensate Filter Demineralizer 394 246 11.8 Condensate and Reactor Feedwater 43 250 Stoage1140 1.9 igue - ondesat 251 11 Table of Contents 15 CHAPTER 12 - STRUCTURES AND SHIELDING 252 12.1 Summary Description 34 253 12.2 Principal Structures and Foundations 227 254 12.2 Table - Principal Structures and Foundations 145 255 12.2 Figure - Principal Structures and Foundations 9257 256 12.3 Shielding and Radiation 49 257 12.3 Table - Shielding and Radiation 32 258 12 Table of Contents 44 262 13.3 Training Programs 35 263 13.4 Preoperational Test Program 115 264 13.4 Figure - Preoperational Test Program 426 265 13.5 Startup and Power Test Program 144 266 13.5 Table - Startup and Power Test Program 68 267 13.5 Figure - Startup and Power Test Program 607 268 13.6 Normal Operations 44 269 13.6 Figure -Normal Operations 390 9

FILE NAME FILE SIZE (KB) 270 13.7 Records 23 271 13.8 Operational Review and Audits 38 272 13.9 Refueling Operations 3 273 13.10 Refueling Test Program 57___

274 13 Table of Contents 39 CHAPTER 14 - PLANT SAFETY SYSTEMS 275 14.1 Plant Safety Analysis 36 276 14.2 Unacceptable Safety Results for Abnormal Operational Transients 32 2h 7 o9S a e4ty.4AnTala sibs3l - p p r o c 97 280 14.4 Figure - Approach to Safety Analysis 532 281 14.5 Analyses of Abnormal Operational Transients 162 282 14.5 Table - Analyses of Abnormal Operational Transients 21 283 14.5 Figure - Analyses of Abnormal Operational Transients 1972 284 14.6 Analysis of Design Basis Accidents 106 285 Dsig4.6Tabe Bais Acidnts65 - nalsisof 286 esig 4.6Figue Bais Acidnts781 -Analsisof 2 8 7 1 .7 C n cl u i o n s3 93 288 4.8 naltica Metods462 289 14.8 Table - Analytical Methods 403 290 14.8 Figure - Analytical Methods 433 291 14.9 Dose Sensitivity Evaluation 400 292 14.9 Table - Dose Sensitivity Evaluation 401 293 14.10 Analyses of Abnormal Operational 428 294 ofAbnrma 4.1 Fiure Opeatinal1118 Anlyss 295 Bais 4.1 ccidnts483*

Anaysi of esin 296 esig 4.1 Bais Tabe Acidnts404

-Analsisof 297 14.11 Figure -Analysis of Design Basis Accidents 709 298 14 Table of Contents 25 10

FILE NAME FILE SIZE (KB)

APPENDIX A - CONFORMANCE OF AEC PROPOSED GENERAL DESIGN CRITERIA

-299 Appendix A Conformance to AEC Propose 128

-300 Appendix A Table - Conformance to AEC Propose 52

-303 Appendix C -Structural Qualification 134 304 Table - Appendix C - Structural Qualification 236

-305 Appendix C - Table of Contents 232

-APPENDIX D - QUALITY ASSURANCE PLAN E Sit GaeusRaitAssrneR

-309 iurAppendix late 360

-3108 Appendix E - Table of Contents 327

-APPENDIX F - UNIT SHARING AND INTERACTIONS

-311 Appendix F - Unit Sharing and Interactions 71

-312 Appendix F - Table of Contents 41

-316 Appendix G Marc Plant Nuclear Safety Operational 461 "317 Appendix G - Table of Contents 41

-318 Appendix G STATE A 684

-319 Appendix G STATE B 689 "320Appendix G STATE C 693

-321 Appendix G STATE D 693

-322 Appendix G STATE E 692

'323 Appendix G STATE F 693 11

FILE NAME FILE SIZE (KB)

APPENDIX H - CORE THERMAL DESIGN 324 Appendix H - Core Thermal Design 25 APPENDIX I - IDENTIFICATION-RESOLUTION OF CONSTRUCTION PERMIT CONCERNS 327 Appendix J - Reactor Pressure Vessel Design 39 328 Appendix J - Table of Contents 3 APPENDIX K - REACTOR PRESSURE VESSEL DESIGN

SUMMARY

REPORT - UNIT 2 329 Appendix K- Reactor Pressure Vessel Design 3 330 Appendix K - Table of Contents 3 APPENDIX L - REACTOR PRESSURE VESSEL DESIGN

SUMMARY

REPORT - UNIT 3 331 Appendix L - Reactor Pressure Vessel Design 39

-332 Appendix L - Table of Contents 31 APPENDIX M - REPORT PIPE FAILURES OUTSIDE CONTAINMENT 333 Appendix M- Report on Pipe Failures Outside 113 334 Table - Appendix M- Report on Pipe Failures 25 335 Figure - Appendix M- Report on Pipe Failures 729 3"36 Appendix M - Table of Contents 39 APPENDIX N - RELOAD LICENSE REPORT 337 Appendix N - Unit 1 Cycle 11 Reload Safety Analysis (July, 2014) 1293 338 Appendix N - Unit 2 Cycle 18 Reload Safety Analysis (July, 2014) 16244 3"39 Appendix N - Unit 3 Cycle 17 Reload Analysis - Revision 2 (July, 2014) 726 340 Appendix N - Table of Contents 7 APPENDIX 0 - AGING MANAGEMENT PROGRAM 3"41 Appendix 0 - Introduction 100 342 Appendix 0 - Table of Contents 138 12

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 October 5, 2015 10 CER 50.4 10 CFR 50.71 (e)(4)

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

Subject:

Browns Ferry Nuclear Plant - Updated Final Safety Analysis Report, Amendment 26 Pursuant to the requirements of Title 10 of the Code of Federal Regulations (10 CFR)

Part 50.71, "Maintenance of Records, Making of Reports," paragraph (e)(4), the Tennessee Valley Authority (TVA) is submitting Amendment 26 of the Updated Final Safety Analysis Report (UFSAR) for the Browns Ferry Nuclear Plant (BEN).

Amendment 26 of the BEN UFSAR is a 24-month update that reflects editorial corrections, changes, and analyses pursuant to 10 CFR 50.59, "Changes Tests, and Experiments," to the facility, which were in effect through July 1, 2015. All UFSAR pages issued as a result of this update are clearly delineated with a "BFN-26" in the page header and a revision bar is shown in the right margin next to the specific change.

The summaries of evaluations of these changes made under the provisions of 10 CFR 50.59, will be provided to the NRC in separate correspondence by November 30, 2015.

The Enclosure to this letter contains one (1) Compact Disk - Recordable (CD-R) of the BEN UFSAR, Amendment 26. TVA has also included a listing of the UFSAR files and their size in the enclosure. Please note that some files do not have embedded fonts; however, these files are from a scanned source and are thus consistent with Section 2.23 of the "Guidance for Electronic Submissions to the NRC, Revision 6.1."

TVA has determined that some of the informration contained on the CD-R is considered to be sensitive unclassified non-safeguards information. As such, the enclosed CD-R is labeled "Tennessee Valley Authority, Browns Ferry Nuclear Plant, Updated Final Safety Analysis,*.

Report, Amendment 26, Non-Public Version." This CD'R submittal of the entire BEN UFSAR {, LL supersedes all previously submitted hard copies and CD-R copies of the document.

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

U.S. Nuclear Regulatory Commission Page 2 October 5, 2015 Pursuant to the requirements of 10 CFR 54.37, "Additional Records and Recordkeeping Requirements," paragraph (b), TVA has determined that there are no newly identified systems, structures, and components that would have been subject to an aging management review or evaluation of time-limited aging analyses.

I certify that I am duly authorized by TVA, and that, to the best of my knowledge and belief, the information contained herein accurately presents changes made since the previous submittal, necessary to reflect information and analyses submitted to the Commission or prepared pursuant to Commission requirements.

There are no new regulatory commitments contained in this letter. If you have any questions, please contact Jamie L. Paul, Nuclear Site Licensing Manager, at (256) 729-2636.

Enclosures:

CD-ROM Containing Browns Ferry Nuclear Plant Updated Final Safety Analyses Report, Amendment 26, Non-Public Version and a Listing of the UFSAR Files, Including File Size cc (w/ Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

U.S. Nuclear Regulatory Commission Page 3 October 5, 2015 JLP:MWO Enclosure bcc (wi Enclosure):

NRC Project Manager - Browns Ferry Nuclear Plant bcc (w/o Enclosure):

M. A. Balduzzi G. A. Boerschig S. M. Bono K. H. Bronson C. R. Church D. M. Czufin S. M. Douglas J. P. Grimes E. K. Henderson D. L. Hughes S. W. Hunnewell G. W. Maudlin J. L. Paul E. D. Schrull J. W. Shea P. B. Summers, Jr.

S. A. Vance B. A. Wetzel G. R. Williams P. R. Wilson CNL-1 5-199

Security-Related Information Withhold Under 10 CFR 2.390.

This letter is decontrolled when separated from the enclosure.

Enclosure Tennessee Valley Authority Browns Ferry Nuclear Plant Units 1, 2, and 3 Updated Final Safety Analysis Report (UFSAR)

Amendment 26 This enclosure provides the following:

  • One CD-ROM Copy of Amendment 26 of the Browns Ferry Nuclear Plant UFSAR, Non-Public Version
  • Listing of the Browns Ferry Nuclear Plant UFSAR, Amendment 26 files, including file size

BROWNS FERRY NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT (UFSAR), AMENDMENT 26 CHAPTERS 1 THROUGH 14 AND APPENDICES A THROUGH 0 LIST OF FILES ON DISK (ALL LISTED FILES - NON-PUBLIC VERSION)

FILE NAME FILE SIZE (KB) 01 BEN FSAR Amendment 26 EPL 001 BEN FSAR Table of Contents CHAPTER 1 - INTRODUCTION AND

SUMMARY

002 1.1 Project Identification 54 003 1.2 Definitions 66 004 1.2 Figure - Definitions 78 005 1.3 Methods of Technical Presentation 41 006 1.3 Table - Methods of Technical Presentation 85 007 1.3 Figure - Methods of Technical Presentation 80 008 1.4 Class of BWR Systems Criteria Requirements for Safety Evaluation 37 009 1.4 Table - Class of BWR Systems Criteria. Requirements for Safety Evaluation 76 010 1.5 Principal Design Criteria 60 011 1.6 Plant Description 89 012 1.6 Figure - Plant Description 1136 013 1.7 Comparison of Principal Design Characteristics 35 014 1.7 Table - Comparison of Principal Design Characteristics 84 015 1.8 Summary of Radiation Effects 34 016 1.9 Plant Management 29 017 1.10 Quality Assurance Program 31 018 1.11 Identification Resolution of Construction Permit Concern - Summary 37 019 1.11 Table - Identification Resolution of Construction Permit Concern - Summary 59 020 1.12 General Conclusions 35 021 1-Table of Contents 56 CHAPTER 2 - SITE 022 2.1 Summary Description 34 023 2.2 Site Description 13 024 2.2 Table - Site Description 85 025 2.2 Figure - Site Description 362 I

FILE NAME FILE SIZE (KB) 026 2.3 Meteorology 64 027 2.3 Table - Meteorology 122 028 2,3 Figure - Meteorology 1695 029 2.4 Hydrology, Water Quality, and Aquatic Biology 58 030 2.4 Table - Hydrology, Water Quality, and Aquatic Biology 43 031 2.4 Figure - Hydrology, Water Quality, and Aquatic Biology 782 032 2.4A Maximum Possible Flood 69 033 2.4A Table - Maximum Possible Flood 449 034 2-4A-Table 1 Maximum Possible Flood 390 035 2.4A Figure - Maximum Possible Flood 5417 036 2.5 Geology and Seismology 77 037 2.5 Table - Geology and Seismology 49 038 2.5 Figure - Geology and Seismology 2093 039 2.6 Environmental Radiological Monitoring Program 38 040 2 Table of Contents 58 CHAPTER 3 - REACTOR 041 3,1 Summary Description 38 042 3,2 Fuel Mechanical Design 56 043 3,2 Table - Fuel Mechanical Design 27 044 3.2 Figure - Fuel Mechanical Design 387 045 3.3 Reactor Vessel Internals Mechanical Design 74 046 3,3 Table - Reactor Vessel Internals Mechanical Design 38 047 3.3 Figure - Reactor Vessel Internals 1020 048 3.4 Reactivity Control Mechanical Design 102 049 3.4 Figure - Reactivity Control Mechanical Design 1879 050 3,5 Control Rod Drive Housing Supports 46 051 3,5 Figure - Control Rod Drive 420 052 3,6 Nuclear Design 59 053 3.6 table - Nuclear Design 41 054 3,6 Figure - Nuclear Design 390 055 3.7 Thermal and Hydraulic Design 144 056 3.7 Figure - Thermal and Hydraulic Design 624 2

FILE NAME FILE SIZE (KB) 057 3.8 Standby Liquid Control 36 058 3.8 Figure - Standby Liquid Control 472 059 3 Table of Contents 43 CHAPTER 4 - REACTOR COOLANT SYSTEM 060 4.1 Summary Description 39 061 4.2 Reactor Vessel and Appurtenances Mechanical Design 76 062 4.2 Table - Reactor Pressure Vessel Materials 44 063 4.2 Figure - Reactor Vessel and Appurtenances Mechanical Design 1165 064 Sytem57.3 Ractr Reircuatin 065 .3 Sytem46 able- Racto Reircuatin 066 ecicultio

.3 Sytem1269 igue -Reator 067 4.4 Nuclear Steam Pressure Relief System 38 068 4.4 Table -Nuclear Steam Pressure Relief System 38 069 4.4 Figure - Nuclear Steam Pressure Relief System 1230 070 4.5 Main Steam Line Flow Restrictor 40 071 4.5 Figure - Main Steam Line Flow Restrictor 287 072 4.6 Main Steam Isolation Valves 58 073 4.6 Figure - Main Steam Isolation Valves 419 074 4.7 Reactor Core Isolation Cooling 48 075 4.7 Table -Reactor Core Isolation Cooling 41 076 Cve soltionCooing1956

.7 igur - eacor 077 .8 esiual eatRemval63 078 eatRemval44

.8 abl - esidal 079 .8 igue - esiualHeatRemval421 080 4.9 Reactor Water Cleanup 1 081 4.9 Table - Reactor Water Cleanup 4 082 4.9 Figure - Reactor Water Cleanup 20 083 4.10 Nuclear System Leakage Rate Limits 5 084 4.10 Figure - Nuclear System Leakage Rate44 085 4.11 Main Steam Lines, Feedwater Piping and Drains 39 086 4.11 Figure - Main Steam Lines, Feedwater Piping and Drains 67 087 4.12 Inservice Inspection and Testing 38 3

FILE NAME FILE SIZE (KB) 088 4- Tabie of Contents 46 CHAPTER 5 - CONTAINMENT 089 5.1 Summary Description 34

-090 5.2 Primary Containment 496 091 5.2 Table - Primary Containment 58 092 5.2 Figure - Primary Containment 4784

-093 5.3 Secondary Containment 93 094 5.3 Figure - Secondary Containment 1454 095 5 -Table of Contents 41 CHAPTER 6 - EMERGENCY CORE COOLING SYSTEM

-096 6.1 Safety Objective 33

-097 6.2 Safety Design Basis 36 098 6.3 Emergency Core Cooling 38

-099 6.3 Table - Emergency Core Cooling 43 100 6.3 Figure - Emergency Core Cooling 389 101 6.4 Description 39 102 6.4 Table - Description 24 103 6.4 Figure - Description 1890' 104 6.5 Safety Evaluation 62 105 6.5 Table - Safety Evaluation 18 1*06 6.5 Figure -Safety Evaluation 712 107 6.6 Inspection and Testing 41 "1086 Table of Contents 39 CHAPTER 7 - CONTROL AND INSTRUMENTATION 109 7.1 Summary Description 45

-110 7.1 Figure - Summary Description 420 1117.2 Reactor Protection System 78 112 7.2 Table - Reactor Protection System 42 113 7.2 Figure - Reactor Protection System 2278 114 7.3 Primary Containment Isolation 75 115 7.3 Table - Primary Containment Isolation 40

-116 7.3 Figure - Primary Containment Isolation 868 4

FILE NAME FILE SIZE (KB) 117 7.4 Emergency Core Cooling Control 87 118 7.4 Table - Emergency Core Cooling Control 46 119 7.4 Figure - Emergency Core Cooling Control 4788 120 7.5 Neutron Monitoring System 84 121 7.5 Table - Neutron Monitoring System 38 122 7.5 Figure- Neutron Monitoring System 1788 123 7.6 Refueling Interlocks 4 124 7.6 Table - Refueling Interlocks 37 125 7.6 Figure - Refueling Interlocks 450 126 7.7 Reactor Manual Control System 58 127 7.7 Table - Reactor Manual Control System 42 128 7.7 Figure - Reactor Manual Control System 506 129 7.7A_B Rod Sequence Control 32 130 7.8 Reactor Vessel Instrumentation 48 131 7.8 Table - Reactor Vessel Instrumentation 45 132 7.8 Figure - Reactor Vessel Instrumentation 1873 133 7.9 Recirculation Flow Control 57 134 7.9 Figure - Recirculation Flow Control 31 135 7.10 Feedwater Control System 54 136 7.10 Figure - Feedwater Control System 763 137 7.11 Pressure Regulator and Turbine 46 138 7.11 Figure - Pressure Regulator and Turbine 436 139 7.12 Process Radiation Monitoring 55 140 7.12 Table - Process Radiation Monitoring 36 141 7.12 Figure - Process Radiation Monitoring 2423 142 7.13 Area Radiation Monitoring 41 143 7.13 Table -Area Radiation Monitoring 50 144 7.13 Figure - Area Radiation Monitoring 389 145 7.14 Drywell Leak Detection Radiation 37 146 7.15 Health Physics Laboratory 35 147 7.16 Process Computer System 53 148 7.16 Table - Process Computer System 25 5

FILE NAME FILE SIZE (KB) 149 7.16 Figure - Process Computer System 388 150 7.17 Deleted 28 151 7.17 Table -Deleted 24 152 7.17 Figure - Deleted 391 153 7.18 Backup Control System 4 154 7.19 Anticipated Transient Without Scram 45 155 7.20 Instrument Setpoint Methodology 44 156 7.20 Table - Instrument Setpoint Methodology 50 157 7 Table of Contents 53 CHAPTER 8 - ELECTRICAL POWER SYSTEMS 158 8.1 Summary Description 41 159 8.2 Generators 36 160 8.3 Transmission System 16 161 8.3 Figure - Transmission System 4193 162 8.4 Normal Auxiliary Power 79 163 8.4 Table - Normal Auxiliary Power 5 164 8.4 Figure - Normal Auxiliary Power 1385 165 8.5 Standby AC Power Supply 47 166 8.5 Table - Standby AC Power Supply 112 167 8.5 Figure - Standby AC Power Supply 19510 168 8.6 250-V DC Power Supply 49 169 8.6 Table - 250-V DC Power Supply 37 170 8.6 Figure - 250-V DC Power Supply 1296 171 8.7 120-V AC Power Supply 42 172 8.7 Figure-I120-V AC Power Supply 978 173 8.8 Auxiliary DC Power Supply 48 174 8.8 Table -Auxiliary DC Power Supply 37 175 8.8 Figure - Auxiliary DC Power Supply 391 176 8.9 Safety Systems Independence Criteria 57 177 8.9 Table - Safety Systems Independence Criteria 25 178 8.9 Figure - Safety Systems Independence Criteria 389 179 8.10 Station Blackout 39 6

FILE NAME FILE SIZE (KB) 180 8 Table of Contents 20 CHAPTER 9 - RADIOACTIVE WASTE CONTROL SYSTEMS 181 9.1 Summary Description 3 182 9.2 Liquid Radwaste System 26 183 9.2 Table - Liquid Radwaste System 49 184 9.2 Figure - Liquid Radwaste System 3475 185 9.3 Solid Radwaste System 45 187Sytem25.4 Gseos Rawase 190 9.5 Figure - Gaseous Radwaste System - Modified 1432 191 9 Table of Contents 38 CHAPTER 10 - AUXILIARY SYSTEMS 192 10.1 Summary Description 34 193 10.2 New Fuel Storage 40 194 0.2Figue -NewFuelStoage437

-19510.3Spen Fue Stoage66 198 10.4 Table - Tools and Servicing Equipment 36 199 10.4 Figure - Tools and Servicing Equipment 439 200 10.5 Fuel Pool Cooling and Cleanup 52 201 10.5 Table - Fuel Pool Cooling and Cleanup 41 202 10.5 Figure - Fuel Pool Cooling and Cleanup 2086 203 10.6 Reactor Building Closed Cooling 41 204 10.6 Table - Reactor Building Closed Cooling 80 205 10.6 Figure -Reactor Building Closed Cooling 711 206 10.7 Raw Cooling Water 39 207 10.7 Figure -Raw Cooling Water 2671 208 10.8 Raw Service Water 37 209 10.9 RHR Service Water 19 7

FILE NAME FILE SIZE 210 10.9 Figure - RHR Service Water 1313 211 10.10 Emergency Equipment Cooiing Water 44 212 10.10 Figure -Emergency Equipment Cooling Water 1665 213 10.11 Fire Protection System 24 214 10.11 Figure- Fire Protection System 391 215 10.12 Heating Ventilating and Air Conditioning 60 216 10.12 Figure - Heating Ventilating and Air Conditioning 2122 217 10.13 Demineralized Water System 42 218 10.13 Figure - Demineralized Water System 390 219 10.14 Control and Service Air 46 220 10.14 Figure -Control and Service Air 1249 221 10.15 Potable Water and Sanitary 32 222 10.16 Equipment and Floor Drainage 45 223 10.17 Process Sampling Systems 41 224 10.17 Table - Process Sampling Systems 42 225 10.17 Figure - Process Sampling Systems 1204 226 10.18 Plant Communications System 49 227 10.18 Figure -Plant Communications System 726 228 10.19 Lighting System 36 229 10.20 Auxiliary Boiler System 4 230 10.21 Postaccident Sampling 43 231 10.21 Table - Postaccident Sampling 39 232 10.21 Figure - Postaccident Sampling 634 233 10.22 Auxiliary Decay Heat Removal 37 234 10.23 Hydrogen Water Chemistry 44 235 10 Table of Contents 47 CHAPTER 11 - POWER CONVERSION SYSTEMS 236 11.1 Summary Description 30 237 11.1 Figure - Summary Description 910 238 11.2 Turbine Generator 48 239 11.3 Main Condenser 44 240 11.4 Main Condenser Gas Removal 37 8

FILE NAME FILE SIZE 241 11.5 Turbine Bypass 36 242 11.6 Condenser Circulating Water 21 242a 11.6 Table -Condenser Circulating Water 21 243 11.6 Figure - Condenser Circulating Water 1512 244 11.7 Condensate Filter Demineralizer 41 245 11.7 Figure - Condensate Filter Demineralizer 394 246 11.8 Condensate and Reactor Feedwater 43 250 Stoage1140 1.9 igue - ondesat 251 11 Table of Contents 15 CHAPTER 12 - STRUCTURES AND SHIELDING 252 12.1 Summary Description 34 253 12.2 Principal Structures and Foundations 227 254 12.2 Table - Principal Structures and Foundations 145 255 12.2 Figure - Principal Structures and Foundations 9257 256 12.3 Shielding and Radiation 49 257 12.3 Table - Shielding and Radiation 32 258 12 Table of Contents 44 262 13.3 Training Programs 35 263 13.4 Preoperational Test Program 115 264 13.4 Figure - Preoperational Test Program 426 265 13.5 Startup and Power Test Program 144 266 13.5 Table - Startup and Power Test Program 68 267 13.5 Figure - Startup and Power Test Program 607 268 13.6 Normal Operations 44 269 13.6 Figure -Normal Operations 390 9

FILE NAME FILE SIZE (KB) 270 13.7 Records 23 271 13.8 Operational Review and Audits 38 272 13.9 Refueling Operations 3 273 13.10 Refueling Test Program 57___

274 13 Table of Contents 39 CHAPTER 14 - PLANT SAFETY SYSTEMS 275 14.1 Plant Safety Analysis 36 276 14.2 Unacceptable Safety Results for Abnormal Operational Transients 32 2h 7 o9S a e4ty.4AnTala sibs3l - p p r o c 97 280 14.4 Figure - Approach to Safety Analysis 532 281 14.5 Analyses of Abnormal Operational Transients 162 282 14.5 Table - Analyses of Abnormal Operational Transients 21 283 14.5 Figure - Analyses of Abnormal Operational Transients 1972 284 14.6 Analysis of Design Basis Accidents 106 285 Dsig4.6Tabe Bais Acidnts65 - nalsisof 286 esig 4.6Figue Bais Acidnts781 -Analsisof 2 8 7 1 .7 C n cl u i o n s3 93 288 4.8 naltica Metods462 289 14.8 Table - Analytical Methods 403 290 14.8 Figure - Analytical Methods 433 291 14.9 Dose Sensitivity Evaluation 400 292 14.9 Table - Dose Sensitivity Evaluation 401 293 14.10 Analyses of Abnormal Operational 428 294 ofAbnrma 4.1 Fiure Opeatinal1118 Anlyss 295 Bais 4.1 ccidnts483*

Anaysi of esin 296 esig 4.1 Bais Tabe Acidnts404

-Analsisof 297 14.11 Figure -Analysis of Design Basis Accidents 709 298 14 Table of Contents 25 10

FILE NAME FILE SIZE (KB)

APPENDIX A - CONFORMANCE OF AEC PROPOSED GENERAL DESIGN CRITERIA

-299 Appendix A Conformance to AEC Propose 128

-300 Appendix A Table - Conformance to AEC Propose 52

-303 Appendix C -Structural Qualification 134 304 Table - Appendix C - Structural Qualification 236

-305 Appendix C - Table of Contents 232

-APPENDIX D - QUALITY ASSURANCE PLAN E Sit GaeusRaitAssrneR

-309 iurAppendix late 360

-3108 Appendix E - Table of Contents 327

-APPENDIX F - UNIT SHARING AND INTERACTIONS

-311 Appendix F - Unit Sharing and Interactions 71

-312 Appendix F - Table of Contents 41

-316 Appendix G Marc Plant Nuclear Safety Operational 461 "317 Appendix G - Table of Contents 41

-318 Appendix G STATE A 684

-319 Appendix G STATE B 689 "320Appendix G STATE C 693

-321 Appendix G STATE D 693

-322 Appendix G STATE E 692

'323 Appendix G STATE F 693 11

FILE NAME FILE SIZE (KB)

APPENDIX H - CORE THERMAL DESIGN 324 Appendix H - Core Thermal Design 25 APPENDIX I - IDENTIFICATION-RESOLUTION OF CONSTRUCTION PERMIT CONCERNS 327 Appendix J - Reactor Pressure Vessel Design 39 328 Appendix J - Table of Contents 3 APPENDIX K - REACTOR PRESSURE VESSEL DESIGN

SUMMARY

REPORT - UNIT 2 329 Appendix K- Reactor Pressure Vessel Design 3 330 Appendix K - Table of Contents 3 APPENDIX L - REACTOR PRESSURE VESSEL DESIGN

SUMMARY

REPORT - UNIT 3 331 Appendix L - Reactor Pressure Vessel Design 39

-332 Appendix L - Table of Contents 31 APPENDIX M - REPORT PIPE FAILURES OUTSIDE CONTAINMENT 333 Appendix M- Report on Pipe Failures Outside 113 334 Table - Appendix M- Report on Pipe Failures 25 335 Figure - Appendix M- Report on Pipe Failures 729 3"36 Appendix M - Table of Contents 39 APPENDIX N - RELOAD LICENSE REPORT 337 Appendix N - Unit 1 Cycle 11 Reload Safety Analysis (July, 2014) 1293 338 Appendix N - Unit 2 Cycle 18 Reload Safety Analysis (July, 2014) 16244 3"39 Appendix N - Unit 3 Cycle 17 Reload Analysis - Revision 2 (July, 2014) 726 340 Appendix N - Table of Contents 7 APPENDIX 0 - AGING MANAGEMENT PROGRAM 3"41 Appendix 0 - Introduction 100 342 Appendix 0 - Table of Contents 138 12

Retrieved from "https://kanterella.com/w/index.php?title=ML15295A025&oldid=2027776"