ML18024A529
| ML18024A529 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/05/2017 |
| From: | Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18018A778 | List:
|
| References | |
| Download: ML18024A529 (8) | |
Text
BFN-27
1.0 INTRODUCTION
AND
SUMMARY
Project Identification 1.1 Definitions 1.2 Methods of Technical Presentation 1.3 Classification of BWR Systems, Criteria, 1.4 and Requirements for Safety Evaluation Principal Design Criteria 1.5 Plant Description 1.6 Comparison of Principal Design Characteristics 1.7 Summary of Radiation Effects 1.8 Plant Management 1.9 Quality Assurance Program 1.10 Identification-Resolution of Construction Permit 1.11 Concern-Summary General Conclusions 1.12 2.0 SITE Summary Description 2.1 Site Description 2.2 Meteorology 2.3 Hydrology, Water Quality, and Aquatic Biology 2.4 Geology and Seismology 2.5 Environmental Radiological Monitoring Program 2.6 3.0 REACTOR Summary Description 3.1 Fuel Mechanical Design 3.2 Reactor Vessel Internals Mechanical Design 3.3 Reactivity Control Mechanical Design 3.4 Control Rod Drive Housing Supports 3.5 Nuclear Design 3.6 Thermal and Hydraulic Design 3.7 Standby Liquid Control System 3.8 4.0 REACTOR COOLANT SYSTEM Summary Description 4.1 Reactor Vessel and Appurtenances Mechanical Design 4.2 Reactor Recirculation System 4.3 Nuclear System Pressure Relief System 4.4 Main Steam Line Flow Restrictor 4.5 Main Steam Isolation Valves 4.6 1
BFN-27 Reactor Core Isolation Cooling System 4.7 Residual Heat Removal System (RHRS) 4.8 Reactor Water Cleanup System 4.9 Nuclear System Leakage Rate Limits 4.10 Main Steam Lines, Feedwater Piping, and Drains 4.11 Inservice Inspection and Testing 4.12 5.0 CONTAINMENT Summary Description 5.1 Primary Containment System 5.2 Secondary Containment System 5.3 6.0 EMERGENCY CORE COOLING SYSTEMS Safety Objective 6.1 Safety Design Basis 6.2 Summary Description-Emergency Core Cooling Systems 6.3 Description 6.4 Safety Evaluation 6.5 Inspection and Testing 6.6 7.0 CONTROL AND INSTRUMENTATION Summary Description 7.1 Reactor Protection System 7.2 Primary Containment Isolation System 7.3 Emergency Core Cooling Control and Instrumentation 7.4 Neutron Monitoring System 7.5 Refueling Interlocks 7.6 Reactor Manual Control System 7.7 Reactor Vessel Instrumentation 7.8 Recirculation Flow Control System 7.9 Feedwater Control System 7.10 Pressure Regulator and Turbine-Generator Control 7.11 Process Radiation Monitoring 7.12 Area Radiation Monitoring System 7.13 Drywell Leak Detection Radiation Monitoring System 7.14 Health Physics Laboratory Radiation Monitoring Equipment 7.15 Process Computer System 7.16 Deleted 7.17 Backup Control System 7.18 Anticipated Transient Without Scram 7.19 Instrument Setpoint Methodology 7.20 2
BFN-27 8.0 ELECTRICAL POWER SYSTEMS Summary Description 8.1 Generators 8.2 Transmission System 8.3 Normal Auxiliary Power System 8.4 Standby AC Power Supply and Distribution 8.5 250-V DC Power Supply and Distribution 8.6 120-V AC Power Supply and Distribution 8.7 Auxiliary DC Power Supply and Distribution 8.8 Safety Systems Independence Criteria and Bases for Electrical Cable Installation 8.9 Station Blackout 8.10 9.0 RADIOACTIVE WASTE CONTROL SYSTEMS Summary Description 9.1 Liquid Radwaste System 9.2 Solid Radwaste System 9.3 Deleted 9.4 Gaseous Radwaste System (Modified) 9.5 10.0 AUXILIARY SYSTEMS Summary Description 10.1 New Fuel Storage 10.2 Spent Fuel Storage 10.3 Tools and Servicing Equipment 10.4 Fuel Pool Cooling and Cleanup System 10.5 Reactor Building Closed Cooling Water System 10.6 Raw Cooling Water System 10.7 Raw Service Water System 10.8 RHR Service Water System 10.9 Emergency Equipment Cooling Water System 10.10 Fire Protection Systems 10.11 Heating, Ventilating and Air-Conditioning Systems 10.12 Demineralized Water System 10.13 Control and Service Air Systems 10.14 Potable Water and Sanitary Systems 10.15 Equipment and Floor Drainage Systems 10.16 Process Sampling Systems 10.17 Plant Communications System 10.18 Lighting System 10.19 Auxiliary Boiler System 10.20 Postaccident Sampling System 10.21 Auxiliary Decay Heat Removal System 10.22 Hydrogen Water Chemistry System (HWC) 10.23 3
BFN-27 11.0 POWER CONVERSION SYSTEMS Summary Description 11.1 Turbine-Generator 11.2 Main Condenser System 11.3 Main Condenser Gas Removal and Turbine Sealing Systems 11.4 Turbine Bypass System 11.5 Condenser Circulating Water System 11.6 Condensate Filter-Demineralizer System 11.7 Condensate and Reactor Feedwater Systems 11.8 Condensate Storage and Transfer Systems 11.9 12.0 STRUCTURES AND SHIELDING Summary Description 12.1 Principal Structures and Foundations 12.2 Shielding and Radiation Protection 12.3 13.0 CONDUCT OF OPERATIONS Organizational Structure for the Conduct of Operations 13.1 Organization and Responsibility 13.2 Training Programs 13.3 Preoperational Test Program 13.4 Startup and Power Test Program 13.5 Normal Operations 13.6 Deleted 13.7 Operational Review and Audits 13.8 Refueling Operations 13.9 Refueling Test Program 13.10 14.0 PLANT SAFETY ANALYSIS Analytical Objective 14.1 Unacceptable Safety Results for Abnormal Operational Transients 14.2 Unacceptable Safety Results for Accidents 14.3 Approach to Safety Analysis 14.4 Analyses of Abnormal Operational Transients - Uprated 14.5 Analysis of Design Basis Accidents - Uprated 14.6 Conclusions 14.7 Analytical Methods 14.8 Dose Sensitivity Evaluation Using Assumptions of the AEC/DRL (Incorporated with TID 14844) 14.9 4
BFN-27 Analyses of Abnormal Operational Transients - Pre-uprated 14.10 Analysis of Design Basis Accidents - Pre-Uprated 14.11 APPENDIX A. CONFORMANCE TO AEC PROPOSED GENERAL DESIGN CRITERIA Summary Description A.1 Criterion Conformance A.2 APPENDIX B. TECHNICAL SPECIFICATIONS AND TECHNICAL REQUIREMENTS MANUAL APPENDIX C. STRUCTURAL QUALIFICATION OF SUBSYSTEMS AND COMPONENTS Scope C.1 Loading Conditions, Definitions, and Overview C.2 Piping and Pipe Supports C.3 Major Components C.4 Primary Containment System and Penetrations C.5 Equipment C.6 Heating, Ventilation, and Air Conditioning (HVAC) Ductwork and Supports C.7 Control of Heavy Loads C.8 References C.9 APPENDIX D. QUALITY ASSURANCE PLAN FOR THE BROWNS FERRY NUCLEAR PLANT Quality Assurance During Design and Construction D.1 General Electric Quality System for BWR Nuclear Steam Supply Projects D.2 Quality Assurance Program for Station Operation D.3 APPENDIX E. SITE GASEOUS RELEASE RATE LIMIT CALCULATION Site Gaseous Release Rate Limits E.1 5
BFN-27 APPENDIX F. UNIT SHARING AND INTERACTIONS Introduction F.1 Scope F.2 References F.3 Criteria F.4 List of Shared Features F.5 Description of Shared Conventional Systems F.6 Description of Shared Class I Seismic Features, Structures, Safeguards Systems and Supporting Auxiliary Systems F.7 APPENDIX G. PLANT NUCLEAR SAFETY OPERATIONAL ANALYSIS Analytical Objective G.1 Basis for Selecting Operational Requirements G.2 Basis for Selection of Surveillance Test Frequencies for Nuclear Safety Systems and Engineered Safeguards G.3 Method of Analysis G.4 Analysis and Results G.5 Conclusion G.6 APPENDIX H. CORE THERMAL DESIGN Deleted APPENDIX I. IDENTIFICATION-RESOLUTION OF CONSTRUCTION PERMIT CONCERNS Summary Description I.1 Areas Specified in the Browns Ferry AEC-ACRS Construction Permit Reports I.2 Areas Specified in the AEC-Staff Construction Permit-Safety Evaluation Reports I.3 Areas Specified in Other Related AEC-ACRS Construction Permit and Operating License Reports I.4 Areas Specified in Other Related AEC-Staff Construction Permit or Operating License Safety Evaluation Reports I.5 Summary Conclusions I.6 References I.7 6
BFN-27 APPENDIX J. REACTOR PRESSURE VESSEL DESIGN
SUMMARY
REPORT - UNIT 1 Introduction J.1 Design and Fabrication Requirements -
Summary J.2 APPENDIX K. REACTOR PRESSURE VESSEL DESIGN
SUMMARY
REPORT - UNIT 2 Introduction K.1 Design and Fabrication Requirements -
Summary K.2 APPENDIX L. REACTOR PRESSURE VESSEL DESIGN
SUMMARY
REPORT - UNIT 3 Introduction L.1 Design and Fabrication Requirements -
Summary L.2 APPENDIX M. REPORT ON PIPE FAILURES OUTSIDE CONTAINMENT IN THE BROWNS FERRY NUCLEAR PLANT Introduction M.1 Pressure and Analyses M.2 Piping Design Philosophy M.3 Breaks Postulated and Loading Effects Considered M.4 Pipe Rupture Loads M.5 Pipe Break Assumptions, Analysis, and Break Locations M.6 Structural Analysis M.7 Effects on Safety-Related Components and Structures M.8 Additional Work M.9 Summary and Conclusions M.10 References M.11 APPENDIX N. RELOAD LICENSING REPORT Browns Ferry, Unit 1 Cycle 10 Reload Safety Report N.1-1 Browns Ferry Unit 2 Cycle 18 Reload Safety Analysis N.2-1 Browns Ferry Unit 3 Cycle 16 Reload Safety Analysis N.3-1 7
BFN-27 APPENDIX O. AGING MANAGEMENT PROGRAMS Aging Management Programs O.1 Plant Specific Aging Management Programs O.2 Time-Limited Aging Management Programs O.3 10 CFR 54.37(b) New Identified Systems, Structures, and Components O.4 References O.5 8