ML18024A355
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BFN-26 PLANT SAFETY ANALYSIS TABLE OF CONTENTS Page No 14.0 PLANT SAFETY ANALYSIS ..................................................................................................... 14.1-1 14.1 Analytical Objective ................................................................................................................... 14.1-1 14.2 Unacceptable Safety Results for Abnormal Operational Transients .......................................... 14.2-1 14.3 Unacceptable Safety Results For Accidents.............................................................................. 14.3-1 14.4 Approach to Safety Analysis...................................................................................................... 14.4-1 14.4.1 General ...................................................................................................................... 14.4-1 14.4.2 Abnormal Operational Transients .............................................................................. 14.4-2 14.4.3 Accidents ................................................................................................................... 14.4-4 14.4.4 Barrier Damage Evaluations ...................................................................................... 14.4-6 14.5 Analyses of Abnormal Operational Transients - Uprated........................................................... 14.5-1 14.5.1 Objective .................................................................................................................... 14.5-1 14.5.2 Events Resulting in a Nuclear System Pressure Increase ......................................... 14.5-9 14.5.3 Events Resulting in a Reactor Vessel Water Temperature Decrease ........................ 14.5-18 14.5.4 Events Resulting in a Positive Reactivity Insertion..................................................... 14.5-18 14.5.5 Events Resulting in a Reactor Vessel Coolant Inventory Decrease ........................... 14.5-26 14.5.6 Events Resulting in a Core Coolant Flow Decrease .................................................. 14.5-34 14.5.7 Events Resulting in a Core Coolant Flow Increase .................................................... 14.5-38 14.5.8 Event Resulting in Excess of Coolant Inventory......................................................... 14.5-42 14.5.9 Loss of Habitability of the Control Room .................................................................... 14.5-45 14.6 Analysis of Design Basis Accidents - Uprated ........................................................................... 14.6-1 14.6.1 Introduction ................................................................................................................ 14.6-1 14.6.2 Control Rod Drop Accident (CRDA) ........................................................................... 14.6-2 14.6.3 Loss of Coolant Accident (LOCA) .............................................................................. 14.6-10 14.6.4 Refueling Accident ..................................................................................................... 14.6-26 14.6.5 Main Steam Line Break Accident ............................................................................... 14.6-32 14.7 Conclusions ............................................................................................................................... 14.7-1 14.8 Analytical Methods .................................................................................................................... 14.8-1 14.8.1 Nuclear Excursion Analysis........................................................................................ 14.8-1 14.8.2 Reactor Vessel Depressurization Analysis................................................................. 14.8-2 14.8.3 Reactor Core Heatup Analysis ................................................................................... 14.8-9 14.8.4 Containment Response Analysis ............................................................................... 14.8-13 14.8.5 Analytical Methods for Evaluating Radiological Effects .............................................. 14.8-16 14.0-i
BFN-26 PLANT SAFETY ANALYSIS TABLE OF CONTENTS (Cont'd)
Page No 14.9 Dose Sensitivity Evaluation Using Assumptions Of The AEC/DRL (Incorporated with TID 14844) .................................................................................................. 14.9-1 14.9.1 Loss-of-Coolant Accident (183 meter release height) ................................................ 14.9-1 14.9.2 Refueling Accident (183 meter release height) .......................................................... 14.9-2 14.9.3 Steam Line Break Accident (ground level release) .................................................... 14.9-2 14.9.4 Control Rod Drop Accident (ground level release) ..................................................... 14.9-3 14.9.5 Radiological Consequences....................................................................................... 14.9-3 14.9.6 Discussion of Assumptions ........................................................................................ 14.9-4 14.10 Analyses of Abnormal Operational Transients - Pre-uprated .................................................... 14.10-1 14.10.1 Events Resulting in a Nuclear System Pressure Increase ........................................ 14.10-1 14.10.2 Events Resulting in a Reactor Vessel Water Temperature Decrease ....................... 14.10-5 14.10.3 Events Resulting in a Positive Reactivity Insertion ................................................... 14.10-7 14.10.4 Events Resulting in a Reactor Vessel Coolant Inventory Decrease .......................... 14.10-9 14.10.5 Events Resulting in a Core Coolant Flow Decrease ................................................. 14.10-13 14.10.6 Events Resulting in a Core Coolant Flow Increase ................................................... 14.10-15 14.10.7 Event Resulting in Excess of Coolant Inventory ....................................................... 14.10-17 14.10.8 Loss of Habitability of the Control Room ................................................................... 14.10-18 14.11 Analysis of Design Basis Accidents - Pre-Uprated .................................................................... 14.11-1 14.11.1 Introduction ............................................................................................................... 14.11-1 14.11.2 Control Rod Drop Accident (CRDA) .......................................................................... 14.11-2 14.11.3 Loss of Coolant Accident (LOCA) ............................................................................. 14.11-15 14.11.4 Refueling Accident .................................................................................................... 14.11-28 14.11.5 Main Steam Line Break Accident .............................................................................. 14.11-35 14.0-ii
BFN-26 LIST OF TABLES Table Title 14.4-1 Summary of Abnormal Operational Transients 14.4-2 Results of Design Basis Accidents 14.5-1 Transient Analyses Power/Flow State Points 14.5-2 Transient Analyses Initial Conditions 14.6-1 Control Rod Drop Accident - Fission Product Release to Environment 14.6-2 (Deleted) 14.6-3 Summary of Power Uprate Input Parameters used for all Containment Analyses 14.6-4 Summary of Power Uprate Input Parameters used for DBA-LOCA Short Term Containment
Response
14.6-5 Summary of Power Uprate Input Parameters used for DBA-LOCA Long Term Containment
Response
14.6-6 DBA-LOCA Short Term Pressure and Temperature Response 14.6-7 Inventory in Primary Containment Available for Leakage 14.6-8 Values for X/Q for Accident Dose Calculations 14.6-9 (Deleted) 14.8-1 Characteristics of Nuclear Excursions - Water-Moderated Oxide Cores 14.8-2 Dose Computational Methods Wind Direction Persistence 14.8-3 Meteorology Applicable to Design Basis Accidents 14.8-4 Calculated Air Concentration for 183 Meter Release Height 14.8-5 Calculated Air Concentration for 183 Meter Release Height 14.8-6 Thyroid Dose Conversion Factors 14.9-1 Design Basis Accident Radiological Doses (REM) 14.9-2 Sensitivity of Doses to Variation of Assumptions 14.11-1 Control Rod Drop Accident - Fission Product Release Rate to Environs 14.11-2 Control Rod Drop Accident - Radiological Effects 14.11-3 Loss-of-Coolant Accident - Primary Containment Response Summary 14.11-4 Inventory in Primary Containment Available for Leakage 14.11-5 (Deleted) 14.11-6 (Deleted) 14.11-7 (Deleted) 14.11-8 Values for X/Q for Accident Dose Calculations 14.11-9 (Deleted) 14.0-iii
BFN-26 LIST OF TABLES Table Title 14.11-10 (Deleted) 14.11-11 Steam Line Break Accident - Radiological Effects 14.0-iv
BFN-26 PLANT SAFETY ANALYSIS LIST OF ILLUSTRATIONS Figure Title 14.4-1 Plant Safety Analysis - Method for Identifying and Evaluating Abnormal Operational Transients 14.4-2 Plant Safety Analysis - Method for Identifying and Evaluating Accidents 14.5-1 (Deleted) 14.5-2 (Deleted) 14.5-3 (Deleted) 14.5-4 (Deleted) 14.5-5 Generator Trip (TCV Fast Closure) With Bypass Valve Failure 100P/105F 14.5-6 Load Rejection No Bypass with EOC-RPT-OOS 100P/105F 14.5-7a Loss of Condenser Vacuum 102P/105F 14.5-7b Loss of Condenser Vacuum 102P/105F 14.5-8 Turbine Stop Valve Closure/Turbine Trip 102P/105F 14.5-9 Bypass Failure Following Turbine Trip, High Power 100P/105F 14.5-10a Bypass Failure Following Turbine Trip, Low Power 30P/50F 14.5-10b Bypass Failure Following Turbine Trip, Low Power 30P/50F 14.5-11 Closure of All Main Steam Line Isolation Valves 102P/105F 14.5-12a Closure of One Main Steam Line Isolation Valve 102P/105F 14.5-12b Closure of One Main Steam Line Isolation Valve 102P/105F 14.5-13a Loss of Feedwater Heater 102P/81F 14.5-13b Loss of Feedwater Heater 102P/81F 14.5-14a Inadvertent Pump Start 102P/81F 14.5-14b Inadvertent Pump Start 102P/81F 14.5-15a Pressure Regulator Failure Open 102P/100F 14.5-15b Pressure Regulator Failure Open 102P/100F 14.5-16a Inadvertent Opening of a Relief Valve 102P/100F 14.5-16b Inadvertent Opening of a Relief Valve 102P/100F 14.5-17a Loss of Feedwater Flow,Short Term 102P/100F 14.5-17b Loss of Feedwater Flow, Short Term 102P/100F 14.5-17c Loss of Feedwater Flow, Long Term 102P/100F 14.5-18a Loss of Auxiliary Power Transformers 102P/100F 14.5-18b Loss of Auxiliary Power Transformers 102P/100F 14.5-19a Loss of Auxiliary Power - All Grid Connections 102P/105F 14.5-19b Loss of Auxiliary Power - All Grid Connections 102P/105F 14.5-20a Recirculation Flow Control Failure-Decreasing Flow 102P/100F 14.0-v
BFN-26 PLANT SAFETY ANALYSIS LIST OF ILLUSTRATIONS Figure Title 14.5-20b Recirculation Flow Control Failure-Decreasing Flow 102P/100F 14.5-21a One Recirculation Pump Trip 102P/100F 14.5-21b One Recirculation Pump Trip 102P/100F 14.5-22a (Deleted) 14.5-22b (Deleted) 14.5-22c Two Recirculation Pump Trip with VFDs 100P/100F 14.5-22d Two Recirculation Pump Trip with VFDs 100P/100F 14.5-22e Two Recirculation Pump Trip with VFDs 100P/100F 14.5-22f Two Recirculation Pump Trip with VFDs 100P/100F 14.5-23a One Recirculation Pump Seizure 102P/100F 14.5-23b One Recirculation Pump Seizure 102P/100F 14.5-24a (Deleted) 14.5-24b (Deleted) 14.5-24c Recirculation Flow Control Failure-Increasing Flow 75P/52F - VFD Speed Control 14.5-24d Recirculation Flow Control Failure-Increasing Flow 75P/52F - VFD Speed Control 14.5-24e Recirculation Flow Control Failure-Increasing Flow 75P/52F - VFD Speed Control 14.5-24f Recirculation Flow Control Failure-Increasing Flow 75P/52F - VFD Speed Control 14.5-25a (Deleted) 14.5-25b (Deleted) 14.5-25c Idle Recirculation Loop Startup 75P/52F - VFD Speed Control 14.5-25d Idle Recirculation Loop Startup 75P/52F - VFD Speed Control 14.5-25e Idle Recirculation Loop Startup 75P/52F - VFD Speed Control 14.5-25f Idle Recirculation Loop Startup 75P/52F - VFD Speed Control 14.5-26a Idle Recirculation Loop Startup 75P/52F Coupler Position 11%
14.5-26b Idle Recirculation Loop Startup 75P/52F Coupler Position 11%
14.5-27a Idle Recirculation Loop Startup 30P/52F Coupler Position 19%
14.5-27b Idle Recirculation Loop Startup 30P/52F Coupler Position 19%
14.5-28 Feedwater Control Failure-Maximum Demand 75P/108F 14.5-29 Feedwater Control Failure-Maximum Demand 100P/105F 14.5-30 Feedwater Control Failure-Maximum Demand. EOC-RPT-OOS 100P/105F 14.5-31 Feedwater Control Failure-Maximum Demand. TBP-OOS 100P/105F 14.6-1 DBA-LOCA Short Term Containment Temperature Response (102% of Uprated Power, 81% CF) 14.0-vi
BFN-26 PLANT SAFETY ANALYSIS LIST OF ILLUSTRATIONS Figure Title 14.6-2 DBA-LOCA Short Term Containment Pressure Response (102% of Uprated Power, 81% CF) 14.6-3 DBA-LOCA Long Term Wetwell Temperature Response (102% of Uprate Power, 100% CF) 14.6-4 DBA-LOCA Long Term Drywell Temperature Response (102% of Uprated Power, 100% CF) 14.6-5 DBA-LOCA Long Term Pressure Response (102% of Uprated Power, 100% CF) 14.6-6 Loss-of-Coolant Accident, Primary Containment Capability for Metal-Water Reaction 14.6-7 Main Steamline Break Accident, Break Location 14.6-8 Main Steamline Break Accident, Mass of Coolant Lost Through Break 14.6-9 Main Steamline Break Accident, Normalized Core Inlet Flow 14.6-10 Main Steamline Break Accident - Minimum Critical Heat Flux Ratio 14.6-11 (Deleted) 16.6-12 (Deleted) 14.6-13 (Deleted) 16.6-14 (Deleted) 14.6-15 (Deleted) 16.6-16 (Deleted) 16.6-17 (Deleted) 14.6-18 (Deleted) 14.8-1 Fuel Rod and Fuel Bundle Details 14.10-1 Transient Results, Turbine Trip from High Power without Bypass and Loss of Condenser Vacuum 14.10-2 Transient Results, Turbine Trip 14.10-3 Transient Results, Turbine Trip from Low Power without Bypass 14.10-4 Transient Results, Closure of all Main Steam Isolation Valves 14.10-5 Transient Results, Closure of One Main Steam Isolation Valve 14.10-6 Transient Results, Loss of Feedwater Heater 14.10-7 Transient Results, Continuous Rod Withdrawal During Power Range Operation 14.10-8 Transient Results, Pressure Regulator Failure 14.10-9 Transient Results, Inadvertent Opening of a Relief Valve or Safety Valve 14.10-10 Transient Results, Loss of Feedwater Flow 14.10-10a Water Level Response with RCIC for Loss of Feedwater Flow Event 14.10-10b Pressure Response with RCIC for Loss of Feedwater Flow Event 14.10-11 Transient Results, Loss of Auxiliary Power 14.10-12 Transient Results, Loss of Auxiliary Power - All Grid Connections 14.0-vii
BFN-26 PLANT SAFETY ANALYSIS LIST OF ILLUSTRATIONS Figure Title 14.10-13 Water Level vs. Time Following Loss of Auxiliary Power (RCIC Only) 14.10-14 Transient Results, Trip of One Recirculation System Generator Field Breaker 14.10-15 Transient Results, Trip of Two Recirculation Pump M-G Set Drive Motors 14.10-16 Transient Results, Recirculation Pump Seizure 14.10-17 Transient Results, Recirculation Flow Control Failure Increasing Flow 14.10-18 Transient Results, Startup of Idle Recirculation Pump 14.10-19 Transient Results, Feedwater Controller Failure Maximum Demand 14.10-20 Generator Load Rejection Without Bypass, EOC2,RPT 14.11-1 Maximum Rod Worth Versus Moderator Density 14.11-2 Maximum Rod Worth Versus Power Level 14.11-3 Rod Drop Accident (Cold, Critical) Peak Fuel Enthalpy 14.11-4 Rod Drop Accident (Hot, Critical) Peak Fuel Enthalpy 14.11-5 Rod Drop Accident (Power Range) Peak Fuel Enthalpy 14.11-6 (Deleted) 14.11-7 (Deleted) 14.11-8 (Deleted) 14.11-9 (Deleted) 14.11-10 Loss-of-Coolant Accident, Primary Containment Pressure Response 14.11-11 Loss-of-Coolant Accident, Drywell Temperature Response 14.11-12 Loss-of-Coolant Accident, Pressure Suppression Pool Temperature Response 14.11-13 (Deleted) 14.11-14 Loss-of-Coolant Accident, Primary Containment Capability for Metal-Water Reaction 14.11-15 Main Steamline Break Accident, Break Location 14.11-16 Main Steamline Break Accident, Mass of Coolant Lost Through Break 14.11-17 Main Steamline Break Accident, Normalized Core Inlet Flow 14.11-18 Main Steamline Break Accident - Minimum Critical Heat Flux Ratio 14.0-viii