ML23110A193
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PBAPS UFSAR SECTION 14 - PLANT SAFETY ANALYSIS TABLE OF CONTENTS SECTION TITLE 14.0 PLANT SAFETY ANALYSIS 14.0.1 Analytical Objective 14.0.2 Frequency Classification 14.1 UNACCEPTABLE SAFETY RESULTS FOR INCIDENTS OF MODERATE FREQUENCY - ANTICIPATED (EXPECTED)
OPERATIONAL TRANSIENTS 14.2 UNACCEPTABLE SAFETY RESULTS FOR INFREQUENT INCIDENTS - ABNORMAL (UNEXPECTED) OPERATIONAL TRANSIENTS 14.3 UNACCEPTABLE SAFETY RESULTS FOR LIMITING FAULTS -
DESIGN BASIS ACCIDENTS 14.4 APPROACH TO SAFETY ANALYSIS 14.4.1 General 14.4.2 Analytical Categories 14.4.3 Accidents 14.4.4 Barrier Damage Evaluations 14.4.4.1 Fuel Damage 14.4.4.2 Nuclear System Process Barrier Damage 14.4.4.3 Containment Damage 14.4.5 Licensing Basis Versus Emergency Procedure Guidelines 14.5 ANALYSES OF ABNORMAL OPERATIONAL TRANSIENTS 14.5.1 Events Resulting in a Nuclear System Pressure Increase 14.5.1.1 Electrical Load Rejection (Turbine Control Valve Fast Closure) with Bypass Failure 14.5.1.2 Turbine Trip (Turbine Stop Valve Closure) 14.5.1.2.1 Turbine Trip from High Power with Bypass 14.5.1.2.2 Turbine Trip from High Power without Bypass 14.5.1.2.3 Turbine Trip from Lower Power without Bypass 14.5.1.3 Main Steam Line Isolation Valve Closure 14.5.1.3.1 Closure of All Main Steam Line Isolation Valves 14.5.1.3.2 Closure of One Main Steam Line Isolation Valve 14.5.2 Events Resulting in a Reactor Vessel Water Temperature Decrease 14.5.2.1 Inadvertent Pump Start (HPCIS) 14.5.2.2 Feedwater Controller Failure - Maximum Demand 14.5.2.3 Loss of Feedwater Heating CHAPTER 14 14-i REV. 28, APRIL 2021
PBAPS UFSAR TABLE OF CONTENTS (cont'd)
SECTION TITLE 14.5.2.4 Shutdown Cooling (RHRS) Malfunction - Decreasing Temperature 14.5.3 Events Resulting in a Positive Reactivity Insertion 14.5.3.1 Continuous Rod Withdrawal During Power Range Operation 14.5.3.2 Continuous Rod Withdrawal During Reactor Startup 14.5.3.3 Control Rod Removal Error During Refueling 14.5.3.4 Fuel Assembly Insertion Error During Refueling 14.5.3.5 Misoriented Bundle Event 14.5.4 Events Resulting in a Reactor Vessel Coolant Inventory Decrease 14.5.4.1 Pressure Regulator Failure 14.5.4.2 Inadvertent Opening of a Relief Valve or Safety Valve 14.5.4.3 Loss of Feedwater Flow 14.5.4.4 Loss of Auxiliary Power 14.5.5 Events Resulting in a Core Coolant Flow Decrease 14.5.5.1 Recirculation Flow Control Failure - Decreasing Flow 14.5.5.2 Trip of One Recirculation Pump 14.5.5.3 Trip of Two Adjustable Speed Drives (ASDs) 14.5.5.4 Recirculation Pump Seizure 14.5.6 Events Resulting in a Core Coolant Flow Increase 14.5.6.1 Recirculation Manual Control Station Failure -
Increasing Flow 14.5.6.2 Startup of Idle Recirculation Pump 14.6 ANALYSIS OF DESIGN BASIS ACCIDENTS 14.6.1 Introduction 14.6.2 Control Rod Drop Accident 14.6.2.1 Initial Conditions 14.6.2.2 Excursion Analysis Assumptions 14.6.2.3 Fuel Damage 14.6.2.4 Fission Product Release From Fuel 14.6.2.5 Fission Product Transport 14.6.2.6 Fission Product Release to Environs 14.6.2.7 Radiological Effects 14.6.2.8 Elimination of Main Steamline Scram and Primary Containment High Radiation 14.6.3 Loss of Coolant Accident 14.6.3.1 Initial Conditions and Assumptions 14.6.3.2 Nuclear System Depressurization and Core Heatup 14.6.3.3 Primary Containment Response 14.6.3.3.1 Initial Conditions and Assumptions CHAPTER 14 14-ii REV. 28, APRIL 2021
PBAPS UFSAR TABLE OF CONTENTS (cont'd)
SECTION TITLE 14.6.3.3.2 Containment Response 14.6.3.3.3 Metal-Water Reaction Effects on the Primary Containment 14.6.3.4 Fission Products Released to Primary Containment 14.6.3.5 Fission Product Release to Secondary Containment 14.6.3.6 Fission Product Release to Environs 14.6.3.7 Radiological Effects 14.6.4 Refueling Accident 14.6.4.1 Accident Scenario Assumptions 14.6.4.2 Fuel Damage 14.6.4.3 Fission Product Release From Fuel 14.6.4.4 Fission Product Release to Secondary Containment 14.6.4.5 Fission Product Release to Environs 14.6.4.6 Radiological Effects 14.6.5 Main Steam Line Break Accident 14.6.5.1 Nuclear System Transient Effects 14.6.5.1.1 Assumptions 14.6.5.1.2 Sequence Events 14.6.5.1.3 Coolant Loss and Reactor Vessel Water Level 14.6.5.2 Radioactive Material Release 14.6.5.2.1 Assumptions 14.6.5.2.2 Fission Product Release from Break 14.6.5.2.3 Steam Cloud Movement 14.6.5.3 Radiological Effects
14.7 CONCLUSION
S 14.8 ANALYTICAL METHODS 14.8.1 Nuclear Excursion Analysis 14.8.1.1 Introduction 14.8.1.2 Description 14.8.2 Reactor Vessel Depressurization Analysis 14.8.2.1 Introduction 14.8.2.2 Theoretical Development 14.8.2.2.1 Mass Balance 14.8.2.2.2 Mass Rate of Change in Vessel 14.8.2.2.3 Rate of Change of Energy in Vessel 14.8.2.2.4 Flashing Rate in Vessel 14.8.2.2.5 Vessel Depressurization Rate 14.8.2.2.6 Mass Flow Rates 14.8.2.3 Numerical Solution 14.8.3 Reactor Core Heatup Analysis 14.8.3.1 Introduction CHAPTER 14 14-iii REV. 28, APRIL 2021
PBAPS UFSAR TABLE OF CONTENTS (cont'd)
SECTION TITLE 14.8.3.2 Theoretical Development 14.8.3.2.1 Heat Sources 14.8.3.2.2 Conduction Heat Transfer 14.8.3.2.3 Convection Heat Transfer 14.8.3.2.4 Radiation 14.8.3.3 Method of Solution 14.8.4 Containment Response Analysis 14.8.4.1 Short-Term Containment Response 14.8.4.1.1 Introduction 14.8.4.1.2 Theoretical Development 14.8.4.1.3 Solution 14.8.4.2 Long-Term Containment Pressure Response 14.8.5 Analytical Methods for Calculating Radiological Effects 14.8.5.1 Introduction 14.8.5.2 Meteorological Diffusion Evaluation Methods 14.8.5.2.1 General 14.8.5.2.2 Height of Release 14.8.5.2.3 Diffusion Conditions 14.8.5.2.4 Applied Meteorology 14.8.5.2.5 Cloud Dispersion Calculations 14.8.5.2.6 Cloud Depletion and Group Deposition 14.8.5.2.7 Air Concentration Calculation 14.8.5.3 Radiological Effects Calculation 14.8.5.3.1 Passing Cloud Dose 14.8.5.3.2 Inhalation Dose 14.9 EVALUATIONS USING AEC METHOD 14.9.1 Evaluation of Plant Systems Using TID-14844 Source Terms 14.9.1.1 Source Term Assumptions 14.9.1.2 Standby Gas Treatment System 14.9.1.3 Core Standby Cooling System Components 14.9.1.4 Electrical Penetrations 14.9.1.5 Control Room 14.9.1.6 Conclusion 14.9.2 Current Licensing Basis Evaluations Using the Alternative Source Term (RG 1.183) 14.9.2.1 Loss-of-Coolant Accident (LOCA) 14.9.2.2 Refueling Accident 14.9.2.3 Main Steam Line Break Accident (MSLB) 14.9.2.4 Control Rod Drop Accident (CRDA) 14.9.2.5 Conclusion CHAPTER 14 14-iv REV. 28, APRIL 2021
PBAPS UFSAR TABLE OF CONTENTS (cont'd)
SECTION TITLE 14.10 ANALYSIS OF CONTAINMENT RESPONSE 14.10.1 Methodology 14.10.2 Short-Term Containment Pressure Response 14.10.3 Drywell and Suppression Pool Temperature 14.10.4 Long-Term Bulk Suppression Pool Temperature Response - Design Basis Accidents 14.10.4.1 Suppression Pool Temperature Response - DBLOCA 14.10.4.2 Suppression Pool Temperature Response - Small Steam Break LOCA 14.10.4.3 Suppression Pool Temperature Response - Non-Accident Unit 14.10.4.4 Suppression Pool Temperature Response - Loss of RHR Normal Shutdown Cooling Function Event 14.10.5 Long-Term Bulk Suppression Pool Temperature Response - Special Events 14.10.5.1 Station Blackout 14.10.5.2 Appendix R Fire Safe Shutdown 14.10.5.3 Anticipated Transient Without Scram (ATWS)
CHAPTER 14 14-v REV. 28, APRIL 2021
PBAPS UFSAR SECTION 14.0 - PLANT SAFETY ANALYSIS LIST OF TABLES TABLE TITLE 14.4.1 Plant Safety Analysis Results of Operational Transients 14.4.2 Plant Safety Analysis Results of Design Basis Accidents 14.5.1 Deleted 14.5.2 Continuous Rod Withdrawal Sequence of Events 14.6.1 Control Rod Drop Accident, Fission Product Release Rate to Environs 14.6.2 Control Rod Drop Accident 14.6.3 Primary Containment Response Summary 14.6.4 Loss-of-Coolant Accident, Primary Containment Airborne Fission Product Inventory 14.6.5 Loss-of-Coolant Accident, Secondary Containment Airborne Fission Product Inventory 14.6.6 Loss-of-Coolant Accident, Fission Product Release Rate to Environs 14.6.7 Loss-of-Coolant Accident 14.6.8 Refueling Accident, Secondary Containment Airborne Fission Product Inventory 14.6.9 Refueling Accident, Fission Product Release Rate to Environs 14.6.10 Refueling Accident 14.6.11 Steam Line Break Accident 14.8.1 Calculated Air Concentration for 0-Meter Release Height CHAPTER 14 14-vi REV. 28, APRIL 2021
PBAPS UFSAR LIST OF TABLES (cont'd)
TABLE TITLE 14.8.2 Calculated Air Concentration for 30-Meter Release Height 14.8.3 Calculated Air Concentration for 152-Meter Release Height 14.9.1 Activity, Mass Loading, and Heat Loading at Various Locations for TID Release Assumptions 14.9.2 Standby Gas Treatment System Performance 14.9.3 Doses for Various Equipment or Locations Based on TID-14844 Fission Product Release Assumptions 14.9.4 Gamma Ray Energy Spectrum of Fission Products in the Secondary Containment 14.9.5 Biological Dose Rate at the Center of the Control Room Floor Following a Loss-of-Coolant Accident 14.9.6 Integrated Dose in the Control Room 14.9.7 Design Basis Accident Radiological Doses 14.9.8 Sensitivity of Doses to Variation of Assumptions, Loss-of-Coolant Accident 14.9.9 AST Radionuclide and Magnitude 14.9.10 Parameters And Assumptions Used In Post-LOCA Radiological Consequence Analysis 14.9.11 Parameters And Assumptions Used In Fuel Handling Accident Radiological Consequence Analysis 14.9.12 Parameters And Assumptions Used In Control Rod Drop Accident Radiological Consequence Analysis 14.10.1 DBA Containment Response Key Analysis Input Values 14.10.2 Containment Response Results Dual-Unit Interaction 14.10.3 Special Event Containment Response Key Analysis Input Values CHAPTER 14 14-vii REV. 28, APRIL 2021
PBAPS UFSAR SECTION 14.0 - PLANT SAFETY ANALYSIS LIST OF FIGURES 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 14.5.1A Transient Results - Electrical Load Rejection Without Bypass (Unit 3) 14.5.1AA Peach Bottom Response to LRNBP (Unit 2) 14.5.1B Deleted 14.5.1BB Peach Bottom Response to TTNBP (Unit 2) 14.5.2 Deleted 14.5.3 Transient Results - Closure of All Main Steam Isolation Valves (Unit 3) 14.5.3A Deleted 14.5.3B Deleted 14.5.4 Deleted 14.5.4A Deleted 14.5.5 Transient Results - Feedwater Controller Failure, Maximum Demand 14.5.6 Deleted 14.5.7A Deleted 14.5.7B Deleted CHAPTER 14 14-viii REV. 28, APRIL 2021
PBAPS UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE 14.5.8 Transient Results - Pressure Regulator Failure -
Open 14.5.9 Transient Results - Inadvertent Opening of a Relief Valve or Safety Valve (Unit 3) 14.5.10 Transient Results - Loss of FW Flow (Unit 2) 14.5.11A Deleted 14.5.11B Deleted 14.5.11C Deleted 14.5.12a Deleted 14.5.12b Deleted 14.5.13 Deleted 14.5.14 Transient Results - Single Recirculation Flow Controller Failure - Increasing Flow 14.5.15 Transient Results - Startup of Idle Recirculation Pump 14.5.16 Deleted 14.5.17 Deleted 14.5.18 Deleted 14.5.19 Deleted 14.6.1 Maximum Rod Worth Versus Moderator Density 14.6.2 Control Rod Worth As a Function of Core Power 14.6.3 Rod Drop Accident (Cold, Critical) Peak Fuel Enthalpy 14.6.4 Rod Drop Accident (Hot, Critical) Peak Fuel Enthalpy CHAPTER 14 14-ix REV. 28, APRIL 2021
PBAPS UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE 14.6.5 Rod Drop Accident (Power Range) Peak Fuel Enthalpy 14.6.6 LOCA - Humboldt Primary Containment Pressure
Response
14.6.7 LOCA - Bodega Bay Primary Containment Pressure
Response
14.6.8 LOCA - Bodega Bay Primary Containment Pressure
Response
14.6.9 LOCA - Comparison of Calculated and Measured Peak Drywell Pressure for Bodega Bay and Humboldt Tests 14.6.10 LOCA - Primary Containment Pressure Response 14.6.10A Pressure Response as a Function of Time at 3696 MWt and 100% Core Flow 14.6.10B Long Term Containment Pressure Response - Normal ECCS Flow 14.6.10C Drywell Pressure Response - Small Steam Line Break (1.0 sqft) 14.6.11 LOCA - Drywell Temperature Response 14.6.11A Temperature Response as a Function of Time at 3696 MWt and 100% Core Flow 14.6.11B Long Term Drywell Airspace Temperature Response -
Normal ECCS Flow 14.6.11C Drywell Airspace Temperature Response - Small Steam Line Break (0.25 sqft) 14.6.12 LOCA - Suppression Pool Temperature Response 14.6.12A Long Term Suppression Pool Temperature Response -
Normal ECCS Flows 14.6.13 Primary Containment Leak Rate 14.6.14 Primary Containment Capability Index for Metal-Water Reaction CHAPTER 14 14-x REV. 28, APRIL 2021
PBAPS UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE 14.6.15 Main Steam Line Break Accident - Break Location 14.6.16 Main Steam Line Break Accident - Mass of Coolant Through Break 14.6.17 Main Steam Line Break Accident - Normalized Core Inlet Flow 14.6.18 Main Steam Line Break Accident - Minimum Critical Heat Flux Ratio 14.8.1 Fuel Rod and Fuel Bundle Details 14.10.1 Design Case Short-Term RSLB DBLOCA Containment Pressure Response - Design Case 14.10.2 Design Case Short-Term RSLB DBLOCA Containment Temperature Response - Design Case 14.10.3 Bonding Case Short-Term RSLB DBLOCA Containment Pressure Response - Bounding Case 14.10.4 Bounding Case Short-Term RSLB DBLOCA Containment Temperature Response - Bounding Case 14.10.5 Reference Case Short-Term RSLB DBLOCA Containment Pressure Response - Reference Case 14.10.6 Reference Case Short-Term RSLB DBLOCA Containment Temperature Response - Reference Case 14.10.7 Long-Term Small Steam Break LOCA Drywell Temperature Response 14.10.8 SP and WW Temperature Response to RSLB DBLOCA (CIC) 14.10.8A SP and WW Airspace Temperature Response to RSLB DBLOCA Dual-Unit Interaction (CIC) 14.10.9 DW and WW Airspace Temperature Response to DBLOCA (CIC) 14.10.9A DW and WW Airspace Temperature Response to DBLOCA Dual-Unit Interaction (CIC)
CHAPTER 14 14-xi REV. 28, APRIL 2021
PBAPS UFSAR LIST OF FIGURES (cont'd)
FIGURE TITLE 14.10.10 Long-Term Small Break LOCA Suppression Pool Temperature Response 14.10.10A Long-Term Small Break LOCA Suppression Pool Temperature Response Dual-Unit Interaction 14.10.11 SP Temperature Response of Non-Accident Unit During Safe Shutdown 14.10.12 SP Temperature Response to Loss of Normal RHR Shutdown Cooling Event (CIC) 14.10.12A SP Temperature Response to Loss of Normal RHR Shutdown Cooling Event Dual-Unit Interaction (CIC) 14.10.13 Deleted 14.10.14 Deleted CHAPTER 14 14-xii REV. 28, APRIL 2021