ML081500289
| ML081500289 | |
| Person / Time | |
|---|---|
| Site: | South Texas  |
| Issue date: | 04/30/2008 |
| From: | South Texas |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NOC-AE-08002283, STI: 32284244 | |
| Download: ML081500289 (19) | |
Text
STPEGS UFSAR TABLE OF CONTENTS
CHAPTER 15 ACCIDENT ANALYSES Section Title Page TC 15-1 Revision 13 15.0 ACCIDENT ANALYSIS 15.0-1 15.0.1 Classification of Plant Conditions 15.0-1 15.0.2 Optimization of Control Systems 15.0-5 15.0.3 Plant Characteristics and Initial Conditions Assumed in the Accident Analyses 15.0-5 15.0.4 Reactivity Coefficients Assumed in the Accident Analyses 15.0-8 15.0.5 Rod Cluster Control Assembly Insertion Characteristics 15.0-8 15.0.6 Trip Points and Time Delays to Trip Assumed in Accident Analyses 15.0-9 15.0.7 Instrumentation Drift and Calorimetric Errors 15.0-9 15.0.8 Plant Systems and Components Available for Mitigation of Accident Effects 15.0-10 15.0.9 Residual Decay Heat 15.0-12 15.0.10 Computer Codes Utilized 15.0-12 15.0.11 Summary of Accident Results 15.0-14 15.1 INCREASE IN HEAT REMO VAL BY THE SECONDARY SYSTEM 15.1-l 15.1.1 Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature 15.1-l 15.1.2 Feedwater System Malfunctions Causing an Increase in Feedwater Flow 15.1-3 15.1.3 Excessive Increase in Secondary Steam Flow 15.1-5 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve Causing a Depres-surization of the Main Steam System 15.1-8 15.1.5 Spectrum of Steam System Piping Failure Inside and Outside Containment 15.1-10 15.2 DECREASE IN HEAT REMO VAL BY THE SECONDARY SYSTEM 15.2-l 15.2.1 Steam Pressure Regulator Malfunction or Failure That Results in Decreasing Steam Flow 15.2-l 15.2.2 Loss of External Electrical Load 15.2-l 15.2.3 Turbine Trip 15.2-3 STPEGS UFSAR TABLE OF CONTENTS (Continued)
CHAPTER 15 Section Title Page TC 15-2 Revision 13 15.2.4 Inadvertent Closure of Main Steam Isolation Valves 15.2-8 15.2.5 Loss of Condenser Vacuum and Other Events Causing a Turbine Trip 15.2-8 15.2.6 Loss of Nonemergency AC Power to the Plant Auxiliaries (Loss of Offsite Power) 15.2-8 15.2.7 Loss of Normal Feedwater Flow 15.2-10 15.2.8 Feedwater System Pipe Break 15.2-13 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 15.3-l 15.3.l Partial Loss of Forced Reactor Coolant Flow 15.3-l 15.3.2 Complete Loss of Forced Reactor Coolant Flow 15.3-3 15.3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor) 15.3-6 15.3.4 Reactor Coolant Pump Shaft Break 15.3-10 15.4 REACTIVITY AND POWER DI STRIBUTION ANOMALIES 15.4-l 15.4.l Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from a Subcritical or Low Power Startup Condition 15.4-l 15.4.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power 15.4-5 15.4.3 Rod Cluster Control Assembly Misoperation 15.4-10 15.4.4 Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature 15.4-16 15.4.5 A Malfunction or Failure of the Flow Controller in a BWR Loop That Results in an Increased Reactor Coolant Flow Rate 15.4-18 15.4.6 Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant 15.4-18 15.4.7 Inadvertent Loading of a Fuel Assembly into an Improper Position 15.4-23 15.4.8 Spectrum of Rod Cluster Control Assembly Ejection Accidents 15.4-25 15.4.9 Spectrum of Rod Drop Accidents in a BWR 15.4-36 STPEGS UFSAR TABLE OF CONTENTS (Continued)
CHAPTER 15 Section Title Page TC 15-3 Revision 13 15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5-l 15.5.1 Inadvertent Operation of ECCS During Power Operation 15.5-l 15.5.2 Chemical and Volume Control Systems Malfunction That Increases Reactor Coolant Inventory 15.5-2 15.6 DECREASE IN REACTOR COOLANT INVENTORY 15.6-1 15.6.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve 15.6-1 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6-3 15.6.3 Steam Generator Tube Rupture 15.6-5 15.6.4 Radiological Consequences of Main Steam Line Failure Outside Containment (BWR) 15.6-11 15.6.5 Loss of Coolant Accidents 15.6-11 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7-1 15.7.1 Waste Gas System Failure 15.7-1 15.7.2 Postulated Radioactive Releases due to Liquid-Containing Tank Failure (Release to Atmosphere) 15.7-1 15.7.3 Postulated Radioactive Releases Due to Liquid-Containing Tank Failures (Ground Release) 15.7-1 15.7.4 Design Basis Fuel Handling Accidents 15.7-2 15.7.5 Spent Fuel Cask Drop Accident 15.7-6 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM 15.8-1 APPENDIX FISSION PRODUCT INVENTORIES 15.A-1 15.A 15.A.l Activities in the Core 15.A-1 15.A.2 Activities in the Fuel Pellet-Cladding Gap 15.A-1 15.A.3 Concentrations in the Coolants 15.A-1 15.A.4 The Impact of Extended Burnup Fuel on Source Terms 15.A-2 15.A.5 Applicability of V5H Fuel Upgrade 15.A-2 15.A.6 The Impact of Operating at a Reduced Feedwater Temperature on Source Terms 15.A-2 STPEGS UFSAR TABLE OF CONTENTS (Continued)
CHAPTER 15 Section Title Page TC 15-4 Revision 13 15.A.7 The Impact of Replacement Steam Generators on Source Terms 15.A-3 APPENDIX DOSE MODELS 15.B-1 15.B 15.B.l General Accident Parameters 15.B-1 15.B.2 Offsite Radiological Consequences Calculational Models 15.B-1 15.B.2.l Accident Release Pathways 15.B-1 15.B.2.2 Single-Region Release Model 15.B-2 15.B.2.3 Two Region Spray Model in Containment-LOCA 15.B-3 15.B.2.4 Offsite Thyroid Dose Calculation Model 15.B-4 15.B.2.5 Offsite Beta Skin Dose Calculational Model 15.B-4 15.B.2.6 Offsite Gamma Body Dose Calculational Model 15.B-5 15.B.2.7 Offsite Beta-Skin Dose Calculational Model for SGTR 15.B-5 15.B.2.8 Offsite Gamma Whole-B ody Dose Calculational Model for SGTR 15.B-5 15.B.3 Control Room Radiological Consequences Calculation Models 15.B-6 Radioactive Cloud External to the Control Room 15.B-6 15.B.3.1 Integrated Activity in the Control Room 15.B-6 15.B.3.2 Integrate Activity Concentration in the Control Room from Single-Region System 15.B-6 15.B.3.3 Control Room Thyroid Dose Calculational Model 15.B-7 15.B.3.4 Control Room Beta Skin Dose Calculational Model 15.B-7 15.B.3.5 Control Room Gamma Body Dose Calculation 15.B-8 15.B.4 Dose Conversion Factors 15.B-8 STPEGS UFSAR LIST OF TABLES
CHAPTER l5 Table Title Page TC 15-5 Revision 13 15.0-1 Nuclear Steam Supply System Power Ratings 15.0-18 15.0-2 Summary of Initial Conditions and Computer Codes 15.0-19 15.0-3 Nominal Values of Pertinent Plant Parameters Supported by the Accident Analyses 15.0-23 15.0-4 Trip Points and Time Delays to Trip Assumed in Accident Analyses 15.0-24 15.0-6 Plant Systems and Equipment Available for Transient and Accident Conditions 15.0-26 15.0-7 Single Failures Assumed in Accident Analyses 15.0-31 15.1-1 Time Sequence of Events for Incidents Which Cause an Increase in Heat Removal by the Secondary System 15.l-19 15.1-2 Parameters Used in Stea m Line Break Analysis 15.l-21 15.1-3 Doses Resulting From Steam Line Break 15.l-22 15.2-1 Time Sequence of Events for Incidents Which Cause a Decrease in Heat Removal by the Secondary System 15.2-20 15.3-l Time Sequence of Events for Incidents Which Result in a Decrease in Reactor Coolant System Flow 15.3-13 15.3-2 Summary of Results for Locked Rotor Transients without Offsite Power 15.3-15 15.3-3 Parameters Used in Reactor Coolant Pump Shaft Seizure Accident Analysis 15.3-16 15.3-4 Doses Resulting from Reactor Coolant Pump Shaft Seizure Accident 15.3-17 15.4-l Time Sequence of Events for Incidents Which Cause Reactivity and Power Distribution Anomalies 15.4-39 STPEGS UFSAR LIST OF TABLES (Continued)
CHAPTER l5 Table Title Page TC 15-6 Revision 13 15.4-3 Parameters Used in the Rod Cluster Control Assembly Ejection Accident Analysis 15.4-42 15.4-4 Parameters Used in Rod Ej ection Accident Analysis 15.4-43 15.4-5 Doses Resulting from Rod Ejection Accident 15.4-45 15.5-1 CVCS Malfunction Time Sequences 15.5-7 15.6-1 Time Sequence of Events for Incidents Which Cause a Decrease in Reac tor Coolant Inventory 15.6-28 15.6-2 Parameters Used in Sample Line Failure Radiological Analysis 15.6-29 15.6-3 Parameters Used in Stea m Generator Tube Rupture Radiological Analyses 15.6-30 15.6-4 Doses Resulting from Steam Generator Tube Rupture 15.6-31 15.6-5 Input Parameters Used in Small Break LOCA Analyses 15.6-32 15.6-5a Input Parameters Used in Large Break LOCA Analyses 15.6-34 15.6-6 Large Break Time Sequence of Events 15.6-35 15.6-7 Large Break LOCA Analysis Results 15.6-36 15.6-8 Small Break (2-inch) - Time Sequence of Events 15.6-37 15.6-8a Small Break (2-inch) - Analysis Results 15.6-38 15.6-9 Small Break (3-inch) - Time Sequence of Events 15.6-39 15.6-9a Small Break (3-inch) - Analysis Results 15.6-40 15.6-10 Parameters Used in Analysis of Loss-of-Coolant Accident Offsite Doses 15.6-41 15.6-11 Doses Resulting From Large Break Loss-of-Coolant Accident 15.6-43 15.6-12 Maximum Potential Recirculation Loop Leakage External to Containment 15.6-44 15.6-13 Parameters Used in CVCS Le tdown Line Failure Radiological Analysis 15.6-45 15.6-14 Dose from Small Line Breaks Outside Containment 15.6-46 15.7-3 Parameters Used in Anal ysis of Liquid-Containing Tank Failure 15.7-7 15.7-4 Activity Available for Release 15.7-8 15.7-7 Activities in the Peak Inventory Discharged Assembly (Fuel Handling Accident in the FHB) 15.7-9 15.7-9 Parameters Used for the Fuel Handling Accident in the FHB 15.7-10 15.7-10 Doses Resulting from a Fuel Handling Accident in the FHB 15.7-11 15.7-11 Activities in the Peak Inventory Discharged Assembly (Fuel Handling Accident in the RCB) 15.7-12 STPEGS UFSAR LIST OF TABLES (Continued)
CHAPTER l5 Table Title Page TC 15-7 Revision 13 15.7-12 Parameters Used for the Fuel Handling Accident in the RCB 15.7-13 15.7-13 Doses Resulting from a Fuel Handling Accident in the RCB 15.7-14 15.A-1 Core and Gap Activities Based upon Full-Power Operation for 900 Days 15.A-5 15.A-1A Core Activities for V5H Fuel Upgrade and the Parameters Used in Calculation 15.A-6 15.A-2 Coolant Concentrations - Design Basis 15.A-7 15.A-4 Reactor Coolant Iodine Concentrations 15.A-8 15.A-5 Secondary Coolant Io dine Concentration 15.A-9 15.A-6 Iodine Appearance Rates in the Reactor Coolant 15.A-10 15.B-1 Dispersion Factors 15.B-11 15.B-2 Breathing Rates for an Individual Offsite 15.B-12 15.B-3 Dose Conversion Factors Used in Accident 15.B-13 Analysis 15.B-4 Average Gamma and Beta Energy for Noble Gases and Iodines 15.B-14 STPEGS UFSAR LIST OF FIGURES
CHAPTER l5Figure Number Title TC 15-8 Revision 13 15.0-1A Range of Programmed T avg Considered in the Safety Analysis 15.0-1B Nominal Pressurizer Water Level Assumptions 15.0-1C Illustration of Overtemperatur e and Overpower Delta T Protection 15.0-2 Doppler Power Coefficient Used in Accident Analysis 15.0-3 RCCA Position versus Time to Dashpot 15.0-4 Minimum Trip Reactivity versus Rod Position 15.0-5 Normalized RCCA Negative Reactivity Insertion versus Time 15.0-6 Moderator Density Coefficient 15.0-6a Abbreviations and Symbols Used in Sequence Diagrams 15.0-7 Excessive Heat Removal Due to Feedwater System Malfunction 15.0-8 Excessive Load Increase Incident 15.0-9 Accidental Depressurization of the Main Steam System 15.0-10 Loss of External Electrical Load 15.0-11 Loss of Offsite Power to the Station Auxiliaries 15.0-12 Loss of Normal Feedwater 15.0-13 Major Rupture of a Main Feedwater Line 15.0-14 Loss of Forced Reactor Coolant Flow 15.0-15 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal 15.0-16 Dropped Rod Cluster Control Assembly 15.0-17 Single Rod Cluster Control Assembly Withdrawal at Full Power 15.0-18 Startup of an Inactive Reactor Coolant Loop
STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5
Figure Number Title TC 15-9 Revision 13 15.0-19 Uncontrolled Boron Dilution 15.0-20 Rupture of Control Rod Drive Mechanism Housing 15.0-21 Inadvertent ECCS Operation at Power 15.0-22 Accidental Depressurization of the Reactor Coolant System 15.0-23 Sample Line Failure 15.0-24 Steam Generator Tube Rupture 15.0-25 Loss of Coolant Accident 15.0-27 Liquid Waste Processing System Tank(s) Failure 15.0-28 Fuel Handling Accident in the Fuel Building 15.0-29 Fuel Handling Accident in the Containment 15.0-30 Spent Fuel Cask Drop (From Ht < 30 ft) 15.0-31 CVCS Letdown Line Break Outside Containment 15.1-1 Feedwater Control Valve Malfunction, Manual Rod Control - DNBR and Power 15.1-2 Feedwater Control Valve Malfunction, Manual Rod Control - Pressurizer Pressure and Vessel Temperature 15.1-3 Ten Percent Step Load Increase, Minimum Moderator Feedback, Manual Reactor Control 15.1-4 Ten Percent Step Load Increase, Minimum Moderator Feedback, Manual Reactor Control 15.1-5 Ten Percent Step Load Increase, Maximum Moderator Feedback, Manual Reactor Control 15.1-6 Ten Percent Step Load Increase, Maximum Moderator Feedback, Manual Reactor Control
STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-10 Revision 13 15.1-7 Ten Percent Step Load Increase, Minimum Moderator Feedback, Automatic Reactor Control 15.1-8 Ten Percent Step Load Increase, Minimum Moderator Feedback, Automatic Reactor Control 15.1-9 Ten Percent Step Load Increase, Maximum Moderator Feedback, Automatic Reactor Control 15.1-10 Ten Percent Step Load Increase, Maximum Moderator Feedback, Automatic Reactor Control 15.1-11 Keff versus Core Average Temperature 15.1-l4 Doppler Power Feedback 15.1-15 l.4 ft 2 Steam Line Rupture, Offsite Power Available - Heat Flux, Average Temperature, Plant Steam Flow 15.1-16 l.4 ft 2 Steam Line Rupture, Offsite Power Available - RCS Pressure and Pressurizer Water Volume 15.1-17 l.4 ft 2 Steam Line Rupture, Offsite Power Available - Boron Concentration and Reactivity 15.2-1 Turbine Trip Accident with Pr essurizer Spray and Power-Operated Relief Valves - Power and Vessel Temperature 15.2-1A Turbine Trip Accident with Pressure Operated Spray and Power-Operated Relief Valves 15.2-2 Turbine Trip Accident with Pr essurizer Spray and Power-Operated Relief Valves - Pressurizer Pressure and Volume 15.2-2A Turbine Trip Accident with Pressure Operated Spray and Power Operated Relief Valves 15.2-3 Turbine Trip Accident with Pr essurizer Spray and Power-Operated Relief Valves - DNBR and Steam Pressure 15.2-3A Turbine Trip Accident with Pressure Operated Spray and Power-Operated Relief Valves
STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-11 Revision 13 15.2-4 Turbine Trip Accident without Pr essurizer Spray or Power-Operated Relief Valves - Power and Vessel Temperature 15.2-5 Turbine Trip Accident without Pr essurizer Spray or Power-Operated Relief Valves - Pressurizer Pressure and Volume 15.2-6 Turbine Trip Accident without Pr essurizer Spray or Power-Operated Relief Valves - Steam Pressure 15.2-9A Loss of Normal Feedwater wit hout Offsite Power - Power and Core Heat Flux 15.2-9B Loss of Normal Feedwater wi thout Offsite Power - RCS Flow 15.2-9C Loss of Normal Feedwater without Offsite Power - Loop Temperatures 15.2-9D Loss of Normal Feedwater without Offsite Power - Pressurizer Pressure and Volume 15.2-10 Loss of Normal Feedwater without Offsite Power - SG Pressure and Mass 15.2-11 Feedline Break with Offsite Po wer - Core Power and Break Flow 15.2-12 Feedline Break with Offsite Power - Pressurizer Pressure and Volume 15.2-13 Feedline Break with Offsite Power - SG Pressure and Mass 15.2-14 Feedline Break with Offsite Power - Loop Temperatures 15.2-15 Feedline Break without Offsite Power - Power and Break Flow 15.2-l6 Feedline Break without Offsite Power - Pressurizer Pressure and Volume 15.2-l7 Feedline Break without Offsite Power - SG Pressure and Mass 15.2-l8 Feedline Break without Offs ite Power - Loop Temperatures 15.3-1 Flow Transients for Partial Loss of Flow, Four Loops in Operation, One Pump Coasting Down l5.3-2 Nuclear Power Transient and Pressu rizer Pressure Transient for Partial Loss of Flow, Four Loops in Operation, One Pump Coasting Down STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-12 Revision 13 15.3-3 Average and Hot Channel Heat Flux Transients for Partial Loss of Flow, Four Loops in Operation, One Pump Coasting Down 15.3-4 DNBR versus Time for Partial Loss of Flow, Four Loops in Operation, One Pump Coasting Down 15.3-9 Core Flow Coastdown versus Time for Four Loops in Operation, Four Pumps Coasting Down, Complete Loss of Flow 15.3-10 Nuclear Power Transient and Pressu rizer Pressure Transient for Four Loops in Operation, Four Pumps Coasting Down, Complete Loss of Flow 15.3-11 Average and Hot Channel Heat Fl ux Transients for Four Loops in Operation, Four Pumps Coasting Down, Complete Loss of Flow 15.3-12 DNBR versus Time for Four Loops in Operation, Four Pumps Coasting Down, Complete Loss of Flow 15.3-l7 Flow Transients for Four Loops in Operation, One Locked Rotor 15.3-l8 Reactor Coolant System Pressure Transient for Four Loops in Operation, One Locked Rotor 15.3-l9 Nuclear Power Transient, Av erage and Hot Channel Heat Flux Transients for Four Loops in Operation, One Locked Rotor 15.3-20 Maximum Clad Temperature at Hot Spot for Four Loops in Operation, One Locked Rotor 15.4-1 Neutron Power Transient for Uncontrolled Rod Withdrawal from a Subcritical Condition 15.4-2 Thermal Flux Transient for Uncontrolled Rod Withdrawal from a Subcritical Condition 15.4-3 Fuel and Clad Temperature Tr ansients for Uncontrolled Rod Withdrawal from a Subcritical Condition 15.4-4 Nuclear Power Transient and Core Heat Flux Transient for Uncontrolled Rod Withdrawal from Full Power with Minimum Feedback and 3 pcm/sec Withdrawal Rate STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-13 Revision 13 15.4-5 Pressurizer Pressure and Water Volume Transients for Uncontrolled Rod Withdrawal from Full Power with Minimum Feedback and 3 pcm/sec Withdrawal Rate 15.4-6 Core Average Temperature Transient and DNBR Transient for Uncontrolled Rod Withdrawal from Full Power with Minimum Feedback and 3 pcm/sec Withdrawal Rate 15.4-10 Minimum DNBR verses Reactivity Insertion Rate - Rod Withdrawal at Power (l00% Power) 15.4-11 Minimum DNBR verses Reactivity Insertion Rate - Rod Withdrawal at Power (60% Power) 15.4-12 Minimum DNBR verses Reactivity Insertion Rate - Rod Withdrawal at Power (l0% Power) 15.4-13 Nuclear Power and Core Heat Flux Transients for Rod Cluster Control Assembly Misoperation (2 sheets) 15.4-14 Pressurizer Pressure Transient and Core Average Temperature Transient for Rod Cluster Control Assembly Misoperation (2 sheets) 15.4-16 Nuclear Power Transient Duri ng Startup of an Inactive Loop 15.4-17 Average and Hot Channel Heat Fl ux Transients During Startup of an Inactive Loop
STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-14 Revision 13 15.4-l8 Core Flow During Startup of an Inactive Loop 15.4-l9 Pressurizer Pressure Transient and Core Average Temperature Transient During Startup of an Inactive Loop 15.4-20 DNBR Transient During Startup of an Inactive Loop 15.4-2l Interchange Between Region l and Region 3 Assembly 15.4-22 Interchange Between Region l and Region 2 Assembly, Burnable Absorber Rods Being Retained by the Region 2 Assembly 15.4-23 Interchange Between Region l and 2 Assembly, Burnable Absorber Rods Being Transferred to the Region l Assembly 15.4-24 Enrichment Error: a Region 2 Asse mbly Loaded into the Core Central Position 15.4-25 Loading a Region 2 Assembly into a Region l Position Near Core Periphery 15.4-26 Nuclear Power Transient, BOL HZP Rod Ejection Accident 15.4-27 Hot Spot Fuel and Clad Temperature versus Time, BOL HZP Rod Ejection Accident 15.4-28 Nuclear Power Transient, EOL HFP Rod Ejection Accident 15.4-29 Hot Spot Fuel and Clad Temperature versus Time, EOL HFP Rod Ejection Accident 15.4-30 BOL HFP Rod Ejection Accident FQ versus Ejected Rod Worth 15.4-31 EOL HZP Rod Ejection Accident FQ versus Ejected Rod Worth 15.5-1A Charging System Malfunction with Pressurizer Spray and Heaters, Minimum Reactivity Feedback 15.5-1B Charging System Malfunction with Pressurizer Spray and Heaters, Minimum Reactivity Feedback
STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-15 Revision 13 15.5-2A Charging System Malfunction with Pressurizer Spray and Heaters, Maximum Reactivity Feedback 15.5-2B Charging System Malfunction with Pressurizer Spray and Heaters, Maximum Reactivity Feedback 15.6-1 Nuclear Power Transient for Inadvertent Opening of a Pressurizer Safety Valve 15.6-2 Pressurizer Pressure Transient and Core Average Temperature Transient for Inadvertent Opening of a Pressurizer Safety Valve 15.6-3 DNBR Transient for Inadvertent Open ing of a Pressurizer Safety Valve 15.6-4 Steam Generator Tube Rupture Margin to Overfill Analysis - Ruptured SG Water Volume 15.6-5 Sequence of Events for Large Break Loss of Coolant Analysis 15.6-6 Code Interface Description for Large Break Model 15.6-7 Code Interface Description for Small Break Model 15.6-8 Peak Clad Temperature - DECLG, C D = 0.6, Min SI, High T AVG , non-IFBA 15.6-9 Core Pressure - DECLG, C D = 0.6, Min SI, High T AVG , non-IFBA 15.6-10 Reflood Transient Downcomer and Core Water Levels - DECLG, C D = 0.6, Min SI, High T AVG , non-IFBA 15.6-11 Reflood Transient Core Inlet Fluid Velocity - DECLG, C D = 0.6, Min SI, High T AVG , non-IFBA 15.6-12 Core Power Transient - DECLG, C D = 0.6, Min SI, High T AVG , non-IFBA 15.6-13 Containment Pressure - DECLG, C D = 0.6, Min SI, High T AVG , non-IFBA 15.6-14 Peak Clad Temperature - DECLG, C D = 0.8, Min SI, High T AVG , non-IFBA STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-16 Revision 13 15.6-14a Peak Clad Temperature - DECLG, C D = 0.8, Min SI, Low TAVG , non-IFBA 15.6-14b Peak Clad Temperature - DECLG, C D = 0.8, Max SI, High T AVG , IFBA 15.6-15 Core Pressure - DECLG, C D = 0.8, Min SI, High T AVG , non-IFBA 15.6-15a Core Pressure - DECLG, C D = 0.8, Min SI, Low T AVG , non-IFBA 15.6-15b Core Pressure - DECLG, C D = 0.8, Max SI, High T AVG , IFBA 15.6-16 Reflood Transient Downcomer and Core Water Levels - DECLG, C D = 0.8, Min SI, High T AVG , non-IFBA 15.6-16a Reflood Transient Downcomer and Core Water Levels - DECLG, C D = 0.8, Min SI, Low TAVG , non-IFBA 15.6-16b Reflood Transient Downcomer and Core Water Levels - DECLG, C D = 0.8, Max SI, High T AVG , IFBA 15.6-17 Reflood Transient Core Inlet Fluid Velocity - DECLG, C D = 0.8, Min SI, High T AVG , non-IFBA 15.6-17a Reflood Transient Core Inlet Fluid Velocity - DECLG, C D = 0.8, Min SI, Low T AVG , non-IFBA 15.6-17b Reflood Transient Core Inlet Fluid Velocity - DECLG, C D = 0.8, Max SI, High T AVG , IFBA 15.6-18 Core Power Transient - DECLG, C D = 0.8, Min SI, High T AVG , non-IFBA 15.6-18a Core Power Transient - DECLG, C D = 0.8, Min SI, Low TAVG , non-IFBA 15.6-18b Core Power Transient - DECLG, C D = 0.8, Max SI, High T AVG , IFBA STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-17 Revision 13 15.6-19 Containment Pressure - DECLG, C D = 0.8, Min SI, High T AVG , non-IFB A 15.6-19a Containment Pr essure - DECLG, C D = 0.8, Min SI, Low TAVG , non-IFBA 15.6-19b Containment Pressure - DECLG, C D = 0.8, Max SI, High T AVG , IFBA 15.6-20 Peak Clad Temperature - DECLG, C D = 1.0, Min SI, High T AVG , non-IFBA 15.6-21 Core Pressure - DECLG, C D = 1.0, Min SI, High T AVG , non-IFBA 15.6-22 Reflood Transient Downcomer and Core Water Levels - DECLG, C D = 1.0, Min SI, High T AVG , non-IFBA 15.6-23 Reflood Transient Core Inlet Fluid Velocity - DECLG, C D = 1.0, Min SI, High T AVG , non-IFBA 15.6-24 Core Power Transient - DECLG, C D = 1.0, Min SI, High T AVG , non-IFBA 15.6-25 Containment Pressure - DECLG, C D = 1.0, Min SI, High T AVG , non-IFBA 15.6-26 Peak Clad Temperature - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-27 Core Pressure - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-28 Reflood Transient Downcomer and Core Water Levels - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-29 Reflood Transient Core Inlet Fluid Velocity - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-30 Core Power Transient - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-31 Containment Pressure - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-32 Core Flow, Inlet and Outlet - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-33 Heat Transfer Coefficient - DECLG, C D = 0.8, Min SI, High T AVG , IFBA STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-18 Revision 13 15.6-34 Fluid Temperature at Hot Spot - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-35 Break Mass Flow Rate - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-36 Break Energy Flow Rate to Containment - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-37 Fluid Quality at Hot Spot - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-38 Mass Flux at Hot Spot - DECLG, C D = 0.8, Min SI, High T AVG , IFBA 15.6-39 Accumulator Flow Rate (Blowdown) - DECLG, C D = 0.8, Minimum SI, High T AVG , IFBA 15.6-40 ECCS Injection Rate During Reflood - DECLG, C D = 0.8, Minimum SI, High T AVG , IFBA 15.6-41 RCS Depressurization Transient (2-in. Break, High Loop Temperature, High MFW Temperature) 15.6-42 Core Mixture Height (2-in. Break, High Loop Temperature, High MFW Temperature) 15.6-43 Core Steam Flow (2-in. Break, High Loop Temperature, High MFW Temperature) 15.6-44 Clad Temperature Transient (2-in. Break, High Loop Temperature, High MFW Temperature) 15.6-45 Hot Spot Fluid Temperature in. Break, High Loop Temperature, High MFW Temperature 15.6-46 Rod Film Heat Transfer Co efficient in. Break, High Loop Temperature, High MFW Temperature 15.6-47 Core Power After Reactor Trip 15.6-48 RCS Depressurization Transient -
2-in. Break, High Loop Temperature, Low MFW Temperature
STPEGS UFSAR LIST OF FIGURES (Continued)
CHAPTER l5 Figure Number Title TC 15-19 Revision 13 15.6-49 Core Mixture Height in. Break, High Loop Temperature, Low MFW Temperature 15.6-50 Clad Temperature Transient in. Break, High Loop Temperature, Low MFW Temperature 15.6-5l RCS Depressurization Transient in. Break, Low Loop Temperature, High MFW Temperature 15.6-52 Core Mixture Height in. Break, Low Loop Temperature, High MFW Temperature 15.6-53 Clad Temperature Transient in. Break, Low Loop Temperature, High MFW Temperature 15.6-54 RCS Depressurization Transient in. Break, Low Loop Temperature, Low MFW Temperature 15.6-55 Core Mixture Height in. Break, Low Loop Temperature, Low MFW Temperature 15.6-56 Clad Temperature Transient in. Break, Low Loop Temperature, Low MFW Temperature 15.6-57 Small Break Safety Injection Flow Rate vs. RCS Pressure 15.6-58 Small Break Power Shape 15.B-2 Release Pathways 15.B-3 Control Room Pathways