ML19170A347
| ML19170A347 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/20/2019 |
| From: | Mahesh Chawla Plant Licensing Branch III |
| To: | Exelon Generation Co |
| Chawla M | |
| Shared Package | |
| ML18355A456 | List: |
| References | |
| Download: ML19170A347 (739) | |
Text
B/B-UFSAR CHAPTER 15.0 - ACCIDENT ANALYSES TABLE OF CONTENTS PAGE 15.0 ACCIDENT ANALYSES 15.0-1 15.0.1 Classification of Plant Conditions 15.0-1 15.0.1.1 Condition I - Normal Operation and Operational Transients 15.0-2 15.0.1.2 Condition II - Faults of Moderate Frequency 15.0-3 15.0.1.3 Condition III - Infrequent Faults 15.0-5 15.0.1.4 Condition IV - Limiting Faults 15.0-5 15.0.1.5 Summary of Results 15.0-6 15.0.2 Optimization of Control Systems 15.0-7 15.0.3 Plant Characteristics and Initial Conditions Assumed in the Accident Analyses 15.0-7 15.0.3.1 Design Plant Conditions 15.0-7 15.0.3.2 Initial Conditions 15.0-8 15.0.3.3 Power Distribution 15.0-9 15.0.4 Reactivity Coefficients Assumed in the Accident Analyses 15.0-10 15.0.5 Rod Cluster Control Assembly Insertion Characteristics 15.0-10 15.0.6 Trip Points and Time Delays to Trip Assumed in Accident Analyses 15.0-11 15.0.7 Instrumentation Drift and Calorimetric Errors -
Power Range Neutron Flux 15.0-12a 15.0.8 Plant Systems and Components Available for Mitigation of Accident Effects 15.0-13 15.0.9 Fission Product Inventories 15.0-14 15.0.9.1 Activities in the Core 15.0-14 15.0.9.2 Activities in the Fuel Pellet Cladding Gap 15.0-14 15.0.9.3 Activities in the Primary Coolant 15.0-15 15.0.9.4 Activities in the Secondary Coolant 15.0-15a 15.0.10 Residual Decay Heat 15.0-15a 15.0.10.1 Total Residual Heat 15.0-15a 15.0.10.2 Distribution of Decay Heat Following Loss of Coolant Accident 15.0-15a 15.0.11 Computer Codes Utilized 15.0-16 15.0.11.1 FACTRAN 15.0-16 15.0.11.2 LOFTRAN 15.0-16 15.0.11.3 TWINKLE 15.0-17 15.0.11.4 Deleted 15.0-17 15.0.11.5 Deleted 15.0-17 15.0.11.6 RADTRAD 15.0-17a 15.0.11.7 ARCON96 15.0-17a 15.0.11.8 PAVAN 15.0-17b 15.0.11.9 VIPRE 15.0-17b 15.0.11.10 ANC 15.0-17b 15.0.11.11 LOFTTR2 15.0-17c 15.0.12 Radiological Consequences 15.0-17c 15.0.13 Limiting Single Failures 15.0-19b 15.0.14 Operator Actions 15.0-20 15.0.15 References 15.0-23a 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.1-1 15.1.1 Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature 15.1-1 15.0-i REVISION 17 - DECEMBER 2018
B/B-UFSAR TABLE OF CONTENTS (Cont'd)
PAGE 15.1.1.1 Identification of Causes and Accident Description 15.1-1 15.1.1.2 Analysis of Effects and Consequences 15.1-2 15.1.1.3 Radiological Consequences 15.1-3 15.1.1.4 Conclusions 15.1-3 15.1.2 Feedwater System Malfunctions Causing an Increase in Feedwater Flow 15.1-3 15.1.2.1 Identification of Causes and Accident Description 15.1-3 15.1.2.2 Analysis of Effects and Consequences 15.1-4 15.1.2.3 Radiological Consequences 15.1-6 15.1.2.4 Conclusions 15.1-6 15.1.3 Excessive Increase in Secondary Steam Flow 15.1-7 15.1.3.1 Identification of Causes and Accident Description 15.1-7 15.1.3.2 Analysis of Effects and Consequences 15.1-7 15.1.3.3 Radiological Consequences 15.1-9 15.1.3.4 Conclusions 15.1-9 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1-10 15.1.5 Steam System Piping Failure at Zero Power 15.1-14 15.1.5.1 Identification of Causes and Accident Description 15.1-14 15.1.5.2 Analysis of Effects and Consequences 15.1-16 15.1.5.3 Radiological Consequences of a Postulated Steamline Break using AST 15.1-21 15.1.5.4 Conclusions 15.1-23 15.1.6 Steam System Piping Failure at Full Power 15.1-23 15.1.6.1 Identification of Causes and Accident Description 15.1-23 15.1.6.2 Analysis of Effects and Consequences 15.1-23 15.1.7 References 15.1-23b 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2-1 15.2.1 Steam Pressure Regulator Malfunction or Failure That Results in Decreasing Steam Flow 15.2-1 15.2.2 Loss of External Load 15.2-1 15.2.2.1 Identification of Causes and Accident Description 15.2-1 15.2.2.2 Analysis of Effects and Consequences 15.2-3 15.2.2.3 Radiological Consequences 15.2-4 15.2.2.4 Conclusions 15.2-4 15.2.3 Turbine Trip 15.2-4 15.2.3.1 Identification of Causes and Accident Description 15.2-4 15.2.3.2 Analysis of Effects and Consequences 15.2-5 15.0-ii REVISION 17 - DECEMBER 2018
B/B-UFSAR TABLE OF CONTENTS (Cont'd)
PAGE 15.2.3.3 Radiological Consequences 15.2-8 15.2.3.4 Conclusions 15.2-9 15.2.4 Inadvertent Closure of Main Steam Isolation Valves 15.2-9 15.2.5 Loss of Condenser Vacuum and Other Events Causing a Turbine Trip 15.2-9 15.2.6 Loss of Nonemergency AC Power to the Plant Auxiliaries (Loss of Offsite Power) 15.2-13a 15.2.6.1 Identification of Causes and Accident Description 15.2-13a 15.2.6.2 Analysis of Effects and Consequences 15.2-13c 15.2.6.3 Radiological Consequences 15.2-13g 15.2.6.4 Conclusions 15.2-13g 15.2.7 Loss of Normal Feedwater Flow 15.2-16a 15.2.7.1 Identification of Causes and Accident Description 15.2-16a 15.2.7.2 Analysis of Effects and Consequences 15.2-16b 15.2.7.3 Radiological Consequences 15.2-16e 15.2.7.4 Conclusions 15.2-16e 15.2.8 Feedwater System Pipe Break 15.2-24a 15.2.8.1 Identification of Causes and Accident Description 15.2-24a 15.2.8.2 Analysis of Effects and Consequences 15.2-24d 15.2.8.3 Radiological Consequences 15.2-24i 15.2.8.4 Conclusions 15.2-24i 15.2.9 References 15.2-24i 15.0-iii REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE OF CONTENTS (Cont'd)
PAGE 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 15.3-1 15.3.1 Partial Loss of Forced Reactor Coolant Flow 15.3-1 15.3.1.1 Identification of Causes and Accident Description 15.3-1 15.3.1.2 Analysis of Effects and Consequences 15.3-2 15.3.1.3 Radiological Consequences 15.3-3 15.3.1.4 Conclusions 15.3-4 15.3.2 Complete Loss of Forced Reactor Coolant Flow 15.3-4 15.3.2.1 Identification of Causes and Accident Description 15.3-4 15.3.2.2 Analysis of Effects and Consequences 15.3-5 15.3.2.3 Radiological Consequences 15.3-6 15.3.2.4 Conclusions 15.3-6 15.3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor) 15.3-6 15.3.3.1 Identification of Causes and Accident Description 15.3-7 15.3.3.2 Analysis of Effects and Consequences 15.3-7 15.3.3.3 Conclusions 15.3-9b 15.3.3.4 Radiological Consequences 15.3-10 15.3.3.4.1 Source Term 15.3-11 15.3.3.5 Radiological Conclusions 15.3-11 15.3.3.6 Deleted 15.3-l2 15.3.4 Reactor Coolant Pump Shaft Break 15.3-l3 15.3.5 Locked Rotor With Loss of Offsite Power 15.3-17 15.3.6 References 15.3-17 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4-1 15.4.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Subcritical or Low Power Startup Condition 15.4-1 15.4.1.1 Identification of Causes and Accident Description 15.4-1 15.4.1.2 Analysis of Effects and Consequences 15.4-3 15.4.1.3 Radiological Consequences 15.4-6 15.4.1.4 Conclusions 15.4-6 15.4.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power 15.4-6 15.4.2.1 Identification of Causes and Accident Description 15.4-6 15.4.2.2 Analysis of Effects and Consequences 15.4-8 15.4.2.3 Radiological Consequences 15.4-11 15.4.2.4 Conclusions 15.4-11 15.4.3 Rod Cluster Control Assembly Misoperation (System Malfunction or Operator Error) 15.4-12 15.4.3.1 Identification of Causes and Accident Description 15.4-12 15.4.3.2 Analysis of Effects and Consequences 15.4-14 15.4.3.3 Radiological Consequences 15.4-18 15.4.3.4 Conclusions 15.4-19 15.4.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature 15.4-19 15.0-iv REVISION 17 - DECEMBER 2018
B/B-UFSAR TABLE OF CONTENTS (Cont'd)
PAGE 15.4.5 A Malfunction or Failure of the Flow Controller in a BWR Loop that Results in an Increased Reactor Coolant Flow Rate 15.4-22 15.4.6 Chemical and Volume Control System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant 15.4-22 15.4.6.1 Identification of Causes and Accident Description 15.4-22 15.4.6.2 Analysis of Effects and Consequences 15.4-24 15.4.6.3 Results and Conclusions 15.4-27 15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 15.4-29 15.4.7.1 Acceptance Criteria 15.4-29 15.4.7.2 Identification of Causes and Accident Description 15.4-30 15.4.7.3 Evaluation 15.4-31 15.4.7.4 Conclusions 15.4-31a 15.4.8 Spectrum of Rod Cluster Control Assembly Ejection Accidents 15.4-32 15.4.8.1 Identification of Causes and Accident Description 15.4-32 15.4.8.1.1 Design Precautions and Protection 15.4-32 15.4.8.1.2 Limiting Criteria 15.4-34 15.4.8.2 Analysis of Effects and Consequences 15.4-35 15.4.8.2.1 Calculation of Basic Parameters 15.4-36 15.4.8.2.2 Results 15.4-39 15.4.8.3 Radiological Consequences of a Postulated Rod Ejection Accident (AST) 15.4-41 15.4.8.4 Conclusions 15.4-44 15.4.9 References 15.4-44 15.5 INCREASE IN REACTOR COOLANT INVENTORY 15.5-1 15.5.1 Inadvertent Operation of Emergency Core Cooling System During Power Operation 15.5-1 15.5.1.1 Identification of Causes and Accident Description 15.5-1 15.5.1.2 Analysis of Effects and Consequences 15.5-2 15.5.1.3 Radiological Consequences 15.5-7 15.5.1.4 Conclusions 15.5-7 15.5.2 Chemical and Volume Control Systems Malfunction that Increases Reactor Coolant Inventory 15.5-8 15.5.3 A Number of BWR Transients 15.5-8 15.5.4 References 15.5-8 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 15.6.1.1 Identification of Causes and Accident Description 15.6-1 15.6.1.2 Analysis of Effects and Consequences 15.6-2 15.6.1.3 Radiological Consequences 15.6-3 15.6.1.4 Conclusions 15.6-3 15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6-3 15.0-v REVISION 17 - DECEMBER 2018
B/B-UFSAR TABLE OF CONTENTS (Cont'd)
PAGE 15.6.2.1 Identification of Causes and Accident Description 15.6-3 15.6.2.2 Analysis of Effects and Consequences 15.6-4 15.6.2.3 Radiological Consequences 15.6-4 15.6.3 Steam Generator Tube Rupture 15.6-5 15.6.3.1 Identification of Causes and Accident Description 15.6-5 15.6.3.2 Analysis of Effects and Consequences 15.6-9 15.6.3.3 Radiological Consequences 15.6-12a 15.6.3.4 Conclusions 15.6-12g 15.6.4 Spectrum of BWR Steam System Piping Failures Outside of Containment 15.6-12g 15.6.5 Loss-of-Coolant Accidents Resulting From a Spectrum of Postulated Piping Breaks Within The Reactor Coolant Pressure Boundary 15.6-12g 15.6.5.1 General 15.6-13 15.6.5.1.1 Performance Criteria for Emergency Core Cooling System 15.6-13 15.6.5.1.2 Accident Description 15.6-14 15.6.5.2 Thermal Analysis 15.6-14a 15.6.5.2.1 Large Break Analysis 15.6-15 15.6.5.2.1.1 General 15.6-15 15.6.5.2.1.2 Design Basis Accident 15.6-17 15.6.5.2.1.3 Analysis Assumptions 15.6-17 15.6.5.2.1.4 Method of Analysis, Large Break 15.6-17 15.6.5.2.2 Small Break LOCA Analysis 15.6-18 15.6.5.2.2.1 Description of Analysis and Assumptions 15.6-18 15.6.5.2.2.2 Method of Analysis, Small Break 15.6-19a 15.6.5.2.3 Results 15.6-20a 15.6.5.2.3.1 Results of Large Break LOCA Analysis 15.6-20a 15.6.5.2.3.1.1 Deleted 15.6-20a 15.6.5.2.3.1.2 Deleted 15.6.21b 15.6.5.2.3.1.2.1 Deleted 15.6-21b 15.6.5.2.3.1.2.2 Deleted 15.6.21b 15.6.5.2.3.1.2.3 Deleted 15.6.21c 15.6.5.2.3.1.3 Deleted 15.6.21c 15.6.5.2.3.1.4 Deleted 15.6.21d 15.6.5.2.3.1.5 Deleted 15.6-21e 15.6.5.2.3.2 Results of Small Break LOCA Analysis 15.6-21e 15.6.5.2.3.2.1 Unit 1 Analysis 15.6-21e 15.6.5.2.3.2.2 Unit 2 Analysis 15.6-21f 15.6.5.2.3.2.3 Safety Injection Evaluation 15.6-21g 15.6.5.2.3.3 Post Analysis of Record Evaluations 15.6-21h 15.6.5.2.3.3.1 Large Break LOCA 15.6-21h 15.6.5.2.3.3.2 Small Break LOCA - Unit 1 15.6-21i 15.6.5.2.3.3.3 Small Break LOCA - Unit 2 15.6-21i 15.6.5.2.4 Post-LOCA Long-Term Core Cooling/
Subcriticality 15.6-21i 15.6.5.2.5 Conclusions - Thermal Analysis 15.6-21k 15.6.5.3 Radiological Consequences of a Postulated Loss-of-Coolant Accident 15.6-22 15.6.5.4 Deleted 15.6-25 15.6.6 BWR Transient 15.6-26 15.6.7 References 15.6-26 15.0-vi REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE OF CONTENTS (Cont'd)
PAGE 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7-1 15.7.1 Gas Waste System Leak or Failure 15.7-1 15.7.1.1 Identification of Causes and Accident Description 15.7-1 15.7.1.2 Analysis of Effects and Consequences 15.7-1 15.7.1.3 Radiological Consequences of a Postulated Waste Gas Decay Tank Rupture 15.7-1 15.7.2 Radioactive Liquid Waste System Leak or Failure (Atmospheric Release) 15.7-2 15.7.2.1 Identification of Causes and Frequency Classification 15.7-2 15.7.2.2 Sequence of Events and Systems Operation 15.7-3 15.7.2.3 Modeling of Accident Sequence 15.7-4 15.7.2.3.1 Mathematical Model 15.7-4 15.7.2.3.2 Input Parameters and Initial Conditions 15.7-4 15.7.2.3.3 Results 15.7-5 15.0-vi (cont'd) REVISION 6 - DECEMBER 1996
B/B-UFSAR TABLE OF CONTENTS (Cont'd)
PAGE 15.7.3 Postulated Radioactive Release Due to Liquid Tank Failure (Ground Release) 15.7-5 15.7.3.1 Identification of Causes of Frequency Classification 15.7-5 15.7.3.2 Sequence of Events and Systems Operation 15.7-5 15.7.3.3 Modeling of Accident Sequence 15.7-5 15.7.3.3.1 Mathematical Model 15.7-5 15.7.3.3.2 Input Parameters and Initial Conditions 15.7-6 15.7.3.4 Radiological Consequences 15.7-6 15.7.4 Fuel Handling Accidents 15.7-6 15.7.4.1 Accident Description 15.7-6 15.7.4.2 Analysis of Effects and Consequences 15.7-6 15.7.4.2.1 Fuel Handling Accident Inside Spent Fuel Storage Building 15.7-7 15.7.4.2.1.1 Identification of Causes and Frequency Classification 15.7-7 15.7.4.2.1.2 Sequence of Events and Systems Operation 15.7-7 15.7.4.2.1.3 System Operation 15.7-8 15.7.4.2.1.4 Radiation Monitoring System 15.7-8a 15.7.4.2.2 Fuel Handling Accident Inside Containment 15.7-9 15.7.4.2.2.1 Identification of Causes and Frequency Classification 15.7-9 15.7.4.2.2.2 Sequence of Events and Systems Operation 15.7-9a 15.7.4.2.2.3 System Operation 15.7-10 15.7.4.2.2.4 Radiation Monitoring System 15.7-11 15.7.4.3 Radiological Consequences of a Postulated Fuel Handling Accident (AST) 15.7-12 15.7.5 Spent Fuel Cask Drop Accident 15.7-12c 15.7.5.1 Identification of Causes and Frequency Classification 15.7-12c 15.7.5.2 Evaluation and Analysis 15.7-13 15.7.5.3 Barrier Performance 15.7-14a 15.7.5.4 Modeling of Accident Analysis 15.7-15 15.7.5.4.1 Mathematical Model 15.7-15 15.7.5.4.2 Input Parameters and Initial Conditions 15.7-15 15.7.5.4.3 Results 15.7-15 15.7.5.5 Radiological Consequences 15.7-15 15.7.6 References 15.7-15a 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) 15.8-1 15.8.1 References 15.8-2 ATTACHMENT 15A - DOSE MODELS USED TO EVALUATE THE RADIOLOGICAL CONSEQUENCES OF ACCIDENTS 15.0-vii REVISION 16 - DECEMBER 2016
B/B-UFSAR CHAPTER 15.0 - ACCIDENT ANALYSES LIST OF TABLES NUMBER TITLE PAGE 15.0-1 Nuclear Steam Supply System Power Ratings 15.0-25 15.0-2 Summary of Initial Conditions and Computer Codes Used 15.0-26 15.0-3 Nominal Values of Pertinent Plant Parameters Utilized in Accident Analyses (BWI Steam Generators) 15.0-34 15.0-4 Nominal Values of Pertinent Plant Parameters Utilized in Accident Analyses (D5 Steam Generators) 15.0-35 15.0-5 Trip Points and Time Delays to Trip Assumed in Accident Analyses 15.0-36 15.0-6 Determination of Maximum Overpower Trip Point Power Range Neutron Flux Channel -
Based on Nominal Setpoint Considering Inherent Instrument Errors 15.0-37 15.0-7 Plant Systems and Equipment Credited for Transients and Accident Conditions 15.0-39 15.0-8 Iodine and Noble Gas Inventory in Reactor Core (For use with TID-14844 Based Analyses) 15.0-43 15.0-9 Primary Coolant Noble Gas Activity Based on 1% Fuel Defects (For use with TID-14844 Based Analyses and AST Analyses Using Regulatory Guide 1.183) 15.0-44 15.0-10 Iodine Activity in the Primary and Secondary Coolant (For use with TID-14844 Based Analyses and AST Analyses Using Regulatory Guide 1.183) 15.0-45 15.0-11 Potential Doses Due to Accidents (Byron) 15.0-46 15.0-12 Potential Doses Due to Accidents (Braidwood) 15.0-48 15.0-13 Accident Atmospheric Dilution Factors
(/Q) at Exclusion Area Boundary and Low Population Zone for the Byron Station (Using TID-14844 Analyses) 15.0-50 15.0-14 Accident Atmospheric Dilution Factors
(/Q) at Exclusion Area Boundary and Low Population Zone for the Braidwood Station (Using TID-14844 Analyses) 15.0-51 15.0-15 Single Failures Assumed in Accident Analyses 15.0-52 15.0-16 Reactor Core Nuclide Inventory Using AST 15.0-54 15.0-17 Byron-Braidwood Maximum Site-Unit /Q Summary 15.0-55 15.1-1 Time Sequence of Event for Incidents Which Cause an Increase in Heat Removal by the Secondary System 15.1-24 15.1-2 Equipment Required Following a Break of a Main Steam Line 15.1-26 15.1-3 Input Parameters for the MSLB Radiological Consequence Analyses Using AST 15.1-29 15.1-3a Deleted 15.1-30a 15.1-4 Deleted 15.1-31 15.1-4a Deleted (Byron) 15.1-32 15.1-4a Deleted (Braidwood) 15.1-33 15.1-4b Deleted 15.1-33a 15.0-viii REVISION 15 - DECEMBER 2014
B/B-UFSAR LIST OF TABLES (Cont'd)
NUMBER TITLE PAGE 15.2-1 Time Sequence of Events for Incidents Which Cause a Decrease in Heat Removal by the Secondary System 15.2-25 15.2-2 Deleted 15.2-30 15.2-3 Deleted 15.2-31 15.2-4 Deleted 15.2-32 15.3-1 Time Sequence of Events for Incidents Which Result in a Decrease in Reactor Coolant System Flow 15.3-18 15.3-2 Summary of Results for Locked Rotor/Shaft Break Transient 15.3-20 15.3-3 Deleted 15.3-21 15.3-3a Deleted 15.3-22 15.3-4 Input Parameters for the Locked Rotor Accident Radiological Consequence Analysis Using AST 15.3-23 15.4-1 Time Sequence of Events for Incidents Which Cause Reactivity and Power Distribution Anomalies 15.4-46 15.4-2 Deleted 15.4-49 15.4-3 Parameters Used in the Analysis of the Rod Cluster Control Assembly Ejection Accident 15.4-50 15.4-4 Input Parameters for the CREA Radiological Consequences Analysis Using AST 15.4-52 15.4-5 Deleted 15.4-55 15.4-6 Flux Map Review Criteria 15.4-56 15.5-1 Time Sequence of Events for Increase in Reactor Coolant Inventory Events 15.5-9 15.6-1 Time Sequence of Events for Inadvertent Opening of a Pressurizer Safety Valve 15.6-29 15.6-1a Time Sequence of Events for Large Break LOCA Analysis, Limiting PCT Case 15.6-30 15.6-1b Time Sequence of Events for Small Break LOCA Analysis, Unit 1, Nominal Tavg 15.6-31 15.6-1c Time Sequence of Events for Small Break LOCA Analysis, Unit 1, Reduced Tavg 15.6-31a 15.6-1d Time Sequence of Events for Small Break LOCA Analysis, Unit 2, Nominal Tavg 15.6-31b 15.6-1e Time Sequence of Events for Small Break LOCA Analysis, Unit 2, Reduced Tavg 15.6-31c 15.6-2 Input Parameters Used in the SBLOCA Analysis 15.6-32 15.6-2a Key LBLOCA Parameters and Initial Transient Assumptions 15.6-32b 15.6-3 LBLOCA Confirmatory Cases PCT Results Summary 15.6-33 15.6-3a Overall Large Break LOCA Results for the Byron/Braidwood Nuclear Stations 15.6-33b 15.6-3b Plant Operating Range Allowed by the Best-Estimate Large Break LOCA Analysis (Byron/Braidwood Nuclear Stations) 15.6-33c 15.6-3c Deleted 15.6-33f 15.6-3d Total Minimum Injected Safety Injection Flow Used in Best-Estimate Large-Break LOCA Analysis for Byron/Braidwood 15.6-33g 15.6-3e Deleted 15.6-33h 15.6-3f Deleted 15.6-33i 15.0-ix REVISION 15 - DECEMBER 2014
B/B-UFSAR LIST OF TABLES (Cont'd)
NUMBER TITLE PAGE 15.6-4 Small Break LOCA Results for ZIRLO Fuel Cladding - NOTRUMP EM-Unit 1 Nominal Tavg 15.6-34 15.6-4a Small Break LOCA Results for ZIRLO Fuel Cladding - NOTRUMP EM-Unit 1 Reduced Tavg 15.6-34a 15.6-4b Small Break LOCA Results for ZIRLO Fuel Cladding - NOTRUMP EM-Unit 2 Nominal Tavg 15.6-34b 15.6-4c Small Break LOCA Results for ZIRLO Fuel Cladding - NOTRUMP EM-Unit 2 Reduced Tavg 15.6-34c 15.6-4d Small Break LOCA Results for Zircaloy-4 Fuel Cladding - NOTRUMP EM-Unit 2 Reduced Tavg 15.6-34d 15.6-5 SGTR Initial Conditions 15.6-35 15.6-5a SGTR Operator Actions 15.6-36 15.6-6 Input Parameters for the SGTR Radiological Consequence Analysis Using AST 15.6-37 15.6-6a Sequence of Events for SGTR (Typical Offsite Dose Case) 15.6-37a 15.6-6b Sequence of Events for SGTR (Margin to Overfill Case) 15.6-37b 15.6-7 Input Parameters for the LOCA Radiological Consequence Analysis Using AST 15.6-38 15.6-8 Deleted 15.6-40 15.6-9 Deleted 15.6-41 15.0-ix (Cont'd) REVISION 12 - DECEMBER 2008
B/B-UFSAR LIST OF TABLES (Cont'd)
NUMBER TITLE PAGE 15.6-10 Deleted 15.6-43 15.6-11 Deleted 15.6-44 15.6-12 Deleted 15.6-45 15.6-13 Maximum Recirculation Loop Leakage External to Containment 15.6-46 15.6-14 Activity Releases to Atmosphere from Recirculation Loop Leakage Following a LOCA 15.6-47 15.6-15 Peak Clad Temperature Including All Penalties and Benefits 15.6-48 15.7-1 Parameters Used in Waste Gas Decay Tank Rupture Analyses (using TID-14844) 15.7-16 15.7-2 Waste Gas Decay Tank Inventory (One Unit)
(using TID-14844) 15.7-17 15.7-3 Spent Resin Storage Tank Inventory (using TID-14844) 15.7-18 15.7-4 Deleted 15.7-19 15.7-5 Deleted 15.7-20 15.7-6 Bounding Isotopic Core Inventory 15.7-21 15.7-7 Input Parameters for the FHA Radiological Consequence Analysis Using AST 15.7-22 15.7-8 Deleted 15.7-24 15A-1 Physical Data for Isotopes 15A-4 15.0-x REVISION 15 - DECEMBER 2014
B/B-UFSAR LIST OF FIGURES NUMBER TITLE 15.0-1 Illustration of Overtemperature and Overpower T Protection 15.0-2 Moderator Temperature Coefficient Used in Analysis 15.0-3 Doppler Power Coefficient Used in Accident Analysis 15.0-4 RCCA Position Versus Time to Dashpot 15.0-5 Normalized Rod Worth Versus Percent Inserted 15.0-6 Normalized RCCA Bank Reactivity Worth vs. Normalized Drop Time 15.0-7 Symbols Used in Accident Sequences 15.0-8 Excessive Heat Removal Due to Feedwater System Malfunctions 15.0-9 Excessive Load Increase 15.0-10 Depressurization of Main Steam System 15.0-11 Loss of External Load 15.0-12 Loss of Offsite Power to Station Auxiliaries 15.0-13 Loss of Normal Feedwater 15.0-14 Major Break of a Main Feedwater Line 15.0-15a Partial and Complete Loss of Forced Reactor Coolant Flow 15.0-15b Reactor Coolant Pump Shaft Seizure (Locked Rotor) 15.0-16 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal 15.0-17 Dropped Rod Cluster Control Assembly 15.0-18 Single Rod Cluster Control Assembly Withdrawal at Full Power 15.0-19 Deleted 15.0-20 Boron Dilution 15.0-21 Rupture of Control Rod Drive Mechanism Housing 15.0-22 Inadvertent ECCS Operation at Power 15.0-23 Accidental Depressurization of Reactor Coolant System 15.0-24 Steam Generator Tube Rupture 15.0-25a Small Break Loss-of-Coolant Accident 15.0-25b Large Break Loss-of-Coolant Accident 15.1-1a Nuclear Power Transient Pressurizer Pressure Transient for Feedwater Temperature Reduction 15.1-1b Reactor Coolant Loop Delta T Transient Core Average Temperature Transient and DNBR Transient for Feedwater Temperature Reduction 15.1-2a Nuclear Power Transient Pressurizer Pressure Transient for Feedwater Control Valve Malfunction 15.1-2b Reactor Coolant Loop Delta T Transient Core Average Temperature Transient and DNBR Transient for Feedwater Control Valve Malfunction 15.1-3 Ten Percent Step Load Increase, Minimum Moderator Feedback, Manual Reactor Control 15.1-4 Ten Percent Step Load Increase, Minimum Reactivity Feedback, Manual Reactor Control 15.0-xi REVISION 11 - DECEMBER 2006
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.1-5 Ten Percent Step Load Increase, Maximum Reactivity Feedback, Manual Reactor Control 15.1-6 Ten Percent Step Load Increase, Maximum Reactivity Feedback, Manual Reactor Control 15.1-7 Ten Percent Step Load Increase, Minimum Reactivity Feedback, Automatic Reactor Control 15.1-8 Ten Percent Step Load Increase, Minimum Reactivity Feedback, Automatic Reactor Control 15.0-xi (cont'd) REVISION 7 - DECEMBER 1998
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.1-9 Ten Percent Step Load Increase, Maximum Reactivity Feedback, Automatic Reactor Control 15.1-10 Ten Percent Step Load Increase, Maximum Reactivity Feedback, Automatic Reactor Control 15.1-11 Keff vs. Coolant Average Temperature 15.1-11a Deleted 15.1-12 Deleted 15.1-13 Deleted 15.1-14 Doppler Power Feedback 15.1-14a Deleted 15.1-15 Deleted 15.1-16 Deleted 15.1-17 Deleted 15.1-18 1.4 ft2 Steamline Break, Offsite Power Available, Unit 22 15.1-19 1.4 ft Steamline Break, Offsite Power Available, Unit 22 15.1-20 1.4 ft Steamline Break, Offsite Power Available, Unit 2 15.1-21 Deleted 15.1-22 Deleted 15.1-23 Deleted 15.1-24 Deleted 15.1-25 Deleted 15.1-26 Deleted 15.1-27 Steam System Piping Failure at Full Power - 0.95 Ft2 Break, Unit 1 15.1-28 Steam System Piping Failure at Full Power - 0.95 Ft2 Break, Unit 1 15.1-29 Steam System Piping Failure at Full Power - 0.95 Ft2 Break, Unit 1 15.1-30 Deleted 15.1-31 Deleted 15.1-32 Deleted 15.2-1 Loss of Load/Turbine Trip RCS Pressure Case, Unit 1 15.2-1a Loss of Load/Turbine Trip MSS Pressure Case, Unit 1 15.2-2 Loss of Load/Turbine Trip RCS Pressure Case, Unit 1 15.2-2a Loss of Load/Turbine Trip MSS Pressure Case, Unit 1 15.2-3 Loss of Load/Turbine Trip DNB Case, Unit 1 15.2-4 Loss of Load/Turbine Trip DNB Case, Unit 1 15.2-5 Loss of Load/Turbine Trip RCS Pressure Case, Unit 2 15.2-5a Loss of Load/Turbine Trip MSS Pressure Case, Unit 2 15.2-6 Loss of Load/Turbine Trip RCS Pressure Case, Unit 2 15.2-6a Loss of Load/Turbine Trip MSS Pressure Case, Unit 2 15.2-7 Loss of Load/Turbine Trip DNB Case, Unit 2 15.2-8 Loss of Load/Turbine Trip DNB Case, Unit 2 15.2-9 Deleted 15.2-9a Pressurizer Pressure and Water Volume Transients for Loss of Offsite Power (Byron) 15.2-9a Pressurizer Pressure and Water Volume Transients for Loss of Offsite Power (Braidwood) 15.2-10 Deleted 15.2-10a Loop Temperatures and Steam Generator Pressure for Loss of Offsite Power (Byron) 15.2-10a Loop Temperatures and Steam Generator Pressure for Loss of Offsite Power (Braidwood) 15.0-xii REVISION 16 - DECEMBER 2016
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.2-11 Pressurizer Pressure and Water Volume Transients for Loss of Normal Feedwater 15.2-12 Loop Temperatures and Steam Generator Pressure for Loss of Normal Feedwater 15.0-xii (Cont'd) REVISION 15 - DECEMBER 2014
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.2-13 Nuclear Power, Total Core Reactivity, and Feedline Break Flow Transients for Main Feedline Break With Offsite Power Available 15.2-14 Pressurizer Pressure, Pressurizer Relief, and Pressurizer Water Volume Transients for Main Feedline Break With Offsite Power Available 15.2-15 Faulted and Intact Loop Coolant Temperature Transients for Main Feedline Break With Offsite Power Available 15.2-16 Core Heat Flux and Steam Generator Pressure Transients for Main Feedline Break With Offsite Power Available 15.2-17 Nuclear Power, Total Core Reactivity, and Feedline Break Transients for Main Feedline Break Without Offsite Power Available 15.2-18 Pressurizer Pressure, Pressurizer Relief and Pressurizer Water Volume Transients for Main Feedline Break Without Offsite Power Available 15.2-19 Faulted and Intact Loop Coolant Temperature Transients for Main Feedline Break Without Offsite Power Available 15.2-20 Core Heat Flux and Steam Generator Pressure Transients for Main Feedline Break Without Offsite Power Available 15.3-1 Partial Loss of Forced Reactor Coolant Flow 15.3-2 Partial Loss of Forced Reactor Coolant Flow 15.3-3 Complete Loss of Forced Reactor Coolant Flow 15.3-4 Complete Loss of Forced Reactor Coolant Flow 15.3-5 Single Reactor Coolant Pump Locked Rotor 15.3-6 Single Reactor Coolant Pump Locked Rotor 15.0-xiii REVISION 7 - DECEMBER 1998
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.4-1 Neutron Flux Transients 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 Transients for Uncontrolled Rod Withdrawal from a Subcritical Condition 15.4-4 Uncontrolled RCCA Bank Withdrawal from Full Power with Minimum Reactivity Feedback (80 PCM/SEC Withdrawal Rate) 15.4-5 Uncontrolled RCCA Bank Withdrawal from Full Power With Minimum Reactivity Feedback (80 PCM/SEC Withdrawal Rate) 15.4-6 Uncontrolled RCCA Bank Withdrawal from Full Power With Minimum Reactivity Feedback (3.0 PCM/SEC Withdrawal Rate) 15.4-7 Uncontrolled RCCA Bank Withdrawal from Full Power With Minimum Reactivity Feedback (3.0 PCM/SEC Withdrawal Rate) 15.0-xiv REVISION 15 - DECEMBER 2014
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.4-8 Uncontrolled RCCA Bank Withdrawal from 100% Power; Minimum DNBR vs. Reactivity Insertion Rate 15.4-9 Uncontrolled RCCA Bank Withdrawal from 60% Power; Minimum DNBR vs. Reactivity Insertion Rate 15.4-10 Uncontrolled RCCA Bank Withdrawal from 10% Power; Minimum DNBR vs. Reactivity Insertion Rate 15.4-11 (Deleted) 15.4-12 (Deleted) 15.4-12a Typical Transient Response to a Dropped RCCA (or RCCAs) in Automatic Control 15.4-13 (Deleted) 15.4-14 (Deleted) 15.4-15 (Deleted) 15.4-16 (Deleted) 15.4-17 (Deleted) 15.4-18 Nuclear Power Transient, BOL HFP Rod Ejection Accident 15.4-19 Hot Spot Fuel and Clad Temperature vs. Time, BOL HFP Rod Ejection Accident 15.4-20 Nuclear Power Transient, EOL HZP Rod Ejection Accident 15.4-21 Hot Spot Fuel and Clad Temperatures vs. Time, EOL HZP Rod Ejection Accident 15.5-1 Nuclear Power Transient Core Average Water Temperature Transient for Inadvertent Actuation of ECCS During Power Operation 15.5-2 Pressurizer Pressure Transient Pressurizer Water Volume Transient for Inadvertent Operation of ECCS During Power Operation 15.5-3 DNBR and Steam Flow Transients for Inadvertent Operation of ECCS During Power Operation 15.6-1 Inadvertent Opening of a Pressurizer Safety Valve 15.6-2 Inadvertent Opening of a Pressurizer Safety Valve 15.6-3 Sequence of Events for Large Break LOCA Analysis 15.6-3a Pressurizer and Steam Generator Pressure with Ruptured Tube Flow Rates SGTR Results (Offsite Dose Case)- Unit 1 15.6-3b Steam Generator Release Rates and Mass Retained SGTR Results (Offsite Dose Case) - Unit 1 15.0-xv REVISION 15 - DECEMBER 2014
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.6-3c Pressurizer and Steam Generator Pressure and Ruptured Tube Flow Rate SGTR Results (Offsite Dose Case) -
Unit 2 15.6-3d Steam Generator Release Rates and Mass Retained SGTR Results (Offsite Dose Case) - Unit 2 15.6-3e Pressurizer and Steam Generator Pressure and Ruptured Tube Flow Rate SGTR Results (Margin to Overfill Case)- Unit 1 15.6-3f Steam Generator Release Rates and Mass Retained SGTR Results (Margin to Overfill Case)- Unit 1 15.6-3g Pressurizer and Steam Generator Pressure and Ruptured Tube Flow Rate SGTR Results (Margin to Overfill Case)
- Unit 2 15.6-3h Steam Generator Release Rates and Mass Retained SGTR Results (Margin to Overfill Case) - Unit 2 15.6-3i Break Flow Flashing Fraction SGTR Results (Offsite Dose Case) - Unit 1 15.6-3j Break Flow Flashing Fraction SGTR Results (Offsite Dose Case) - Unit 2 15.6-4 Deleted 15.6-5 Code Interface Description for Small Break Model 15.0-xv (Cont'd) REVISION 15 - DECEMBER 2014
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.6-6 Deleted 15.6-6a Deleted 15.6-7 PBOT/PMID Operating Limits 15.6-8a Hot Rod Peak Cladding Temperature for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8b Break Flow on Vessel Side of Broken Cold Leg for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8c Break Flow on Loop Side of Broken Cold Leg for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8d Void Fraction at the Intact and Broken Loop Pump Inlet for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8e Vapor Flow Rate at Top of Core Average Channel During Blowdown for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8f Vapor Flow Rate at Top of Hot Assembly Channel During Blowdown for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8g Collapsed Liquid Level in Lower Plenum for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8h Accumulator Mass Flow Rate for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8i Total SI Mass Flow Rate for One Intact Loop for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8j Collapsed Liquid Level in Core Average Channel for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8k Collapsed Liquid Level in Downcomer Channel for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8l Vessel Fluid Mass for ECCS LBLOCA PCT Limiting Case -
Unit 1 15.6-8m Hot Rod Peak Cladding Temperature Location for ECCS LBLOCA PCT Limiting Case - Unit 1 15.6-8n Deleted 15.6-8o Deleted 15.6-8p Deleted 15.6-9a Hot Rod Peak Cladding Temperature for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9b Break Flow on Vessel Side of Broken Loop for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9c Break Flow on Loop Side of Broken Cold Leg for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9d Void Fraction at the Intact and Broken Loop Pump Inlet for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9e Vapor Flow Rate at Top of Core Average Channel During Blowdown for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9f Vapor Flow Rate at Top of Hot Assembly Channel During Blowdown for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9g Collapsed Liquid Level in Lower Plenum for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9h Accumulator Mass Flow Rate for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9i Total SI Mass Flow Rate for One Intact Loop for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9j Collapsed Liquid Level in Core Average Channel for ECCS LBLOCA PCT Limiting Case - Unit 2 15.0-xvi REVISION 14 - DECEMBER 2012
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.6-9k Collapsed Liquid Level in Downcomer Channel for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9l Vessel Fluid Mass for ECCS LBLOCA Reference Transient -
Unit 2 15.6-9m Hot Rod Peak Cladding Temperature Location for ECCS LBLOCA PCT Limiting Case - Unit 2 15.6-9n Deleted 15.6-9o Deleted 15.6-9p Deleted 15.6-10 Deleted Figures 15.6-10a through 15.6-10p 15.6-11 Deleted Figures 15.6-11a through 15.6-11p 15.6-12 Deleted Figures 15.6-12a through 15.6-12p 15.6-13a Intact Loop Safety Injection Flow Rates Used in Small Break 15.6-13b Broken Loop Safety Injection Flow Rates Used in Small Break 15.6-13c Core Power Shape Used in Small Break 15.6-13d Recirculation Phase Intact Loop ECCS Flow Rates Used in Small Break LOCA 15.6-13e Recirculation Phase Broken Loop ECCS Flow Rates Used in Small Break LOCA 15.6-14a Peak Clad Temperature - Unit 1 (1.5 Inch, Reduced Tavg) 15.6-14b Fluid Temperature - Unit 1 (1.5 Inch, Reduced Tavg) 15.6-14c Heat Transfer Coefficient - Unit 1 (1.5 Inch, Reduced Tavg) 15.6-14d RCS Pressure - Unit 1 (1.5 Inch, Reduced Tavg) 15.6-14e Core Mixture Level - Unit 1 (1.5 Inch, Reduced Tavg) 15.6-14f Core Steam Flow Rate - Unit 1 (1.5 Inch, Reduced Tavg) 15.6-14g Break Flow - Unit 1 (1.5 Inch, Reduced Tavg) 15.6-14h Peak Clad Temperature - Unit 1 (1.5 Inch, Nominal Tavg) 15.6-14i Fluid Temperature - Unit 1 (1.5 Inch, Nominal Tavg) 15.6-14j Heat Transfer Coefficient - Unit 1 (1.5 Inch, Nominal Tavg) 15.6-14k RCS Pressure - Unit 1 (1.5 Inch, Nominal Tavg) 15.6-14l Core Mixture Level - Unit 1 (1.5 Inch, Nominal Tavg) 15.6-14m Core Steam Flow Rate - Unit 1 (1.5 Inch, Nominal Tavg) 15.6-14n Break Flow - Unit 1 (1.5 Inch, Nominal Tavg) 15.6-15a Peak Clad Temperature - Unit 1 (2 Inch, Reduced Tavg) 15.6-15b Fluid Temperature - Unit 1 (2 Inch, Reduced Tavg) 15.6-15c Heat Transfer Coefficient - Unit 1 (2 Inch, Reduced Tavg) 15.6-15d RCS Pressure - Unit 1 (2 Inch, Reduced Tavg) 15.6-15e Core Mixture Level - Unit 1 (2 Inch Reduced Tavg) 15.6-15f Core Steam Flow Rate - Unit 1 (2 Inch, Reduced Tavg) 15.6-15g Break Flow - Unit 1 (2 Inch, Reduced Tavg) 15.6-15h Peak Clad Temperature - Unit 1 (2 Inch, Nominal Tavg) 15.6-15i Fluid Temperature - Unit 1 (2 Inch, Nominal Tavg) 15.6-15j Heat Transfer Coefficient - Unit 1 (2 Inch, Nominal Tavg) 15.6-15k RCS Pressure- Unit 1 (2 Inch, Nominal Tavg) 15.6-15l Core Mixture Level - Unit 1 (2 Inch, Nominal Tavg) 15.6-15m Core Steam Flow Rate - Unit 1 (2 Inch, Nominal Tavg) 15.6-15n Break Flow - Unit 1 (2 Inch, Nominal Tavg) 15.6-16a Peak Clad Temperature - Unit 1 (3 Inch, Reduced Tavg) 15.6-16b Fluid Temperature - Unit 1 (3 Inch, Reduced Tavg) 15.6-16c Heat Transfer Coefficient - Unit 1 (3 Inch, Reduced Tavg) 15.0-xvii REVISION 14 - DECEMBER 2012
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.6-16d RCS Pressure - Unit 1 (3 Inch, Reduced Tavg) 15.6-16e Core Mixture Level - Unit 1 (3 Inch, Reduced Tavg) 15.6-16f Core Steam Flow Rate - Unit 1 (3 Inch, Reduced Tavg) 15.6-16g Break Flow - Unit 1 (3 Inch, Reduced Tavg) 15.6-16h Peak Clad Temperature - Unit 1 (3 Inch, Nominal Tavg) 15.6-16i Fluid Temperature - Unit 1 (3 Inch, Nominal Tavg) 15.6-16j Heat Transfer Coefficient - Unit 1 (3 Inch, Nominal Tavg) 15.6-16k RCS Pressure - Unit 1 (3 Inch, Nominal Tavg) 15.6-16l Core Mixture Level - Unit 1 (3 Inch, Nominal Tavg) 15.6-16m Core Steam Flow Rate - Unit 1 (3 Inch, Nominal Tavg) 15.6-16n Break Flow - Unit 1 (3 Inch, Nominal Tavg) 15.6-17a Peak Clad Temperature - Unit 1 (4 Inch, Reduced Tavg) 15.6-17b Fluid Temperature - Unit 1 (4 Inch, Reduced Tavg) 15.6-17c Heat Transfer Coefficient - Unit 1 (4 Inch, Reduced Tavg) 15.0-xviia REVISION 9 - DECEMBER 2002
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.6-17d RCS Pressure - Unit 1 (4 Inch, Reduced Tavg) 15.6-17e Core Mixture Level - Unit 1 (4 Inch, Reduced Tavg) 15.6-17f Core Steam Flow Rate - Unit 1 (4 Inch, Reduced Tavg) 15.6-17g Break Flow - Unit 1 (4 Inch, Reduced Tavg) 15.6-17h Peak Clad Temperature - Unit 1 (4 Inch, Nominal Tavg) 15.6-17i Fluid Temperature - Unit 1 (4 Inch, Nominal Tavg) 15.6-17j Heat Transfer Coefficient - Unit 1 (4 Inch, Nominal Tavg) 15.6-17k RCS Pressure - Unit 1 (4 Inch, Nominal Tavg) 15.6-17l Core Mixture Level - Unit 1 (4 Inch, Nominal Tavg) 15.6-17m Core Steam Flow Rate - Unit 1 (4 Inch, Nominal Tavg) 15.6-17n Break Flow - Unit 1 (4 Inch, Nominal Tavg) 15.6-18a Peak Clad Temperature - Unit 2 (1.5 Inch, Reduced Tavg) 15.6-18b Fluid Temperature - Unit 2 (1.5 Inch, Reduced Tavg) 15.6-18c Heat Transfer Coefficient - Unit 2 (1.5 Inch, Reduced Tavg) 15.6-18d RCS Pressure - Unit 2 (1.5 Inch, Reduced Tavg) 15.6-18e Core Mixture Level - Unit 2 (1.5 Inch, Reduced Tavg) 15.6-18f Core Steam Flow Rate - Unit 2 (1.5 Inch, Reduced Tavg) 15.6-18g Break Flow - Unit 2 (1.5 Inch, Reduced Tavg) 15.6-18h Peak Clad Temperature - Unit 2 (1.5 Inch, Nominal Tavg) 15.6-18i Fluid Temperature - Unit 2 (1.5 Inch, Nominal Tavg) 15.6-18j Heat Transfer Coefficient - Unit 2 (1.5 Inch, Nominal Tavg) 15.6-18k RCS Pressure - Unit 2 (1.5 Inch, Nominal Tavg) 15.6-18l Core Mixture Level - Unit 2 (1.5 Inch, Nominal Tavg) 15.6-18m Core Steam Flow Rate - Unit 2 (1.5 Inch, Nominal Tavg) 15.6-18n Break Flow - Unit 2 (1.5 Inch, Nominal Tavg) 15.6-19a Peak Clad Temperature - Unit 2 (2 Inch, Reduced Tavg) 15.6-19b Fluid Temperature - Unit 2 (2 Inch, Reduced Tavg) 15.6-19c Heat Transfer Coefficient - Unit 2 (2 Inch, Reduced Tavg) 15.6-19d RCS Pressure - Unit 2 (2 Inch, Reduced Tavg) 15.6-19e Core Mixture Level - Unit 2 (2 Inch, Reduced Tavg) 15.6-19f Core Steam Flow Rate - Unit 2 (2 Inch, Reduced Tavg) 15.6-19g Break Flow - Unit 2 (2 Inch, Reduced Tavg) 15.6-19h Peak Clad Temperature - Unit 2 (2 Inch, Nominal Tavg) 15.6-19i Fluid Temperature - Unit 2 (2 Inch, Nominal Tavg) 15.6-19j Heat Transfer Coefficient - Unit 2 (2 Inch, Nominal Tavg) 15.6-19k RCS Pressure - Unit 2 (2 Inch, Nominal Tavg) 15.6-19l Core Mixture Level - Unit 2 (2 Inch, Nominal Tavg) 15.6-19m Core Steam Flow Rate - Unit 2 (2 Inch, Nominal Tavg) 15.6-19n Break Flow - Unit 2 (2 Inch, Nominal Tavg) 15.6-20a Peak Clad Temperature - Unit 2 (3 Inch, Reduced Tavg) 15.6-20b Fluid Temperature - Unit 2 (3 Inch, Reduced Tavg) 15.6-20c Heat Transfer Coefficient - Unit 2 (3 Inch, Reduced Tavg) 15.6-20d RCS Pressure - Unit 2 (3 Inch, Reduced Tavg) 15.6-20e Core Mixture Level - Unit 2 (3 Inch, Reduced Tavg) 15.6-20f Core Steam Flow Rate - Unit 2 (3 Inch, Reduced Tavg) 15.0-xviii REVISION 9 - DECEMBER 2002
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.6-20g Break Flow - Unit 2 (3 Inch, Reduced Tavg) 15.6-20h Peak Clad Temperature - Unit 2 (3 Inch, Nominal Tavg) 15.6-20i Fluid Temperature - Unit 2 (3 Inch, Nominal Tavg) 15.6-20j Heat Transfer Coefficient - Unit 2 (3 Inch, Nominal Tavg) 15.6-20k RCS Pressure - Unit 2 (3 Inch, Nominal Tavg) 15.6-20l Core Mixture Level - Unit 2 (3 Inch, Nominal Tavg) 15.6-20m Core Steam Flow Rate - Unit 2 (3 Inch, Nominal Tavg) 15.6-20n Break Flow - Unit 2 (3 Inch, Nominal Tavg) 15.0-xviiia REVISION 9 - DECEMBER 2002
B/B-UFSAR LIST OF FIGURES (Cont'd)
NUMBER TITLE 15.6-21a Peak Clad Temperature - Unit 2 (4 Inch, Reduced Tavg) 15.6-21b Fluid Temperature - Unit 2 (4 Inch, Reduced Tavg) 15.6-21c Heat Transfer Coefficient - Unit 2 (4 Inch, Reduced Tavg) 15.6-21d RCS Pressure - Unit 2 (4 Inch, Reduced Tavg) 15.6-21e Core Mixture Level - Unit 2 (4 Inch, Reduced Tavg) 15.6-21f Core Steam Flow Rate - Unit 2 (4 Inch, Reduced Tavg) 15.6-21g Break Flow - Unit 2 (4 Inch, Reduced Tavg) 15.6-21h Peak Clad Temperature - Unit 2 (4 Inch, Nominal Tavg) 15.6-21i Fluid Temperature - Unit 2 (4 Inch, Nominal Tavg) 15.6-21j Heat Transfer Coefficient - Unit 2 (4 Inch, Nominal Tavg) 15.6-21k RCS Pressure - Unit 2 (4 Inch, Nominal Tavg) 15.6-21l Core Mixture Level - Unit 2 (4 Inch, Nominal Tavg) 15.6-21m Core Steam Flow Rate - Unit 2 (4 Inch, Nominal Tavg) 15.6-21n Break Flow - Unit 2 (4 Inch, Nominal Tavg) 15.7-1 Spent Fuel Storage Pool Plan 15.7-2 Spent Fuel Storage Pool DRAWINGS CITED IN THIS CHAPTER*
- The listed drawings are included as General References only; i.e., refer to the drawings to obtain additional detail or to obtain background information. These drawings are not part of the UFSAR. They are controlled by the Controlled Documents Program.
DRAWING* SUBJECT 108D685-15 Auxiliary Feedwater Pumps Startup Diagram 15.0-xix REVISION 6 - DECEMBER 1996
B/B-UFSAR CHAPTER 15.0 - ACCIDENT ANALYSES 15.0 ACCIDENT ANALYSES This chapter addresses the representative initiating events listed on pages 15-10, 15-11, and 15-12 of Regulatory Guide 1.70, as they apply.
Certain items in the guide warrant comment, as follows:
Items 1.3 and 2.1 - There are no pressure regulators in the nuclear steam supply system (NSSS) pressurized water reactor (PWR) design whose malfunction or failure could cause a steam flow transient.
In compliance with the requirements of Regulatory Guide 1.70, a failure mode and effects analysis (FMEA) has been provided for each safety system needed to mitigate the consequences of the accidents analyzed in Chapter 15.0. A FMEA for the emergency core cooling system is provided in Section 6.3.2.5; for the residual heat removal system in Subsection 5.4.7.2.5; for the engineered safety features actuation system in Subsection 7.3.2; for the control rod drive system in Subsection 4.6.2; for the reactor trip system in Subsection 7.2.2.1; for the chemical and volume control system in Subsection 9.3.4.1.3.9; for the containment spray system in Subsection 6.5.2.2; and, for the auxiliary feedwater system in Subsection 10.4.9.3.
Once actuated, the equipment of a safety system operates in the same manner regardless of initiating accident. However, performance depends on the type of accident that is being mitigated.
15.0.1 Classification of Plant Conditions Since 1970, the American Nuclear Society (ANS) classification of plant conditions has been used which divides plant conditions into four categories in accordance with anticipated frequency of occurrence and potential radiological consequences to the public.
The four categories are as follows:
Condition I: Normal Operation and Operational Transients.
Condition II: Faults of Moderate Frequency.
Condition III: Infrequent Faults.
Condition IV: Limiting Faults.
The basic principle applied in relating design requirements to each of the conditions is that the most probable occurrences should yield the least radiological risk to the public and those extreme situations having the potential for the greatest risk to the public shall be those least likely to occur. Where applicable, reactor trip system and engineered safeguards functioning is assumed to the extent allowed by considerations 15.0-1 REVISION 4 - DECEMBER 1992
B/B-UFSAR such as the single failure criterion, in fulfilling this principle.
15.0.1.1 Condition I - Normal Operation and Operational Transients Condition I occurrences are those which are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant. As such, Condition I occurrences are accommodated with margin between any plant parameter and the value of that parameter which would require either automatic or manual protective action. Inasmuch as Condition I occurrences occur frequently or regularly, they must be considered from the point of view of affecting the consequences of fault conditions (Conditions II, III and IV).
In this regard, analysis of each fault condition described is generally based on a conservative set of initial conditions corresponding to adverse conditions which can occur during Condition I operation.
A typical list of Condition I events is listed below:
- a. Steady-state and shutdown operations
- 1. Power operation (>5 to 100% of rated thermal power),
- 2. Startup (keff > 0.99 to 5% of rated thermal power),
- 3. Hot standby (subcritical, residual heat removal system isolated),
- 4. Hot shutdown (subcritical, residual heat removal system in operation),
- 5. Cold shutdown (subcritical, residual heat removal system in operation), and
- 6. Refueling
- b. Operation with permissible deviations Various deviations which may occur during continued operation as permitted by the plant Technical Specifications must be considered in conjunction with other operational modes. These include:
- 1. Operation with components or systems out of service,
- 2. Radioactivity in the reactor coolant, due to leakage from fuel with cladding defects, 15.0-2 REVISION 5 - DECEMBER 1994
B/B-UFSAR (a) Fission products (b) Corrosion products (c) Tritium
- 3. Operation with steam generator leaks up to the maximum allowed by the Technical Specification 3.4.13, and
- 4. Testing as allowed by Technical Specifications and the Technical Requirements Manual (TRM)
- c. Operational transients
- 1. Plant heatup and cooldown (up to 100F/hour for the reactor coolant system; 200F/hour for the pressurizer during cooldown and 100F/hour for the pressurizer during heatup),
- 2. Step load changes (up to +/- 10%),
- 3. Ramp load changes (up to 5%/minute), and
- 4. Load rejection up to and including design full load rejection transient 15.0.1.2 Condition II - Faults of Moderate Frequency These faults, at worst, result in the reactor trip with the plant being capable of returning to operation. By definition, these faults (or events) do not propagate to cause a more serious fault, i.e., Condition III or IV events. In addition, Condition II events are not expected to result in fuel rod failures or reactor coolant system or secondary system overpressurization.
For the purposes of this report, the following faults are included in this category:
- a. Feedwater system malfunctions that result in a decrease in feedwater temperature (Subsection 15.1.1),
- c. Excessive increase in secondary steam flow (Subsection 15.1.3),
- d. Inadvertent opening of a steam generator relief or safety valve (Subsection 15.1.4),
15.0-3 REVISION 8 - DECEMBER 2000
B/B-UFSAR
- e. Loss of external electrical load (Subsection 15.2.2),
- f. Turbine trip (Subsection 15.2.3),
- g. Inadvertent closure of main steam isolation valves (Subsection 15.2.4),
- h. Loss of condenser vacuum and other events resulting in turbine trip (Subsection 15.2.5),
- i. Loss of nonemergency ac power to the station auxiliaries (Subsection 15.2.6),
- j. Loss of normal feedwater flow (Subsection 15.2.7),
- k. Partial loss of forced reactor coolant flow (Subsection 15.3.1),
- l. Uncontrolled rod cluster control assembly bank withdrawal at a subcritical or low power startup condition (Subsection 15.4.1),
- m. Uncontrolled rod cluster control assembly bank withdrawal at power (Subsection 15.4.2),
- n. Rod cluster control assembly misalignment (dropped full length assembly, dropped full length assembly bank, or statically misaligned full length assembly)
(Subsection 15.4.3),
- o. Deleted
- p. Chemical and volume control system malfunction that results in a decrease in the boron concentration in the reactor coolant (Subsection 15.4.6),
- q. Inadvertent operation of the emergency core cooling system during power operation (Subsection 15.5.1),
- r. Chemical and volume control system malfunction that increases reactor coolant inventory (Subsection 15.5.2),
- s. Inadvertent opening of a pressurizer safety or relief valve (Section 15.6.1), and
- t. Break in instrument line or other lines from reactor coolant pressure boundary that penetrate containment (Subsection 15.6.2).
15.0-4 REVISION 5 - DECEMBER 1994
B/B-UFSAR 15.0.1.3 Condition III - Infrequent Faults By definition Condition III occurrences are faults which may occur very infrequently during the life of the plant. They will be accommodated with the failure of only a small fraction of the fuel rods although sufficient fuel damage might occur to preclude resumption of the operation for a considerable outage time. The release of radioactivity will not be sufficient to interrupt or restrict public use of those areas beyond the exclusion radius.
A Condition III fault will not, by itself, generate a Condition IV fault or result in consequential loss of function of the reactor coolant system or containment barriers. For the purposes of this report the following faults are included in this category:
- a. Steam system piping failure from zero power and full power (minor) (Subsections 15.1.5 and 15.1.6),
- b. Complete loss of forced reactor coolant flow (Subsection 15.3.2),
- c. Rod cluster control assembly misalignment (single rod cluster control assembly withdrawal at full power)
(Subsection 15.4.3),
- d. Inadvertent loading and operation of a fuel assembly in an improper position (Subsection 15.4.7),
- e. Loss of coolant accidents resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary (small break) (Subsection 15.6.5),
- f. Gaseous radwaste system leak or failure (Subsection 15.7.1),
- g. Liquid radwaste system leak or failure (Subsection 15.7.2),
- h. Postulated radioactive releases due to liquid tank failures (Subsection 15.7.3), and
- i. Spent fuel cask drop accidents (Subsection 15.7.5).
15.0.1.4 Condition IV - Limiting Faults Condition IV occurrences are faults which are not expected to take place, but are postulated because their consequences would include the potential of the release of significant amounts of radioactive material. They are the most drastic which must be designed against and represent limiting design cases. Condition IV faults are not to cause a fission product release to the environment resulting in an undue risk to public health and 15.0-5 REVISION 7 - DECEMBER 1998
B/B-UFSAR safety in excess of guidelines values of 10 CFR 100 for TID-14844 based dose analyses and 10 CFR 50.67 for AST based analyses. A single Condition IV fault is not to cause a consequential loss of required functions of systems needed to cope with the fault including those of the emergency core cooling system and the containment. For the purposes of this report, the following faults have been classified in this category:
- a. Steam system piping failure from zero power and full power (major) (Subsections 15.1.5 and 15.1.6),
- b. Feedwater system pipe break (Subsection 15.2.8),
- c. Reactor coolant pump shaft seizure (locked rotor)
(Subsection 15.3.3),
- d. Reactor coolant pump shaft break (Subsection 15.3.4),
- e. Spectrum of rod cluster control assembly ejection accidents (Subsection 15.4.8),
- f. Steam generator tube failure (Subsection 15.6.3),
- g. Loss of coolant accidents resulting from the spectrum of postulated piping breaks within the reactor coolant pressure boundary (large break)
(Subsection 15.6.5), and
- h. Design basis fuel handling accidents (Subsection 15.7.4).
15.0.1.5 Summary of Results For all Condition II transients analyzed, the calculated minimum DNBR was greater than the limit value. For each of these transients, the peak RCS pressure was less than the safety limit of 110% of design pressure (2750 psia) and there was no failed fuel as a result of the transients. Since DNB does not occur for any Condition II transients, peak cladding temperature does not increase sufficiently above nominal values for these events.
For all of the applicable Condition III transients, the minimum DNBR was greater than the limit value and there was no failed fuel except for a single RCCA withdrawal at full power. For this transient, the upper bound of the number of fuel rods experiencing DNBR less than the limit value was 5% of the total rods in the core. All of the applicable Condition III transients experienced a peak RCS pressure less than 2750 psia. The only Condition III transient for which cladding temperature was calculated was the small LOCA and the peak value was less than 2200F.
15.0-6 REVISION 13 - DECEMBER 2010
B/B-UFSAR All the applicable Condition IV transients analyzed met the applicable condition IV acceptance criteria. For the locked rotor event, DNB was assumed to occur at the initiation of the transient and the peak cladding temperature was calculated to be less than the limit of 2375F (associated with Optimized ZIRLO' fuel cladding which bounds the 2700F limit for ZIRLO fuel cladding, Reference 21). For the LOCA the amount of failed fuel calculated was equal or less than 100% and for rod ejection it was equal or less than 10%. The locked rotor accident results in cladding failure in less than 2% of the fuel rods. There was no failed fuel predicted for the feed line break, the steam line break, or the steam generator tube rupture. All of the applicable Condition IV transients experienced a peak RCS pressure less than 2750 psia. The peak cladding temperature calculated for LOCA was less than 2200. The average fuel pellet enthalpy at the hot spot for rod ejection was less than 200 cal/gm.
15.0.2 Optimization of Control Systems A control system setpoint study is performed in order to simulate performance of the reactor control and protection systems. In this study, emphasis is placed on the development of a control system which will automatically maintain prescribed conditions in the plant even under a conservative set of reactivity parameters with respect to both system stability and transient performance.
For each mode of plant operation, a group of optimum controller setpoints is determined. In areas where the resultant setpoints are different, compromises based on the optimum overall performance are made and verified. A consistent set of control system parameters is derived satisfying plant operational requirements throughout the core life and for various levels of power operation.
The study comprises an analysis of the following control systems:
rod cluster control assembly, steam dump, steam generator, level, pressurizer pressure, and pressurizer level.
15.0.3 Plant Characteristics and Initial Conditions Assumed in the Accident Analyses Each of the four RCS loops are equipped with loop isolation valves. However, the stations are not currently licensed for less than all loops in operation. Therefore, the UFSAR presents the licensing basis for four-loop operation only.
15.0.3.1 Design Plant Conditions Table 15.0-1 lists the principal power rating values which are assumed in analyses performed in this report.
15.0-7 REVISION 16 - DECEMBER 2016
B/B-UFSAR Allowances for errors in the determination of the steady-state power level are made as described in Subsection 15.0.3.2. The thermal power values used for each transient analyzed are given in Table 15.0-2.
The values of other pertinent plant parameters utilized in the accident analyses are given in Tables 15.0-3 (Unit 1) and 15.0-4 (Unit 2).
15.0.3.2 Initial Conditions For most accidents which are DNB limited, nominal values of initial conditions are assumed. The allowances on power, temperature, and pressure are determined on a statistical basis and are included in the limit DNBR, as described in WCAP-11397 (Reference 10). This procedure is known as the "Revised Thermal Design Procedure," and is discussed more fully in Section 4.4.
For accidents which are not DNB limited, or in which the Revised Thermal Design Procedure is not employed, the initial conditions are obtained by adding the maximum steady state errors to rated values. The following conservative steady state errors were assumed in the analysis:
- a. Core power 0.345% allowance for calorimetric error 15.0-8 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- b. Average reactor 9.1 -7.6F allowance coolant system for controller deadband and temperature measurement error
- c. Pressurizer 43 pounds per square inch pressure (psi) allowance for steady state fluctuations and measurement error Table 15.0-2 summarizes initial conditions and computer codes used in the accident analysis, and shows which accidents employed a DNB analysis using the revised thermal design procedure.
15.0.3.3 Power Distribution The transient response of the reactor system is dependent on the initial power distribution. The nuclear design of the reactor core minimizes adverse power distribution through the placement of control rods and operating instructions. Power distribution may be characterized by the radial factor (FH) and the total peaking factor (FQ). The peaking factor limits are given in the technical specifications.
For transients which may be DNB limited, the radial peaking factor is of importance. The radial peaking factor increases with decreasing power level due to rod insertion. This increase in FH is included in the core limits illustrated in Figure 15.0-1. All transients that may be DNB limited are assumed to begin with a FH consistent with the initial power level defined in the technical specifications.
The axial power shape used in the DNB calculation is discussed in Section 4.4.
The radial and axial power distributions described above are input to the VIPRE Code as described in Section 4.4.
For transients which may be overpower limited, the total peaking factor (FQ) is of importance. All transients that may be overpower limited are assumed to begin with plant conditions including power distributions which are consistent with reactor operation as defined in the technical specifications.
For overpower transients which are slow with respect to the fuel rod thermal time constant, for example, the chemical volume control system malfunction that results in a decrease in the boron concentration in the reactor coolant incident which lasts many minutes, and the excessive increase in secondary steam flow incident which may reach equilibrium without causing a reactor trip, the fuel rod thermal evaluations are performed as discussed in Section 4.4. For overpower transients which are fast with respect to the fuel rod thermal time constant, 15.0-9 REVISION 15 - DECEMBER 2014
B/B-UFSAR for example, the uncontrolled rod cluster control assembly bank withdrawal from subcritical or low power startup and rod cluster control assembly ejection incidents which result in a large power rise over a few seconds, a detailed fuel heat transfer calculation must be performed. Although the fuel rod thermal time constant is a function of system conditions, fuel burnup and rod power, a typical value at beginning-of-life for high power rods is approximately 5 seconds.
15.0.4 Reactivity Coefficients Assumed in the Accident Analyses The transient response of the reactor system is dependent on reactivity feedback effects, in particular the moderator temperature coefficient and the Doppler power coefficient. These reactivity coefficients and their values are discussed in detail in Chapter 4.0.
In the analysis of certain events, conservatism requires the use of large reactivity coefficient values whereas in the analysis of other events, conservatism requires the use of small reactivity coefficient values. Some analyses such as loss of reactor coolant from cracks or breaks in the reactor coolant system do not depend on reactivity feedback effects. The values are given in Table 15.0-2. Figure 15.0-2 shows the Technical Specifications limit for the moderator temperature coefficient as a function of power level. Figure 15.0-3 shows the least negative (upper bound), least negative at End-of-Life, and most negative (lower bound) bound Doppler power coefficients as a function of power, used in the transient analysis. The justification for use of conservatively large versus small reactivity coefficient values are treated on an event-by-event basis. In some cases conservative combinations of parameters are used to bound the effects of core life, although these combinations may not represent possible realistic situations.
15.0.5 Rod Cluster Control Assembly Insertion Characteristics The negative reactivity insertion following a reactor trip is a function of the position versus time of the rod cluster control assemblies and the variation in rod worth as a function of rod position. With respect to accident analyses, the critical parameter is the time of insertion up to the dashpot entry or approximately 85% of the rod cluster travel.
The rod cluster control assembly position versus time assumed in accident analyses is shown in Figure 15.0-4. The rod cluster control assembly insertion time to dashpot entry is taken as 2.7 seconds. Drop time testing requirements are dependent on the type of cluster control assemblies actually used in the plant and are specified in the plant technical specifications.
Figure 15.0-5 shows the fraction of total negative reactivity insertion versus normalized rod position for a core where the axial distribution is skewed to the lower region of the core.
An axial distribution which is skewed to the lower region of 15.0-10 REVISION 15 - DECEMBER 2014
B/B-UFSAR the core can arise from an unbalanced xenon distribution. This curve is used to compute the negative reactivity insertion versus time following a reactor trip which is input to all point kinetics core models used in transient analyses. The bottom skewed power distribution itself is not input into the point kinetics core model.
There is inherent conservatism in the use of Figure 15.0-5 in that it is based on a skewed flux distribution which would exist relatively infrequently. For cases other than those associated with unbalanced xenon distributions, significant negative reactivity would have been inserted due to the more favorable axial distribution existing prior to trip.
The normalized rod cluster control assembly negative reactivity insertion versus time is shown in Figure 15.0-6. The curve shown in this figure was obtained from Figures 15.0-4 and 15.0-5. A total negative reactivity insertion following a trip of 4% k is assumed in the transient analyses except where specifically noted otherwise. This assumption is conservative with respect to the calculated trip reactivity worth available as shown in Table 4.3-3. For Figures 15.0-4 and 15.0-6, the rod cluster control assembly drop time is normalized to 2.7 seconds, unless otherwise noted for a particular event.
The normalized rod cluster control assembly negative reactivity insertion versus time curve for an axial power distribution skewed to the bottom (Figure 15.0-6) is used in those transient analyses for which a point kinetics core model is used. Where special analyses require use of three dimensional or axial one dimensional core models, the negative reactivity insertion resulting from the reactor trip is calculated directly by the reactor kinetics code and is not separable from the other reactivity feedback effects. In this case, the rod cluster control assembly position versus time of Figure 15.0-4 is used as code input.
15.0.6 Trip Points and Time Delays to Trip Assumed in Accident Analyses A reactor trip signal acts to open two trip breakers connected in series feeding power to the control rod drive mechanisms.
The loss of power to the mechanism coils causes the mechanisms to release the rod cluster control assemblies which then fall by gravity into the core. There are various instrumentation delays associated with each trip function, including delays in signal actuation, in opening the trip breakers, and in the release of the rods by the mechanisms. The total delay to trip is defined as the time delay from the time that trip conditions are reached to the time the rods are free and begin to fall.
Limiting trip setpoints assumed in accident analyses and the time delay assumed for each trip function are given in Table 15.0-5.
15.0-11 REVISION 9 - DECEMBER 2003
B/B-UFSAR Reference is made in that table to overtemperature and overpower T trips shown in Figure 15.0-1. This figure presents the allowable reactor coolant loop average temperature and T for the design flow and power distribution, as described in Section 4.4, as a function of primary coolant pressure. The boundaries of operation defined by the overpower T trip and the overtemperature T trip are represented as "protection lines" on this diagram. The protection lines are drawn to include all adverse instrumentation and setpoint errors so that under nominal conditions trip would occur well within the area bounded by these lines. The utility of this diagram is in the fact that the limit imposed by any given DNBR can be represented as a line. The DNB lines represent the locus of conditions for which the DNBR equals the limit value (1.37 for the thimble cell and the typical cell).
All points below and to the left of a DNB line for a given pressure have a DNBR greater than the limit value. The diagram shows that DNB is prevented for all cases if the area enclosed with the maximum protection lines is not traversed by the applicable DNBR line at any point.
The area of acceptable operation during transient conditions (power, pressure and temperature) is bounded by the combination of reactor trips: high neutron flux (fixed setpoint); high pressure (fixed setpoint); low pressure (fixed setpoint);
overpower and overtemperature T (variable setpoints).
A Tavg coastdown/power coastdown strategy may be used at end of cycle. The first part of the strategy consists of a Tavg coastdown from full power conditions to a specified reduced Tavg.
In this portion of the strategy, extended full power capability is achieved by the increase in core reactivity that occurs with the increased moderator density due to the reduced Tavg.
Following the Tavg coastdown, a power coastdown is then performed to extend plant operation at end of cycle. The Chapter 15 analyses support a constant full power Tavg window. An underlying operating strategy modeled in the safety analyses is that the OTT and OPT trip functions which are dependent on the full power Tavg are appropriately scaled and/or calibrated to the specified cycle-specific value.
To support end of cycle Tavg coastdown/power coastdown operating strategy, the events that are protected by the OTT and OPT reactor trips were reanalyzed or evaluated with a low Tavg without rescaling the OTT and OPT reactor trips. The analyses and evaluations concluded that these events continue to meet the acceptance criteria for the respective events.
The difference between the limiting trip point assumed for the analysis and the normal trip point represents an allowance for instrumentation channel error and setpoint error. Nominal trip setpoints are specified in the plant technical specifications.
During plant startup tests, it will be demonstrated that actual instrument time delays are equal to or less than the assumed values. Additionally, protection system channels are calibrated and instrument response times determined periodically in accordance with the technical specifications.
15.0-12 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.0.7 Instrumentation Drift and Calorimetric Errors - Power Range Neutron Flux The instrumentation drift and calorimetric errors used in establishing the power range high neutron flux setpoint are presented in Table 15.0-6. The calorimetric errors presented in Table 15.0-6 are applicable for the limiting condition in which the feedwater flow venturis are used in the determination of core thermal power. The calorimetric errors associated with the determination of core thermal power using the Leading Edge Flow Meters are bounded by the errors associated with the feedwater venturis.
The calorimetric error is the error assumed in the determination of core thermal power as obtained from secondary plant measurements. The total ion chamber current (sum of the top and bottom sections) is calibrated (set equal) to this measured power on a periodic basis.
The secondary power is obtained from measurement of feedwater flow, feedwater inlet temperature to the steam generators and steam pressure. High accuracy instrumentation is provided for 15.0-12a REVISION 15 - DECEMBER 2014
B/B-UFSAR these measurements with accuracy tolerances much tighter than those which would be required to control feedwater flow.
The nuclear instrumentation system is calibrated by comparing individual power level indications with the power obtained calorimetrically. The error assumed in the calorimetric determination of power level (for the purpose of establishing the maximum overpower trip setpoint) is 2% at the 95% confidence level, as given in Table 15.0-6.
An analysis has been prepared which took the effects of measurement errors and their confidence levels into account and provided the basis for the selection of protection system setpoints. The analysis was performed in accordance with NRC requirements.
15.0.8 Plant Systems and Components Available for Mitigation of Accident Effects The NSSS is designed to afford proper protection against the possible effects of natural phenomena, postulated environmental conditions and dynamic effects of postulated accidents. In addition, the design incorporates features which minimize the probability and effects of fires and explosions. Chapter 17.0 discusses the quality assurance program which has been implemented to assure that the NSSS will satisfactorily perform its assigned safety functions. The incorporation of these features in the NSSS, coupled with the reliability of the design, ensures that the normally operating systems and components listed in Table 15.0-7 will be available for mitigation of the events discussed in Chapter 15.0. Table 15.0-7 identifies plant systems and equipment credited for transients and accident conditions.
In determining which systems are necessary to mitigate the effects of these postulated events, the classification system of ANSI-N18.2-1973 is utilized. The design of "systems important to safety" (including protection systems) is consistent with IEEE Standard 379-1972 and Regulatory Guide 1.53 in the application of the single failure criterion.
In the analysis of the Chapter 15.0 events, control system action is considered only if that action results in more severe accident results. No credit is taken for control system operation if that operation mitigates the results of an accident.
For some accidents, the analysis is performed both with and without control system operation to determine the worst case.
The response times for the air-operated and motor-operated valves in the auxiliary feedwater and main steam systems are verified during preoperational testing.
15.0-13 REVISION 11 - DECEMBER 2006
B/B-UFSAR 15.0.9 Fission Product Inventories 15.0.9.1 Activities in the Core For the accidents evaluated using TID-14844, the calculation of the core iodine fission product inventory was modeled using the computer code ORIGEN2 (Reference 1) which is a versatile point-depletion and radioactive-decay code for use in simulating nuclear fuel cycles and calculating the nuclide compositions and characteristics of materials contained therein. This code takes into account the transmutation of all the isotopes in the fuel.
The core fission product inventories were determined for end-of-cycle conditions, assuming an equilibrium fuel cycle. These inventories are given in Table 15.0-8. The isotopes included in Table 15.0-8 are the controlling isotopes from considerations of thyroid dose (iodines) and from external dose due to immersion (noble gases).
The isotopic yields used in the calculations are from the data of NEDO-12154-1, utilizing the isotopic yield data for thermal fissioning of U-235 as the sole fissioning source. The change in fission product inventory resulting from the fissioning of other fissionable atoms has been reviewed. The results of this review indicated that inclusion of all fission source data would result in a small (less than 10%) change in the isotopic inventories.
For the accidents evaluated using Alternative Source Terms, the calculation of the core iodine fission product inventory was also modeled using computer code ORIGEN2 (Reference 1), based on reactor operation at 3586.6 MWth and an equilibrium 542.9 Effective Full Power Days (EFPD) eighteen month cycle design.
The maximum of the 100 EFPD and End of Cycle values for each isotope were selected to generate the bounding isotopic core inventory activity and composition results. The output from ORIGEN2 is adjusted to 3658.3 MWt such that the inputs to the accident dose analyses remain bounded for the MUR core licensed thermal power level including measurement uncertainties.
15.0.9.2 Activities in the Fuel Pellet Cladding Gap The fuel clad gap activities are defined differently for the different accidents which model gap activity releases. These accidents include the locked pump rotor, rod ejection, and fuel handling accidents. The specific gap activity model used for each of these events is described in the associated accident analysis discussion.
The NRC staff has completed its review of the revised Westinghouse fuel rod internal pressure design criteria and has decided on an acceptable amended criterion:
"The internal pressure of the lead fuel rod in the reactor will be limited to a value below that which could cause (1) the diametrical gap to increase due to outward cladding creep during steady-state operation, and (2) extensive DNB propagation to occur."
15.0-14 REVISION 15 - DECEMBER 2014
B/B-UFSAR WCAP-8963, "Safety Analysis for the Revised Fuel Internal Design Basis," was found to be acceptable to support the conclusion that an insignificant number of additional DNB events would occur during transients and accidents as a result of operating with fuel rod pressure (1) greater than nominal system pressure, and (2) limited by the above criterion.
For all Condition III and IV overpower events, the number of rods that are assumed to fail is less than 10%. Therefore, the analyses for the Byron/Braidwood OFA amendment (Amendment 30) are bounded by the analysis presented in the WCAP. The results presented in the WCAP are based on the detailed probability analysis performed to determine the maximum extent of core damage that could lead to DNB propagation. It was shown that the propagation mechanism causes only a small incremental increase in the percentage of rods in DNB. In view of the conservative nature of the failure propagation scheme and the small percentage increase in the number of failed rods, the potential increase in site release is inconsequential.
Although this effect, resulting from the revised fuel rod internal pressure design criterion, is small, it was factored into the number of rods predicted to fail.
15.0.9.3 Activities in the Primary Coolant The accident dose analyses consider the releases of radioactive iodines and noble gases. The noble gas activity in the primary coolant is based on operation with 1.0 percent fuel defects and with no credit for stripping of gases from the primary coolant for processing by the gaseous waste processing system. The primary coolant noble gas activities are provided in Subsection 11.1 and are also repeated here in Table 15.0-9.
While Subsection 11.1 identifies the iodine concentrations associated with operation with 1.0 percent fuel defects, the iodine activity in the primary coolant is controlled to an equilibrium concentration of no more than 1.0 µCi/g of dose equivalent (DE) I-131. The determination of DE I-131 is based on five iodine isotopes: I-131, I-132, I-133, I-134, and I-135. In the event of an accident involving depressurization of the primary coolant system or reactor trip, the rate at which iodine enters the primary system is assumed to increase. This is referred to as an accident-initiated or concurrent iodine spike.
There is also the potential for an accident to occur during the time in which an iodine spike has already occurred, initiated by changes in plant operation (e.g., power reduction). This is referred to as a pre-existing iodine spike. The maximum iodine concentration associated with a pre-existing iodine spike is 60
Table 15.0-10 lists the equilibrium iodine concentrations associated with 1.0 µCi/g of DE I-131, the equilibrium iodine appearance rates into the primary coolant, the iodine spike 15.0-15 REVISION 15 - DECEMBER 2014
B/B-UFSAR appearance rates (500 times equilibrium), and the iodine concentrations associated with 60 µCi/g of DE I-131.
15.0.9.4 Activities in the Secondary Coolant It is assumed that there is primary to secondary coolant leakage that transports activity to the secondary system. However, there is no significant amount of noble gas in the secondary coolant since non-condensable gases entering the secondary system are rapidly expelled to the environment. The iodine concentration in the secondary coolant is controlled to be less than 0.1 µCi/g of DE I-131 (values for individual isotopes are identified in Table 15.0-10).
15.0.10 Residual Decay Heat 15.0.10.1 Total Residual Heat Residual heat in a subcritical core is calculated for the loss of coolant accident per the requirements of Appendix K of 10 CFR 50.46 (Reference 3) as described in (References 4 and 5).
These requirements include assuming infinite irradiation time before the core goes subcritical to determine fission product decay energy. For all other accidents, unless otherwise noted in the text, the same models are used except that fission product decay energy is based on core average exposure at the end of the equilibrium cycle.
15.0.10.2 Distribution of Decay Heat Following Loss of Coolant Accident During a loss-of-coolant accident, the core is rapidly shut down by void formation or rod cluster control assembly insertion, or both, and a large fraction of the heat generation to be considered comes from fission product decay gamma rays.
This heat is not distributed in the same manner as steady state fission power. Local peaking effects which are important for the neutron dependent part of the heat generation do not apply to the gamma ray contribution. The steady state factor of 97.4% which represents the fraction of heat generated within the clad and pellet drops to 95% for the hot rod in a loss-of-coolant accident.
For example, in the transient resulting from the postulated double ended break of the largest reactor coolant system pipe, 1/2 second after the break, about 30% of the heat generated in the fuel rods is from gamma ray absorption. The power shape, reducing the energy deposited in the hot rod at the expense of adjacent colder rods. A conservative estimate of this effect is a reduction of 10% of the gamma ray contribution or 3% of the total. Since the water density is considerably reduced at this time, an average of 98% of the available heat is deposited in the fuel rods, the remaining 2% being absorbed by water, thimbles, sleeves and grids. The net effect is a factor of 0.95 rather than 0.974, to be applied to the heat production in the hot rod.
15.0-15a REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.0.11 Computer Codes Utilized Summaries of some of the principal computer codes used in transient analyses are given below. Other codes, in particular very specialized codes in which the modeling has been developed to simulate one given accident, such as those used in the analysis of the reactor coolant system pipe break (Section 15.6),
are summarized in their respective accident analyses sections.
The codes used in the analysis of each transient have been listed in Table 15.0-2.
15.0.11.1 FACTRAN FACTRAN calculates the transient temperature distribution in a cross section of a metal clad UO2 fuel rod and the transient heat flux at the surface of the clad using as input the nuclear power and time-dependent coolant parameters (pressure, flow, temperature, and density). The code uses a fuel model which exhibits the following features simultaneously.
- a. A sufficiently large number of radial space increments to handle fast transients such as rod ejection accidents.
- b. Material properties which are functions of temperature and a sophisticated fuel-to-clad gap heat transfer calculation.
- c. The necessary calculations to handle post-DNB transients: film boiling heat transfer correlations, zirconium-water reaction and partial melting of the materials.
FACTRAN is further discussed in Reference 6.
15.0.11.2 LOFTRAN The LOFTRAN program is used for studies of transient response of a PWR system to specified perturbations in process parameters. LOFTRAN simulates a multiloop system by a model containing reactor vessel, hot and cold leg piping, steam generator (tube and shell sides) and the pressurizer. The pressurizer heaters, spray, relief and safety valves are also considered in the program. Point model neutron kinetics, and reactivity effects of the moderator, fuel, boron and rods are included. The secondary side of the steam generator utilizes a homogeneous, saturated mixture for the thermal transients and a water level correlation for indication and control. The reactor protection system is simulated to include reactor trips on high neutron flux, overtemperature T, overpower T, high and low pressure, low flow, and high pressurizer level. Control systems are also simulated including rod control, steam dump, feedwater control and pressurizer pressure control. The 15.0-16 REVISION 6 - DECEMBER 1996
B/B-UFSAR emergency core cooling system, including the accumulators is also modeled.
LOFTRAN is a versatile program which is suited to both accident evaluation and control studies as well as parameter sizing.
LOFTRAN also has the capability of calculating the transient value of DNBR based on the input from the core limits illustrated on Figure 15.0-1. The core limits represents the minimum value of DNBR as calculated for typical or thimble cell.
LOFTRAN is further discussed in Reference 7.
15.0.11.3 TWINKLE The TWINKLE program is a multi-dimensional spatial neutron kinetics code, which was patterned after steady state codes presently used for reactor core design. The code uses an implicit finite-difference method to solve the two-group transient neutron diffusion equations in one, two and three dimensions. The code uses six delayed neutron groups and contains a detailed multi-region fuel-clad-coolant heat transfer model for calculating pointwise Doppler and moderator feedback effects. The code handles up to 2000 spatial points, and performs its own steady state initialization. Aside from basic cross section data and thermal-hydraulic parameters, the code accepts as input basic driving functions such as inlet temperature, pressure, flow, boron concentration, control rod motion, and others. Various edits are provided e.g., channel-wise power, axial offset, enthalpy, volumetric surge, pointwise power, and fuel temperatures.
The TWINKLE Code is used to predict the kinetic behavior of a reactor for transients which cause a major perturbation in the spatial neutron flux distribution.
TWINKLE is further discussed in Reference 8.
15.0.11.4 Deleted 15.0.11.5 Deleted 15.0-17 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.0.11.6 RADTRAD RADTRAD is used to determine accident doses at the appropriate dose points cited in Regulatory Guide 1.183; the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and Control Room (CR).
RADTRAD is a simplified model of RADionuclide Transport and Removal And Dose Estimation developed for, and endorsed by, the NRC as an acceptable methodology for reanalysis of the radiological consequences of design basis accidents.
In accident analyses RADTRAD is used to estimate the releases using the Alternative Source Term assumptions. The RADTRAD code uses a combination of tables and/or numerical models of source term reduction phenomena to determine the time-dependent dose at user-specified locations for a given accident scenario. The code system also provides the inventory, decay chain, and dose conversion factor tables needed for the dose calculation. The technical basis for the RADTRAD code is documented in Reference 15.
15.0.11.7 ARCON96 ARCON96 was developed to calculate relative concentrations in plumes from nuclear power plants at control room air intakes in the vicinity of the release point.
ARCON96 implements a straight-line Gaussian dispersion model with dispersion coefficients that are modified to account for low wind meander and building wake effects. Hourly, normalized concentrations (/Q) are calculated from hourly meteorological data. The hourly values are averaged to form /Qs for periods ranging from 2 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> in duration. The calculated values for each period are used to form cumulative frequency distributions.
The technical basis for the ARCON96 code is documented in Reference 16.
15.0-17a REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.0.11.8 PAVAN PAVAN estimates down-wind ground-level air concentrations for potential accidental releases of radioactive material from nuclear facilities. Options can account for variation in the location of release points, additional plume dispersion due to building wakes, plume meander under low wind speed conditions, and adjustments to consider non-straight trajectories. It computes an effective plume height using the physical release height which can be reduced by inputted terrain features.
Using joint frequency distributions of wind direction and wind speed by atmospheric stability, the program provides relative air concentration (/Q) values as functions of direction for various time periods at the exclusion area boundary (EAB) and the outer boundary of the low population zone (LPZ). Calculations of /Q values can be made for assumed ground-level releases or elevated releases from free-standing stacks. The /Q calculations are based on the theory that material released to the atmosphere will be normally distributed (Gaussian) about the plume centerline. A straight-line trajectory is assumed between the point of release and all distances for which /Q values are calculated.
The technical basis for the PAVAN code is documented in Reference 17.
15.0.11.9 VIPRE The VIPRE computer program performs thermal-hydraulic calculations. The code calculates coolant density, mass velocity, enthalpy, void fractions, static pressure and DNBR distributions along flow channels within a reactor core.
The VIPRE code is described in Reference 18.
15.0.11.10 ANC ANC is an advanced nodal code capable of two-dimensional and three-dimensional neutronics calculations. ANC is the reference model for certain safety analysis calculations, power distributions, peaking factors, critical boron concentrations, control rod worths, reactivity coefficients, etc. In addition, three-dimensional ANC validates one-dimensional and two dimensional results and provides information about radial (x-y) peaking factors as a function of axial position. It can calculate discrete pin powers from nodal information as well.
The ANC code is described in Reference 19.
15.0-17b REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.0.11.11 LOFTTR2 LOFTTR2 is a derivative of the LOFTRAN program discussed in Subsection 15.0.11.2 with a more realistic break flow model, a 2-region steam generator secondary side, and an improved capability to simulate operator actions during a steam generator tube rupture (SGTR) event.
LOFTTR2 is further discussed in Reference 20.
15.0.12 Radiological Consequences This chapter also analyzes the effects of postulated accidents with respect to radiological consequences. The analysis considers a broad spectrum of events. The radiological consequences of each accident are shown to be within the guidelines of 10 CFR 100, or, for those accidents evaluated using Alternative Source Terms, the limits of 10 CFR 50.67 and the guidance of Regulatory Guide 1.183.
Chapter 15 also documents a set of radiological consequence analyses utilizing Alternative Source Terms (AST) methodology per Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors for the following accidents:
Loss of Coolant Accident (LOCA),
Fuel Handling Accident (FHA),
Control Rod Ejection Accident (CREA),
Locked Rotor Accident (LRA)
Main Steam Line Break Accident (MSLBA),
Steam Generator Tube Rupture Accident (SGTRA),and Boron Recycle Holdup Tank Failure (Atmospheric Release) 15.0-17c REVISION 15 - DECEMBER 2014
B/B-UFSAR The analyses use design-basis assumptions and parameters to demonstrate the adequacy of the plant design with regard to the guidelines of 10 CFR 100 and 10 CFR 50.67.
The parameters and assumptions used for each analysis, as well as the results, are presented in tabular form for each limiting accident. The sequence of events for each transient is listed in tables in the applicable sections of this chapter. In addition, the accident sequences are provided in Figure 15.0-7 through 15.0-25. Tables 15.0-11 and 15.0-12 present the accident doses for the Exclusion Area Boundary (EAB), the Low Population Zone (LPZ),
and the control room for the Byron/Braidwood Stations respectively.
This format of table identification is used consistently within the text of Chapter 15.0 to identify Braidwood Station.
Figures 15.0-24, 15.0-25a and 15.0-25b symbolically show event sequences of steam generator tube rupture and loss-of-coolant accidents. Included with the events illustrated are operator actions and systems necessary to mitigate the effects of the accidents. The figures were prepared using a protection sequence event diagram format to illustrate accident sequences. A description of operator actions required for system operation along with a reference to a listing of process instrumentation available to the operator following an accident is noted on the diagrams.
The sequence of events for a steam generator tube rupture is given in Tables 15.6-6a and 15.6-6b. This information is included in Table 15.6-1 for large and small break loss-of-coolant accidents.
Tables 15.0-13 and 15.0-14 list the atmospheric dilution factors
(/Q) used in the TID-14844 analyses. The 5th percentile atmospheric dilution factors are used in the analyses.
The assumptions and methodology for the TID-14844 radiological dose analysis are discussed in Attachment 15A.
Table 15.0-17 lists the /Q factors used in the AST analyses, as derived in Section 2.3. The AST radiological dose analyses assumptions and methodology are discussed in the applicable accidents Radiological Consequences sections of this chapter.
15.0-18 REVISION 15 - DECEMBER 2014
B/B-UFSAR The effects of Optimized and VANTAGE 5 fuel on the radiological consequence evaluations presented in Chapter 15.0 were assessed.
The effects of extended fuel burnup were considered.
Analyses of core and fuel gap fission product inventories were performed for burnups of 33,000 MWd/Mtu and 60,000 MWd/Mtu (Reference 13), as well as 33,000 MWd/Mtu and 48,000 MWD/Mtu (Reference 14).
15.0-18a REVISION 12 - DECEMBER 2008
B/B-UFSAR Analyses of core inventories show that increasing burnup from 33,000 MWd/Mtu to 60,000 MWd/Mtu produces negligible changes in the fission product inventories of short-lived noble gases and iodines. The effect of increased burnup on long-lived Kr-85 is essentially a linear increase in core inventory (Reference 14).
Since Kr-85 is not a significant contributor to the radiological impact of postulated accidents, the results presented in Chapter 15.0 change only slightly due to the use of extended burnup fuel and remain well within the NRC regulatory limits.
Analyses show that the fraction of an isotope that is released to the fuel rod gap can increase with fuel burnup increasing from 33,000 to 60,000 MWd/Mtu (Reference 13). This analysis was performed assuming the peak fuel rod in terms of burnup during normal operations and was applied to the fuel handling accidents in Section 15.7. Radiological consequences analyses, of the six Alternative Source Term (AST) Design Basis Accidents (DBAs) that result in control room and offsite exposure, were performed to support a full-scope implementation of AST as described in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors. The results of the plant-specific AST analyses used the guidance in RG 1.183 and met the requirements of 10CFR50.67 Accident source term. Technical Information Document (TID) 14844, Calculation of Distance Factors for Power and Test Reactor Sites, will continue to be used as the radiation dose basis for equipment qualification. Support of full scope implementation of AST consisted of the following steps.
Analysis of the atmospheric dispersion for the radiological propagation pathways Calculation of offsite Exclusion Area Boundary (EAB) and Low Population Zone (LPZ), Control Room (CR), and, for LOCA only, Technical Support Center (TSC) personnel Total Effective Dose Equivalent (TEDE) doses Identification of the AST based on plant-specific analysis of a bounding core fission product inventory Calculation of fission product deposition rates and transport and removal mechanisms Calculation of the release fractions for the DBAs that result in the most significant CR and offsite doses (i.e.,
LOCA, FHA, CREA, LRA, MSLB, and SGTR)
The analysis assumptions for the transport, reduction, and release of the radioactive material from the fuel and the reactor coolant are consistent with the guidance provided in applicable appendices of RG 1.183 for the analyzed DBAs.
AST calculations for the LOCA, FHA, CREA, LRA, MSLB, and SGTR 15.0-19 REVISION 12 - DECEMBER 2008
B/B-UFSAR were prepared for the simulation of the radionuclide release, transport, removal, and dose estimates associated with the postulated accidents. The RADTRAD computer code developed for and endorsed by the NRC for AST analyses was used in the calculations.
The RADTRAD program is a radiological consequence analysis code used to estimate post-accident doses at plant offsite locations and in the control room.
Vendor fuel supply data were used in the calculation of the reactor core fission products for RADTRAD analysis. The inventories were determined based on the licensed core power level prior to the MUR power uprate of 3586.6 megawatts thermal (MWt) and further adjusted to 102% (3658.3 MWt) in support of the AST evaluations. This core power level bounds the MUR core licensed power level including measurement uncertainties.
Control room and offsite atmospheric dispersion factors (/Qs) developed for the releases from each plant at the containment wall, plant vent, Steam Generator (SG) power operated relief valves (PORVs)/safety valves, and through a main steam line break were utilized. Offsite /Qs were calculated with the PAVAN computer code, using the guidance of Regulatory Guide 1.145, and control room /Qs were calculated with the ARCON96 computer code. The PAVAN and ARCON96 codes calculate relative concentrations in plumes from nuclear power plants at offsite locations and control room air intakes, respectively (refer to section 2.3). All of these codes have previously been used by the NRC in their safety reviews.
The Byron Station and Braidwood Station meteorological databases for the five-year period 1994-1998 were applied in the ARCON96 modeling analysis. Wind measurements at Byron Station were taken at 30 ft and 250 ft; and the vertical temperature difference was measured between 250 ft and 30 ft. Braidwood Station wind measurements were taken at 34 ft and 203 ft; and the vertical temperature difference was measured between 199 ft and 30 ft.
Calm wind speeds at both stations were assigned a value of 0.4 mph (i.e., one-half the threshold value) per UFSAR Section 2.3.4.
The minimum wind speed (i.e. wind threshold) was set to the ARCON96 default value of 0.5 m/sec in accordance with RG 1.194, Table A-2.
Other ARCON96 parameters are provided in section 2.3.
The atmospheric dispersion factors (/Qs) utilized are as found in Sections 2.3 and Table 15.0-17.
The key inputs/assumptions used in the AST analysis for each of the six major accidents and their results/conclusions are presented in the applicable UFSAR sections.
Hence the results of the radiological consequence evaluations presented in Chapter 15.0 remain acceptable as a result of the use of extended burnup fuel.
15.0-19a REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.0.13 Limiting Single Failures The most limiting single failure of safety-related equipment, where one exists, is identified in each analysis description, and the consequences of this failure are described therein. In some instances, because of redundancy in protection equipment, no single failure which could adversely affect the consequences of the transient has been identified. The failure assumed in each analysis is listed in Table 15.0-15.
All accident and safety analyses in the UFSAR that require safety valves actuation assume operation of three pressurizer safety valves to limit increases in RCS pressure. The overpressure protection analysis is also based on operation of three safety valves.
15.0-19b REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.0.14 Operator Actions For most of the events analyzed in Chapter 15.0, the plant will be in a safe and stable hot standby condition following the automatic actuation of reactor trip. This condition will in fact be similar to plant conditions following any normal, orderly shutdown of the reactor. At this point, the actions taken by the operator would be no different than normal operating procedures. The exact actions taken, and the time these actions would occur, will depend on what systems are available (e.g., steam dump system, main feedwater system) and the plans for further plant operation. As a minimum, to maintain the hot stabilized condition, decay heat must be removed via the steam generators. The main feedwater system and the steam dump or atmospheric relief system could be used for this purpose. Alternatively, the auxiliary feedwater system and the steam generator safety valves, both of which are safety grade systems, may be used. Although auxiliary feed may be started manually, it will be automatically actuated, if needed, by one of the signals shown in Drawing 108D685, Sheet 15, such as low steam generator level. If hot standby conditions are maintained for an extended period of time, operator action may be required to transfer the auxiliary feedwater system to the long-term source of auxiliary feedwater. The time when such action is required depends on the design of the auxiliary feed system, but will be sufficiently long to permit operator action. Also, if the hot standby condition is maintained for an extended period of time (greater than approximately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />), operator action may be required to add boric acid via the CVCS to compensate for xenon decay and to maintain shutdown margin. The CVCS could also be used to control pressurizer level according to the operating procedures. Again, the actions taken by the operator would be no different than during a normal plant shutdown.
For several events involving breaks in the reactor coolant system or secondary system piping, additional requirements for operator action can be identified as described in the following.
Following the hypothetical steamline break incident, a safety injection signal (generated a few seconds after the break) will cause main feedwater isolation to occur. The only source of water available to the faulted steam generator is then the auxiliary feedwater system. Following steamline isolation, steam pressure in the steamline with the faulted steam generator will continue to fall rapidly, while the pressure stabilizes in the remaining three steamlines. The indication of the different steam pressures will be available to the operator within a few seconds of steamline isolation. This will provide the information necessary to identify the faulted steam generator so that auxiliary feedwater to it can be isolated. Manual controls are provided in the control room for start and stop of the auxiliary feedwater pumps and for the control valves associated with the auxiliary feedwater system.
The means for detecting the faulted steam generator and isolating auxiliary feedwater to it requires only the use of 15.0-20 REVISION 12 - DECEMBER 2008
B/B-UFSAR safety grade equipment available following the break. The removal of decay heat in the long-term, (following the initial cooldown) using the remaining steam generators, requires only the auxiliary feedwater system as a water source and the secondary system safety valves to relieve steam.
For a feedwater line break, emergency operating procedures provide guidance to the operator during long-term cooling to prevent pressurization of the reactor coolant system. The operator is provided with the necessary controls to start, operate, and shutdown the units to assure safe and reliable operation under normal and accident conditions.
For the boron dilution transient, the operator is required to open RWST isolation valves LCA-112D and E, and close CVCS isolation valves LCV-112B and C. This action will allow 2300 ppm borated water to enter the RCS and terminate the RCS dilution.
Operator action (both short-term and long-term) required for the various modes of ECCS operation to mitigate the consequences of a loss-of-coolant accident (LOCA) or steamline break, as well as other accident conditions, are presented in the Emergency Operating Procedures. These procedures discuss 15.0-20a REVISION 12 - DECEMBER 2008
B/B-UFSAR the alarm/indications available to the operator to lead him to take the appropriate actions. The discussion provided below, constitutes an outline of the operator action required following a LOCA or steamline break.
The primary function of the safety injection system (SIS) is to provide emergency core cooling (ECC) in the event of a LOCA resulting from a break in the primary reactor coolant system (RCS) or to provide emergency boration in the event of a steamline break accident resulting from a break in the secondary steam system.
ECC following a LOCA is divided into three phases:
- a. Short-Term Core Cooling/Cold Leg Injection Phase The cold leg injection phase is defined as that period during which borated water is delivered from the refueling water storage tank (RWST) and accumulators to the RCS cold legs. During this phase, no operator actions are required to ensure proper ECCS operation.
- b. Long-Term Core Cooling/Cold Leg Recirculation The cold leg recirculation phase is that period during which borated water is recirculated from the containment sump to the RCS cold legs. Operator actions are required to establish the cold leg recirculation phase. These actions are detailed in Table 6.3-7 and are not required prior to 10 minutes following event initiation.
- c. Long-Term Core Cooling/Hot Leg Recirculation Phase The hot leg recirculation phase is that period during which borated water is recirculated from the containment sump to both the RCS hot legs and RCS cold legs. Operator actions required to establish hot leg recirculation are detailed until approximately 6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> following event initiation.
The emergency boration following a steam break accident would occur only during the injection phase. The function of the SIS during this phase would be to inject borated water into the RCS with sufficient shutdown reactivity to compensate for the change in RCS volume and counteract any reactivity increase caused by the resulting cooldown. The SIS would continue to inject borated water from the RWST until the RCS conditions have stabilized, the accident has been identified as a steam break, and the criteria for safety injection termination are satisfied. The operator should then take action to terminate ECCS operation.
15.0-21 REVISION 10 - DECEMBER 2004
B/B-UFSAR For a secondary system break, such as a steamline or feedwater line break, the operator is instructed to complete specific actions in the Westinghouse Owners Group Emergency Response Guidelines (ERG). These actions instruct the operator to identify the faulted steam generator and terminate auxiliary feedwater to it. When acceptable plant conditions exist (i.e.,
RCS subcooling, RCS pressure, pressurizer level, secondary heat sink), the operator is instructed to terminate SI. Postaccident monitoring instrumentation which meets appropriate safety criteria is provided to monitor the course of various postulated accidents, including feedwater line break and steamline break (see Section 7.5).
In regard to overpressurizing the reactor coolant system, the small break LOCA is less limiting than the steamline break for the following reasons:
- a. A small break LOCA is a relatively slow transient with the RCS liquid remaining at saturated condition for an extended period of time.
- b. In all cases due to a break in the RCS, the system depressurizes.
- c. Even for very small break sizes, a large enough amount of water leaks from the system so that level would not return to the pressurizer for a relatively long period of time.
- d. For small and large LOCAs, the operator utilizes the emergency core cooling system for long-term cooling so that the operator does not initiate normal cooldown procedures.
The Byron/Braidwood abnormal and emergency operating procedures provide instructions to the operator on appropriate post-LOCA manual actions, as follows:
- a. The operator is instructed to align the ECCS to cold leg recirculation when the RWST level reaches the auto switchover level setpoint and to hot leg recirculation at 6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
15.0-22 REVISION 10 - DECEMBER 2004
B/B-UFSAR
- b. The operator is instructed to control the steam generator and pressure levels utilizing the respective level indications.
- c. If acceptable plant conditions exist (i.e., RCS subcooling, RCS pressure, pressurizer level, secondary heat sink), the operator is instructed to terminate SI.
- d. If leakage past the PORVs is indicated by valve position or by conditions in the pressurizer relief tank, the operator is instructed to isolate the pressurizer PORVs.
- e. If the RCS low pressure setpoint for the reactor coolant pumps is reached and SI flow is being delivered, the operator is instructed to trip the RCPs.
All of the previous indications are located inside the control room (see Section 7.5).
For a steam generator tube rupture (SGTR), the operator is instructed to complete specific actions per the plant-specific emergency procedures addressing the SGTR accident. The emergency procedures are based on the latest revision of the Westinghouse Owners Group generic emergency response guidelines and are, therefore, consistent with the latest guidelines regarding SGTR mitigation. The transient is identified as an SGTR in the E-0 procedure. Procedure E-3 directs the operator to mitigate the transient and terminate the break flow for the purpose of preventing SG overfill. The tube rupture results in a reactor trip and safety injection due to low pressurizer pressure, and the E-3 procedure instructs the operator to perform the following major actions:
- 1. The operator identifies the accident as a SGTR in the E-0 procedure based on steam generator level rising in an uncontrolled manner and isolates AFW flow to the ruptured steam generator. After all remaining post-trip checks have been performed, the operator enters the E-3 procedure based on abnormal secondary plant radiation levels to mitigate and recover from the accident. The operator then completes the isolation of the secondary side of the ruptured steam generator.
Manual controls for the isolation of the ruptured steam generator are located on the main control board.
- 2. The next major step has the operator cool down the RCS below the saturation temperature at the ruptured steam generator pressure by opening the steam dumps or PORVs on the intact steam generators.
15.0-23 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- 3. The steam generator PORV controls and RCS temperature indication are both available in the control room to allow the operator to cool down the RCS.
- 4. After verifying adequate RCS subcooling exists, the operator uses the pressurizer spray valve or PORV to depressurize the RCS below the ruptured steam generator pressure to terminate the break flow.
- 5. The final major step is for the operator to terminate the safety injection flow to prevent repressurization of the RCS, reestablishment of the break flow, and potential overfilling of the ruptured steam generator.
- 6. After completing these actions, the tube rupture break flow has been terminated and the operator is ready to transition the plant to cold shutdown conditions utilizing the appropriate recovery procedure.
See Subsection 15.6.3 for further information regarding the SGTR analysis performed.
15.0.15 References
- 1. CCC-371, ORIGEN2.1: Isotope Generation and Depletion Code -
Matrix Exponential Method, RSIC Computer Code Collection, Oak Ridge National Laboratory, February 1996.
- 2. Deleted
- 3. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors,"
10 CFR 50.46 and Appendix K of 10 CFR 50. Federal Register, Volume 39, Number 3, January 4, 1974.
- 4. Bordelon, F. M., et al., "SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss of Coolant,"
WCAP-8302 (Proprietary), and WCAP-8306 (Nonproprietary),
June 1974.
- 5. Bordelon, F. M., et al., "LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Nonproprietary), June 1974.
- 6. Hargrove H. G., "FACTRAN - A Fortran-IV Code for Thermal Transients in a UO2 Fuel Rod," WCAP-7908-A, December 1989.
- 7. Burnett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary), WCAP-7907-A (Nonproprietary), April 1984.
15.0-23a REVISION 15 - DECEMBER 2014
B/B-UFSAR
- 8. Risher, D. H., Jr. and Barry, R. F., "TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code,"
WCAP-7979-P-A (Proprietary), and WCAP-8028-A (Non-Proprietary), January 1975.
- 9. "Westinghouse Nuclear Energy Systems Division Quality Assurance Plan," WCAP-8370-A.
- 10. Friedland, A. J., and Ray, S., "Revised Thermal Design Procedure," WCAP-11397-P-A and WCAP-11397-A, April 1989.
- 11. DELETED.
- 12. DELETED.
- 13. "Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors," NUREG/CR-5009, February 1988.
- 14. "Extended Burnup Evaluation of Westinghouse Fuel," WCAP-10125-P-A, December 1985.
- 15. NUREG/CR-6604, RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation, April 1998.
- 16. NUREG-6331, Atmospheric Relative Concentrations in Building Wakes, Revision 1, May 1997.
- 17. NUREG-2858, PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations, November 1982.
- 18. Y.X. Sung, et. al., WCAP-14565-P-A (Proprietary)/WCAP-15306-A (Non-Proprietary), VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, October 1999.
- 19. Y.S. Liu, et. al., WCAP-10965-P-A, ANC: A Westinghouse Advanced Nodal Computer Code, September 1986.
- 20. Lewis, Huang, Behnke, Fittante, Gelman, SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, WCAP-10698-P-A (Proprietary) and WCAP-10750-A (Non-Proprietary), August 1987.
- 21. Shah, H. H., Optimized ZIRLO', WCAP-12610-P-A &
CENPD-404-P-A Addendum 1-A, July 2006.
15.0-24 REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.0-1 NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS
- 1. NSSS thermal power output (MWt) 3672.0 (nominal pump power)
- 2. Nominal thermal power generated by 14 the reactor coolant pumps (MWt)
Maximum thermal power generated by 20 the reactor coolant pumps (MWt)
- 3. Core thermal power (MWt) 3658.0/3648.0(a)
(a) The Byron/Braidwood MUR-PU increased the licensed reactor core power level from 3586.6 MWt to 3645.0 MWt. A Conservative nominal value of 3658.0 MWt was used in the transient analyses supporting the MUR-PU. A conservative nominal value of 3648.0 MWt was used in the DNB analyses supporting the MUR-PU.
15.0-25 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED (REACTIVITY COEFFICIENTS ASSUMED)
REVISED MODERATOR MODERATOR THERMAL COMPUTER TEMPERATURE DENSITY DNB DESIGN FAULTS CODES UTILIZED (K/F) (/gm/cc) DOPPLER CORRELATION PROCEDURE 15.1 Increase in Heat Removal by the Secondary System
- Feedwater System Mal- LOFTRAN, VIPRE, ANC NA 0.43 Minimum EOL* WRB-2 Yes function Causing a Decrease in Feedwater Temperature
- Feedwater System Mal- LOFTRAN, VIPRE, ANC NA 0.43 Minimum EOL* WRB-2 Yes function Causing an Increase in Feedwater Flow
- Excessive Increase in LOFTRAN NA 0 and 0.54 Minimum* WRB-2 Yes Secondary Steam and Maximum Flow 15.0-26 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED (REACTIVITY COEFFICIENTS ASSUMED)
REVISED MODERATOR MODERATOR THERMAL COMPUTER TEMPERATURE DENSITY DNB DESIGN FAULTS CODES UTILIZED (K/F) (/gm/cc) DOPPLER CORRELATION PROCEDURE
- Steam System Piping VIPRE, ANC, LOFTRAN NA Function of Moderator See Subsection WLOP No Failure at Zero Power Density, See 15.1.5 Subsection 15.1.5 (Figure 15.1-11)
- Steam System Piping VIPRE, ANC, LOFTRAN NA 0.43 Minimum EOL WRB-2 Yes Failure at Full Power 15.2 Decrease in Heat Removal by the Secondary System 15.0-26a REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REVISED MODERATOR MODERATOR THERMAL COMPUTER TEMPERATURE DENSITY DNB DESIGN FAULTS CODES UTILIZED (K/F) (/gm/cc) DOPPLER CORRELATION PROCEDURE
- Loss of External LOFTRAN 0 NA Minimum* WRB-2 Yes (DNB)
Electrical Load NA (press.)
and/or Turbine Trip
- Loss of Non-Emergency LOFTRAN 0 NA Maximum* NA NA A-C Power to the Station Auxiliaries
- Loss of Normal Feedwater LOFTRAN 0 NA Maximum* NA NA Flow
- Feedwater System Pipe LOFTRAN 0 NA Minimum* NA NA Break 15.3 Decrease in Reactor Coolant System Flow Rate
- Partial and Complete LOFTRAN, VIPRE Figure 15.0-2 NA Maximum* WRB-2 Yes
- Loss of Forced Reactor Coolant Flow
- Reactor Coolant Pump LOFTRAN, VIPRE Figure 15.0-2 NA Maximum* WRB-2 Yes Shaft Seizure (Locked)
Rotor 15.0-27 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REVISED MODERATOR MODERATOR THERMAL COMPUTER TEMPERATURE DENSITY DNB DESIGN FAULTS CODES UTILIZED (K/F) (/gm/cc) DOPPLER CORRELATION PROCEDURE 15.4 Reactivity and Power Distribution Anomalies
- Uncontrolled Rod Cluster TWINKLE, Refer to Refer to Refer to ABB-NV (first No Control Assembly Bank FACTRAN, Subsection Subsection Subsection grid span)
Withdrawal from a Sub- VIPRE 15.4.1.2 15.4.1.2 15.4.1.2 WRB-2 (remaining critical or Low Power grid spans)
Startup Condition
- Uncontrolled Rod Cluster LOFTRAN Figure 15.0-2 N/A Maximum and WRB-2 Yes Assembly Bank Withdrawal Minimum*
at Power 15.0-28 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REVISED MODERATOR MODERATOR THERMAL COMPUTER TEMPERATURE DENSITY DNB DESIGN FAULTS CODES UTILIZED (K/F) (/gm/cc) DOPPLER CORRELATION PROCEDURE
- Spectrum of Rod Cluster TWINKLE, Refer to Refer to Refer to NA NA Control Assembly Ejection FACTRAN Subsection Subsection Subsection Accidents 15.4.8 15.4.8 15.4.8.2.1 min., max. min., max.
feedback feedback 15.5 Increase in Coolant Inventory
- Inadvertent Operation LOFTRAN 0 NA Maximum and WRB-2 Yes(DNB) of ECCS During Power Minimum* NA(PZR-Fill)
Operation 15.6 Decrease in Reactor Coolant Inventory
- Inadvertent Opening of a LOFTRAN 0 NA Minimum* WRB-2 Yes Pressurizer Safety or Relief Valve
- Steam Generator Tube LOFTTR2 0 NA Maximum* NA NA Rupture Offsite Dose Case
- Steam Generator Tube LOFTTR2 0 NA Minimum* NA NA Rupture Margin To Overfill Case 15.0-29 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REACTOR VESSEL PRESSURIZER***
INITIAL NSSS VESSEL AVERAGE PRESSURIZER WATER FEEDWATER **
THERMAL POWER OUTPUT COOLANT TEMPERATURE PRESSURE VOLUME TEMPERATURE FAULTS (MWt) FLOW (GPM) (F) (PSIA) (ft3) (F) 15.1 Increase in Heat Removal by the Secondary System
- Feedwater System 3672.0 386,000 589.5 2250 1061.2 449.2 Malfunction Causing a Decrease in Feedwater Temperature
- Feedwater System 3672.0 386,000 589.5 2250 1061.2 449.2 Malfunction Causing an Increase in Feed-water Flow
- Excessive Increase in 3672.0 386,000 589.5 2250 1061.2 449.2 Secondary Steam Flow
- Steam System Piping 0 368,000 557 2250 693.84 100.0 Failure at Zero Power (Subcritical)
- Steam System Piping 3672.0 386,000 589.5 2250 979.6 449.2 Failure at Full Power 15.0-30 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REACTOR VESSEL PRESSURIZER***
INITIAL NSSS VESSEL AVERAGE PRESSURIZER WATER FEEDWATER **
THERMAL POWER OUTPUT COOLANT TEMPERATURE PRESSURE VOLUME TEMPERATURE FAULTS (MWt) FLOW (GPM) (F) (PSIA) (ft3) (F) 15.2 Decrease in Heat Removal by the Secondary System
- Loss of External 3672.0 (DNB) 386,000 (DNB) 589.5 (DNB) 2250 (DNB) 1142.9 (DNB) 449.2 (DNB)
Electrical Load and/or 3672.6 368,000 597.1 2207 1149.9 446.6 Turbine Trip (RCS press.) (RCS press.) (RCS press. high) (RCS press.) (RCS press.) (RCS press.)
578.9 (RCS press. low) 3672.0 368,000 597.1 2207 1142.9 449.2 (MSS press.) (MSS press.) (MSS press.) (MSS press.) (MSS press.) (MSS press.)
- Loss of Non-Emergency 3672.6 368,000 597.1 (high) 2293 (high) 1149.9 448.4 A-C Power to the 567.4 (low) 2207 (low)
Station Auxiliaries 15.0-30a REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REACTOR VESSEL PRESSURIZER***
INITIAL NSSS VESSEL AVERAGE PRESSURIZER WATER FEEDWATER **
THERMAL POWER OUTPUT COOLANT TEMPERATURE PRESSURE VOLUME TEMPERATURE FAULTS (MWt) FLOW (GPM) (F) (PSIA) (ft3) (F)
- Loss of Normal 3672.6 368,000 597.1 (high) 2293 (high) 1149.9 448.4 Feedwater Flow 567.4 (low) 2207 (low)
- Feedwater System Pipe 3672.6 368,000 597.1 2293 (high) 1067 448.4 Break 2207 (low) 15.3 Decrease in Reactor Coolant System Flow Rate
- Partial and Complete 3672.0 386,000 589.5 2250 1061.2 449.2 Loss of Forced Reactor Coolant Flow
- Reactor Coolant Pump 3672.6 368,000 597.1 2293 1149.9 446.6 Shaft Seizure/Locked Rotor (Peak RCS Pressure and Peak Clad Temperature Core)
Reactor Coolant Pump 3672.0 386,000 589.5 2250 1061.2 449.2 Shaft Seizure/Locked Rotor (Rods-in-DNB core) 15.4 Reactivity and Power Distribution Anomalies
- Uncontrolled Rod Cluster 0 163,686 557.0 2207 NA NA Control Assembly Bank Withdrawal from a Sub-critical or Low Power Startup Condition 15.0-31 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REACTOR VESSEL PRESSURIZER***
INITIAL NSSS VESSEL AVERAGE PRESSURIZER WATER FEEDWATER **
THERMAL POWER OUTPUT COOLANT TEMPERATURE PRESSURE VOLUME TEMPERATURE FAULTS (MWt) FLOW (GPM) (F) (PSIA) (ft3) (F)
- Uncontrolled Rod 3672.0/2203.2/367.2 386,000 589.5/577.1/ 2250 1061.2/832.6/ 433.0/379.5/
Cluster Assembly Bank 561.6 546.9 264.0 Withdrawal at Power (DNB)
Uncontrolled Rod 293.8 368,000 569.2 2207 541.8 272 Cluster Assembly Bank Withdrawal at Power (Peak Pressure)
- Spectrum of Rod 0 and 3672.6 163,686 and 557 2207 NA NA Cluster Control 368,000 and Assembly Ejection 597.1 Accidents 15.0-32 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-2 (Cont'd)
REACTOR VESSEL PRESSURIZER***
INITIAL NSSS VESSEL AVERAGE PRESSURIZER WATER FEEDWATER **
THERMAL POWER OUTPUT COOLANT TEMPERATURE PRESSURE VOLUME TEMPERATURE FAULTS (MWt) FLOW (GPM) (F) (PSIA) (ft3) (F) 15.5 Increase in Coolant Inventory
- Inadvertent Operation 3672.0 386,000(DNB) 589.5(DNB) 2250(DNB) 1142.9 449.2 of ECCS During Power 358,800 572.0 2207 Operation (PZR-Fill) (PZR-Fill) (PZR-Fill) 15.6 Decrease in Reactor Coolant Inventory
- Inadvertent Opening 3672.0 386,000 589.5 2250 1061.2 435.0 of a Pressurizer Safety or Relief Valve
- Steam Generator 3672 368,000 588.0 2207 1061.2 UNIT 1 433.0 UNIT 1 Tube Rupture 1066.6 UNIT 2 435.0 UNIT 2 Offsite Dose Case
- Steam Generator 3672 368,000 580.0 UNIT 1 2207 1061.2 433.0 UNIT 1 Tube Rupture 575.0 UNIT 2 435.0 UNIT 2 Margin to Overfill Case
____________________
- Reference Figure 15.0-3 Maximum refers to lower curve, minimum refers to upper curve, and minimum EOL refers to least negative EOL curve.
- Evaluations or analyses have been performed to support a nominal feedwater temperature window of 433F to 449.2F for Unit 1 and 435F to 449.2F for Unit 2.
- Evaluations have been performed to allow a decrease in nominal pressurizer water volume from 1066.6 ft3 to 1061.2 ft3. This reduction in volume is due to updated volume calculation, not a physical change. The evaluations also addressed initial volumes that included uncertainties based upon the reduced nominal volume.
NA - Not Applicable BOC - Beginning of Cycle EOC - End of Cycle 15.0-33 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-3 NOMINAL VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN ACCIDENT ANALYSES (BWI STEAM GENERATORS)
Thermal Output of NSSS (MWt)* 3672.0 Core Inlet Temperature (ºF) 541.4 (Low) 555.1 (high)
Vessel Average Temperature (ºF) **575.0 (low) 588.0(high)
Reactor Coolant System Pressure 2250 (psia)
Reactor Coolant Flow per Loop 92,000 (TDF)
(gpm) 96,500 (MMF)
Total Reactor Coolant Flow 140.0 (TDF, low Tavg)
(106 lb/hr) 137.4 (TDF, high Tavg) 146.8 (MMF, low Tavg) 144.1 (MMF, high Tavg)
Steam Flow from NSSS (106 lb/hr) 15.98/16.36 (low Tavg, low/high Tfeed, 0% SGTP) 15.98/16.35 (low Tavg, low/high Tfeed, 5% SGTP) 16.06/16.43 (high Tavg, low/high Tfeed, 0% SGTP) 16.05/16.43 (high Tavg, low/high Tfeed, 5% SGTP)
Steam Pressure at Steam 897 (low Tavg, 0% SGTP)
Generator Outlet (psia) 891 (low Tavg, 5% SGTP) 1008 (high Tavg, 0% SGTP) 1002 (high Tavg, 5% SGTP)
Assumed Feedwater Temperature 433.0 (low) at Steam Generator Inlet (ºF) 449.2 (high)
Average Core Heat Flux 210,815 (Btu/hr-ft2)
Maximum Average Steam Generator 5 Tube Plugging Level (%)
Maximum Steam Generator Tube 5 Plugging Level in any Steam Generator (%)
- See Table 15.0-2
- Low Vessel Average Temperature is limited to 580.0 ºF per SGTR analysis.
15.0-34 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-4 NOMINAL VALUES OF PERTINENT PLANT PARAMETERS UTILIZED IN ACCIDENT ANALYSES (D5 STEAM GENERATORS)
Thermal Output of NSSS (MWt)* 3672.0 Core Inlet Temperature (ºF) 541.4 (low) 555.1 (high)
Vessel Average Temperature (ºF) 575.0 (low) 588.0(high)***
Reactor Coolant System Pressure 2250 (psia)
Reactor Coolant Flow per Loop 92,000 (TDF)
(gpm) 96,500 (MMF)
Total Reactor Coolant Flow 140.0 (TDF, low Tavg)
(106 lb/hr) 137.4 (TDF, high Tavg) 146.8 (MMF, low Tavg) 144.1 (MMF, high Tavg)
Steam Flow from NSSS (106 lb/hr) 16.01/16.34 (low Tavg, low/high Tfeed, 0% SGTP) 16.00/16.32 (low Tavg, low/high Tfeed, 10% SGTP) 16.09/16.42 (high Tavg, low/high Tfeed, 0% SGTP) 16.07/16.39 (high Tavg, low/high Tfeed, 10% SGTP)
Steam Pressure at Steam 837/829 (low Tavg, low/high Tfeed, 0% SGTP)
Generator Outlet (psia) 809/802 (low Tavg, low/high Tfeed, 10% SGTP) 953/945 (high Tavg, low/high Tfeed, 0% SGTP) 923/914 (high Tavg, low/high Tfeed, 10% SGTP)
Assumed Feedwater Temperature 435.0 (low) at Steam Generator Inlet (ºF) 449.2 (high)
Average Core Heat Flux 210,815 (Btu/hr-ft2)
Maximum Average Steam Generator 10**
Tube Plugging Level (%)
Maximum Steam Generator Tube 10**
Plugging Level in any Steam Generator (%)
- See Table 15.0-2
15.0-35 REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.0-5 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES LIMITING TRIP TRIP POINT ASSUMED TIME DELAYS FUNCTION IN ANALYSIS (SECONDS)
Power range high neutron 118% 0.5 flux, high setting Power range high neutron 35% 0.5 flux, low setting Overtemperature T Variable see 8.0*
Figure 15.0-1 Overpower T Variable see 8.0*
Figure 15.0-1 High pressurizer pressure 2456 psig 2.0 Low pressurizer pressure 1845 psig 2.0 Low reactor coolant flow 85.1% loop flow 1.0 (From loop flow detectors)
Underfrequency trip 54.0 Hz 0.6 Undervoltage trip Not Applicable 1.5 Turbine trip Not applicable 2.0 Low-low steam generator 10% of narrow range 2.0 level** level span(Unit 1 loss of normal feed) 0% of narrow range level span (Unit 1 feedline break) 28.6% of narrow range level span (Unit 2 loss of normal feed) 18.6% of narrow range level span (Unit 2 feedline break) 15.0-36 REVISION 7 - DECEMBER 1998
B/B-UFSAR TABLE 15.0-5 (Cont'd)
TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES LIMITING TRIP TRIP POINT ASSUMED TIME DELAYS FUNCTION IN ANALYSIS (SECONDS)
High steam generator 100% of narrow 2.5 level trip of the range level span feedwater pumps and closure of feedwater system valves, and turbine trip
____________________
- Total response time (including RTD time response, trip circuit delay time, electronic filtering and channel electronics delay) from the time the temperature difference in the coolant loops exceeds the trip setpoint until the rods are free to fall, including time for the trip breakers to open.
- See Subsections 15.2.6, 15.2.7, and 15.2.8 for details.
15.0-36a REVISION 7 - DECEMBER 1998
B/B-UFSAR TABLE 15.0-6 DETERMINATION OF MAXIMUM OVERPOWER TRIP POINT - POWER RANGE NEUTRON FLUX CHANNEL - BASED ON NOMINAL SETPOINT CONSIDERING INHERENT INSTRUMENT ERRORS ACCURACY OF EFFECT ON MEASUREMENT THERMAL POWER OF VARIABLE DETERMINATION VARIABLE (% ERROR) (% ERROR)
(ESTIMATED) (ASSUMED)
Calorimetric Errors in the Measurement of Secondary System Thermal Power:
Feedwater temperature 0.5 Feedwater pressure 0.5 0.3 (small correction on enthalpy)
Steam pressure (small 2 correction on enthalpy)
Feedwater flow 1.25 1.25 Assumed Calorimetric Error
(% of rated power) 2 = a Axial power distribution effects on total ion chamber current Estimated Error 3
(% of rated power)
Assumed Error 5 = b
(% of rated power) 15.0-37
B/B-UFSAR TABLE 15.0-6 (Cont'd)
ACCURACY OF EFFECT ON MEASUREMENT THERMAL POWER OF VARIABLE DETERMINATION VARIABLE (% ERROR) (% ERROR)
(ESTIMATED) (ASSUMED)
Instrumentation channel drift and setpoint repro-ducibility Estimated Error 1
(% of rated power)
Assumed Error 2 = c
(% of rated power)
Total assumed error in setpoint 9
= a + b + c Percent of Rated Power Nominal Setpoint 109 Maximum overpower trip point 118 assuming all individual errors are simultaneously in the most adverse direction 15.0-38
B/B-UFSAR TABLE 15.0-7 PLANT SYSTEMS AND EQUIPMENT CREDITED FOR TRANSIENTS AND ACCIDENT CONDITIONS INCIDENT REACTOR TRIP FUNCTIONS ESF ACTUATION FUNCTIONS OTHER EQUIPMENT ESF EQUIPMENT 15.1 INCREASE IN HEAT REMOVED BY THE SECONDARY SYSTEM Feedwater system Power range high neutron SI on low pressurizer Feedwater isolation NA malfunction that results flux, overpower pressure, valves in a decrease in feedwater T, overtemperature T feedwater isolation on SI temperature Feedwater system mal- Power range high neutron Feedwater isolation and Feedwater isolation NA functions that result flux, overpower turbine trip on hi-hi steam valves, steam in an increase in feed- T, overtemperature T generator water level generator safety water flow valves Excessive increase in Power range high neutron NA Pressurizer safety NA secondary steam flow flux, overtemperature T, valves, steam overpower T, low generator safety valves pressurizer pressure Steam system piping Power range high neutron SI on low steamline Feedwater isolation Auxiliary Feedwater failure flux, SI, low pressurizer pressure or low pressurizer valves, steamline System, Safety pressure, overpower T pressure, steamline isolation valves Injection System, isolation on low steamline Emergency Power pressure, feedwater System isolation on SI 15.0-39 REVISION 11 - DECEMBER 2006
B/B-UFSAR TABLE 15.0-7 (Cont'd)
INCIDENT REACTOR TRIP FUNCTIONS ESF ACTUATION FUNCTIONS OTHER EQUIPMENT ESF EQUIPMENT 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM Loss of external elec- High pressurizer NA Pressurizer safety NA trical load/turbine pressure overtemperature valves, steam gen-trip T, lo-lo steam erator safety valves generator water level Loss of non-emergency Lo-lo steam generator AF initiation on lo-lo Steam generator safety Auxiliary Feedwater ac power to the station water level steam generator water level valves, pressurizer System auxiliaries safety valves Loss of normal feedwater Lo-lo steam generator AF initiation on lo-lo Steam generator safety Auxiliary Feedwater flow water level steam generator water level valves, pressurizer System safety valves Feedwater system pipe Lo-lo steam generator SI on low steamline Steamline isolation Auxiliary Feedwater break water level, SI pressure or low pressurizer valves, feedwater System, Safety pressure, AF initiation on isolation valves, Injection System lo-lo steam generator water pressurizer safety level, steamline isolation valves, steam genera-on SI, feedwater isolation tor safety valves on SI 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE Partial and complete Low reactor coolant NA Steam generator NA loss of forced reactor loop flow, reactor safety valves coolant flow coolant pump undervoltage and underfrequency Reactor coolant pump Low reactor coolant loop NA Pressurizer safety NA shaft seizure (locked flow valves, steam genera-rotor) tor safety valves 15.0-40 REVISION 11 - DECEMBER 2006
B/B-UFSAR TABLE 15.0-7 (Cont'd)
INCIDENT REACTOR TRIP FUNCTIONS ESF ACTUATION FUNCTIONS OTHER EQUIPMENT ESF EQUIPMENT 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES Uncontrolled rod cluster Power range high neutron NA NA NA control assembly bank flux (low setpoint),
withdrawal from a sub- source range high neutron critical or low power flux startup condition Uncontrolled rod cluster Power range high neutron NA Pressurizer safety NA control assembly bank flux, overtemperature T, valves, steam withdrawal at power high pressurizer pres- generator safety sure valves Dropped rod cluster Manual NA Pressurizer safety NA control assembly valves, steam generator safety valves Single rod cluster Overtemperature T NA NA NA control assembly withdrawal at full power Rod cluster control Manual NA NA NA assembly misalignment Chemical and volume Source range high neutron NA Boron Dilution Alert NA control system mal- flux, power range high alarms function that results neutron flux (high and in a decrease in boron low setpoint),
concentration in the overtemperature T reactor coolant Spectrum of rod cluster Power range high neutron NA Pressurizer safety NA control assembly flux (high and low valves ejection accidents setpoint), high positive neutron flux rate 15.0-41 REVISION 11 - DECEMBER 2006
B/B-UFSAR TABLE 15.0-7 (Cont'd)
INCIDENT REACTOR TRIP FUNCTIONS ESF ACTUATION FUNCTIONS OTHER EQUIPMENT ESF EQUIPMENT 15.5 INCREASE IN REACTOR COOLANT INVENTORY Inadvertent operation Low pressurizer pres- SI on low pressurizer Pressurizer Power Safety Injection of the ECCS during sure, SI pressure, AF initiation on Operated Relief Valves System, Auxiliary power operation SI and Safety Valves, Steam Feedwater System Generator Safety Valves 15.6 DECREASE IN REACTOR COOLANT INVENTORY Inadvertent opening Low pressurizer pressure, NA NA NA of a pressurizer overtemperature T safety or relief valve Steam generator tube Low pressurizer SI on low pressurizer Service Water, Com- Emergency Core Cool-rupture pressure, pressure, AF initiation on ponent Cooling Water, ing System, Auxiliary overtemperature T, SI or loss of offsite power Steam Generator Safety Feedwater System, SI Valves and Power Emergency Power Operated Relief Valves, System, AF Main Steam Isolation Accumulator Tanks Valves, Pressurizer Power Operated Relief Valves, Main Steam-Line Area Radiation Monitors, Instrument Air System 15.0-42 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-7 (Cont'd)
INCIDENT REACTOR TRIP FUNCTIONS ESF ACTUATION FUNCTIONS OTHER EQUIPMENT ESF EQUIPMENT Loss of coolant acci- Low pressurizer pressure SI on low pressurizer Service Water System, Emergency Core Cool-dents resulting from pressure or high Component Cooling ing System, Auxiliary the spectrum of postu- containment pressure, AF Water System, Steam Feedwater System, lated piping breaks initiation on SI Generator Safety Containment Heat Re-within the reactor and/or Relief Valves moval System, Emer-coolant pressure boun- gency Power System dary 15.0-42a REVISION 11 - DECEMBER 2006
B/B-UFSAR TABLE 15.0-8 IODINE AND NOBLE GAS INVENTORY IN REACTOR CORE*
(For use with TID-14844 Based Analyses)
CORE ACTIVITY ISOTOPE (CURIES)
Security - Related Information Withheld Under 10 CFR 2.390
- Based on end of equilibrium fuel cycle and power level of 3658.3 MWt. This core power level bounds the MUR core licensed power level including measurement uncertainties.
15.0-43 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.0-9 PRIMARY COOLANT NOBLE GAS ACTIVITY BASED ON 1% FUEL DEFECTS (For Use with TID-14844 Based Analyses and AST Analyses Using Regulatory Guide 1.183)
CONCENTRATION ISOTOPE (Ci/gm)
Kr-85m 1.8 Kr-85 7.11 Kr-87 1.15 Kr-88 3.35 Xe-131m 3.31 Xe-133m 3.65 Xe-133 251 Xe-135m 0.488 Xe-135 7.72 Xe-138 0.663 15.0-44 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.0-10 IODINE ACTIVITY IN THE PRIMARY AND SECONDARY COOLANT (For use with TID-14844 Based Analysis and AST Analyses Using Regulatory Guide 1.183)
Primary Coolant Isotope Equilibrium Equilibrium Iodine Spike Pre-Existing Concentration Appearance Appearance Iodine Spike (1.0 µCi/gm Rate, Rate,* (60 µCi/gm DE DE I-131) Ci/min Ci/min I-131)
µCi/gm µCi/gm I-131 0.742 0.416 208 44.5 I-132 0.979 1.754 877 58.7 I-133 1.350 0.923 462 81.0 I-134 0.243 0.926 463 14.6 I-135 0.842 0.826 413 50.5 Secondary Coolant Isotope Equilibrium Concentraion (0.1 µCi/gm DE I-131)
µCi/gm I-131 0.0742 I-132 0.0979 I-133 0.1350 I-134 0.0243 I-135 0.0842
- Based on spike factor of 500. The SGTR analysis uses a spike factor of 335 consistent with Regulatory Guide 1.183.
15.0-45 REVISION 15 - DECEMBER 2014
BYRON-UFSAR TABLE 15.0-11 POTENTIAL DOSES DUE TO ACCIDENTS POTENTIAL DOSES DUE TO ACCIDENTS USING AST AST Dose Results (rem TEDE)
Exclusion Area Boundary Low Population Zone Control Room Accident Dose Limit Dose Limit Dose Limit Loss of Coolant 15.03 25 4.99 25 4.94 5.0 Accident Main Steam (1) 0.146 25 (1) 0.083 (1) 25 (1) 0.264 (1)
Line Break (2) 5.0 0.201 2.5 (2) 0.459 (2) 2.5 (2) 0.654 (2)
Accident Control Rod Ejection 5.358 6.3 2.278 6.3 4.538 5.0 Accident Locked Rotor 1.679 2.5 0.602 2.5 2.790 5.0 Accident Steam (1)
Generator 3.7 25 (1) 0.69 (1) 25 (1) 2.0 (1)
(3) 5.0 Tube Rupture 2.1 2.5 (3) 0.41 (3) 2.5 (3) 0.56 (3)
Accident Fuel Handling Accident 5.31 6.3 0.94 6.3 4.40 5.0 Recycle Holdup Tank 0.852 6.3 0.152 6.3 0.459 5.0 Failure Note (1): Pre-accident 60 Ci/gm DEI spike Note (2): Accident initiated 500 times equilibrium iodine release rate spike Note (3): Accident initiated 335 times equilibrium iodine release rate spike 15.0-46 REVISION 17 - DECEMBER 2018
BYRON-UFSAR TABLE 15.0-11 (Cont'd)
POTENTIAL DOSES DUE TO ACCIDENTS POTENTIAL DOSES DUE TO ACCIDENTS USING TID-14844 (10 CFR 100):
DOSE (2 HOURS) AT EXCLUSION DOSE (COURSE OF ACCIDENT) AT AREA BOUNDARY LOW POPULATION ZONE (445 meters) (4828 meters)
UFSAR POSTULATED ACCIDENT SECTION THYROID (REM) WHOLE BODY (REM) THYROID (REM) WHOLE BODY (REM)
Process Gas System Rupture 15.7.1 0 5.4 (-1) 0 2.0 (-2)
Radioactive Liquid Waste System Failure 15.7.3 Spent Resin Tank 4.5 (-1) 1.6 (-4) 1.4 (-2) 4.7 (-6) 10CFR100 limits 300 25 300 25
________________________
Note: 2.89 (+1) = 2.89 x 101 15.0-47 REVISION 15 - DECEMBER 2014
BRAIDWOOD-UFSAR TABLE 15.0-12 POTENTIAL DOSES DUE TO ACCIDENTS POTENTIAL DOSES DUE TO ACCIDENTS USING AST AST Dose Results (rem TEDE)
Exclusion Area Boundary Low Population Zone Control Room Accident Dose Limit Dose Limit Dose Limit Loss of Coolant 15.03 25 4.99 25 4.94 5.0 Accident Main Steam (1) 0.146 25 (1) 0.083 (1) 25 (1) 0.264 (1)
Line Break (2) 5.0 0.201 2.5 (2) 0.459 (2) 2.5 (2) 0.654 (2)
Accident Control Rod Ejection 5.358 6.3 2.278 6.3 4.538 5.0 Accident Locked Rotor 1.679 2.5 0.602 2.5 2.790 5.0 Accident Steam (1)
Generator 3.7 25 (1) 0.69 (1) 25 (1) 2.0 (1)
(3) 5.0 Tube Rupture 2.1 2.5 (3) 0.41 (3) 2.5 (3) 0.56 (3)
Accident Fuel Handling Accident 5.31 6.3 0.94 6.3 4.40 5.0 Recycle Holdup Tank 0.852 6.3 0.152 6.3 0.459 5.0 Failure Note (1): Pre-accident 60 Ci/gm DEI spike Note (2): Accident initiated 500 times equilibrium iodine release rate spike Note (3): Accident initiated 335 times equilibrium iodine release rate spike 15.0-48 REVISION 17 - DECEMBER 2018
BRAIDWOOD-UFSAR TABLE 15.0-12 (Cont'd)
POTENTIAL DOSES DUE TO ACCIDENTS Potential Doses Due To Accidents Using TID-14844 (10 CFR 100):
DOSE (2 HOURS) AT EXCLUSION DOSE (COURSE OF ACCIDENT) AT AREA BOUNDARY LOW POPULATION ZONE (485 meters) (1811 meters)
UFSAR POSTULATED ACCIDENT SECTION THYROID (REM) WHOLE BODY (REM) THYROID (REM) WHOLE BODY (REM)
Process Gas System Rupture 15.7.1 0 7.3 (-1) 0 7.0 (-2)
Radioactive Liquid Waste System Failure 15.7.3 Spent Resin Tank 6.1 (-1) 2.1 (-4) 5.6 (-2) 2.0 (-5) 10CFR100 limits 300 25 300 25
________________________
Note: 3.90 (+1) = 3.90 x 101 15.0-49 REVISION 15 - DECEMBER 2014
BYRON-UFSAR TABLE 15.0-13 ACCIDENT ATMOSPHERIC DILUTION FACTORS (X/Q)* AT EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE FOR THE BYRON STATION (using TID-14844 Analyses)
EXCLUSION AREA BOUNDARY (445 Meters) LOW POPULATION ZONE (4828 Meters)
TIME PERIOD (Hours) 5th PERCENTILE 50th PERCENTILE 5th PERCENTILE 50th PERCENTILE 0-2 5.7 (-4) 6.5 (-5) 0-8 1.7 (-5) 1.5 (-6) 8-24 2.4 (-6) 3.1 (-7) 24-96 1.1 (-6) 1.5 (-7)96-720 7.6 (-7) 1.6 (-7)
- X/Q values, expressed in sec/m3, are based on hourly onsite meteorological data for the period of record, January 1974 - December 1976.
Note: 5.7 (-4) = 5.7 x 10-4 15.0-50 REVISION 12 - DECEMBER 2008
BRAIDWOOD-UFSAR TABLE 15.0-14 ACCIDENT ATMOSPHERIC DILUTION FACTORS (X/Q)* AT EXCLUSION AREA BOUNDARY AND LOW POPULATION ZONE FOR THE BRAIDWOOD STATION (using TID-14844 Analyses)
EXCLUSION AREA BOUNDARY (485 Meters) LOW POPULATION ZONE (1811 Meters)
TIME PERIOD (Hours) 5th PERCENTILE 50th PERCENTILE 5th PERCENTILE 50th PERCENTILE 0-2 7.7 (-4) 7.1 (-5) 0-8 7.1 (-5) 6.6 (-6) 8-24 1.4 (-5) 1.8 (-6) 24-96 7.1 (-6) 8.5 (-7)96-720 4.1 (-6) 1.0 (-6)
- X/Q values, expressed in sec/m3, are based on hourly onsite meteorological data for the period of record, January 1974 - December 1976.
Note: 7.7 (-4) = 7.7 x 10-4 15.0-51 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.0-15 SINGLE FAILURES ASSUMED IN ACCIDENT ANALYSES EVENT DESCRIPTION WORST FAILURE ASSUMED NSSS:
Feedwater temperature reduction One Protection Train Excessive feedwater flow One Protection Train Excessive steam flow (1)
Inadvertent secondary depressurization One Safety Injection Train Steam system piping failure One Safety Injection Train Steam pressure regulatory malfunction (2)
Loss of external load One Protection Train Turbine Trip One Protection Train Inadvertent closure of MSIV One Protection Train Loss of condenser vacuum One Protection Train Loss of ac power One Aux, FW Train Loss of normal feedwater One Aux, FW Train Feedwater system pipe break One Aux, FW Train Partial loss of forced reactor coolant flow One Protection Train Complete loss of forced reactor coolant flow One Protection Train RCP locked rotor One Protection Train RCP shaft break One Protection Train RCCA bank withdrawal from subcritical One Protection Train RCCA bank withdrawal at power One Protection Train Dropped RCCA, dropped RCCA bank Nuclear Instrumentation System Statically misaligned RCCA (3)
Single RCCA withdrawal One Protection Train Flow controller malfunction (2)
Uncontrolled boron dilution Standby charging pump is operating Improper fuel loading (3)
RCCA ejection One Protection Train Inadvertent ECCS operation at power One Protection Train Increase in RCS inventory One Protection Train BWR transients (2)
Inadvertent RCS depressurization One Protection Train Failure of small lines carrying primary coolant outside containment (3)
BWR piping failures (2) 15.0-52 REVISION 7 - DECEMBER 1998
B/B-UFSAR TABLE 15.0-15 (Cont'd)
EVENT DESCRIPTION WORST FAILURE ASSUMED Steam generator tube rupture Offsite dose case One SG PORV on ruptured steam generator (F.O.)
Margin to overfill case One SG PORV on intact steam generator (F.C.) (6)
Spectrum of LOCA Small Break One Electrical Train Large Break One SI Train (5)
For BOP:
Break in instrument line or other lines from reactor coolant pressure boundary that penetrate containment (Subsection 15.6.2) (3)
Radioactive gas waste system leak or failure (Subsection 15.7.1) (1)
Radioactive Liquid Waste System Leak or Failure (Atmospheric Release)
Spent Resin Storage Tank (1)
Boron Recycle Holdup Tank One Protection Train (5)
Postulated radioactive releases due to liquid tanks failures (Ground Release)
(Subsection 15.7.3) (1)
Postulated fuel handling accident inside Spent Fuel Storage Building (Subsection 15.7.4.2.1) One Protection Train (4)
Postulated fuel handling accident inside containment (Subsection 15.7.4.2.2) One Protection Train (4)
Spent fuel cask drop accidents (Subsection 15.7.5) (1)
____________________
NOTES:
(1) No protective action required.
(2) Not applicable to Byron/Braidwood.
(3) No transient analysis involved.
(4) Only one of the two filtration trains was considered in determining offsite doses.
(5) For determining control room dose one train of the control room filtration system is assumed to fail to realign to the emergency mode of operation.
(6) Failures, active or passive, that result in only two SG PORVs available for RCS cooldown on intact steam generators are bounding.
15.0-53 REVISION 16 - DECEMBER 2016
B/B-UFSAR B/B - UFSAR TABLE 15.0-16 REACTOR CORE NUCLIDE INVENTORY USING AST Security - Related Information Withheld Under 10 CFR 2.390 15.0-54 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.0-17 BYRON-BRAIDWOOD MAXIMUM SITE-UNIT /Q
SUMMARY
Release Path Recommended /Q (sec/m3)
Release Receptor/
Model 0-2 hr 2-8 hr(1) 8-24 hr 1-4 day 4-30 day Notes Point Intake Containment CR Fresh ARCON96 1.73E-03 1.24E-03 5.23E-04 3.55E-04 2.62E-04 Wall Air Diffuse area CR source per RG Turbine 1.194 Sec.
Containment ARCON96 Building 1.01E-03 7.25E-04 3.07E-04 2.07E-04 1.46E-04 3.2.4 Wall Emergency Air CR Fresh ARCON96 Plant Vent 2.22E-03 1.80E-03 7.20E-04 4.75E-04 3.81E-04 Building area Air perpendicular CR to wind Turbine utilized per ARCON96 Plant Vent Building 2.46E-03 1.92E-03 8.14E-04 5.52E-04 4.40E-04 RG 1.194 Emergency Table A-2 Air PORVs/Safety CR Fresh Includes ARCON96 1.77E-03 1.52E-03 6.98E-04 4.72E-04 3.50E-04 Valves Air factor of 5 reduction for CR vertical Turbine uncapped PORVs/Safety 8.14E-04 ARCON96 Building 6.98E-04 3.12E-04 1.95E-04 1.67E-04 release per Valves (2)
Emergency RG 1.194 Sec.
Air 6 "Taut String Length" to Fresh Air Intake and CR Fresh ARCON96 MSLB 3.205E-03 2.735E-03 1.1920E-03 8.205E-04 6.60E-04 building area Air perpendicular to wind utilized per RG 1.194 CR Building area Turbine perpendicular ARCON96 MSLB Building 1.70E-02 1.46E-02 6.68E-03 4.48E-03 3.31E-03 to wind Emergency utilized per Air RG 1.194 Outer 5.36E-04 2.65E-04 1.89E-04 9.04E-05 3.16E-05 PAVAN Containment EAB (N) (SE) (SE) (SE) (SE) In compliance Wall with RG 1.145 Midpoint (3) 9.32E-05 4.50E-05 3.12E-05 1.41E-05 4.54E-06 PAVAN between two LPZ (ESE) (ESE) (ESE) (ESE) (ESE) reactors Outer 6.18E-04 3.08E-04 2.17E-04 1.02E-04 3.44E-05 PAVAN Containment EAB (SE) (SE) (SE) (SE) (SE) In compliance Wall with RG 1.23 Midpoint 1.10E-04 5.13E-05 3.51E-05 1.53E-05 4.68E-06 Rev. 1(4)
PAVAN between two LPZ (W) (W) (W) (W) (W) reactors (1) /Q values for the PAVAN results are for 0-8 hour time period.
(2) A slightly more conservative value of 8.16E-04 was used in the MSLB, LRA, and CREA DBA calculations.
(3) /Q values using the wind speed categories provided in RG 1.23 Rev. 0.
(4) Based on a commitment made to the NRC, finer wind speed categories from RG 1.23 Rev. 1 have been used in the latest revisions to the LOCA, MSLB, CREA, LRA, SGTR, and FHA offsite dose consequences analyses.
15.0-55 REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM A number of events have been postulated which could result in an increase in heat removal from the reactor coolant system by the secondary system. Analyses are presented for several such events which have been identified as limiting cases.
Discussions of the following reactor coolant system cooldown events are presented in this section:
- c. Excessive increase in secondary steam flow.
- d. Inadvertent opening of a steam generator relief or safety valve.
- e. Steam system piping failure.
The above are considered to be ANS Condition II events, with the exception of a major steam system pipe break, which is considered to be an ANS Condition IV event. Subsection 15.0.1 contains a discussion of ANS classification and applicable acceptance criteria.
15.1.1 Feedwater System Malfunctions Causing a Reduction in Feedwater Temperature 15.1.1.1 Identification of Causes and Accident Description Reductions in feedwater temperature will cause an increase in core power by decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and of the reactor coolant system (RCS). The overpower - overtemperature protection (neutron overpower, overtemperature and overpower T trips) prevent any power increase which could lead to a DNBR less than the limit value.
A reduction in feedwater temperature may be caused by the accidental opening of a feedwater bypass valve which diverts flow around a portion of the feedwater heaters. In the event of an accidental opening of the bypass valve, there could be a sudden reduction in feedwater inlet temperature to the steam generators.
At power, this increased subcooling will create a greater load demand on the RCS.
With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator temperature coefficient of reactivity. However, the rate of 15.1-1
B/B-UFSAR energy change is reduced as load and feedwater flow decrease, so the no-load transient is less severe than the full power case.
The net effect on the RCS due to a reduction in feedwater temperature would be similar to the effect of increasing secondary steam flow, i.e., the reactor will reach a new equilibrium condition at a power level corresponding to the new steam generator T.
A decrease in normal feedwater temperature is classified as an ANS Condition II event, fault of moderate frequency. See Subsection 15.0.1 for a discussion of Condition II events.
The protection credited for mitigating the consequences of a decrease in feedwater temperature is the same as that for an excessive steam flow increase, as discussed in Section 15.0.8 and listed in Table 15.0-7.
15.1.1.2 Analysis of Effects and Consequences Method of Analysis The reduction in feedwater temperature due to a feedwater system malfunction transient is analyzed by using the detailed digital computer code, LOFTRAN (Reference 1). This code simulates a multiloop system, neutron kinetics, the pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperature, pressure, and power level.
The system is analyzed to demonstrate plant behavior following a reduction in feedwater temperature due to the simultaneous opening of the low pressure feedwater heater string bypass valve and isolation of a low pressure feedwater heater string due to a high-two level in a first stage feedwater heater. The loss of a low pressure feedwater heater string causes a redistribution of flow to the other two parallel heater strings and the bypass line. The bypass line takes 40% of the total flow, which effectively increases the temperature differential since less flow passes through the heaters. This reduction in feedwater temperature results in cascading feedwater heater instability with the end result being a loss of all feedwater heating downstream of the fourth stage feedwater heaters. This bypass case results in the most severe decrease in feedwater temperature. For the purpose of this analysis, it is conservatively assumed that the temperature reduction that occurs is due to the loss of multiple trains of feedwater heaters along with the opening of the feedwater heater bypass valves.
This accident is analyzed with the revised thermal design procedure as described in WCAP-11397-P-A (Reference 8). The following assumptions are made:
- a. Initial reactor power, pressure, and RCS temperatures are assumed to be at their normal values.
15.1-2 REVISION 11 - DECEMBER 2006
B/B-UFSAR
- b. For the main feedwater malfunction transient, the limiting temperature reduction failure is the loss of an entire train of feedwater heaters. To conservatively bound this failure, a loss of multiple trains of feedwater heaters is modeled along with the opening of the feedwater heater bypass valves. It is assumed that 40% of the feedwater flows through the bypass line. A conservative feedwater temperature of 200oF is used after the initiating failure for this event. This event is only analyzed at hot full power (HFP) conditions.
- c. The rate of feedwater flow to the steam generators is assumed to remain the same as it was immediately preceding the accident.
- d. Changes in turbine performance due to changes in RCS performance and process steam extraction rates are neglected.
- e. The temperature transient resulting from the bypass of an entire string of low pressure feedwater heaters is terminated by an overpower T trip signal, which trips the reactor, and by a safety injection system low pressurizer pressure signal, which closes the feedwater control and isolation valves.
- f. In order to determine the most limiting feedwater temperature reduction case for this accident, both of the Byron/Braidwood steam generator designs (BWI and D5) are analyzed for this event. The most limiting case of the two steam generator designs is documented and presented in the Results section that follows.
Plant characteristics and initial conditions are discussed in Subsection 15.0.3.
Normal reactor control systems and engineered safety systems are not required to function. The reactor protection system may function to trip the reactor due to an overpower condition. No single active failure will prevent operation of the reactor protection system.
Results Following the initiation of this event as described above, feedwater temperature is conservatively modeled to instantaneously decrease from 449.2oF to 200oF. This reduction in feedwater temperature increases the thermal load on the primary system. The resultant temperature and power transient causes a reactor trip on an overpower T signal. When the pressurizer pressure reaches the low pressure setpoint, the safety injection system is actuated, the feedwater control and isolation valves are closed, and feedwater isolation occurs.
15.1-2a REVISION 15 - DECEMBER 2014
B/B-UFSAR Transient results for the most limiting feedwater temperature reduction case, as shown in Figures 15.1-1a and 15.1-1b, show the increase in nuclear power and loop T associated with the increased thermal load on the reactor. Following reactor trip and feedwater isolation, the plant will approach a stabilized condition, which in this case, is hot standby. Normal plant operating procedures may then be followed. The operating procedures would call for operator action to control reactor coolant system boron concentration and pressurizer level using the chemical volume and control system (CVCS) and to maintain steam generator level through control of the main or auxiliary feedwater system.
Since the power level rises during the reduction in feedwater temperature event, the fuel temperatures will also rise until after reactor trip occurs. The core heat flux lags behind the neutron flux due to fuel rod thermal time constant and the peak linear rod power reached is limited to a value below that which would result in exceeding the fuel melting temperature. Hence, fuel melting is precluded for this event.
The transient results show that departure from nucleate boiling (DNB) does not occur at any time during the feedwater temperature reduction transient; thus, the ability of the primary coolant to remove heat from the fuel rod is not reduced. The fuel cladding temperature, therefore, does not rise significantly above its initial value during the transient.
The calculated sequence of events for the most limiting feedwater temperature reduction case for this accident is shown in Table 15.1-1.
15.1.1.3 Radiological Consequences The radiological consequences will be less severe than the steamline break accident analyzed in Subsection 15.1.5.3.
15.1.1.4 Conclusions The results show that the DNB ratios (DNBRs) encountered for the reduction in feedwater temperature at power are above the limit value at all times; therefore, no fuel or clad damage is predicted. The radiological consequences of this event are bounded by the steamline break accident analyzed in Subsection 15.1.5.3.
15.1-2b REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.1.2 Feedwater System Malfunctions Causing an Increase in Feedwater Flow 15.1.2.1 Identification of Causes and Accident Description Additions of excessive feedwater will cause an increase in core power by decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS. The overpower - overtemperature protection (neutron overpower, overtemperature and overpower T trips) prevent any power increase which could lead to a DNBR less than the limit value.
An example of excessive feedwater flow would be a full opening of one or more feedwater control valves due to a feedwater control system malfunction or an operator error. At power this excess flow causes a greater load demand on the RCS due to increased subcooling in the steam generators. With the plant at no-load conditions, the addition of an excess of feedwater may cause a decrease in RCS temperature and thus a reactivity insertion due to the effects of the negative moderator coefficient of reactivity. Excessive feedwater flow at no-load conditions results in a less severe transient than at full power.
Therefore, only the full power case was analyzed.
Continuous addition of excessive feedwater is prevented by the steam generator high-high level trip, which closes the feedwater valves.
An increase in normal feedwater flow is classified as an ANS Condition II event, a fault of moderate frequency. (See Subsection 15.0.1 for a discussion of ANS Condition II events.)
15.1-3 REVISION 7 - DECEMBER 1998
B/B-UFSAR Plant systems and equipment credited for mitigating the effects of the accident are discussed in Subsection 15.0.8 and listed in Table 15.0-7.
15.1.2.2 Analysis of Effects and Consequences Method of Analysis The excessive heat removal due to a feedwater system malfunction transient is analyzed by using the detailed digital computer code LOFTRAN (Reference 1). This code simulates a multiloop system, neutron kinetics, the pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.
The system is analyzed to demonstrate plant behavior in the event that excessive feedwater addition, due to a control system malfunction or operator error which allows one or more feedwater control valves to open fully occurs. The limiting case analyzed is the accidental opening of four feedwater control valves with the reactor in automatic control at full power.
This accident is analyzed with the revised thermal design procedure as described in WCAP-11397-P-A. Plant characteristics and initial conditions are discussed in Subsection 15.0.3. The reactivity insertion rate following a feedwater system malfunction is calculated with the following assumptions:
- a. Initial reactor power, pressure, and RCS temperatures are assumed to be at their nominal values.
Uncertainties in initial conditions are included in the limit DNBR as described in WCAP-11397-P-A.
- b. For the limiting excessive feedwater flow accident at full power, four feedwater control valves are assumed to malfunction resulting in a step increase to 129% of nominal feedwater flow to each steam generator. The temperature of the feedwater was reduced from 449.2oF to 360oF to all four steam generators.
- c. No credit is taken for the heat capacity of the RCS and steam generator thick metal in attenuating the resulting plant cooldown.
- d. The excessive feedwater flow results in rising steam generator level which is terminated by a steam generator high-high level trip signal which closes all feedwater control and isolation valves, trips the main feedwater pumps, and trips the turbine.
- e. In order to determine the most limiting excessive feedwater flow case for this accident, both of the Byron/Braidwood steam generator designs (BWI and D5) are analyzed for this event. The most limiting case of the two steam generator designs is documented and presented in the Results section that follows.
15.1-4 REVISION 16 - DECEMBER 2016
B/B-UFSAR Plant characteristics and initial conditions are further discussed in Subsection 15.0.3.
15.1-4a REVISION 9 - DECEMBER 2002
B/B-UFSAR Normal reactor control systems and engineered safety systems are not required to function. The reactor protection system may function to trip the reactor due to an overpower condition. No single active failure will prevent operation of the reactor protection system.
Results The accidental opening of four feedwater control valves at full power (maximum moderator reactivity feedback and minimum end-of-life Doppler-only power coefficients, automatic rod control) gives the largest reactivity feedback and results in the greatest power increase. Assuming the reactor to be in the manual rod control mode results in a slightly less severe transient. The rod control system is, however, not required to function for an excessive feedwater flow event.
The reactor trips on a overpower T signal. When the steam generator water level in the faulted loop(s) reach the high-high level setpoint, all feedwater isolation valves and feedwater pump discharge valves are automatically closed and the main feedwater pumps are tripped. This prevents continuous addition of the feedwater.
15.1-5 REVISION 16 - DECEMBER 2016
B/B-UFSAR Transient results for the most limiting excessive feedwater flow case, see Figures 15.1-2a and 15.1-2b, show the increase in nuclear power and T associated with the increased thermal load on the reactor. Following reactor trip and feedwater isolation, the plant will approach a stabilized condition at hot standby.
Normal plant operating procedures may then be followed. The operating procedures would call for operator action to control reactor coolant system boron concentration and pressurizer level using the CVCS and to maintain steam generator level through control of the main or auxiliary feedwater system. Any action required of the operator to maintain the plant in a stabilized condition will be in a time frame in excess of ten minutes following reactor trip.
Since the power level rises during the excessive feedwater flow event, the fuel temperatures will also rise until after the reactor trip occurs. The core heat flux lags behind the neutron flux due to fuel rod thermal time constant and the peak linear rod power reached is limited to a value below that which would result in exceeding the fuel melting temperature. Hence, fuel melting is precluded for this event.
The transient results show that DNB does not occur at any time during the excessive feedwater flow incident; thus, the ability of the primary coolant to remove heat from the fuel rod is not reduced. The fuel cladding temperature, therefore, does not rise significantly above its initial value during the transient.
The calculated sequence of events for the most limiting excessive feedwater flow case for this accident is shown in Table 15.1-1.
15.1.2.3 Radiological Consequences The radiological consequences will be less severe than the steamline break accident analyzed in Subsection 15.1.5.3.
15.1.2.4 Conclusions The results of the analysis show that the DNBRs encountered for an excessive feedwater addition at power are at all times above the limit value; hence, no fuel or clad damage is predicted.
The radiological consequences of this event will be 15.1-6 REVISION 15 - DECEMBER 2014
B/B-UFSAR less than the steam line break accident analyzed in Subsection 15.1.5.3.
15.1.3 Excessive Increase In Secondary Steam Flow 15.1.3.1 Identification of Causes and Accident Description An excessive increase in secondary system steam flow (excessive load increase incident) is defined as a rapid increase in steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. The reactor control system is designed to accommodate a 10% step load increase of a 5% per minute ramp load increase in the range of 15% to 100% of full power. Any loading rate in excess of these values may cause a reactor trip actuated by the reactor protection system. Steam flow increases greater than 10% are analyzed in Subsections 15.1.4 and 15.1.5.
This accident could result from either an administrative violation such as excessive loading by the operator or an equipment malfunction in the steam dump control or turbine speed control.
During power operation, steam dump to the condenser is controlled by reactor coolant condition signals, i.e., high reactor coolant temperature indicates a need for steam dump. A single controller malfunction does not cause steam dump; an interlock is provided which blocks the opening of the valves unless a large turbine load decrease or a turbine trip has occurred.
Protection against an excessive load increase accident is provided by the following reactor protection system (RPS) signals:
- a. Low pressurizer pressure,
- b. Overtemperature T, and
- c. Power range high neutron flux.
An excessive load increase incident is considered to be an ANS Condition II event, a fault of moderate frequency. See Subsection 15.0.1 for a discussion of Condition II events.
15.1.3.2 Analysis of Effects and Consequences Method of Analysis This accident is analyzed using the LOFTRAN Code (Reference 1).
The code simulates the neutron kinetics, reactor coolant system, (RCS) pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, steam generator safety valves, and feedwater system. The code computes pertinent plant variables including temperatures, pressures, and power level.
15.1-7 REVISION 9 - DECEMBER 2002
B/B-UFSAR Four cases are analyzed to demonstrate the plant behavior following a 10% step load increase from rated load. These cases are as follows:
- a. reactor control in manual with minimum reactivity feedback,
- b. reactor control in manual with maximum reactivity feedback,
- c. reactor control in automatic with minimum reactivity feedback, and
- d. reactor control in automatic with maximum reactivity feedback.
For the minimum reactivity feedback cases, the core has a zero moderator temperature coefficient of reactivity and therefore, the least inherent transient capability. For the maximum reactivity feedback cases, the moderator temperature coefficient of reactivity has its highest absolute value. This results in the largest amount of reactivity feedback due to changes in coolant temperature.
A conservative limit on the turbine valve opening is assumed, and all cases are studied without credit taken for pressurizer heaters.
This accident is analyzed with the revised thermal design procedure as described in WCAP-11397-P-A. Plant characteristics and initial conditions are discussed in Subsection 15.0.3.
Initial reactor power, pressure, and RCS temperatures are assumed to be at their nominal values. Uncertainties in initial conditions are included in the limit DNBR as described in WCAP-11397-P-A.
Normal reactor control systems and engineered safety systems are not required to function. The reactor protection system is assumed to be operable; however, reactor trip is not encountered for many cases due to the error allowances assumed in the setpoints. No single active failure will prevent the reactor protection system from performing its intended function.
The cases which assume automatic rod control are analyzed to ensure that the worst case is presented. The automatic rod control system is not required to mitigate the consequences of an accident.
Results Figures 15.1-3 through 15.1-6 illustrate the transient with the reactor in manual rod control mode. As expected, for the minimum reactivity feedback case there is a slight power increase, and the average core temperature shows a large decrease. This results in a DNBR which increases above its initial value. For the maximum reactivity feedback manually controlled case, there is a much 15.1-8 REVISION 6 - DECEMBER 1996
B/B-UFSAR larger increase in reactor power due to the moderator feedback.
A reduction in DNBR is experienced, but DNBR remains above the limit value. For these cases, the plant rapidly reaches a stabilized condition at the higher power level. Normal plant operating procedures would then be followed to reduce power.
Figures 15.1-7 through 15.1-10 illustrate the transient assuming the reactor is in the automatic rod control mode and no reactor trip signals occur. Both the minimum and maximum reactivity feedback cases show that core power increases, thereby reducing the rate of decrease in coolant average temperature and pressurizer pressure. For both of these cases, the minimum DNBR remains above the limit value.
The excessive load increase incident is an overpower transient for which the fuel temperatures will rise. Reactor trip may not occur for some of the cases analyzed, and the plant reaches a new equilibrium condition at a higher power level corresponding to the increase in steam flow.
Since the DNBR limit is not violated at any time during the excessive load increase transients, the ability of the primary coolant to remove heat from the fuel rod is not reduced. Thus, the fuel cladding temperature does not rise significantly above its initial value during the transient.
The calculated sequence of events for the excessive load increase incident is shown on Table 15.1-1.
As can be seen from Figures 15.1-4, 15.1-6, 15.1-8, and 15.1-10, for the excessive increase in secondary steam flow accident analysis, the conservative assumptions made in this analysis result in no reactor trip for all cases analyzed. The plant analyses show that a new steady-state condition is reached.
15.1.3.3 Radiological Consequences There are no radiological consequences associated with this event and activity is contained within the fuel rods and reactor coolant system within design limits.
15.1.3.4 Conclusions The analysis discussed above shows that for a 10% step load increase, the DNBR remains above the limit value, thereby precluding fuel or clad damage. The plant rapidly reaches a stabilized condition following the load increase.
15.1-9 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.1.4 Inadvertent Opening of a Steam Generator Relief or Safety Valve The inadvertent opening of a steam generator relief or safety valve event (i.e., the credible steamline break) creates a depressurization of the secondary side with an effective opening size that is within the spectrum of break sizes analyzed by the hypothetical steamline break event. Therefore, the credible steamline break is bounded by the hypothetical steamline break discussed in Subsections 15.1.5 and 15.1.6.
15.1-10 REVISION 7 - DECEMBER 1998
B/B-UFSAR Pages 15.1-11 through 15.1-13 have been deleted intentionally.
15.1-11 through 15.1-13 REVISION 7 - DECEMBER 1998
B/B-UFSAR 15.1.5 Steam System Piping Failure at Zero Power 15.1.5.1 Identification of Causes and Accident Description The steam release arising from a break of a main steamline would result in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive rod cluster control assembly (RCCA) is assumed stuck in its fully withdrawn position after reactor trip, there is an increased possibility that the core will become critical and return to power. A return to power following a steamline break is a potential problem mainly because of the high power peaking factors which exist assuming the most reactive RCCA to be stuck in its fully withdrawn position. The core is ultimately shut down by the boric acid injection delivered by the safety injection system.
The analysis of a main steamline break is performed to demonstrate that the following criteria are satisfied:
- a. Assuming a stuck RCCA with or without offsite power, and assuming a single failure in the engineered safety features, the core remains in place and intact.
Radiation doses shall meet the applicable regulatory criteria.
- b. Although DNB and possible clad perforation following a steam pipe break are not necessarily unacceptable, the following analysis, in fact, shows that DNBR never falls below the analysis limit for any break assuming the most reactive assembly stuck in its fully withdrawn position. The DNBR design basis is discussed in Section 4.4.
A major steamline break is classified as an ANS Condition IV event. See Subsection 15.0.1 for a discussion of Condition IV events.
Effects of minor secondary system pipe breaks are bounded by the analysis presented in this section. Minor secondary system pipe breaks are classified as Condition III events, as described in Subsection 15.0.1.3.
15.1-14 REVISION 15 - DECEMBER 2014
B/B-UFSAR The major break of a steamline is the most limiting cooldown transient and is analyzed at zero power with no decay heat.
Decay heat would retard the cooldown thereby reducing the return to power. A detailed analysis of this transient with the most limiting break size, a double ended break, is presented here.
The following functions provide the protection for a steamline break:
- a. Safety injection system actuation from any of the following:
- 1. Two-out-of-three low steamline pressure signals in any one loop
- 2. Two-out-of-four low pressurizer pressure signals.
- 3. Two-out-of-three high-1 containment pressure signals.
- b. The overpower reactor trips (neutron flux and T) and the reactor trip occurring in conjunction with receipt of the safety injection signal.
- c. Redundant isolation of the main feedwater lines.
Sustained high feedwater flow would cause additional cooldown. Therefore, in addition to the normal control action which will close the main feedwater valves a safety injection signal will rapidly close all feedwater control valves and backup feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.
- d. Trip of the fast acting steamline stop valves (designed to close in less than 5 seconds) on:
- 1. Two-out-of-three low steamline pressure signals in any one loop.
- 2. Two-out-of-three high-2 containment pressure signals.
- 3. Two-out-of-three high negative steamline pressure rate signals in any one loop (used only during cooldown and heatup operations).
The blocking of safety injection from the low pressurizer pressure and/or low steamline pressure signals is permitted following receipt of the P-11 permissive to allow the plant to be intentionally cooled down without the initiation of safety injection. To ensure that the hot zero power steamline break analysis is bounding when these automatic signals are blocked, the RCS must be borated to ensure subcriticality at 200F. Prior to manually blocking these automatic signals, meeting the subcriticality boron concentration for 200F and meeting the normal shutdown margin boron concentration at the current core conditions are both required.
15.1-15 REVISION 10 - DECEMBER 2004
B/B-UFSAR For breaks downstream of the isolation valves, closure of all valves would completely terminate the blowdown. For any break, in any location, no more than one steam generator would experience an uncontrolled blowdown even if one of the isolation valves fails to close. A description of steamline isolation is included in Chapter 10.0.
15.1-15a REVISION 10 - DECEMBER 2004
B/B-UFSAR Steam flow is measured by monitoring dynamic head in nozzles located in the throat of the steam generator. The effective throat area of the nozzles is 1.1 square feet for Unit 1 and 1.4 square feet for Unit 2, which is considerably less than the main steam pipe area; thus, the nozzles also serve to limit the maximum steam flow for a break at any location.
Table 15.1-2 lists the equipment required in the recovery from a high energy line break. Not all equipment is required for any one particular break, since the requirements will vary depending upon postulated break location and details of balance of plant design and pipe break criteria as discussed elsewhere in this application. Design criteria and methods of protection of safety-related equipment from the dynamic effects of postulated piping breaks are provided in Section 3.6.
15.1.5.2 Analysis of Effects and Consequences Method of Analysis The analysis of the steam pipe break has been performed to determine:
- a. The core heat flux and RCS temperature and pressure resulting from the cooldown following the steamline break. The LOFTRAN code (Reference 1) has been used.
- b. The thermal and hydraulic behavior of the core following a steamline break. A detailed thermal and hydraulic digital-computer code, VIPRE, has been used to determine if DNBR falls below the safety analysis limit for the core conditions computed in item a above.
The following conditions were assumed to exist at the time of a main steam break accident:
- a. End-of-life shutdown margin at no-load, equilibrium xenon conditions, and the most reactive RCCA stuck in its fully withdrawn position. Operation of the control rod banks during core burnup is restricted in such a way that addition of positive reactivity in a steamline break accident will not lead to a more adverse condition than the case analyzed.
- b. A negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive RCCA in the fully withdrawn position. The variation of the coefficient with temperature and pressure has been included. The Keff versus temperature at 1150 psi corresponding to the negative moderator temperature coefficient used is shown in Figure 15.1-11. The effect of power generation in the core on overall reactivity is shown in Figure 15.1-14.
15.1-16 REVISION 15 - DECEMBER 2014
B/B-UFSAR The core properties associated with the sector nearest the affected steam generator and those associated with the remaining sector were conservatively combined to obtain average core properties for reactivity feedback calculations.
Further, it was conservatively assumed that the core power distribution was uniform. These two conditions cause underprediction of the reactivity feedback in the high power region near the stuck rod. To verify the conservatism of this method, the reactivity as well as the power distribution was checked for the limiting statepoints for the cases analyzed.
This core analysis, performed with the ANC code, considered the Doppler reactivity from the high fuel temperature near the stuck RCCA, moderator feedback from the high water enthalpy near the stuck RCCA, power redistribution and non-uniform core inlet temperature effects. For cases in which steam generation occurs in the high flux regions of the core, the effect of void formation was also included.
It was determined that the reactivity employed in the kinetics analysis was always larger than the reactivity calculated including the above local effects for the statepoints. These results verify conservatism; i.e., underprediction of negative reactivity feedback from power generation.
- c. Minimum capability for injection of concentrated boric acid (2,300 ppm) solution corresponding to the most restrictive single failure in the high head safety injection (HHSI) system. The emergency core cooling system (ECCS), consists of three systems:
(1) the passive accumulators, (2) the residual heat removal system (RHRS), and (3) the low head safety injection system (LHSIS), and the HHSI system. Only the HHSI system is modeled for the steamline break accident analysis.
The actual modeling of the HHSI system in LOFTRAN is described in Reference 1. The flow corresponds to that delivered by one charging pump delivering its full flow to the cold leg header. No credit has been taken for the low concentration borated water, which must be swept from the lines downstream of the refueling water storage tank prior to the delivery of concentrated boric acid to the reactor coolant loops.
For the cases where offsite power is assumed, the sequence of events in the HHSI system is the following. After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the charging pump starts. In 17 seconds, the 15.1-17 REVISION 15 - DECEMBER 2014
B/B-UFSAR valves are assumed to be in their final position and the pump is assumed to be at full speed. This does not include sequential transfer of high head safety injection pump suction from the VCT to the RWST. The additional 10 seconds for valves CV112B and C to close after CV112D and E are open has been evaluated and is consistent with the accident analysis results.
Transfer of the pump suction would be completed in 27 seconds. The volume containing the low concentration borated water is swept before the 2,300 ppm borated water reaches the core. This delay, described above, is inherently included in the modeling.
In cases where offsite power is not available, an additional 13-second delay is assumed to start the diesels and to load the necessary safety injection equipment onto them.
- d. Design value of the steam generator heat transfer coefficient including allowance for fouling factor.
- e. Since the steam generators are provided with integral flow restrictors with a 1.1 square foot throat area for Unit 1 and a 1.4 square foot throat area for Unit 2, any break with a break area greater than the area of the flow restrictor, regardless of location, would have the same effect on the NSSS as the break equal to the area of the flow restrictor. The following cases have been considered in determining the core power and RCS transients:
Case 1: Complete severance of a pipe, with the plant initially at no-load conditions, full reactor coolant flow with offsite power available.
Case 2: Case 1 with loss of offsite power coincident with the steamline break. Loss of offsite power results in reactor coolant pump coastdown, which is assumed to begin at 3 seconds.
- f. Power peaking factors corresponding to one stuck RCCA and nonuniform core inlet coolant temperatures are determined at end of core life. The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod. The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return to power phase following the steamline break.
This void in conjunction with the large negative moderator coefficient partially offsets the effect of the stuck assembly. The power peaking factors depend upon the core power, temperature, pressure, and flow, and, thus, are different for each case studied.
15.1-18 REVISION 9 - DECEMBER 2002
B/B-UFSAR The core parameters used for both with and without offsite power cases correspond to values determined from the respective transient analysis.
Both cases above assume initial hot shutdown conditions at time zero since this represents the most pessimistic initial condition. The hot shutdown initial conditions were considered for cases assuming full power operation at both the high (588.0oF) and low (575.0oF) HFP Tavg conditions. Should the reactor be just critical or operating at power at the time of a steamline break, the reactor will be tripped by the normal overpower protection system when power level reaches a trip point. Following a trip at power, the RCS contains more stored energy than at no-load, the average coolant temperature is higher than at no-load and there is appreciable energy stored in the fuel.
Thus, the additional stored energy is removed via the cooldown caused by the steamline break before the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load condition at time zero. A spectrum of steamline breaks at various power levels has been analyzed in Reference 4.
- g. In computing the steam flow during a steamline break, the Moody Curve (Reference 3) for f(L/D) = 0 is used.
- h. Perfect moisture separation in the steam generator is assumed.
These assumptions are discussed more fully in Reference 4.
Results The calculated sequence of events for the limiting case (Unit 2, low Tavg, offsite power available) is shown in Table 15.1-1.
The results presented are a conservative indication of the events which would occur assuming a steamline break since it is postulated that all of the conditions described above occur simultaneously.
Core Power and Reactor Coolant System Transient Figures 15.1-18 through 15.1-20 for Unit 2 show the RCS transient and core heat flux following a main steamline break (complete severance of a pipe) at initial no-load condition.
Offsite power is assumed available so that full reactor coolant flow exists. The transient shown assumes an uncontrolled steam release from only one steam generator. Should the core be critical at near zero power when the break occurs, the initiation of safety injection by low steamline pressure will 15.1-19 REVISION 9 - DECEMBER 2002
B/B-UFSAR trip the reactor. Steam release from more than one steam generator will be prevented by automatic trip of the fast acting isolation valves in the steamlines by low steamline pressure signals, high containment pressure signals, or high negative steamline pressure rate signals. Even with the failure of one valve, release is limited to no more than 10 seconds for the other steam generators while the one generator blows down. The steamline stop valves are designed to be fully closed in less than 5 seconds from receipt of a closure signal.
As shown in Figure 15.1-20 the core attains criticality with the RCCAs inserted (with the design shutdown assuming one stuck RCCA) before boron solution at 2,300 ppm enters the RCS. A peak core power lower than the nominal full power value is attained.
The calculation assumes the boric acid is mixed with, and diluted by, the water flowing in the RCS prior to entering the reactor core. The concentration after mixing depends upon the relative flow rates in the RCS and in the HHSI system. The variation of mass flow rate in the RCS due to water density changes is included in the calculation as is the variation of flow rate in the HHSI system due to changes in the RCS pressure. The HHSI system flow calculation includes the line losses in the system as well as the pump head curve.
The loss of offsite power case corresponds to the case discussed above with additional loss of offsite power at the time the safety injection signal is generated. The safety injection system delay time includes 13 seconds to start the diesel in addition to 17 seconds to start the safety injection pump and open the valves. An additional 10 seconds is required to close valves CV112B and C after CV112D and E are open to transfer the high head safety injection pump suction from the VCT to the RWST.
This additional 10 second delay has been evaluated and is consistent with the accident analysis results. In 40 seconds, the diesel and pump are assumed to start and the valves are assumed to be in their final position with the pump suction transferred from the VCT to the RWST. Criticality is achieved later and the core power increase is slower than in the similar case with offsite power available. The ability of the emptying steam generator to extract heat from the RCS is reduced by the decreased flow in the RCS. The peak power remains well below the nominal full power value.
It should be noted that following a steamline break only one steam generator blows down completely. Thus, the remaining steam generators are still available for dissipation of decay heat after the initial transient is over. In the case of loss of offsite power this heat is removed to the atmosphere via the steamline safety valves.
15.1-20 REVISION 9 - DECEMBER 2002
B/B-UFSAR Margin to Critical Heat Flux DNB analyses were performed for both units, with and without offsite power and High and Low Tavg programs. The minimum DNBR is greater than the limit value in all cases and the limiting case is that for Unit 2 (D5 SGs), low HFP Tavg case, with offsite power available. The results of this case are presented herein.
15.1.5.3 Radiological Consequences of a Postulated Steamline Break Using AST The key inputs and assumptions used in the Main Steam Line Break (MSLB) radiological consequence analysis are summarized below and provided in Table 15.1-3. Although this analysis is presented in the section describing a steam system piping failure at zero power, it is also applicable for the full power event (described in section 15.1.6).
The MSLB accident is postulated as a break of one of the large steam lines leading from a steam generator. This break results in the release of radioactive material from the Byron and Braidwood containment system. For the three intact Steam Generator (SG) loops, primary to secondary coolant leakage transfers activity into the Secondary Coolant. This makes it available for release into the environment via steaming through the SG Power-Operated Release Valves (PORV). For the coolant loop with the broken steam line (referred to as the faulted steam generator), primary to secondary coolant leakage is assumed to be released from the RCS directly into the environment without passing through any secondary coolant. This is due to assumed "dry-out" conditions in the faulted steam generator. Consistent with Regulatory Guide (RG) 1.183, two reactor transients that maximize the radioactivity available for release were modeled.
In addition to these two transients, the release of the maximum allowed operational concentration of iodine activity in the secondary coolant system, 0.1 µCi/gm is analyzed. This Case simulates the initial blowdown of all fluid in the faulted SG (assuming a 2-minute duration), and the PORV release of secondary coolant activity of the intact SGs. The dose consequence of this simulation is added to each of the other modeled cases.
Case 1: Dose Due to Pre-accident Iodine Spike The first case involves a 60 uCi/gm pre-accident Iodine spike.
This 60 µCi/gm spike is consistent with the Technical Specification operational Reactor Coolant System (RCS) activity concentration limit for an assumed spike. In this scenario, it is assumed that all of the spike activity is homogeneously mixed in the primary coolant, prior to accident initiation.
15.1-21 REVISION 12 - DECEMBER 2008
B/B-UFSAR Case 2: Dose Due to Accident Initiated Concurrent Iodine Spike The second case involves an accident initiated iodine spike that occurs concurrently with the release of fluid from the primary and secondary coolant systems. Regulatory guidance specifies that this spike should result in a release rate from the operating limit defective fuel fraction that is 500 times the normal rate.
Case 3: Dose Due to Equilibrium Secondary Coolant System Iodine The third case simulates the dose contribution that results from the initial blowdown of all fluid in the faulted SG (assuming a 2-minute duration), and the PORV release of secondary coolant through the intact SGs. These releases of specifically secondary coolant activity, existing prior to the MSLB accident, are analyzed, and the dose is added to each of the other modeled cases.
Fuel Damage and Core Source Term The design basis assumes no fuel damage for the postulated main steam line break event. For this MSLB accident, the source terms are defined by the Technical Specification activity release rates from a maximum failed fuel fraction assumed during operation, which are characterized by the equilibrium 1.0 µCi/gm Dose Equivalent(DE) I-131 iodine activity concentration in the primary reactor coolant system. The noble gas inventory in the RCS is based on operation with a conservative worst-case 1% core fuel defects. Because no fuel damage is assumed for this accident, only iodine and noble gas isotopes are modeled to contribute to dose, as given in Table 15.0-16. To identify the worst-case MSLB accident, however, two different cases of iodine spiking are analyzed, per regulatory guidance.
Case 1: Pre-Accident Iodine Spike Source Terms The first case is simply identified as a reactor pre-accident, transient induced, iodine spike, which raises the primary coolant iodine concentration to the maximum 60 µCi/gm DE I-131 value permitted by Technical Specifications at full power operations, prior to the initiation of the accident. Therefore this case is termed the pre-accident iodine spike case.
15.1-21a REVISION 12 - DECEMBER 2008
B/B-UFSAR Case 2: Concurrent Iodine Spike Source Terms The second case assumes that the postulated MSLB event causes a primary reactor system transient. This transient, in turn, is associated with an iodine spike which assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the release rate corresponding to the 1.0 µCi/gm DE I-131 equilibrium iodine concentration as given in the Technical Specifications. This 500 times activity release rate spike is assumed to occur for a duration of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, as this period has been shown to conservatively deplete the available gap activity in the assumed operating damaged fuel fraction. Also, this assumption has historically been used as the design basis for this accident at Byron and Braidwood. In RADTRAD a Nuclide Inventory File (NIF) is designed to input the total isotopic iodine activity that is associated with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of activity release at the 500 times rate specified. Then, this NIF is used in conjunction with a modified Release Fraction and Timing (RFT) file, which defines the complete release of this activity over a 6-hour period.
Case 3: Equilibrium Secondary Coolant System Iodine Source Terms The case 3 source term consists simply of the 0.1 µCi/gm DE I-131 equilibrium secondary coolant activity concentration limit in Technical Specifications.
Activity Removal Mechanisms in Containment The design basis MSLB releases activity directly into the primary RCS, therefore no plateout, or other activity deposition, is credited.
Decay Credited:
Decay of radioactivity is credited in all compartments, prior to release. This is implemented in RADTRAD using the half-lives in the NIFs. The RADTRAD decay option is used.
Depletion from Leakage Credited:
For analyses of doses due to release from the RCS volume, the dose results from leakage. It is reasonable to credit the small amount of depletion from the available RCS activity inventory associated with this leakage. This is calculated inherently by the RADTRAD code.
15.1-21b REVISION 12 - DECEMBER 2008
B/B-UFSAR Release Rates, Steaming Rates, and Partitioning Factors:
Activity that originates in the primary RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. This design basis leak rate value is 0.218 gpm, per intact SG, totaling 0.654 gpm, and 0.5 gpm for the faulted SG with the broken steam line. For input into RADTRAD these rates were converted from gallons per minute to cubic feet per minute, making them 0.02914 cfm, per intact SG, totaling 0.08743 cfm, and 0.06684 cfm for the faulted SG.
Primary to secondary coolant leakage through the faulted steam generator conservatively goes directly to the environment, without mixing with any secondary coolant. Therefore, under the assumed dry-out conditions, no partitioning of any nuclides is expected to occur in this release pathway.
For all post-accident releases through the PORVs of the intact SG loops, the mechanism for release to the environment is steaming of the secondary coolant. Because of this release dynamic, RG 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water. For Iodine, the partitioning factor of 0.01 was taken directly from the suggested guidance of RG 1.183. Reviewing the specified AST release fractions, it is concluded that the only nuclides other than iodines to be released from the core source term are noble gas nuclides.
Because of the volatility of noble gases, no partitioning is assumed for any such isotopes.
The methodology used to model steaming of activity through PORVs following the postulated MSLB event, assumes an average cumulative release rate through the SG valves that is reduced in steps. The partitioning factors are applied to these release rates, which were derived from the total time increment mass releases. Incremental steam mass releases are in pounds.
Release rates were derived by dividing these totals by the time increment. This data was then converted using the assumption of cooled liquid conditions (i.e., 62.4 lbm/ft3), as specified by the applicable guidance of RG 1.183. The steaming release and primary-to-secondary coolant leakage is postulated to end at 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, when the RCS and secondary loop have equilibrated. The following table below shows the time steps, isotopic partitioning factors, and associated release rates:
15.1-21c REVISION 12 - DECEMBER 2008
B/B-UFSAR MSLB Partitioning Factors And Associated Release Rates Time Total Iodine Noble Gas Steam Steam Interval Steam Partitioning Partitioning Release Release (hrs) Mass Factor Factor Rate for Rate for Release Iodines Noble Gases (lbm) (cfm) (cfm) 0 - 2.0 447,000 0.01 1 5.9696E-01 5.9696E+01 2.0 - 40 3,279,000 0.01 1 2.3047E-01 2.3047E+01 For the loop with the broken steam line, i.e., the faulted SG, it is postulated that the entire release of the secondary coolant of that loop will take 2 minutes. Therefore, for input into RADTRAD the faulted SG coolant volume of 2675 ft3 is divided by 2 minutes to arrive at a design basis value of 1.3375E+03 cfm.
/Q Calculations (Meteorology)
Releases from the SG PORVs were considered elevated releases due to the high steaming rates and the associated /Qs were reduced by a factor of 5 per guidance in RG 1.194, as described in Section 15.4.8.3. The atmospheric dispersion factors are given in Table 15.0-17.
This event was reanalyzed using the finer wind speed categories of Regulatory Guide 1.23, Revision 1.
Assumptions and Inputs The following inputs and assumptions were used in the MSLB analysis.
- a. Core inventory is based on a DBA power level of 3658.3 MWt. This power level bounds the MUR power uprate Rated Thermal Power level including measurement uncertainties.
- b. There is no fuel damage as a result of the postulated main steam line break accident.
- c. In the case of a postulated Iodine activity release rate spike, the spike release is assumed to occur for a period of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, when the activity available for release from the fuel has been conservatively depleted.
- d. The activity released from the fuel is assumed to be instantaneously mixed with the RCS.
- e. All iodine released from the SGs is assumed to be of the elemental species. This is done for RADTRAD simulation considerations, and is consistent with the RG 1.183 specification of 97% elemental and 3%
organic, because elemental and organic iodine are treated identically by the computer model.
15.1-21d REVISION 16 - DECEMBER 2016
B/B-UFSAR
- f. The Control Room HVAC system is realigned to the emergency mode of operation 30 minutes after the initiation of this design basis accident.
- g. The faulted steam generator is assumed to be in a "dry-out" condition, and does not inhibit activity release from the RCS through that coolant loop.
- h. It is conservatively assumed that blowdown of the faulted steam generator 167,000 lbm fluid takes two minutes to complete.
Dose Results Radiological doses resulting from a design basis MSLB for a control room operator and a person located at EAB or LPZ are to be less than the regulatory dose limits as given below.
Regulatory Dose Limits - MSLB Dose Type Control Room EAB and LPZ (rem) (rem)
Case 1 TEDE Dose 5a 25b Case 2 TEDE Dose 5a 2.5b Notes:
a 10 CFR 50.67 b
10 CFR 50.67 as modified by Regulatory Guide 1.183 (Table 6, Page 1.183-20)
The table below provides the results from the Case 1 and Case 2 simulations that were modeled using the RADTRAD 3.03 code. The total dose for the two cases includes the dose result from the Case 3 simulations. Therefore for convenience, the doses shown in the Table include the Case 3 dose.
Main Steam Line Break Accident Radiological Analysis Results Case 1: Pre-Accident 60 uCi/gm DE I-131 Spike Dose Assessment Results Control Room EAB LPZ (REM TEDE) (REM TEDE) (REM TEDE) 0.264 0.146 0.083 Case 2: Accident Initiated 500 times Equilibrium Iodine Release Rate Spike Dose Assessment Results Control Room EAB LPZ (REM TEDE) (REM TEDE) (REM TEDE) 0.654 0.201 0.459 15.1-21e REVISION 16 - DECEMBER 2016
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.1-22 REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.1.5.4 Conclusions The analysis has shown that the criteria stated in Subsection 15.1.5.1 are satisfied for operation of all units at the uprated power conditions. Although DNB and possible cladding perforation following a steam pipe break are not necessarily unacceptable and not precluded by the criteria, the above analysis, in fact, shows that the DNB design basis is met as stated in Section 4.4.
The radiological consequences of this event are within the dose acceptance criteria of 10 CFR 50.67 and Regulatory Guide 1.183.
15.1.6 Steam System Piping Failure at Full Power 15.1.6.1 Identification of Causes and Accident Description The steam system piping failure accident analysis described in Subsection 15.1.5 was performed assuming hot zero power conditions with the control rods fully inserted in the core with the exception for the most reactive rod. Such a condition could occur while the reactor is in hot shutdown at the minimum required shutdown margin or after the plant has been tripped automatically by the reactor protection system or manually by the operator. For an at-power steamline break, the analysis of Subsection 15.1.5 represents the limiting condition with respect to core protection for the period following reactor trip.
Analysis of a steam system piping failure occurring from at-power initial conditions is performed to demonstrate that core protection is maintained prior to and immediately following reactor trip.
Depending on the size of the break, this event is classified as either a Condition III (infrequent fault) or Condition IV (limiting fault) event. The acceptance criteria for this event are defined in Subsection 15.0.1.
15.1.6.2 Analysis of Effects and Consequences Method of Analysis The analysis of the steamline break at-power was performed as follows:
- a. The LOFTRAN code (Reference 1) was used to calculate the nuclear power, core heat flux, and reactor coolant system temperature and pressure transients resulting from the cooldown following the steamline break.
- b. The core radial and axial peaking factors were determined using the thermal-hydraulic conditions from LOFTRAN as input to the nuclear core models. A detailed thermal-hydraulic code, VIPRE, was used to calculate the DNBR for the limiting time during the transient.
15.1-23 REVISION 15 - DECEMBER 2014
B/B-UFSAR The analysis was performed with the revised thermal design procedure as described in WCAP-11397-P-A (Reference 8). Plant characteristics and initial conditions area discussed in Subsection 15.0.3.
Assumptions
- a. Initial conditions - The initial core power, reactor coolant temperature, and reactor coolant system pressure were assumed to be at their nominal full-power values at uprated power conditions. Cases assuming full power operation at the high (588oF) HFP Tavg conditions with uniform initial loop flow are analyzed.
In addition, cases assuming low (575 oF) HFP Tavg and asymmetric initial loop flow conditions are considered and determined to be non-limiting. The asymmetric flow cases assume a maximum 5% loop-to-loop asymmetric flow variation.
- b. Break size - The limiting break size was calculated to be 0.95 ft2 for Unit 1. The results for this case bound all other break sizes for both Unit 1 and Unit 2.
- c. Break flow - In computing the steam flow during a steamline break, the Moody curve for f(L/D)=0 is used.
- d. Reactivity coefficients - The analysis assumed maximum moderator reactivity feedback and least negative Doppler power feedback to maximize the power increase following the break.
- e. Protection system - The protection system features that mitigate the effects of a steamline break are described in Subsection 15.1.5. This analysis only considers the initial phase of the transient from at-power conditions. Protection in this phase of the transient is provided by reactor trip, if necessary. Subsection 15.1.5 presents the analysis of the bounding transient following reactor trip, where other protection system features are actuated to mitigate the effects of the steamline break.
- f. Control systems - The pressurizer sprays are modeled to minimize RCS pressure, which is conservative with respect to DNBR. Other control systems were not credited in mitigating the effects of the transient since doing so would not make the results more limiting.
15.1-23a REVISION 16 - DECEMBER 2016
B/B-UFSAR Results The sequence of events for the limiting case (Unit 1) is shown in Table 15.1-1. Although a break spectrum was analyzed, plots from only one break size (0.95 ft2 break) are shown. Figures 15.1-27 through 15.1-29 show the transient responses for Unit 1.
Conclusions The 0.95 ft2 break with symmetric RCS flow for Unit 1 is the most limiting case for both kW/ft and DNB considerations. The Unit 1 results bound the results of Unit 2.
For radiological consequences of a postulated steamline break, see section 15.1.5.3.
15.1.7 References
- 1. Burnett, T. W. T., et al., "LOFTRAN Code Description,"
WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary),
April 1984.
- 2. Deleted.
- 3. Moody, F. S., "Transactions of the ASME," Journal of Heat Transfer, Figure 3, page 134, February 1965.
- 4. Scherder, W.J. (Editor), et. al. "Reactor Core Response to Excessive Secondary Steam Releases," WCAP-9226-P-A, Revision 1, (Proprietary), February 1998, and WCAP-9227, Revision 1, (Non-Proprietary), January 1978.
- 5. Deleted.
- 6. Deleted.
- 7. NUREG-0800, "Standard Review Plan," Subsection 15.1.5, Appendix A, "Radiological Consequences of Main Steamline Failure Outside Containment for a PWR," Revision 2, July 1981.
- 8. Freidland, A. J., et al., "Revised Thermal Design Procedure,"
WCAP-11397-P-A (Proprietary) and WCAP-11397-A (Nonproprietary), April 1989.
15.1-23b REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.1-1 TIME SEQUENCE OF EVENT FOR INCIDENTS WHICH CAUSE AN INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM TIME ACCIDENT EVENT (SEC.)
Limiting Excessive FW Four main feedwater valves fail 0.0 Flow Case open Multi-Loop Excessive Overpower T reactor trip 36.0 FW Flow to Model BWI setpoint reached steam generator -
automatic rod control Rod motion occurs 44.0 Minimum DNBR occurs 44.5 High-High Steam Generator Water 69.5 Level Setpoint Reached Feedwater isolation occurs 76.5 Limiting Feedwater Feedwater heater bypass valves 0.0 Temperature Reduction fail open and a loss of multiple Case trains of feedwater heaters occurs Feedwater Temperature Overpower T reactor trip 5.9 Reduction to Model D5 setpoint reached steam generator -
manual rod control Rod motion occurs 13.9 Minimum DNBR occurs 14.5 Low Pressurizer Pressure SI 34.8 setpoint reached Feedwater isolation occurs 41.8 15.1-24 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.1-1 (Contd)
TIME ACCIDENT EVENT (SEC.)
Excessive Increase in Secondary Steam Flow
- 1. Manual Reactor 10% step load increase 0.0 Control (Minimum moderator feedback)
Equilibrium conditions 300 reached (approximate time only)
- 2. Manual Reactor 10% step load increase 0.0 Control (Maximum moderator feedback)
Equilibrium conditions 150 reached (approximate time only)
- 3. Automatic Reactor 10% step load increase 0.0 Control (Minimum moderator feedback)
Equilibrium conditions 300 reached (approximate time only)
- 4. Automatic Reactor 10% step load increase 0.0 Control (Maximum moderator feedback)
Equilibrium conditions 200 reached (approximate time only)
Steam system piping failure at zero power
- 1. Unit 2, low Tavg Steamline breaks 0.0 (Offsite power available) Criticality attained 26.8 Pressurizer empties ~24.8 Boron reaches core ~132.8 15.1-24a REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.1-1 (Cont'd)
TIME ACCIDENT EVENT (sec.)
Steam system piping failure at full power
- 1. Unit 1 - 0.95 ft2 Steamline breaks 0.0 break with uniform flow Overpower T reactor 8.46 trip setpoint reached Rods begin to drop 16.46 Peak core heat flux 17.10 occurs 15.1-25 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.1-2 EQUIPMENT REQUIRED FOLLOWING A BREAK OF A MAIN STEAM LINE SHORT TERM (REQUIRED FOR MITIGATION OF ACCIDENT) HOT STANDBY REQUIRED FOR COOLDOWN Reactor trip and safeguards Auxiliary feedwater system Steam generator power actuator channels including including pumps, water supply, operated relief valves sensors, circuitry, and and system valves and piping (can be manually operated processing equipment (the (this system must be placed locally).
protection circuits used in service to supply water to to trip the reactor on operable steam generators no undervoltage, underfrequency, later than 10 minutes after and turbine trip the incident).
may be excluded).
HHSI system including the Reactor containment Residual heat removal system pumps, the refueling water ventilation cooling units. including pumps, heat storage tank, and the exchanger, and system valves systems valves and piping. Capability for obtaining a and piping necessary reactor coolant system sample. to cool and maintain the reactor coolant system in a cold shutdown condition.
Standby diesel generators and Class IE power distribution equipment.
Essential service water and plant component cooling water system.
Containment safeguards cooling equipment.
15.1-26 REVISION 3 - DECEMBER 1991
B/B-UFSAR TABLE 15.1-2 (Cont'd)
SHORT TERM (REQUIRED FOR MITIGATION OF ACCIDENT) HOT STANDBY REQUIRED FOR COOLDOWN Auxiliary Feedwater System including pumps, water supplies, piping and valves.
Main feedwater control valves (trip closed feature).
Bypass feedwater control valves (trip closed feature).
Primary and secondary safety valves.
Circuits and/or equipment required to trip the main feedwater pumps.
Main feedwater isolation valves (trip closed feature).
Main steam line stop valves (trip closed feature).
Main steam line stop valve bypass valves (trip closed feature).
Steam generator blowdown isolation valves (automatic closure feature).
15.1-27
B/B-UFSAR TABLE 15.1-2 (Cont'd)
SHORT TERM (REQUIRED FOR MITIGATION OF ACCIDENT) HOT STANDBY REQUIRED FOR COOLDOWN Batteries (Class 1E).
Control Room air conditioning.
Control Room equipment must not be damaged to an extent where any equipment will be spuriously actuated or any of the equipment contained elsewhere in this list cannot be operated.
Post Accident Monitoring Systema.
ESF and HHSI/charging pump cubicle unit coolers a
See Section 7.5 for a discussion of the Postaccident Monitoring System.
15.1-28
B/B-UFSAR TABLE 15.1-3 INPUT PARAMETERS FOR THE MSLB RADIOLOGICAL CONSEQUENCE ANALYSIS USING AST Parameter Unit Value Notes Steam Released The total leakage to be evenly to Environment: divided for the four steam generators is 1 gpm. For Faulted SG lbm 167,000 events involving a faulted (in 2 minutes) steam generator, 0.5 gpm leak rate shall be used for faulted Intact SGs: generator and 0.218 gpm shall 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> lbm 447,000 be used for each of the intact generators.
Assumed to be based on water 2 - 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> lbm 3,279,000 at cold conditions.
Primary to Secondary Leakage:
Faulted SG gpm 0.5 Intact SG (each) gpm 0.218 Duration of hr 40 steam releases from intact SGs Duration of hr 40 activity release due to leakage of primary coolant to the faulted SG Accident Iodine Factor of 500 In addition to pre-accident Spike iodine spike case.
Primary and secondary coolant volumes:
Primary gm 2.063E8 Secondary:
3 Intact SGs gm 1.000E8 Faulted SG gm 7.575E7 15.1-29 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.1-3 INPUT PARAMETERS FOR THE MSLB RADIOLOGICAL CONSEQUENCE ANALYSIS USING AST (continued)
Parameter Unit Value Notes Noble gas These values, based on releases through operation with 1% fuel the faulted SG defects.
due to primary The values are also consistent to secondary with a RCS volume of 2.063E8 leakage: gm.
KR-85m Curies 3.713E2 KR-85 Curies 1.467E3 KR-87 Curies 2.372E2 KR-88 Curies 6.911E2 XE-131m Curies 6.829E2 XE-133m Curies 7.530E2 XE-133 Curies 5.178E4 XE-135m Curies 1.007E2 XE-135 Curies 1.593E3 XE-138 Curies 1.368E2 15.1-29a REVISION 12 - DECEMBER 2008
B/B-UFSAR This page has been intentionally deleted.
15.1-30 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.1-3a 15.1-30a REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.1-4 Table 15.1-4 has been deleted intentionally.
15.1-31 REVISION 7 - DECEMBER 1998
BYRON-UFSAR TABLE 15.1-4a This page has been intentionally deleted.
15.1-32 REVISION 9 - DECEMBER 2002
BRAIDWOOD-UFSAR TABLE 15.1-4a 15.1-33 REVISION 9 - DECEMBER 2002
BRAIDWOOD-UFSAR TABLE 15.1-4b This page have been intentionally deleted.
15.1-33a REVISION 7 - DECEMBER 1998
B/B-UFSAR 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM A number of transients and accidents have been postulated which could result in a reduction of the capacity of the secondary system to remove heat generated in the reactor coolant system (RCS). These events are discussed in this section. Detailed analyses are presented for several such events which have been identified as more limiting than the others.
Discussions of the following RCS coolant heatup events are presented in Section 15.2:
- a. steam pressure regulator malfunction,
- b. loss of external load,
- c. turbine trip,
- d. inadvertent closure of main steam isolation valves,
- e. loss of condenser vacuum and other events resulting in turbine trip,
- f. loss of nonemergency a-c power to the station auxiliaries,
- g. loss of normal feedwater flow, and
- h. feedwater system pipe break.
The above items are considered to be ANS Condition II events, with the exception of a feedwater system pipe break, which is considered to be an ANS Condition IV event. Subsection 15.0.1 contains a discussion of ANS classification and applicable acceptance criteria.
15.2.1 Steam Pressure Regulator Malfunction or Failure That Results in Decreasing Steam Flow There are no pressure regulators whose failure or malfunction could cause a steam flow transient.
15.2.2 Loss of External Load 15.2.2.1 Identification of Causes and Accident Description A major load loss on the plant can result from loss of external electrical load due to some electrical system disturbance.
Offsite ac power remains available to operate plant components such as the reactor coolant pumps; as a result, the onsite emergency diesel generators are not required to function for this event. Following the loss of generator load, an immediate fast closure of the turbine control valves will occur. This will cause a sudden reduction in steam flow, resulting in an increase 15.2-1
B/B-UFSAR in pressure and temperature in the steam generator shell. As a result, the heat transfer rate in the steam generator is reduced, causing the reactor coolant temperature to rise, which in turn causes coolant expansion, pressurizer insurge, and RCS pressure rise.
For a loss of external electrical load without subsequent turbine trip, no direct reactor trip signal would be generated, as the plant would be expected to trip from the reactor protection system if a safety limit were approached. A continued steam load of approximately 5% would exist after total loss of external electrical load because of the steam demand of plant auxiliaries.
In the event that a safety limit is approached, protection would be provided by the high pressurizer pressure and overtemperature T trips. Voltage and frequency relays associated with the reactor coolant pump provide no additional safety function for this event. Following a complete loss of load, the maximum turbine overspeed would be approximately 8% to 9%, resulting in an overfrequency of less than 6 Hz. This resulting overfrequency is not expected to damage the turbine protection trip sensors in any way. Testing of turbine overspeed protection equipment is required by Technical Requirements Manual (TRM) 3.3.g. Any degradation in their performance could be ascertained at that time. Any frequency increase to the reactor coolant pump motors will result in slightly increased flow rate and subsequent additional margin to safety limits. For postulated loss of load and subsequent turbine generator overspeed, any overfrequency condition is not seen by safety-related pump motors, reactor protection system equipment, or other safeguards loads. Safe-guards loads are supplied from offsite power or, alternatively, from emergency diesels. Reactor protection system equipment is supplied from the inverters; the inverters are supplied from a d-c bus energized from batteries or by a rectified ac voltage from safeguards buses.
In the event the steam dump valves fail to open following a large loss of load, the steam generator safety valves may lift and the reactor may be tripped by the high pressurizer pressure signal, the high pressurizer water level signal, or the overtemperature T signal. The steam generator shell side pressure and reactor coolant temperatures will increase rapidly. The pressurizer safety valves and steam generator safety valves are, however, sized to protect the reactor coolant system (RCS) and steam generator against overpressure for all load losses without assuming the operation of the steam dump system, pressurizer spray, pressurizer power-operated relief valves, or automatic rod cluster control assembly control.
15.2-2 REVISION 15 - DECEMBER 2014
B/B-UFSAR A more complete discussion of overpressure protection can be found in Reference 1.
A loss of external load is classified as an ANS condition II event, fault of moderate frequency. See Subsection 15.0.1 for a discussion of condition II events.
A loss of external load event results in an NSSS transient that is less severe than the turbine trip event as analyzed in Subsection 15.2.3. Therefore, a detailed transient analysis is not presented for the loss of external load.
The primary-side transient is caused by a decrease in heat transfer capability from primary to secondary due to a rapid termination of steam flow to the turbine, accompanied by an automatic reduction of feedwater flow. (Should feed flow not be reduced, a larger heat sink would be available and the transient would be less severe.) Termination of steam flow to the turbine following a loss of external load occurs due to automatic fast closure of the turbine control valves. Following a turbine trip event, termination of steam flow occurs via turbine stop valve closure. The analysis presented in Subsection 15.2.3 assumes a valve closure time which is conservatively fast for both turbine stop valves and control valves. Therefore, the results of that analysis apply to both the loss of external load event and the turbine trip event.
The protection credited for mitigating the consequences of a loss of external load is the same as that for a turbine trip, as listed in Table 15.0-7.
15.2.2.2 Analysis of Effects and Consequences Method of Analysis Refer to Subsection 15.2.3.2 for the method used to analyze the limiting transient (turbine trip) in this grouping of events.
The results of the turbine trip event analysis are bounding for those expected for the loss of external load, as discussed in Subsection 15.2.2.1.
Normal reactor control systems and engineered safety systems are not required to function. The auxiliary feedwater system may, however, be automatically actuated following a loss of main feedwater; this will further mitigate the effects of the transient.
15.2-3 REVISION 15 - DECEMBER 2014
B/B-UFSAR The reactor protection system may be required to function following a complete loss of external load to terminate core heat input and prevent departure from nucleate boiling (DNB).
Depending on the magnitude of the load loss, pressurizer safety valves and/or steam generator safety valves may be required to open to maintain system pressure below allowable limits. No single active failure will prevent operation of any system required to function.
15.2.2.3 Radiological Consequences Loss of external load from full power would result in the operation of the steam dump system. This system keeps the main turbine generator operating to supply auxiliary electrical loads. Operation of the steam dump system results in bypassing steam to the condenser. If steam dumps are not available, steam generator safety and relief valves relieve to the atmosphere.
The radiological consequences will be less severe than those for the steamline break accident analyzed in Subsection 15.1.5.3.
15.2.2.4 Conclusions Based on results obtained for the turbine trip event (Subsection 15.2.3) and considerations described in Subsection 15.2.2.1, the applicable acceptance criteria for a loss of external load event are met.
15.2.3 Turbine Trip 15.2.3.1 Identification of Causes and Accident Description For a turbine trip event, the turbine stop valves close rapidly (typically 0.1 sec.) on loss of trip fluid pressure actuated by one of a number of possible turbine trip signals. Turbine trip initiation signals include:
- a. low condenser vacuum,
- b. low bearing oil pressure,
- c. turbine thrust bearing failure,
- d. turbine overspeed,
- e. manual trip,
- f. Low emergency trip header pressure, and
- g. Loss of both redundant controllers.
Upon initiation of stop valve closure, steam flow to the turbine stops abruptly. Sensors on the stop valves detect the turbine trip and initiate steam dump. The loss of steam flow results in an almost immediate rise in secondary system temperature and 15.2-4 REVISION 12 - DECEMBER 2008
B/B-UFSAR pressure with a resultant primary system transient as described in Subsection 15.2.2.1 for the loss of external load event. For a turbine trip, the reactor would be tripped directly (unless below approximately 30% (P-8) power) on a signal from the turbine stop valves.
The automatic steam dump system would normally accommodate the excess steam generation. Reactor coolant temperatures and pressure do not significantly increase if the steam dump system and pressurizer pressure control system are functioning properly.
If the turbine condenser was not available, the excess steam generation would be dumped to the atmosphere and main feedwater flow would be lost. For this situation, feedwater flow would be maintained by the auxiliary feedwater system to ensure adequate residual and decay heat removal capability. Should the steam dump system fail to operate, the steam generator safety valves may lift to provide pressure control. See Subsection 15.2.2.1 for a further discussion of the transient.
A turbine trip is classified as an ANS condition II event, fault of moderate frequency. See Subsection 15.0.1 for a discussion of condition II events.
A turbine trip event is bounding for loss of external load, loss of condenser vacuum, inadvertent closure of the main steam isolation valves, and other turbine trip events. As such, this event has been analyzed in detail. Results and discussion of the analysis are presented in Subsection 15.2.3.2.
The plant systems and equipment credited for mitigating the consequences of a turbine trip are discussed in Subsection 15.0.8 and listed in Table 15.0-7.
15.2.3.2 Analysis of Effects and Consequences Method of Analysis In this analysis, the behavior of the unit is evaluated for a complete loss of steam load from full power primarily to show the adequacy of the pressure relieving devices and also to demonstrate core protection margins. The reactor is not tripped until conditions in the RCS result in a trip. No credit is taken for steam dump. Main feedwater flow is terminated at the time of turbine trip, with no credit taken for auxiliary feedwater to mitigate the consequences of the transient.
The turbine trip transients are analyzed by employing the detailed digital computer program LOFTRAN (Reference 3). The program simulates the neutron kinetics, RCS pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The program 15.2-5 REVISION 15 - DECEMBER 2014
B/B-UFSAR computes pertinent plant variables including temperatures, pressures, and power level.
This accident is analyzed for three specific cases, one for maximum RCS pressure, a second for maximum MSS pressure, and a third for minimum DNBR. For the pressure cases, the analysis is performed using the standard thermal design procedure. For the DNB case, the revised thermal design procedure is used as described in WCAP-11397-P-A (Reference 5). Plant characteristics and initial conditions are discussed in Subsection 15.0.3. This accident has also been analyzed separately for Unit 1 and for Unit 2 due to differences in steam generator designs.
Major assumptions used in all three cases are summarized below:
- a. Moderator and Doppler Coefficients of Reactivity -
minimum reactivity feedback is assumed conservatively in all cases. The analysis is performed conservatively at full power conditions assuming a moderator temperature coefficient of 0 pcm/oF and the least negative Doppler-only power and Doppler temperature coefficients. These conditions are bounding conservatively for all operating conditions anticipated throughout each cycle.
- b. Reactor Control - from the standpoint of the adverse conditions of concern for this event, it is conservative to assume that the reactor is in manual control. If the reactor were in automatic control, the control rod banks would move prior to trip and reduce the severity of the transient.
- c. Steam Release - no credit is taken for the operation of the steam dump system or steam generator power-operated relief valves. When the steam generator pressure rises to the safety valve setpoint, the steam release through the safety valve limits secondary steam pressure.
- d. Feedwater Flow - main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip. No credit is taken for auxiliary feedwater flow since a stabilized plant condition will be reached before auxiliary feedwater initiation is normally assumed to occur. The auxiliary feedwater flow would remove core decay heat following plant stabilization.
- e. Reactor Trip - is actuated by the first reactor protection system trip setpoint reached. Trip signals are expected due to high pressurizer pressure and overtemperature T.
- f. Main Steam Safety Valves - The analysis assumes all main steam safety valves are operable. If one or more 15.2-6 REVISION 15 - DECEMBER 2014
B/B-UFSAR main steam safety valves become inoperable, Technical Specification 3.7.1 requires a reduction in power as well as a reduction in the power range neutron flux -
high reactor trip setpoint. The required reductions are based on a heat balance equation as described in Reference 6. A sensitivity study was performed to demonstrate that the reductions are adequate.
Additional major assumptions for the RCS pressure case include the following:
- a. Initial Operating Conditions - initial reactor power is assumed to be at the nominal value plus uncertainties.
The nominal full power RCS temperature plus uncertainties, including the RCS average temperature bias, is modeled. Initial RCS pressure is assumed to be at its nominal value minus uncertainties. The RCS flow rate corresponds to thermal design flow. Maximum steam generator tube plugging is assumed.
- b. Pressurizer Spray and Power-Operated Relief Valves - no credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are operable.
Additional major assumptions for the MSS pressure case include the following:
- a. Initial Operating Conditions - initial reactor power is assumed to be at the nominal value plus uncertainties.
The nominal full power RCS temperature plus uncertainties, including the RCS average temperature bias, is modeled. Initial RCS pressure is assumed to be at its nominal value minus uncertainties. The RCS flow rate corresponds to thermal design flow. Minimum steam generator tube plugging is assumed.
- b. Pressurizer Spray and Power-Operated Relief Valves -
full credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are also available.
Additional major assumptions for the DNB case include the following:
- a. Initial Operating Conditions - initial reactor power and pressure are assumed to be at their nominal values.
With the exception of the RCS average temperature bias, which is explicitly modeled in the analysis, uncertainties in initial conditions are included in the limit DNBR as described in WCAP-11397-P-A (Reference 5).
The RCS flow rate corresponds to minimum measured flow.
Maximum steam generator tube plugging is assumed.
- b. Pressurizer Spray and Power-Operated Relief Valves -
full credit is taken for the effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are also available.
15.2-6a REVISION 15 - DECEMBER 2014
B/B-UFSAR Margin Between High Pressurizer Pressure Reactor Trip and Opening of Pressurizer Safety Valves The Technical Specifications allow a 2% tolerance around the nominal setting of 2460 psig for the pressurizer safety valves, with the lowest allowed lifting pressure being 2411 psig. The assumed high pressurizer pressure reactor trip setpoint in the peak pressure cases, including instrument uncertainty, is higher than the lowest setting for the pressurizer safety valves. As a result, the pressurizer safety valves could lift prior to reaching the high pressurizer pressure reactor trip setpoint and the reactor trip might be delayed.
The pressurizer safety valves are assumed to lift at the highest setpoint in the peak pressure cases to maximize RCS and secondary side pressure. A sensitivity study is performed assuming the lowest pressurizer safety valve lift setpoint to demonstrate that, with a delayed reactor trip, all applicable acceptance criteria are met.
15.2-6b REVISION 15 - DECEMBER 2014
B/B-UFSAR Except as discussed above, normal reactor control system and engineered safety systems are not required to function.
The reactor protection system may be required to function following a turbine trip. Pressurizer safety valves and/or steam generator safety valves may be required to open to maintain system pressures below allowable limits. No single active failure will prevent operation of any system required to function.
Results The transient responses for a turbine trip from full power operation are shown for three cases for each unit. The calculated sequence of events for the accident is shown in Table 15.2-1.
RCS Pressure Case The transient responses for the total loss of steam load from full power for the RCS overpressure case are shown in Figures 15.2-1 and 15.2-2 for Unit 1 and Figures 15.2-5 and 15.2-6 for Unit 2, respectively. No credit is taken for the pressurizer spray, pressurizer power-operated relief valves, or for the steam dump. The reactor is tripped by the high pressurizer pressure trip channel. The pressurizer safety valves are actuated and the primary system pressure remains below the 110% design value.
MSS Pressure Case The transient responses for the total loss of steam load from full power for the MSS overpressure case are shown in Figures 15.2-1a and 15.2-2a for Unit 1 and Figures 15.2-5a and 15.2-6a for Unit 2, respectively. Full credit is taken for the pressurizer spray and pressurizer power-operated relief valves. No credit is taken for the steam dump. The reactor is tripped by the overtemperature T trip channel. The steam generator safety valves maintain the secondary side steam pressure below 110% of the steam generator shell design pressure.
DNB Case The transient responses for the total loss of steam load from full power for the DNB case are shown in Figures 15.2-3 and 15.2-4 for Unit 1 and Figures 15.2-7 and 15.2-8 for Unit 2, respectively. Full credit is taken for the pressurizer spray and pressurizer power-operated relief valves. No credit is taken for the steam dump. The reactor is tripped by the overtemperature T trip channel. The minimum DNBR remains well above the limit value.
15.2-7 REVISION 15 - DECEMBER 2014
B/B-UFSAR In the event that feedwater flow is not terminated at the time of turbine trip for these cases, flow would continue under automatic control with the reactor at a reduced power. The operator would take action to terminate the transient and bring the plant to a stabilized condition. If no action were taken by the operator, the reduced power operation would continue until the condenser hotwell is emptied. A low-low steam generator water level reactor trip would be generated along with auxiliary feedwater initiation signals. Auxiliary feedwater would then be used to remove decay heat with the results less severe than those presented in Subsection 15.2.7.
15.2-7a REVISION 15 - DECEMBER 2014
B/B-UFSAR For Unit 2, Reference 1 presents additional results of the analysis for a complete loss of heat sink including loss of main feedwater. This analysis shows the overpressure protection that is afforded by the pressurizer and steam generator safety valves.
15.2.3.3 Radiological Consequences The turbine trip transient and steam released for this event are similar to the loss of load transient described in Subsection 15.2.2.3.
There are only minimal radiological consequences associated with this event, therefore, this event is not limiting. The radio-logical consequences resulting from atmosphere steam dump are less severe than the steamline break event analyzed in Subsection 15.1.5.3.
15.2-8 REVISION 9 - DECEMBER 2002
B/B-UFSAR 15.2.3.4 Conclusions Results of the analyses show that the plant design is such that a turbine trip presents no hazard to the integrity of the RCS or the main steam system. Pressure relieving devices incorporated in the two systems are adequate to limit the maximum pressures to within the design limits.
The integrity of the core is maintained by operation of the reactor protection system, i.e., the DNBR will be maintained above the limit value. The applicable acceptance criteria as listed in Subsection 15.0.1 have been met. The above analysis demonstrates the ability of the NSSS to safely withstand a full load rejection. The radiological consequences in this event will be less than the steam break event analyzed in Subsection 15.1.5.3.
15.2.4 Inadvertent Closure of Main Steam Isolation Valves Inadvertent closure of the main steam isolation valves would result in a turbine trip. Turbine trips are discussed in Subsection 15.2.3.
15.2.5 Loss of Condenser Vacuum and Other Events Causing a Turbine Trip Loss of condenser vacuum is one of the events that can cause a turbine trip. Turbine trip initiating events are described in Subsection 15.2.3. A loss of condenser vacuum would preclude the use of steam dump to the condenser; however, since steam dump is assumed not to be available in the turbine trip analysis, no additional adverse effects would result if the turbine trip were caused by loss of condenser vacuum. Therefore, the analysis results and conclusion contained in Subsection 15.2.3 apply to loss of condenser vacuum. In addition, analyses for the other possible causes of a turbine trip, as listed in Subsection 15.2.3.1 are covered by Subsection 15.2.3. Possible over-frequency effects due to a turbine overspeed condition are discussed in Subsection 15.2.2.1 and are not a concern for this type of event.
15.2-9 REVISION 15 - DECEMBER 2014
B/B-UFSAR Pages 15.2-10 through 15.2-13 have been deleted intentionally.
15.2-10 through 15.2-13 REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.2.6 Loss of Nonemergency AC Power to the Plant Auxiliaries 15.2.6.1 Identification of Causes and Accident Description A complete loss of nonemergency ac power may result in the loss of all power to the plant auxiliaries, i.e., the reactor coolant pumps, condensate pumps, etc. The loss of power may be caused by a complete loss of the offsite grid accompanied by a turbine generator trip at the station, or by a loss of the onsite ac distribution system.
This transient is more severe than the turbine trip event analyzed in Subsection 15.2.3 because for this case the decrease in heat removal by the secondary system is accompanied by a flow coastdown which further reduces the capacity of the primary coolant to remove heat from the core. The reactor will trip:
(1) upon reaching one of the trip setpoints in the primary and secondary systems as a result of the flow coastdown and decrease in secondary heat removal; or (2) due to loss of power to the control rod drive mechanisms as a result of the loss of power to the plant.
Following a loss of ac power with turbine and reactor trips, the sequence described below will occur:
- a. The standby diesel generators, started on loss of voltage on the plant emergency buses, begin to supply plant vital loads.
- b. Plant vital instruments are supplied from emergency d-c power sources.
- c. As the steam system pressure rises following the trip, the steam generator power-operated relief valves may be automatically opened to the atmosphere. The condenser is assumed not to be available for steam dump. If the steam flow rate through the power relief valves is not available, the steam generator safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in the reactor.
- d. As the no load temperature is approached, the steam generator power-operated relief valves (or safety valves, if the power operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot standby condition.
The auxiliary feedwater system is started automatically as follows:
15.2-13a REVISION 16 - DECEMBER 2016
B/B-UFSAR One motor driven and one diesel-driven auxiliary feedwater pump are started on any of the following:
- a. Low-low level in any steam generator,
- b. Any safety injection signal,
- c. Loss of offsite power, and
- d. Manual actuation.
The motor-driven auxiliary feedwater pump is supplied power by the diesel generators. The diesel-driven auxiliary feedwater pump is driven by its own diesel engine. Both type pumps are designed to supply rated flow within approximately one minute of the initiating signal even if a loss of all nonemergency a-c power occurs simultaneously with loss of normal feedwater.
The motor-driven auxiliary feedwater pump is supplied power by the diesel generators. The diesel-driven auxiliary feedwater pump is driven by its own diesel engine. Both type pumps are designed to supply rated flow within 63 seconds of a low-low steam generator signal if a loss of all nonemergency a-c power occurs with a loss of normal feedwater. The motor driven auxiliary feedwater pump delay and startup sequence times after the low-low steam generator level is reached and assuming a loss of offsite power occurs when the reactor trip occurs are given below. The starting sequence is within the 63 seconds of the low-low steam generator signal assumed for establishing rated auxiliary feedwater flow with a loss of offsite power.
Start of rod insertion and loss of offsite power occurs 2.0 seconds CV-7 Undervoltage Relay Allowable Response Time 5.9 Diesel Starting 10.0 CV-7 Overvoltage Relay Operation 2.0 Auxiliary Feedwater Pump Sequence Delay 35.0 Auxiliary Feedwater Pump Maximum Acceleration Time 6.1 Miscellaneous Auxiliary Relay/Breaker Operation 1.0 Total 62.0 seconds The pumps take suction from the condensate storage tank for delivery to the steam generators. If the condensate storage tank is not available, then the source for the AFW is switched to the Essential Service Water system causing an additional delay of 12 seconds in the initiation of AFW so that sufficient water is provided to the AFW system to operate the pump.
Upon the loss of power to the reactor coolant pumps, coolant flow necessary for core cooling and the removal of residual heat is maintained by natural circulation in the reactor coolant loops.
A loss of nonemergency ac power to the station auxiliaries is classified as an ANS Condition II event, fault of moderate frequency. See Subsection 15.0.1 for a discussion of Condition II events.
15.2-13b REVISION 16 - DECEMBER 2016
B/B-UFSAR The first few seconds after the loss of power to the reactor coolant pumps will closely resemble the simulation of the complete loss of flow incident (see Subsection 15.3.2, where it is shown that the DNBR is maintained above the limit value). Therefore, the DNB aspects of this event were not reevaluated for this analysis.
Following the reactor coolant pump coastdown caused by the loss of ac power, the natural circulation capability of the RCS will remove residual and decay heat from the core, aided by auxiliary feedwater in the secondary system. An analysis is presented here to show that the natural circulation flow in the RCS following a loss of ac power event is sufficient to remove residual heat from the core.
Like the analysis of the Loss of Normal Feedwater Flow event described in UFSAR Section 15.2.7, the primary purpose of the analysis for the Loss of Non-emergency AC Power to the Plant Auxiliaries described herein is to demonstrate the long-term heat removal capacity of the auxiliary feedwater system. Following a loss of offsite power signal and resultant loss of instrument air, the letdown isolation valves close, the normal charging path is isolated, and the charging pumps are initiated. Unlike the letdown and normal charging flow paths, the reactor coolant pump seal injection flow path is not isolated following a loss of instrument air; thus, coolant would be added to the RCS via seal injection flow. These actions would provide a net addition of relatively cold water to the reactor coolant system inventory. Since this addition of colder water would provide a benefit relative to the post-trip primary-side temperature increase, it is conservative not to credit these actions in demonstrating the heat removal capacity of the auxiliary feedwater system. However, the potential effect of the reactor coolant pump seal injection flow on the reactor coolant system inventory following a loss of offsite power are considered as part of the overall evaluation of this event.
The plant systems and equipment credited for mitigating the consequences of a loss of ac power event are discussed in Subsection 15.0.8 and listed in Table 15.0-7.
15.2.6.2 Analysis of Effects and Consequences An analysis was performed for both Unit 1 and Unit 2. However, the Unit 1 analysis is more limiting and, therefore, is presented here.
Method of Analysis A detailed analysis using the LOFTRAN Code (Reference 3) is performed to obtain the plant transient following a station loss of nonemergency a-c power. The simulation describes the plant thermal kinetics, reactor coolant system (RCS) including the natural circulation, pressurizer, steam generators and feedwater system.
The digital program computes pertinent variables including the steam generator mass, pressurizer water level, and reactor coolant average temperature.
15.2-13c REVISION 16 - DECEMBER 2016
B/B-UFSAR Assumptions made in the analysis are:
- a. The initial power level bounds 100% of the nominal NSSS power level of 3672 MWt (including power uncertainty), which includes the nominal reactor coolant pump heat.
- b. The initiating event is a loss of all non-emergency ac power that results in the loss of power supply for the condensate pumps. The loss of the condensate pumps results in a loss of normal feedwater.
- c. The RCPs are conservatively assumed to operate until the time of reactor trip providing a constant reactor coolant volumetric flow equal to the Thermal Design Flow value.
This is to maximize the amount of stored energy in the RCS.
The loss of power to the RCPs is not assumed to occur until after the start of rod motion following the reactor trip on a low-low steam generator water level condition.
- d. No credit is taken for the immediate insertion of the control rods because of the loss of ac power to the station auxiliaries.
- e. Cases are analyzed assuming initial HFP reactor vessel average coolant temperatures at the upper and lower ends of the reactor vessel average temperature window. The vessel average temperature assumed at the upper end of the temperature window is 588oF plus an uncertainty of 9.1oF, which includes a bias of 1.5oF. The average temperature assumed at the lower end of the average temperature window is 575oF minus an uncertainty of 7.6oF.
The limiting Unit 1 case assumes a vessel average temperature of 575oF minus an uncertainty of 7.6oF.
- f. Initial pressurizer pressure is assumed to be 2250 psia with an initial pressurizer pressure uncertainty of +/-43 psi.
Cases are considered with the pressure uncertainty applied in both the positive and negative direction to conservatively encompass the potential operating conditions.
The limiting Unit 1 case assumes an initial pressurizer pressure of 2250 psia plus the uncertainty of 43 psi.
- g. Reactor trip occurs on steam generator low-low water level at 10% of narrow range span for the limiting case (Unit 1).
- h. The worst single failure, which is modeled in the analysis, is the loss of the diesel-driven AFW pump. This results in the availability of one motor-driven AFW pump supplying a minimum total AFW flow of 560 gpm distributed equally to each of the four steam generators.
15.2-13d REVISION 16 - DECEMBER 2016
B/B-UFSAR
- i. Cases are analyzed assuming different sources for AFW resulting in different delay times to the initiation of AFW.
When the AFW source is the condensate storage tank, AFW flow is assumed to be initiated 63 seconds following a low-low steam generator water level signal to allow time for sensor response, signal processing, startup of the emergency diesel generators and the AFW pump. AFW flow is assumed to be initiated 75 seconds following a low-low steam generator water level signal when the condensate storage tank is not available and the AFW source is switched to the Essential Service Water system. The additional 12 seconds allows time for the Essential Service Water system to provide sufficient water to operate the pump. The limiting case (Unit 1) assumes a delay of 75 seconds in AFW initiation because the Essential Service Water system is the AFW source.
- j. A maximum steam generator blowdown valve leak rate of 10 gpm per steam generator is accounted for in the analysis.
- k. The pressurizer sprays and PORVs are assumed to be operable to maximize the pressurizer water volume. These control systems are not credited for event mitigation since the pressurizer safety valves alone would prevent the RCS pressure from exceeding the RCS design pressure limit during this transient. An evaluation was performed modeling the PORVs as inoperable. The evaluation proved that it is conservative to model the PORVs as inoperable and acceptable results were obtained.
- l. Secondary system steam relief is achieved through the steam generator safety valves that are modeled assuming a +4% lift point tolerance.
- m. A conservative core residual heat generation based upon long term operation at the initial power level preceding the trip is assumed in the analysis. This core residual heat generation model is based on the 1979 version of ANS 5.1 (Reference 4). ANSI/ANS-5.1-1979 is a conservative representation of the decay energy release rates.
- n. A steam generator tube plugging level of 0% is modeled for the limiting case (Unit 1).
- o. Cases are analyzed assuming a maximum AFW enthalpy of 91.12 Btu/lbm when the AFW source is condensate storage tank and assuming a maximum AFW enthalpy of 78.2 Btu/lbm for Byron and 80.2 Btu/lbm for Braidwood (AF temperature of 104F) when the AFW source is the Essential Service Water system. An AFW line purge volume of 160 ft3 is also modeled. The limiting case (Unit 1) modeled an AFW enthalpy of 78.2 Btu/lbm for Byron and 80.2 Btu/lbm for Braidwood (AF temperature of 104F).
- p. A heat transfer coefficient in the steam generators associated with RCS natural circulation is assumed following the RCP coastdown.
15.2-13e REVISION 16 - DECEMBER 2016
B/B-UFSAR The assumptions used in the analysis are essentially identical to the loss of normal feedwater flow incident (Subsection 15.2.7),
except that power is assumed to be lost to the reactor coolant pumps at the time of reactor trip and the pressurizer heaters are not operable during a loss of offsite power event.
Plant characteristics and initial conditions are further discussed in Subsection 15.0.3.
Results The limiting results were calculated for Unit 1 modeling the Essential Service Water system as the AFW source. The transient response of the RCS following a loss of ac power is shown in Figures 15.2-9a and 15.2-10a. The calculated sequence of events for this event is listed in Table 15.2-1.
The first few seconds after the loss of power to the reactor coolant pumps will closely resemble a simulation of the complete loss of flow incident (see Subsection 15.3.2) i.e., core damage due to rapidly increasing core temperatures is prevented by promptly tripping the reactor. After the reactor trip, stored and residual decay heat must be removed to prevent damage to either the RCS or the core.
For the purpose of demonstrating adequate heat removal capacity of the emergency auxiliary feedwater system, operation of the charging pumps, initiated on a loss of offsite power signal, are not assumed to function for this event since their operation is a benefit with respect to long term core decay heat removal. A loss of offsite power will lead to increased reactor coolant system inventory because the following events occur: 1) the charging pumps actuate and mass is added to the reactor coolant system via reactor coolant seal injection flow, and 2) the letdown isolation valves close due to a loss of instrument air.
This scenario is examined separately. From the evaluation of this scenario, it was determined that appropriate operator actions can be taken within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of event initiation to terminate water relief through the pressurizer safety valves and preclude valve damage. These operator actions are included in the plant emergency response guidelines following a loss of offsite power event.
15.2-13f REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.2.6.3 Radiological Consequence A loss of nonessential ac power to plant auxiliaries would result in a turbine and reactor trip and loss of condenser vacuum. Heat removal from the secondary system would occur through the steam generator power-operated relief valves or safety valves. The radiological consequences will be less severe than the steamline break accident analyzed in Subsection 15.1.5.3.
15.2.6.4 Conclusions Analysis of the natural circulation capability of the reactor coolant system has demonstrated that sufficient heat removal capability exists following RCP coastdown to prevent fuel or clad damage. The radiological consequences of this event would be less severe than the steamline break event analyzed in Subsection 15.1.5.3.
15.2-13g REVISION 16 - DECEMBER 2016
B/B-UFSAR Pages 15.2-14 through 15.2-16 have been deleted intentionally.
15.2-14 through 15.2-16 REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.2.7 Loss of Normal Feedwater Flow 15.2.7.1 Identification of Causes and Accident Description A loss of normal feedwater (from pump failures, valve malfunctions, or loss of offsite ac power) results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If an alternative supply of feedwater were not supplied to the plant, core residual heat following reactor trip would heat the primary system water to the point where water relief from the pressurizer would occur, resulting in a substantial loss of water from the reactor coolant system (RCS). Since the plant is tripped well before the steam generator heat transfer capability is reduced, the primary system variables never approach a DNB condition.
The following occur upon loss of normal feedwater (assuming main feedwater pump failures or valve malfunctions):
- a. As the steam system pressure rises following the trip, the steam generator power-operated relief valves are automatically opened to the atmosphere.
Steam dump to the condenser is assumed not to be available. If the steam flow rate through the power-operated relief valves is not available, the steam generator safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in the reactor.
- b. As the no load temperature is approached, the steam generator power-operated relief valves (or safety valves, if the power-operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot standby condition.
A loss of normal feedwater is classified as an ANS Condition II event, fault of moderate frequency. See Subsection 15.0.1 for a discussion of Condition II events.
The auxiliary feedwater system is started automatically as discussed in Subsection 15.2.6.1. Both the motor-driven auxiliary feedwater pump and the independently powered diesel-driven auxiliary feedwater pump take suction directly from the condensate storage tank for delivery to the steam generators.
If the condensate storage tank is not available, then the source for the AFW is switched to the Essential Service Water system causing an additional delay of 12 seconds in the initiation of AFW so that sufficient water is provided to the AFW system to operate the pump.
15.2-16a REVISION 16 - DECEMBER 2016
B/B-UFSAR An analysis of the system transient is presented below to show that following a loss of normal feedwater, the auxiliary feedwater system is capable of removing the stored and residual heat, thus preventing either overpressurization of the RCS or loss of water from the reactor core, and returning the plant to a safe condition.
15.2.7.2 Analysis of Effects and Consequences An analysis was performed for both Unit 1 and Unit 2. Because the Unit 2 analysis is more limiting, it is presented here.
Method of Analysis A detailed analysis using the LOFTRAN Code (Reference 3) is performed in order to obtain the plant transient following a loss of normal feedwater. The simulation describes the plant thermal kinetics, RCS, pressurizer, steam generators and feedwater system. The digital program computes pertinent variables including the steam generator mass, pressurizer water level, and reactor coolant average temperature.
Assumptions made in the analysis are:
- a. The initial power level bounds 100% of the nominal NSSS power level of 3672 MWt (including power uncertainty), which includes applicable uncertainties and a maximum reactor coolant pump heat of 16.65 MWt.
- b. The RCPs are assumed to operate continuously throughout the transient providing a constant reactor coolant volumetric flow equal to the Thermal Design Flow.
- c. Cases are analyzed assuming initial HFP reactor vessel average coolant temperatures at the upper and lower ends of the reactor vessel average temperature window.
The vessel average temperature assumed at the upper end of the temperature window is 588oF plus an uncertainty of 9.1oF, which includes a bias of 1.5oF.
The average temperature assumed at the lower end of the average temperature window is 575oF minus an uncertainty of 7.6oF.
The limiting Unit 2 case assumes a vessel average temperature of 588oF plus an uncertainty of 9.1oF.
- d. Initial pressurizer pressure is assumed to be 2250 psia with an initial pressurizer pressure uncertainty of +/-43 psi. Cases are considered with the pressure uncertainty applies in both the positive and negative direction to conservatively encompass the potential operating conditions.
15.2-16b REVISION 17 - DECEMBER 2018
B/B-UFSAR The limiting Unit 2 case assumes an initial pressurizer pressure of 2250 psia plus the uncertainty of 43 psi.
- e. Reactor trip occurs on steam generator low-low water level at 28.6% of narrow range span for the limiting case (Unit 2).
- f. The worst single failure, which is modeled in the analysis, is the loss of the diesel-driven AFW pump.
This results in the availability of one motor-driven AFW pump supplying a minimum total AFW flow of 560 gpm distributed equally among each of the four steam generators.
- g. Cases are analyzed assuming different sources for AFW resulting in different delay times to the initiation of AFW. AFW flow is assumed to be initiated 55 seconds following a low-low steam generator water level signal when the AFW sources is the condensate storage tank to allow time for sensor response, signal processing, and startup of the AFW pump. AFW flow is assumed to be initiated 67 seconds following a low-low steam generator water level signal when the condensate storage tank is not available and the AFW source is switched to the Essential Service Water system to provide sufficient water to operate the pump. The limiting case (Unit 2) assumes a delay of 55 seconds in AFW initiation because the condensate storage tank is the AFW source.
- h. A maximum steam generator blowdown valve leak rate of 3 gpm per steam generator is accounted for in the analysis.
- i. In order to conservatively maximize the pressurizer water volume, the pressurizer sprays are assumed to be operable and the pressurizer PORVs are assumed to be inoperable. For event mitigation, the pressurizer safety valves alone would prevent the RCS pressure from exceeding the RCS design pressure limit during this transient.
- j. Secondary system steam relief is achieved through the steam generator safety valves that are modeled assuming at +3% lift point tolerance.
- k. A conservative core residual heat generation based upon long term operation at the initial power level preceding the trip is assumed in the analysis. This core residual heat generation model is based on the 1979 version of ANS 5.1 (Reference 4). ANSI/ANS-5.1-1979 is a conservative representation of the decay energy release rates.
15.2-16c REVISION 16 - DECEMBER 2016
B/B-UFSAR
- l. A steam generator tube plugging level of 0% is modeled for the limiting case (Unit 2).
- m. Cases are analyzed assuming a maximum AFW enthalpy of 91.12 Btu/lbm when the AFW source is the condensate storage tank and assuming a maximum AFW enthalpy of 78.2 Btu/lbm for Byron and 80.2 Btu/lbm for Braidwood when the AFW source is the Essential Service Water system. An AFW line purge volume of 160 ft3 is also modeled. The limiting case (Unit 2) modeled an AFW enthalpy of 91.12 Btu/lbm.
The loss of normal feedwater analysis is performed to demonstrate the adequacy of the reactor protection and engineered safeguards systems (e.g., the auxiliary feedwater system) in removing long term decay heat and preventing excessive heatup of the RCS with possible resultant RCS overpressurization or loss of RCS water.
As such, the assumptions used in this analysis are designed to minimize the energy removal capability of the system and to maximize the possibility of water relief from the coolant system by maximizing the coolant system expansion, as noted in the assumptions listed above.
For the loss of normal feedwater transient, the reactor coolant volumetric flow remains at its normal value and the reactor trips via the low-low steam generator level trip. The reactor coolant pumps may be manually tripped at some later time to reduce heat addition to the RCS.
Plant characteristics and initial conditions are further discussed in Subsection 15.0.3.
Plant systems and equipment credited for mitigating the effects of a loss of normal feedwater accident are discussed in Subsection 15.0.8 and listed in Table 15.0-7. Normal reactor control systems are not required to function. The reactor protection system is required to function following a loss of normal feedwater as analyzed here. The auxiliary feedwater system is required to deliver a minimum auxiliary feedwater flowrate. No single active failure will prevent operation of any system required to function.
Results The limiting results were calculated for Unit 2, modeling the condensate storage tank as the AFW source. Figures 15.2-11 and 15.2-12 show the significant plant parameters following a loss of normal feedwater.
15.2-16d REVISION 16 - DECEMBER 2016
B/B-UFSAR Following the reactor and turbine trip from full load, the water level in the steam generators will fall due to the reduction of steam generator void fraction and because steam flow through the safety valves continues to dissipate the stored and generated heat. Approximately one minute following the initiation of the low-low level trip, at least one auxiliary feedwater pump is automatically started, reducing the rate of water level decrease.
The capacity of the auxiliary feedwater pumps combined with the large steam generator inventory is such that the water level in the steam generators being fed does not recede below the lowest level at which sufficient heat transfer area is available to dissipate core residual heat without water relief from the RCS safety valves. Figure 15.2-11 shows that at no time is there water relief from the pressurizer.
The calculated sequence of events for this accident is listed in Table 15.2-1. As shown in Figures 15.2-11 and 15.2-12, the plant approaches a stabilized condition following reactor trip and auxiliary feedwater initiation. Plant procedures may be followed to further cool down the plant.
15.2.7.3 Radiological Consequences The steam release and resulting radiological consequence from this transient would be the same as that for the loss of offsite ac power, and similarly, radiological consequences resulting from this transient are less severe than the steamline break accident analyzed in Subsection 15.1.5.3.
15.2.7.4 Conclusions Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system and the auxiliary feedwater capacity is such that reactor coolant water is not relieved from the pressurizer relief or safety valves. The radiological consequences of this event would be less severe than the steamline break accident analyzed in Subsection 15.1.5.3.
15.2-16e REVISION 16 - DECEMBER 2016
B/B-UFSAR Pages 15.2-16f through 15.2-24 have been deleted intentionally.
15.2-16f through 15.2-24 REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.2.8 Feedwater System Pipe Break It should be noted that the results and plots presented in this section are for Unit 2 and are not intended to be representative of Unit 1. The cooldown that occurs during a Unit 1 feedline break event is much more severe than that for Unit 2. This is because of the difference between the main feedwater system design in the Unit 1 (feedring) and Unit 2 (preheat) steam generators. The cooldown portion of the event is bounded by the steamline break event, as analyzed in Subsections 15.1.5 and 15.1.6. The results of the heatup portion of the feedline break show that the Unit 2 analysis is more limiting than the Unit 1 analysis with respect to hot leg saturation margin. Therefore, the Unit 1 results are not presented here.
15.2.8.1 Identification of Causes and Accident Description A major feedwater line break is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the steam generators to maintain shell side fluid inventory in the steam generators. If the break is postulated in a feedline between the check valve and the steam generator, fluid from the steam generator may also be discharged through the break. (A break upstream of the feedline check valve would affect the nuclear steam supply system NSSS only as a loss of feedwater.
This case is covered by the evaluation in Subsection 15.2.7.)
15.2-24a REVISION 16 - DECEMBER 2016
B/B-UFSAR Depending upon the size of the break and the plant operating conditions at the time of the break, the break could cause either a RCS cooldown (by excessive energy discharge through the break) or a RCS heatup. Potential RCS cooldown resulting from a secondary pipe break is evaluated in Subsection 15.1.5.
Therefore, only the RCS heatup effects are evaluated for a feedwater line break.
A feedwater line break reduces the ability of the steam generators to remove heat generated by the core from the RCS for the following reasons:
- a. Feedwater flow to the steam generators is reduced.
Since feedwater is subcooled, its loss may cause reactor coolant temperatures to increase prior to reactor trip.
- b. Fluid in the steam generator may be discharged through the break, and would then not be available for decay heat removal after trip.
- c. The break may be large enough to prevent the addition of any main feedwater to the steam generators.
An auxiliary feedwater system is provided to assure that adequate feedwater will be available such that:
- a. No substantial overpressurization of the RCS shall occur.
- b. Sufficient liquid in the RCS shall be maintained in order to provide adequate decay heat removal.
Refer to Subsection 10.4.9 for a description of the auxiliary feedwater system interfaces.
Although overpressurization of the RCS is stated as a concern, the main feedwater line break event is bounded by the loss of load and turbine trip event (see subsections 15.2.2 and 15.2.3), in which assumptions are made to conservatively calculate the RCS pressure transient.
A major feedwater line break is classified as an ANS Condition IV event. See Subsection 15.0.1 for a discussion of Condition IV events.
The severity of the feedwater line break transient depends on a number of system parameters including break size, initial reactor power, and credit taken for the functioning of various control and safety systems. A number of cases of feedwater line breaks have been analyzed. Based on these analyses, it has been shown that the most limiting feedwater line breaks are the double ended break of the largest feedwater line, occurring at full power with and without loss of offsite power. These cases are discussed below.
The following provides the necessary protection for a main feedwater break:
15.2-24b REVISION 16 - DECEMBER 2016
B/B-UFSAR
- a. A reactor trip on any of the following conditions:
- 1. High pressurizer pressure.
- 2. Overtemperature T.
- 3. Low-low steam generator water level in any steam generator.
- 4. Safety injection signals from any of the following:
a) 2/3 low steam line pressure in any loop.
b) 2/3 high containment pressure.
(Refer to Chapter 7.0 for a description of the actuation system.)
- b. An auxiliary feedwater system to provide an assured source of feedwater to the steam generators for decay heat removal. The auxiliary feedwater system can draw water from either the condensate storage tank or the Essential Service Water system. If the condensate storage tank is not available, then the auxiliary feedwater source is switched to the Essential Service Water System causing an additional delay of 12 seconds in the initiation of auxiliary feedwater because of the time required to provide sufficient water to operate the AFW pump. (Refer to Subsection 10.4.9 for a description of the auxiliary feedwater system.)
Receipt of a low-low steam generator water level signal in at least one steam generator starts the motor-driven and diesel-driven auxiliary feedwater pumps, which then deliver auxiliary feedwater flow to the steam generators. Similarly, receipt of a low steamline pressure signal in at least one steam line initiates a steamline isolation signal which closes the main steamline isolation valves in all steam lines. This signal also gives a safety injection signal which initiates flow of borated water into the RCS. The amount of safety injection flow is a function of RCS pressure.
Emergency operating procedures call for the following actions to be taken by the reactor operator following a secondary system line break:
- a. Isolate feedwater flow spilling out the break of the faulted steam generator.
- b. Stop high head safety injection charging pumps if (1) wide range reactor coolant pressure is stable or increasing, (2) pressurizer water level is greater than 12% of span (normal containment) and 28% of span (adverse containment) for Byron (Unit 1 and Unit 2) 15.2-24c REVISION 16 - DECEMBER 2016
B/B-UFSAR and 14% of span (normal containment) and 28% of span (adverse containment) for Braidwood (Unit 1 and Unit 2), (3) total feed flow to an intact steam generator is > 500 gpm or the narrow range level in at least one intact steam generator is > 10% of narrow range span (normal containment) and 31% of span (adverse containment) for Unit 1 (Byron and Braidwood) and 14%
of narrow range span (normal containment) and 34% of span (adverse containment) for Unit 2 (Byron and Braidwood), and (4) RCS subcooling is acceptable.
Isolating feedwater flow through the break allows additional auxiliary feedwater flow to be diverted to the intact steam generators.
Subsequent to recovery of level in the intact steam generators, the high head safety injection pumps will be turned off and plant operating procedures will be followed in cooling the plant to hot shutdown conditions.
15.2.8.2 Analysis of Effects and Consequences Method of Analysis A detailed analysis using the LOFTRAN Code (Reference 3) is performed in order to determine the plant transient following a feedwater line break.
The code describes the plant neutron kinetics, RCS including natural circulation, pressurizer, steam generators and feedwater system, and computes pertinent variables including the pressurizer pressure, pressurizer water level, and reactor coolant average temperature.
The limiting cases analyzed (with and without offsite power) assume a double ended break of the largest feedwater pipe at full power. Major assumptions made in the analyses are as follows:
- a. The initial power level bounds 100% of the nominal NSSS power level of 3672 MWt (including power uncertainty).
- b. Initial reactor coolant average temperature is 9.1F above the nominal value which includes a 1.5F bias. The initial pressurizer pressure is 43 psi below the nominal value.
- c. No credit is taken for the pressurizer spray. To determine that sufficient decay heat removal capability is maintained, this analysis does assume operation of the pressurizer power-operated relief valve in order to minimize RCS pressure (Tsat).
- d. Initial pressurizer level is at the nominal programmed value for the cases that assume the condensate storage tank for the auxiliary feedwater source. An evaluation of the effect of the pressurizer level 15.2-24d REVISION 16 - DECEMBER 2016
B/B-UFSAR uncertainty concluded that this analysis remains valid. For the cases that assume the Essential Service Water system for the auxiliary feedwater source, the initial pressurizer level is at the nominal programmed value plus 5% span.
- e. Initial steam generator water level is at the nominal value plus 5% NRS in the faulted steam generator and at the nominal value minus 5% NRS in the intact steam generators.
- f. Main feedwater flow to all steam generators is assumed to be lost at the time the break occurs (all main feedwater spills out through the break).
- g. A double-ended break area of 0.223 ft2 is assumed based on the maximum flow through the steam generator flow orifices.
- h. A conservative feedline break discharge quality is assumed prior to the time the reactor trip occurs, thereby maximizing the time the trip setpoint is reached. After the trip occurs, a saturated liquid discharge is assumed until all the water inventory is discharged from the faulted steam generator. This minimizes the heat removal capability of the faulted steam generator.
- i. Reactor trip is assumed to be initiated when the steam generator water level reaches the low-low steam generator water level setpoint of 18.6% of the narrow range span in the faulted steam generator, to allow for error.
- j. The auxiliary feedwater system is actuated by the low-low steam generator water level signal. The auxiliary feedwater system is assumed to supply a total of 453 gpm for all steam generator pressures. This is consistent with the auxiliary feedwater system control valves functioning as designed (to control flow to 160 gpm less measurement uncertainties).
The air supply to the control valves has been modified to provide a 30 minute safety related supply of air.
After the 30 minute supply of air is exhausted, the control valves will fail open. Since the supply of air is limited, the case in which the valves are assumed to conservatively fail (open) at event initiation has also been considered. Flow may be diverted out the break, but more flow may be provided to the intact steam generators at lower pressure.
This case has been found to be less limiting. For this case, credit is taken for operators to isolate the faulted steam generator 20 minutes after the reactor trip.
15.2-24e REVISION 16 - DECEMBER 2016
B/B-UFSAR
- k. Separate cases are analyzed assuming different sources for auxiliary feedwater resulting in different delay times for the initiation of auxiliary feedwater flow.
For the cases where the condensate storage tank is the auxiliary feedwater source and offsite power is available, a 55 second delay is assumed following the low-low steam generator water level setpoint being reached to allow time for sensor response, signal processing, and startup of the pump. For the cases where offsite power is lost, the delay time assumed is 63 seconds to account for additional delays for startup of the emergency diesel generators. In all of the cases analyzed, there is an additional 181 seconds before the feedwater lines are purged and the relatively cold (120°F) auxiliary feedwater enters the intact steam generators.
Additional cases assume the condensate storage tank is not available such that a switch is made to the Essential Service Water system as the source for the auxiliary feedwater system, which causes an additional 12-second delay to initiate the auxiliary feedwater flow. For the case with offsite power available, a 67 second delay is assumed following the low-low steam generator water level setpoint being reached to allow time for sensor response, signal processing, provision of sufficient water from the Essential Service Water system to operate the pump, and startup of the pump.
For the case where offsite power is lost, the delay time assumed is 75 seconds to account for additional delays for startup of the emergency diesel generators.
In all of the cases analyzed, there is an additional 181 seconds before the feedwater lines are purged and the relatively cold (102°F for Byron and 104°F for Braidwood) auxiliary feedwater enters the intact steam generators.
The limiting results were calculated for Unit 2 modeling the condensate storage tank as the AFW source.
- l. Credit is taken for heat energy deposited in some portions of the RCS metal during the RCS heatup.
- m. No credit is taken for charging or letdown.
- n. Steam generator heat transfer area is assumed to decrease as the shell side liquid inventory decreases.
- o. Conservative core residual heat generation is assumed based upon long-term operation at the initial power level preceding the trip (Reference 4).
15.2-24f REVISION 16 - DECEMBER 2016
B/B-UFSAR
- p. No credit is taken for the following potential protection logic signals to mitigate the consequences of the accident:
- 1. high pressurizer pressure,
- 2. overtemperature T,
- 3. high pressurizer level, and
- 4. high containment pressure.
- q. To account for potential variations associated with steam generator tube plugging, the analysis considers conservatively a maximum loop-to-loop flow variation of 7% and a maximum tube plugging level of 10% in any steam generator.
- r. Pressurizer heaters are modeled in the cases with offsite power available since their operation slightly reduces the margin to the acceptance criterion.
- s. A maximum steam generator blowdown valve leak rate of 3 gpm per steam generator was accounted for in the analysis.
Plant characteristics and initial conditions are further discussed in Subsection 15.0.3.
Only reactor control systems that reduce the margin to the acceptance criteria are assumed to function including the power-operated relief valves as previously discussed and, for cases with offsite power available, pressurizer heaters. The reactor protection system is required to function following a feedwater line break as analyzed here. No single active failure will prevent operation of this system.
The engineered safety systems that are assumed to function are the safety injection system, auxiliary feedwater system, and the main steamline isolation system. One train of safety injection has been assumed to be available. For the auxiliary feedwater system, the worst-case configuration has been assumed, i.e.,
three intact steam generators receive auxiliary feedwater following the break. The diesel-driven auxiliary feedwater pump has been assumed to fail also. The motor-driven auxiliary feedwater pump is assumed to deliver flow to each of the three intact steam generators with the auxiliary feedwater control valves operating properly. This configuration normally controls the flow to each steam generator at 160 gpm; however, after measurement uncertainties are considered, the minimum flow of 151 gpm is assumed to be provided to each of the three intact steam generators. Receipt of a low steamline pressure signal in at least one steam line initiates a steamline isolation signal that closes the main steamline isolation valves in all of the steam lines.
15.2-24g REVISION 16 - DECEMBER 2016
B/B-UFSAR Following the trip of the reactor coolant pumps for the feedline break without offsite power, there will be a flow coastdown until flow in the loops reaches the natural circulation value. The natural circulation capability of the RCS has been shown in Subsection 15.2.6, for the loss of ac power transient, to be sufficient to remove core decay heat following reactor trip.
Pump coastdown characteristics are demonstrated in Subsections 15.3.1 and 15.3.2 for single and multiple reactor coolant pump trips, respectively.
A detailed description and analysis of the safety injection system is provided in Section 6.3. The auxiliary feedwater system is described in Subsection 10.4.9. The main steamline isolation system is described in Subsection 10.3.2.
Operator Action The only operator action for which credit is taken in this transient is ensuring an adequate supply of auxiliary feedwater to maintain the secondary heat sink. This includes isolation of the faulted steam generator 20 minutes after reactor trip. This is not modeled in the limiting case, because the control valves are assumed to be operating correctly. The only other possible operator action before 20 minutes is the termination of safety injection (SI) according to the plant emergency operating procedures. However, the manual termination of SI is not required for this event since continued operation of SI would not degrade the results of this analysis.
Results The limiting results were calculated for Unit 2 modeling the condensate storage tank AFW source. Calculated plant parameters following a major feedwater line break analyzed to maximize the potential for losing RCS inventory are shown in Figures 15.2-13 through 15.2-20. Results for the case with offsite power available are presented in Figures 15.2-13 through 15.2-16.
Results for the case where offsite power is lost are presented in Figures 15.2-17 through 15.2-20. The calculated sequence of events for both cases analyzed is listed in Table 15.2-1.
Figures 15.2-13 and 15.2-17 show that following reactor trip, the plant remains subcritical. RCS pressure will be maintained at the power-operated relief valve setpoint until safety injection flow is terminated by the operator, as mentioned in Subsection 15.2.8.2. However, the reactor core remains covered with water throughout the transient, as water relief due to thermal expansion is limited by the heat removal capability of the auxiliary feedwater system and makeup is provided by the safety injection system.
The major difference between the two cases can be seen in the plots of hot and cold leg temperatures, Figure 15.2-15 (with offsite power available) and Figure 15.2-19 (without offsite power). The case with offsite power results in a more severe rise in temperature. The pressurizer fills for the case with 15.2-24h REVISION 16 - DECEMBER 2016
B/B-UFSAR power due to the increased coolant expansion resulting from the reactor coolant pump heat addition; hence, water is relieved for the case with power. As previously stated, however, the core remains covered with water for both cases.
15.2.8.3 Radiological Consequences The feedwater line break with the most significant consequences would be one that occurred inside the containment between a steam generator and the feedwater check valve. In this case, the contents of the steam generator would be released to the containment. The radioactivity released would be less than that for the steamline break, as analyzed in Subsection 15.1.5.3.
Furthermore, automatic isolation of the containment would further reduce any radiological consequences from this postulated accident.
15.2.8.4 Conclusions Results of the analyses show that for the postulated feedwater line break, the assumed auxiliary feedwater system capacity is adequate to remove decay heat, to prevent overpressurizing the RCS, and to prevent uncovering the reactor core. Radiological doses from the postulated feedwater line break would be less than those previously presented for the postulated steam line break.
15.2.9 References
- 1. Mangan, M. A., "Overpressure Protection for Westinghouse Pressurized Water Reactors," WCAP-7769, Rev. 1, June, 1972.
- 2. Deleted.
- 3. Burnett, T. W. T., et al., "LOFTRAN Code Description,"
WCAP-7907-P-A (Proprietary), WCAP 7907-A (Non-Proprietary),
April 1984.
- 4. "American National Standard for Decay Heat Power in Light Water Reactors," ANSI/ANS-5.1-1979, August 1979.
- 5. Freidland, A. J., et al., "Revised Thermal Design Procedure," WCAP-11397-P-A (Proprietary), and WCAP-11397-A (Non-Proprietary), April 1989.
- 6. NRC Information Notice 94-60, Potential Overpressurization of the Main Steam System, August 22, 1994.
15.2-24i REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.2-1 TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM TIME ACCIDENT EVENT (sec)
- 1. Unit 1 Turbine trip, loss of main 0.0 RCS Pressure Case feedwater flow High pressurizer pressure 5.4 reactor trip setpoint reached Rods begin to drop 7.4 Peak pressurizer pressure 7.6 Initiation of steam release 9.8 from steam generator safety valves
- 2. Unit 1 Turbine trip, loss of main 0.0 MSS Pressure Case feedwater flow Initiation of steam release 3.4 from steam generator safety valves Overtemperature T reactor 3.7 trip setpoint reached Rods begin to drop 11.7 Maximum SG pressure occurs 15.6
- 3. Unit 1 Turbine trip, loss of main 0.0 DNB Case feedwater flow Overtemperature T reactor 5.4 trip setpoint reached Initiation of steam release 6.0 from steam generator safety valves 15.2-25 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.2-1(Cont'd)
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM TIME ACCIDENT EVENT (sec)
Rods begin to drop 13.4 Minimum DNBR occurs 14.1
- 4. Unit 2 Turbine trip, loss of main 0.0 RCS Pressure Case feedwater flow High pressurizer pressure 4.5 reactor trip setpoint reached Initiation of steam release 5.9 from steam generator safety valves Peak pressurizer pressure 6.3 occurs Rods begin to drop 6.5
- 5. Unit 2 Turbine trip, loss of main 0.0 MSS Pressure Case feedwater flow Overtemperature T reactor 1.8 trip setpoint reached Initiation of steam release 4.8 from steam generator safety valves Rods begin to drop 9.8 Maximum SG pressure occurs 15.1 15.2-25a REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.2-1 (Cont'd)
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM TIME ACCIDENT EVENT (sec)
- 6. Unit 2 Turbine trip, loss of 0.0 DNB Case main feedwater flow Overtemperature T reactor 3.5 trip setpoint reached Initiation of steam release 7.9 from steam generator safety valves Rods begin to drop 11.5 Minimum DNBR occurs 12.6 15.2-26 REVISION 15 - DECEMBER 2014
BRAIDWOOD-UFSAR TABLE 15.2-1 (Cont'd)
TIME ACCIDENT EVENT (sec)
Loss of Nonemergency Main feedwater flow 10.1 a-c Power stops Low-low steam generator 67.4 water level trip Rods begin to drop 69.4 Reactor coolant pumps 71.4 begin to coastdown Opening of SG safety valves ~75.6 Four steam generators 142.4 begin to receive auxiliary feedwater from one motor driven auxiliary feedwater pump Core decay heat decreases ~328 to auxiliary feedwater heat removal capacity Cold auxiliary feedwater 696.0 is delivered to the steam generators Peak water level in 800.0 pressurizer occurs 15.2-27 REVISION 16 - DECEMBER 2016
BYRON-UFSAR TABLE 15.2-1 (Cont'd)
TIME ACCIDENT EVENT (sec)
Loss of Nonemergency Main feedwater flow 10.1 a-c Power stops Low-low steam generator 67.4 water level trip Rods begin to drop 69.4 Reactor coolant pumps 71.4 begin to coastdown Opening of SG safety valves ~75.6 Four steam generators 142.4 begin to receive auxiliary feedwater from one motor driven auxiliary feedwater pump Core decay heat decreases ~328 to auxiliary feedwater heat removal capacity Cold auxiliary feedwater 696.0 is delivered to the steam generators Peak water level in 790.0 pressurizer occurs 15.2-27a REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.2-1 (Cont'd)
TIME ACCIDENT EVENT (sec)
Loss of Normal Feed- Main feedwater flow 10 water Flow stops Low-low steam generator 51.3 water level trip Rods begin to drop 53.3 Four steam generators 106.3 begin to receive auxiliary feedwater from one motor-driven auxiliary feedwater pump Cold auxiliary feedwater 302 is delivered to the steam generators Peak water level in 1774 pressurizer occurs Core decay heat plus pump ~1950 heat decreases to auxiliary feedwater heat removal capacity 15.2-27b REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.2-1 (Cont'd)
TIME ACCIDENT EVENT (sec)
Feedwater System Pipe Break
- 1. With offsite power Main feedline break 10 available occurs Low-low steam generator 32 water level reactor trip setpoint reached in faulted steam generator Rods begin to drop 34 One motor-driven auxiliary 87 feedwater pump starts and supplies three intact steam generators Low steam line pressure 204 setpoint reached in faulted steam generator All main steam line 212 isolation valves close Cold auxiliary feedwater 268 is delivered to intact steam generators Steam generator safety 510 valve setpoint reached in intact steam generators Pressurizer water relief 1234 begins Core decay heat and ~4900 pump heat decreases to auxiliary feedwater heat removal capacity 15.2-28 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.2-1 (Cont'd)
TIME ACCIDENT EVENT (sec)
- 2. Without offsite Main feedline break 10 power occurs Low-low steam generator 32 level reactor trip setpoint reached in faulted steam generator Rods begin to drop 34 Power lost to the 36 reactor coolant pumps.
One auxiliary feedwater 95 pump starts and supplies three intact steam generators Low steam line pressure 256 setpoint reached in faulted steam generator All main steam line 264 isolation valves close Cold auxiliary feedwater 276 is delivered to intact steam generators Steam generator safety 795 valve setpoint reached in intact steam generators Core decay heat decreases ~1800 to auxiliary feedwater heat removal capacity 15.2-29 REVISION 15 - DECEMBER 2014
B/B-UFSAR Tables 15.2-2 and 15.2-3 have been deleted intentionally 15.2-30 and 15.2-31 REVISION 7 - DECEMBER 1998
B/B-UFSAR TABLE 15.2-4 This page has been intentionally deleted.
15.2-32 REVISION 9 - DECEMBER 2002
B/B-UFSAR 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE A number of faults are postulated which could result in a decrease in reactor coolant system flow rate. These events are discussed in this section. Detailed analyses are presented for the most limiting of these events.
Discussions of the following flow decrease events are presented in Section 15.3:
- a. partial loss of forced reactor coolant flow,
- b. complete loss of forced reactor coolant flow,
- c. reactor coolant pump shaft seizure (locked rotor), and
- d. reactor coolant pump shaft break.
Item a. above is considered to be an ANS Condition II event, item
- b. an ANS Condition III event, and items c. and d. ANS Condition IV events. Subsection 15.0.1 contains a discussion of ANS classifications.
15.3.1 Partial Loss of Forced Reactor Coolant Flow 15.3.1.1 Identification of Causes and Accident Description A partial loss of coolant flow accident can result from a mechanical or electrical failure in a reactor coolant pump, or from a fault in the power supply to the pump or pumps supplied by a reactor coolant pump bus. If the reactor is at power at the time of the accident, the immediate effect of loss of coolant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor is not tripped promptly.
Normal power for two of the reactor coolant pumps is supplied through individual buses connected to the generator, whereas the other two reactor coolant pumps are supplied from offsite power.
When a generator trip occurs, the buses which are normally fed from the generator are automatically transferred to an offsite power supply. The pumps will continue to supply coolant flow to the core. Following any turbine trip where there are no electrical faults or thrust bearing failure, which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The two reactor coolant pumps normally fed from the generator remain connected to the generator thus ensuring full flow for approximately 30 seconds after the reactor trip before any transfer is made.
This event is classified as an ANS Condition II incident (an incident of moderate frequency) as defined in Subsection 15.0.1.
15.3-1
B/B-UFSAR The necessary protection against a partial loss of coolant flow accident is provided by the low primary coolant flow reactor trip signal which is actuated in any reactor coolant loop by two out of three low flow signals. Above Permissive 8 (refer to Table 7.2-2 for a discussion of permissives), low flow in any loop will actuate a reactor trip. Between approximately 10% power (Permissive 7) and the power level corresponding to Permissive 8, low flow in any two loops will actuate a reactor trip. A reactor trip signal from the pump breaker position is provided as a backup to the low flow signal. When operating above Permissive 7, a breaker open signal from any two pumps will actuate a reactor trip. Reactor trip on reactor coolant pump breakers open signal is blocked below Permissive 7.
15.3.1.2 Analysis of Effects and Consequences Method of Analysis One case has been analyzed:
Loss of two pumps with four loops in operation.
This transient is analyzed by two digital computer codes. First, the LOFTRAN Code (Reference 1) is used to calculate the loop and core flow during the transient, the time of reactor trip based on the calculated flows, the nuclear power transient, and the primary system pressure and temperature transients. The VIPRE code (see Section 4.4) is then used to calculate the hot-channel heat flux transient and DNBR, based on the nuclear power and RCS temperature (enthalpy), pressure, and flow from LOFTRAN. The DNBR transients presented represent the minimum of the typical or thimble cell.
This accident is analyzed with the revised thermal design procedure as described in WCAP-11397-P-A (Reference 3). Plant characteristics and initial conditions are discussed in Subsection 15.0.3.
Initial Conditions Initial reactor power (consistent with uprated power conditions) and pressure are assumed to be at their nominal values. The initial temperature is assumed to be at the nominal value for the high Tavg program plus a 1.5oF bias. Uncertainties in initial conditions are included in the limit DNBR as described in WCAP-11397-P-A.
Reactivity Coefficients A conservatively large absolute value of the Doppler-only power coefficient is used (see Figure 15.0-3). This is equivalent to a total integrated Doppler reactivity from 0% to 100% power of 0.016 k.
15.3-2 REVISION 15 - DECEMBER 2014
B/B-UFSAR A moderator temperature coefficient (MTC) of 0 pcm/oF is assumed in the analysis. The use of this MTC is consistent with the analysis initial conditions assumptions and corresponds to the applicable MTC limit at hot full power (HFP) initial conditions.
The HFP analysis results using a 0 pcm/oF MTC bound those for part-power initial conditions with a positive MTC at the licensed allowable MTC limit.
Flow Coastdown The flow coastdown analysis is based on a momentum balance around each reactor coolant loop and across the reactor core. This momentum balance is combined with the continuity equation, a pump momentum balance and the pump characteristics and is based on high estimates of system pressure losses.
Plant systems and equipment credited for mitigating the effects of the accident are discussed in Subsection 15.0.8 and listed in Table 15.0-7. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.
Results Figures 15.3-1 and 15.3-2 show the transient response for the loss of two reactor coolant pumps with four loops in operation.
The reactor is assumed to be tripped on a low flow signal.
Figure 15.3-2 shows the DNBR to be greater than the limit value.
For the case analyzed, since DNB does not occur, the ability of the primary coolant to remove heat from the fuel rod is not greatly reduced. Thus, the average fuel and clad temperatures do not increase significantly above their respective initial values.
The calculated sequence of events for the case analyzed is shown on Table 15.3-1. The affected reactor coolant pumps will continue to coast down, and the core flow will reach a new equilibrium value corresponding to the number of pumps still in operation. With the reactor tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed.
15.3.1.3 Radiological Consequences A partial loss of reactor coolant flow from full load would result in a reactor and turbine trip. Assuming that the condenser is not available, atmospheric steam dump may be required.
15.3-3 REVISION 11 - DECEMBER 2006
B/B-UFSAR There are only minimal radiological consequences associated with this event. Therefore, this event is not limiting.
The radiological consequences resulting from atmospheric steam dump would be less severe than the steamline break event analyzed in Subsection 15.1.5.3.
15.3.1.4 Conclusions The analysis shows that the DNBR will not decrease below the limit value at any time during the transient. Thus, no fuel or clad damage is predicted, and all applicable acceptance criteria are met.
The radiological consequences of this event would be less than the steamline break event analyzed in Subsection 15.1.5.3.
15.3.2 Complete Loss of Forced Reactor Coolant Flow 15.3.2.1 Identification of Causes and Accident Description A complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all reactor coolant pumps. If the reactor is at power at the time of the accident, the immediate effect of loss of coolant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly.
Normal power for two of the reactor coolant pumps is supplied through buses from a transformer connected to the generator, whereas the other two reactor coolant pumps are supplied from offsite power. When a generator trip occurs, the buses which are normally fed from the generator are automatically transferred to an offsite power supply. The pumps will continue to supply coolant flow to the core. Following any turbine trip where there are no electrical faults or thrust bearing failure which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The two reactor coolant pumps normally fed from the generator remain connected to the generator thus ensuring full flow for 30 seconds after the reactor trip before any transfer is made.
This event is classified as an ANS Condition III incident (an infrequent incident) as defined in Subsection 15.0.1.
The following signals provide the necessary protection against a complete loss of flow accident:
- a. Reactor coolant pump power supply undervoltage or underfrequency.
- b. Low reactor coolant loop flow.
15.3-4 REVISION 9 - DECEMBER 2002
B/B-UFSAR The reactor trip on reactor coolant pump bus undervoltage is provided to protect against conditions which can cause a loss of voltage to all reactor coolant pumps, i.e., loss of nonemergency a-c power. This function is blocked below approximately 10%
power (Permissive 7).
The reactor trip on reactor coolant pump underfrequency is provided to trip the reactor for an underfrequency condition, resulting from frequency disturbances on the power grid.
Reference 4 provides analyses of grid frequency disturbances and the resulting nuclear steam supply system (NSSS) protection requirements which are generally applicable.
The reactor trip on low primary coolant loop flow is provided to protect against loss of flow conditions which affect only one reactor coolant loop. This function is generated by two out of three low flow signals per reactor coolant loop. Above Permissive 8, low flow in any loop will actuate a reactor trip.
Between approximately 10% power (Permissive 7) and the power level corresponding to Permissive 8, low flow in any two loops will actuate a reactor trip.
15.3.2.2 Analysis of Effects and Consequences The following cases for complete loss of forced reactor coolant flow are analyzed:
- 1. Complete loss of all four reactor coolant pumps (RCPs) with four loops in operation; and
- 2. Frequency decay event resulting in a complete loss of forced reactor coolant flow.
Case 1 of the complete loss of flow event assumes that the RCPs begin to coastdown upon reaching a undervoltage trip setpoint (modeled to occur at t=0 seconds in this analysis). Rod motion following the undervoltage trip is modeled to occur at t=1.5 seconds, reflecting an undervoltage trip time delay of 1.5 seconds. For the underfrequency complete loss of flow event (Case 2), a frequency decay of 5 Hz/sec is assumed to occur at t=0 seconds, decreasing RCS flow to all loops. At t=1.2 seconds, the underfrequency trip setpoint of 54.0 Hz is reached. Rod motion occurs at t=1.8 seconds, following a 0.6 second underfrequency trip time delay.
These transients are analyzed by two digital computer codes.
First, the LOFTRAN code (Reference 1) is used to calculate the loop and core flow transients, the nuclear power transient, and the primary system pressure and temperature transients. The flow coastdown analysis performed by LOFTRAN is based on a momentum balance around each reactor coolant loop and across the reactor core.
15.3-5 REVISION 15 - DECEMBER 2014
B/B-UFSAR This momentum balance is combined with the continuity equation, a pump momentum balance, and the as-built pump characteristics and is based on conservative system pressure loss estimates.
The VIPRE code (see Section 4.4) is then used to calculate the hot-channel heat flux transient and DNBR, based on the nuclear power and RCS temperature (enthalpy), pressure, and flow from LOFTRAN. The DNBR results are based on the minimum of the typical and thimble cells.
Results Figures 15.3-3 and 15.3-4 show the transient response for the limiting case; the frequency decay complete loss of flow event (Case 2). The reactor is assumed to be tripped on an underfrequency signal. Figure 15.3-4 shows the DNBR to be greater than the limit value.
Since DNB does not occur for the cases analyzed, the ability of the primary coolant to remove heat from the fuel rod is not greatly reduced. Thus, the average fuel and clad temperatures do not increase significantly above their respective initial values.
15.3-5a REVISION 15 - DECEMBER 2014
B/B-UFSAR The calculated sequence of events for the case analyzed is shown on Table 15.3-1. The speed of the reactor coolant pumps will continue to decrease from the 5 hz/sec frequency decay until a pump trip occurs on an underfrequency condition. Following pump trip, the reactor coolant pumps will continue to coast down, and natural circulation flow will eventually be established, as demonstrated in Subsection 15.2.6. With the reactor tripped, a stable plant condition would be attained. Normal plant shutdown may then proceed.
15.3.2.3 Radiological Consequences A complete loss of reactor coolant flow from full load results in a reactor and turbine trip. Assuming, in addition, that the condenser is not available, atmospheric steam dump would be required. The quantity of steam released would be the same as for a loss of offsite power.
There are only minimal radiological consequences associated with this event. Therefore, this event is not limiting.
The radiological consequences resulting from atmospheric steam dump would be less severe than the steamline break analyzed in Subsection 15.1.5.
15.3.2.4 Conclusions The analysis performed has demonstrated that for the complete loss of forced reactor coolant flow, the DNBR does not decrease below the limit value at any time during the transient. Thus, no fuel or clad damage is predicted, and all applicable acceptance criteria are met.
15.3.3 Reactor Coolant Pump Shaft Seizure (Locked Rotor) 15.3-6 REVISION 9 - DECEMBER 2002
B/B-UFSAR 15.3.3.1 Identification of Causes and Accident Description A transient analysis is performed for the instantaneous seizure of an RCP rotor (locked rotor). Flow through the affected reactor coolant loop is rapidly reduced, leading to a reactor trip on a low flow signal. Following the trip, heat stored in the fuel rods continues to pass into the core coolant, causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generator is reduced, first because the reduced flow results in a decreased tube side film coefficient, and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with the reduced heat transfer in the steam generator causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer causes a pressure increase which in turn actuates the automatic spray system, opens the power-operated relief valves (PORVs), and opens the pressurizer safety valves. The sequence of events initiated by the insurge depends on the rate of insurge and pressure increase. The PORVs are designed for reliable operation and would be expected to function properly during the accident. However, for conservatism, their pressure-reducing effect as well as the pressure-reducing effect of the spray is not included in this analysis.
This event is classified as an ANS Condition IV incident (a limiting fault) as defined in Subsection 15.0.1.
The consequences of a locked rotor (i.e., an instantaneous seizure of a pump shaft) are very similar to those of a pump shaft break. The initial rate of the reduction in coolant flow is slightly greater for the locked rotor event. However, with a broken shaft, the impeller could conceivably be free to spin in the reverse direction. The effect of reverse spinning is to decrease the steady-state core flow when compared to the locked rotor scenario. Only one analysis, which permits reverse spinning but no forward flow, has been performed and represents the most limiting condition for the locked rotor and pump shaft break accidents.
15.3.3.2 Analysis of Effects and Consequences Method of Analysis Two digital computer codes are used to analyze this transient.
The LOFTRAN Code (Reference 1) is used to calculate the resulting loop and core flow transients following the pump seizure, the time of reactor trip based on the loop flow transients, the nuclear power following reactor trip, and the peak RCS pressure.
The thermal behavior of the fuel located at the core hot spot is investigated using the FACTRAN Code (Reference 2), which uses the core flow and the nuclear power values calculated by LOFTRAN.
The FACTRAN Code includes a film boiling heat transfer coefficient.
15.3-7 REVISION 7 - DECEMBER 1998
B/B-UFSAR One case is analyzed: one locked rotor/shaft break with four loops in operation, concurrent with a loss of offsite power at the time of trip.
Initial Operating Conditions At the beginning of the postulated locked rotor accident, the plant is assumed to be operating at nominal reactor power consistent with uprated power conditions and pressure. Initial temperature is assumed to be nominal plus a 1.5oF bias. The revised thermal design procedure as described in WCAP-11397-P-A (Reference 3) is used. Plant characteristics and initial conditions are discussed in Subsection 15.0.3.
For the peak pressure and peak clad temperature evaluations, one analysis is performed and the initial pressure is conservatively estimated as 43 psi above the nominal pressure of 2250 psia to allow for errors in the pressurizer pressure measurement and control channels. This is done to obtain the highest possible rise in the coolant pressure during the transient. To obtain the maximum pressure in the primary side, conservatively high loop pressure drops are added to the calculated pressurizer pressure.
The pressure response shown in Figure 15.3-6 is at the point in the RCS having the maximum pressure (i.e., the outlet of the faulted loop's RCP). The remainder of the plant is assumed to be operating under the most adverse steady-state operating condition, e.g., 100% of the NSSS MUR thermal power including applicable uncertainties and the maximum steady-state coolant average temperature, including uncertainties and a 1.5oF bias.
For a conservative analysis of fuel rod behavior, the hot spot evaluation assumes that DNB occurs at the initiation of the transient and continues throughout the event. This assumption reduces heat transfer to the coolant and results in conservatively high hot spot temperatures.
The reactor coolant flow coastdown analysis is based on a momentum balance around each reactor coolant loop and across the reactor core. This momentum balance is combined with continuity equation, a pump momentum balance, and the as-built pump characteristics and is based on high estimates of system pressure losses. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.
A conservatively large absolute value of the Doppler-only power coefficient is used (see Figure 15.0-3). The total integrated Doppler reactivity from 0 to 100% power is assumed to be 0.016 K.
A moderator temperature coefficient (MTC) of 0 pcm/oF is assumed in the analysis. The use of this MTC is consistent with the analysis initial conditions assumptions and corresponds to the applicable MTC limit at hot full power (HFP) initial conditions.
The HFP analysis results using a 0 pcm/oF MTC bound those for part-power initial conditions with a positive MTC at the licensed allowable MTC limit. For this analysis, the curve of trip reactivity versus time (Figure 15.0-6) was used with a 4% K trip reactivity, which includes the most reactive rod cluster control assembly (RCCA) stuck out of the core.
15.3-8 REVISION 15 - DECEMBER 2014
B/B-UFSAR Evaluation of the Pressure Transient A detailed model was used to determine the peak pressure in the RCS under postulated accident conditions and to obtain the neutron flux response as a function of time, which is used later in the analysis.
After pump seizure, neutron flux is rapidly reduced because of the control rod insertion upon plant trip. In this analysis, rod motion is assumed to begin one second after the flow in the affected loop reached 85.1% of nominal flow.
No credit was taken for the pressure-reducing effect of the pressurizer relief valves, pressurizer spray, steam dump or controlled feedwater flow after plant trip. Although these operations are expected to occur and would result in a lower peak pressure, an additional degree of conservatism is provided by ignoring their effect.
Upon actuation of the pressurizer safety valves at an opening pressure of 2549.9 psia (including 2% allowance for drift and 1%
for pressure shift), purge of the water in the safety valve loop seal occurs and full valve relief capacity is achieved within 1 second. The pressurizer safety valves capacity for steam relief is described in Chapter 5.
Evaluation of DNB in the Core During the Accident For this accident, DNB is assumed to occur in the core.
Therefore, an evaluation of the consequences with respect to fuel rod thermal transients is performed. Results obtained from analysis of this hot spot condition represent the upper limit with respect to clad temperature and zirconium water reaction.
In the evaluation, the rod power at the hot spot is assumed to be 2.6 times the average rod power (i.e., FQ=2.6) at the initial core power level.
Calculation of the extent of DNB in the core during the locked rotor event is performed with the VIPRE computer code (See Section 4.4). The VIPRE code calculates the hot-channel heat flux transient and DNBR, based on the nuclear power and RCS temperature (enthalpy), pressure, and flow from LOFTRAN.
Film Boiling Coefficient The film boiling coefficient is calculated in the FACTRAN Code using the Bishop-Sandberg-Tong film boiling correlation (Reference 5). The fluid properties are evaluated at film temperature (average between wall and bulk temperatures). The program calculates the film coefficient at every time step based upon the actual heat transfer conditions at the time. The neutron flux, system pressure, bulk density, and mass flow rate as a function of time are used as program input.
15.3-9 REVISION 15 - DECEMBER 2014
B/B-UFSAR For this analysis, the initial values of the pressure and bulk density are used throughout the transient since they are the most conservative with respect to the clad temperature response. As indicated earlier, DNB was assumed to start at the beginning of the accident.
Fuel-Clad Gap Coefficient The magnitude and time dependence of the heat transfer coefficient between fuel and cladding (gap coefficient) have a pronounced influence on the thermal results. The larger the value of the gap coefficient, the more heat is transferred between pellet and clad. For the first part of the transient, a high gap coefficient produces higher clad temperatures since the heat stored and generated in the fuel pellet tries to redistribute itself in the cooler clad. This effect of the gap coefficient is reversed when the clad temperature exceeds the pellet temperature in cases where a zirconium-steam reaction is present. The gap coefficient was taken to be the conservatively large value of 10,000 Btu/hr-ft2-F which is greater than the highest value calculated during core life.
Zirconium-Steam Reaction The zirconium-steam reaction can become significant above 1800F (clad temperature). In order to take this phenomenon into account, the following correlation, which defines the rate of the zirconium-steam reaction, was introduced into the model (see Reference 4):
d(w 2)
(33.3 x 106)e-(45500/1.986T) dt where w = amount reacted, mg/cm2 t = time, seconds T = temperature, K Results The transient results without offsite power available are shown in Figures 15.3-5 through 15.3-6. The peak RCS pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits. Also, the peak clad surface temperature is considerably less than the limit of 2375F (associated with Optimized ZIRLO' fuel cladding which bounds the 2700F limit for ZIRLO fuel cladding, Reference 7).
It should be noted that the clad temperature was conservatively calculated assuming the DNB occurs at the initiation of the transient. The results of these calculations (peak pressure, peak clad temperature, and zirconium-steam reaction) are also summarized in Table 15.3-2.
The calculated sequence of events is shown in Table 15.3-1.
Figure 15.3-6 shows that the core flow rapidly coasts down to a new equilibrium value. With the reactor tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed.
15.3-9a REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.3.3.3 Conclusion The results of the transient analysis show that 0.10% of the fuel rods will experience DNBR values below the safety analysis limit value. This is considerably less than the transient limit of 2.0%.
Since the peak RCS pressure reached during the transient is less than that which would cause stresses to exceed the faulted condition stress limits, the integrity of the primary coolant system is not endangered. Also, since the peak clad average temperature calculated for the hot spot during the worst transient remains considerably less than the limit of 2375F (associated with Optimized ZIRLO' fuel cladding which bounds the 2700F limit for ZIRLO fuel cladding, Reference 7) the core will remain in place and intact with no loss of core cooling capability.
15.3-9b REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.3.3.4 Radiological Consequences The evaluation of the radiological consequences of a postulated seizure of a reactor coolant pump rotor (Locked Rotor Accident-LRA) assumes that the reactor has been operating with a small percent of defective fuel and leaking steam generator tubes for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant. A concurrent PORV failure is also assumed.
It is assumed conservatively that, as a result of the locked rotor accident, 2% of the fuel rods in the core undergo sufficient clad damage to result in the release of their gap activity.
As a result of the accident, radionuclides carried by the primary coolant to the steam generators, via the leaking tubes, are released to the environment via the steam line safety or power operated relief valves.
The key inputs and assumptions used in the AST Locked Rotor Accident (LRA) radiological consequence analysis using AST methodology are summarized below and provided in Table 15.3-4.
This LRA analysis postulates the instantaneous seizure of a Reactor Coolant Pump (RCP) rotor, where the reactor is tripped on the subsequent low flow signal. Following the trip, heat stored in fuel rods continues to pass into the reactor coolant, causing the coolant to expand. At the same time, heat transfer to the shell side of the Steam Generator (SG) is reduced, first because the reduced flow results in a decreased tube side film coefficient, and then because the primary reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with the reduced heat transfer in the SG, causes an insurge of coolant into the pressurizer and a pressure increase throughout the Reactor Coolant System (RCS). This insurge into the pressurizer causes a pressure increase, which in turn actuates the automatic spray system, opens the power-operated relief valves (PORVs), and also opens the pressurizer safety valves.
The SG PORVs are designed for reliable operation and would be expected to function properly during the accident. However, for conservatism, their pressure reducing effect is ignored, and an SG PORV failure, in the open position at the onset of accident releases, is assumed. In addition, the pressure reducing effect of the spray is also ignored for this analysis.
This evaluation of the radiological consequences of an LRA assumes that the reactor has been operating with a small percent of defective fuel and leaking SG tubes. The reactor is assumed to have been operating in this condition for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and secondary coolant. Additionally, prior to 15.3-10 REVISION 12 - DECEMBER 2008
B/B-UFSAR accident initiation, the reactor is postulated to experience an Iodine spike, thereby increasing the RCS iodine activity above that of equilibrium levels.
As a result of this accident, radionuclides carried by the primary coolant to the Steam Generators, via the leaking tubes, are released to the environment via the steam line safety valves or PORVs.
This LRA dose assessment is analyzed using two modeled simulations. The first simulation, Case 1, is modeled to calculate the doses due to the activity that was instantaneously released into the primary RCS from the postulated damaged fuel fraction, and the activity resulting from a pre-accident 60 Ci/gm Dose Equivalent (DE) I-131 spike. Leak and steaming rates are used to model the transport of activity from the RCS to the environment, through the PORVs of intact SGs, and through the failed PORV of the faulted SG, which is the postulated single-failure for this analysis.
The second simulation, Case 2, is modeled to calculate the doses due to 0.1 Ci/gm DE I-131 equilibrium activity existing in the secondary coolant prior to accident initiation. This iodine activity is released using the same partitioned steaming rates that are associated with Case 1 SG PORV release to the environment. For the intact SGs this iodine activity is released the same partitioned steaming rates that are associated with Case
- 1. However, for the SG with the failed PORV, referred to as the faulted SG, it is postulated that the activity initially contained in the faulted SG is released to the environment for 20 minutes. The failed SG PORV is isolated by locally closing the associated isolation valve. The operator can identify the failed SG PORV when the faulted SG pressure drops below SG PORV reset value. There is also a positive valve position indication of the open SG PORV on the control board. As a result of the failed SG PORV, the operator would enter the emergency procedure for "Faulted Steam Generator Isolation". The dispatch, travel time, and local isolation of this valve by an operator was conservatively assumed to require 20 minutes, from the time the SG PORV fails open.
Fuel Damage and Core Source Term The LRA core source terms are those associated with a DBA power level of 3658.3 MWt. This power level bounds the MUR power uprate Rated Thermal Power level including measurement uncertainties.
The instantaneous seizure of the RCP rotor associated with the LRA results in the damage of 2% of the core. The design basis of this accident assumes that no fuel melt is postulated to occur.
Therefore for Case 1, the source term available for release is associated with this fraction of damaged fuel and the fraction of core activity existing in the gap, plus the iodine in the RCS due to a design basis pre-accident 60 Ci /gm DE I-131 spike, and the noble gas activity associated with 1% fuel defects.
15.3-10a REVISION 15 - DECEMBER 2014
B/B-UFSAR The additional source activity modeled in the second case consists simply of the 0.1 Ci/gm DE I-131 equilibrium secondary coolant activity concentration limit from the Technical Specifications. The total activity available for release from both the intact SGs and the SG with a failed SG PORV are input to the RADTRAD NIF for Case 2.
Activity Release Fractions Release fractions and transport fractions are per Regulatory Guide 1.183, Appendix G and Table 3. To comply with this regulatory guidance, 5% of the core inventory of iodine and noble gas is assumed to be in the fuel-clad gap, excluding I-131 and Kr-85, where 8% and 10% are assumed, respectively. Additionally, Table 3 of RG 1.183 shows that 12% of the core cesium and rubidium should be assumed to be in the fuel-clad gap. However, to accommodate the consideration of extended fuel burnup, in excess of the RG 1.183 assumptions, all RG 1.183 Table 3 Other Noble Gases, Other Halogens, and Alkali Metals isotopic release fractions are doubled. Although analyses have shown that isotopic activity fractions in the fuel-clad gap may in fact decrease when "burning" the fuel longer than the 62 GWd/MTU specified in RG 1.183, this 100% increase in the gap fractions is used as an accepted and conservative means of bounding all extended burnup phenomena. All of the gap activity in the damaged fuel is released in its entirety, instantaneously into the RCS and mixed homogeneously therein.
These activity release fractions are input to the RADTRAD code through the use of the Release Fractions and Timing (RFT) file.
Airborne Activity Removal Mechanisms in Containment As discussed below only decay and leakage are credited.
Decay Credited:
Decay of radioactivity is credited in all compartments, prior to release. This is implemented in RADTRAD using the half-lives in the Nuclide Inventory File (NIF). The RADTRAD decay plus daughter option is used. In reality, daughter products such as xenon from iodines or iodines from tellurium are unlikely to readily escape from the matrix in which the parent iodine or tellurium is contained. Nevertheless, the RADTRAD feature to include daughter effects is selected for conservatism.
Depletion due to Leakage Credited:
For analyses of doses due to release from the RCS volume, the dose results from leakage, and it is reasonable to credit the small amount of depletion from the available RCS activity inventory associated with this leakage. This is calculated inherently by the RADTRAD code.
15.3-10b REVISION 12 - DECEMBER 2008
B/B-UFSAR Release Rates, Steaming Rates, and Partitioning Factors:
Activity that originates in the primary RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. This design basis leak rate value is 0.218 gpm, per intact SG, totaling 0.654 gpm, and 0.5 gpm for the SG with the failed SG PORV. For input into RADTRAD, these rates were converted from gallons per minute to cubic feet per minute, making them 0.02914 cfm, per intact SG, totaling 0.08743 cfm, and 0.06684 cfm for the SG with the failed SG PORV.
Releases to the environment are associated with the secondary coolant steaming from the Steam Generators with intact SG PORVs and releases directly out of an SG with a failed SG PORV.
Because of the release dynamic of the activity from the intact SG PORVs, RG 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water for this release path.
For Iodine, the partitioning factor of 0.01 was taken directly from the suggested guidance. However, there is no explicit guidance with regard to other particulate nuclides. Reviewing the specified AST release fractions, it is concluded that the only nuclides other than iodines to be released from the core source term are cesiums, rubidium, and noble gases. For cesiums and rubidium, a bounding partitioning factor of 0.0055 is used, as shown in the applicable ANSI Standard, ANS/ANSI-18.1-1999.
This value bounds the actual 0.00529 factor that is shown for Cs-134 in the ANSI Standard, which shows the largest partitioning factor of these such isotopes. Because of the volatility of noble gases, no partitioning is assumed for any such isotopes.
The methodology used to model steaming of activity through intact SG PORVs following the postulated LRA event is applicable to both the Case 1 and Case 2 simulations. This methodology assumes an average cumulative release rate through the SG valves that is reduced in steps. The partitioning factors are applied to these release rates, which were derived from the total time increment mass releases. Incremental steam mass releases are in pounds.
Release rates were derived by dividing these totals by the time increment. Then, this data was converted using the assumption of cooled liquid conditions (i.e., 62.4 lbm/ft3), as specified by the guidance of RG 1.183. The steaming release and primary-to-secondary coolant leakage is postulated to end at 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, when the RCS and secondary loop have equilibrated. The table below shows the time steps, isotopic partitioning factors, and associated release rates:
15.3-10c REVISION 12 - DECEMBER 2008
B/B-UFSAR LRA Partitioning Factors And Associated Release Rates Steam Total Steam Steam Steam Release Time Mass Iodine Cesium Noble Gas Release Rate Release Rate Rate for Interval Release Partitioning Partitioning Partitioning for Iodines for Cesiums Noble Gases (hrs) (lbm) Factor Factor Factor (cfm) (cfm) (cfm) 0 - 2.0 719,000 0.01 0.0055 1.0 9.5977E-01 5.2788E-01 9.5977E+01 2.0 - 1,109,000 0.01 0.0055 1.0 4.9346E-01 2.7140E-01 4.9346E+01 8.0 8.0 - 40 2,664,000 0.01 0.0055 1.0 2.2226E-01 1.2224E-01 2.2226E+01 40 - 64 0 0.01 0.0055 1.0 0 0 0 The release rate through the failed SG PORV is conservatively assumed to be un-partitioned, and therefore no isotopic partitioning factors are applied. The rate at which activity is released from this pathway is therefore equal to the primary-to-secondary coolant leak rate discussed above.
In Case 2, for the intact SGs, the iodine activity is released using the same partitioned steaming rates that are associated with Case 1 SG PORV release to the environment. For the faulted SG the release rate is based on a 20-minute failed SG PORV release. The total mass released is 167,000 lbm. Converted to a volume, and divided by the 20-minute release time, this becomes the 133.75 cfm volumetric flow rate.
/Q Calculations (Meteorology)
Releases from the SG PORVs are considered elevated releases due to the high steaming rates, and the associated /Q's have been reduced by a factor of 5 per guidance in RG 1.194. The atmospheric dispersion factors are given in Table 15.0-17.
Assumptions and Inputs The following inputs and assumptions were used in the LRA analysis.
- a. Core inventory is based on a DBA power level of 3658.3 MWt. This power level bounds the MUR power uprate Rated Thermal Power level including measurement uncertainties.
- b. Two percent (2%) of the fuel is damaged during the initiation of this accident, and is assumed to have failed.
- c. No fuel melts following the postulated LRA.
- d. Five percent (5%) of the core inventory of noble gases and iodines are released from the fuel gap, excluding I-131 and Kr-85, where 8%
and 10% are respectively released. Release fractions of other nuclide groups contained in the fuel gap are detailed in Table 3 of Regulatory Guide 1.183, and to account for gap fraction uncertainty due to expected extended fuel burnup, these fractions from the referenced table are doubled.
15.3-10d REVISION 15 - DECEMBER 2014
B/B-UFSAR
- e. All iodine released from the SGs is assumed to be of the elemental species. This is done for RADTRAD simulation considerations, and is consistent with the RG 1.183 specification of 97% elemental and 3%
organic, because elemental and organic iodine are treated identically by the computer model.
- f. The Control Room HVAC system is realigned to the emergency mode of operation 30 minutes after the initiation of this design basis accident.
- g. The activity released from the fuel from either the gap or from fuel pellets is assumed to be instantaneously mixed with the reactor coolant within the pressure vessel per RG 1.183.
- h. SG PORV releases end at 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, when the RCS has seen a large enough reduction in residual heat to no longer require steaming via the PORVs for temperature reduction.
- i. A failure of one SG PORV, in the open position, that takes 20 minutes to isolate, is assumed to conservatively maximize activity release.
Dose Results Radiological doses resulting from a design basis LRA for a control room operator and a person located at EAB or LPZ are to be less than the regulatory dose limits as given below:
Regulatory Dose Limits - LRA Dose Type Control Room (rem) EAB and LPZ (rem)
TEDE Dose 5a 2.5b Notes:
a 10 CFR 50.67 b
10 CFR 50.67 as modified by Regulatory Guide 1.183 (Table 6, Page 1.183-20) 15.3-10e REVISION 12 - DECEMBER 2008
B/B-UFSAR The table below provides the results from the simulations modeled using the RADTRAD 3.03 code, as well as the summed result:
Locked Rotor Accident Radiological Analysis Results (Maximum for Both Byron or Braidwood)
Case 1: Doses from Iodine Spike and Fuel Damage RCS Activity Dose Assessment Results Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 2.729 1.639 0.594 Case 2: Doses from Equilibrium Secondary Coolant Activity Dose Assessment Results Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 0.061 0.040 0.008 Total Dose from Design Basis Locked Rotor Accident (LRA)
Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 2.790 1.679 0.602 These doses are below the Regulatory Dose Limits, so it is verified that the LRA is sufficiently mitigated at both Byron and Braidwood Stations.
15.3-10f REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.3.3.4.1 Source Term The concentration of nuclides in the primary and secondary system prior to and following the accident are determined as follows:
- a. The iodine concentrations in the reactor coolant will be based upon a preaccident iodine spike and 2% failed fuel.
- 1. Preaccident Spike - A reactor transient has occurred prior to the LRA and has raised the primary coolant iodine concentration to 60 Ci/gm of Dose Equivalent (DE) I-131 (Table 15.0-10).
- 2. Failed Fuel - 2% of the fuel rods in the core suffer clad damage due to the LRA and release all their iodine gap activity to the primary coolant.
- b. The noble gas concentrations in the primary coolant are based on 1 percent defective fuel existing prior to the LRA, plus 5% of the inventory of noble gases and iodines released from the fuel gap of the damaged fuel, excluding I-131 and Kr-85, where 8% and 10% are respectively released. Release fractions of other nuclide groups contained in the fuel gap are detailed in Regulatory Guide 1.183 Table 3. To account for gap fraction uncertainty due to expected extended fuel burnup, fractions from Table 3 are doubled.
15.3.3.5 Radiological Conclusions The resulting radiation doses in the control room, at the exclusion area boundary, and at the low-population zone outer boundary are presented in Tables 15.0-11 and 15.0-12. The doses from this accident are within the NRCs dose acceptance criterion of 10 CFR Part 50.67 guidelines.
15.3-11 REVISION 12 - DECEMBER 2008
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.3-12 REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.3.4 Reactor Coolant Pump Shaft Break Refer to Section 15.3.3 15.3-13 REVISION 9 - DECEMBER 2002
B/B-UFSAR Pages 15.3-14 through 15.3-16 have been deleted intentionally.
15.3-14 through 15.3-16 REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.3.5 Locked Rotor with Loss of Offsite Power Subsection 15.3.3 assumes a locked-rotor/shaft break concurrent with a loss of offsite power at time of trip.
15.3.6 References
- 1. Burnett, T. W. T., et al., "LOFTRAN Code Description",
WCAP-7907-P-A (proprietary), WCAP 7907-A (non-proprietary),
April 1984.
- 2. Hargrove, H. G., "FACTRAN - A Fortran - IV Code for Thermal Transients in a U02 Fuel Rod", WCAP-7908-A, December 1989.
- 3. Friedland, A.J. and Ray, S., "Revised Thermal Design Procedure," WCAP-11397-P-A (proprietary) and WCAP-11397-A (non-proprietary), April 1989.
- 4. Baldwin, M. S., Merrian, M. M., Schenkel, H. S. and Van De Walle, D. J., "An Evaluation of Loss of Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs," WCAP-8424, Revision 1, June 1975.
- 5. Bishop, A. A., Sandberg, R. O. and Tong, L. S., "Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31, August 1965.
- 6. NUREG-0800, Standard Review Plan Section 15.3.3-15.3.4, Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break, Revision 2, July 1981.
- 7. Shah, H. H., Optimized ZIRLO', WCAP-12610-P-A &
CENPD-404-P-A Addendum 1-A, July 2006.
15.3-17 REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.3-1 TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN REACTOR COOLANT SYSTEM FLOW ACCIDENT EVENT TIME (sec.)
Partial Loss of Forced Reactor Coolant Flow Four loops operating two pumps coasting down Coastdown begins 0.0 Low flow reactor trip 1.7 Rods begin to drop 2.7 Minimum DNBR occurs 3.8 Complete Loss of Forced Reactor Coolant Flow (Underfrequency)
Frequency decay to all four RCPs begins 0.0 Underfrequency trip setpoint is reached 1.2 Rods begin to drop 1.8 Minimum DNBR occurs 3.9 15.3-18 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.3-1 (Cont'd)
ACCIDENT EVENT TIME (sec.)
Four Loop Operation Reactor Coolant Pump Locked Rotor/Shaft Break Rotor on one pump locks 0.0 Low flow trip point reached 0.04 Rods begin to drop 1.04 Maximum RCS pressure occurs 3.60 Maximum clad temperature occurs 3.7 15.3-19 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.3-2
SUMMARY
OF RESULTS FOR LOCKED ROTOR/SHAFT BREAK TRANSIENT FOUR LOOPS OPERATING INITIALLY Maximum Reactor Coolant System Pressure (psia) 2736 Maximum Cladding Temperature (F)
Core Hot Spot 1954 Zr-H2O reaction at core hot spot
(% by weight) 0.54%
15.3-20 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.3-3 THIS TABLE HAS BEEN INTENTIONALLY DELETED.
15.3-21 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.3-3a 15.3-22 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.3-4 INPUT PARAMETERS FOR THE LOCKED ROTOR ACCIDENT RADIOLOGICAL CONSEQUENCE ANALYSIS USING AST Parameter Unit Value Notes Fraction of Core % 2.0 The value of 2% is used, the Experiencing reload limit for safety Cladding Damage with analysis, bounding the 0.1%
Failed-Open PORV value in the previous design basis calculation.
Gap Fractions: (2 times the As a significant number of following): fuel assemblies not
-
I-131 0.08 qualifying for AST due to
-
Kr-85 0.10 their containing fuel rods
-
Other Noble Gases 0.05 with maximum linear heat
-
Other Halogens 0.05 generation rates exceeding
-
Alkali Metals 0.12 6.3 kilowatts per foot peak rod average power for burnups exceeding 54 GWD/MTU, the fuel will be treated as having gap fractions a factor of 2 greater than the Regulatory Guide 1.183 values.
SG Iodine Partition Credited only for non-
-
Factor _ 0.01 faulted SG, with PORV
-
SG Aerosol Carryover _ 0.001 failure assumed in faulted SG Steam Released to Per RG 1.183, release Environment: duration is until cold 0 - 2 hrs lbm 719,000 shutdown is established and 2 - 8 hrs lbm 1,109,000 releases from the steam 8 - 40 hrs lbm 2,664,000 generators have been 40 - 64 hrs lbm 0 terminated.
RHR Cut-in Time hours 40 Time for termination of release due to PORV steaming.
Chemical form of radioiodine released from the fuel:
Cesium iodide % 95 Elemental iodide % 4.85 Organic iodide % 0.15 Iodine Releases from steam generator to environment:
Elemental % 97 Organic % 3 15.3-23 REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES A number of faults have been postulated which could result in reactivity and power distribution anomalies. Reactivity changes could be caused by control rod motion or ejection, boron concentration changes, or addition of cold water to the reactor coolant system (RCS). Power distribution changes could be caused by control rod motion, misalignment, or ejection, or by static means such as fuel assembly mislocation. These events are discussed in this section. Detailed analyses are presented for the most limiting of these events.
Discussion of the following incidents is presented in Section 15.4:
- a. Uncontrolled rod cluster control assembly bank withdrawal from a subcritical or low power startup condition,
- b. Uncontrolled rod cluster control assembly bank withdrawal at power,
- c. Rod cluster control assembly misalignment,
- d. Startup of an inactive reactor coolant pump at an incorrect temperature (detailed analysis is deleted),
- e. Chemical and volume control system malfunction that results in a decrease in boron concentration in the reactor coolant,
- f. Inadvertent loading and operation of a fuel assembly in an improper position, and
- g. Spectrum of rod cluster control assembly ejection accidents.
Items a, b, c, d and e are considered to be ANS Condition II events, Item f and ANS Condition III event, and Item g of ANS Condition IV event. Item c entails both Condition II and III events. Subsection 15.0.1 contains a discussion of ANS classifications.
15.4.1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Subcritical or Low Power Startup Condition 15.4.1.1 Identification of Causes and Accident Description A rod cluster control assembly (RCCA) withdrawal accident is defined as an uncontrolled addition of reactivity to the reactor core caused by withdrawal of RCCAs resulting in a power excursion. Such a transient could be caused by operator action or by a malfunction of the reactor control or rod control systems.
15.4-1 REVISION 5 - DECEMBER 1994
B/B-UFSAR This could occur with the reactor either subcritical, hot zero power or at power. The "at power" case is discussed in Subsection 15.4.2.
Although the reactor is normally brought to power from a subcritical condition by means of RCCA withdrawal, initial startup procedures with a clean core call for boron dilution.
The maximum rate of reactivity increase in the case of boron dilution is less than that assumed in this analysis (see Subsection 15.4.6).
The RCCA drive mechanisms are wired into preselected bank configurations which are not altered during reactor life. These circuits prevent the RCCAs from being automatically withdrawn in other than their respective banks. Power supplied to the banks is controlled such that no more than two banks can be withdrawn at the same time and in their proper withdrawal sequence. The RCCA drive mechanisms are of the magnetic latch type and coil actuation is sequenced to provide variable speed travel. The maximum reactivity insertion rate analyzed in the detailed plant analysis is that occurring with the simultaneous withdrawal of the combination of two sequential control banks having the maximum combined worth at maximum speed.
This event is classified as an ANS Condition II incident (an incident of moderate frequency) as defined in Subsection 15.0.1.
The neutron flux response to a continuous reactivity insertion is characterized by a very fast rise terminated by the reactivity feedback effect of the negative Doppler coefficient. This self limitation of the power excursion is of primary importance since it limits the power to a tolerable level during the delay time for protective action. Should a continuous RCCA withdrawal accident occur, the transient will be terminated by the following automatic features of the reactor protection system:
- a. Source Range High Neutron Flux Reactor Trip - actuated when either of two independent source range channels indicates a neutron flux level above a preselected manually adjustable setpoint. This trip function may be manually bypassed only after an intermediate range flux channel indicates a flux level above a specified level. It is automatically reinstated when both intermediate range channels indicate a flux level below a specified level.
- b. Intermediate Range High Neutron Flux Reactor Trip -
actuated when either of two independent intermediate range channels indicates a flux level above a preselected manually adjustable setpoint. This trip function may be manually bypassed only after two of the four power range channels are reading above approximately 10% of full power and is automatically 15.4-2
B/B-UFSAR reinstated when three of the four channels indicate a power level below this value.
- c. Power Range High Neutron Flux Reactor Trip (Low Setting) - actuated when two out of the four power range channels indicate a power level above approximately 25% of full power. This trip function may be manually bypassed when two of the four power range channels indicate a power level above approximately 10% of full power and is automatically reinstated only after three of the four channels indicate a power level below this value.
- d. Power Range High-Neutron Flux Reactor Trip (High Setting) - actuates when two out of the four power range channels indicate a power level above a preset setpoint. This trip function is always active.
- e. High Nuclear Flux Rate Reactor Trip - actuated when the positive rate of change of neutron flux on two out of four nuclear power range channels indicate a rate above the preset setpoint. This trip function is always active.
In addition, control rod stops on high intermediate range flux level (one of two) and high power range flux level (one out of four) serve to discontinue rod withdrawal and prevent the need to actuate the intermediate range flux level trip and the power range flux level trip, respectively.
15.4.1.2 Analysis of Effects and Consequences Method of Analysis The analysis of the uncontrolled RCCA bank withdrawal from subcritical accident is performed in three stages: first, an average core nuclear power transient calculation, then an average core heat transfer calculation, and finally the DNBR calculation.
The average core nuclear calculation is performed using spatial neutron kinetics methods TWINKLE (Reference 1) to determine the average power generation with time including the various total core feedback effects, i.e., Doppler reactivity and moderator reactivity. The average heat flux and temperature transients are determined by performing a fuel rod transient heat transfer calculation in FACTRAN (Reference 2). The average heat flux is next used in VIPRE (described in Section 4.4) for transient DNBR calculation.
Plant characteristics and initial conditions are discussed in Subsection 15.0.3. In order to give conservative results for a startup accident, the following assumptions are made:
- a. Since the magnitude of the power peak reached during the initial part of the transient for any given rate 15.4-3 REVISION 15 - DECEMBER 2014
B/B-UFSAR of reactivity insertion is strongly dependent on the Doppler coefficient, conservatively low values as a function of power are used. The Doppler reactivity defect is determined as a function of power level using a one-dimensional steady-state computer code with a Doppler weighting factor of 1.0. The Doppler coefficient used does not directly correlate with Figure 15.0-3 because the TWINKLE code, on which the neutronic analysis is based, is a diffusion-theory code, rather than a point-kinetics approximation. The Doppler defect used as an initial condition is 965 pcm.
- b. Contribution of the moderator reactivity coefficient is negligible during the initial part of the transient because the heat transfer time between the fuel and the moderator is much longer than the neutron flux response time. However, after the initial neutron flux peak, the succeeding rate of power increase is affected by the moderator reactivity coefficient. A highly conservative value is used in the analysis to yield the maximum peak heat flux.
- c. The reactor is assumed to be just critical at hot zero power (no load) Tavg. This assumption is more conservative than that of a lower initial system temperature. The higher initial system temperature yields a large fuel-water heat transfer coefficient, larger specific heats, and a less negative (smaller absolute magnitude) Doppler coefficient, all of which tend to reduce the Doppler feedback effect thereby increasing the neutron flux peak. The initial effective multiplication factor is assumed to be 1.0 since this results in the worst nuclear power transient.
- d. Reactor trip is assumed to be initiated by power range high neutron flux (low setting). The most adverse combination of instrument and setpoint errors, as well as delays for trip signal actuation and rod cluster control assembly release, is taken into account. A 10% increase is assumed for the power range flux trip setpoint raising it from the nominal value of 25% to 35%. Since the rise in the neutron flux is so rapid, the effect of errors in the trip setpoint on the actual time at which the rods are released is negligible. In addition, the reactor trip insertion characteristic is based on the assumption that the highest worth rod cluster control assembly is stuck in its fully withdrawn position. See Subsection 15.0.5 for rod cluster control assembly insertion characteristics.
15.4-4 REVISION 9 - DECEMBER 2002
B/B-UFSAR
- e. The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the combination of the two sequential control banks having the greatest combined worth at maximum speed (45 inches/minute). Control rod drive mechanism design is discussed in Section 4.6.
15.4-4a REVISION 6 - DECEMBER 1996
B/B-UFSAR
- f. The most limiting axial and radial power shapes associated with having the two highest combined worth banks in their high worth position are assumed in the DNB analysis.
- g. The initial power level was assumed to be below the power level expected for any shutdown condition (10-9 of nominal power). This combination of highest reactivity insertion rate and lowest initial power produces the highest peak heat flux.
- h. Two reactor coolant pumps are assumed to be in operation. This is conservative with respect to DNB.
Plant systems and equipment credited for mitigating the effects of the accident are discussed in Subsection 15.0.8 and listed in Table 15.0-7. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.
Results Figures 15.4-1 through 15.4-3 show the transient behavior for the uncontrolled RCCA bank withdrawal incident, with the accident terminated by reactor trip at 35% of nominal power. The reactivity insertion rate used is greater than that calculated for the two highest worth sequential control banks, both assumed to be in their highest incremental worth region. Figure 15.4-1 shows the neutron flux transient.
The energy release and the fuel temperature increases are relatively small. The thermal flux response, of interest for DNB considerations, is shown on Figure 15.4-2. The beneficial effect of the inherent thermal lag in the fuel is evidenced by a peak heat flux much less than the full power nominal value. There is a large margin to DNB during the transient since the rod surface heat flux remains below the design value, and there is a high degree of subcooling at all times in the core. Figure 15.4-3 shows the response of the hot spot fuel average temperature and the hot spot clad temperature. The average fuel temperature increases to a value lower than the nominal full power value.
The minimum DNBR at all times remains above the limit value.
The calculated sequence of events for this accident is shown on Table 15.4-1. With the reactor tripped, the plant returns to a stable condition. The plant may subsequently be cooled down further by following normal plant shutdown procedures.
The operating procedures would call for operator action to control RCS boron concentration and pressurizer level using the CVCS, and to maintain steam generator level through control of the main or auxiliary feedwater system. Any action required of the operator to maintain the plant in a stabilized condition will be in a time frame in excess of 10 minutes following reactor trip.
15.4-5 REVISION 11 - DECEMBER 2006
B/B-UFSAR 15.4.1.3 Radiological Consequences There are no radiological consequences associated with an uncontrolled rod cluster control assembly bank withdrawal from a subcritical or low power start-up condition event since radioactivity is contained within the fuel rods and reactor coolant system within design limits.
15.4.1.4 Conclusions In the event of a RCCA withdrawal accident from the subcritical condition, the core and the reactor coolant system are not adversely affected, since the combination of thermal power and the coolant temperature result in a DNBR greater than the limit value. Thus, no fuel or clad damage is predicted as a result of DNB.
15.4.2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power 15.4.2.1 Identification of Causes and Accident Description Uncontrolled rod cluster control assembly (RCCA) bank withdrawal at power results in an increase in the core heat flux. Since the heat extraction from the steam generator lags behind the core power generation until the steam generator pressure reaches the relief or safety valve setpoint, there is a net increase in the reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise could eventually result in DNB. Therefore, in order to avert damage to the fuel clad, the reactor protection system is designed to terminate any such transient before the DNBR falls below the limit value.
This event is classified as an ANS Condition II incident (an incident of moderate frequency) as defined in Subsection 15.0.1.
The automatic features of the reactor protection system which prevent core damage following the postulated accident include the following:
- a. Power range neutron flux instrumentation actuates a reactor trip if two-of-four channels exceed an overpower setpoint.
- b. Reactor trip is actuated if any two-out-of-four channels exceed a positive neutron flux rate setpoint.
- c. Reactor trip is actuated if any two-out-of-four T channels exceed an overtemperature T setpoint. This setpoint is automatically varied with axial power imbalance, coolant temperature and pressure to protect against DNB.
15.4-6 REVISION 9 - DECEMBER 2002
B/B-UFSAR
- d. Reactor trip is actuated if any two-out-of-four T channels exceed an overpower T setpoint. This setpont is automatically varied with coolant average temperature so that the allowable heat generation rate (kw/ft) is not exceeded.
15.4-6a REVISION 9 - DECEMBER 2002
B/B-UFSAR
- e. A high pressurizer pressure reactor trip actuated from any two-out-of four pressure channels which is set at a fixed point.
- f. A high pressurizer water level reactor trip actuated from any two-out-of-three level channels when the reactor power is above approximately 10%
(Permissive-7).
In addition to the above listed reactor trips, there are the following RCCA withdrawal blocks:
- a. high neutron flux (one-out-of-four power range),
- b. Overpower T (two-out-of-four), and
- c. Overtemperature T (two-out-of-four).
The manner in which the combination of overpower and overtemperature T trips provide protection over the full range of reactor coolant system conditions is described in Chapter 7.0.
Figure 15.0-1 presents allowable reactor coolant loop average temperature and T for the design power distribution and flow as a function of primary coolant pressure. The boundaries of operation defined by the overpower T trip and the over temperature T trip are represented as "protection lines" on this diagram. The protection lines are drawn to include all adverse instrumentation and setpoint errors so that under nominal conditions trip would occur well within the area bounded by these lines. The utility of this diagram is in the fact that the limit imposed by a given DNBR can be represented as a line. The DNB lines represent the locus of conditions for which the DNBR equals the limit value. All points below and to the left of a DNB line for a given pressure have a DNBR greater than the limit value.
The diagram shows that DNB is prevented for all cases if the area enclosed with the maximum protection lines is not traversed by the applicable DNBR line at any point.
The area of permissible operation (power, pressure and temperature) is bounded by the combination of reactor trips:
high neutron flux (fixed setpoint); high pressure (fixed setpoint); low pressure (fixed setpoint); overpower and overtemperature T (variable setpoints).
15.4-7 REVISION 11 - DECEMBER 2006
B/B-UFSAR 15.4.2.2 Analysis of Effects and Consequences Method of Analysis The transient is analyzed by the LOFTRAN Code (Reference 3).
This code simulates the neutron kinetics, reactor coolant system, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperature, pressures, and power level. The core limits as illustrated in Figure 15.0-1 are used as input to LOFTRAN to determine the minimum DNBR during the transient.
Bryon/Braidwood Unit 1 have the BWI steam generators while Byron/Braidwood Unit 2 have Westinghouse D5 steam generators.
For this reason, the limiting cases are analyzed separately using input for each steam generator design to insure limiting results are calculated.
This accident is analyzed with the Revised Thermal Design Procedure as described in WCAP-11397-P-A for minimum DNBR. Plant characteristics and initial conditions are discussed in Subsection 15.0.3.
- a. Initial reactor power, pressure, and RCS temperatures are assumed to be at their nominal values. With the exception of the RCS average temperature bias, which is explicitly modeled in the analysis, uncertainties in initial conditions are included in the limit DNBR as described in WCAP-11397-P-A.
- b. Reactivity Coefficients - Two cases are analyzed:
- 1. Minimum Reactivity Feedback. The most positive moderator coefficient of reactivity allowed by the Technical Specifications is assumed corresponding to the beginning of core life. A variable Doppler power coefficient with core power is used in the analysis. A conservatively small (in absolute magnitude) value is assumed.
- 2. Maximum Reactivity Feedback. A conservatively large positive moderator density coefficient and a large (in absolute magnitude) negative Doppler power coefficient are assumed.
- c. The reactor trip on high neutron flux is assumed to be actuated at a conservative value of 118% of nominal full power. The T trips include all adverse instrumentation and setpoint errors; the delays for trip actuation are assumed to be the maximum values.
- d. The RCCA trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.
15.4-8 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- e. The maximum positive reactivity insertion rate is greater than that for the simultaneous withdrawal of the combinations of the two control banks have the maximum combined worth at maximum speed.
- f. The analysis assumes all main steam safety valves are operable. If one or more main steam safety valves become inoperable, Technical Specification 3.7.1 requires a reduction in power as well as a reduction in the power range neutron flux - high reactor trip setpoint. The required reductions are based on a heat balance equation as described in Reference 12. A sensitivity study was performed to demonstrate that the reductions are adequate.
The effect of RCCA movement on the axial core power distribution is accounted for by causing a decrease in overtemperature T trip setpoint proportional to a decrease in margin to DNB.
The peak RCS pressure case for this event is analyzed using the standard thermal design procedure. Initial reactor power, pressure, and RCS temperatures are assumed to be at their nominal values with uncertainties applied in the conservative direction.
No credit is taken for the effect of pressurizer spray and power operated relief valves in reducing or limiting the coolant pressure. Pressurizer safety valves are operable. The Technical Specifications allow a 2% tolerance around the nominal setting of 2460 psig for the pressurizer safety valves, with the lowest allowed lifting pressure being 2411 psig. The assumed high pressurizer pressure reactor trip setpoint in the peak pressure cases, including instrument uncertainty, is higher than the lowest setting for the pressurizer safety valves. As a result, the pressurizer safety valves could lift prior to reaching the high pressurizer pressure reactor trip setpoint and the reactor trip might be delayed.
The pressurizer safety valves are assumed to lift at the highest setpoint in the peak pressure cases to maximize RCS and secondary side pressure. A sensitivity study is performed assuming the lowest pressurizer safety valve lift setpoint to demonstrate that, with a delayed reactor trip, all applicable acceptance criteria are met.
15.4-8a REVISION 11 - DECEMBER 2006
B/B-UFSAR Plant systems and equipment credited for mitigating the effects of the accident are discussed in Subsection 15.0.8 and listed in Table 15.0-7. No single active failure in any of these systems or equipment will adversely offset the consequences of the accident.
Results Figures 15.4-4 and 15.4-5 show the transient response for a rapid RCCA withdrawal incident starting from full power. Reactor trip on high neutron flux occurs shortly after the start of the accident. Since this is rapid with respect to the thermal time constants of the plant, small changes in Tavg and pressure result and margin to the DNBR limit is maintained.
The transient response for a slow RCCA withdrawal from full power is shown in Figures 15.4-6 and 15.4-7. Reactor trip on overtemperature T occurs after a longer period and the rise in temperature and pressure is consequently larger than for rapid RCCA withdrawal. Again, the minimum DNBR is greater than the limit value.
Figure 15.4-8 shows the minimum DNBR as a function of reactivity insertion rate from initial full power operation for minimum and maximum reactivity feedback. It can be seen that two reactor trip channels provide protection over the whole range of reactivity insertion rates. These are the high neutron flux and overtemperature T channels. The minimum DNBR is never less than the limit value.
Figures 15.4-9 and 15.4-10 show the minimum DNBR as a function of reactivity insertion rate for RCCA withdrawal incidents starting at 60% and 10% power, respectively. In neither case does the DNBR fall below the limit value.
The shape of the curves of minimum DNB ratio versus reactivity insertion rate in the reference figures is due both to reactor core and coolant system transient response and to protection system action in initiating a reactor trip.
Referring to Figure 15.4-10, for example, it is noted that:
- a. For high reactivity insertion rates (i.e., between
~ 2.0 x 10-4 K/sec and 11.0 x 10-4 K/sec) reactor trip is initiated by the high neutron flux trip for the minimum reactivity feedback cases. For the higher insertion rates in this range, the neutron flux level in the core rises rapidly while core heat flux and coolant system fluid lag behind due to the thermal capacity of the fuel and coolant system fluid. Thus, the reactor is tripped prior to any significant increase in heat flux or coolant temperature resulting in higher minimum DNBRs for these reactivity insertion rates and more margin to the applicable safety analysis limit.
15.4-9 REVISION 15 - DECEMBER 2014
B/B-UFSAR As reactivity insertion rate decreases, core heat flux and coolant temperatures can remain more nearly in equilibrium with the neutron flux; minimum DNB ratio during the transient thus decreases with decreasing insertion rate.
- b. The overtemperature T reactor trip circuit initiates a reactor trip when measured coolant loop T exceeds a setpoint based on measured reactor coolant system average temperature and pressure. This trip circuit is described in detail in Chapter 7.0; however, it is important in this context to note that the average temperature contribution to the circuit is lead-lag compensated in order to account for the effect of the thermal capacity of the reactor coolant system in response to power increases.
- c. With further decrease in reactivity insertion rate, the overtemperature T and high neutron flux trips become equally effective in terminating the transient (e.g., at ~ 2.0 x 10-4 K/sec reactivity insertion rate).
For reactivity insertion rates between ~ 2.0 x 10-4 K/sec and ~ 5.0 x 10-5 K/sec the effectiveness of the overtemperature T trip increases (in terms of increased minimum DNB ratio) due to the fact that with lower insertion rates the power increase rate is slower, the rate of rise of average coolant temperature is slower and the system lags and delays become less significant.
For reactivity insertion rates less than ~ 5.0 x 10-5 K/sec, the rise in the reactor coolant temperature is sufficiently high so that the setpoint for the first bank of steam generator safety valves is reached before the reactor trip setpoint. This delays reactor trip due to dynamic compensation associated with the overtemperature T trip resulting in lower minimum DNB ratios.
15.4-10 REVISION 15 - DECEMBER 2014
B/B-UFSAR Figures 15.4-8, 15.4-9 and 15.4-10 illustrate minimum DNBR calculated for minimum and maximum reactivity feedback for the BWI steam generators which bound the D5 steam generators.
Since the RCCA withdrawal at power incident is an overpower transient, the fuel temperatures rise during the transient until after reactor trip occurs. For high reactivity insertion rates, the overpower transient is fast with respect to the fuel rod 15.4-10a REVISION 15 - DECEMBER 2014
B/B-UFSAR thermal time constant, and the core heat flux lags behind the neutron flux response. Due to this lag, the peak core heat flux does not exceed 118% of its nominal value (i.e., the high neutron flux trip setpoint assumed in the analysis). Taking into account the effect of the RCCA withdrawal on the axial core power distribution, the peak fuel temperature will still remain below the fuel melting temperature.
For slow reactivity insertion rates, the core heat flux remains more nearly in equilibrium with the neutron flux. The overpower transient is terminated by the overtemperature T reactor trip before a DNB condition is reached. The peak heat flux again is maintained below 118% of its nominal value. Taking into account the effect of the RCCA withdrawal on the axial core power distribution, the peak clad centerline temperature will remain below the fuel melting temperature.
Since DNB does not occur at any time during the RCCA withdrawal at power transient, the ability of the primary coolant to remove heat from the fuel rod is not reduced. Thus, the fuel cladding temperature does not rise significantly above its initial value during the transient.
The calculated sequence of events for this accident is shown on Table 15.4-1. With the reactor tripped, the plant eventually returns to a stable condition. The plant may subsequently be cooled down further by following normal plant shutdown procedures.
The peak RCS pressure for this event is analyzed using the same range of reactivity insertion rates as the DNB cases discussed above. A limit is placed on the maximum reactivity insertion rate during an RWAP event to demonstrate compliance with the TS RCS pressure SL. Subsequently, the plant-specific core design must ensure this maximum reactivity insertion rate will not be exceeded as part of the reload evaluation process. This ensures the RWAP RCS overpressure analysis remains valid on a cycle specific basis.
15.4.2.3 Radiological Consequences There are only minimal radiological consequences associated with an uncontrolled rod cluster control assembly bank withdrawal at power event. The reactor trip causes a turbine trip, and heat is removed from the secondary system through the steam generator power relief valves or safety valves. The radiological consequences associated with atmospheric steam release from this event are less severe than the steamline break accident analyzed in Subsection 15.1.5.
15.4.2.4 Conclusions The high neutron flux and overtemperature T trip channels provide adequate protection over the entire range of possible reactivity insertion rates, i.e., the minimum value of DNBR is always larger than the limit value. The radiological consequences would be less severe than the steamline break accident analyzed in Subsection 15.1.5.
15.4-11 REVISION 13 - DECEMBER 2010
B/B-UFSAR 15.4.3 Rod Cluster Control Assembly Misoperation (System Malfunction or Operator Error) 15.4.3.1 Identification of Causes and Accident Description Rod cluster control assembly (RCCA) misoperation accidents include:
- a. One or more dropped RCCAs within the same group,
- b. A dropped RCCA bank,
- c. Statically misaligned RCCA,
- d. Withdrawal of a single RCCA.
Each RCCA has a position indicator channel which displays position of the assembly. The displays of assembly positions are grouped for the operator's convenience. Fully inserted assemblies are further indicated by a rod at bottom signal, which actuates a local alarm and a control room annunciator. Group demand position is also indicated.
Full length RCCAs are always moved in preselected banks, and the banks are always moved in the same preselected sequence. Each bank of RCCAs is divided into two groups. The rods comprising a group operate in parallel through multiplexing thyristors. The two groups in a bank move sequentially such that the first group is always within one step of the second group in the bank. A definite schedule of actuation (or deactuation of the stationary gripper, movable gripper, and lift coils of a mechanism) is required to withdraw the RCCA that is attached to the mechanism.
Since the stationary gripper, movable gripper, and lift coils associated with the four RCCAs of a rod group are driven in parallel, any single failure which would cause rod withdrawal would affect a minimum of one group. Mechanical failures are in the direction of insertion, or immobility.
The dropped RCCAs, dropped RCCA bank, and statically misaligned RCCA events are classified as ANS Condition II incidents (incidents of moderate frequency) as defined in Subsection 15.0.1. The single RCCA withdrawal incident is classified as an ANS Condition III event, as discussed below.
No single electrical or mechanical failure in the rod control system could cause the accidental withdrawal of a single rod cluster control assembly (RCCA) from the inserted bank at full power operation. The operator could deliberately withdraw a single RCCA in a control bank since this feature is necessary in order to retrieve an assembly should one be accidentally dropped.
The event analyzed must result from multiple wiring failures (probability for single random failure is on the order of 10-4/year refer to Subsection 7.7.2.2) or multiple deliberate operator actions and subsequent and repeated operator disregard 15.4-12 REVISION 9 - DECEMBER 2002
B/B-UFSAR of event indication. The probability of such a combination of conditions is considered so low that the limiting consequences may include slight fuel damage.
Thus, consistent with the philosophy and format of ANSI N18.2, the event is classified as a Condition III event. By definition "Condition III occurrences include incidents, any one of which may occur during the lifetime of a particular plant," and "shall not cause more than a small fraction of fuel elements in the reactor to be damaged..."
This selection of criterion is not in violation of GDC 25 which states, "The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods."
(Emphases have been added.) It has been shown that single failures resulting in RCCA bank withdrawals do not violate specified fuel design limits. Moreover, no single malfunction can result in the withdrawal of a single RCCA. Thus, it is concluded that the criterion established for the single rod withdrawal at power is appropriate and in accordance with GDC 25.
A dropped RCCA or RCCA bank is detected by:
- a. Sudden drop in the core power level as seen by the nuclear instrumentation system;
- b. Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples;
- c. Rod at bottom signal;
- d. Rod deviation alarm; and
- e. Rod position indication.
Misaligned assemblies are detected by:
- a. Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples;
- b. Rod deviation alarm; and
- c. Rod position indicators.
The resolution of the rod position indicator channel is 6 steps at full accuracy and 12 steps at half accuracy. Deviation of any assembly from its group by twice this distance (24 steps) will not cause power distributions worse than the design limits.
The deviation alarm alerts the operator to rod deviation with respect to the group position in excess of 12 steps. If the rod deviation alarm is not operable, procedures require the operator to take appropriate actions.
15.4-13 REVISION 9 - DECEMBER 2002
B/B-UFSAR If one or more rod position indicator channels should be out of service, detailed operating instructions shall be followed to assure the alignment of the non-indicated RCCAs. The operator is also required to take action as required by the Technical Specifications.
In the extremely unlikely event of simultaneous electrical failures which could result in single RCCA withdrawal, rod deviation and rod control urgent failure would both be displayed on the plant annunciator, and the rod position indicators would indicate the relative positions of the RCCAs in the bank. The urgent failure alarm also inhibits automatic rod motion in the group in which it occurs. Withdrawal of a single RCCA by operator action, whether deliberate or by a combination of errors, would result in activation of the same alarm and the same visual indications. Withdrawal of a single RCCA results in both positive reactivity insertion tending to increase core power, and an increase in local power density in the core area associated with the RCCA. Automatic protection for this event is provided by the overtemperature T reactor trip, although due to the increase in local power density it is not possible in all cases to provide assurance that the core safety limits will not be violated.
Plant systems and equipment credited for mitigating the effects of the various control rod misoperations are discussed in Subsection 15.0.8 and listed in Table 15.0-7. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.
15.4.3.2 Analysis of Effects and Consequences
- a. Dropped RCCAs, dropped RCCA bank, and statically misaligned RCCA.
Method of Analysis
- 1. One or more dropped RCCAs from the same group.
For evaluation of the dropped RCCA event, the transient system response is calculated using the LOFTRAN code. The code simulates the neutron kinetics, reactor coolant system, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level. Statepoints are calculated and nuclear models are used to obtain a hot channel factor consistent with the primary system conditions and reactor power. By incorporating the primary conditions from the transient and the hot channel factor from the nuclear analysis, the 15.4-14 REVISION 11 - DECEMBER 2006
B/B-UFSAR DNB design basis is shown to be met using the VIPRE code. The transient response, nuclear peaking factor analysis, and DNB design basis confirmation are performed in accordance with the methodology described in Reference 9.
- 2. Dropped RCCA Bank A LOFTRAN calculation is not necessary for the dropped RCCA event. Westinghouse WCAP 11394-P-A concludes that sufficient DNB margin exists, subject to plant/cycle specific analysis.
- 3. Statically Misaligned RCCA.
Steady state power distributions are analyzed using the computer codes as described in Table 4.1-2. The peaking factors are then compared to the peaking factor limit determined by the VIPRE code at the safety analysis DNBR limit.
Results
- 1. One or more dropped RCCAs.
Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion.
When detected, the reactor is manually tripped following the drop of the RCCAs. The core is not adversely affected during this period, since power is decreasing rapidly. Following manual reactor trip, normal shutdown procedures are followed.
RCCA may be manually retrieved by following approved operating procedures.
For those dropped RCCAs which do not result in a manual reactor trip, power may be reestablished either by reactivity feedback or control bank withdrawal. Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition. The equilibrium process without control system interaction is monotonic, thus removing power overshoot as a concern and establishing the automatic rod control mode of operation as the limiting case.
For a dropped RCCA event in the automatic rod control mode, the rod control system detects the drop in power and initiates control bank withdrawal. Power overshoot may occur due to this action by the automatic rod controller after which the control system will insert the control bank to restore nominal power. Figure 15.4-12a 15.4-15 REVISION 15 - DECEMBER 2014
B/B-UFSAR shows a typical transient response to a dropped RCCA (or RCCAs) in automatic control.
Uncertainties in the initial condition are included in the DNB evaluation as described in Reference 11. In all cases, the minimum DNBR remains above the limit value.
- 2. Dropped RCCA Bank.
A dropped RCCA bank typically results in a reactivity insertion greater than 500 pcm. The reactor is manually tripped following the drop of a RCCA bank. The core is not adversely affected during this period, since power is decreasing rapidly. Following reactor trip, normal shutdown procedures are followed to further cool down the plant. Any action required of the operator to maintain the plant in a stabilized condition will be in a time frame in excess of ten minutes following the incident.
- 3. Statically Misaligned RCCA The most severe misalignment situations with respect to DNBR at significant power levels arise from cases in which one RCAA is fully inserted, or where bank D is fully inserted with one RCCA fully withdrawn. Multiple independent alarms, including a bank insertion limit alarm, alert the operator well before the postulated conditions are approached. The bank can be inserted to its insertion limit with any one assembly fully withdrawn without the DNBR falling below the limit value.
The insertion limits in the Technical Specifications may vary from time to time depending on a number of limiting criteria. It is preferable, therefore, to analyze the misaligned RCCA case at full power for a position of the control bank as deeply inserted as the criteria on minimum DNBR and power peaking factor will allow.
The full power insertion limits on control bank D must then be chosen to be above that position and will usually be dictated by other criteria.
Detailed results will vary from cycle to cycle depending on fuel arrangements.
For the RCCA misalignment in which control bank D is inserted to its full power insertion limit and one RCCA is fully withdrawn, the DNBR does not fall below the limit value. This case was analyzed 15.4-16 REVISION 9 - DECEMBER 2002
B/B-UFSAR assuming the initial reactor power, pressure, and RCS temperatures were at their nominal values, but with the increased radial peaking factor associated with the misaligned RCCA.
Uncertainties in initial conditions were included as described in WCAP-11397-P-A (Reference 11).
DNB calculations have not been performed specifically for assemblies missing from other banks; however, power shape calculations have been done as required for the RCCA ejection analysis. Inspection of the power shapes shows that the DNB and peak kW/ft situation is less severe than the bank D case discussed above assuming insertion limits on the other banks equivalent to a bank D full-in insertion limit.
For the RCCA misalignments with one RCCA fully inserted, the DNBR does not fall below the limit value. This case was analyzed assuming the initial reactor power, pressure, and RCS temperatures are at their nominal values, but with the increased radial peaking factor associated with the misaligned RCCA.
Uncertainties in the initial conditions are included as described in WCAP-11397-P-A.
DNB does not occur for the RCCA misalignment incident and thus the ability of the primary coolant to remove heat from the fuel rod is not reduced. The peak fuel temperature corresponds to a linear heat generation rate based on the radial peaking factor penalty associated with the misaligned RCCA and the design axial power distribution. The resulting linear heat generation is well below that which would cause fuel melting.
Following the identification of a RCCA group misalignment condition by the operator, the operator is required to take action as required by the plant Technical Specifications and operating instructions.
- b. Single RCCA Withdrawal Method of Analysis Power distributions within the core are calculated using the computer codes as described in Table 4.1-2.
The peaking factors are then used by VIPRE to calculate the minimum DNBR for the event. The case of the worst rod withdrawn from control bank D inserted at the insertion limit, with the reactor 15.4-17 REVISION 15 - DECEMBER 2014
B/B-UFSAR initially at full power, was analyzed. This incident is assumed to occur at beginning-of-life since this results in the minimum value of moderator temperature coefficient. This assumption maximizes the power rise and minimizes the tendency of increased moderator temperature to flatten the power distribution.
Results For the single rod withdrawal event, two cases have been considered as follows:
- 1. If the reactor is in the manual control mode, continuous withdrawal of a single RCCA results in both an increase in core power and coolant temperature, and an increase in the local hot channel factor in the area of the withdrawing RCCA. In terms of the overall system response, this case is similar to those presented in Subsection 15.4.2; however, the increased local power peaking in the area of the withdrawn RCCA results in lower minimum DNBRs than for the withdrawn bank cases. Depending on initial bank insertion and location of the withdrawn RCCA, automatic reactor trip may not occur sufficiently fast to prevent the minimum core DNBR from falling below the limit value. Evaluation of this case at the power and coolant conditions at which the overtemperature T trip would be expected to trip the plant shows that an upper limit for the number of rods with a DNBR less than the limit value is 5%.
- 2. If the reactor is in the automatic control mode, the multiple failures that result in the withdrawal of a single RCCA will result in the immobility of the other RCCAs in the controlling bank. The transient will then proceed in the same manner as Case 1 described above.
For such cases as above, a reactor trip will ultimately ensue, although not sufficiently fast in all cases to prevent a minimum DNBR in the core of less than the limit value. Following reactor trip, normal shutdown procedures are followed.
15.4.3.3 Radiological Consequences The most limiting rod cluster control assembly misoperation, accidental withdrawal of a single RCCA, is predicted to result in limited fuel damage (< 5% of the total). The subsequent reactor and turbine trip would result in atmospheric steam release, assuming the condenser was not available for use. The radiological consequences from 15.4-18 REVISION 9 - DECEMBER 2002
B/B-UFSAR this event would be no greater than the locked rotor event, analyzed in Subsection 15.3.3.
15.4.3.4 Conclusions For cases of dropped RCCAs or dropped banks, for which the reactor is manually tripped, there is no reduction in the margin to core thermal limits, and consequently the DNB design basis is met. It is shown for all cases which do not result in reactor trip that the DNBR remains greater than the limit value and, therefore, the DNB design is met.
For all cases of any RCCA fully inserted, or control bank D inserted to its rod insertion limits with any single RCCA in that bank fully withdrawn (static misalignment), the DNBR remains greater than the limit value.
For the case of the accidental withdrawal of a single RCCA, with the reactor in the automatic or manual control mode and initially operating at full power with control bank D at the insertion limit, an upper bound of the number of fuel rods experiencing DNB is 5 percent of the total fuel rods in the core.
15.4.4 Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature The Technical Specifications require that all four reactor coolant pumps be operating for reactor power operation; therefore, operation with an inactive loop is precluded. This event was originally included in the FSAR licensing basis when operation with a loop out of service was considered. Based on the Technical Specifications, which prohibit at-power operation with an inactive loop, and changes to the Technical Specifications that deleted all references to three-loop operation, this event has been deleted from the UFSAR.
15.4-19 REVISION 9 - DECEMBER 2002
B/B-UFSAR Pages 15.4-20 and 15.4-21 have been deleted intentionally.
15.4-20 and 15.4-21 REVISION 5 - DECEMBER 1994
B/B-UFSAR 15.4.5 A Malfunction or Failure of the Flow Controller in a BWR Loop That Results in an Increased Reactor Coolant Flow Rate (Not applicable in PWRs) 15.4.6 Chemical and Volume Control System Malfunction That Results in a Decrease in Boron Concentration in the Reactor Coolant 15.4.6.1 Identification of Causes and Accident Description The principal means of causing an inadvertent boron dilution are the opening of the primary water makeup control valve and failure of the blend system, either by controller or mechanical failure.
The CVCS and RMCS are designed to limit, even under various postulated failure modes, the potential rate of dilution to values which, with indication by alarms and instrumentation, will allow sufficient time for operator response, depending on the mode of operation, to terminate the dilution. An inadvertent dilution from the RMCS may be terminated by closing the primary water makeup control valve, CV-111A. All expected sources of dilution may be terminated by closing isolation valves in the CVCS, LCV-112B and C. The lost shutdown margin (SDM) may be regained by the opening of isolation valves to the RWST, LCV-112D and E, thus allowing the addition of 2300 ppm borated water to the RCS.
15.4-22 REVISION 9 - DECEMBER 2002
B/B-UFSAR It is assumed that the addition rate of unborated water to the RCS is limited to 205 gpm by the capacity of the primary water makeup pumps and two charging pumps for Modes 1 and 2. The addition rate of unborated water to the RCS is assumed to be limited to 168 gpm by the high charging flow alarm in Modes 3, 4, and 5. Flows higher than 168 gpm are not considered in these modes of operation since the VCT will be filled faster and the boron dilution alert alarms will be initiated earlier.
Unborated water sources are isolated from the RCS if conditions prescribed by Technical Specifications 3.3.9 cannot be met.
Generally, to dilute, the operator must perform two distinct actions:
- a. Switch control of the RMCS from the automatic makeup mode to the dilute mode and
- b. Take RMCS control switch to start.
Failure to carry out either of the above actions prevents initiation of dilution. Also during normal operation the operator may add borated water to the RCS by blending boric acid from the boric acid storage tanks with primary grade water. This requires the operator to determine the concentration of the addition and setting the blended flow rate and the boric acid flow rate. The makeup controller will then limit the sum of the boric acid flow rate and primary grade water flow rate to the blended flow rate, i.e., the controller determines the primary grade water flow rate, after the RMCS control switch is taken to start.
The status of the RCS makeup is continuously available to the operator by:
- a. Indication of the boric acid and blended flow rates,
- c. Deviation alarms if the boric acid or blended flow rates deviate by more than 10 percent from the preset values,
- d. Source range neutron flux - when reactor is subcritical;
- 1. high flux at shutdown alarm,
- 2. indicated source range neutron flux count rates,
- 3. audible source range neutron flux count rate, and
- 4. source range neutron flux - doubling alarm.
15.4-23 REVISION 9 - DECEMBER 2002
B/B-UFSAR
- e. With the reactor critical
- 1. Axial flux difference alarm (reactor power 50 percent RTP),
- 2. Control rod insertion limit low and low-low alarms,
- 3. Overtemperature T alarm (at power),
- 4. Overtemperature T turbine runback (at power),
and
- 5. Overtemperature T reactor trip.
- f. Power Range Neutron Flux - High, both high and low setpoint reactor trips.
- g. Boron Dilution Alert alarm(s) when reactor is subcritcal
- 1. VCT high level,
- 2. Divert valve CV112A is not in VCT position, and
- 3. Source Range neutron flux-doubling.
This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Subsection 15.0.1.
15.4.6.2 Analysis of Effects and Consequences To cover all phases of plant operation, boron dilution during refueling, cold shutdown, hot shutdown, hot standby, startup, and power modes of operation are considered in this analysis.
Conservative values for necessary parameters were used, i.e.,
high RCS critical boron concentrations, high boron worths, minimum shutdown margins, and lower than actual RCS volumes.
These assumptions result in conservative determinations of the time available for operator or system response after detection of a dilution transient in progress.
Conservative analysis methods are used to analyze a CVCS malfunction that results in a decrease in boron concentration in the reactor coolant. Minimum reactor coolant volumes and maximum dilution flow rates are conservatively assumed for each case analyzed. The result is a logarithmic decrease in coolant boron concentration according to the equation:
d CB Q
= - in CB dt V where CB = Boron concentration in the RCS 15.4-24 REVISION 9 - DECEMBER 2002
B/B-UFSAR Qin = Maximum dilution flow rate V = Active volume in RCS.
This equation can be solved for the time at which the core would become critical or all shutdown margin would be lost. The rate of reactivity insertion due to the dilution can be calculated from the dilution rate and the differential boron worth. The results of this analysis are conservative for all cases analyzed.
A comprehensive review of the primary system has been completed.
This review showed that a single failure in the CVCS system in the cold shutdown condition would not result in a boron dilution of the reactor coolant system and that the CVCS malfunction represents the most limiting potential source of dilution. Based on this review, it is clear that the analysis results presented in the UFSAR bound all potential sources of inadvertent dilution under all modes of operation.
Dilution During Refueling An uncontrolled boron dilution transient cannot occur during this mode of operation. Inadvertent dilution is prevented by administrative controls which isolate the RCS from the potential source of unborated water. CVCS valves, specified in Technical Specification 3.9.2 Bases will be verified closed and secured in position by mechanical stops or by removal of air or electrical power. These valves block all automatic flow paths that could allow unborated makeup water to reach the RCS. Any makeup which is required during refueling will be borated water supplied from the RWST. Administrative controls will strictly control and limit the volume of unborated water which can be manually added to the refueling pool, such as for decontamination activities, in order to prevent diluting the refueling cavity or canal below the required refueling boron concentrations.
Dilution During Cold Shutdown During this mode, the reactor is shutdown and meets the minimum shutdown margin described by the Core Operating Limits Report (COLR) per Technical Specification 3.1.1. The following conditions are assumed for inadvertent boron dilution while in this operating mode:
a) The minimum ratio of initial to critical boron concentration required to ensure the core does not reach criticality in this mode is 1.065 for Byron Unit 1 and Braidwood Unit 1 and 1.075 for Byron Unit 2 and Braidwood Unit 2. The initial boron concentrations used for these ratios must meet the shutdown margin requirements as specified in the COLR.
b) Dilution flow is limited to a maximum dilution rate of 168 gpm, which is equivalent to the high charging flow alarm plus uncertainties. Any additional dilution flow would increase the VCT level, without an immediate decrease in RCS boron concentration.
15.4-25 REVISION 16 - DECEMBER 2016
B/B-UFSAR c) The minimum RCS water volume of 10583 ft3 for Byron Unit 1 and Braidwood Unit 1 and 9260 ft3 for Byron Unit 2 and Braidwood Unit 2 is used. This is a conservative estimate of the active volume of the RCS with one RCP running and all four loop stop valves open.
Dilution During Hot Shutdown During this mode, the reactor is shutdown and meets the minimum shutdown margin described by the Core Operating Limits Report (COLR) per Technical Specification 3.1.1. The following conditions are assumed for inadvertent boron dilution while in this operating mode:
a) The minimum ratio of initial to critical boron concentration required to ensure the core does not reach criticality in this mode is 1.071 for Byron Unit 1 and Braidwood Unit 1 and 1.082 for Byron Unit 2 and Braidwood Unit 2. The initial boron concentrations used for these ratios must meet the shutdown margin requirements as specified in the COLR.
b) Dilution flow is limited to a maximum dilution rate of 168 gpm, which is equivalent to the high charging flow alarm plus uncertainties. Any additional dilution flow would increase the VCT level, without an immediate decrease in RCS boron concentration.
c) The minimum RCS water volume of 10583 ft3 for Byron Unit 1 and Braidwood Unit 1 and 9260 ft3 for Byron Unit 2 and Braidwood Unit 2 is used. This is a conservative estimate of the active volume of the RCS with one RCP running and all four-loop stop valves open.
Dilution During Hot Standby During the mode, the reactor is shutdown and meets the minimum shutdown margin described by the Core Operating Limits Report (COLR) per Technical Specification 3.1.1. The following conditions are assumed for inadvertent boron dilution while in this operating mode:
a) The minimum ratio of initial to critical boron concentration required to ensure the core does not reach criticality in this mode is 1.088 for Byron Unit 1 and Braidwood Unit 1 and 1.101 for Byron Unit 2 and Braidwood Unit 2. The initial boron concentrations used for these ratios must meet the shutdown margin requirements as specified in the COLR.
15.4-26 REVISION 13 - DECEMBER 2010
B/B-UFSAR b) Dilution flow is limited to a maximum dilution rate of 168 gpm, which is equivalent to the high charging flow alarm plus uncertainties. Any additional dilution flow would increase the VCT level, without an immediate decrease in RCS boron concentration.
c) The minimum RCS water volume of 10583 ft3 for Byron Unit 1 and Braidwood Unit 1 and 9260 ft3 for Byron Unit 2 and Braidwood Unit 2 is used. This is a conservative estimate of the active volume of the RCS with one RCP running and all four-loop stop valves open.
Dilution During Startup Startup is a transitory mode of operation. In this mode the plant is being taken from one long-term mode of operation, hot standby, to another, power operation. The plant is maintained in the startup mode only for the purpose of startup testing at the beginning of each cycle.
During this mode of operation the plant is in manual control, i.e., Tavg/rod control is in manual. All normal actions required to change power level, either up or down, require operator initiation. The Technical Specifications and TRM require a SDM as specified in the Core Operating Limits Report and four reactor coolant pumps operating. Other conditions assumed are:
- a. Dilution flow is limited by the reactor makeup water system. The maximum anticipated flow rate of unborated primary water to the RCS is 205 gpm.
- b. A minimum RCS water volume of 9941.9 ft3 for Byron Units 1 and 2 and Braidwood Units 1 and 2. This is a conservative estimate of the active RCS volume, minus the pressurizer volume.
- c. The CB for criticality (ARI-1) is very conservatively assumed with a very conservative, constant boron worth.
15.4-26a REVISION 13 - DECEMBER 2010
B/B-UFSAR Dilution During Full Power Operation The plant may be operated at power two ways, automatic Tavg/rod control and under operator control. The TRM requires an available trip reactivity as specified in the Core Operating Limiting Report. Technical Specifications require operation with four reactor coolant pumps. With the plant at power and the RCS at pressure, the dilution rate is limited by the capacity of the reactor makeup water system with two centrifugal charging pumps in operation (analysis is performed assuming two charging pumps are in operation even though normal operation is with one pump).
Conditions assumed for this mode are:
- a. For manual and automatic reactor control at full power conditions, the dilution flow is limited by the reactor makeup water system (205 gpm).
- b. A minimum RCS water volume of 9941.9 ft3 for Byron Units 1 and 2 and Braidwood Units 1 and 2. This is a conservative estimate of the active RCS volume, minus the pressurizer volume.
- c. The CB for criticality (ARI-1) is very conservatively assumed with a very conservative, constant boron worth.
15.4.6.3 Results and Conclusions Dilution During Refueling Dilution during this mode has been precluded through administrative control of valves in the possible dilution flow paths, see Subsection 15.4.6.2.
Dilution During Cold Shutdown In this mode of operation, unborated water sources will be isolated from the RCS if the conditions prescribed by Technical Specifications 3.3.9 cannot be met. In the event of an inadvertent boron dilution transient while in this mode of operation, the boron dilution alert alarms will sound upon detection of VCT high level, and the operator will administratively align CVCS valves to terminate dilution and start boration (note that this alarm could also be initiated by the flux doubling signal or CV112A valve not in VCT position, but these signals are not credited in the analysis). Valves LCV-112D and E (isolation valves to the RWST) are opened to supply 2300 ppm borated water to the suction of the charging pumps and valves LCV-112B and C (isolation valves in the CVCS) are closed to terminate the dilution. These actions are carried out to minimize the approach to criticality and regain lost shutdown margin. The operator has at least 15 minutes to complete these actions from the time of the alarms until shutdown margin is lost.
15.4-27 REVISION 13 - DECEMBER 2010
B/B-UFSAR Dilution During Hot Shutdown and Hot Standby In this mode of operation, unborated water sources will be isolated from the RCS if the conditions prescribed by Technical Specifications 3.3.9 cannot be met. In the event of an inadvertent boron dilution transient while in this mode of operation, the boron dilution alert alarms will sound upon detection of VCT high level, and the operator will administratively align CVCS valves to terminate dilution and start boration (note that this alarm could also be initiated by the flux doubling signal or CV112A valve not in VCT position, but these signals are not credited in the analysis). Valves LCV-112D and E (isolation valves to the RWST) are opened to supply 2300 ppm borated water to the suction of the charging pumps and valves LCV-112B and C (isolation valves in the CVCS) are closed to terminate the dilution. These actions are carried out to minimize the approach to criticality and regain lost shutdown margin. The operator has at least 15 minutes to complete these actions from the time of the alarm(s) until shutdown margin is lost.
Dilution During Startup This mode of operation is a transitory mode to go to power and is the operational mode in which the operator intentionally dilutes and withdraws control rods to take the plant critical. During this mode the plant is in manual control with the operator required to maintain a very high awareness of the plant status.
For a normal approach to criticality the operator must manually initiate a limited dilution and subsequently manually withdraw the control rods, a process that takes several hours. The plant Technical Specifications require that the operator determine the estimated critical position of the control rods prior to approaching criticality thus assuring that the reactor does not go critical with the control rods below the insertion limits.
Once critical, the power escalation must be sufficiently slow to allow the operator to manually block the source range reactor trip (nominally at 105 cps) after receiving P-6 from the intermediate range. Too fast a power escalation (due to an unknown dilution) would result in reaching P-6 unexpectedly leaving insufficient time to manually block the source range reactor trip. Failure to perform this manual action results in a reactor trip and immediate shutdown of the reactor.
However, in the event of an unplanned approach to criticality or dilution during power escalation while in the startup mode, the plant status is such that minimal impact will result. The plant will slowly escalate in power to a reactor trip on the power range neutron flux - high, low setpoint (nominally 25 percent RTP). After reactor trip, there are more than 15 minutes for operator action prior to return to criticality. The required operator action is to initiate and continue boration until adequate shutdown margin is restored and to terminate the dilution.
15.4-28 REVISION 12 - DECEMBER 2008
B/B-UFSAR Dilution During Full Power Operation With the reactor in manual control and no operator action taken to terminate the transient, the power and temperature rise will cause the reactor to reach the overtemperature T or Power Range High Neutron Flux trip setpoint resulting in a reactor trip.
After reactor trip, there are more than 15 minutes for operator action prior to return to criticality. The required operator action is to initiate and continue boration until adequate shutdown margin is restored and to terminate the dilution. The boron dilution transient in this case is essentially the 15.4-28a REVISION 9 - DECEMBER 2002
B/B-UFSAR equivalent to an uncontrolled rod withdrawal at power. The maximum reactivity insertion rate for a boron dilution transient is conservatively estimated to be 2.0 pcm/sec and is within the range of insertion rates analyzed for uncontrolled rod withdrawal at power. It should be noted that prior to reaching the overtemperature T reactor trip, the operator will have received an alarm on overtemperature T turbine runback.
Thus with the reactor in automatic rod control, a boron dilution will result in a power and temperature increase such that the rod controller will attempt to compensate by slow insertion of the control rods. This action by the controller will result in at least one of three alarms to the operator:
- a. rod insertion limit - low level alarm,
- b. rod insertion limit - low-low level alarm if insertion continued after (1), and
- c. axial flux difference alarm (I outside of the target band).
Given the many alarms, indications, and the inherent slow process of dilution at power, the operator has sufficient time for action. For example, the operator has more than 15 minutes from the rod insertion limit low-low alarm until l.3 percent K/K is inserted at beginning-of-life. The time would be significantly longer at end-of-life, due to the low initial boron concentration, when shutdown margin is a concern.
The above results demonstrate that in all modes of operation an inadvertent boron dilution is precluded, or sufficient time is available for operator action to terminate the transient.
Following termination of the dilution flow and initiation of boration the reactor is in a stable condition with the operator regaining the required shutdown margin.
15.4.7 Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position 15.4.7.1 Acceptance Criteria This event is identified as a Condition III event (Infrequent Fault) as defined in Subsection 15.0.1. The specific acceptance criteria for this event are as follows:
- a. To meet the requirements of General Design Criteria 13, plant operating procedures should include a provision requiring that reactor instrumentation be used to search for potential fuel loading errors after fueling operations.
15.4-29 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- b. In the event the error is not detectable by the instrumentation system and fuel rod failure limits could be exceeded during normal operations, the offsite consequences should be a small fraction of the 10 CFR Part 100 guidelines.
15.4.7.2 Identification of Causes and Accident Description The Inadvertent Loading Event comprises core misloading scenarios such as the loading of one or more fuel assemblies into improper positions, loading a fuel rod during manufacture with one or more pellets of the wrong enrichment, or the loading of a full fuel assembly during manufacture with pellets of the wrong enrichment.
In addition to these scenarios, misloading events involving burnable absorbers are theoretically possible, scenarios such as the placement of a cluster of 20 burnable absorbers into a core location slated to have 24 burnable absorbers. All of these misloading scenarios potentially result in a core reactivity distribution that differs from the intended core reactivity distribution. As a result, the core power distribution and peaking factors may differ from predictions. Specifically, misloading errors can lead to increased local power peaking at the location of the misloading if the misloading results in a local reactivity increase relative to the intended pattern. If the misloading results in a local reactivity decrease, power peaking increases away from the location of the misloading are possible due to unintended power tilts. These kinds of increases, however, are generally distributed over a large core volume and are small relative to those where the local reactivity is increased.
Fuel misloads are prevented by the manufacturing controls employed to build the fuel and the core loading controls used to assemble the core. The manufacturing controls include checks on fuel rod weight to confirm fuel enrichments, pellet stack lengths, pellet types, and the absence of pellet gaps during fuel manufacturing, and bar coding of each fuel rod to confirm its proper placement in the fuel assembly.
To reduce the probability of core loading errors during fuel loading, each fuel assembly and core component is marked with an identification number and loaded in accordance with a core loading diagram. During core loading, the identification numbers are checked before each assembly is moved into the core.
Identification numbers read during fuel movement are subsequently recorded on the loading diagram as a further check on proper placement after loading is completed. These procedures make the likelihood of core misloadings very small.
15.4-30 REVISION 15 - DECEMBER 2014
B/B-UFSAR The severity and detectability of fuel misloads are influenced by several factors: the local reactivity perturbation relative to the intended core loading pattern, the core position of the misload, the local environment of the misloaded fuel assembly, and the number of operable incore detector locations and their proximity to the misload location. Should misloadings occur, the incore system of movable flux detectors, which is used to verify power distributions during startup and throughout the operating cycle, is capable of revealing enrichment errors or misloadings which would cause the kind of substantial power distribution perturbation that would be necessary to induce large numbers of fuel rod failures. In addition, thermocouples and excore detectors can provide additional indications of power distribution anomalies. The instrumentation, along with the startup testing performed each cycle, make the detection of severe misloadings highly likely.
15.4.7.3 Evaluation The incore moveable detector system is used to search for potential fuel misloads at the start of each operating cycle.
Following fuel loading and low power physics testing, an initial core power distribution measurement is made. The core power level of this initial flux map is typically between ~30% and ~50%
of rated thermal power. This initial power distribution measurement is used to confirm that the measured power distribution is consistent with the predicted power distribution.
Observed flux map deviations in excess of the flux map review criteria (See Table 15.4-6) would prompt an investigation of a possible core anomaly. This satisfies the first acceptance criterion given above.
In Reference 13, a larger number of misloads were evaluated for representative core designs employing current fuel types and fuel failures. The simulated misloads, involving one or two fuel assemblies, covered a wide range of local reactivity perturbations and core positions. The resulting hot full power (HFP) FH peaking factors ranged from benign to severe. Severe misloads with peaking factors that exceed the FH limit for DNB at normal operation conditions have the potential for fuel failure if they remain undetected. The simulated misloads were assessed with respect to severity and detectability.
The detectability assessments of Reference 13 demonstrated that the incore detector system is very robust with respect to detection of misloads severe enough to fail fuel during normal operation. By examining a large number of moveable detector thimble patterns, the detectability assessments considered the effect of inoperable moveable detector thimbles on misload detectability. Even when the minimum number of operable detector locations allowed per the plant licensing bases was assumed, the incore detector system was capable of reliably detecting misloads severe enough to fail fuel during normal operation.
Fuel misloads involving a single fuel rod or fuel pellet were not evaluated as part of Reference 13. Such misloads, in general, 15.4-31 REVISION 15 - DECEMBER 2014
B/B-UFSAR will not be detectable using the incore detector system due to the very small power distribution perturbation. In terms of increased peaking factors and reduced DNBR values, however, the consequences of such misloads will be very small and limited to the affected fuel rod and the immediately adjacent fuel rods.
Detection of fuel misloads is, in part, a function of the number of available incore detector locations. Reference 13 demonstrated that the flux map review criteria of Table 15.4-6 are effective in detecting fuel misloads that could lead to fuel failures during normal operation. To enhance the probability that significant misloads will be detected, tighter review criteria are employed when the number of available detector locations is reduced. These review criteria will be used for startup and subsequent at-power flux maps.
The detectability assessments of Reference 13 confirm that the moveable detector system can reliably detect fuel misloads that could fail fuel during normal operation when the Table 15.4-6 review criteria are employed. Specifically, Reference 13 demonstrated that only a small fraction of 1% of misloads severe enough to fail fuel during normal operation would be undetected at startup using these limited review criteria. Furthermore, it was judged that even these undetected misloads would very likely be detected if other attributes of the startup power distribution measurement (e.g., tilts and reaction rate error contours) were considered along with the results of low power physics testing. Given that detection of >99% of misloads severe enough to fail fuel is expected using these review criteria, a radiological consequences analysis is deemed unnecessary.
Failures in fresh fuel during startup would have negligible radiological consequences since there is only a small fission product inventory. Following startup, any fuel rod failures would occur gradually and would be detected by coolant activity monitoring. Since the number of potential fuel rod failures due to a core misload would be extremely small and such failures would occur gradually, any coolant activity releases would initially be well within the cleanup capacity of the plant. Any trend in increased coolant activity would warrant further investigation and evaluation. Therefore, the second acceptance criterion for this event would be satisfied since failures would be gradual, detectable, and the operations would be maintained within Technical Specification coolant activity guidance.
15.4.7.4 Conclusions Fuel misloads are prevented by manufacturing controls and core loading controls. In the unlikely event that a fuel misload should occur, the incore moveable detector system is capable of reliably detecting misloads that could fail fuel at normal operation conditions. Exceeding the review criteria herein would initiate an investigation to identify potential core anomalies.
Any failures associated with an undetected fuel misload would be gradual, detectable, and the operations would be maintained within Technical Specification coolant activity guidelines.
15.4-31a REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.4.8 Spectrum of Rod Cluster Control Assembly Ejection Accidents 15.4.8.1 Identification of Causes and Accident Description This accident is defined as the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of a rod cluster control assembly (RCCA) and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.
15.4.8.1.1 Design Precautions and Protection Certain features are intended to preclude the possibility of rod ejection accident, or to limit the consequences if the accident were to occur. These include a sound, conservative mechanical design of the rod housings, together with a thorough quality control (testing) program during assembly, and a nuclear design which lessens the potential ejection worth of RCCAs, and minimizes the number of assemblies inserted at high power levels.
Mechanical Design The mechanical design is discussed in Section 4.6. Mechanical design and quality control procedures intended to preclude the possibility of a RCCA drive mechanism housing failure are listed below:
- a. Each full length control rod drive mechanism housing is completely assembled and shop tested at 4100 psi.
- b. The mechanism housings are individually hydrotested after they are attached to the head adapters in the reactor vessel head, and checked during the hydrostatic test of the completed reactor coolant system.
- c. Stress levels in the mechanism are not affected by anticipated system transients at power, or by the thermal movement of the coolant loops. Moments induced by the design-basis earthquake can be accepted within the allowable primary working stress range specified by the ASME Code,Section III, for Class 1 components.
15.4-32
B/B-UFSAR
- d. The latch mechanism housings are each a single length of forged or cast Type-304 stainless steel. The rod travel housings are each a single length of forged Type-304 stainless steel. This material exhibits excellent notch toughness at all temperatures that will be encountered.
A significant margin of strength in the elastic range together with the large energy absorption capability in the plastic range gives additional assurance that gross failure of the housing will not occur. The joints between the latch mechanism housing and head adapter, and between the latch mechanism housing and rod travel housing, are threaded joints reinforced by canopy type welds. Administrative regulations require periodic inspections of these (and other) welds.
Nuclear Design Even if a rupture of a RCCA drive mechanism housing is postulated, the operation of a plant utilizing chemical shim is such that the severity of an ejected RCCA is inherently limited.
In general, the reactor is operated with the RCCAs inserted only far enough to permit load follow. Reactivity changes caused by core depletion and xenon transients are compensated by boron changes. Further, the location and grouping of control RCCA banks are selected during the nuclear design to lessen the severity of a RCCA ejection accident. Therefore, should a RCCA be ejected from its normal position during full power operation, only a minor reactivity excursion, at worst, could be expected to occur.
However, it may be occasionally desirable to operate with larger than normal insertions. For this reason, a rod insertion limit is defined as a function of power level. Operation with the RCCAs above this limit guarantees adequate shutdown capability and acceptable power distribution. The position of all RCCAs is continuously indicated in the control room. An alarm will occur if a bank of RCCAs approaches its insertion limit or if one RCCA deviates from its bank. Operating instructions may require boration, as necessary, at the low level alarm. The low-low alarm alerts the operator to stop diluting if in progress, and verify Shutdown Margin is within the limits specified in the COLR or initiate boration to restore Shutdown Margin to within limit.
Reactor Protection The reactor protection in the event of a rod ejection accident has been described in Reference 5. The protection for this accident is provided by high neutron flux trip (high and low setting) and high positive rate neutron flux trip. These protection functions are described in detail in Section 7.2.
15.4-33 REVISION 17 - DECEMBER 2018
B/B-UFSAR Effects on Adjacent Housings Disregarding the remote possibility of the occurrence of a RCCA mechanism housing failure, investigations have shown that failure of a housing due to either longitudinal or circumferential cracking would not cause damage to adjacent housings. However, even if damage is postulated, it would not be expected to lead to 15.4-33a REVISION 8 - DECEMBER 2000
B/B-UFSAR a more severe transient since RCCAs are inserted in the core in symmetric patterns, and control rods immediately adjacent to worst ejected rods are not in the core when the reactor is critical. Damage to an adjacent housing could, at worst, cause that RCCA not to fall on receiving a trip signal; however, this is already taken into account in the analysis by assuming a stuck rod adjacent to the ejected rod.
15.4.8.1.2 Limiting Criteria This event is classified as an ANS Condition IV incident. See Subsection 15.0.1 for a discussion of ANS classifications. Due to the extremely low probability of a RCCA ejection accident, some fuel damage could be considered an acceptable consequence.
Comprehensive studies of the threshold of fuel failure and of the threshold or significant conversion of the fuel thermal energy to mechanical energy, have been carried out as part of the SPERT project by the Idaho Nuclear Corporation (Reference 6).
Extensive tests of UO2 zirconium clad fuel rods representative of those in pressurized water reactor type cores have demonstrated failure thresholds in the range of 240 to 257 cal/gm. However, other rods of a slightly different design have exhibited failures as low as 225 cal/gm. These results differ significantly from the TREAT (Reference 7) results, which indicated a failure threshold of 280 cal/gm. Limited results have indicated that this threshold decreases by about 10% with fuel burnup. The clad failure mechanism appears to be melting for zero burnup rods and brittle fracture for irradiated rods. Also important is the conversion ratio of thermal to mechanical energy. This ratio becomes marginally detectable above 300 cal/gm for unirradiated rods and 200 cal/gm for irradiated rods; catastrophic failure, (large fuel dispersal, large pressure rise) event for irradiated rods, did not occur below 300 cal/gm.
In view of the above experimental results, criteria are applied to ensure that there is little or no possibility of fuel dispersal in the coolant, gross lattice distortion, or severe shock waves. These criteria are:
- a. Average fuel pellet enthalpy at the hot spot below 225 cal/gm for unirradiated fuel and 200 cal/gm for irradiated fuel.
- b. Peak reactor coolant pressure less than that which could cause stresses to exceed the faulted condition stress limits.
- c. Fuel melting will be limited to less than ten percent of the fuel volume at the hot spot even if the 15.4-34 REVISION 6 - DECEMBER 1996
B/B-UFSAR average fuel pellet enthalpy is below the limits of criterion 1 above.
It should be noted that the FSAR included an additional criterion that the average clad temperature at the hot spot must remain below 2,700F. The elimination of this criterion as a basis for evaluating the RCCA ejection accident results is consistent with the revised Westinghouse acceptance criteria for this event (Reference 10).
15.4.8.2 Analysis of Effects and Consequences Method of Analysis The calculation of the RCCA ejection transient is performed in two stages, first an average core channel calculation and then a hot region calculation. The average core calculation is performed using spatial neutron kinetics methods to determine the average power generation with time including the various total core feedback effects, i.e., Doppler reactivity and moderator reactivity. Enthalpy and temperature transients in the hot spot are then determined by multiplying the average core energy generation by the hot channel factor and performing a fuel rod transient heat transfer calculation. The power distribution calculated without feedback is pessimistically assumed to persist throughout the transient.
A detailed discussion of the method of analysis can be found in Reference 5.
Average Core Analysis The spatial kinetics computer code, TWINKLE (Reference 1), is used for the average core transient analysis. This code solves the two group neutron diffusion theory kinetic equation in one, two or three spatial dimensions (rectangular coordinates) for six delayed neutron groups and, up to 2000 spatial points. The computer code includes a detailed multiregion, transient fuel-clad-coolant heat transfer model for calculation of pointwise Doppler and moderator feedback effects. In this analysis, the code is used as a one dimensional axial kinetics code since it allows a more realistic representation of the spatial effects of axial moderator feedback and RCCA movement.
However, since the radial dimension is missing, it is still necessary to employ very conservative methods (described in the following) of calculating the ejected rod worth and hot channel factor. Further description of TWINKLE appears in Subsection 15.0.11.
Hot Spot Analysis In the hot spot analysis, the initial heat flux is equal to the nominal times the design hot channel factor. During the transient, the heat flux hot channel factor is linearly increased 15.4-35 REVISION 6 - DECEMBER 1996
B/B-UFSAR to the transient value in 0.1 second, the time for full ejection of the rod. Therefore, the assumption is made that the hot spots before and after ejection are coincident. This is very conservative since the peak after ejection will occur in or adjacent to the assembly with the ejected rod, and prior to ejection the power in this region will necessarily be depressed.
15.4-35a REVISION 6 - DECEMBER 1996
B/B-UFSAR The hot spot analysis is performed using the detailed fuel-and cladding transient heat transfer computer code, FACTRAN (Reference 2). This computer code calculates the transient temperature distribution in a cross section of a metal clad UO2 fuel rod, and the heat flux at the surface of the rod, using as input the nuclear power versus time and the local coolant conditions. The zirconium-water reaction is explicitly represented, and all material properties are represented as functions of temperature. A conservative pellet radial power distribution is used within the fuel rod.
FACTRAN uses the Dittus-Boelter or Jens-Lottes correlation to determine the film heat transfer before DNB, and the Bishop-Sandburg-Tong correlation (Reference 8) to determine the film boiling coefficient after DNB. The BST correlation is conservatively used assuming zero bulk fluid quality. The DNB ratio is not calculated, instead the code is forced into DNB by specifying a conservative DNB heat flux. The gap heat transfer coefficient can be calculated by the code; however, it is adjusted in order to force the full power steady-state temperature distribution to agree with the fuel heat transfer design codes. Further description of FACTRAN appears in Subsection 15.0.11.
System Overpressure Analysis Because safety limits for fuel damage specified earlier are not exceeded, there is little likelihood of fuel dispersal into the coolant. The pressure surge may therefore be calculated on the basis of conventional heat transfer from the fuel and prompt heat generation in the coolant.
The pressure surge is calculated by first performing the fuel heat transfer calculation to determine the average and hot spot heat flux versus time. Using this heat flux data, a THINC (Section 4.4) calculation is conducted to determine the volume surge. Finally, the volume surge is simulated in a plant transient computer code. This code calculates the pressure transient taking into account fluid transport in the reactor coolant system and heat transfer to the steam generators. No credit is taken for the possible pressure reduction caused by the assumed failure of the control rod pressure housing.
15.4.8.2.1 Calculation of Basic Parameters Input parameters for the analysis are conservatively selected on the basis of values calculated for this type of core. The more important parameters are discussed below. Table 15.4-3 presents the parameters used in this analysis.
Ejected Rod Worths and Hot Channel Factors The values for ejected rod worths and hot channel factors are calculated using either three dimensional static methods or by a 15.4-36
B/B-UFSAR synthesis method employing one dimensional and two dimensional calculations. Standard nuclear design codes are used in the analysis. No credit is taken for the flux flattening effects of reactivity feedback. The calculation is performed for the maximum allowed bank insertion at a given power level, as determined by the rod insertion limits. Adverse xenon distributions are considered in the calculation.
Appropriate margins are added to the ejected rod worth and hot channel factors to account for any calculational uncertainties, including an allowance for nuclear power peaking due to densification.
Power distributions before and after ejection for a "worst case" can be found in Reference 5. During plant startup physics testing, ejected rod worths and power distributions are measured in the zero and full power rodded configurations and compared to values used in the analysis. It has been found that the ejected rod worth and power peaking factors are consistently overpredicted in the analysis.
Reactivity Feedback Weighting Factors The largest temperature rises, and hence the largest reactivity feedbacks occur in channels where the power is higher than average. Since the weight of a region is dependent on flux, these regions have high weights. This means that the reactivity feedback is larger than that indicated by a simple channel analysis. Physics calculations have been carried out for temperature changes with a flat temperature distribution, and with a large number of axial and radial temperature distributions. Reactivity changes were compared and effective weighting factors determined. These weighting factors take the form of multipliers which when applied to single channel feedbacks correct them to effective whole core feedbacks for the appropriate flux shape. In this analysis, since a one dimensional (axial) spatial kinetics method is employed, axial weighting is not necessary if the initial condition is made to match the ejected rod configuration. In addition, no weighting is applied to the moderator feedback. A conservative radial weighting factor is applied to the transient fuel temperature to obtain an effective fuel temperature as a function of time accounting for the missing spatial dimension. These weighting factors have also been shown to be conservative compared to three dimensional analysis (Reference 5).
Moderator and Doppler Coefficient The critical boron concentrations at the beginning of life and end of life are adjusted in the nuclear code in order to obtain moderator density coefficient curves which are conservative compared to actual design conditions for the plant. As discussed above, no weighting factor is applied to these results.
15.4-37 REVISION 1 - DECEMBER 1989
B/B-UFSAR The Doppler reactivity defect is determined as function of power level using a one dimensional steady-state computer code with a Doppler weighting factor of 1.0. The Doppler defect used is 0.965% for the BOL cases and 0.893% for the EOL cases.
The Doppler weighting factor will increase under accident conditions, as discussed above.
Delayed Neutron Fraction, Calculations of the effective delayed neutron fraction (eff) typically yield values no less than 0.70% at beginning of life and 0.50% at end of life for the first cycle. The accident is sensitive to if the ejected rod worth is equal to or greater than as in zero power transients. In order to allow for future cycles, pessimistic estimates of of 0.55% at beginning of cycle and 0.44% at end of cycle were used in the analysis.
Trip Reactivity Insertion The trip reactivity insertion assumed is given in Table 15.4-3 and includes the effect of one stuck RCCA. These values are reduced by the ejected rod reactivity. The shutdown reactivity was simulated by dropping a rod of the required worth into the core. The start of rod motion occurred 0.5 seconds after the high neutron flux trip point was reached. This delay is assumed to consist of 0.2 seconds for the instrument channel to produce a signal, 0.15 seconds for the trip breaker to open and 0.15 seconds for the coil to release the rods. A curve of trip rod insertion versus time was used which assumed that insertion to the dashpot does not occur until 2.7 seconds after the start of fall. The choice of such a conservative insertion rate means that there is over one second after the trip point is reached before significant shutdown reactivity is inserted into the core.
This is a particularly important conservatism for hot full power accidents.
The minimum design shutdown available for this plant at HZP may be reached only at end of life in the equilibrium cycle. This value includes an allowance for the worst stuck rod, adverse xenon distribution conservative Doppler and moderator defects, and an allowance for calculational uncertainties. Physics calculations for this plant have shown that the effect of two stuck RCCAs (one of which is the worst ejected rod) is to reduce the shutdown by about an additional one percent k. Therefore, following a reactor trip resulting from an RCCA ejection accident, the reactor will be subcritical when the core returns to HZP.
Depressurization calculations have been performed for a typical four-loop plant assuming the maximum possible size break (2.75 inch diameter) located in the reactor pressure vessel head. The results show a rapid pressure drop and a decrease in system water mass due to the break. The safety injection system is actuated by the low pressurizer pressure trip within one minute after the break. The reactor coolant pressure continues to drop and 15.4-38 REVISION 6 - DECEMBER 1996
B/B-UFSAR reaches saturation (1100 to 1300 psi depending on the system temperature) in about two to three minutes. Due to the large thermal inertia of primary and secondary system, there has been no significant decrease in the reactor coolant system temperature below no-load by this time, and the depressurization itself has caused an increase in shutdown margin by about 0.2% k due to the pressure coefficient. The cooldown transient could not absorb the available shutdown margin until more than ten minutes after the break. The addition of high borated (2300 ppm) safety injection flow starting one minute after the break is much more than sufficient to ensure that the core remains subcritical during the cooldown.
Reactor Protection As discussed in Subsection 15.4.8.1.1, reactor protection for a rod ejection is provided by high neutron flux trip (high and low setting) and high positive rate neutron flux trip. These protection functions are part of the reactor trip system. No single failure of the reactor trip system will negate the protection functions required for the rod ejection accident, or adversely affect the consequences of the accident.
15.4.8.2.2 Results Cases are presented for both beginning and end of life at zero and full power.
- 1. Beginning of Cycle, Full Power Control bank D was assumed to be inserted to its insertion limit. The worst ejected rod worth and hot channel factor were conservatively calculated to be 0.2% k and 6.10, respectively. The peak hot spot clad average temperature was 2434F. The peak hot spot fuel center temperature reached melting, conservatively assumed at 4900F.
However, melting was restricted to less than 10%
of the pellet.
- 2. Beginning-of-Cycle, Zero Power For this condition, control bank D was assumed to be fully inserted and banks B and C were at their insertion limits. The worst ejected rod is located in control bank D and has a worth of 0.765% k and a hot channel factor of 11.5. The peak hot spot clad average temperature reached 2348F, the fuel center temperature was 3616F.
- 3. End of Cycle, Full Power Control bank D was assumed to be inserted to its insertion limit. The ejected rod worth and hot 15.4-39 REVISION 9 - DECEMBER 2002
B/B-UFSAR channel factors were conservatively calculated to be 0.25% k and 6.40, respectively. This resulted in a peak clad average temperature of 2235F. The peak hot spot fuel temperature reached melting conservatively assumed at 4800F. However, melting was restricted to less than 10% of the pellet.
- 4. End of Cycle, Zero Power The ejected rod worth and hot channel factor for this case were obtained assuming control bank D to be fully inserted and banks C and B at its insertion limit. The results were 0.8% k and 23.0, respectively. The peak clad average and fuel center temperatures were 2337 and 3479F.
The Doppler weighting factor for this case is significantly higher than for the other cases due to the very large transient hot channel factor.
A summary of the cases presented above is given in Table 15.4-3.
The nuclear power and hot spot fuel and clad temperature transients for the worst cases are presented in Figures 15.4-18 through 15.4-21. (Beginning of life full power and end of life full power.)
The calculated sequence of events for the worst case rod ejection accidents, as shown in Figures 15.4-18 through 15.4-21, is presented in Table 15.4-1. For all cases, reactor trip occurs very early in the transient, after which the nuclear power excursion is terminated. As discussed previously in Subsection 15.4.8.2.2, the reactor will remain subcritical following reactor trip.
The ejection of an RCCA constitutes a break in the reactor coolant system, located in the reactor pressure vessel head. The effects and consequences of loss of coolant accidents are discussed in Subsection 15.6.5. Following the RCCA ejection, the operator would follow the same emergency instructions as for any other loss of coolant accident to recover from the event.
Fission Product Release It is assumed that fission products are released from the gaps of all rods entering DNB. In all cases considered, less than 10% of the rods entered DNB based on a detailed three dimensional THINC analysis (Reference 5).
Pressure Surge A detailed calculation of the pressure surge for an ejection worth of one dollar at beginning of life, hot full power, indicates that the peak pressure does not exceed that which would cause stress to exceed the faulted condition stress limits (Reference 5). Since the severity of the present analysis does 15.4-40 REVISION 9 - DECEMBER 2002
B/B-UFSAR not exceed the "worst case" analysis, the accident for this plant will not result in an excessive pressure rise or further damage to the reactor coolant system.
Lattice Deformations A large temperature gradient will exist in the region of the hot spot. Since the fuel rods are free to move in the vertical direction, differential expansion between separate rods cannot produce distortion. However, the temperature gradients across individual rods may produce a differential expansion tending to bow the midpoint of the rods toward the hotter side of the rod.
Calculations have indicated that this bowing would result in a negative reactivity effect at the hot spot since Westinghouse cores are under-moderated, and bowing will tend to increase the under-moderation at the hot spot. Since the 17 x 17 fuel design is also under-moderated, the same effect would be observed. In practice, no significant bowing is anticipated, since the structural rigidity of the core is more than sufficient to withstand the forces produced. Boiling in the hot spot region would produce a net flow away from that region. However, the heat from the fuel is released to the water relatively slowly, and it is considered inconceivable that cross flow will be sufficient to produce significant lattice forces. Even if massive and rapid boiling, sufficient to distort the lattice, is hypothetically postulated, the large void fraction in the hot spot region would produce a reduction in the total core moderator to fuel ratio, and a large reduction in this ratio at the hot spot. The net effect would therefore be a negative feedback. It can be concluded that no conceivable mechanism exists for a net positive feedback resulting from lattice deformation. In fact, a small negative feedback may result. The effect is conservatively ignored in the analysis.
15.4.8.3 Radiological Consequences of a Postulated Rod Ejection Accident (AST)
The key inputs and assumptions used in the AST Control Rod Ejection Accident (CREA) radiological consequence analysis are summarized below and provided in Table 15.4-4.
Two cases are considered when analyzing the radioactive release due to a CREA.
Case 1: Containment Leakage For Case 1, the ejected control rod is assumed to breach the reactor pressure vessel (RPV), effectively causing the equivalent of a small break loss of coolant accident. In this case, all activity from damaged fuel that has been mixed with the primary coolant of the reactor coolant system (RCS) leaks directly to the containment volume. This flashed release is assumed to instantaneously and homogeneously mix with the containment atmosphere and subsequently be available for release to the environment via a containment leak rate limit, or La, conservatively assumed to be 0.2% per day for this accident analysis. In accordance with RG 1.183 guidance, the leak rate 15.4-41 REVISION 12 - DECEMBER 2008
B/B-UFSAR is reduced by 50% after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, based on the containment pressure decreasing over time.
Case 2: Steam Generator PORV Release For Case 2, no breach of the RPV is assumed following the rod ejection. In this case, reactor coolant system (RCS) integrity is maintained and all activity from damaged fuel that has been mixed with the RCS leaks to the secondary side through the steam generator (SG) tubes at a conservative rate of 1.0 gpm total leakage. From here, activity is available for release to the environment by steaming of the SG power operated relief valves (PORVs). An average rate of release is assumed. In addition to the activity released from the primary to secondary coolant, iodine activity in the secondary coolant at the TS limit (i.e.,
0.1 Ci/gm Dose Equivalent (DE) I-131) is also assumed to be released.
Fuel Damage and Core Source Term The CREA core source terms are those associated with a DBA power level of 3658.3 MWt. This power level bounds the MUR power uprate Rated Thermal Power level including measurement uncertainties.
For the radiological dose analysis, the sudden rod ejection and localized temperature spike associated with the CREA is assumed to damage of 10% of the core fuel rods. Only 2.5 % of the damaged core fuel rods release melted fuel activity (i.e., 0.25%
of the total core melts). Therefore for both cases, the source term available for release is associated with this fraction of melted fuel and the fraction of core activity existing in the gap.
The damaged fuel is assumed to have operated at a radial peaking factor of 1.7.
Activity Release Fractions Release fractions and transport fractions conform to RG 1.183, Appendix H and Table 3. To conform with this regulatory guidance, 10% of the core inventory of iodine and noble gas is assumed to be in the fuel-clad gap. Additionally, Table 3 of Regulatory Guide 1.183 shows that 12% of the core cesium and rubidium should be assumed to be in the fuel-clad gap and should be released in its entirety from the damaged 10% of the total core. However, to account for fuel burnup in excess of the referenced assumptions, the cesium and rubidium release fraction is doubled. Although analyses have shown that isotopic activity fractions in the fuel-clad gap may in fact decrease when "burning" the fuel longer than the 54 GWd/MTU specified in Regulatory Guide 1.183, this 100% increase in the gap fractions is used as an accepted and conservative means of bounding all extended burnup phenomena. With regard to the fraction released from melted fuel, it is assumed that 90% of the core inventory of iodine and noble gas, and 76% of the core cesium and rubidium 15.4-41a REVISION 15 - DECEMBER 2014
B/B-UFSAR remain available for release due to melting (i.e., these are the remaining fractions of activity that are not in the fuel-clad gap). Again, in conformance with RG 1.183, it would be assumed that 100% of the noble gases, 25% of the iodines, and 50% of the cesium and rubidium (i.e., considered particulate/aerosol nuclides) released from the melted fuel would be available for release from containment. However, for this analysis the assumption of 25% of the iodines being available for release was increased to 50%. This was done to prevent a "double counting" of the iodine removal due to plate-out in containment, because this analysis credits Powers' Natural Deposition model for plateout, as opposed to the historically assumed 50% plateout.
These activity release fractions are input to the RADTRAD code through the use of the Release Fractions and Timing (RFT) file.
Airborne Activity Removal Mechanisms in Containment As discussed below only natural deposition, decay, and leakage are credited.
Natural Deposition:
The RADTRAD computer program, including the Powers Natural Deposition algorithm based on NUREG/CR-6189, is used for modeling aerosol deposition in containment. No natural deposition is assumed for elemental or organic iodine. The RADTRAD lower bound level (i.e., 10 percent) of deposition credit is used.
Decay Credited:
Decay of radioactivity is credited in all compartments, prior to release. This is implemented in RADTRAD using the half-lives in the Nuclide Inventory File (NIF). The RADTRAD decay plus daughter option is used. In reality, daughter products such as xenon from iodines or iodines from tellurium are unlikely to readily escape from the matrix in which the parent iodine or tellurium is contained. Nevertheless, the RADTRAD feature to include daughter effects is selected for conservatism.
Steaming Release Rates and Partition Factors:
Activity that originates in the RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate.
This assumed leak rate value is a total of 1.0 gpm. For input into RADTRAD, this rate is converted from gallons per minute to cubic feet per minute, making it equal to 0.1337 cfm.
For Case 2, the release to the environment is associated with the secondary coolant steaming from the SGs. Because of this release dynamic, RG 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water. For iodine, the partition factor of 0.01 was taken directly from the suggested guidance. However, there is no explicit guidance with 15.4-41b REVISION 12 - DECEMBER 2008
B/B-UFSAR regard to other particulate nuclides. Reviewing the specified AST release fractions, it is concluded that the only nuclides other than iodines to be released from the core source term are cesiums, rubidium, and noble gases. For cesiums and rubidium, a partition factor of 0.0055 is used which bounds the value of 0.00529 shown in ANSI Standard, ANS/ANSI 18.1 - 1999, for Cs-134 which has the largest partition factor of these isotopes. Because of their volatility, 100%
of the noble gases are assumed to be released.
The methodology used to model steaming of activity through PORVs following the postulated CREA event of Case 2 assumes an average cumulative release rate through the SG PORVs that, for simplicity and conservatism, is reduced in steps. The partition factors discussed above are applied to these release rates, which were derived from the total time increment mass releases. The following table shows the time steps, isotopic partition factors, and associated release rates; conversion of this data from mass to volumetric flow rates was performed based on cooled liquid conditions (i.e., 62.4 lbm/ft3), as specified by RG 1.183.
CREA Partition Factors and Release Rates Total Steam Steam Steam Steam Time Mass Iodine Cesium Noble Gas Release Rate Release Rate Release Rate Interval Release Partition Partition Partition for Iodines for Cesiums for Noble (hrs) (lbm) Factor Factor Factor (cfm) (cfm) Gases (cfm) 0 - 0.0556 600,000 0.01 0.0055 1 2.8833E+01 1.5858E+01 2.8833E+03 0.0556 - 1,900,000 0.01 0.0055 1 4.8055E+00 2.6430E+00 4.8055E+02 1.1111 1.1111 - 720 0 0.01 0.0055 1 0 0 0
/Q Calculations (Meteorology)
Releases from the SG PORVs are considered elevated releases due to the high steaming rates, and the associated /Qs have been reduced by a factor of five per guidance in RG 1.194 for energetic releases from steam relief valves, if (1) the release point is uncapped and vertically oriented, and (2) the time-dependent vertical velocity exceeds the 95th-percentile wind speed, at the release point height.
Byron Station and Braidwood Station meet these criteria.
15.4-41c REVISION 12 - DECEMBER 2008
B/B-UFSAR Assumptions and Inputs The following inputs and assumptions were used in the CREA analysis.
- a. Core inventory is based on a DBA power level of 3658.3 MWt.
This power level bounds the MUR power uprate Rated Thermal Power level including measurement uncertainties.
- b. 10% of the fuel is damaged during the initiation of this accident, and is assumed to have failed.
- c. The damaged fuel is assumed to have operated at a radial peaking factor of 1.7.
- d. 10% of the core inventory of noble gases and iodines are released from the fuel gap (RG 1.183, Appendix H). Release fractions of other nuclide groups contained in the fuel gap are detailed in Table 3 of RG 1.183, and to account for gap fraction uncertainty. Due to fuel burnup, fractions from the referenced table are doubled.
- e. 2.5% of the damaged fuel rods will experience melting during the CREA.
- f. 100% of noble gases and 50% of the iodines contained in the melted fuel fraction are assumed to be released to the reactor coolant in accordance with Appendix H of RG 1.183. Fractions of other nuclides released from the melted fuel are used from Table 2 of RG 1.183. Though these are described as LOCA values for fuel melt release, they are conservatively used for the other nuclide groups.
- g. The activity released from the fuel from either the gap or from fuel pellets is assumed to be instantaneously mixed with the reactor coolant within the pressure vessel.
- h. All iodine released from the SGs is assumed to be of the elemental species. This is done for RADTRAD simulation considerations, and is consistent with the RG 1.183 specification of 97% elemental and 3% organic, because elemental and organic iodine are identically treated by the computer model.
- i. The CR ventilation system is assumed to realign to the emergency mode of operation 30 minutes after the initiation of this design basis accident.
- j. For the containment leakage case, all leakage is assumed to be at an La of 0.2% per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and 0.1% per day thereafter.
15.4-41d REVISION 15 - DECEMBER 2014
B/B-UFSAR Dose Results Radiological doses resulting from a design basis CREA for a CR operator and a person located at the EAB or LPZ are to be less than the regulatory dose limits as given in the Table below.
Regulatory Dose Limits - CREA Dose Type Control Room (rem) EAB and LPZ (rem)
TEDE Dose 5a 6.3b Notes:
a 10 CFR 50.67 b
10 CFR 50.67 as modified by Regulatory Guide 1.183 (Table 6, Page 1.183-20)
The table below provides the results from the simulations modeled using the RADTRAD code.
Control Rod Ejection Accident Radiological Analysis Results (Maximum of Either Byron or Braidwood)
Case 1: Containment Leakage CREA Dose Assessment Results Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 4.538 5.358 2.278 Case 2: Steam Generator PORV Release CREA Dose Assessment Results Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 0.424 1.707 0.304 For the cases analyzed in this calculation, it is shown that a Case 1 CREA that breaches the RPV, and causes a containment release, would be the bounding CREA scenario. All doses are below the Regulatory Dose Limits, so it is verified that this design basis Control Rod Ejection Accident is sufficiently mitigated at both Byron and Braidwood Stations.
15.4-41e REVISION 16 - DECEMBER 2016
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.4-42 REVISION 12 - DECEMBER 2008
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.4-43 REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.4.8.4 Conclusions Conservative analyses indicate that the described fuel and cladding limits are not exceeded. It is concluded that there is no danger of sudden fuel dispersal into the coolant. Since the peak pressure does not exceed that which would cause stresses to exceed the faulted condition stress limits, it is concluded that there is no danger of further consequential damage to the reactor coolant system. The analyses have demonstrated that the number of fuel rods entering DNB is limited to less than 10% of the fuel rods in the core. The radiological consequences of this event, based on 10% of the fuel rods being damaged, are well within the dose limits of 10 CFR Part 50.67.
15.4.9 References
- 1. Risher, D. H., Jr. and Barry, R. F., "TWINKLE - A Multi-Dimensional Neutron Kinetics Computer Code," WCAP-7979-A (Proprietary) and WCAP-8028-A (Non-Proprietary), January 1975.
- 2. Hargrove, H. G., "FACTRAN - A Fortran-IV Code for Thermal Transients in a UO2 Fuel Rod," WCAP-7908-A, December 1989.
- 3. Burnett, T. W. T., et al., "LOFTRAN Code Description,"
WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary),
April 1984.
- 4. NUREG-0800, Standard Review Plan Section 15.4.8, Appendix A, Radiological Consequences of a Control Rod Ejection Accident (PWR), Revision 1, July 1981.
15.4-44 REVISION 12 - DECEMBER 2008
B/B-UFSAR
- 5. Risher, D. H., Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1-A, January 1975.
- 6. Taxelius, T. G. (Ed), "Annual Report - SPERT Project, October 1968, September 1969," Idaho Nuclear Corporation IN-1370, June 1970.
- 7. Liimataninen, R. C. and Testa, F. J., "Studies in TREAT of Zircaloy-2-Clad, UO2-Core Simulated Fuel Elements" ANL-7225, January - June 1966, p. 177, November 1966.
- 8. Bishop, A. A., Sandberg, R. O., and Tong, L. S., "Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31, August 1965.
- 9. Hassler, R. L., et al, Methodology for the Analysis of the Dropped Rod Event, WCAP-11394-P-A (Proprietary), Revision 1, January 1990, and WCAP-11395 (Non-Proprietary), April 1987.
- 10. Letter from W. J. Johnson of Westinghouse to R. C. Jones of the NRC, "Use of 2700F PCT Acceptance Limit in Non-LOCA Accidents," NS-NRC-89-3466, October 1989.
- 11. Friedland, A.J., and Ray, S., Revised Thermal Design Procedure, WCAP-11397-P-A (Proprietary), WCAP-11397-A (Non-Proprietary), April 1989.
- 12. NRC Information Notice 94-60, Potential Overpressurization of the Main Steam System, August 22, 1994.
- 13. Ankney, R. D. and Grover, J. L., Analysis Update for the Inadvertent Loading Event, WCAP-16676-NP, March 2009.
15.4-45 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.4-1 TIME SEQUENCE OF EVENTS FOR INCIDENT WHICH CAUSE REACTIVITY AND POWER DISTRIBUTION ANOMALIES TIME ACCIDENT EVENT (sec.)
Uncontrolled Rod Initiation of uncontrolled 0.0 Cluster Control rod withdrawal from 10-9 of Assembly Bank nominal power Withdrawal from a Subcritical or Low Power Startup Condition 1 Power range high neutron flux low setpoint reached 11.1 2 Peak nuclear power occurs 11.3 3 Rod begin to fall into core 11.6 4 Peak heat flux occurs 13.6 5 Minimum DNBR occurs 13.6 6 Peak average clad temperature occurs 13.9 7 Peak average fuel temperature occurs 14.0 Uncontrolled RCCA bank withdrawal at power
- 1. Case A Initiation of uncontrolled RCCA 0 withdrawal at a high reactivity insertion rate (80 pcm/sec)
Power range high neutron flux 1.6 high trip point reached Rods begin to fall into core 2.1 Minimum DNBR occurs 3.2 15.4-46 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.4-1 (Cont'd)
- 2. Case B Initiation of uncontrolled RCCA 0 withdrawal at a small reactivity insertion rate (3.0 pcm/sec)
Overtemperature T trip signal 35.7 initiated Rods begin to fall into core 43.7 Minimum DNBR occurs 44.3 Rod Cluster Control Assembly Ejection
- 1. Beginning-of- Initiation of rod ejection 0.0 Life, Full Power Power range high neutron flux 0.05 setpoint reached Peak nuclear power occurs 0.13 Rods begin to fall into core 0.55 Peak fuel average temperature 2.28 occurs Peak clad temperature occurs 2.38 Peak heat flux occurs 2.39 15.4-47 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.4-1 (Cont'd)
- 2. End-of-Life, Initiation of rod ejection 0.0 Full Power Power range high neutron flux 0.04 setpoint reached Peak nuclear power occurs 0.13 Rods begin to fall into core 0.54 Peak fuel average temperature 2.31 occurs Peak clad temperature occurs 2.40 Peak heat flux occurs 2.41 15.4-48 REVISION 9 - DECEMBER 2002
B/B-UFSAR Table 15.4-2 has been deleted.
15.4-49
B/B-UFSAR TABLE 15.4-3 PARAMETERS USED IN THE ANALYSIS OF THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT BOL-HFP BOL-HZP EOL-HFP EOL-HZP TIME IN LIFE BEGINNING BEGINNING END END Power Level, % 100 (including 0 100 (including 0 uncertainties) uncertainties)
Ejected rod worth,%WK 0.2 0.765 0.25 0.8 Delayed neutron fraction, % 0.55 0.55 0.44 0.44 Feedback reactivity weighting 1.30 2.07 1.30 3.55 Trip reactivity,%WK 4.0 2.0 4.0 2.0 Fq before rod ejection 2.60 2.60 Fq after rod ejection 6.10 11.5 6.40 23.0 Number of operational pumps 4 2 4 2 15.4-50 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.4-3 (Cont'd)
BOL-HFP BOL-HZP EOL-HFP EOL-HZP TIME IN LIFE BEGINNING BEGINNING END END Max. fuel pellet average 4128 3123 4044 3056 temperature, F Max. fuel center temperature, F >4900 3616 >4800 3479 Max. clad average temperature, F 2434 2348 2369 2337 Max. fuel store energy, cal/gm 181 130 177 127
% Fuel Melt <10% 0 <10% 0 15.4-51 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.4-4 INPUT PARAMETERS FOR THE CREA RADIOLOGICAL CONSEQUENCES ANALYSIS USING AST Parameter Unit Value Notes Portion of Core Fuel % 10 Bounds value predicted by safety Rods Experiencing analysis. The acceptance criteria for Cladding Damage fuel melting is a maximum of 10%.
Melted fuel % of 0.25 Based on 50% of the rods that violate core the DNB limit having melting in the inner 10% over 50% of the axial length.
Gap Activity As a significant number of fuel assemblies not qualifying for AST due Alkali Metals % 0.24 to their containing fuel rods with maximum linear heat generation rates exceeding 6.3 kilowatts per foot peak rod average power for burnups exceeding 54 GWD/MTU, the fuel will be treated as having gap fractions a factor of 2 greater than the RG 1.183 values.
Fraction of activity released to containment:
From gap inventory Iodine - 1.0 Noble Gases - 1.0 From melted fuel Iodine - 0.5 Noble gas - 1.0 Iodine plateout onto - 0.5 containment surfaces Iodine Chemical Species .
in Containment:
Aerosol % 95 Elemental % 4.85 Organic % 0.15 Iodine Chemical Species in Release from SG to Environment:
Elemental % 97 Organic % 3 15.4-52 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.4-4 (Contd)
INPUT PARAMETERS FOR THE CREA RADIOLOGICAL CONSEQUENCES ANALYSIS USING AST Parameter Unit Value Notes Iodine Removal N/A See Typically, no credit is taken for Coefficients in Notes continuing iodine removal in the Containment containment for the rod ejection accident, however under provisions allowed by the AST governing RG 1.183, Power's model for particulate deposition removal may be credited, if 50% plateout is not credited.
Containment Leak Rate: The design basis containment leak 0-24 Hours weight 0.20 rate at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is consistent with
> 24 Hours %/day 0.10 guidance of RG 1.183.
Fraction of activity released to primary coolant (for primary to secondary leakage pathway):
From gap inventory Iodine - 1.0 Noble Gases - 1.0 From melted fuel Iodine - 0.5 Noble gas - 1.0 Iodine Chemical Species in Primary Coolant:
Elemental % 100 Organic % 0 Particulate % 0 Steam Released to Bounds release predicted by small Environment: break LOCA analysis.
0 - 200 sec lb/sec 3000 200 - 4000 sec lb/sec 500 Chemical form of radioiodine released to the containment atmosphere:
Cesium iodine % 95 Elemental iodine % 4.85 Organic iodide % 0.15 15.4-53 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.4-4 (Contd)
ASSUMPTIONS USED FOR THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT THIS PAGE HAS BEEN INTENTIALLY DELETED 15.4-54 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.4-5 This table has been intentionally deleted.
15.4-55 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.4-6 Flux Map Review Criteria Measured vs. Predicted Number of Available Detector Symmetric Thimble Reaction Detector Reaction Rate Locations Rate Comparison**
Comparison*
55 to 58 10% 7%
49 to 54 8% 5%
44 to 48 6% 5%
- The review criterion is the table value (%) or an absolute normalized reaction rate difference equal to the table value divided by 100% (e.g., 10% / 100% = 0.1), whichever is greater.
- Applicable to symmetric thimbles with normalized reaction rates above 0.7. The review criterion is relative to the expected reaction rate difference 15.4-56 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.5 INCREASE IN REACTOR COOLANT INVENTORY Discussion and analysis of the following events is presented in this section:
- a. inadvertent operation of emergency core cooling system during power operation,
- b. chemical and volume control system malfunction that increases reactor coolant inventory, and
- c. a number of BWR transients (not applicable to the Byron/Braidwood Stations).
These events, considered to be ANS Condition II, cause an increase in reactor coolant inventory. Subsection 15.0.1 contains a discussion of ANS classifications.
15.5.1 Inadvertent Operation of Emergency Core Cooling System During Power Operation 15.5.1.1 Identification of Causes and Accident Description Inadvertent operation of the emergency core cooling system (ECCS) at power could be caused by operator error, test sequence error, or a false electrical actuation signal. A spurious signal initiated after the logic circuitry in one solid-state protection system train for any of the following engineered safety feature (ESF) functions could cause this incident by actuating the ESF equipment associated with the affected train.
- a. High containment pressure,
- b. Low pressurizer pressure, or
- c. Low steamline pressure.
Following the actuation signal, the suction of the coolant charging pumps diverts from the volume control tank to the refueling water storage tank. Simultaneously, the valves isolating the charging pumps from the injection header automatically open and the normal charging line isolation valves close. The charging pumps force the borated water from the refueling water storage tank (RWST) through the pump discharge header, the injection line, and into the cold leg of each loop.
The passive accumulator tank safety injection and low head system are available. However, they do not provide flow when the reactor coolant system (RCS) is at normal pressure.
A safety injection (SI) signal normally results in a direct reactor trip and a turbine trip. However, any single fault that actuates the ECCS will not necessarily produce a reactor trip.
15.5-1 REVISION 7 - DECEMBER 1998
B/B-UFSAR If an SI signal generates a reactor trip, the operator should determine if the signal is spurious. If the SI signal is determined to be spurious, the operator should terminate SI and maintain the plant in the hot-standby condition as determined by appropriate recovery procedures. If repair of the ESF actuation system instrumentation is necessary, future plant operation will be in accordance with the Technical Specifications. If the SI results in discharge of coolant through the pressurizer safety valves, the operators will bring the plant to cold shutdown in order to inspect the valves.
If the reactor protection system does not produce an immediate trip as a result of the spurious SI signal, the reactor experiences a negative reactivity excursion due to the injected boron, which causes a decrease in reactor power. The power mismatch causes a drop in Tavg and consequent coolant shrinkage.
The pressurizer pressure and water level decrease. Load decreases due to the effect of reduced steam pressure on load after the turbine throttle valve is fully open. If automatic rod control is used, these effects will lessen until the rods have moved out of the core. The transient is eventually terminated by the reactor protection system low pressurizer pressure trip or by manual trip.
The time to trip is affected by initial operating conditions.
These initial conditions include the core burnup history which affects initial boron concentration, rate of change of boron concentration, and Doppler and moderator coefficients.
15.5.1.2 Analysis of Effects and Consequences Method of Analysis Inadvertent operation of the ECCS is analyzed using the LOFTRAN computer code (Reference 1). The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, the feedwater system, the steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.
Inadvertent operation of the ECCS at power is classified as a Condition II event, a fault of moderate frequency. The criteria established for Condition II events include the following.
- a. Pressure in the reactor coolant and main steam systems should be maintained below 110% of the design values,
- b. Fuel cladding integrity shall be maintained by ensuring that the minimum departure from nucleate boiling ratio (DNBR) remains above the DNBR limit, derived at a 95% confidence level and 95% probability, and
- c. An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently.
15.5-2 REVISION 15 - DECEMBER 2014
B/B-UFSAR The inadvertent ECCS actuation at power event is analyzed to determine the maximum RCS pressure encountered throughout the accident. The most limiting case with respect to RCS pressure is an SI at Hot Full Power coincident with a reactor trip. Because of the pressure reduction from the reactor trip, the SI flow is maximized. The SI flow refills the pressurizer until the pressurizer is water solid, and the SI flow results in liquid discharge through the pressurizer safety valves.
The performance of the pressurizer safety valve system and the loads on pressurizer safety valves, associated piping, and supports as a result of liquid discharge through the pressurizer safety valves, was determined to be acceptable (References 4 and 5).
The Inadvertent Operation of the ECCS During Power Operation event does not progress into a stuck open Pressurizer Safety Valve LOCA event. All three valves may lift in response to the event, but they will reclose. The resulting leakage from up to three pressurizer safety valves that are seated is bounded by flow through one fully open valve. The consequences of the event are bounded by the analysis described in UFSAR Section 15.6.1, Inadvertent opening of Pressurizer Safety or Relief Valve (References 6 and 7). This event is also classified as an event of moderate frequency.
American Nuclear Society standard 51.1/N18.2-1973 (Reference 2) describes example 15 of a condition II event as a "minor reactor coolant system leak which would not prevent orderly reactor shutdown and cooldown assuming makeup is provided by normal makeup systems only." In Reference 2, normal makeup systems are defined as those systems normally used to maintain reactor coolant inventory under respective conditions of startup, hot standby, power operation, or cooldown, using onsite power. Since the cause of the water relief is the ECCS flow, the magnitude of the leak will be less than or equivalent to that of the ECCS (i.e., operation of the ECCS maintains RCS inventory during the postulated event and establishes the magnitude of the subject leak). Therefore, the above example of a Condition II event is met.
The inadvertent ECCS actuation at power event is also analyzed to determine the minimum DNBR value.
15.5-3 REVISION 15 - DECEMBER 2014
B/B-UFSAR The most limiting case is a minimum reactivity feedback condition with the plant assumed to be in manual rod control. Because of the power and temperature reduction during the transient, operating conditions do not approach the core limits.
The minimum DNBR was obtained at time zero for both units. The Unit 1 specific results are presented here. However, they are representative of the results for both Unit 1 and Unit 2.
The analysis assumptions for the DNBR evaluation are as follows:
- a. Initial Operating Conditions The event is analyzed with the revised thermal design procedure as described in WCAP-11397-P-A (Reference 3). Initial reactor power, RCS pressure and temperature are assumed to be at the nominal full power values. With the exception of the RCS average temperature bias, which is explicitly modeled in the analysis, uncertainties in initial conditions are included in the limit DNBR as described in Reference 3.
- b. Moderator and Doppler Coefficients of Reactivity The analysis assumes a zero moderator temperature coefficient and a low absolute value Doppler power coefficient at beginning of life.
- c. Reactor Control The reactor is assumed to be in manual rod control.
15.5-4 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- d. Pressurizer Pressure Control Pressurizer heaters are assumed to be inoperable.
This assumption yields a higher rate of pressure decrease which is conservative. Pressurizer spray and PORVs are assumed available in order to minimize RCS pressure.
- e. Boron Injection At the initiation of the event, two charging pumps inject borated water into the cold leg of each loop.
The analysis assumes zero injection line purge volume for calculational simplicity; thus, the boration transient begins immediately in the analysis. The positive displacement charging pump is assumed to be inoperable at event initiation.
- f. Turbine Load The turbine load remains constant until the governor drives the throttle valve wide open. After the throttle valve is full open, turbine load decreases as steam pressure drops.
- g. Reactor Trip Reactor trip is initiated by a low pressurizer pressure signal at 1860 psia.
- h. Decay Heat The decay heat has no impact on the DNB case (i.e.,
minimum DNBR occurs prior to reactor trip). A conservative core residual heat generation based upon long-term operation at the initial power level is assumed.
15.5-5 REVISION 10 - DECEMBER 2004
B/B-UFSAR
- i. Operator Action Time Operator action is not required to mitigate the consequences of this event. Operator action is assumed to occur after the event to stabilize the plant in accordance with approved procedures to bring the plant to the applicable condition.
- j. Pressurizer Safety Valves The safety valves setpoints do not impact the minimum DNBR since the PORVs are assumed available to maintain low RCS pressure; this assumption is conservative with respect to DNBR.
- k. Auxiliary Feedwater Auxiliary feedwater was not credited.
- l. Main Steam Safety Valves The main steam safety valves are assumed conservatively to open at +5% above their nominal set pressure for the DNB case. No credit for steam dump is assumed in this analysis.
Plant systems and equipment credited for mitigating the effects of the accident are discussed in Subsection 15.0.8 and listed in Table 15.0-7. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.
Results The transient response is shown in Figures 15.5-1 through 15.5-3.
Table 15.5-1 shows the calculated sequence of events.
Nuclear power starts decreasing immediately due to boron injection, but steam flow does not decrease until later in the transient when the turbine throttle valve is wide open. The mismatch between load and nuclear power causes Tavg, pressurizer 15.5-6 REVISION 11 - DECEMBER 2006
B/B-UFSAR water level, and pressurizer pressure to drop. The reactor trips and control rods start moving into the core when the pressurizer pressure reaches the pressurizer low pressure trip setpoint. The DNBR increases throughout the transient.
15.5.1.3 Radiological Consequences There are only minimal radiological consequences associated with inadvertent ECCS operation. The reactor trip causes a turbine trip and heat is removed from the secondary system through the steam generator power relief valves or safety valves. Since no fuel damage is postulated to occur for this transient, the radiological consequences associated with an atmospheric steam release from this event would be less severe than the steamline break event analyzed in Subsection 15.1.5.3.
Water relief from the pressurizer PORVs and safeties may result in overpressurization of the pressurizer relief tank (PRT),
breaching the rupture disk and spilling contaminated fluid into containment. The radiological releases (offsite doses) resulting from breaking the PRT rupture disk are limited by isolation of the containment.
15.5.1.4 Conclusions Results of the analysis show that spurious ECCS operation at measurement uncertainty recapture conditions without immediate reactor trip does not present any hazard to the integrity of the core or the RCS with respect to DNBR. The minimum DNBR is never less than the initial value. If the reactor does not trip immediately, the low pressurizer pressure reactor trip will provide protection. This trips the turbine and prevents excess cooldown, which expedites recovery from the incident.
With respect to pressurizer filling, RCS pressure will stabilize well below the RCS pressure safety limit of 2735 psig. The performance of the pressurizer safety valve system and the loads on pressurizer safety valves, associated piping, and supports will be within acceptable limits.
15.5-7 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.5.2 Chemical and Volume Control System Malfunction That Increases Reactor Coolant Inventory An increase in reactor coolant inventory which results from the addition of cold, unborated water to the reactor coolant system is analyzed in Subsection 15.4.6, chemical and volume control system malfunction that results in a decrease in boron concentration in the reactor coolant. An increase in reactor coolant inventory which results from the injection of highly borated water into the reactor coolant system is analyzed in Subsection 15.5.1, inadvertent operation emergency core cooling system during power operation.
15.5.3 A Number of BWR Transients (Not applicable to the Byron/Braidwood Stations.)
15.5.4 References
- 1. Burnett, T. W. T., et al., "LOFTRAN Code Description,"
WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary),
April, 1984.
- 2. ANS-51.1/Nl8.2-1973, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."
- 3. Freidland, A. J. and Ray, S., "Revised Thermal Design Procedure," WCAP-11397-P-A (Proprietary), WCAP-11397-A (Non-Proprietary) April 1989.
- 4. Technical Evaluation Report, TMI Action Plan - NUREG-0737 (II.D.1) Braidwood Unit 1 & 2. Docket No. 50-456, 50-457, G. K. Miller et. al, Idaho National Engineering Laboratory, January 1988.
- 5. NRC Letter from Mr. Leonard N. Olshan to ComEd Henry E.
Bliss, dated August 18, 1988,
Subject:
NUREG-0737, Item II.D.1, Performance Testing on Relief and Safety Valves for Byron Station, Units 1 and 2 (TAC Nos. 56200 and 63240) transmitting Technical Evaluation Report (TER) providing the results of the NRCs review on Byron Units 1 and 2 response to NUREG-0737, Item II.D.1.
- 6. NRC Letter from Mr. George F. Dick (NRR) to Oliver D.
Kingsley (Exelon), dated May 4, 2001.
Subject:
Issuance of Amendments; Increase in Reactor Power, Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (TAC Nos.
MA9428, MA9429, MA9426 and MA9427).
- 7. NRC Letter from Mr. George F. Dick (NRR) to Mr. Christopher M. Crane (Exelon), dated August 26, 2004.
Subject:
Byron Station, Units 1 and 2, and Braidwood Station Units 1 and 2
- Issuance of Amendments, RE: Pressurizer Safety Valve Setpoints, (TAC Nos. MB9762, MB9763, MB9760 and MB9761).
15.5-8 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.5-1 TIME SEQUENCE OF EVENTS FOR INCREASE IN REACTOR COOLANT INVENTORY EVENTS ACCIDENT EVENT TIME (sec.)
Inadvertent Actuation of ECCS During Power Operation Spurious SI signal generated; two charging pumps begin injecting borated water 0 Turbine throttle valve wide open, load begins to drop with steam pressure 51.5 Low pressurizer pressure reactor trip setpoint reached 74.1 Control Rod Motion Begins 76.1 Minimum DNBR occurs (*)
(*) - DNBR does not decrease below its initial value 15.5-9 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.6 DECREASE IN REACTOR COOLANT INVENTORY Events which result in a decrease in reactor coolant inventory are discussed in this section, as follows:
- a. inadvertent opening of a pressurizer safety or relief valve,
- b. failure of small lines carrying primary coolant outside containment,
- c. steam generator tube rupture,
- d. BWR piping failure outside containment (not applicable),
and
- e. loss-of-coolant accident resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary.
15.6.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve 15.6.1.1 Identification of Causes and Accident Description An accidental depressurization of the reactor coolant system could occur as a result of an inadvertent opening of a pressurizer relief or safety valve. Since a safety valve is sized to relieve approximately twice the steam flow-rate of a relief valve, and will therefore allow a much more rapid depressurization upon opening, the most severe core conditions resulting from an accidental depressurization of the reactor coolant system are associated with an inadvertent opening of a pressurizer safety valve. Initially the event results in a rapidly decreasing reactor coolant system pressure that could reach the hot leg saturation pressure if a reactor trip did not occur. The pressure continues to decrease throughout the transient. The effect of the pressure decrease is to decrease power via the moderator density feedback, but the reactor control system (if in the automatic mode) functions to maintain the power and average coolant temperature until reactor trip occurs. Pressurizer level increases initially due to expansion caused by depressurization and then decreases following reactor trip.
The reactor may be tripped by the following reactor protection system signals:
- a. overtemperature T and
- b. pressurizer low pressure.
15.6-1 REVISION 6 - DECEMBER 1996
B/B-UFSAR An inadvertent opening of a pressurizer safety valve is classified as an ANS Condition II event, a fault of moderate frequency. See Subsection 15.0.1 for a discussion of Condition II events.
15.6.1.2 Analysis of Effects and Consequences Method of Analysis The accidental depressurization transient is analyzed by employing the detailed digital computer code LOFTRAN (Reference 1). The code simulates the neutron kinetics, reactor coolant system, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.
This accident is analyzed with the revised thermal design procedure as described in WCAP-11397-P-A. Plant characteristics and initial conditions are discussed in Subsection 15.0.3.
In order to give conservative results in calculating the DNBR during the transient, the following assumptions are made:
- a. Initial reactor power, pressure, and RCS temperatures (consistent with the uprated power conditions) are assumed to be at their nominal values. With the exception of the RCS average temperature bias, which is explicitly modeled in the analysis, uncertainties in initial conditions are included in the limit DNBR as described in WCAP-11397-P-A.
- b. A zero moderator temperature coefficient is assumed.
The spatial effect of void due to local or subcooled boiling is not considered in the analysis with respect to reactivity feedback or core power shape.
- c. A least negative Doppler-only power coefficient is assumed (see Figure 15.0-3) such that the resultant amount of negative feedback is conservatively low in order to maximize any power increase due to moderator reactivity feedback.
D. Cases assuming both D5 and BWI steam generator models at maximum steam generator tube plugging levels with both minimum and maximum feedwater temperatures were analyzed.
Plant systems and equipment credited for mitigating the effects of a reactor coolant system depressurization caused by an inadvertent safety valve opening are discussed in Subsection 15.0.8 and listed in Table 15.0-7.
Normal reactor control systems are not required to function. The reactor protection system 15.6-2 REVISION 15 - DECEMBER 2014
B/B-UFSAR functions to trip the reactor on the appropriate signal. No single active failure will prevent the reactor protection system from functioning properly.
Results The most limiting case (D5 steam generators at the maximum steam generator tube plugging level and minimum feedwater temperature) for an inadvertent opening of a pressurizer safety or relief valve is shown on Figures 15.6-1 and 15.6-2. Figure 15.6-1 illustrates the nuclear power transient following the depressurization. Nuclear power is maintained at the initial value until reactor trip occurs on low pressurizer pressure. The pressure decay transient and average temperature transient following the accident are given in Figure 15.6-2. Pressure drops more rapidly while core heat generation is reduced via the trip, and would then slow once saturation temperature is reached in the hot leg. The DNBR decreases initially, but increases rapidly following the trip, as shown in Figure 15.6-1. The DNBR remains above the limit value throughout the transient.
The calculated sequence of events for the inadvertent opening of a pressurizer safety valve incident is shown on Table 15.6-1.
15.6.1.3 Radiological Consequences An inadvertent opening of a pressurizer safety or relief valve releases primary coolant to the pressurizer relief tank; however, even assuming a direct release to the containment atmosphere, the radiological consequences of this event would be substantially less than that of a LOCA (Subsection 15.6.5) because less primary coolant is released and the activity is lower as fuel damage is not predicted as a result of this event.
15.6.1.4 Conclusions The results of the analysis show that the pressurizer low pressure and the overtemperature T reactor protection system signals provide adequate protection against the RCS depressurization event. No fuel or clad damage is predicted for this accident. The radiological consequences of this event would be substantially less than that of the LOCA analyzed in Subsection 15.6.5.
15.6.2 Failure of Small Lines Carrying Primary Coolant Outside Containment 15.6.2.1 Identification of Causes and Accident Description An assumed CVCS line break outside containment releases primary coolant at a rate of 140 gpm. Primary coolant activities are given in Tables 15.0-9 and 15.0-10. Iodine spiking is taken into consideration.
An accident which results from a break in small sample lines connected to the primary coolant system and penetrating the 15.6-3 REVISION 15 - DECEMBER 2014
B/B-UFSAR containment will cause expulsion of the coolant at a rate which can be accommodated by a charging pump. The charging pump would maintain an operational water level in the pressurizer, permitting the operator to conduct an orderly shutdown. The release contains the radionuclide concentration of the primary coolant.
The maximum break size for which the normal makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the reactor coolant system (RCS) through the postulated break against the charging pump makeup flow at normal RCS pressure, i.e., 2250 psia. A makeup flow rate from one centrifugal charging pump is adequate to sustain pressurizer level and a pressure of 2250 psia for a break through a 0.375 inch diameter hole. This break results in a loss of approximately 17.5 lb/sec, and, due to the use of a 0.245 inch restriction, is the maximum flow available for all reactor coolant sample line breaks outside of the containment. In addition, all such lines meet the requirements of General Design Criterion 55 of Appendix A 10 CFR 50. There are no instrument lines which pass through the containment and connect directly to the RCS. A failure of a small line carrying primary coolant outside containment is classified as an ANS Condition II event, a fault of moderate frequency. See Subsection 15.0.2 for a discussion of Condition II events.
15.6.2.2 Analysis of Effects and Consequences A break assumed to occur immediately downstream of valve CV8152 would result in low level in the volume control tank if other remedial action were not taken. Fifteen minutes is allowed for the operator to determine what has happened and to close the valve. Spring closure of the valve is assumed to take 5 seconds.
The break of a small sample line does not result in a leakage rate greater than the capacity of a charging pump and pressurizer level does not decrease, normal shutdown procedures can be employed. There are no significant consequences to the reactor or its essential auxiliary systems.
15.6.2.3 Radiological Consequences A partition factor of 0.1 and a DF of 10 in the building filtration system were applied to the release of radioiodine assuming a CVCS line break outside containment.
Conservative (fifth percentile) values of /Q at the exclusion area boundary (Tables 15.0-13 and 15.0-14) were used in the analysis.
The following radiological doses are calculated as a consequence of this accident:
15.6-4 REVISION 17 - DECEMBER 2018
B/B-UFSAR TWO HOUR DOSE AT EXCLUSION AREA BOUNDARY (rem)
WHOLE BODY THYROID Byron 0.03 1.0 Braidwood 0.04 1.4 These doses meet the dose acceptance criterion of being a small fraction of the 10 CFR Part 100 dose guidelines where small fraction is defined as being 10% (Reference 41).
There are no unusual features associated with the plant design which would prevent the limitation of the radiological consequences of a failure of a small sample line carrying primary coolant outside containment to an acceptable level by utilizing the appropriate limits or reactor coolant activity concentrations or isolation valve closing time and leak rates.
15.6.3 STEAM GENERATOR TUBE RUPTURE The SGTR analysis was performed to evaluate the two major SGTR potential consequences of concern, specifically:
margin-to-overfill (MTO) case - the potential for overfilling the ruptured steam generator before the AFW can be isolated and the break flow terminated.
offsite dose case - the potential to release primary system activity through the secondary side in excess of 10 CFR 50.67 limits.
The background information and results of the SGTR analysis are presented in the following sections.
15.6.3.1 Identification of Causes and Accident Description The accident examined is the complete severance of a single steam generator tube. The accident is assumed to take place at steady-state reactor power with the reactor coolant system activity at the allowable technical specification operating limit. The accident leads to an increase in secondary system activity due to the leakage of radioactive primary system water into the ruptured steam generator. In the event of a coincident loss of offsite power and subsequent failure of the steam dump system, there is a discharge of activity to the atmosphere through the steam generator safety and/or power-operated relief valves (PORVs).
15.6-5 REVISION 15 - DECEMBER 2014
B/B-UFSAR Since the steam generator tube material is Inconel, a highly ductile material, the assumption of a complete tube severance is conservative.
Steam generator operating experience has demonstrated the most probable mode of failure would be one or more minor tube leaks of varying sizes of undetermined origin. An aggressive nondestructive examination policy is implemented during refueling outages to identify failing steam generator tubes and repair those with indications exceeding conservatively established limits. In addition, the steam and power conversion system is continually monitored for activity indicating tube leakage and an accumulation of minor leaks exceeding the technical specification limits is not permitted during operation.
The response of the operator and his ability to implement recovery actions is critical in mitigating the consequence of a SGTR event. The operator actions assumed in these two analyses are based upon the plant-specific emergency procedures addressing the SGTR accident. The emergency procedures are based upon the latest revision of the Westinghouse Owners Group generic emergency response guidelines and are, therefore, consistent with the latest guidelines regarding SGTR mitigation.
The operator action times used in this analysis are based upon the Westinghouse WCAP-10698-P-A SGTR analysis methodology and operator response times observed during initial and requalification operator license simulator training. The operator action times used in the two analyses are shown in Table 15.6-5a.
The MTO analysis assumption time intervals in Table 15.6-5a are validated on the simulator. The operator action times for simulator verification do not include the calculated plant thermal hydraulic response.
A simple validation of the operator action times modeled in the MTO analyses requires that the operators perform each of the actions within the time listed in Table 15.6-5a. However, evaluations can be performed to show that acceptable results continue to be obtained if some of these action times are traded off. For example, performing the earlier actions (e.g. isolating AFW) in a shorter time than analyzed allows for the possibility that the later times (e.g. starting the depressurization) may be increased over the values used in the analyses and still demonstrate acceptable results.
The SGTR offsite dose case does not require operator simulator verification, since the scenario is mitigated by isolating the stuck open SG PORV within 30 minutes and the remaining required operator actions are the same as the MTO case. All of the individual, assumed analysis operator action time intervals for the UFSAR design basis SGTR cases are discussed in detail for reference purposes.
15.6-6 REVISION 15 - DECEMBER 2014
B/B-UFSAR There are six major recovery phases during the SGTR event. The first five phases require timely completion to terminate the primary to secondary tube leakage and ensure overfill does not occur or offsite and Control Room dose limits are not exceeded.
These six recovery phases are given below and are discussed in detail later for both the margin-to-overfill and offsite dose case analyses.
- 1) Identify The Ruptured Steam Generator The first critical step is for the operator to identify a SGTR exists and determine which steam generator is ruptured. Since the SGTR generates a reactor trip and safety injection, the operator enters the E-0 procedure to verify proper automatic system responses, assess the plant conditions, and determine the appropriate recovery procedure. While in the E-0 procedure, the SGTR identification can be made by several possible indications.
High secondary activity on the steam jet air ejector radiation monitor or other secondary monitors would positively identify a SGTR. These monitors would lose direct sampling capability at the time of the loss of offsite power (LOOP), but the recorder readouts and residual meter indication would show above normal secondary activity. For scenarios where the reactor trip occurs some time after the tube rupture, there is a definite possibility the operator will identify the SGTR prior to trip based upon a reduction in feedwater flow to the ruptured steam generator or steam generator level deviation. The main identification factor for analysis purposes is the increase in the ruptured steam generator level and recovery rate compared to the intact steam generators following the reactor trip. This is due to the addition of break flow and auxiliary feedwater (AFW) flow. The identification of the SGTR directs the operator to enter the E-3 SGTR procedure where the remaining recovery actions are performed.
- 2) Isolation of the Ruptured Steam Generator The next major step is to isolate the ruptured steam generator.
The operator first isolates AFW flow. Isolation is completed by closing the MSIV and MSIV bypass valves after ensuring the PORV and the blowdown valves are closed.
The isolation of the ruptured steam generator provides three functions: 1) it minimizes feedwater accumulation in the steam generator and decreases the potential for overfill, 2) it isolates the ruptured steam generator from the intact steam generators, preventing blowdown and release from the ruptured steam generator when the intact steam generator PORVs are used to cool the RCS, 3) it maximizes the ruptured steam generator pressure, thereby reducing the amount of RCS depressurization needed to terminate the primary to secondary leakage.
15.6-7 REVISION 12 - DECEMBER 2008
B/B-UFSAR
- 3) Cooldown of the Reactor Coolant System For this step, the RCS must be cooled down below the saturation temperature of the ruptured steam generator. Cooldown would normally be performed by using the steam dumps. However, the LOOP makes the condenser unavailable, and the operator initiates the RCS cooldown by opening the intact steam generator PORVs.
The cooldown ensures the RCS remains adequately subcooled when the primary system is depressurized below the ruptured SG pressure.
- 4) Depressurization of the Reactor Coolant System As stated previously, the RCS must then be depressurized to equalize pressure with the ruptured steam generator and terminate the primary to secondary leakage. The operator would normally use the pressurizer spray system to depressurize the RCS; however, in the advent of a LOOP, the sprays are not available since the RCPs are tripped. Therefore, the operator opens the pressurizer PORV to decrease the RCS pressure below the ruptured steam generator. The RCS depressurization increases the rate at which SI flow can restore the RCS mass inventory and terminates the primary to secondary leakage.
- 5) Termination of the Safety Injection Flow Unless terminated, the SI flow continues repressurizing the RCS until the break flow reinitiates. Therefore, the E-3 procedure specifies that the operator must verify RCS pressure is stable or increasing and then terminate all SI flow except for one centrifugal charging pump. The operator then establishes normal charging flow and adjusts it as necessary to maintain adequate pressurizer level.
- 6) Place the Reactor in the Cold Shutdown Condition At this point, the primary to secondary leakage has been terminated and the RCS is stable and ready for transition to the cold shutdown condition. This requires depressurization and cooldown of the ruptured steam generator and cooldown of the RCS to less than 200F. The three different methods available per the E-3 series procedures to perform the transition to cold shutdown are:
post-SGTR cooldown using backfill - this method drains the ruptured steam generator through the ruptured tube into the RCS by incrementally decreasing RCS pressure below the steam generator pressure.
post-SGTR cooldown using blowdown - this method depressurizes the ruptured steam generator by draining it through the blowdown lines.
15.6-8 REVISION 4 - DECEMBER 1992
B/B-UFSAR post-SGTR cooldown using steam dump - this method depressurizes the ruptured steam generator by dumping steam via the steam dumps to the condenser or via the SG PORV.
The transition to the cold shutdown method used is determined by the TSC personnel based upon available plant equipment and system conditions.
15.6.3.2 Analysis of Effects and Consequences The analyses for the offsite and Control Room dose and margin-to-overfill cases were performed utilizing the most limiting plant parameters as identified in Table 15.6-5 and the limiting single failures as identified in Table 15.0-15.
The following analysis assumptions were made for both the offsite and Control Room dose and margin-to-overfill transient cases.
Conservative operator action times for the two cases were used as shown in Table 15.6-5a.
Loss of offsite power (LOOP) is assumed to occur concurrent with the reactor trip.
No credit is taken for the chemical and volume control system charging or letdown flow.
Prior to the reactor trip, the SG level control system maintains the faulted steam generator water level essentially constant, compensating for the additional break flow by reducing the feed flow.
The operator throttles the AFW flow to the intact SGs as necessary to maintain adequate narrow range level.
During the RCS depressurization, the operator ensures the pressurizer level stays within the limits specified in the E-3 procedure.
Offsite and Control Room Dose Case The major sequence of operator actions for the offsite and Control Room dose case is the same as the above summary, but there is one extra operator manipulation required during the isolation of the ruptured SG. The limiting single failure involves the PORV on the ruptured SG failing in the full open position. This causes the SG to be both ruptured and faulted (ruptured/faulted).
The time sequence of events for the offsite and Control Room dose case is shown in Table 15.6-6a and the key transient parameter responses are shown in Figures 15.6-3a, 15.6-3b, and 15.6-3I for Unit 1 and Figures 15.6-3c, 15.6-3d, and 15.6-3J for Unit 2.
15.6-9 REVISION 15 - DECEMBER 2014
B/B-UFSAR Automatic Actions After the tube rupture occurs, reactor coolant immediately begins flowing from the primary system into the secondary side of the ruptured steam generator causing the RCS pressure to decrease until a reactor trip occurs on overtemperature Delta-T. The reactor trip signal closes the turbine stop valves isolating steam flow to the turbine. The normal feedwater flow to the SGs is also isolated due to the reactor trip.
The LOOP occurs coincident with the reactor trip causing the RCPs to trip and the main condenser to become unavailable when the circulating water pumps are lost.
After the reactor trips, the core power quickly decreases to decay heat levels. The steam dump system cannot be used to dissipate the core decay heat due to the unavailable condenser.
Therefore, the secondary pressure increases in the SGs until the PORVs open.
The RCS pressure continues decreasing, and a low pressurizer pressure safety injection (SI) signal is generated. The RCS break flow steadily decreases from time zero as the RCS pressure drops, and it reduces substantially following the reactor trip as the pressure increases in the ruptured SG secondary side, thereby reducing the primary to secondary pressure differential. The RCS pressure then begins increasing again, as the ECCS injection flow exceeds the RCS break flow restoring RCS mass inventory.
The RCS temperature drops after the reactor trip. With the RCPs tripped, the reactor coolant system evolves into the natural circulation cooling mode.
15.6-10 REVISION 15 - DECEMBER 2014
B/B-UFSAR Major Operator Actions The operator isolates AFW flow to the ruptured SG and then closes the ruptured SG MSIV. The ruptured SG PORV is assumed to fail open at the time the MSIV is closed.
The ruptured SG failed PORV is isolated when the block valve is manually closed thirty minutes after the PORV fails open.
After the isolation of the ruptured SG, the operator initiates the RCS cooldown by opening the intact SG PORVs. The operator continues the cooldown until the RCS temperature is subcooled below the saturation temperature at the ruptured SG pressure.
The operator initiates the RCS depressurization by opening the pressurizer PORV. The operator maintains the pressurizer PORV open until the RCS pressure decreases below the pressure in the ruptured SG. The pressurizer PORV is reclosed and the depressurization terminated.
After the RCS depressurization is completed the operator terminates ECCS flow.
Margin-To-Overfill Case The sequence of events for the margin-to-overfill case is shown in Table 15.6-6b, and the key parameter transient responses are shown in Figures 15.6-3e and 15.6-3f for Unit 1 and 15.6-3g and 15.6-3h for Unit 2.
For the tube rupture event, the overtemperature T reactor trip setpoint is reached earlier than the low pressurizer pressure reactor trip setpoint. Since the time from tube rupture to auxiliary feedwater isolation is an assumed operator action time, an early reactor trip results in auxiliary feedwater going to the ruptured steam generator for a longer period of time which leads to a reduction in margin-to-overfill.
15.6-11 REVISION 15 - DECEMBER 2014
B/B-UFSAR Automatic Actions The initial sequence for the MTO case is similar to the OD case as the tube rupture occurs at time, t = 0, and the primary to secondary break flow initiates. The reactor trips on overtemperature T. The reactor trip signal closes the turbine stop valves isolating steam flow to the turbine.
The loss of offsite power occurs coincident with the reactor trip, resulting in the RCPs tripping and the condenser becoming unavailable due to the loss of the circulating water pumps. The loop also results in a trip of the main feedwater pumps.
The auxiliary feedwater pumps automatically start on loss-of-offsite power, and the auxiliary feedwater flow immediately begins entering the steam generators. The RCS pressure continues decreasing and initiates a safety injection signal on low pressurizer pressure.
15.6-11a REVISION 15 - DECEMBER 2014
B/B-UFSAR After the trip, the core power quickly decreases to decay heat levels. However, since the steam dump system cannot be used to dissipate the core decay heat due to the unavailable condenser, the secondary pressure increases until the ruptured SG PORV setpoint is reached. The valve relieves pressure produced by the decay heat and the rising liquid level (see Figures 15.6-3f and 15.6-3h).
The liquid volume in the steam generator with the tube rupture begins increasing immediately after the reactor trip and continues to increase up to the termination of the transient.
Major Operator Actions The operator is able to identify and isolate AFW flow to the ruptured SG within 9 minutes after tube rupture occurs. The ruptured SG water level continues increasing after isolation due to break flow, but at a reduced rate.
The operator then initiates the RCS cooldown by opening the intact SG PORVs. Since one of the intact SG PORVs fails to open, there is a resulting reduced cooldown rate and a longer time to reach the E-3 procedure RCS subcooling requirement. The operator terminates the RCS cooldown after meeting the subcooling requirement.
The operator initiates the RCS depressurization by opening the pressurizer PORV.
After termination of the RCS depressurization, the operator halts the ECCS injection flow and the break flow begins steadily decreasing the RCS pressure. Detailed operator actions involved in termination of ECCS flow and subsequent actions to reduce RCS pressure and terminate break flow are not modeled in the analysis. This leads to a conservatively long period of continued break flow after ECCS termination is modeled before the RCS and ruptured SG pressures equalize.
15.6-12 REVISION 15 - DECEMBER 2014
B/B-UFSAR The SGTR Margin to Overfill transient is terminated when the RCS pressure is equalized with the ruptured SG.
Results In the SGTR offsite and Control Room dose case there is no concern for steam generator overfill since the failed open PORV reduces the secondary mass and results in a much lower SG level at termination. At the time the SGTR margin-to-overfill case is terminated, the ruptured SG still contains more than 50 cubic feet and more than 250 cubic feet of gas volume margin for Unit 1 and Unit 2, respectively. This demonstrates there are sufficient indications and controls which provide the operator adequate time to terminate the break flow before steam generator overfill occurs.
15.6.3.3 Radiological Consequences It was determined that the doses resulting from a SGTR at Unit 2 are higher than those from a SGTR at Unit 1. The Unit 2 analysis is presented.
The key inputs and assumptions used in the AST Steam Generator Tube Rupture (SGTR) radiological consequence analysis are summarized below and provided in Table 15.6-6.
The SGTR accident is postulated as a complete severance of a single Steam Generator (SG) tube. This is a conservative assumption because tube material is Inconel, a highly ductile metal alloy, and the most probable mode of failure would be one or more minor tube leaks of varying sizes and undetermined origin.
The tube rupture results in the release of radioactive material from the Containment system. For the three intact SGs, primary to secondary coolant leakage continues to transfer activity into the Secondary Coolant side. This makes it available for release into the environment via steaming through the SG Power Operated Release Valves (PORV). For the SG with the ruptured tube, referred to as the ruptured SG, coolant release will take two forms:
Break Flow - un-flashed release of RCS coolant directly into the secondary loop, and made available for steaming release to the environment through the SG PORV.
Flashed Break Flow - RCS coolant that flashes directly to steam when released from the ruptured tube, and is sent through the SG PORV to the environment.
15.6-12a REVISION 15 - DECEMBER 2014
B/B-UFSAR SGTR accident mitigation can be described in recovery phases.
The major phases that this analysis uses to model the dose consequences from this event are shown in Table 15.6-6, along with the approximate time increments that are associated with the sequence of events:
Consistent with RG 1.183, two reactor transients that maximize the radioactivity available for release were modeled.
Case 1: Dose Due to Pre-accident Iodine Spike The first case involves a 60 Ci/gm pre-accident Iodine spike.
This 60-Ci/gm spike is consistent with the Technical Specification operational Reactor Coolant System (RCS) activity concentration limit for an assumed spike. In this scenario, it is assumed that all of the spike activity is homogeneously mixed in the primary coolant, prior to accident initiation.
Case 2: Dose Due to Accident Initiated Concurrent Iodine Spike The second case involves an accident initiated iodine spike that occurs concurrently with the release of fluid from the primary and secondary coolant systems. Regulatory guidance specifies that this spike should result in a release rate from defective fuel that is 335 times the normal rate, and lasts for an 8-hour duration.
15.6-12b REVISION 15 - DECEMBER 2014
B/B-UFSAR Fuel Damage and Core Source Term The design basis assumes no fuel damage for the postulated steam generator tube rupture event. For this SGTR accident, the source terms are defined by the Technical Specification Dose Equivalent (DE) I-131 iodine and DE Xe-133 noble gas activity in the primary reactor coolant system. Because no fuel damage is assumed for this accident, only iodine and noble gas isotopes are modeled to contribute to dose. To identify the worst-case SGTR accident, however, the two different cases of iodine spiking described above are analyzed, per regulatory guidance.
In addition, for both cases, a 0.1 Ci/gm DE I-131 equilibrium secondary coolant activity concentration limit from Technical Specifications, is included.
Activity Removal Mechanisms in Containment The design basis SGTR releases activity directly into the primary RCS, therefore no plateout, or other activity deposition, is credited.
Decay Credited:
Decay of radioactivity is credited, prior to release to the environment.
Depletion from Leakage Credited:
For analyses of doses due to release from the RCS volume, the dose results from leakage. It is reasonable to credit the small amount of depletion from the available RCS activity inventory associated with this leakage. This is calculated inherently by the RADTRAD code.
Release Rates and Partitioning Factors:
As discussed above, a number of modes of release are indicative of this particular accident scenario. Therefore, the varying releases associated with the timing and sequence of events of this accident was derived.
Activity that originates in the primary RCS is released to the secondary coolant by means of the primary-to-secondary coolant leak rate. This design basis leak rate value is 0.218 gpm, per intact SG, totaling 0.654 gpm. This volumetric flow rate is converted to mass flow using the assumption of cooled liquid conditions (i.e., 62.4 lbm/ft3), as specified by RG 1.183.
15.6-12c REVISION 15 - DECEMBER 2014
B/B-UFSAR The methodology used to model steaming of activity through SG PORVs following the postulated SGTR event assumes an average cumulative release rate through the SG valves. The partitioning factors are applied to these release rates. For the time increments used in this accident scenario, release rates were derived by taking the averages of these rates over each specified time increment.
The ruptured/faulted steam generator sees two simultaneous release mechanisms. Primary to secondary coolant leakage through the ruptured tube that flashes conservatively goes directly to the environment, without mixing with any secondary coolant.
Therefore, with this release mechanism, no partitioning of iodine is expected to occur in this release. However, leakage that does mix with the volume of coolant in the ruptured SG is released by flashing to the environment, and the applicable partition factor is applied, as discussed below.
For all post-accident releases through the SG PORVs, the mechanism for release to the environment is steaming of the coolant in the secondary system. Because of this release dynamic, RG 1.183 allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water. For Iodine, the partitioning factor of 0.01 was taken directly from the suggested guidance of RG 1.183. Reviewing the specified AST release fractions, it is concluded that the only nuclides to be released from the core source term, other than iodines, are noble gas nuclides, and because of the volatility of noble gases, no partitioning is assumed for any such isotopes.
In addition to the steam released through the SG PORVs, the steam release through the Condenser until the time of Reactor trip and loss of offsite power is also accounted for. For steam flow that is released through the Condenser an additional 0.01 factor is applied to model partitioning in this pathway.
/Q Calculations (Meteorology)
Releases from the SG PORVs are considered elevated releases due to the high steaming rates, and the associated /Qs have been reduced by a factor of 5, per guidance in RG 1.194, as described in Section 15.4.8.3.
This event was reanalyzed using the finer wind speed categories of Regulatory Guide 1.23, Revision 1.
15.6-12d REVISION 16 - DECEMBER 2016
B/B-UFSAR Assumptions and Inputs The following inputs and assumptions were used in the SGTR analysis.
- a. There is no fuel damage as a result of the postulated steam generator tube rupture accident.
- b. In the case of a postulated iodine activity release rate spike, the spike release is assumed to occur for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
- c. The activity released from the fuel is assumed to be instantaneously mixed with the RCS.
- d. All iodine released from the SGs is modeled consistent with the RG 1.183 specification of 97% elemental and 3% organic.
- e. The Control Room HVAC system is realigned to the emergency mode of operation 30 minutes after the initiation of this design basis accident.
- f. The break flow modeled in the dose analysis is shown in Figure 15.6-3C.
- g. The break flashing fraction modeled in the dose analysis is shown in Figure 15.6-3J.
- h. The PORV steam releases until break flow termination modeled in the dose analysis are shown in Figure 15.6-3D.
- i. In addition to PORV steam releases, the full power steam release through the Condenser until the time of reactor trip and loss of offsite power is accounted for.
15.6-12e REVISION 15 - DECEMBER 2014
B/B-UFSAR Dose Results Radiological doses resulting from a design basis SGTR for a control room operator and a person located at EAB or LPZ are to be less than the regulatory dose limits as given below.
Regulatory Dose Limits - SGTR Dose Type Control Room EAB and LPZ (rem) (rem)
Case 1 TEDE Dose 5a 25b Case 2 TEDE Dose 5a 2.5b Notes:
a 10 CFR 50.67 b
10 CFR 50.67 as modified by Regulatory Guide 1.183 (Table 6, Page 1.183-20)
The table below provides the results.
Steam Generator Tube Rupture Accident Radiological Analysis Results Case 1: Pre-Accident 60 Ci/gm DE I-131 Spike Dose Assessment Results Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 2.0 3.7 0.69 Case 2: Accident Initiated 335 times Equilibrium Iodine Release Rate Spike Dose Assessment Results Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 0.56 2.1 0.41 These doses are below the Regulatory Dose Limits, so it is verified that this design basis Steam Generator Tube Rupture accident is sufficiently mitigated at both Byron and Braidwood Stations.
15.6-12f REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.6.3.4 Conclusions A steam generator tube rupture will cause no subsequent damage to the reactor coolant system or the reactor core. An orderly recovery from the accident can be completed before steam generator overfill occurs even assuming simultaneous loss of offsite power. The radiological consequences are within the 10 CFR 50.67 and RG 1.183 guidelines.
15.6.4 Spectrum of BWR Steam System Piping Failures Outside of Containment This section is not applicable to the Byron/Braidwood Stations.
15.6.5 Loss of Coolant Accidents Resulting From a Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary A comprehensive safety analysis of postulated pipe breaks within the reactor coolant system (RCS) boundary has been performed.
This analysis has included cases of the loss-of-coolant accident (LOCA) resulting from a broad spectrum of small and large pipe breaks up to and including the double-ended break of the largest RCS pipe.
15.6-12g REVISION 15 - DECEMBER 2014
B/B-UFSAR The objective of the analysis was to determine the conditions of the RCS, core, and containment in the event of a postulated LOCA and to demonstrate that the emergency core cooling system (ECCS) has the capability to mitigate each LOCA.
For the analysis reported here, a major pipe break (large break) is defined as a break with a total cross-sectional area equal to or greater than 1.0 ft2. This event is considered an ANS Condition IV event, a limiting fault that is not expected to occur during the lifetime of the plant, but is postulated as a conservative design basis.
A minor pipe break (small break) is defined as a break of the reactor coolant pressure boundary with a total cross sectional area less than 1.0 ft2. This event is considered an ANS Condition III event, an infrequent event fault which may occur during the life of the plant.
15.6.5.1 General 15.6.5.1.1 Performance Criteria for Emergency Core Cooling System The reactor is designed to withstand thermal effects caused by a loss-of-coolant accident including the double-ended severance of the largest reactor coolant system pipe. The reactor core and internals together with the emergency core cooling system are designed so that the reactor can be safely shut down and the essential heat transfer geometry of the core is preserved following the accident. The emergency core cooling system, even when operating during the injection mode with the most severe single active failure, is designed to meet the requirements of 10 CFR 50.46 (Reference 2). The requirements are:
15.6-13 REVISION 16 - DECEMBER 2016
B/B-UFSAR
- a. The calculated maximum fuel element cladding temperature shall not exceed 2200F.
- b. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
- c. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
- d. Calculated changes in core geometry shall be such that the core remains capable of being cooled.
- e. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptable low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
For the Best Estimate LB LOCA analysis, it is noted that criteria a through c above are satisfied by ensuring that there is a high level of probability that when uncertainties in the analysis method and inputs are accounted for, the criteria are not exceeded.
15.6-13a REVISION 13 - DECEMBER 2010
B/B-UFSAR 15.6.5.1.2 Accident Description A LOCA would result from a break of the RCS piping or of any line connected to that system up to the first closed valve. The charging pumps have the capability to make up for leakage resulting from ruptures of a small cross section, thus permitting an orderly shutdown. The coolant released would remain in the containment.
For a postulated break, reactor trip is initiated when the pressurizer low pressure setpoint is reached while the safety injection system (SI) signal is actuated by pressurizer low pressure or containment high pressure. The consequences of the accident are limited in two ways:
- a. Reactor trip and borated water injection supplement void formation in causing rapid reduction of the nuclear power to a residual level corresponding to the delayed fission product decay; and
- b. Injection of borated water ensures sufficient flooding of the core to prevent excessive temperatures.
Before the reactor trip occurs, the reactor is in a equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. After reactor trip and turbine trip, core heat and heat from hot internals and the vessel is transferred to the RCS fluid and then to the secondary system.
The secondary system pressure increases and steam dump may occur.
Makeup to the secondary side is provided automatically by the auxiliary feedwater pumps. The SI signal stops normal feedwater flow by closing the main feedwater line isolation valves and initiates auxiliary feedwater flow by starting the auxiliary feedwater pumps. If offsite power is available, the steam is dumped to the condenser; if not, the steam is dumped to the atmosphere. The secondary flow aids in the reduction of RCS pressure. When the RCS pressure falls below 600 psia, the accumulators begin to inject borated water.
Reactor coolant pump (RCP) operation is assumed as follows. For small break LOCA analysis, the RCPs are assumed to trip, with attendant time delay, at the time of loss of offsite power, which is assumed to occur simultaneously with low pressurizer pressure reactor trip. Large break LOCA analyses cases for both with and without loss of offsite power are performed. Large break LOCA cases with loss of offsite power assume RCP trip at the transient initiation, with no time delay, and the effects of RCP coastdown are included in the blowdown analysis. Large break LOCA cases without loss of offsite power assume powered RCP operation throughout the transient.
15.6-14 REVISION 9 - DECEMBER 2002
B/B-UFSAR 15.6.5.2 Thermal Analysis The analysis specified by 10 CFR 50.46 is presented in this subsection. The time sequence of events for the LOCA analysis is provided in Tables 15.6-1a, 15.6-1b, 15.6-1c, 15.6-1d and 15.6-1e. The results of the loss-of-coolant accident analysis are shown in Tables 15.6-3, 15.6-3a, 15.6-4, 15.6-4a, 15.6-4b and 15.6-4c and show compliance with acceptance criteria.
15.6-14a REVISION 9 - DECEMBER 2002
B/B-UFSAR The SBLOCA analysis is based on reactor conditions shown in Table 15.6-2. The detailed description of the SBLOCA analysis methodology is given in References 11, 13, and 34. These documents describe the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the acceptance criteria. The NOTRUMP (References 11 and 13) and LOCTA-IV (Reference 8) codes are used to assess the core heat transfer geometry and to determine if the core remains capable of being cooled throughout the SBLOCA. The input assumptions in the Best Estimate large break analysis are summarized in Table 15.6-2a and 15.6-3b, as discussed in further detail in Section 15.6.5.2.1.3. The codes, interfaces, and modeling consideration associated with the Best Estimate large break analysis are described in detail in Section 15.6.5.2.1.4.
The conditions of Table 15.6-2 reflect four-loop operation.
Containment parameters used in the large break analysis are given in Section 6.2.1.5.
The method of analysis to determine peak clad temperature is divided into two types of analysis: 1) large break LOCA, and 2) small break LOCA. The methods of analysis for large and small break LOCA are described below and results are given.
15.6.5.2.1 Large Break Analysis 15.6.5.2.1.1 General When the Final Acceptance Criteria (FAC) governing the loss-of-coolant accident (LOCA) for Light Water Reactors was issued in Appendix K of 10 CFR 50.46, both the Nuclear Regulatory Commission (NRC) and the industry recognized that the stipulations of Appendix K were highly conservative. That is, using the then accepted analysis methods, the performance of the Emergency Core Cooling System (ECCS) would be conservatively underestimated, resulting in predicted Peak Clad Temperatures (PCTs) much higher than expected. At that time, however, the degree of conservatism in the analysis could not be quantified.
As a result, the NRC began a large-scale confirmatory research program with the following objectives:
- 1) Identify, through separate effects and integral effects experiments, the degree of conservatism in those models permitted in the Appendix K rule. In this fashion, those areas in which a purposely prescriptive approach was used in the Appendix K rule could be quantified with additional data so that a less prescriptive future approach might be allowed.
- 2) Develop improved thermal-hydraulic computer codes and models so that more accurate and realistic accident analysis calculations could be performed. The purpose of this research was to develop an accurate predictive capability so that the uncertainties in the ECCS performance and the degree of conservatism with respect to the Appendix K limits could be quantified.
15.6-15 REVISION 14 - DECEMBER 2012
B/B-UFSAR Since that time, the NRC and the nuclear industry have sponsored reactor safety research programs directed at meeting the above two objectives. The overall results have quantified the conservatism in the Appendix K rule for LOCA analyses and confirmed that some relaxation of the rule can be made without a loss in safety to the public. It was also found that some plants were being restricted in operating flexibility by the overly conservative Appendix K requirements. In recognition of the Appendix K conservatism that was being quantified by the research programs, the NRC adopted an interim approach for evaluation methods. This interim approach is described in SECY-83-472 (Reference 5). The SECY-83-472 approach retained those features of Appendix K that were legal requirements, but permitted applicants to use best-estimate thermal-hydraulic models in their ECCS evaluation model. Thus, SECY-83-472 represented an important step in basing licensing decisions on realistic calculations, as opposed to those calculations prescribed by Appendix K.
In 1998, the NRC Staff amended the requirements of 10 CFR 50.46 and Appendix K, "ECCS Evaluation Models" (Reference 2), to permit the use of a realistic evaluation model to analyze the performance of the ECCS during a hypothetical LOCA. This decision was based on an improved understanding of LOCA thermal-hydraulic phenomena gained by extensive research programs. Under the amended rules, best-estimate thermal-hydraulic models may be used in place of models with Appendix K features. The rule change also requires, as part of the LOCA analysis, an assessment of the uncertainty of the best-estimate calculations. It further requires that this analysis uncertainty be included when comparing the results of the calculations to the prescribed acceptance criteria of 10 CFR 50.46. Further guidance for the use of best-estimate codes is provided in Regulatory Guide 1.157 (Reference 37).
To demonstrate use of the revised ECCS rule, the NRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology (Reference 38). This method outlined an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis.
A LOCA evaluation methodology for three- and four-loop Pressurized Water Reactor (PWR) plants based on the revised 10 CFR 50.46 rules was developed by Westinghouse with the support of EPRI and Consolidated Edison and has been approved by the NRC (Reference 36).
More recently, Westinghouse developed an alternative uncertainty methodology called ASTRUM, which stands for Automated Statistical TReatment of Uncertainty Method (Reference 4). This method is still based on the CQD methodology and follows the steps in the CSAU methodology (Reference 38). However, the uncertainty analysis (Element 3 in the CSAU) is replaced by a technique based on order statistics. The ASTRUM methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are 15.6-15a REVISION 14 - DECEMBER 2012
B/B-UFSAR simultaneously sampled for each case. The ASTRUM methodology has received NRC approval for referencing in licensing calculations in Reference 4.
The three 10 CFR 50.46 criteria (peak clad temperature, maximum local oxidation, and core-wide oxidation) are satisfied by running a sufficient number of WCOBRA/TRAC calculations (sample size). In particular, the statistical theory predicts that 124 calculations are required to simultaneously bound the 95th percentile values of three parameters with a 95-percent confidence level.
The Byron/Braidwood analysis is in accordance with the applicability limits and usage conditions defined in Section 13-3 of Reference 4, as applicable to the ASTRUM methodology. Section 13-3 of Reference 4 was found to acceptably disposition each of the identified conditions and limitations related to WCOBRA/TRAC and the CQD uncertainty approach per Section 4.0 of the ASTRUM Final Safety Evaluation Report.
Two analytical models for the Byron/Braidwood Stations were developed for the best estimate large break LOCA analyses. One analysis (Unit 1) utilized the BWI Replacement Steam Generator geometry and the other (Unit 2) utilized the Westinghouse D-5 Steam Generator Geometry. The remaining analysis parameters were identical. Very slight differences were observed in the vessel design and these differences were blended together in a generally conservative fashion to create one common vessel model as characterized in detail in Section 3 of Reference 40.
Input parameters used for the Byron/Braidwood minimum containment pressure analysis are presented in Section 6.2.1.5. Mass and energy releases from the WCOBRA/TRAC Reference transient were utilized to confirm the containment back pressure utilized in the Reference Transient as described in detail in Section 6.2.1.5.
15.6-16 REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.6.5.2.1.2 Design Basis Accident The Byron/Braidwood PCT-limiting transients are double-ended cold leg guillotine breaks which analyze conditions that fall within those listed in Table 15.6-3b. Traditionally, cold leg breaks have been limiting for large break LOCA. This location is the one where flow stagnation in the core appears most likely to occur. Scoping studies with WCOBRA/TRAC have confirmed that the cold leg remains the limiting break location (Reference 36). The design basis accident is described in more detail in Section 15.6.5.2.3.1.
15.6.5.2.1.3 Analysis Assumptions The expected PCT and its uncertainty developed are valid for a range of plant operating conditions. The range of variation of the operating parameters has been accounted for in the uncertainty evaluation. Table 15.6-3b summarizes the operating ranges for Byron/Braidwood as defined for the proposed operating conditions, which are supported by the Best-Estimate LBLOCA analysis. Section 6.2.1.5 summarizes the LBLOCA containment data used for calculating containment pressure. If operation is maintained within these ranges, the LBLOCA results using WCOBRA/TRAC are considered to be valid. Note that some of these parameters vary over their range during normal operation (accumulator temperature) and other ranges are fixed for a given operational condition (Tavg) (see Table 15.6-2a).
15.6.5.2.1.4 Method of Analysis. Large Break The methods used in the application of WCOBRA/TRAC to the large break LOCA with ASTRUM are described in Reference 36 and Reference 4. A detailed assessment of the computer code WCOBRA/TRAC was made through comparisons to experimental data.
These assessments were used to develop quantitative estimates of the code's ability to predict key physical phenomena in a PWR large break LOCA. Modeling of a PWR introduces additional uncertainties which are identified and quantified in the plant-specific analysis. WCOBRA/TRAC MOD7A was used for the execution of ASTRUM for Byron/Braidwood.
WCOBRA/TRAC combines two-fluid, three-field, multi-dimensional fluid equations used in the vessel with one-dimensional drift-flux equations used in the loops to allow a complete and detailed simulation of a PWR. This best-estimate computer code contains the following features:
- 1) Ability to model transient three-dimensional flows in different geometries inside the vessel.
- 2) Ability to model thermal and mechanical non-equilibrium between phases
- 3) Ability to mechanistically represent interfacial heat, mass, 15.6-17 REVISION 14 - DECEMBER 2012
B/B-UFSAR and momentum transfer in different flow regimes.
- 4) Ability to represent important reactor components such as fuel rods, steam generators, reactor coolant pumps, etc.
A typical calculation using WCOBRA/TRAC begins with the establishment of a steady-state, initial condition with all loops intact. Following the establishment of an acceptable steady-state condition, the transient calculation is initiated by introducing a break into one of the loops. The evolution of the transient through blowdown, refill, and reflood proceeds continuously, using the same computer code (WCOBRA/TRAC) and the same modeling assumptions. Containment pressure is modeled with the BREAK component using a time dependent pressure table.
Containment pressure is calculated using the COCO code (Reference
- 7) and mass and energy releases from the WCOBRA/TRAC calculation.
The final step of the best-estimate methodology, in which all uncertainties of the LOCA parameters are accounted for to estimate a PCT, Local Maximum Oxidation (LMO), and CoreWide Oxidation (CWO) at 95-percent probability, is described in the following sections.
- 1) Plant Model Development:
In this step, a WCOBRA/TRAC model of the plant is developed.
A high level of nodding detail is used in order to provide an accurate simulation of the transient. However, specific guidelines are followed to ensure that the model is consistent with models used in the code validation. This results in a high level of consistency among plant models, except for specific areas dictated by hardware differences, such as in the upper plenum of the reactor vessel or the ECCS injection configuration.
- 2) Determination of Plant Operating Conditions:
In this step, the expected or desired operating range of the plant to which the analysis applies is established. The parameters considered are based on a "key LOCA parameters" list that was developed as part of the methodology. A set of these parameters, at mostly nominal values, is chosen for input as initial conditions to the plant model. A transient is run utilizing these parameters and is known as the "initial transient".
The list of these parameters is given in Table 15.6-2a.
Next, several confirmatory runs are made, which vary a subset of the key LOCA parameters over their expected operating range in one-at-a-time sensitivities. Because certain parameters are not included in the uncertainty analysis, these parameters are set at their bounding condition. This analysis is commonly referred to as the confirmatory analysis. The most limiting input conditions, based on these confirmatory runs, are then combined into the model that will represent the limiting state for the plant, which is the starting point for the assessment of 15.6-17a REVISION 14 - DECEMBER 2012
B/B-UFSAR uncertainties. The results of the confirmatory analyses are given in Table 15.6-3.
- 3) Assessment of Uncertainty:The ASTRUM methodology is based on order statistics. The technical basis of the order statistics is described in Section 11 of Reference 4. The determination of the PCT uncertainty, LMO uncertainty, and CWO uncertainty relies on a statistical sampling technique.
According to the statistical theory, 124 WCOBRA/TRAC calculations are necessary to assess against the three 10 CFR 50.46 criteria (PCT, LMO, CWO).
The uncertainty contributors are sampled randomly from their respective distributions for each of the WCOBRA/TRAC calculations. The list of uncertainty parameters, which are randomly sampled for each time in the cycle, break type (split or double-ended guillotine), and break size for the split break are also sampled as uncertainty contributors within the ASTRUM methodology.
Results from the 124 calculations are tallied by ranking the PCT from highest to lowest. A similar procedure is repeated for LMO and CWO. The highest rank of PCT, LMO, and CWO will bound 95 percent of their respective populations with 95-percent confidence level.
- 4) Plant Operating Range:
The plant operating range over which the uncertainty evaluation applies is defined. Depending on the results obtained in the above uncertainty evaluation, this range may be the desired range or may be narrower for some parameters to gain additional margin.
Note that the LBLOCA analysis was performed with ZIRLO cladding. However, Reference 6 concluded that the LOCA ZIRLO cladding models are acceptable for application to Optimized ZIRLO' cladding in the Large Break analysis, and that no additional calculations are necessary for evaluating the use of Optimized ZIRLO cladding provided that plant specific ZIRLO calculations were previously performed.
15.6-17b REVISION 16 - DECEMBER 2016
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-17c REVISION 14 - DECEMBER 2012
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-17d REVISION 14 - DECEMBER 2012
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-17e REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.6.5.2.2 Small Break LOCA Analysis 15.6.5.2.2.1 Description of Analysis and Assumptions The small break LOCA analysis was performed using the Westinghouse ECCS small break evaluation model (Reference 12) which utilizes the NOTRUMP (References 11 and 13) and LOCTA-IV (Reference 8) computer codes. Figure 15.6-5 shows the interaction of the computer codes used to evaluate the small break cases. Two complete break spectrums were analyzed, one for Unit 1 (1.5, 2, 3, and 4-inch diameter) and one for Unit 2 (1.5, 2, 3, and 4-inch diameter). These spectrums were selected in order to ensure that the limiting PCT is bounded for the small break LOCA event.
In addition to the description of the NOTRUMP code in References 11 and 13, safety injection was explicitly modeled in the broken loop. Figure 15.6-13b presents the broken loop safety injection flow from one safety injection line for Unit 1 and 2. The charging pump line discharges to the intact loop at RCS pressure, consistent with the implemented COSI methodology (Reference 34).
15.6-18 REVISION 9 - DECEMBER 2002
B/B-UFSAR The core power level utilized in the small break LOCA analysis was 3659 MWt. Since accumulator water volume is not a significant parameter in small beak LOCA analyses, the large break assumption was employed. The core power and peaking factors used in the analyses are also given in Table 15.6-2.
Additionally the main steam safety valves were set 5% above the Technical Specifications setpoint value and required an additional 3% accumulation before being fully open.
Figure 15.6-13c presents the hot rod power shape utilized to perform the small break analysis. This actual power shape was chosen because it provides that distribution of power versus core height which will maximize peak clad temperature given an upper limit of +13% for the axial flux difference.
Reference 12 covered a range of break locations for the small break LOCA. This study determined that the bottom of the cold leg was the most limiting location. Therefore, only the spectrum of cold leg breaks covering the range of 1.5-, 2-, 3-, and 4-inch diameter breaks for Unit 1 and Unit 2 were analyzed in order to determine the most limiting break size. Once the limiting break size was determined, the range of RCS operating temperatures was analyzed. The operating range of temperatures is the nominal Tavg of 588.0oF plus uncertainties to the reduced Tavg of 575.0oF minus uncertainties. These breaks were analyzed following the method presented in Subsection 15.6.5.2.2.2.
The limiting single failure in the small break LOCA analysis is the loss of one diesel, which results in the loss of one train of SI pumps under loss of offsite power.
The limiting conditions of burnup for small break LOCA need to be specifically determined due to the effects of burst upon the small break LOCA peak clad temperature. For peak clad temperatures (PCTs) less than 1700oF, a beginning-of-life (BOL) burnup is assumed. No burnup spectrum study is performed since the burst PCT at higher burnups will not exceed BOL PCT. If the PCT exceeds 1700oF, a burnup study is performed to obtain the limiting time in life for rod burst. For Unit 1 and Unit 2, a burnup study was not performed since the PCT was less than 1700oF (Unit 1 - 1624oF, Unit 2 - 1627oF).
Figures 15.6-13a and 15.6-13b provide the intact and broken loop safety injection (total ECCS) flows assumed in the small break LOCA analysis. Consistent with the implemented COSI model (Reference 34), the intermediate safety injection and charging lines for Unit 1 and Unit 2 discharge to the broken and intact loop at RCS pressure. The safety injection flows also incorporate a 5% flow reduction for future pump degradation.
15.6-19 REVISION 15 - DECEMBER 2014
B/B-UFSAR The small break LOCA analysis was performed with the assumptions appearing on Table 15.6-2. The analysis assumed operation at a vessel thermal design flow rate of 368,000 gpm, 5% steam generator tube plugging for Unit 1 and 10% steam generator tube plugging for Unit 2, and a 40-second delay in delivery of pumped ECCS assuming loss of offsite power. Finally, to address the flexibility of a range of RCS operating temperatures, a sensitivity was performed to compare reduced RCS operating temperature to nominal RCS operating temperature. Table 15.6-2 identifies the nominal RCS operating temperature (upper range),
and the reduced RCS operating temperature (lower range) supported by the small break LOCA analysis.
Note that the SBLOCA analysis was performed with ZIRLO cladding.
However, Reference 6 concluded that the LOCA ZIRLO cladding models are acceptable for application to Optimized ZIRLO' cladding in the Small Break analysis, and that no additional calculations are necessary for evaluating the use of Optimized ZIRLO cladding provided that plant specific ZIRLO calculations were previously performed.
15.6.5.2.2.2 Method of Analysis, Small Break The NOTRUMP and LOCTA-IV computer codes are used to perform the analysis of loss-of-coolant accidents due to small breaks in the reactor coolant system (RCS). The NOTRUMP computer code, approved for this use by the Nuclear Regulatory Commission, is 15.6-19a REVISION 16 - DECEMBER 2016
B/B-UFSAR used to calculate the transient depressurization of the RCS as well as to describe the mass and enthalpy of the flow through the reactor core and break. This code is a state-of-the-art one-dimensional general network code incorporating a number of advanced features. Among these new features are the utilization of a nonequilibrium thermal calculation in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stack fluid nodes, and regime-dependent heat transfer correlations. The NOTRUMP small break LOCA emergency core cooling system evaluation model was developed to determine the RCS response to design basis small break LOCAs and to address the NRC concerns expressed in NUREG-0611, "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants."
In NOTRUMP, the RCS is subdivided into fluid-filled control volumes (fluid nodes) and metal nodes interconnected by flowpaths and heat transfer links. The transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum applied to these nodes. The broken loop is modeled explicitly, and the intact loops are lumped into a second loop. A detailed description of the NOTRUMP code is provided in References 11 and 13.
In the NOTRUMP model, the reactor core is represented as a vertical stack of heated control volumes with an associated bubble rise model to permit a transient mixture height calculation. The multi-node capability of the program enables the explicit and detailed spatial representation of various system components. In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant accident.
The COSI condensation model was used in the NOTRUMP portion of the Unit 1 and Unit 2 small break LOCA analyses. As explained in Reference 34, the COSI model is an improved condensation model based on data that modeled the configuration of the SI piping to the RCS cold leg. With this model, improved condensation of steam in the intact loops results in lower RCS pressure and larger SI flow rates. Further, increased condensation of steam by the SI water in the intact cold legs results in additional warming of the SI water prior to reaching the core. Warmer water entering the core results in increased steaming in the core, increased mixture void fraction, and increased mixture level.
Additionally, improvements in condensation in the broken loop by the broken loop SI can further decrease RCS pressure and may partially or completely offset any "plugging" effect on the break.
15.6-20 REVISION 8 - DECEMBER 2000
B/B-UFSAR Clad thermal analyses are performed with the LOCTA-IV computer code which uses as input the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core, and mixture height history from the NOTRUMP hydraulic calculations as input. For all computations, the NOTRUMP and LOCTA-IV calculations were terminated slightly after the time the core mixture level returned to the top of the core following core uncovery.
A schematic representation of the computer code interfaces is given in Figure 15.6-5 15.6.5.2.3 Results 15.6.5.2.3.1 Results of Large Break LOCA Analysis The large break LOCA transient can be divided into convenient time periods in which specific phenomena occur, such as various hot assembly heatup and cooldown transients. For a typical large break, the blowdown period can be divided into the Critical Heat Flux (CHF) phase, the upward core flow phase, and the downward core flow phase. These are followed by the refill, reflood, and long-term cooling periods. Specific important transient phenomena and heat transfer regimes are discussed below, with the transient results shown in Figures 15.6-8a through 15.6-8m (Unit
- 1) and 15.6-9a through 15.6-9m (Unit 2). (The PCT-limiting case was chosen to show a conservative representation of the response to a large break LOCA.)
- 1) Critical Heat Flux (CHF) Phase:
Immediately following the cold leg rupture, the break discharge rate is subcooled and high (Figures 15.6-8b, -8c and Figure 15.6-9b, -9c). The regions of the RCS with the highest initial temperatures (core, upper plenum, upper head, and hot legs) begin to flash to steam, the core flow reverses and the fuel rods begin to go through departure from nucleate boiling (DNB). The fuel cladding rapidly heats up (Figures 15.6-8a and 9a) while the core power shuts down due to voiding in the core. This phase is terminated when the water in the lower plenum and downcomer begins to flash (Figures 15.6-8g, 8k and Figures 15.6-9g, 9k respectively). The mixture swells and intact loop pumps, still rotating in single-phase liquid, push this two-phase mixture into the core.
- 2) Upward Core Flow Phase:
Heat transfer is improved as the two-phase mixture is pushed into the core. This phase may be enhanced if the pumps are not degraded, or if the break discharge rate is low due to saturated fluid conditions at the break. If pump degradation is high or the break flow is large, the cooling effect due to upward flow may not be significant.
Figures 15.6-8d and 15.6-9d show the void fraction for one intact loop pump and the broken loop pump. The figure shows that the intact loop remains in single-phase liquid 15.6-20a REVISION 14 - DECEMBER 2012
B/B-UFSAR flow for several seconds, resulting in enhanced upward core flow cooling. This phase ends as the lower plenum mass is depleted, the loop flow becomes two-phase, and the pump head degrades.
- 3) Downward Core Flow Phase:
The loop flow is pushed into the vessel by the intact loop pumps and decreases as the pump flow becomes two-phase.
The break flow begins to dominate and pulls flow down through the core, up the downcomer to the broken loop cold leg, and out the break. While liquid and entrained liquid flow provide core cooling, the top of core vapor flow (Figures 15.6-8e, 15.6-8f, 15.6-ge and 15.6-9f) best illustrates this phase of core cooling. Once the system has depressurized to the accumulator pressure, the accumulators begin to inject cold borated water into the intact cold legs (Figures 15.6-8h and 15.6-9h). During this period, due to steam upflow in the downcomer, a portion of the injected ECCS water is calculated to be bypassed around the downcomer and out the break. As the system pressure continues to fall, the break flow, and consequently the downward core flow, are reduced. The core begins to heat up as the system pressure approaches the containment pressure and the vessel begins to fill with ECCS water (Figures 15.6-81 and 15.6-91).
- 4) Refill Period:
As the refill period begins, the core begins a period of heatup and the vessel begins to fill with ECCS water (Figures 15.6-81 and 15.6-91). This period is characterized by a rapid increase in cladding temperatures at all elevations due to the lack of liquid and steam flow in the core region. This period continues until the lower plenum is filled and the bottom of the core begins to reflood and entrainment begins.
- 5) Reflood Period:
During the early reflood phase, the accumulators begin to empty and nitrogen enters the system. This forces water into the core, which then boils, causing system repressurization, and the lower core region begins to quench (Figures 15.6-8m and 15.6-9m). During this time, core cooling may increase due to vapor generation and liquid entrainment. During the reflood period, the core flow is oscillatory as cold water periodically rewets and quenches the hot fuel cladding, which generates steam and causes system re-pressurization. The steam and entrained water must pass through the vessel upper plenum, the hot legs, the steam generators, and the reactor coolant pumps before it is vented out of the break. This flow path resistance is overcome by the downcomer water elevation head, which provides the gravity driven reflood force.
From the later stage of blowdown to the beginning of reflood, the accumulators rapidly discharge borated 15.6-20b REVISION 14 - DECEMBER 2012
B/B-UFSAR cooling water into the RCS, filling the lower plenum and contributing to the filling of the downcomer. The pumped ECCS water (Figures 15.6-8i and 15.6-9i) aids in the filling of the downcomer and subsequently supplies water to maintain a full down comer and complete the reflood period. As the quench front progresses up the core (Figures 15.6-8j and 15.6-9j), the PCT location moves higher into the top core region (Figures 15.6-8m and 15.6-9m). As the vessel continues to fill, the PCT location is cooled the early reflood period is terminated.
A second cladding heatup transient may occur due to boiling in the downcomer. The mixing of ECCS water with hot water and steam from the core, in addition to the continued heat transfer from the hot vessel and vessel metal, reduces the subcooling of ECCS water in the lower plenum and downcomer. The saturation temperature is dictated by the containment pressure. If the liquid temperature in the downcomer reaches saturation, subsequent heat transfer from the vessel and other structures will cause boiling and level swell in the downcomer. The downcomer liquid will spill out of the broken cold leg and reduce the driving head, which can reduce the reflood rate, causing a late reflood heatup at the upper core elevations. Figures 15.6-8k and 15.6-9k shows only a slight reduction in downcomer level and indicates that a late reflood heatup does not occur.
Additionally, it must be demonstrated that there is a high level of probability that the limits set forth in 10 CFR 50.46 are met.
The demonstration that these limits are met is as follows:
(b)(1) The limiting PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the resulting PCT for the limiting cases are 1913 OF for Unit 1 and 2041 OF for Unit 2, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(1), i.e., "Peak Clad Temperature less than 2200°F, is demonstrated. The results are shown in Table 15.6-3a.
(b)(2) The limiting cladding oxidation corresponds to a bounding estimate of the 95th percentile LMO at the 95-percent confidence level. Since the resulting LMO for the limiting cases are 5.51% for Unit 1 and 8.27% for Unit 2, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Local Maximum Oxidation of the cladding less than 17 percent", is demonstrated. The results are shown in Table 15.6-3a.
(b)(3) The limiting core-wide oxidation corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. The limiting Hot Assembly Rod (HAR) total maximum oxidation is 0.25 percent for Unit 1 and 0.33 percent for Unit 2. A detailed CWO calculation takes advantage of the 15.6-21 REVISION 14 - DECEMBER 2012
B/B-UFSAR core power census that includes many lower power assemblies. Because there is significant margin to the regulatory limit, the CWO value can be conservatively chosen as that calculated for the limiting HAR. A detailed CWO calculation is therefore not needed because the outcome will always be less than 0.25 percent for Unit 1 and 0.33 for Unit 2. Since the resulting CWO is 0.25 percent for Unit 1 and 0.33 percent for Unit 2, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(3), i.e., "Core-Wide Oxidation less than 1 percent", is demonstrated. The results are shown in Table 15.6-3a.
(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains amenable to cooling.
This criterion has historically been satisfied by adherence to criteria (b)(1) and (b)(2), and by assuring that fuel deformation due to combined LOCA and seismic loads is specifically addressed. It has been demonstrated that the PCT and maximum cladding oxidation limits remain in effect for Best-Estimate LOCA applications. The approved methodology (Reference 36) specifies that effects of LOCA and seismic loads on core geometry do not need to be considered unless grid crushing extends beyond the 44 assemblies in the low-power channel.
This situation has not been calculated to occur for Byron/Braidwood. Therefore, acceptance criterion (b)(4) is satisfied.
(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS.
Long-term cooling is dependent on the demonstration of continued delivery of cooling water to the core.
The actions, automatic or manual, that are currently in place at these plants to maintain long-term cooling remain unchanged with the application of the ASTRUM methodology (Reference 4).
Based on the ASTRUM Analysis results (Table 15.6-3a), it is concluded that Byron/Braidwood continue to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.
15.6-21a REVISION 14 - DECEMBER 2012
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-21b REVISION 14 - DECEMBER 2012
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-21c REVISION 14 - DECEMBER 2012
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-21d REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.6.5.2.3.2 Results of Small Break LOCA Analysis 15.6.5.2.3.2.1 Unit 1 Analysis The time sequence of events and results of interest for the 1.5, 2, 3 and 4-inch diameter breaks at the reduced Tavg conditions are shown in Tables 15.6-1c and 15.6-4a, respectively.
The analysis included both Zircaloy-4 and ZIRLO fuel assemblies in accordance with the licensing guidelines of Reference 33.
Zircaloy-4 results for the limiting case have been provided in Tale 15.6-4d.
ZIRLO is a trademark or registered trademark in the United States of Westinghouse Electric LLC, its subsidiaries and/or its affiliates. This mark may also be used and/or registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.
15.6-21e REVISION 15 - DECEMBER 2014
B/B-UFSAR Figures 15.6-14a through 15.6-17n show the pertinent small break LOCA parameters during the transient.
Figures 15.6-14a - 15.6-17a (reduced) Peak Clad Temperature 15.6-14h - 15.6-17h (nominal)
Figures 15.6-14b - 15.6-17b (reduced) Local fluid 15.6-14i - 15.6-17i (nominal) temperatures at the peak clad temperature elevation Figures 15.6-14c - 15.6-17c (reduced) Local heat transfer 15.6-14j - 15.6-17j (nominal) coefficient at the peak clad temperature elevation Figures 15.6-14d - 15.6-17d (reduced) RCS pressure 15.6-14k - 15.6-17k (nominal)
Figures 15.6-14e - 15.6-17e (reduced) Core mixture height 15.6-14l - 15.6-17l (nominal)
Figures 15.6-14f - 15.6-17f (reduced) Core steam flow 15.6-14m - 15.6-17m (nominal)
Figures 15.6-14g - 15.6-17g (reduced) Break flow 15.6-14n - 15.6-17n (nominal)
The PCT for the limiting break, 2 inches in diameter at reduced Tavg was 1624oF, occurring at an assembly average burnup of 0 MWD/MTU with either Zircaloy-4 or ZIRLO fuel assembly materials.
The maximum local zirconium oxidation of 1.47%, which is less than the 17% criterion, and the core wide oxidation was less than 0.22% (Refer to Reference 35), which meets the 1% criterion.
These results indicate that a coolable geometry was maintained for the small break LOCAs and that, therefore, long-term cooling is assured by continued operation of the ECCS. Sometimes the small break LOCA PCT occurs at a later time in life due to rod burst and an associated zirconium-water driven spike in PCT.
However, since the PCT was less than 1700oF, burnup studies were not performed on the limiting 2-inch, reduced Tavg case. Instead, a BOL burnup of 0 MWD/MTU was assumed.
15.6.5.2.3.2.2 Unit 2 Analysis The time sequence of events and results of interest for the 1.5-,
2-, 3-, and 4-inch diameter breaks at the nominal Tavg conditions are shown in Tables 15.6-1d and 15.6-4b, respectively.
The analysis included both Zircaloy-4 and ZIRLO fuel assemblies in accordance with the licensing guidelines of Reference 33.
Zircaloy-4 fuel assemblies were found to be more limiting than the ZIRLO fuel assemblies for the 3-inch case at a burnup of 0 Mwd/Mtu. As shown in Table 15.6-4d, a burnup of 6000 Mwd/Mtu was assumed for Zircaloy-4 fuel to make ZIRLO clad fuel limiting.
15.6-21f REVISION 9 - DECEMBER 2002
B/B-UFSAR Figures 15.6-18a through 15.6-21n show the pertinent small break LOCA parameters during the transient.
Figures 15.6-18a - 15.6-21a (reduced) Peak clad temperature 15.6-18h - 15.6-21h (nominal)
Figures 15.6-18b - 15.6-21b (reduced) Local fluid temperature at 15.6-18i - 15.6-21I (nominal) the peak clad temperature elevation Figures 15.6-18c - 15.6-21c (reduced) Local heat transfer 15.6-18j - 15.6-21j (nominal) coefficient at the peak clad temperature elevation Figures 15.6-18d - 15.6-21d (reduced) RCS pressure 15.6-18k - 15.6-21k (nominal)
Figures 15.6-18e - 15.6-21e (reduced) Core mixture height 15.6-18l - 15.6-21l (nominal)
Figures 15.6-18f - 15.6-21f (reduced) Core steam flow 15.6-18m - 15.6-21m (nominal)
Figures 15.6-18g - 15.6-21g (reduced) Break flow 15.6-18n - 15.6-21n (nominal)
The PCT for the limiting break, 2-inch diameter at reduced Tavg, and ZIRLO fuel assembly materials was 1627F occurring at an assembly average burnup of 0 MWd/MTU. The maximum local zirconium oxidation was 1.59% and the core wide oxidation was less than 0.24% (See Reference 35) which meets the 1% criteria. These results indicate that a coolable geometry was maintained for small break LOCAs and therefore, long-term core cooling is assured by continued operation of the ECCS. Sometimes the small break LOCA PCT occurs at a later time in life due to rod burst and an associated Zirconium-water driven spike in PCT.
15.6.5.2.3.2.3 Safety Injection Evaluation Subsequent to the completion of the break spectrum study, a non-conservative discrepancy was discovered in the safety injection flows used in the analysis. Two data points were not included in the original cases and thus the flow modeled were non-conservatively higher in the analysis. These points more appropriately modeled the shutoff pressure of the high pressure safety injection pumps. The effect of this is greater on the 2-inch breaks and negligible for the 3-inch breaks and larger, greatly due to the increased dependency on safety injection for the smaller break sizes. The result for Unit 1 was a more limiting Peak Cladding Temperature (PCT) and the result for Unit 2 was a shift in break size and conditions to the 2-inch Low Tavg break.
The prior Unit 2 limiting case was the 3-inch High Tavg case.
Tables 15.6-1c, 15.6-1e, 15.6-4a, and 15.6-4c have been 15.6-21g REVISION 9 - DECEMBER 2002
B/B-UFSAR updated to reflect the results. Also, Figures 15.6-15a through 15.6-15g and Figures 15.6-18h through 15.6-18n have been updated to reflect the new 2-inch Low Tavg limiting case for Unit 1 and Unit 2, respectively. Note that for the Zirc-4 cladding evaluation, the results for Unit 2 are based on the High Tavg 3-inch prior limiting case. Although the new Unit 2 Low Tavg 2-inch limiting case has not been performed with Zirc-4 cladding, the prior results are conservatively being applied to the new limiting case because the effects are expected to be similar.
The impact of the SI discrepancy has also been evaluated on the other break sizes and resulted in a negligible impact on those.
Thus, the original results demonstrated herein remain applicable to those break sizes.
15.6.5.2.3.3 Post Analysis of Record Evaluations In addition to the analyses presented in Subsections 15.6.5.2.1 and 15.6.5.2.2, evaluations and reanalyses may be performed as needed to address emergent issues or to support plant changes.
The issues or changes are evaluated, and the impact on the PCT is determined. The resultant increase or decrease in PCT is added to the analysis of record.
The peak clad temperatures, including all penalties and benefits are presented in Table 15.6-15.
15.6.5.2.3.3.1 Large Break LOCA As documented in the preceding section two complete large break LOCA analyses were performed for both the Byron and Braidwood Units with a FTQ of 2.6 and FNH of 1.70 as reported in Table 15.6.3b. The 95 percentile PCT results are reported in Table 15.6-3a for these analyses.
In addition, evaluations/reanalysis are also performed from time to time to address the various issues or to support plant changes as they arise. On a cycle specific basis, if axial power shapes are beyond those reflected in Figure 15.6-7, the violating axial shapes are evaluated. The plant changes or the issues are evaluated and the impact on the PCT determined. The resultant increase or decrease in PCT is added to the analysis of record PCT. These issues and their evaluations are reported to the NRC via the normal 10CFR50.46 reporting requirement. The latest 10 CFR 50.46 report is publicly available on the NRC website in the Agencywide Documents Access and Management System (ADAMS).
Between UFSAR updates the latest PCT is tracked by the cognizant organization.
The current PCT (including PCT penalties/benefits associated with all the evaluations) for the four Byron and Braidwood Units are maintained in Table 15.6-15.
15.6-21h REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.6.5.2.3.3.2 Small Break LOCA - Unit 1 As documented in the preceding section a complete break spectrum small break LOCA analysis was performed for Byron and Braidwood Unit 1. The results of this analysis are reported in Tables 15.6-4 and 15.6.4a. The limiting break was determined to be the 2-inch diameter break at reduced Tavg conditions with a PCT of 1624F, the analysis of record. In addition, evaluations/reanalysis are performed from time to time to address the various issues or to support plant changes as they arise.
The plant changes or the issues are evaluated and the impact on the PCT determined. The resultant increase or decrease in PCT is added to the analysis of record PCT (1624F). These issues and their evaluations are reported to the NRC via the normal 10CFR50.46 reporting requirement. The latest 10 CFR 50.46 report is publicly available on the NRC website in the Agencywide Documents Access and Management System (ADAMS).
An evaluation of reduced recirculation flows was performed to support GSI-191 modifications using the flows provided in Figures 15.6-13d for intact loop and 15.6-13e for broken loop. The current PCT (including PCT penalties/benefits associated with all the evaluations) for the two Byron and Braidwood Unit 1s are maintained in Table 15.6-15.
15.6.5.2.3.3.3 Small Break LOCA - Unit 2 As documented in the preceding section a complete break spectrum small break LOCA analysis was performed for Byron and Braidwood Unit 2. The results of this analysis are reported in Tables 15.6-4b and 15.6.4c. The limiting break was determined to be the 2-inch diameter break at reduced Tavg conditions with a PCT of 1627F, the analysis of record.
In addition, evaluations/reanalysis are also performed from time to time to address the various issues or to support plant changes as they arise. The plant changes or the issues are evaluated and the impact on the PCT determined. The resultant increase or decrease in PCT is added to the analysis of record PCT (1627F)
These issues and their evaluations are reported to the NRC via the normal 10CFR50.46 reporting requirement. The latest 10 CFR 50.46 report is publicly available on the NRC website in the Agencywide Documents Access and Management System (ADAMS).
An evaluation of reduced recirculation flows was performed to support GSI-191 modifications using the flows provided in Figures 15.6-13d for intact loop and 15.6-13e for broken loop. The current PCT (including PCT penalties/benefits associated with all the evaluations) for the two Byron and Braidwood Unit 2s are maintained in Table 15.6-15.
15.6.5.2.4 Post-LOCA Long-Term Core Cooling/Subcriticality 10CFR50.46, Acceptance Criteria for Emergency Core Cooling 15.6-21i REVISION 15 - DECEMBER 2014
B/B-UFSAR Systems for Light Water Nuclear Power Reactors paragraph (b) item (5) sets forth the requirements for post-LOCA long-term core cooling. To satisfy the requirements, the core is maintained in a shutdown state solely by the soluble boron contained in the ECCS water after a LOCA because credit for the shutdown provided by the control rods was not taken for breaks > 1.0 ft2. Since safety injection flow is drawn from the sump following switchover from the RWST, the containment sump post-LOCA boron concentration must be higher than the boron concentration required to ensure subcritical conditions.
To determine if the requirements for post-LOCA long-term core cooling subcriticality are met, a calculation is performed for each reload to determine the boron concentration required to keep the core subcritical (Keff<1.0) and the mixed mean boron concentration (MMBC) of the post-LOCA sump water. This calculation, documented in the cycle-specific Reload Safety Analysis Checklist (RSAC), confirms that the post-LOCA sump MMBC exceeds the core critical boron concentration, thereby ensuring the reload core remains subcritical post-LOCA. Note: The post-LOCA long-term core cooling critical boron concentration is determined at the most reactive time in life, assuming an all rods out (ARO) no Xenon condition and a post-LOCA fluid temperature range of 68-212F. All sources of water that may eventually reside in the containment sump at cold leg recirculation switchover time and their respective pre-accident boron concentrations are considered.
Westinghouse has identified a potential safety issue concerning core recriticality following a large break cold leg break LOCA (Reference 43). The potential safety issue is that during hot leg switchover the core will be flushed with a diluted sump solution, which may cause the core to return to criticality. The sump solution would become diluted as boron accumulates in the core during the cold leg recirculation phase due to core boiling.
The accumulation of boron in the core prevents the boron from being displaced to the sump which leads to a diluted sump solution.
However, a generic assessment (Reference 43) concludes that for any given plant and fuel cycle, the boron worth of the inserted control rods plus the equivalent boron worth due to the presence of Xenon at the time of hot leg switchover, would offset any reasonable calculation of sump dilution. As documented in WCAP-15704 (Reference 44), it was demonstrated that control rods will insert following a licensing basis cold leg LOCA for 3-loop and 4-loop Westinghouse plant designs. Control rod insertion will result in negative reactivity benefits on the order of 400 ppm or more (boron equivalent rod worth). Thus, for Byron and Braidwood, the negative reactivity credit associated with control rods can be applied when evaluating recriticality at the time of switchover to hot leg ECCS recirculation. Note that for non-Westinghouse fuel further assessment is required before credit for control rod insertion can be taken to address the post-LOCA subcriticality issue. See section 7.4 of Reference 44 for details.
15.6-21j REVISION 12 - DECEMBER 2008
B/B-UFSAR Consequently, post-LOCA subcriticality will continue to be confirmed using sump boron calculations that do not consider sump dilution, but also use the conservative assumptions of ARO and no Xenon at the most reactive time in life, with an assumed post-LOCA core/fluid temperature in the range of 68-212F.
15.6.5.2.5 Conclusions - Thermal Analysis For breaks up to and including the double-ended severance of a reactor coolant pipe, the emergency core cooling system will meet the acceptance criteria as presented in Subsection 15.6.5.1.2 for a total core peaking factor, FTQ , of 2.60, and an FN H of 1.70.
15.6-21k REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.6.5.3 Radiological Consequences of a Postulated Loss-of-Coolant Accident The results of analyses presented in this section demonstrate that the amounts of radioactivity released to the environment in the event of a loss-of-coolant accident do not result in doses which exceed the guideline values specified in 10 CFR 50.67.
The key inputs and assumptions used in the AST LOCA radiological consequence analysis are summarized below and provided in Table 15.6-7.
EAB and LPZ Dose Model The EAB and LPZ /Qs have been determined, and the worst-case values are used for either of the Byron and Braidwood Stations as located at 445m and 1810m, respectively, from the postulated release locations.
Control Room Dose Model The following Control Room (CR) dose model, with assumed CR HVAC system operation, is used to calculate CR personnel dose for AST re-analyzed accident scenarios.
CR HVAC System Operation:
Actuation of the CR filtration system places the system in the emergency mode of operation. Actuation of the system to the emergency mode of operation starts the makeup fan, opens the turbine building intake damper, isolates the normal intake from outside dampers, isolates the purge dampers (if open), opens the recirculation charcoal adsorber dampers, and closes the recirculation charcoal adsorber bypass dampers. The operating supply and return fans continue to operate. Outside air from the turbine building is filtered and added to the air being recirculated through the CR.
Following CR emergency mode of operation isolation, the credited CR filtration is 99% for the HEPA and 95% for the charcoal filters. The maximum intake rate is used because sensitivity analyses have shown that it is conservative to maximize the potentially contaminated emergency mode of operation make-up airflow into the CR.
The CR HVAC system emergency mode of operation is not credited during the first 30 minutes of the accident, and therefore, unfiltered make-up air is assumed. This assumption is intended to allow for realignment of the CR HVAC system to the emergency mode of operation. During this 30-minute period, it is conservative not to consider the potential filtered intake entering the CR (i.e., the flow from the unused CR make-up train). This assumption is used because filtered flow would act to "clean" the unfiltered air being brought in by the other unfiltered flows (i.e., normal intake and unfiltered inleakage) during these 30 minutes. Sensitivity analyses using RADTRAD 15.6-22 REVISION 12 - DECEMBER 2008
B/B-UFSAR confirmed that such additional filtered intake during the first 30 minutes of an accident (i.e., with the CR filtration system in the normal mode of operation) would lower the CR dose consequences.
The CRs are modeled using a conservatively reduced recirculation train flow rate of 39,150 cfm. This recirculation train filtration consists of a 90% elemental and organic iodine filter (i.e., charcoal filter). No reduction in the efficiency of this filter is proposed.
For additional conservatism, an unfiltered inleakage rate allowance of 500 cfm is modeled for the accident duration.
The following table summarizes the CR dose model inputs and assumptions.
Intake Intake E Recirc. Recirc. E Unfiltered Flow to CR (cfm) (%) (cfm) (%) Inleakage UCSR Exfiltration Time (cfm) (cfm) (cfm)
Byron:
0-30 6424 0 39,150 0 500 1319 6924 min Braidwood:
2430 30 99 Byron:
min HEPA 90 1319 to 8575 95 39,150 Charcoal 500 Braidwood: 9075 30 Charcoal 2430 days Where:
E = Efficiency UCSR = Upper Cable Spreading Room Assumptions and Inputs Input parameters used for the LOCA analysis are given in Table 15.6-7.
The following primary assumptions from previous LOCA analyses continue to apply:
- a. Two release pathways are considered: containment leakage and ECCS recirculation leakage and
- b. A single electrical train failure is assumed to remove the following equipment from service: one containment spray train, one ECCS train, and two of the four reactor containment fan coolers (RCFCs).
15.6-22a REVISION 13 - DECEMBER 2010
B/B-UFSAR Primary Containment and ECCS Leakage Primary Containment Leakage:
Primary containment leakage is the main contributor to LOCA doses.
The impacts of the design basis LOCA are mitigated by: (1) a controlling design basis leak rate; and (2) operation of the containment spray system. While most penetrations are into the auxiliary building that has an ESF filtered exhaust, this filtration system is not credited for primary containment leakage.
Primary containment leakage is modeled as a diffuse area source in conformance with RG 1.194.
The assumed containment leak rate is 0.2% per day. This leak rate is assumed to be reduced to one-half the initial value after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> due to expected reductions in containment pressure.
Containment spray removal coefficients continue to be based on Standard Review Plan 6.5.2, Containment Spray as a Fission Product Cleanup System, Draft Revision 3, April 1996, with particulate removal coefficients applied to aerosols. Spray timing reflects AST-caused differences in time to reach decontamination factor (DF) credit limits.
Dispersion factors, developed in conformance with the guidance in RG 1.145 and RG 1.194 are used. Dispersion factors for primary containment leakage for the CR are based on a diffuse area source.
ECCS Leakage:
ECCS leakage is a minor contributor to LOCA doses. The ECCS leakage rate assumed in the AST LOCA analysis is 276,000 cc/hr. The ECCS leakage flashing fractions assumption is 10% for the duration of the accident.
Fuel Damage and Core Source Term The LOCA core source terms are those associated with a DBA power level of 3658.3 MWt. This power level bounds the MUR power uprate Rated Thermal Power level including measurement uncertainties.
The AST values used in this analysis were derived using guidance outlined in RG 1.183. A list of 60 core isotopic nuclides and their curie per megawatt activities was extracted from the RADTRAD NIF files. The release fractions associated with all of these nuclide groups, as detailed in RG 1.183, were applied to their given groups, and input into the RADTRAD RTF files. RADTRAD uses these files combined with the power of 3658.3 MWth to develop the source terms for the DBA LOCA.
15.6-22b REVISION 16 - DECEMBER 2016
B/B-UFSAR As discussed below, containment spray, natural deposition, decay, and leakage are credited as airborne activity removal mechanisms.
Removal by Containment Spray Iodine removal by containment spray is in accordance with Standard Review Plan Chapter 6.5.2. The major impacts of application of AST are that no initial plateout fraction is assumed and the duration of credited spray is modified.
Sprayed and Unsprayed Volumes, and Air Exchange Rate:
The containment volume is 2.85E6 cubic feet with 82.5% of the containment volume sprayed; i.e., the sprayed volume is 2.35125E6 cubic feet and the unsprayed volume is 4.9875E5 cubic feet.
These values are rounded to 2.35E6 and 5.0E5 cubic feet, respectively, in the analysis. Initial activity distribution is accordingly, 82.5% in the sprayed region, and 17.5% in the unsprayed region.
Transfer between these two volumes is assumed to be limited to that provided by the Reactor Containment Fan Coolers (RCFCs).
Even without the RCFCs, there would be significant mixing induced by the containment sprays and by the combination of steaming and heat transfer. Two of four RCFCs are credited in the analysis.
The assumed flow rate per RCFC is 65,000 cfm for a total of 130,000 cfm.
Spray Removal Coefficients for Aerosols (SRP 6.5.2 particulate removal):
From SRP 6.5.2, the first order removal coefficient for particulate (or, effectively, aerosols) may be estimated by:
p = 3 (h) (F) (E) 2 (V) (D) where:
p = spray removal constant, hr-1 h = drop fall height = 141 ft F = volume flow rate of sprays, ft3/hr
= 2950 gpm (applicable to both injection and recirculation phases)
= (2950 gpm)(60 min/hr) / (7.4805 gal/ft3)
= 23,661.5 ft3/hr V = sprayed volume, ft3
= (0.825)(2.85E6 ft3)
= 2.35125E6 ft3 E/D = ratio of a dimensionless collection efficiency E to the average spray drop diameter D
= 10 m-1 for Mo/Mt < 50
= 1 m-1 for Mo/Mt < 50 where Mo/Mt is the ratio of the initial aerosol mass to the aerosol mass at time t (note that this ratio also defines the DF achieved) 15.6-22c REVISION 12 - DECEMBER 2008
B/B-UFSAR P1 = 3 (141 ft) (23,661.5 ft3/hr) (10 m-1) (0.3048 m/ft) = 6.491 hr-1 2 (2.35E6 ft3)
P2 = 0.1 x 6.491 hr-1 = 0.65 hr -1 (The calculated P1 and P2 values above are conservatively reduced to 6.0 hr -1and 0.6 hr -1, respectively).
It is assumed that after the end of the core activity release process the aerosols would continue to be removed at a of 6.0 hr -1 until an overall DF of 50 is achieved.
Spray Removal Coefficients for Elemental Iodine (same as SRP 6.5.2 based elemental iodine removal):
The current SRP 6.5.2 based assessment of elemental iodine removal coefficients during containment spray are used. The design basis derivation of the spray removal coefficient is 20 hr -1,per SRP 6.5.2.
Elemental iodine removal is limited to a DF of 100.
Spray Timing:
The injection spray is initiated at 90 seconds and continues for 20.9 minutes from the time of initiation. There may be a delay of as much as ten minutes between the termination of containment spray during the ECCS injection phase and the initiation of containment spray during the ECCS recirculation phase.
For aerosol removal, the DF of 50 is reached at 2.21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />. From that point until eight hours, the removal coefficient of 0.6 hr -1 is used.
For elemental iodine removal, the DF of 100 is reached at 1.926 hours0.0107 days <br />0.257 hours <br />0.00153 weeks <br />3.52343e-4 months <br />.
After that time, no elemental iodine removal is credited.
Natural Deposition:
The RADTRAD computer program, including the Powers Natural Deposition algorithm based on NUREG/CR-6189, is used for modeling aerosol deposition in containment. No natural deposition is assumed for elemental or organic iodine. The lower bound (i.e., -10%) level of deposition credit is used.
Decay Credited:
Decay of radioactivity is credited in all compartments prior to release. This is implemented in RADTRAD using the half-lives in the NIF files. The RADTRAD decay plus daughter option is used. In reality, daughter products such as xenon from iodines or iodines from tellurium are unlikely to readily escape from the medium in which the parent iodine or tellurium is contained. The RADTRAD feature to include daughter effects is selected for conservatism.
15.6-22d REVISION 12 - DECEMBER 2008
B/B-UFSAR Depletion from Leakage Credited:
For analyses of doses due to leakage from containment, it is reasonable to credit the small amount of depletion from the containment inventory associated with this leakage. This is done inherently by RADTRAD.
Mechanisms for Environmental Releases:
The applicable release paths are leakage of airborne activity from primary containment, and leakage from ECCS system carrying reactor coolant outside containment into the Auxiliary Building.
These release paths are discussed below.
Containment source dose contributions are not significantly impacted by AST. Airborne containment activity is comparable or slightly smaller with AST due to time dependent release.
Therefore, no change in the containment component is assumed.
Auxiliary Building contained sources such as ECCS piping are not expected to see an increase in dose. However, airborne activity may be increased since the containment leak rate allowable has been doubled. Since dose contributors are not itemized, the values in UFSAR Tables E.20-1and E.20-2 were doubled for conservatism.
Control Room Direct Gamma Dose The contributors to Control Room doses due to gamma shine are Containment Building airborne activity; the post-LOCA plume surrounding the Control Room; and radioactivity accumulated on the Control Room Filter.
The pre-AST containment activity dose of 0.023 rem whole body is slightly conservative for AST conditions, and therefore is treated as a 0.023 rem TEDE dose contribution.
The pre-AST external plume shine dose of 0.003 rem whole-body to control room personnel is multiplied by a factor of 5 to yield 0.015 rem TEDE. Generally these doses are noble gas dominated.
This conservative multiplier accounts for the factor of 2 increase in primary containment leak rate assumptions, and other increase contributors such as ECCS iodine.
Fission product filter loading was reanalyzed using AST assumptions, and compared with those that would be determined using RG 1.3 assumptions. The pre-AST source terms were found to be bounding. Therefore, the pre-AST 0.013 rem whole body dose will continue to be used as an 0.013 rem TEDE dose. The total Direct Gamma Shine Dose is 0.05 rem TEDE.
15.6-22e REVISION 12 - DECEMBER 2008
B/B-UFSAR
/Q Calculations (Meteorology)
Control Room The CR /Q values input to RADTRAD were taken from the ARCON96 results of the design bases analyses. The limiting /Q values used are conservatively the worst-case combination of values from all four units at Byron Station and Braidwood Station. The atmospheric dispersion factors are given in Table 15.0-17.
Releases from primary containment to the environment and subsequently to the CR, utilize the /Q value associated with a diffuse area source in conformance with Regulatory Guide 1.194 (Diffuse Area Source Guidance).
Activity released during the initial 30 minutes of the accident is introduced to the CR via the normal CR fresh air intake.
After this period, when the CR is assumed to be in the emergency mode of operation, the emergency air intake located in the turbine building is used.
The CR, EAB, and LPZ /Q two-hour values are selected such that they coincide with the release period that caused the highest doses.
Dose Results Radiological doses resulting from a design basis LOCA to a control room operator and a person located at EAB or LPZ are to be less than the regulatory dose limits given below.
Regulatory Dose Limits - LOCA Dose Type Control Room EAB and LPZ (rem) (rem)
TEDE Dose 5a 25b Notes:
a 10 CFR 50.67 b
10 CFR 50.67 as modified by Regulatory Guide 1.183 (Table 6, Page 1.183-20) 15.6-22f REVISION 12 - DECEMBER 2008
B/B-UFSAR Loss of Coolant Accident Radiological Analysis Results Control Room EAB LPZ Source (rem TEDE) (rem TEDE) (rem TEDE)
Primary Containment 3.35 13.50 3.21 Leakage Maximum 2 hrs ECCS Leakage 1.54 1.53 1.78 Maximum 2 hrs Direct Dose from Containment, 0.05 None None External Plume,
& CR Filters Total Dose 4.94 15.03 4.99 The total doses are below the Regulatory Dose Limits, so it is verified that this design basis LOCA is sufficiently mitigated at both Byron Station and Braidwood Station.
15.6-22g REVISION 17 - DECEMBER 2018
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-23 REVISION 12 - DECEMBER 2008
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-24 REVISION 12 - DECEMBER 2008
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-25 REVISION 12 - DECEMBER 2008
B/B-UFSAR 15.6.6 BWR Transient This is not applicable to Byron/Braidwood.
15.6.7 References
- 1. Burnett, T. W. T., et al., "LOFTRAN Code Description,"
WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-proprietary),
April 1984.
- 2. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50. Federal Register, Volume 39, Number 3, January 4, 1974, as amended through 61 FR 39299, July 29, 1996.
- 3. "Reactor Safety Study - An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants," WASH-1400, NUREG-75/014, October 1975.
- 4. Information Report from W.J. Dircks to the Commissioners, Emergency Core Cooling System Analysis Methods, SECY 472, November 1983.
- 5. Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), WCAP-16009-P-A, January 2005.
15.6-26 REVISION 14 - DECEMBER 2012
B/B-UFSAR
- 6. Shah, H. H. and Schueren, P., Optimized ZIRLO',
WCAP-12610-P-A & CENPD-404-P-A Addendum 1-A, July 2006.
- 7. Bordelon, F. M. and Murphy, E. T., "Containment Pressure Analysis Code (COCO)," WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974.
- 8. Bordelon, F. M., et al., "LOCTA-IV Program: Loss of Coolant Transient Analysis," WCAP-8301 (Proprietary) and WCAP-8305 (Non-Proprietary), June 1974.
- 9. Deleted.
- 10. Deleted.
- 11. Meyer, P.E., NOTRUMP, A Nodal Transient Small Break and General Network Code, WCAP-10079-P-A, August 1985.
- 12. Lee, H. Rupprecht, S. D., Tauche, W. D., Schwarz, W. R.,
Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, WCAP-10054-P-A, August 1985.
- 13. Rupprecht, S. D., Osterrieder, R. A., Wills, M. E., Willis, J. M., Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study With The NOTRUMP Code, WCAP-11145-P-A, October 1986.
- 14. Deleted.
15.6-27 REVISION 16 - DECEMBER 2016
B/B-UFSAR
- 15. Deleted.
- 16. Deleted.
- 17. Deleted.
- 18. Deleted.
- 19. Letter from T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-2030, January 1979.
- 20. Deleted.
- 21. Deleted.
- 22. Deleted.
- 23. Deleted.
- 24. Deleted.
- 25. Deleted.
15.6-28 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- 26. Deleted.
- 27. Weiner, R.A., et al., "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations,"
WCAP-10851-P-A, August 1988.
- 28. Deleted.
- 29. Deleted.
- 30. Deleted.
- 31. Deleted.
- 32. Byron/Braidwood Increased Steam Generator Tube Plugging (24% Average/30% Peak) Analysis Program Report, "WCAP-14414, Revision 2, November 8, 1995.
- 33. Davidson, S. L., and Ryan, T. L., VANTAGE + Fuel Assembly Reference Core Report, WCAP-12610-P-A, April 1995.
- 34. Thompson, C. M., et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code:
Safety Injection into the Broken Loop and COSI Condensation Model," WCAP-10054-P-A, July 1997.
- 35. Humphries, B. S., (Westinghouse) Letter to S. Ahmed (Commonwealth Edison), "Commonwealth Edison Byron and Braidwood Units 1 and 2 Core Average Zircaloy-Water Reaction,"
CAE-97-165/CCE-97-224, July 7, 1997.
- 36. Code Qualification Document for Best Estimate LOCA Analysis, WCAP-12945-P-A (Proprietary), Volume I, Revision 2, and Volumes II-V, Revision 1, and WCAP-14747 (Non-Proprietary),
Westinghouse Electric Company, March 1998.
- 37. USNRC Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performances, May, 1989.
- 38. Qualifying Reactor Safety Margins: Application of Code Scaling Applicability and Uncertainty (CSAU) Evaluation Methodology to a Large Break Loss-of-Coolant-Accident, NUREG/CR-5249, 1989.
- 39. Letter, R. C. Jones (USNRC) to N. J. Liparulo (W), Acceptance for Referencing of the Topical Report WCAP-12945 (P),
Westinghouse Code Qualification Document for Best Estimate Loss-of-Coolant Analysis, June 28, 1996.
15.6-28a REVISION 16 - DECEMBER 2016
B/B-UFSAR
- 40. Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Byron/Braidwood Units 1 and 2 Nuclear Plant, WCAP-16841-P, Westinghouse Electric Company LLC, November 2007.
- 41. NUREG-0800, Standard Review Plan Section 15.6.2, Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant Outside Containment, Revision 2, July 1981.
- 42. Deleted.
- 43. Core Recriticality During Hot Leg Switchover, NSAL-94-016, Revision 2, March 18, 2002.
- 44. Control Rod Insertion Following a Cold Leg LOCA, Generic Analyses for 3-Loop and 4-Loop Plants, WCAP-15704, October, 2001.
15.6-28b REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE 15.6-1 TIME SEQUENCE OF EVENTS FOR INADVERTENT OPENING OF A PRESSURIZER SAFETY VALVE TIME EVENT (SEC)
Safety valve opens fully 0.0 Low Pressurizer Pressure 29.59 reactor trip setpoint reached Rods begin to drop 31.59 Minimum DNBR occurs 32.20 15.6-29 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-1a TIME SEQUENCE OF EVENTS FOR LARGE BREAK LOCA ANALYSIS LIMITING PCT CASE EVENT Unit 1 Unit 2 Start 0.0 0.0 Reactor trip N/A N/A signal Safety injection 6 6 signal Accumulator 14 12 injection begins End-of-bypass N/A N/A End-of-blowdown 26 24 Pump injection 46 46 begins
____________________
Note: All times are in seconds.
15.6-30 REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE 15.6-1a (Cont'd)
TIME SEQUENCE OF EVENTS FOR LARGE BREAK LOCA ANALYSIS LIMITING PCT CASE EVENT Unit 1 Unit 2 Bottom of core 36 33 recovery Hot rod burst 36 32 Peak clad 102 96 temperature Occurs Accumulators 46 56 empty
____________________
Note: All times are in seconds.
15.6-30a REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE 15.6-1b TIME SEQUENCE OF EVENTS FOR SMALL BREAK LOCA ANALYSIS UNIT 1 NOMINAL TAVG Event Time (sec) 1.5 Inch 2 Inch 3 Inch 4 Inch Break Initiation 0 0 0 0 Reactor Trip Signal 147.1 82.3 54.4 24.7 Safety Injection 159.3 93.9 66.5 35.6 Signal Safety Injection 199.3 133.9 106.5 75.6 Delivered Top of Core 15020 2112 863 637 Uncovered Accumulator N/A N/A 2002 920 Injection Begins Peak Clad 22680.6 3434.5 1834.7 1019.1 Temperature Occurs Top of Core >TMAX >TMAX 2960 2200 Recovered
____________________
Note: All times are in seconds.
15.6-31 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-1c TIME SEQUENCE OF EVENTS FOR SMALL BREAK LOCA ANALYSIS UNIT 1 REDUCED TAVG Event Time (sec) 1.5 Inch 2 Inch 3 Inch 4 Inch Break Initiation 0 0 0 0 Reactor Trip 79.0 41.8 17.7 10.2 Signal Safety Injection 123.3 65.2 27.1 14.1 Signal Safety Injection 163.3 105.2 67.1 54.1 Delivered Top of Core 16040 2268 1032 731.1 Uncovered Accumulator N/A N/A 2119 991 Injection Begins Peak Clad 22894.8 3455.5 2013.5 1099.7 Temperature Occurs Top of Core >TMAX >TMAX 2955 2150 Recovered
____________________
Note: All times are in seconds.
15.6-31a REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-1d TIME SEQUENCE OF EVENTS FOR SMALL BREAK LOCA ANALYSIS UNIT 2 NOMINAL TAVG Event Time (sec) 1.5 Inch 2 Inch 3 Inch 4 Inch Break Initiation 0 0 0 0 Reactor Trip 142.1 80.2 59.0 24.6 Signal Safety Injection 154.6 91.8 71.6 36.4 Signal Safety Injection 194.6 131.8 111.6 76.4 Delivered Top of Core 9810 1809.3 771 510 Uncovered Accumulator N/A N/A 1732 990.8 Injection Begins Peak Clad 16234.8 2804.5 1618.5 889.0 Temperature Occurs Top of Core 37098 4740 2754 2378 Recovered
____________________
Note: All times are in seconds.
15.6-31b REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-1e TIME SEQUENCE OF EVENTS FOR SMALL BREAK LOCA ANALYSIS UNIT 2 REDUCED TAVG Event Time (sec) 1.5 Inch 2 Inch 3 Inch 4 Inch Break Initiation 0 0 0 0 Reactor Trip 77.7 41.3 17.6 10.2 Signal Safety Injection 117.8 59.6 27.2 14.3 Signal Safety Injection 157.8 67.2 96.6 54.3 Delivered Top of Core 10750 1930 717.9 614.5 Uncovered Accumulator N/A N/A 1928.1 882.6 Injection Begins Peak Clad 17491.8 3071.7 1805.6 992.3 Temperature Occurs Top of Core 36950 5543 2826 2103 Recovered
____________________
Note: All times are in seconds.
15.6-31c REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-2 INPUT PARAMETERS USED IN THE SBLOCA ANALYSIS(c)
Analyzed Core Power(a), (MWt) 3659 Total peaking factor, FQ 2.60 Axial peaking factor, FZ (FQ/FH) 1.53 Power shape See Figure 15.6-13c Accumulator water volume, nominal 950 (ft3/accumulator)
Accumulator tank volume, nominal 1350 (ft3/accumulator)
Accumulator gas pressure, minimum (psia) 600 Accumulator water temperature (F) 130 Safety injection pumped flow See Figures 15.6-13a and 15.6-13b Pressurizer pressure reactor trip signal (psia) 1857 Pressurizer pressure SI 1715 signal (psia) 15.6-32 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-2 (Cont'd)
INPUT PARAMETERS USED IN THE SBLOCA ANALYSIS(c)
Initial vessel flowrate (lb/sec)
Reduced Tavg 39,416 (368,000 gpm)
Nominal Tavg 37,595 (368,000 gpm)
Vessel Average Temperature (F)(b)
Reduced Tavg 565 Nominal Tavg 598.0 Average reactor coolant pressure (psia) 2300 Steam generator pressure (psia)
SBLOCA (Nominal Tavg) 1108 (Unit 1) 1012 (Unit 2)
SBLOCA (Reduced Tavg) 824 (Unit 1) 727 (Unit 2)
Steam generator tube plugging level (%)
SBLOCA 5 (Unit 1) 10 (Unit 2)
____________________
(a) Two percent is included in this power to account for calorimetric error. Therefore, the analyzed core power is bounding for the MUR power uprate up to a total core power(licensed core power plus calorimetric uncertainties) of 3659 MWt.
(b) The vessel average temperature used in the LOCA analyses is based on the Tavg window of 575.0 F through 588.0 F 10 F in uncertainties.
(c) The analysis is applicable for VANTAGE 5/VANTAGE+ fuel by accounting for the following features: IFMs, 100 and 200 psig IFBA, 275 psig Non-IFBA, Zircaloy-4 and ZIRLO cladding, and solid/mid-enriched annular blankets.
15.6-32a REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-2a KEY LBLOCA PARAMETERS AND INITIAL TRANSIENT ASSUMPTIONS Parameter Initial Transient Uncertainty or Bias 1.0 Plant Physical Description
- a. Dimensions Nominal -
- b. Flow resistance Nominal Sample**
- c. Pressurizer location On Intact loop Bounded
- d. Hot assembly location Under limiting location Bounded
- e. Hot assembly type 17x17 Vantage+ w/ZIRLOTM clad -
- f. SG tube plugging level High (U1 5%, U2 10% ****) Bounded*
2.0 Plant Initial Operating Conditions 2.1 Reactor Power
- a. Core average linear heat Nominal - Based on 3658.33 MWt Sample**
rate (AFLUX) (102% of 3586.6 MWt)
- b. Hot Rod Peak linear heat Derived from desired Tech Spec Sample**
rate (PLHR) (TS) limit FQ = 2.60 and maximum baseload FQ
- c. Hot rod average linear Derived from TS FH = 1.70 Sample**
heat rate (HRFLUX)
- d. Hot assembly average heat HRFLUX/1.04 Sample**
rate (HAFLUX)
- e. Hot assembly peak heat PLHR/1.04 Sample**
rate (HAPHR)
- f. Axial power distribution Figure 15.6-7 Sample**
(PBOT, PMID)
- g. Low Power region relative High (0.62) Bounded*
power (PLOW)
- h. Hot assembly burnup 2000 MWD/MTU Sample**
- i. Prior operating history Equilibrium decay heat Bounded
- j. Moderator Temperature Tech Spec Maximum (0)
Coefficient (MTC) Bounded
- Confirmatory Sequence Parameter 15.6-32b REVISON 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.6-2a (Cont)
KEY LBLOCA PARAMETERS AND INITIAL TRANSIENT ASSUMPTIONS Parameter Initial Transient Uncertainty or Bias 2.0 Plant Initial Operating Conditions (cont.)
2.2 Fluid Conditions
- a. TAVG High Nominal (588.0oF***) Bounded*, Sample**
- b. Pressurizer pressure Nominal (2250.0 psia) Sample**
- c. Loop flow 92,000 gpm Bounded
- d. TUH TCOLD 0
- e. Pressurizer level Nominal (60% span) 0
- f. Accumulator temperature Nominal (95oF) Sample**
- g. Accumulator pressure Nominal 639.5 psia) Sample**
- h. Accumulator liquid Nominal (950 ft3) Sample**
volume
- i. Accumulator line Nominal Sample**
resistance
- j. Accumulator boron Minimum (2150 ppm) Bounded 3.0 Accident Boundary Conditions
- a. Break location Cold leg Bounded
- b. Break type Guillotine (DEGCL) Sample**
- c. Break Size Nominal (cold leg area) Sample**
- d. Offsite Power Unavailable (RCS pumps Bounded*
de-energized at break initiation) 15.6-32c REVISION 16 - DECEMBER 2016
B/B - UFSAR TABLE 15.6-2a (Cont)
KEY LBLOCA PARAMETERS AND INITIAL TRANSIENT ASSUMPTIONS Parameter Initial Transient Uncertainty or Bias 3.0 Accident Boundary Conditions (Cont.)
- e. Safety injection flow Minimum Bounded
- f. Safety injection temperature Nominal (76oF) Sample**
- g. Safety injection delay Max delay (40.0 sec Loop Initial Bounded Transient)
- h. Containment pressure Bounded - See Section 6.2.1.5 Bounded
- i. Single failure ECCS: Loss of 1 SI train. Bounded Containment pressure: all trains operating
- j. Control rod drop time No control rods Bounded 4.0 Model Parameters
- a. Critical flow Nominal (CD = 1.0) Sample**
- b. Resistance uncertainties in Nominal (as coded) Sample**
broken loop
- c. Initial stored energy/fuel Nominal (as coded) Sample**
rod behavior
- d. Core heat transfer Nominal (as coded) Sample**
- e. Delivery and bypassing of ECC Nominal (as coded) Conservative
- f. Steam binding/entrainment Nominal (as coded) Conservative
- g. Non-condensable Nominal (as coded) Conservative bases/accumulator nitrogen
- h. Condensation Nominal (as coded) Sample**
Notes:
- Bounded by Confirmatory Analysis for Each Plant
- Sampling distribution defined in Table 15.6-3b
15.6-32d REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.6-3 LBLOCA CONFIRMATORY CASES PCT RESULTS
SUMMARY
Results from WCOBRA/TRAC Confirmatory Studies for Byron/Braidwood Unit 1 Transient Description PCT (oF)
Initial Transient (High Tavg, High SGTP, High PLOW, LOOP) 1660 SGTP Confirmatory Transient (High Tavg, Low SGTP, High PLOW, LOOP) 1616 PLOW Confirmatory Transient (High Tavg, High SGTP, Low PLOW, LOOP) 1601 No-LOOP Confirmatory Transient (High Tavg, High SGTP, High PLOW, 1412 no-LOOP)
Tavg Confirmatory Transient (Low Tavg, High SGTP, High PLOW, LOOP) 1451 Reference Transient (High Tavg, High SGTP, High PLOW, LOOP) 1660 Results from WCOBRA/TRAC Confirmatory Studies for Byron/Braidwood Unit 2 Transient Description PCT (oF)
Initial Transient (High Tavg, High SGTP, High PLOW, LOOP) 1578 SGTP Confirmatory Transient (High Tavg, Low SGTP, High PLOW, LOOP) 1543 PLOW Confirmatory Transient (High Tavg, High SGTP, Low PLOW, LOOP) 1609 No-LOOP Confirmatory Transient (High Tavg, High SGTP, High PLOW, 1485 no-LOOP)
Tavg Confirmatory Transient (Low Tavg, High SGTP, High PLOW, LOOP) 1509 Reference Transient (High Tavg, High SGTP, Low PLOW, LOOP) 1609 15.6-33 REVISION 14 - DECEMBER 2012
B/B-UFSAR THIS PAGE WAS INTENTIONALLY DELETED 15.6-33a REVISION 14 - DECEMBER 2012
B/B - UFSAR TABLE 15.6-3a OVERALL LARGE BREAK LOCA RESULTS FOR THE BYRON/BRAIDWOOD NUCLEAR STATIONS Result for Byron/Braidwood Unit 1 IFBA NON-IFBA 95/95 PCT (oF) 1912 1913 95/95 LMO (%) 5.51 5.51 95/95 CWO (%) N/A 0.25 Result for Byron/Braidwood Unit 2 IFBA NON-IFBA 95/95 PCT (oF) 2036 2041 95/95 LMO (%) 7.02 8.27.51 95/95 CWO (%) N/A 0.33 PCT = Peak Cladding Temperature LMO = Local Maximum Oxidation CWO = Core Wide Oxidation 15.6-33b REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE 15.6-3b PLANT OPERATING RANGE ALLOWED BY THE BEST ESTIMATE LARGE BREAK LOCA ANALYSIS (BYRON/BRAIDWOOD NUCLEAR STATIONS)
Parameter As-Analyzed Value or Range 1.0 Plant Physical Description a) Dimensions Nominal b) Pressurizer location On an intact loop(3) c) Hot assembly location Anywhere in core(1) d) Hot assembly type(2) 17X17 Vantage+ fuel with, ZIRLOTM cladding, non-IFBA or IFBA(4) e) Steam generator tube plugging level 5% (U1) or 5%(5) (U2) f) Fuel assembly type(2) 17X17 Vantage+ fuel with, ZIRLOTM cladding, non-IFBA or IFBA(4) 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Core power 3658.3 MWt (including calorimetric uncertainties) b) Peak heat flux hot channel factor (FQ)(2) 2.6 (2) c) Peak hot rod enthalpy rise hot channel factor (FH) 1.70 d) Hot assembly radial peaking factor (PHA)(2) 1.70/1.04 e) Hot assembly heat flux hot channel factor (FQHA) 2.6/1.04 (2) f) Axial power dist (PBOT, PMID) Figure 15.6-7 g) Low power region relative power (PLOW)(2) 0.35 PLOW 0.62 h) Hot assembly burnup 75,000 MWD/MTU, lead rod(1) i) MTC 0 at hot full power (HFP) j) Typical cycle length 18 months (2) k) Minimum core average burnup 10,000 MWD/MTU l) Maximum steady state depletion, FQ(2) 2.1 15.6-33c REVISION 16 - DECEMBER 2016
B/B-UFSAR TABLE 15.6-3b (Cont)
PLANT OPERATING RANGE ALLOWED BY THE BEST ESTIMATE LARGE BREAK LOCA ANALYSIS (BYRON/BRAIDWOOD NUCLEAR STATIONS)
Parameter As-Analyzed Value or Range 2.2 Fluid Conditions a) TAVG 575.0 -10ºF TAVG 588.0 (U1) or 583.5 (U2)(6) +10ºF b) Pressurizer pressure 2250 - 43 psia PRCS 2250 + 43 psia c) Loop flow TDF 92,000 gpm/loop d) Upper head design TCOLD e) Pressurizer level 60.0% of span f) Accumulator temperature 60ºF TACC 130ºF g) Accumulator pressure 587 psia PACC 692 psia h) Accumulator liquid volume 920 ft3 VACC 980 ft3 i) Accumulator fL/D Current Line Configuration j) Minimum accumulator boron 2150 ppm 3.0 Accident Boundary Conditions a) Minimum safety injection flow Table 15.6-3d b) Safety injection temperature 32ºF SI Temp 120ºF c) Safety injection delay 27 seconds (with offsite power) 40 seconds (with LOOP) d) Containment modeling See Section 6.2.1.5 e) Minimum containment pressure 14.2 psia f) Containment spray initiation delay 0 seconds g) Recirculation spray initiation delay Not Applicable h) Single failure Loss of one SI train Notes:
- 1. 44 peripheral locations will not physically be lead power assembly.
- 2. In the Westinghouse Reload Safety Analysis Checklist (RSAC) process, this parameter is identified as a key safety analysis parameter that could be impacted by a fuel reload.
- 3. Analyzing the pressurizer as being located on an intact loop is limiting per Westinghouse methodology.
- 4. Analysis models thimble plugs removed, which is judged to bound plugs installed. Any combination of thimble plugs installed/removed is supported.
15.6-33d REVISION 16 - DECEMBER 2016
BYRON-UFSAR THIS PAGE HAS BEEN INTENTIONALLY DELETED.
15.6-33e REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE 15.6-3c THIS TABLE HAS BEEN INTENTIONALLY DELETED.
15.6-33f REVISION 11 - DECEMBER 2006
BYRON-UFSAR TABLE 15.6-3d TOTAL MINIMUM INJECTED SAFETY INJECTION FLOW USED IN BEST-ESTIMATE LARGE-BREAK LOCA ANALYSIS FOR BYRON/BRAIDWOOD Low Head and Low Head and Intermediate Head Intermediate Head Total Safety Total Safety High Head Injection - LHSI Injection - LHSI RCS Pressure Safety Injection Miniflow Closed* Miniflow Open*
(psia) Flow (ft3/sec) (ft3/sec) (ft3/sec) 14.7 0.671 6.620 5.15 34.7 0.666 6.091 4.621 54.7 0.661 5.348 3.878 74.7 0.657 4.169 2.699 94.7 0.652 2.682 1.212 106.9 0.648 0.867 0.867 120.0 0.644 0.860 0.860 214.7 0.617 0.812 0.812 314.7 0.592 0.760 0.760 414.7 0.567 0.707 0.707 414.8 0.000 0.000 0.000 Note:
- LHSI Miniflow valve is conservatively modeled to close 15 seconds following the SI delay signal.
15.6-33g REVISION 14 - DECEMBER 2012
B/BUFSAR TABLE 15.6-3e THIS TABLE HAS BEEN INTENTIONALLY DELETED.
15.6-33h REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE 15.6-3f THIS TABLE HAS BEEN INTENTIONALLY DELETED.
15.6-33i REVISION 14 - DECEMBER 2012
B/B-UFSAR TABLE 15.6-4 SMALL BREAK LOCA RESULTS FOR ZIRLO FUEL CLADDING - NOTRUMP EM-UNIT 1 NOMINAL TAVG Results 1.5 Inch 2 Inch 3 Inch 4 Inch Peak Clad Temperature 765 1570 1514 1428 (oF)
Peak Clad Temperature 22680.6 3434.5 1834.7 1019.1 Times (s)
Peak Clad Temperature 11.25 11.75 11.5 11.25 Elevation (ft)
Max. Local ZrO2 (%) 0.01 1.17 0.89 0.29 Max. Local ZrO2 11.25 11.75 11.50 11.25 Elevation (ft)
Core-Wide Avg. ZrO2 (%) 0.00 0.17 0.14 0.05 Burst Time(s) N/A N/A N/A N/A Burst Elevation (ft) N/A N/A N/A N/A 15.6-34 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-4a SMALL BREAK LOCA RESULTS FOR ZIRLO FUEL CLADDING - NOTRUMP EM-UNIT 1 REDUCED TAVG Results 1.5 Inch 2 Inch 3 Inch 4 Inch Peak Clad Temperature 731 1624 1457 1292 (oF)
Peak Clad Temperature 22894.8 3455.5 2013.5 1099.7 Times (s)
Peak Clad Temperature 11.50 11.50 11.5 11.25 Elevation (ft)
Max. Local ZrO2 (%) 0.00 1.47 0.62 0.11 Max. Local ZrO2 11.50 11.50 11.50 11.25 Elevation (ft)
Core-Wide Avg. ZrO2 (%) 0.00 0.22 0.10 0.02 Burst Time(s) N/A N/A N/A N/A Burst Elevation (ft) N/A N/A N/A N/A 15.6-34a REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-4b SMALL BREAK LOCA RESULTS FOR ZIRLO FUEL CLADDING - NORTUMP EM-UNIT 2 NOMINAL TAVG Results 1.5 Inch 2 Inch 3 Inch 4 Inch Peak Clad Temperature (oF) 912 1086 1614 1537 Peak Clad Temperature 16234.8 2804.5 1618.5 889.0 Time (s)
Peak Clad Temperature 11.25 11.25 11.50 11.25 Elevation (ft)
Max. Local ZrO2 (%) 0.03 0.06 1.48 0.65 Max. Local ZrO2 Elevation 11.25 11.25 11.50 11.25 (ft)
Core-Wide Avg. ZrO2 (%) 0.00 0.01 0.23 0.11 Burst Time(s) N/A N/A N/A N/A Burst Elevation (ft) N/A N/A N/A N/A 15.6-34b REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-4c SMALL BREAK LOCA RESULTS FOR ZIRLO FUEL CLADDING - NORTUMP EM-UNIT 2 REDUCED TAVG Results 1.5 Inch 2 Inch 3 Inch 4 Inch Peak Clad Temperature (oF) 874 1627 1452 1313 Peak Clad Temperature 17491.8 3071.7 1805.6 992.3 Time (s)
Peak Clad Temperature 11.00 11.50 11.50 11.25 Elevation (ft)
Max. Local ZrO2 (%) 0.03 1.59 0.61 0.14 Max. Local ZrO2 Elevation 11.00 11.50 11.50 11.25 (ft)
Core-Wide Avg. ZrO2 (%) 0.00 0.24 0.09 0.02 Burst Time(s) N/A N/A N/A N/A Burst Elevation (ft) N/A N/A N/A N/A 15.6-34c REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-4d SMALL BREAK LOCA RESULTS FOR ZIRCALOY-4 FUEL CLADDING -
NORTRUMP EM - UNIT 2 REDUCED TAVG Results CAE/CCE CBE/CDE CBE/CDE REDUCED Tavg NOMNIAL Tavg NOMINAL Tavg 2 Inch 3 Inch BOL 3 Inch BU=6K MWD/MTU Peak Clad Temperature (oF) 1601 1615 1601 Peak Clad Temperature 3494.5 1618.5 1624.7 Time (s)
Peak Clad Temperature 11.75 11.50 11.75 Elevation (ft)
Max. Local ZrO2 (%) 1.30 1.5 1.48 Max. Local ZrO2 Elevation 11.75 11.5 11.50 (ft)
Core-Wide Avg. ZrO2 (%) 0.19 0.23 0.45 Burst Time(s) N/A N/A N/A Burst Elevation (ft) N/A N/A N/A 15.6-34d REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-5 SGTR INITIAL CONDITIONS VALUE (Typical) VALUE PARAMETER Offsite Dose Case MTO Case NSSS Power 3672 Mw 3672 Mw RCS Flow 368,000 gpm 368,000 gpm Standard Thermal Standard Thermal Design Design RCS Pressure 2192 psig 2192 psig RCS Tavg 588.0F 580.0F (Unit 1) 575.0F (Unit 2)
RCS Initial Water Mass 5.48E5 lbm (Unit 1) Not 4.88E5 lbm (Unit 2) Calculated Percent Tube Plugging 0% 5% (Unit 1)
(per steam generator) 10% (Unit 2)
S/G Initial Mass 1.37E5 lbm (Unit 1) 1.36E5 lbm (Unit 1) 1.26E5 lbm (Unit 2) 1.22E5 lbm (Unit 2)
Break Location First Row Cold Leg First Row Cold Leg at Tubesheet at Tubesheet (Shortest Tube) (Shortest Tube) 15.6-35 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-5 (Cont'd)
VALUE (Typical) VALUE PARAMETER Offsite Dose Case MTO Case Decay Heat 1.2 X ANS 5.1 1971 ANS 5.1 1979-2sigma AFW Flow 302 gpm per S/G 180 gpm per SG for 40 seconds, 360 gpm/SG after 40 seconds (Unit 1).
263 gpm per SG for 40 seconds, 450.22 gpm/SG after 40 seconds (Unit 2).
AFW Temperature 120F 32F 15.6-35a REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-5A SGTR OPERATOR ACTIONS VALUE VALUE OPERATOR ACTION Offsite Dose Case MTO Case Isolate AFW Flow to the 5 Minutes* 9 Minutes Ruptured Steam Generator After Tube Rupture Occurs After Tube Rupture Occurs Provided Level is Adequate Isolate MSIV for the 18 Minutes 18 Minutes Ruptured Steam Generator After Tube Rupture Occurs After Tube Rupture Occurs Isolate the Failed Open PORV one the 30 Minutes ----------
Ruptured Steam Generator After PORV Fails Open Initiate RCS Cooldown 3 Minutes 3 Minutes After After Failed PORV is Isolated MSIV Isolation Initiate RCS Depressurization 4 Minutes 4 minutes After Cooldown After Cooldown is Completed is Completed Terminate ECCS Flow 3 Minutes 3 Minutes After Depressurization After Depressurization is Completed is Completed
- The assumption of a minimum of 5 minutes from event initiation until AFW isolation used in the input to dose analyses is not a critical operator action time and does not impose a requirement on the operators.
15.6-36 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-6 INPUT PARAMETERS FOR THE SGTR RADIOLOGICAL CONSEQUENCE ANALYSIS USING AST Parameter Unit Value Accident mitigation recovery phases associated with the sequence of events:
Start of event until Reactor trip seconds 0 - 160 Reactor Trip until ruptured SG PORV fails seconds 160 - 1082 open Ruptured SG PORV failure until 15 minutes seconds 1082 - 1982 later 15 minutes after Ruptured SG PORV failure seconds 1982 - 2882 until failed ruptured SG PORV is isolated Ruptured SG PORV Isolation until RCS seconds 2882 - 3062 cooldown is initiated RCS cooldown initiation until Break flow seconds 3062 - 3390 flashing ceases End of Break flow flashing until RCS seconds 3390 - 5164 depressurization initiation RCS depressurization initiation until all seconds 5164 - 6234 Break Flow ceases Iodine Spike Release Period hours 8 Full power steam release through the Condenser until the time of reactor trip and loss of offsite power Ruptured SG lbm 1.830E5 lbm 5.445E5 Intact SGs Primary and secondary coolant masses:
Ruptured SG lbm 6.89E4 Intact SGs lbm 2.06E5 RCS lbm 4.88E5 15.6-37 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-6a SEQUENCE OF EVENTS FOR SGTR (TYPICAL OFFSITE DOSE CASE)
Unit 1 Unit 2 SYSTEM RESPONSE/OPERATOR ACTION Time (sec) Time (sec)
SG Tube Rupture Occurs 0 0 Reactor Trip 217 160 AFW Actuated 280 223 AFW Flow to the Ruptured SG Isolated 300 300 SI Actuated 387 305 Ruptured SG MSIV Closed 1080 1080 Ruptured SG PORV Fails Open 1082 1082 Ruptured SG PORV Isolated 2882 2882 RCS Cooldown Initiated 3062 3062 Break Flow Flashing Terminated 3304 3390 RCS Cooldown Terminated 4284 4922 RCS Depressurization Initiated 4526 5164 RCS Depressurization Terminated 4618 5250 SI Flow Terminated 4798 5430 Break Flow Terminated 5478 6234 15.6-37a REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.6-6b SEQUENCE OF EVENTS FOR SGTR (MARGIN TO OVERFILL CASE)
Unit 1 Unit 2 SYSTEM RESPONSE/OPERATOR ACTION Time (sec) Time (sec)
SG Tube Rupture Occurs 0 0 Reactor Trip 200 139 AFW Actuated 201 140 SI Actuated 474 317 AFW Flow to the Ruptured SG Isolated 540 540 Ruptured SG MSIV Closed 1080 1080 RCS Cooldown Initiated 1260 1260 RCS Cooldown Terminated 1838 1958 RCS Depressurization Initiated 2080 2200 RCS Depressurization Terminated 2178 2302 SI Flow Terminated 2359 2482 Break Flow Terminated 3360 3258 15.6-37b REVISION 15 - DECEMBER 2014
B/B-UFSAR This page has been intentionally deleted.
15.6-37c REVISION 9 - DECEMBER 2002
B/B UFSAR TABLE 15.6-7 INPUT PARAMETERS FOR THE LOCA RADIOLOGICAL CONSEQUENCES ANALYSIS USING AST Parameter Unit Value Notes Core Release Fraction: Per Table 2 in Iodine - RG 1.183 Noble Gases -
Iodine Plate-out Fraction on - Time- Powers natural deposition model in RADTRAD Containment Surfaces dependent Initial Iodine Species in Containment Aerosol % 95 Elemental % 4.85 Organic % 0.15 Iodine released from ECCS to the environment:
Elemental % 97 Organic % 3 Containment Leak Rate: The reduction in the design basis containment 0-24 Hours weight 0.20 leak rate by 50% at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is consistent with
> 24 Hours %/day 0.10 guidance of RG 1.183.
Containment Spray (CS) Flow seconds 90 Biased high for conservatism based on 88.1 or Actuation Time Including Delay 53.1+X* where X* denotes the time at which the containment pressure setpoint is reached (High-3 for CS).
Spray Injection Flow Duration minutes 20.9 This value is based on containment spray switchover at lo-3 RWST level, maximum SI, and one containment spray pump running (because the analysis assumes the failure of one containment spray pump). The duration is from spray initiation.
Containment Spray Parameters:
Containment Spray Flow The minimum amount of delivered flow based Injection gpm 2950 on the minimum delivered flow for 1 spray Recirculation gpm 2950 pump is 3113 gpm. For conservatism, a value of approximately 95 % is used.
The value is based on the most conservative Fraction of Containment Sprayed - 0.825 approach of minimum net sprayed containment volume and maximum net containment volume.
Since the same flow rate is used during injection and recirculation phases, this value is also valid for both phases.
Containment Volume ft3 2.85E6 Sprayed ft3 2.35E6 Unsprayed ft3 5.0E5 This value bounds the minimum height between Containment Spray Fall Height ft 141 the spray ring header (elevation 567 feet), and the operating floor (elevation 426 feet).
15.6-38 REVISION 17 DECEMBER 2018
B/B UFSAR TABLE 15.6-7 (Contd)
INPUT PARAMETERS FOR THE LOCA RADIOLOGICAL CONSEQUENCES ANALYSIS USING AST Parameter Unit Value Notes Conservatively greater than maximum RWST Containment Spray Temperature F 120 temperature in Technical Specification 3.5.4.
Nominal (norm. op.)
Containment Temperature Maximum during LOCA F 260 Number of Deck Fans Operating # 2 of 4 Deck Fan Flow Rate, per fan cfm 65,000 Based on minimum heat removal and service water temperature of 100 ºF.
Time from Spray Cessation Before min 10 Spray Recirculation Starts ECCS Recirculation Leakage cc/hr 276,000 Recirculation Loop Water Volume at ft3 38,979 This value is considered to be the minimum 11.6 minutes containment sump volume at ECCS switchover.
The proposed value assumes the RCS volume for unit 2, which bounds unit 1 and takes into account maximum steam generator tube plugging level. At CS switchover, an additional 19,527 ft3 should be added, for a total of 58,506 ft3.
Fraction of core iodine in sump - 0.4 solution Iodine partition coefficient for - 0.1 recirculation leakage Sump Water Temperature ºF 260 Auxiliary Building Release Path Flashed fluid Iodine is assumed to contain no Iodine Removal Efficiency: aerosol.
Elemental iodine % 90 Organic iodine % 90 Particulates % 99 Passive failure release from the ECCS - N/A Since credit is taken for safety grade filtration recirculation path system in the auxiliary building for ECCS leakage, the passive failure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is not applicable.
Offsite Dose Acceptance Criteria Analysis Release Duration days 30 15.6-39 REVISION 12 DECEMBER 2008
B/B-UFSAR TABLE 15.6-8 This table has been intentionally deleted.
15.6-40 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-9 This page has been intentionally deleted.
15.6-41 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-9 (Cont'd)
This page has been intentionally deleted.
15.6-42 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-10 This page has been intentionally deleted.
15.6-43 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-11 This page has been intentionally deleted.
15.6-44 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-12 THIS TABLE HAS BEEN INTENTIONALLY DELETED 15.6-45 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.6-13 RECIRCULATION LOOP LEAKAGE EXTERNAL TO CONTAINMENT TYPE OF LEAKAGE CONTROL AND LEAKAGE TO LEAKAGE TO UNIT LEAKAGE RATE USED IN ATMOSPHERE VIA LOCAL ATMOSPHERE VIA VENTED ITEMS THE ORIGINAL ANALYSIS LEAKAGE (cc/hr) DRAIN TANK (cc/hr)
- 1. Residual heat removal Mechanical seal 10 cc/hr/seal*(Note 1) 20 0 (low head safety injection) (while pump is in operation)
- 2. Safety injection pumps Same as residual heat removal pump 40 0
- 3. Charging Pumps Same as residual heat removal pump 40 0
- 4. Flanges:
- a. Pumps Gasket - adjusted to zero leakage 0 0 following any test 10 drops/min/gauge
- b. Valves bonnet to body used (30 cc/hr). Due to leaktight 2400 0 (larger than 2 inches) flanges on pumps, no leakage is assumed to atmosphere
- c. Control valves 480 0
- d. Heat exchangers 240 0
- 5. Valves - stem leakoffs Backed seated double packing with 0 50 leakoff- 1 cc/hr/in. of steam diameter used
- 6. Miscellaneous small valves Flanged boyd packed stems - 600 0 1 drop/min used (3 cc/hr)
- 7. Miscellaneous large valves Double packing 1 cc/hr/in. stem 40 0 (Larger than 2 inches) diameter used 3860 50 Total - 3910 cc/hr (Note 2)
- Note 1 During equipment shutdown, mechanical seals may develop minor leakage. However, this leakage typically stops when the pump is returned to operation.
Note 2 The maximum permitted recirculation loop leakage is 138,000 cc/hr. The design basis accident analysis assumes a value two times the maximum permitted recirculation loop leakage (i.e., 276,000 cc/hr) in accordance with the Standard Review Plan.
15.6-46 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.6-14 15.6-47 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.6-15 PEAK CLAD TEMPERATURE INCLUDING ALL PENALTIES AND BENEFITS PEAK CLAD TEMPERATURE ANALYSIS (F)
Large Break LOCA Byron Unit 1 2025 Byron Unit 2 2047 Braidwood Unit 1 2025 Braidwood Unit 2 2047 Small Break LOCA Byron Unit 1 1749 Byron Unit 2 1755 Braidwood Unit 1 1749 Braidwood Unit 2 1755 15.6-48 REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Events which may result in a radioactive release from a subsystem or component are as follows:
- a. waste gas system leak or failure (Subsection 15.7.1),
- b. liquid waste system leak or failure - release to atmosphere (Subsection 15.7.2),
- c. liquid tank failure - ground release (Subsection 15.7.3),
- d. fuel handling accidents (Subsection 15.7.4),
- e. spent fuel cask drop accidents (Subsection 15.7.5).
15.7.1 Gas Waste System Leak or Failure 15.7.1.1 Identification of Causes and Accident Description The gaseous waste processing system, as discussed in Chapter 11.0, is designed to remove fission product gases from the reactor coolant. The system consists of a closed loop with waste gas compressors, hydrogen recombiners, waste gas decay tanks for service at power, and other waste gas decay tanks for service at shutdown and startup.
The accident is defined as the postulated uncontrolled release of the xenon and krypton fission product gases stored in a waste decay tank as a consequence of a failure of a single gas decay tank or associated piping.
15.7.1.2 Analysis of Effects and Consequences The maximum noble gas activities in the gas decay tank assumed to fail, the assumptions on which the activities are based, and the offsite dose consequences of releasing the activity are discussed in Subsection 15.7.1.3.
15.7.1.3 Radiological Consequences of a Postulated Waste Gas Decay Tank Rupture The analysis of the postulated waste gas decay tank rupture is performed based on Regulatory Guide 1.24 (May 1972). The parameters used for this analysis are listed in Table 15.7-1.
The assumptions for the Regulatory Guide 1.24 analysis are:
- a. The reactor has been operating at full power with 1%
defective fuel and a shutdown to cold condition has been conducted near the end of an equilibrium core cycle. As soon as possible after shutdown, all noble 15.7-1 REVISION 9 - DECEMBER 2002
B/B-UFSAR gases have been removed from the reactor coolant system and transferred to the gas decay tank that is assumed to fail.
The inventory of noble gases in the primary coolant conservatively assumes that the unit has been operated for a full cycle without the removal of fission gases from the primary coolant, thus maximizing the coolant noble gas inventory. This coolant noble gas concentration is provided in Table 15.0-9.
- b. Radiological decay is taken into account in the computation only for the minimum time period required to transfer the gases from the reactor coolant system to the decay tank. The noble gas inventories of the tank are given in Table 15.7-2.
- c. The tank rupture is assumed to occur immediately upon completion of the waste gas transfer, releasing the entire contents of the tank at ground level to the outside atmosphere. The assumption of the release of the noble gas inventory from only a single tank is based on the fact that all gas decay tanks will be isolated from each other whenever they are in use.
- d. The 5th percentile atmospheric, diffusion factors given in Tables 15.0-13 and 15.0-14 are used to evaluate the doses from the released activity. Doses are based on the dose models presented in Attachment 15A.
The doses at the exclusion area boundary and the LPZ resulting from this accident are listed in Tables 15.0-11 and 15.0-12. These doses are well within 10 CFR 100 limits.
15.7.2 Radioactive Liquid Waste System Leak or Failure (Atmospheric Release) 15.7.2.1 Identification of Causes and Frequency Classification An unspecified event causes the complete release of the worst case radionuclide inventory in the tanks containing the largest quantities of significant radionuclides in the liquid radwaste system. These are the spent resin storage tank and the boron recycle holdup tanks in the auxiliary building. The airborne radioactivity release during the accident passes directly to the environment via the station vent stack.
Postulated events that could cause release of the radioactive inventory of the spent resin storage tank and the boron recycle 15.7-2 REVISION 9 - DECEMBER 2002
B/B-UFSAR holdup tanks are cracks in the tanks and operator error. The possibility of small cracks and consequent low-level release rates receives primary consideration in system and component design. The spent resin storage tank is designed to operate at a pressure of 125 psi and 150F maximum temperature. The boron recycle holdup tank is designed for a pressure of 15 psig and a temperature of 200F. The possibility of failure of either tank is considered small. A radioactive liquid release from either of these tanks caused by operator error is also considered a remote possibility. Operating techniques and administrative procedures emphasize detailed system and equipment operating instructions.
Should a release of radioactive liquid occur, floor drain sump pumps in the floor of the auxiliary building will receive a high water level alarm, activate automatically, and remove the spilled liquid. This accident is expected to occur with the frequency of a limiting fault.
The probability of a complete rupture of the radwaste system or a complete malfunction accident is considered to be much lower than the rupture of a single tank, as described above. This accident is not analyzed.
Although not built to the standards of Seismic Category I equipment, the liquid radwaste system tanks are constructed in accordance with sound engineering principles and the ASME Code.
Therefore, simultaneous failure of all the tanks is not considered credible.
15.7.2.2 Sequence of Events and Systems Operation The sequence of events expected to occur are as follows; Sequence of Events Elapsed Time
- 1. Event begins--failure 0 occurs
- 2. Area radiation alarms 1 minute alert plant personnel
- 3. Operator actions begin 5 minutes The rupture of the spent resin storage tank or the boron recycle holdup tank would leave little recourse to the operator. No method of recontaining the gaseous phase discharge is available, however, isolation of the auxiliary building would minimize the results. High radiation alarms both in the ventilation exhaust and in the auxiliary building would alert the operator to the failure. No credit is taken for operator action to isolate the auxiliary building ventilation system in evaluating these events.
15.7-3 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.7.2.3 Modeling of Accident Sequence 15.7.2.3.1 Mathematical Models
- a. Spent Resin Storage Tank The release from the station vent stack is considered a ground level release with accident atmospheric dilution factors calculated by the methods discussed in Reference 1. Doses are calculated at the exclusion area boundary and the LPZ boundary using the fifth percentile /Q. Values are given in Tables 15.0-13 and 15.0-14.
The atmospheric dilution factor is assumed constant throughout the duration of the accident. Doses are calculated using the models of Attachment 15A.
- b. Boron Recycle Holdup Tank The doses due to a Boron Recycle Tank failure were calculated using the Alternate Source Term methodology using Regulatory Guide 1.183 with dose acceptance criteria per 10 CFR 50.67. The RADTRAD computer program was used to model the associated radiological releases.
15.7.2.3.2 Input Parameters and Initial Conditions The tank failures are evaluated in accordance with the following sets of assumptions and conditions:
- a. Spent Resin Storage Tank
- 1. One hundred percent of the liquid volume of a spent resin storage tank is released into the spent resin storage tank cubicle.
- 2. The radionuclide inventory is based on a primary coolant concentration of 1.0 µCi/g of Dose Equivalent I-131 (see Table 15.0-10) and letdown demineralizer resins. The tank is assumed to be filled to 80% of capacity with resin.
- 3. The partition factor for radioiodine from water to air is 0.1 and 0.01 for resin to water.
- 4. The airborne radioactivity release into the spent resin storage tank cubicle passes to the environment via the station vent stack.
- 5. The accident (/Q) values are for a ground level release as described in Subsection 15.7.2.3.1 and presented in Tables 15.0-13 and 15.0-14.
15.7-4 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- b. Boron Recycle Holdup Tank A rupture of the Boron Recycle Holdup Tank (HUT) could potentially release activity from the following:
Activity in the water stored in the ruptured HUT.
Activity in the water stored in the un-ruptured HUT that could drain via the cross-tie piping that inter-connects the two tanks.
Activity in gas spaces from both HUTs. The gas spaces in both HUTs are connected to a common header.
Activity from one waste gas decay tank. As described in UFSAR Section 11.3.2.5, one waste gas decay tank is normally aligned to provide cover gas for the HUTs.
Activity from the discharge of RCS fluid from a Residual Heat Removal (RHR) suction relief valve.
The following assumptions and inputs are used in the dose analysis:
- 1. The iodine activity in the primary coolant is the equilibrium concentration based on 1.0 Ci/g dose equivalent I-131 as indicated in UFSAR Table 15.0-10.
- 2. The primary coolant noble gas activity is based on 1% failed fuel. Isotope activity concentrations are indicated in UFSAR Table 15.0-9.
- 3. The HUT rupture event is evaluated for two cases, minimum and maximum initial water level.
- 4. The letdown demineralizer is assumed to have a decontamination factor of 10, so that the initial water collected in the HUT has one tenth of the RCS iodine concentration.
- 5. The initial noble gas activity in the combined air space of the two HUTs at minimum initial water volume is calculated by assuming that the entire inventory of noble gas in the RCS is transferred to the HUTs. This noble gas concentration is also assumed for the maximum initial water volume case.
- 6. In addition to the water initially in the HUTs, the calculated dose assumes the ruptured HUT also releases RCS water from a RH suction relief valve discharge for one hour. The RCS water from a RH suction relief valve discharges directly into one of the HUTs. Therefore, no decontamination factor is credited.
15.7-4a REVISION 15 - DECEMBER 2014
B/B-UFSAR
- 7. The noble gas in one waste gas decay tank is postulated to be released to the environment along with the noble gas released from the HUT air space. The noble gas activity is based on a maximum 50,000 curies of dose equivalent Xe-133 in the waste gas decay tank.
- 8. Following the rupture all of the water in both HUTs, the RH suction relief valve water, all of the noble gas in both HUTs, and the gas in one waste gas decay tank is released to the HUT cubicle.
- 10. All of the activity is released to the environment within a 5 minute time period following the HUT rupture.
- 11. The accident /Q values (using finer wind speed categories from Regulatory Guide 1.23 Revision 1) as described in Subsection 2.6.3.3 and presented in Table 15.0-17.
- 12. The Main Control Room ventilation system is assumed to be manually realigned to the emergency mode of operation 30 minutes after the initiation of the accident.
15.7-4b REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.7.2.3.3 Results The conservative assessment of the tank failures leads to consideration of iodine partitioning from a spill which could be drained in a rapid manner, thereby minimizing the release of iodine to the air.
The auxiliary building exhaust is filtered through HEPA filters for which no reduction in released iodine has been assumed.
Because of the design features incorporated in the Byron/Braidwood Stations, i.e., as most of the tanks are each located in vented and drained cells, the failure of a tank will not result in significant increases in in-plant doses.
- a. Spent Resin Storage Tank Table 15.7-3 lists the contained radionuclide activity for the spent resin storage tank. Tables 15.0-11 and 15.0-12 list the doses at the exclusion area boundary and the LPZ boundary for the release of the spent resin storage tank. These doses are well within 10 CFR 100 limits.
- b. Boron Recycle Holdup Tank The minimum initial water level case is bounding for the offsite doses. The maximum initial water level case is bounding for the control room dose. The results and regulatory dose limits are provided in UFSAR Tables 15.0-11 and 15.0-12.
15.7.3 Postulated Radioactive Release Due to Liquid Tank Failure (Ground Release) 15.7.3.1 Identification of Causes of Frequency Classification This accident is defined as an unexpected and uncontrolled postulated rupture of the boron recycle holdup tank. This tank is located in the Seismic Category I auxiliary building at elevation 346.0 feet. Since grade elevation is 400.0 feet, the only way any effluents can be released accidentally is through postulated cracks in the auxiliary building which would allow the contents of the tank to enter groundwater. This accident is expected to occur with the frequency of a limiting fault (see Subsection 15.7.2.1).
15.7.3.2 Sequence of Events and Systems Operation See Subsection 2.4.12.
15.7.3.3 Modeling of Accident Sequence 15.7.3.3.1 Mathematical Model Subsection 2.4.12 gives the dispersion, dilution, and travel times of accidental releases of liquid effluents in surface water.
15.7-5 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.7.3.3.2 Input Parameters and Initial Conditions The tank failure is evaluated in accordance with the following sets of assumptions and conditions:
- a. One hundred percent of the liquid volume of the boron recycle holdup tank is released into the boron recycle holdup tank cubicle.
- b. The liquid enters the groundwater environment through postulated cracks in the auxiliary building.
15.7.3.4 Radiological Consequences UFSAR Section 2.4.12 provides an evaluation of an accidental release of effluents through postulated cracks in the auxiliary building.
The concentrations of any postulated accidental release of radioactive effluents from the boron recycle holdup tank would not exceed 10 CFR 20 limits at the nearest surface water intake.
15.7.4 Fuel Handling Accidents 15.7.4.1 Accident Description The accident is defined as dropping of a spent fuel assembly onto the spent fuel pool floor or onto the core resulting in the postulated rupture of the cladding of all the fuel rods in the assembly despite many administrative controls and physical limitations imposed on fuel handling operations. All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a supervisor.
15.7.4.2 Analysis of Effects and Consequences The reactor is assumed to have been operating at 3658.3 MWt. The assembly inventory assumes that the reactor has been operating at this power level up to the end of the cycle. It is assumed that 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> has passed from shutdown to the beginning of fuel handling operations. The resulting source term in an average fuel assembly is given in Table 15.7-6.
For the fuel handling accident it is conservatively assumed that the damaged fuel assembly has been operating at 1.7 times average assembly power (this is the lead rod peaking factor) and the activity in the damaged assembly is 1.7 times the values in Table 15.7-6. Use of this peaking factor is conservative since it would not occur in more than a small fraction of the rods in the assembly. The gap model discussed in Regulatory Guide 1.25 (May 1972), as modified by the recommendations of NUREG/CR-5009 (Reference 1), is used to determine the activity that is available for release from the damaged fuel. The gap fractions are listed in Table 15.7-7.
15.7-6 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15.7.4.2.1 Fuel Handling Accident Inside Spent Fuel Storage Building The accident is defined as the dropping of a spent fuel assembly in the spent fuel pool resulting in the rupture of the cladding of fuel rods.
15.7.4.2.1.1 Identification of Causes and Frequency Classification The cause of the event can be identified as any mechanical failure or operating error which results in the dropping of a fuel assembly into the refueling pool during its transfer from one position in the pool to another.
The frequency classification, as defined in Regulatory Guide 1.70, can be categorized as one of the limiting faults. This means that it is an occurrence that is not expected to occur but is postulated because its consequences would include the potential for the release of significant amounts of radioactive material.
15.7.4.2.1.2 Sequence of Events and Systems Operation The bounding fuel handling accident analysis assumes that the fuel handling building and containment are not isolated at the time of the event and remain open to the environment during the event. Control room HVAC system filtration is delayed for the first 30 minutes of the event for conservatism.
As indicated in Regulatory Guide 1.183, if the fuel handling building and containment are not isolated at the time of the event, the release of all activity to the environment is assumed to occur over a 2-hour time period. To assure this, building exhaust rates are set artificially high in the analysis.
The bounding exhaust point is the plant vent, with specific atmospheric dispersion factors (/Qs) for the control room or offsite receptor locations. Alternative release points through major openings such as the fuel handling building inner or outer rail bay roll-up doors or the personnel/equipment hatch from the containment would have lower /Qs and would therefore result in lower calculated doses.
15.7-7 REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.7.4.2.1.3 System Operation During handling of recently irradiated fuel (i.e., fuel that has occupied part of a critical core within the previous 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />),
the fuel handling building ventilation system is required to be operable. The normal supply system is designed to provide 21,000 cfm of outside air to the fuel handling building general area.
The exhaust inlets are located at the pool edge. The shortest distance between the exhaust inlets and the inboard isolation valve is 222 feet.
During movement of irradiated fuel in the fuel handling building, the following compensatory measures are required per the Alternative Source Term license amendment:
- a. At least one train of the fuel handling building ventilation system will be available.
- b. At least one fuel handling building trackway door will be maintained closed.
- c. Within one hour of a fuel handling accident, airflow in the proper direction will be established in the fuel handling building.
- d. Containment and fuel handling building effluents will be monitored consistent with the Technical Specification Radioactive Effluent Controls Program.
15.7-8 REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.7.4.2.1.4 Radiation Monitoring System Redundant GM type gamma detectors are mounted on the walls near the edge of the pool to provide reliable and rapid detection of radioactivity released from the pool surface.
If predetermined levels are exceeded, the monitors alarm locally and in the main control room and initiate control action to route the released activity through the emergency exhaust system.
The monitors have an operating range which extends from 0.1 to 104 mR/hr. The lower range level is chosen to assure that normal operating levels are on scale (provides indication that the instrument is operational). Operating levels below 0.1 or greater than 50 mR/hr are unlikely. Initial setpoints are listed in Table 12.3-3.
15.7-8a REVISION 12 - DECEMBER 2008
B/B-UFSAR The worst case fuel handling accident has the potential of exceeding the 10 R/hr maximum range of the fuel handling accident monitors, but this environment will not prevent the monitor from completing its design function. General Atomics (GA) has tested this monitor to 500 R/hr, and based on this test, they have determined that this monitor will perform its function up to 1000 R/hr.
The monitor has been selected to assure initiation of control action within 6 seconds or less. Commercially available area monitors are suitable for this application.
Two separate and independent (nuclear safety-related) monitors are provided for the spent fuel pool. Two nuclear safety-related recorders are provided in the control room for the spent fuel pool.
15.7.4.2.2 Fuel Handling Accident Inside Containment This accident is defined as the dropping of a spent fuel assembly onto the core during refueling which results in the rupture of the cladding of fuel rods.
15.7.4.2.2.1 Identification of Causes and Frequency Classification The cause of the event can be identified as any event which would result in the dropping of a fuel assembly onto the reactor core during refueling.
The frequency classification, as defined in Regulatory Guide 1.70, can be categorized as one of limiting faults. This means that it is an occurrence that is not expected to occur but is postulated because its consequences would include the potential for the release of significant amounts of radioactive material.
15.7-9 REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.7.4.2.2.2 Sequence of Events and Systems Operation The bounding fuel handling accident analysis assumes that the fuel handling building and containment are not isolated at the time of the event and remain open to the environment during the event. Control room HVAC system filtration is delayed for the first 30 minutes of the event for conservatism.
As indicated in Regulatory Guide 1.183, if the fuel handling building and containment are not isolated at the time of the event, the release of all activity to the environment is assumed to occur over a 2-hour time period. To assure this, building exhaust rates are set artificially high in the analysis.
The bounding exhaust point is the plant vent, with specific atmospheric dispersion factors (/Qs) for the control room or offsite receptor locations. Alternative release points through major openings such as the fuel handling building inner or outer rail bay roll-up doors or the personnel/equipment hatch from the containment would have lower /Qs and would therefore result in lower calculated doses.
15.7-9a REVISION 14 - DECEMBER 2012
B/B-UFSAR 15.7.4.2.2.3 System Operation During handling of recently irradiated fuel assemblies in containment (i.e., fuel that has occupied part of a critical reactor core within the previous 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />), the Technical Specification requirements for containment closure ensure that release of fission product radioactivity within containment will be restricted from escaping to the environment.
15.7-10 REVISION 14 - DECEMBER 2012
B/B-UFSAR During movement of irradiated fuel in the containment, the following compensatory measures are required per the Alternative Source Term license amendment:
- a. At least one train of the fuel handling building ventilation system will be available.
- b. At least one fuel handling building trackway door will be maintained closed.
- c. Within one hour of a fuel handling accident, with the equipment hatch off or with the equipment hatch installed and both personnel air lock doors open, airflow in the proper direction will be established in the containment/fuel handling building and all other penetrations in the containment to the outside air will be closed.
- d. Within one hour of a fuel handling accident, with the equipment hatch installed and at least one of the personnel air lock doors closed, all other penetrations in the containment to the outside air will be closed.
- e. Containment and fuel handling building effluents will be monitored consistent with the Technical Specification Radioactive Effluent Controls Program.
15.7.4.2.2.4 Radiation Monitoring System Redundant GM type gamma detectors are mounted on the shield walls near the edge of the refueling cavity to provide reliable and rapid detection of radioactivity released from the pool surface.
If predetermined levels are exceeded, the monitors alarm in the main control room and initiate control action to close the purge system isolation valves.
15.7-11 REVISION 14 - DECEMBER 2012
B/B-UFSAR The monitors have an operating range which extends from 0.1 to 104 mR/hr. The lower range level is chosen to assure that normal operating levels are on scale (indicates that the instrument is operational). Operating levels below 0.1 or greater than 50 mR/hr are unlikely. Initial setpoints are listed in Table 12.3-3.
The worst case fuel handling accident has the potential of exceeding the 10 R/hr maximum range of the fuel handling accident monitors, but this environment will not prevent the monitor from completing its design function. General Atomics (GA) has tested this monitor to 500 R/hr, and based on this test, they have determined that this monitor will perform its function up to 1000 R/hr.
The monitor has been selected to assure initiation of control action within 6 seconds or less. Commercially available area monitors are suitable for this application.
Two separate and independent (nuclear safety-related) monitors are provided for each reactor unit. Unit 1 is provided with two nuclear safety-related recorders for the fuel building fuel handling accident and the containment building fuel handling 15.7-11a REVISION 12 - DECEMBER 2008
B/B-UFSAR accident. Unit 2 is provided with two nuclear safety-related recorders for the containment building fuel handling accident.
All recorders are located in the control room for each unit.
15.7.4.3 Radiological Consequences of a Postulated Fuel Handling Accident (AST)
The key inputs and assumptions used in the AST Fuel Handling Accident (FHA) radiological consequence analysis are summarized below and provided in Tables 15.7-6 and Table 15.7-7.
This event is defined as the drop of a spent fuel assembly onto the spent fuel pool floor or the core, resulting in the postulated rupture of the cladding of all the fuel rods in one assembly. Consistent with RG 1.183, two potential accident locations were considered; the FHB and the containment. The dropped fuel assembly is assumed to have been subcritical for 48 hrs. The RADTRAD computer program was used to model the associated radiological releases.
The assumed fuel assembly source term is based on the reactor core source terms previously described in the LOCA section. The fraction of the core fuel damaged is based on a postulated rupture of all the fuel rods in one assembly. With 193 fuel assemblies in the core operating at full core power of 3658.3 MWth, the damaged fuel assembly is assumed to have been operating with a 1.7 peaking factor; therefore, the damaged fuel assembly power = 3658.3 MWth x 1.7/193 = 32.22 MWth.
The analysis assumes 23 feet of water above damaged fuel. This value corresponds to the minimum depth of water coverage over the top of irradiated fuel assemblies seated in the spent fuel pool racks as required by TS 3.7.14, Spent Fuel Pool Water Level, and is also assumed for an assembly drop in the core. TS 3.9.7, Refueling Cavity Water Level, requires maintaining at least 23 feet of water above the top of the reactor vessel flange during movement of irradiated fuel assemblies within containment. Since the fuel assemblies are seated below the reactor vessel flange, in actuality, more than 23 feet of water covers the assemblies in the core; therefore, this assumption is conservative for an assembly drop in the core. This assumption is consistent with RG 1.183. As prescribed in Appendix C of RG 1.183, an overall DF of 200 is used as the overall effective iodine DF for this 23-foot water depth, with a DF of 1 for noble gases. Particulate radionuclides are assumed to be retained in the pool water (i.e.,
a DF of infinity).
RG 1.183, Table 3 allows application of the following gap activity fractions for non-LOCA events. These gap activities apply to fuel whose burnup and power are bounded by those specified in RG 1.183, footnote 11.
15.7-12 REVISION 15 - DECEMBER 2014
B/B-UFSAR
- 5% of the noble gases (excluding Kr-85)
- 10% of the Kr-85
- 5% of the iodine inventory (excluding I-131)
- 8% of the I-131
- 12% of the alkali metal inventory Both Byron and Braidwood Stations may utilize some fuel assemblies with linear heat generation rates in excess of RG 1.183 footnote 11 limits. To account for high burnup assemblies, the above gap activity fractions were doubled.
Release modeling uses the refueling floor air space (i.e., in the FHB), with the initial air change rate based on the 525,460 cubic feet (86 x 130 x 47 high) volume exposed to the monitor as developed for the post-accident radiation monitor response time calculation, divided into the spent fuel storage pool total ventilation exhaust rate of 12,400 cfm. The initial air change rate is therefore 0.0236 per minute, assumed to last for the entire period until initiation of the CR emergency mode of operation. Consistent with RG 1.183, the release from the FHB to the environment is assumed over a two-hour time period. To assure this, the refueling floor exhaust rate is set artificially high at five times this value or 0.118 air changes per minute during the CR emergency mode of operation. The postulated exhaust point is the plant vent, with specific dispersion characteristics to the CR or offsite receiving locations which are defined by unique dispersion factors, or /Qs. The alternative release point through a major opening such as a FHB inner or outer trackway roll-up door would have lower /Qs and therefore lower calculated dose results.
For the potential accident location in the containment, the corresponding air change rate based on a containment volume of 2,850,000 cubic feet and containment purge ventilation exhaust rate of 40,000 cfm is 0.0140 air changes per minute, considerably lower than the FHB exhaust rates developed above. The purge exhaust point would again be the plant vent, with the same assumed /Qs for the CR or offsite receiving locations as for the potential accident location in the FHB. The FHB potential accident location would therefore be bounding. For the alternative release points through the major opening of the personnel/equipment hatch from the containment to the outside, lower /Qs apply which would result in lower calculated doses.
For an alternative release point through the major opening of the personnel/equipment hatch from the containment to the auxiliary building, the release would be exhausted through the auxiliary building ventilation system to the plant vent, with the same CR
/Qs as for the FHB potential accident location which would again be bounding.
15.7-12a REVISION 12 - DECEMBER 2008
B/B-UFSAR Inputs and Assumptions The following inputs and assumptions and bounding analyzed conditions are utilized in the analysis.
- a. Core inventory is based on a DBA power level of 3658.3 MWt.
This power level bounds the MUR power uprate Rated Thermal Power level including measurement uncertainties.
- b. Spent fuel source terms are based on reactor core source terms as previously discussed, with a conservative factor of 2.0 multiplier to account for the gap fractions of fuel exceeding 54 GWD/MTU burnup with a maximum linear heat generation rate exceeding the 6.3 kW/ft peak rod average power limit to address RG 1.183 footnote 11.
- c. The damaged fuel is assumed to have operated at a radial peaking factor of 1.7. The damage is assumed to be the rupture of the cladding of all the fuel rods in the one assembly.
- d. Movement of fuel will not occur less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the associated reactor shutdown.
- e. A water depth above the damaged fuel of 23 feet is the limiting case, corresponding to the minimum depth of water coverage over the top of irradiated fuel assemblies seated in the spent fuel pool racks within the spent fuel pool.
- f. No filtration of the airborne radioactivity released from the pool or automatic isolation of the accident location is assumed. Essentially all of the activity released to the environment is assumed to reach the containment refueling floor or spent fuel pool refueling floor airspace within two hours after the accident.
- g. Delayed realignment of the CR HVAC system into the emergency mode of operation with filtered make-up (i.e., with an assumed charcoal filtration efficiency of 95%) at 30 minutes after the accident.
- h. An amount of unfiltered inleakage into the CR of 500 cfm, added continuously throughout the accident duration.
15.7-12b REVISION 15 - DECEMBER 2014
B/B-UFSAR Dose Results Radiological doses resulting from a design basis FHA for a CR operator and a person located at the EAB or LPZ are to be less than the regulatory dose limits as given as indicated below.
Regulatory Dose Limits - FHA Dose Type Control Room (rem) EAB and LPZ (rem)
TEDE Dose 5a 6.3b Notes:
a 10 CFR 50.67 b
10 CFR 50.67 as modified by Regulatory Guide 1.183 (Table 6, Page 1.183-20)
The table below provides the results from the limiting case of the unfiltered FHB FHA simulation modeled using the RADTRAD code and the AST assumptions and design input parameters described above.
Fuel Handling Accident Radiological Analysis Results (Maximum of Byron and Braidwood)
Fuel Handling Accident Dose Assessment Results Control Room EAB LPZ (rem TEDE) (rem TEDE) (rem TEDE) 4.40 5.31 0.94 These doses were below the Regulatory Dose Limit, so it is verified that this design basis FHA is sufficiently mitigated at both Byron Station and Braidwood Station.
An additional FHA analysis was performed for recently irradiated fuel with containment closure established or with the FHB ventilation system operable. The results of this analysis also met the limits of 10 CFR 50.67 assuming a minimum decay time of seven hours. The seven-hour minimum decay time is inconsequential, as it is physically impossible to remove the reactor head and move fuel within the first seven hours after the reactor is subcritical.
15.7.5 Spent Fuel Cask Drop Accident 15.7.5.1 Identification of Causes and Frequency Classification A spent fuel cask will follow the path outlined in Figure 15.7-1.
The fuel handling building overhead crane meets the single-failure proof criteria of ASME NOG-1-2004, NUREG-0554 and NUREG-0612, Appendix C, therefore, a drop of the main hook/lower load block is not credible. The main hoist is classified as a Type I main hoist per ASME NOG-1-2004 - single - failure proof for loads up to 125 tons. The auxiliary hoist is single-failure proof up to 15 tons per NUREG-0554.
15.7-12c REVISION 17 - DECEMBER 2018
B/B-UFSAR The upgraded overhead crane is designed to ensure that a single failure of any one of the main components of the crane system will not result in the loss of capability of the system to safely retain a load.
A failure in the lifting gear (sling) would result in an unsafe condition but would not affect the ability to raise or lower the cask.
The following items are assumed to be incapable of failure:
- a. the cable drum;
- b. any structural component of the crane (bridge beams, rails, etc.);
- c. any support for the crane rails which is tied into the building steel; and
- d. electrical crane interlocks which prevent movement of the cask over the spent fuel pool.
15.7-12d REVISION 14 - DECEMBER 2012
B/B-UFSAR Therefore, the only single failure in the crane which could cause a fuel cask drop is a failure in the main hook, which is not credible since a significant safety factor (10 to 1) is considered in the hook design.
15.7.5.2 Evaluation and Analysis Figure 15.7-1 shows the general arrangement of the fuel handling building and the recommended route for the cask to move the spent fuel.
Guard walls are provided around the cask loading area in the spent fuel storage pool as shown on the plan (Figure 15.7-1) and in Section A-A (Figure 15.7-2). These walls which surround the cask loading area, rise the full height of the pool and are structurally designed to withstand the impact force due to a falling cask. If the cask is positioned over the cask loading area and tips and falls, it will land on the guard walls. Since the center of gravity of the cask is within the loading area, as noted in Figure 15.7-2, the cask cannot tip over into the spent fuel storage pool.
A cask drop incident is not credible since the FHB crane is single-failure-proof in accordance with ASME NOG-1-2004, NUREG-0554, and NUREG-0612, Appendix C and the lifting devices and cask attachment points meet the requirements of NUREG-0612.
In addition, the fuel handling building crane will be restricted from operating over the spent fuel pool storage pool and the cask fill area by providing an electrical interlock for the crane bridge as indicated in Figure 15.7-1 (Interlock #1). The fuel handling building crane will also be restricted from leaving the cask fill area and operating over the spent fuel pool by a second interlock for the crane bridge as indicated in Figure 15.7-1 (Interlock #2). The second electrical interlock also prevents the trolley from traveling over the spent fuel pool.
Any other heavy loads (loads exceeding 2000 pounds) will be carried in the spent fuel pool area in accordance with NUREG-0612.
During new fuel loading, the 15-ton auxiliary hook is used to remove the new fuel from the transport vehicle to the new fuel storage racks or new fuel elevator. It is required to have full freedom of travel horizontally to perform this task, so there are no interlocks or stops to prevent hook movement during this period. The auxiliary hook can travel up to 5 feet 6 inches over the spent fuel pool where it is restrained from further motion by electrical interlocks. The 125-ton main hook (load block) is considered a heavy load, but because the upgraded load block is supported by four redundant wire ropes and the maximum load on each wire rope with the maximum critical load attached does not exceed 10% of the manufacturers published breaking strength, which is much greater than the weight of the main hoist load block, travel over the spent fuel pool is permitted without a load attached. When the crane does travel over the spent fuel pool, the interlocks must be bypassed and the hoist selector 15.7-13 REVISION 14 - DECEMBER 2012
B/B-UFSAR switch must be placed in the Aux Hoist position per procedure to prevent inadvertent lowering of the main hoist. These interlocks may be bypassed with loads of less than 2000 pounds on the crane. The interlocks will be controlled administratively using the guidelines listed in NUREG-0612, Table 2.1-1.
The electrical interlocks (limit switches) on the bridge and trolley prevent the main hook of the crane from travelling over the spent fuel storage area when handling a spent fuel cask.
These interlocks may be bypassed when the crane is used for new fuel handling operations over the spent fuel pool.
15.7-13a REVISION 13 - DECEMBER 2010
B/B-UFSAR New fuel operations and cask handling will not be performed simultaneously, thus minimizing the possibility of improper movement of the cask.
The main hoist/load block is considered a heavy load, but because the reeving of the upgraded load block is supported by four redundant wire ropes, and the maximum load on each wire rope with the maximum critical load attached does not exceed 10% of the manufacturer's published breaking strength, which is much greater than the weight of the main hoist load block, it is very unlikely that the main hook and lower load block could be dropped on the spent fuel. Therefore, it can travel over the spent fuel pool.
Even if such an event were to occur, the resulting damage to the fuel would not result in a release which exceeds the limits of 10 CFR 100. This can be seen by extrapolation of the results of a postulated single fuel element drop in Subsection 15.7.4. This shows that a large number of elements must be damaged to exceed the 10 CFR 100 limits. The lower load block is not large enough to cause this damage.
All potential accidents involving lifting and transporting of loads heavier than a fuel element are addressed in a report submitted in response to NUREG-0612. The fuel handling building crane and loads are included in this report.
The consequences of the drop of loads lighter than a fuel element will be less than the drop of a single fuel element as reported in Chapter 15.0. The design of the tools and the fuel building cranes prevents the tools from dropping onto the fuel from a great height. The heaviest of these loads is the RCC change tool which weighs less than 1100 pounds. This tool is over 30 feet long. Because of the height of the fuel building crane, the RCC change tool can only be carried a few feet above the fuel. With the short vertical drop distance and the low weight per foot of length involved, there is no real probability of damage to the fuel. The burnable poison assembly handling tool, the thimble plug handling tool, and the spent fuel assembly handling tool all have weights under 30 pounds per foot and are not carried high above the fuel. All other tools have gross weights under 100 pounds. The single fuel assembly drop accident in Subsection 15.7.4 is the maximum credible accident involving dropped loads and spent fuel damage.
An unrestrained stack-up configuration is used during the loading or unloading process for transferring the Multi-Purpose Canister (MPC) (loaded with spent fuel assemblies) between the HI-TRAC transfer cask (HI-TRAC) and the HI-STORM 100 Version B overpack (HI-STORM) in the Fuel Handling Building (FHB). In the unrestrained stack-up configuration, the HI-TRAC transfer cask is located on top of the HI-STORM overpack and is bolted to the Mating Device (MD) which is bolted to the top of the HI-STORM (without HI-STORM lid in place). The unrestrained stack-up configuration is shown in Figure 9.1-23.
15.7-14 REVISION 16 - DECEMBER 2016
B/B-UFSAR This stack-up configuration has been demonstrated via analysis (Section 9.1.2.3.11) to maintain its stability (rocking and sliding) during a seismic event without tip-over. In the unlikely event of a tip-over of the casks in the FHB track-way, there is more than 15 feet of separation between the top of the HI-TRAC and the closest edge of the Spent Fuel Pool. Therefore, due to the spatial separation, a tip-over of the HI-TRAC/ HI-STORM stack-up will not result in a compromise to Spent Fuel Pool integrity.
15.7.5.3 Barrier Performance The spent fuel pool walls are designed to withstand increased water pressure caused by a vertical drop of the cask.
If a cask drops on the exterior pool wall, and tips, it will land on the guard walls which are designed to withstand the impact force due to a falling cask. The cask will not fall outside the cask storage well and thus will not affect the fuel in the fuel storage pool.
15.7-14a REVISION 16 - DECEMBER 2016
B/B-UFSAR 15.7.5.4 Modeling of Accident Analysis 15.7.5.4.1 Mathematical Model A free drop of a fuel cask from a height of 30 feet or more onto an unyielding surface (see 10CFR 71) is not a credible event since a single failure proof crane and lifting device will be employed during cask handling activities. Therefore, there will be no resulting damage to the cask that will cause a release of radioactive materials to the public.
15.7.5.4.2 Input Parameters and Initial Conditions There are no input parameters and initial conditions of variables relevant to the evaluation of barrier performance that were not presented in Subsection 15.7.5.3.
15.7.5.4.3 Results Inasmuch as the cask cannot be free dropped onto an unyielding surface from a height of 30 feet or more, the cask will not be damaged and there will be no release of radioactive materials to the public.
Moreover, if damage to the cask compartment liner resulted from a drop of the cask from maximum hook height, the fuel pool would not become dewatered as a result of such damage. This is due to the watertight gate that separates the cask compartment from the main portion of the fuel pool.
Since the cask will always be brought to the pool so as to enter the storage well without passing over the pool, there is no possibility of cask drop into the fuel storage pool.
The fuel pool slab rests directly on rock (Byron) or on stiff soil (Braidwood). This slab was designed for a vertical drop of the cask in the storage well.
15.7.5.5 Radiological Consequences A spent fuel cask drop accident is not a credible event when a single failure proof crane and lifting device are employed to move the cask and therefore, is less severe than the fuel handling accident analyzed in Subsection 15.7.4.
15.7-15 REVISION 14 - DECEMBER 2012
B/B-UFSAR For Braidwood Station a shipment specific radiological analysis was performed for the dropping of an NAC International Shipping Cask with five fuel rods. Impact limiting devices are not employed during the movement of the NAC International Shipping Cask in the fuel handling building. A drop of the cask would result in damage and subsequent release of radioactive materials to the public. Results of the radiological analysis show that the dropping of the loaded NAC International Shipping Cask is not limiting and is much less severe than the fuel handling accident analyzed in Subsection 15.7.4. This analyses is applicable only to Braidwood Station and only for this specific shipment.
15.7.6 References
- 1. NUREG/CR-5009, Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors, February 1988.
- 2. WCAP-7828, Radiological Consequences of a Fuel Handling Accident, December 1971.
- 3. NUREG-0800, Section 15.7.4, Radiological Consequences of Fuel Handling Accidents, Revision 1, July 1981.
15.7-15a REVISION 13 - DECEMBER 2010
B/B-UFSAR TABLE 15.7-1 PARAMETERS USED IN WASTE GAS DECAY TANK RUPTURE ANALYSES (using TID 14844)
REGULATORY GUIDE 1.24 ANALYSIS Fuel defects 1%
Primary Coolant Table 15.0-9 Activity Time of accident Immediately after shutdown at end of equilibrium core cycle Meteorology See Tables 15.0-13 and 15.0-14 Nuclide data Table 15A-1 15.7-16 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.7-2 WASTE GAS DECAY TANK INVENTORY (ONE UNIT)
(using TID 14844)
Security - Related Information Withheld Under 10 CFR 2.390 15.7-17 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.7-3 SPENT RESIN STORAGE TANK IVENTORY (using TID 14844)
Security - Related Information Withheld Under 10 CFR 2.390 15.7-18 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.7-4 This page has been intentionally deleted.
15.7-19 REVISION 15 - DECEMBER 2014
B/B-UFSAR TABLE 15.7-5 This page has been intentionally deleted.
15.7-20 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.7-6 BOUNDING ISOTOPIC CORE INVENTORY Security - Related Information Withheld Under 10 CFR 2.390 15.7-21 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.7-7 INPUT PARAMETERS FOR THE FHA RADIOLOGICAL CONSEQUENCE ANALYSIS USING AST Parameter Unit Value Notes Gap Activity: 2 times As a significant the number of fuel following: assemblies not I-131 % 0.08 qualifying for AST due Kr-85 % 0.10 to their containing Other Noble Gases % 0.05 fuel rods with maximum Other Halogens % 0.05 linear heat generation Alkali Metals % 0.12 rates exceeding 6.3 kilowatts per foot peak rod average power for burnups exceeding 54 GWD/MTU, the fuel will be treated as having gap fractions a factor of 2 greater than the RG 1.183 values.
Number of Assemblies Damaged # 1 264 rods are assumed to be damaged. This is one full assembly.
Pool Scrubbing Factor: Per RG 1.183, based on Overall - 200 23 feet of water Elemental - 500 coverage.
Organic (and Noble Gases) - 1 Release Path Filter Efficiency % 0 Filtration is not credited for AST analysis.
Duration of release hr 2 Chemical form of radioiodine released from the fuel to the spent fuel pool:
Cesium iodide % 95 Elemental iodine % 4.85 Organic iodine % 0.15 Chemical form of radioiodine released from the pool to the building:
Elemental iodine % 57 Organic iodine % 43 Depth of Water above the Top of Reactor Vessel Flange and Fuel Assemblies in Spent Fuel Pool ft 23 15.7-22 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15.7-7 (Cont'd)
This page has been intentionally deleted.
15.7-23 REVISION 9 - DECEMBER 2002
B/B-UFSAR TABLE 15.7-8 This page has been intentionally deleted.
15.7-24 REVISION 9 - DECEMBER 2002
B/B-UFSAR 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS)
An anticipated transient without scram (ATWS) is an anticipated operational occurrence (such as a loss of feedwater, loss of condenser vacuum, or loss of offsite power) that is accompanied by a failure of the reactor trip system to shut down the reactor.
A series of generic studies (References 1 and 2) on ATWS showed that acceptable consequences would result, provided that the turbine trips and that auxiliary feedwater flow is initiated in a timely manner.
The effects of an ATWS are not considered as part of the design basis for transients analyzed in Chapter 15. However, 10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants," the ATWS Rule, requires that each pressurized water reactor have equipment that is diverse from the reactor trip system to automatically initiate the auxiliary feedwater system and initiate a turbine trip under the conditions indicative of an ATWS.
In addition, compliance with the ATWS rule is demonstrated on a cycle-by-cycle basis by focusing on two aspects of WCAP-11992, "ATWS Rule Administration," December 1988 methodology (Reference 3). These aspects are the unfavorable exposure time (UET) and critical trajectory methodologies. The critical trajectories are calculated loci of plant conditions (e.g., power vs. inlet temperature) that provide a peak pressure in the transient analysis of the limiting ATWS event, which is then compared to the specified limit (3200 psig). The UET is the time during the cycle when reactivity feedback is insufficient to maintain pressure under 3200 psig for a given reactor state.
In the application of the UET methodology, the ATWS transient point kinetics information is transferred into steady-state conditions for comparison with cycle-specific core condition evaluation calculations, and the critical trajectories are determined. During peak ATWS pressure conditions, heatup is relatively slow so that steady-state analysis is acceptable. The methodology uses the "base case" conditions from the ATWS submittal, with 100-percent power-operated relief valve capacity, 100-percent auxiliary feedwater, and no control rod insertion.
The cycle-specific calculations are done with appropriate ATWS initial conditions of full power, rods out, equilibrium xenon, and 3200 psig pressure. These calculations are compared to the critical trajectory from the transient analysis. This comparison provides cycle-specific design conditions that would result in transient conditions exceeding 3200 psig. These calculations 15.8-1 REVISION 15 - DECEMBER 2014
B/B-UFSAR show any core design conditions that would result in exceeding the 3200 psig pressure limit. Calculations as a function of the time in cycle and, thus, as a function of moderator temperature coefficient (MTC) show the time during the cycle that the core design critical trajectory is greater than the transient trajectory. From this, the UET is determined. The analysis must show that the UET, given the cycle design, will be less than five percent, or equivalently, that ATWS pressure limit will be met for at least 95 percent of the cycle. If the limit is not met, the core design would be changed until the 95-percent level is achieved.
This 95-percent probability level for the UET is equivalent to the probability level in the reference analyses for the ATWS rule basis. All parameters should be best estimate values with the exception of the MTC initial condition. That was to be at a level not to be exceeded (i.e., not less negative) at full power conditions for at least 95 percent of the cycle.
15.
8.1 REFERENCES
- 1. Burnett, T. W. T., et al., "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974.
- 2. Anderson, T. M., (Westinghouse) letter to S. H. Hanauer (USNRC), "ATWS Submittal," NS-TMA-2182, December 1979.
- 3. Sloane, B. D., et al, Joint Westinghouse Owners Group /
Westinghouse Program: ATWS Rule Administration Process, WCAP 11992, December 1988.
15.8-2 REVISION 15 - DECEMBER 2014
ATTACHMENT 15A DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS
B/B-UFSAR TABLE OF CONTENTS PAGE 15A DOSE MODELS USED TO EVALUATE THE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15A-1 15A.l INTRODUCTION 15A-1 15A.2 ASSUMPTIONS (TID-14844) 15A-1 15A.3 OFFSITE DOSE MODELS (TID-14844) 15A-1a 15A.4 CONTROL ROOM DOSE MODELS (TID-14844) 15A-1c 15A.5 REFERENCES 15A-2 15A-i REVISION 12 - DECEMBER 2008
B/B-UFSAR LIST OF TABLES NUMBER TITLE PAGE 15A-1 Physical Data for Isotopes 15A-4 15A-ii
B/B-UFSAR ATTACHMENT 15A DOSE MODELS USED TO EVALUATE THE RADIOLOGICAL CONSEQUENCES OF ACCIDENTS 15A.1 INTRODUCTION This Attachment identifies the models used to calculate the radiological doses that would result from releases of radioactivity due to various postulated accidents.
The major design basis accidents were analyzed using Alternative Source Term methodology using Regulatory Guide 1.183 with dose acceptance criteria per 10 CFR 50.67. These accidents include:
Loss of Coolant Accident (LOCA)
Main Steam Line Break (MSLB)
Control Rod Ejection Accident (CREA)
Locked Rotor Accident (LRA)
Steam Generator Tube Rupture (SGTR)
Fuel Handling Accident (FHA)
Conformance with Regulatory Guides 1.183 and 1.194 are documented in Appendix A.
Other accidents and events not listed above (including doses related to equipment qualification) were analyzed using the TID-14844 methodology per Regulatory Guide 1.4 with 10 CFR 100 dose acceptance criteria.
15A.2 ASSUMPTIONS (TID-14844)
The following assumptions are basic to all dose models:
- a. Direct radiation from the source point is negligible compared to gamma radiation due to immersion in the semi-infinite radioactive cloud.
- b. All radioactivity releases are treated as ground level releases regardless of the point of discharge.
- c. The dose receptor is a standard man as defined by the International Commission on Radiological Protection (ICRP) (Reference 1).
- d. Radioactive decay from the point of release to the dose receptor is neglected.
- e. The offsite atmospheric dispersion factors used are the 5th percentile values from Table 15.0-13 and 15.0-14.
15A-1 REVISION 15 - DECEMBER 2014
B/B-UFSAR 15A.3 OFFSITE DOSE MODELS (TID-14844)
Whole body (acute) doses and thyroid doses from inhaled radioactive iodine are calculated for the 0 - 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> interval at the exclusion area boundary and for the duration of the accident at the low population zone (LPZ) outer boundary.
15A.3.1 WHOLE BODY DOSE (TID-14844)
The whole body dose delivered to a dose receptor is obtained by considering the dose receptor to be immersed in a radioactive cloud which is infinite in all directions above the ground plane, i.e., an "infinite semispherical cloud." The concentration of radioactive material within this cloud is taken to be uniform and equal to the maximum centerline ground level concentration that would exist in the cloud at the appropriate distance from the point of release.
The whole body dose is a result of external gamma radiation.
15A-1a REVISION 12 - DECEMBER 2008
B/B-UFSAR Two different models are used for calculating the whole body dose. For the noble gases, the model taken from Regulatory Guide 1.109 (Reference 2) is used. The equation is:
D (X / Q) i (Ai)(DCFy i)
Where: D = Whole body dose, rem X/Q = the 5th percentile atmospheric dispersion factor for a given time period and distance (see Tables 15.0-13 and 15.0-14), sec / m3 Ai = the activity of nuclide i released during a given time period, Ci DCFy-i = gamma dose conversion factor for nuclide i (see Table 15A-1), rem
- m3 / Ci
- sec Regulatory Guide 1.109 does not include dose conversion factors submersion in a cloud of radioactive iodine. The whole body dose contribution from the iodine in the cloud of activity is calculated using the following equation from Reference 3.
D (0.25)( X / Q) ( Ai )( E i )
i Where: E-i = average gamma disintegration energy for nuclide i (see Table 15A-1), MeV per disintegration 15A.3.2 THYROID INHALATION DOSE (TID-14844)
The thyroid dose for a given time period is obtained from the following expression:
Dthy B( X / Q) ( Ai )( DCFthy i )
i Where: Dthy = thyroid dose, rem B = breathing rate, m3/sec (from Reference 3), the breathing rate offsite varies with time as follows:
0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.47E-4 m3/sec 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.75E-4 m3/sec
>24 hours 2.32E-4 m3/sec 15A-1b REVISION 12 - DECEMBER 2008
B/B-UFSAR X/Q = the 5th percentile atmospheric dispersion factor for a given time period and distance (see Tables 15.0-13 and 15.0-14), sec / m3 Ai = the activity of iodine isotope i released during a given time period, Ci DCFthy-i = thyroid dose conversion factor for iodine isotope i (see Table 15A-1), rem
/ Ci inhaled 15A.4 CONTROL ROOM DOSE MODELS (TID-14844)
The operators in the control room will accumulate doses due to the activity entering the control room as the result of an accident. The integrated activity in the control room during a given time interval is found by multiplying the release by the appropriate X/Q (see Table 6.4-1a) to give the concentration at the control room intake. This activity is brought into the control room through the filtered intake pathway and by unfiltered inleakage. The activity in the control room atmosphere is reduced by recirculation through filters, by exhausting a portion of the air to the environment (balancing the air inflow and inleakage), and by radioactive decay. The flows and filter efficiencies are provided in Table 6.4-1a.
Using the integrated activity in the control room, the whole body (acute) doses, beta-skin doses, and thyroid doses are calculated for the operators in the control room from the models described in the following subsections.
15A.4.1 WHOLE BODY DOSE (TID-14844)
As with the determination of offsite doses, there are two different models used for calculating the whole body dose. For the noble gases, the model is based on Regulatory Guide 1.109 (Reference 2). The equation is:
D (GF )(OF ) ( Ai )( DCF i )
i Where: D = whole body dose, rem GF = geometry correction factor used to convert the semi-infinite cloud to a finite cloud (from Reference 4, GF =
1173 / V0.338 where V is the control room volume in cubic feet) 15A-1c REVISION 12 - DECEMBER 2008
B/B-UFSAR OF = the occupancy factor (i.e., the fraction of the time period that the operator is assumed to be present in the control room), from Reference 4, the values are:
0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OF = 1.0 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> OF = 0.6 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> OF = 0.4 Ai = the average concentration of nuclide i in the control room atmosphere during a given time period, Ci/m3 DCF-i = whole body dose conversion factor for nuclide i (see Table 15A-1), rem
- m3 /
Ci
- sec Regulatory Guide 1.109 does not include dose conversion factors for sumbersion in a cloud of radioactive iodine. The whole body dose contribution from the iodine in the cloud of activity is calculated using the following equation:
D (GF )(0.25)(OF ) ( Ai )( E i )
i Where: E-i = average gamma disintegration energy for nuclide i (Table 15A-1), MeV per disintegration 15A.4.2 BETA-SKIN DOSE (TID-14844)
As with the whole body dose, two different models are used for calculating the beta-skin dose. For the noble gases, the model is based on Regulatory Guide 1.109 (Reference 2). The equation is:
D OF i (Ci )( DCF i )
Where: D = Beta-skin dose, rem DCF-i = Beta-skin dose conversion factor for nuclide i (see Table 15A-1), rem
- m3 /
Ci
- rem Regulatory Guide 1.109 does not include dose conversion factors for submersion in a cloud of radioactive iodine. The beta-skin dose contribution form the iodine in the cloud of activity is calculated using the following equation:
15A-1d REVISION 12 - DECEMBER 2008
B/B-UFSAR D (0.23)(OF )
i
( Ai )( E 1 )
Where: E-i = average beta disintegration energy for nuclide i (Table 15A-1), MeV per disintegration 15A.4.3 THYROID DOSE (TID-14844)
The equation for the thyroid dose is:
Dthy (OF )( B) i
( Ai )( DCFt hy i )
Where: Dthy = thyroid dose, rem B = breathing rate, m3/sec (from Reference 4, the breathing rate in the control room is 3.47E-4 m3/sec and does not vary with time)
DCFthy-i = thyroid dose conversion factor for nuclide i (see Table 15A-1), rem / Ci inhaled 15A-1e REVISION 12 - DECEMBER 2008
B/B-UFSAR 15A.5 REFERENCES
- 1. "Report of ICRP Committee II on Permissive Dose for Internal Radiation (1959)," Health Physics Vol. 3, pp. 30, 146-153, 1970.
- 2. Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977.
15A-2 REVISION 9 - DECEMBER 2002
B/B-UFSAR
- 3. Regulatory Guide 1.4, Revision 2, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactor," USAEC, June 1974.
- 4. Murphy, K. G. and K. M. Campe, Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19, presented at the 13th AEC Air Cleaning Conference, November 1974.
- 5. ICRP Publication 38, Radionuclide Transformations - Energy and Intensity of Emissions, 1983.
- 6. ICRP Publication 30, Limits for Intakes of Radionuclides by Workers, 1979.
- 7. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.
- 8. NUREG/CR-6604, RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation, April 1998; Supplement 1, June 1999; and Supplement 2, October 2002.
- 9. U.S. Federal Guidance Report No.11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, 1988.
- 10. U.S. Federal Guidance Report No.12, External Exposure to Radionuclides in Air, Water, and Soil, 1993.
- 11. Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, June 2003.
- 12. Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982.
15A-3 REVISION 12 - DECEMBER 2008
B/B-UFSAR TABLE 15A-1 PHYSICAL DATA FOR ISOTOPES Isotope Decay Thyroid Average Gamma Average Beta Constant*, Dose Disintegration Disintegration hr-1 Conversion Energy*, Energy*,
Factors**, Mev/Disintegration Mev/Disintegration rem/Ci I-131 3.59E-3 1.07E6 3.81E-1 3.81E-1 I-132 3.01E-1 6.29E3 2.28E0 2.28E0 I-133 3.33E-2 1.81E5 6.07E-1 6.07E-1 I-134 7.91E-1 1.07E3 2.62E0 2.62E0 I-135 1.05E-1 3.14E4 1.58E0 1.58E0 Isotope Decay Whole-Body Dose Beta-Skin Dose Constant*, Conversion Factor ***, Conversion Factor***,
hr-1 rem-m3/Ci-sec rem-m3/Ci-sec Kr-85m 1.55E-1 3.71E-2 4.63E-2 Kr-85 7.38E-6 5.10E-4 4.25E-2 Kr-87 5.45E-1 1.88E-1 3.085E-1 Kr-88 2.44E-1 4.66E-1 7.51E-2 Xe-131m 2.43E-3 2.90E-3 1.50E-2 Xe-133m 1.32E-2 7.96E-3 3.15E-2 Xe-133 5.51E-3 9.32E-3 9.70E-3 Xe-135m 2.72E0 9.89E-2 2.25E-2 Xe-135 7.63E-2 5.74E-2 5.90E-2 Xe-138 2.93E0 2.80E-1 1.31E-1
- Reference 5
- Reference 6
- Reference 2 15A-4 REVISION 9 - DECEMBER 2002
B/B-UFSAR 5B has been deleted intentionally.
REVISION 1 - DECEMBER 1989
REVISION 15 DECEMBER 2014 80 70
'*\
60 \
\
\
\
\
\
\
\
\
\
\
\
\
30 rfl\T tpotnts: (liiili~~~*
20 560 580 600 620 640 660 Tavg (Deg-F)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.0-1 ILLUSTRATION OF OVERTEMPERATURE AND OVERPOWER ~T PROTECTION
REVISION 6 DECEMBER 1996 10 i
-
LI.
- 9 8
~
-c CD 7
i8 6
! 5 i
!- 4
~ 3 I 2
~ 1 0 "-""'._....................................................--...-..................-......................................._._~i....&,..I,~
0 10 20 30 40 50 60 70 80 90 100 Power (Percent power)
BYRON/BRAIDWOOD STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 1-5.0-2 MODERATOR TEMPERATURE COEF'FICIENT USED IN ANALYSIS
RBVls10N *:.9* !-.
DECEMaeR '~~QZ NOTE J .. "UPPER ClJRVE" LEAST Nt:GATlVE JlOPPLER ONLY POW~tt UEf'.ECT ~ .o.78 llj) (O TO tOOo/~ PO'Wj:R)
NOTE 2 *LEAST NEGATIVE END*OF*LIFE DOPPLER ONLY POWl!;J.lDEFECT.,. ,.0.91 llp (0 1'0 100% POWEf{}
NOTE 3 * "LOWER.CURVE"MOST NJ£GAT1Vtt.bOPPLl!ilt ONLY POWER DEFECT'"' .J.6 Ap (0 TO 100% POWER) 0
- ?
-4
-6
- ~*
0 I:\,
'#.
..a
- 10
~
~ ~
Cl
-12 Ill ~14
-16
- 18 .
..20 0 10 20 30 40 50 60 70 80 90 100 o/o POWER BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE is.o-3 DOPPLER POWER COEFFICIENT USED IN ACCIDENT ANALYSIS
..
~
>,_
Q w ..
~
_, a:
_,w ...U'?
- >Cl)
L\, ~ I-0
..
~
-,_
Cl.
- c *~:c
~
Q ..
<"?
fl)
<(
Q LI.
0
- a. ..
""t
&L.
0 Q.
0 I-
.. WOe
"":
~
..
q 1-w
~!
I-CS!
a: ....
0 Q13.
QC oq wa:NO 0 -'UJ
""z
<(_
"": :Ee>
0 a:w om
~ Za.
0 0 a:
Lt> Q d a:
l&J*
~
....u.
0 <(
CW) w d :&
N d
-t=:
...d
... ...-. ...
""t ~ ~
d co d
.....
d
~
0 in d *d "'d N d
...d 0 0
U.OdHS'VO ~O dO! O.L 3:>N'V.LSI0/03ddOYO 3:>N'V.LSIQ)
NOIJ.ISOd VOOl:f 03Zl1VWl:f0N BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15,0-4 RCCA POSITION VERSUS TIME TODASHPOT
1.0 .
0.9 0.8
...%a:
0 0.7 I:
>
!:: 0.6
>
(,.)
w 0.5 c(
a:
Q w 0.4
~
_,
ct
- e a: 0.3 0
z 0.2 0.1 0
0 0.2 0.4 0.6 0.8 1.0 ROD POSITION (FRACTION INSERTED)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.0*5 NORMALIZED ROD WORTH VERSUS PERCENT INSERtED
~
cc 0.8 i
-t
->~
t-0.6 w
cc c
8cc 0.4 Q
w
!::f
""'c 0.2
- &
cc i
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 1.3 1.4
. TIME AFTER DROP BEGINS NORMALIZED ROD DROP TIME TIME TO TOP OF DASHPOT BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.0-6 NORMALIZED RCCABANK REACTIVITY WORTH VS. NORMALIZED DROP TIME
ABBREVIATlOMS* VS'ED:
AFwS ... AQXJLIARY fEE,DWA1'ER svgEM ECC~ - ~~£RGE~CV C()RE CQO~l"G SYST~M eves -* CHEMICAL .*At4o 'V,OLllME cotli'ROL. :HL *
- HOT 'LEG SYSTEM .CL
- COLD LEG ESFAS. - ENGINEERED SAF~.TY FEATURE~ *~cws
- - :svs'TEM * .* * * *WATER***
COMPONENT *COOUNG ACTUATION SYSTEM
.FW "'FEEOWATER RCS .. ftEAnoR COOL.A"T SY.STEM RTS *~. REACTOR .TR Ip SYSTEM :s.Ws - SERY I CE WATER *;s_ vstbf .
SIS "" SAFETY INJECTION SYSTEM HP I "' HlGft PRESSURE INJtCllOH SI ** SAFETY i NJE'Cr foN '* .* LP I .. LOW .PRESSURE ' INJECTlON RT - REACTOR tRIP c1 .. coNTAINMENf 1soLhi'oN CS - CONTA 1'NM~NT' .SP.RAY S~ .. ~TEAM GENERATQR NOTES:
I. FO~ T~IP INlTIAllON AHO SAFETY SYSTEM ACTµATf~tt MµLTlPLE SIGNALS ARE SHOWN 'eut ONLY A s rN.Gl.E Si GNA{ :1s REQU I REI). THE otH~r(S'iGNAtS ARE BACKUPS. . .
- 2. HO tJM{ffG SEQUENCE IS IMP~IED 8.Y POSIT IQ~ OF VAR10US B~A"CHE$;
REFER TO EVENT '*Tl'MING SEQOENCE*s P~ESE~TED IN TABULA~* FORM iM PERTINENT
~CCIOENT ANALYSIS SECTION OF CHAPTER 15.0 OF THE FSAR.
DIAGRAM SYMBOLS:
c* )* EVENT TITLE
<> ' BRANCH POINT FOR DIFFtRENT PLANt CONOITIONS I ,.- SAF.EJY SYSTEM
( ) .. SAF.ETY ACTION f~n) . - SYSTEM REQUIRED TO MEET SINGLE-FAILURE CRITERIA
....
I PI
-, - MANUAL ACTION REQUIRED DUR I NG SYSTEM OPERATION
\.._,,,
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.0*7
.SVMEiPLS USED IN ACCiOENT SEQUENCES
EXCESSIVE HEAT REMOVAL DUE TO FEEOWATER SYSTEM MALFUNCTIONS HIGH NEUTRON FLUX Sl SIGNAL POWER RANGE 2/4 STEAM HIGH-HIGH SG GENERATOR RTSJ. ovER POWER t.T 2/4 SAFETY (EVELIN ANY SG VALVES c 1 PASSIVE
~**I]\
"ti OVER TEMPERATURE AT 2/4
..,, fTl 0 ~IA. MAIN f'EE:OWATE~
(Tl x ~ SOl.ATION/TURBINEI rrJ (") (Tl m " W£1Y VALVES TRIP . .
OrrJ 0-( OPEH.10 REI.ml£ Ste~ .
REACTOR S'SEU. PRESSURE .
- E (/) :::0 TRIP
)> (/) ..,,o
-I 1--1 BREAKERS (Tl< -z
- o rrJ 1-t
..,, r-CD
~"
(/) :::c G') :::0 'Slf) Cl.OS£ MAJN I'll' ISOLATION c;o VA!.,Y~~,ij_J:BINE
~**~ ..,, 0
"
m ~**~ CONTROL
~;o RODS (Tl (.11
- ~**~0 GRAVllY 31: *3;: )>0 INSERTION CS>
fl!~ I z)> (./) PASSIVE
-rt> CXl Cr- !<~
- z (/) ~
()0 ..... CONTROL ROD
-IC: (1')0 - ~
1-t fTl z CON1ROt.
0 ;o (/)
- Z...;
t/) 0 trj o;;o 0 rrlP'J*****.
(")<
- 0
...;
~Vi
- r:rl"""
mO
- oz.
"°.,.....
s~
- s O')
omrowu iT 2/11
liviilffilPEiAtuat ** 6t.
' PCIWH U.llGE "~ lllUTROil
. ' FM!.~ *. . ~Tl ;
- T ..* . **.
CitNEAATQR
.SAFEN .
,, ,'
i '
- ~**
.~
I
.PASSIVE
'ff;.l..VES
'liUctciii
.TitiP ,$.WJ'f l/Al.VES . . ..
- SAFEf'fVALV£S *** . .*
l~JAKHS Of.'Et:I TC), RELIEVE . PEN TO A£m:v£ S£CONOAR'f
<RCS PRESSURE SYSTEM PRESSURE . . .
- -;. ,
v FJ
'"' 1. 1il1s 01loalM *~'~'" to*o' ciili1iloL liallil MAliUAL AllUUTOidTIC COllTIOL
- q*'Y 1rv
- llOiiU. ..
llllEIT,I.~
l'AIStn
- 2. roil 1111~ r-u**tllT. *r~c10*
..OTlCTiOll SYlfbl FililCTIOllS ARE AllUMEO To 1.£ OPEmlYI.
liit ~. HACTOR TR IP II !lot Emmo.
. . . BYR()N/lflAIQW()OD STATIONS .
UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.0*9 EXCESSIVE LOAD INCREASE
~:'::'
'-'!i'
- *' .*, ' ji.;
§1!1' '.;J... ...*;*.*
a!l1 """""'
J;;.~:1 al.*~*.1:-s:
S..li::*........ **.:a:. ...*. .
I *1s"'l1
!1 1""w"' 1'l
,.
....** *.*.........
i::
.* .. . *. .*.*S :::*..*$.**.ii 1it 11 ts 6 1
1alii1!.*
§c:!atltl!:e
.. ':°' 11 .. ~
.....
. . =-*
"' Ya**
~
!J*~
y,;N ...
~
- 1 ...
- i. I;; EIYAC)N/BRAIDWOOD .ST~'J'l()N$
UPDATEO FINAL SAFETY ANALVSISREP08T Ii i ;;;
FIGURE 15.0"'10 DEPAESSURf%ATjON OF MAIN STEAM SYSTEM
RE\/lStQN . 11 DECEMBER 2006
.HIGH . PtUSUI IZ £Ii 2/~ RYS PAISSUAE ovfrllNPliOiJRE t.T pil£sluilim . :lflAM tEACTOI suin GUE UT.QA 1Af~1Y YA.LYES UIP ~*U e*mt*s P&SSIYl PlSSln
. ,,.
ts r\
- . *. * ~
- conm 'i>>s SUETT Vl~VU SAfrtY Yl~VlS
. GRAVllY . .
.INSlRTION OPEli T~ m IJVE Res .mssuu OPhTO *wm
. $ECCIM~Ah . StSUli Patssuu
- PASSIVE .
CONTROL ROD REACTIVITY cowuo~
BYRON/BRAIDWOOD STAnONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.0*11 LOSS OF EXTERNAL LOAD
R*.E*.*.*v**.*.IS*s*r**.a
[)Ey~MBER . . . N*****. 1*.**1*
2006 s
..
~
..;
$~
! !
!..,..; . ;
- ...... "
.,,._,,
t=; ..
....""
~
,~
' "'"!.-
...
I!.
IYRON/BftAIDWOOD STA110NS UPDATED FINA~S.AfETV ANALYSIS R90ftt.
FIGURE 15.0-12 LOSS OF OFFSITE POWER TO STATION AUXILLi\AIES
REVISION 11 PECE:MaE:R 200.P
.ii
.,;
.--
- -12 B en I a
""'
.
i...
..."'
i :::!
~
~
......_,.. ...w--. 1
~ .
e: !E
. i5g BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFm ANALYSIS REPORT FfGURE 15.0*13 LOSS OF NORMAL FEEDWATER
.
N'
- s!
BVllONl9R~QWOOD ST~noNS UPDATED FINAL SAFm AHALYSJS
- , **--- *; ,_ . .*
REPORT FIGURE 15.0-14
. ~AJQfUIReAK OF A MAlN,FEt:DWATER LINE
LOSS OF FORCED* REACTOR COOLANT FlO' P7
.NO PROTECno:
REQUIRED p,
2/3 lOW RCS now IN. 274 LOOPS STEAM UNDER FREQUENCY 2}4 S1EAM GENERATOR GENERATOR
~fA SAFETY SAFETY VALVES VALVES RE"ACTOR PASSM: ~
.REACTOR PASS'iVE lRIP* TRIP BREAKERS BREAKERS c SAfEIY v..._YES W£!YV~YES
-u DPfH 1D llEU£VE SElXllilWn' OPDflO RELJM**~
0 S'ISIBI PR£5SUllE $\'SIDI..~
~ ~IA ~15'
~ (Tl~
- 0 ...,, ::0 CONTROL CONTROL 0 ;::::! ..... 0 RODS *ROos
..., >.,, Zz GRAVITY GRAYm
...... f!ri:J INSERTION INSERTION o-.,>o PASSIVE PASSIVE ooze U>~
o :;oo ::o >,....
s:.~ 1"11 * .,,0
........
z 0 0{Jl n ...... CCNJRQl. llOO REM:IMlY R£ACl1VIrr COtmilX. -
...,, ::0 ~*jg CONnlOl CCN1lml.
r rrt "'O I
>o
~**~
zO o>~ ..... > (,/)
- EO-iUl dfTlo !:<*~
(/) .......
- o, .......... 0:::0 0 <no z
- 0 (J) ~
~ f'T1 f'l'll;-J
,, ~
0 mo
~
~~
csi-
- ~*
$
O)'.
REVISloN 11 DECEMBER 2006 REACTOR COOLANT PUMP SHAFT SEIZURE {l..OCKED ROTOR)
(FULL POWER)
LOW .RCS .FLOW 2/3 lN ONE LOOP stEAM RESSURIZER GENERAfdR SAFETY SAFFJ'Y VALVES VALVES PASSIVE PASSIVE REACTOR TRIP BREAKERS SAFETY VALVES SAFETY VALVES OPEN TO RELIEVE OPEN TO .RELIEVE SECONDARY RCS PRESSURE SYSTEM PRESSURE CONTROL RODS GRAVITY INSERTION PASSIVE CONTROL ROD R£ACTIVlTY CONTROL BYRON/BRAIDWOOD STATIONS UPDATE FINAL SAFElY ANALYSIS REPORT FIGURE 15.0-15b REACTOR COOLANT PUMP SHAFT SELlURE (LOCKED ROTOR)
IJUUTIGLLEO Ila> cwSTEI CGllHL
&SS0181.Y 1111 wnllilltu&t.
.SlllCIU I IC&l " PCllffl 2/4 . "9u 11111&1 **Gil SIUKE-RbGE Rl&I llEUHm Fl.Ill llf.OTllOI FlUI. I /Z IElCTOI .
REACTOR. 1'11£SSUl1l£1 SIG l/. Of11TtMP£UTOU 41r TRIP WHY StFHT . ....
PafEll IAl&E lllGll IEUllOll .Sl'STOf IJLW£S VALY£S sYSmt FWI. Ullf. 2/'1 P&SSIWE P&SSl'IE ll* IUGR NtSSUllltl PIESSUll
's'1'*' u'- .I
'f*'f*'
' I IHCTDR Jiii' S&Ftn 'fauu tttr*.
c: m*lm mu ro 1tu£;£ at.uas.
- ~*
lC$ PIEsSulf *
~ g ls
"~r* ...
l:;z 8 nlm a<
' ~
cn-t ,.:a CGllHll llGO$ ** -**
- m :JJ -0 Gllltlif
. GRIJIIY S:*O zz US[HIClll Comal UJF >-
,...r- "Tl r-CD PUS.1'1£
-<m :a PISSIYE CDC i5 C'I>>
c
~*.* ~ :u 1:i6 m rn=:
"o OI ~g
- en
_r-: 0
..&
-. ~. o COIHROL ROii c01tloL RGO 5!i Cl) REKTIJllY o-1 IOCOVllf 0::0
- um ~*. !': COllllOl. C*TIOl
>;g -c> ~re UJ::! ,,,~
l8 (iO r-~
z: ~
~. ep* CDo
~ ,. m
- oZ r- 0
- -~
- u .... ..
-ti
- *~__,.
(SI
$
C>
DROPPED ROD CLUSTER CONTROL ASSEMBLY
- MANUAL REACTOR .TRIP BY PROCEDURE SttAM RESSURIZEA GENERATOR
. SAf'ElY SAFETY VALVES VAl.VES REACTOR. PASSIYE PASSIVE TRIP BREAKERS SNEIY VALVES .. .*.. . . . SAFETY V~VE'S. . .. ..... *. .
OPEN ro REUE.Vt OPEN TO REUtllt SECONDAAY' RCS PilESSllRt . SYS'(tll ~"~
t '
lS Ft
' ,
CONTROL RODS GRAVITY IHSERTIOH PASSIVE CONTROL ROD REACTIVITY CONTROL
- TRIP SEQUENCE OCCURS ONLY IF REQUl~ED BY PROCEDURE, O~ IF AN ENTIRE RCCA BANK DROPS, FOR OTHER ROD WORTH, OR FOR MISALIGNMENT, NO AUTOMATIC PROTECTION REQUIRED.
BYRONISRAIDWOOO STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 1s.o.17 DROPPED ROD CLUSTER CONTROL ASSEMBLY
SINGLE ROD CLUSTER CONTROL ASSEM8lY
- .WrrHoRAWA.L it FULL.*POWE*f **
OVERTEMPERATURE AT RE.ACTOR TRIP SYSTEM Ii'S F' I
' I REACTOR TRIP BREAKERS fi'S F' I
'- I CONTROL RODS GRAVITY PASSIVE CONT:lOL ROD REACTIVITY CONTROL BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANAL VSIS REPORT FIGU.RE 15.0*18 SINGLE ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL AT FULL POWER
B/B-UFSAR Figure 15.0-19 has been deleted intentionally.
REVISION 5 - DECEMBER 1994
BYRON/BRAIDWOOD STATIONS .
UPDATED FINAL SAFETY ANAL VSIS REPORT FIGURe 1$.0*20 BORON DILUTION
ii:I
- I'.
. I I
I
..
Ii:
i
! I ~.
- BYRON/BftAIPWOOD STATIONS UPDATED FIN4L. SAFETY ANALYSIS RJPOAT FIGURE 15.0*21 RUPTURE OF CONTROL AOP DRIVE MECHANISM HOUSING
BYRON/B8AIDWOQ[) . STATIONS UPOA1E FINAL SAFETY ANALYSIS REPORT FIGURE 15.0-22' INA.DVERTENT ECCS OPERATION AT POWER
R:EVISlON 1*1 D*ECEM£~ER 2006 ACCIDENTAL DEPlfESSURllATION OF REACtOR *coot.AMT SYSTEM
, ; ., ....... *....... * ... , , ,,,*.,:.-.* . .,._ - . . . . *, .. * .. *, _.
FULL PoWER OYERTEMPERATURE 4T 2/11 PRUSUR IZER .~OW PRESSURE'~ 2/4 RTS
, '
IS F I
' ,
REA.CTOR TRIP BREAKERS fS
~
F'
'
' '
CONTROL ROOS GRAVITY IMSERTIOM PASS IV£ CONTROL ROD REACTIVITY CONTROL POWER > P7 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15~0-23 ACCIDENTAL DEPRESSURIZATfON OF REACTOR COOLANT SYSTEM
111 ff I U
- 11 i . fl I !I ti 1,* i't *1 ti
- I;
~. ,, iiIi J.,*ul h *~* It
- I I{ I, Ill
~ ~ ~
I u
ftI SYRON/BRAIOWOOD STATIONS UPOATEO F'rNAL SAFETY ANALYSIS REPORT FIGURE 15.0-24 STEAM GENERATOR TUBE RUPTURE
~SIONt1 DECEMBER 2006 S8lOCA LOW PlR PRESSURE HI-3 CNl.IT (2 PRESSURE ESFAS 1PERATOR (CI-PHASE A)
LOW PlR (Sl SlCHAI.)
PRESS!m£ OPERATOR (CS)
HI-3 CHUT PRES. (2/4)
~lf1 LO-LO RWST l.&EL (2/4 WEIYY,llol.WS .
Gl'Df ltJ ltEllM * - - . :
Sl$]Dl*l'llE5$llli£.
MAIN fW ISOLAnON CONTROi. OPERATOR ...__...,._ _.
ROD ComROLS AFTER GRAVITY SG l£olEL. 6 HRS; INSERTION ~
PASSIVE
--.-- sws -
litM:IMlY*
-- OlltlltOI.
~
COfllt CGDIJlll c:
--N..
1J
~ ~
"-I sr.m:W5 5
II) a~
.,, ::0 IJ) 0 I
(/)
.... o
--
C'>
.,,l !!:: ....c.... ~~
O)> ::0 Or ITI
- OPERATOR ACTION INCLUDES VALVE Al.IGNMENT AND PUMP or ~~ CONTROL REQUIRED FOR RECIRCULATION.
~CCI UI
- ~*~
z::o 0
.&, :>0 FOOTNOTE: SEE TABLE 7.5-1 FOR INDICATORS AND RECORDERS
- ~ ..., ~f,/) AVAILABLE TO THE OPERATOR FOi.LOWING E.VENT*
0 UI n Q
....
0 ....~.~.....
Pl (I)~
~ :OU>
...,,,
0
~
L8LOCA
~(Cl-PHASE A) CONfAIHWENt
~ . (cs) c:oowc SYSJEM .
UH..O .. c:oNWNuENr lftiSr 1.£11£1.. ~'"' SPRAY' S'IS1EM (JM1.)
Ol'ERA10R -~--
COHlROl.S AFTER SG l.E.VEL. ~
~ * ~EHr SPRAY S'rSJEM
- . (liB:JRC,)
_,,.,,,.
--- ~ff1
-- . ._ , .
cm *a iiue *. r . aaa: ..as . _.., ,
c
-*-- __ __
- ':
"'O CJ
?;j 5 ""m 0-<
..mm=.
-
(/J
(/)
I ...., :!~
zZ 0
- OPERATOR ACTION. INCLUOES VALVE ALIGNMENT AND PUMP 0 ~';]; CONTROL REQUIRED FOR RECIRCULATION.
b~ c
- o OC> 1"1 . .. OPERATOR ALIGNS SPRAY PUMP SUCTION FOR RECIRCULATION QM ~
- ~~
Ul
- ~~ TO CONTAINMENT RECIRCULATION SUMPS
~~ isl 0
-I~ I >0 F001N01£: sEE TABLE 7.S""'."1 FoR. INDICATORS
- AND RECORDERs ;.
N z
(.;" Ul * ~~ AVAltABLE TO lHE OPERATOR F:OUOWINC;* EVENf. ..~:::
0 17
~~ .&
0 -o P1
-
~
U>z
- o Ill
"
....
"'O 0
- 0
-I
REVISION 15 DECEMBER 2014 100 150 200 TIME (SECONDS) w 2300 -
0::::
~ 2200 .
(/)
U1 UJ 2100 a:::
a.....--..
a::: . ~ 2000 ,_
,.
LLJ o_
~.__.. 1900 ...'
a::
- > 1800
(./)
(/)
w 1700
~
0....
1600 0 50 100 150 200 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-10 NUCLEAR POWER TRANSIENT PRESSURIZER PRESSURE TRANSIENT FOR FEEDWATER TEMPERATURE REDUCTION
REVISION 15 DECEMBER 2014 I I ;
' '
0 50 100 150 200 TIME (SECONDS) w a:: soo--~----------------------------------------.
=::i ....
~ 590"':> ..,
M:
~
sso- -~
~ "C:" 570- ,:-
LiJ.!!...-
~
Ct:
550-
....
---,
\
~
~
550-':
540"" ~
-~ ,~--~
- - - - -- - - - - -* -- ----- *-
- *-*
0 -
(_) , I l ' I I I I ' , * -
530 1 1 0 50 100 150 200 TIME (SECONDS) 6 5
Ct:: 4 co
- z Cl 3 1.,.___.....____...._________...._________________,_____-'-________,.
0 50 100 150 200 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1 -1 b REACTOR COOLANT LOOP flT TRANSIENT CORE AVERAGE TEMPERATURE TRANSIENT AND DNBR TRANSIENT FOR FEEDWATER TEMPERATURE REDUCTION
REVISION 15 DECEMBER 2014
~
0 ~
....
0.2- -*
0
- t '
. . I
'
0 50 100 150 200 TIME (SECONDS) 2300 ,_
1100 ......---* ---------,,.--------*----~.-*----i---l--~,-------*----*. . . . .
1 1 0 50 100 150 200 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-20 NUCLEAR POWER TRANSIENT PRESSIRIZER PRESURE TRANSIENT FOR FEEDWATER CONTROL VALVE MALFUNCTION
REVISION 15 DECEMBER 2014 1
.
->---
_____ ___
, ,I 80---------------------------------------------
..
00
~-
WL- i\
o~ 40* 1 a.... \
0 g
200 w
0:::
- >
soo----------------------------------------------
..
< 590- ----..~ .._.
ffi .:* ----- ----
§g 5ao- ~ ---,\
t:= 'G:' :
w~ 570- ..
~ : I, a::
~ 550- . ' -*-,' '*,,.
..........
w a::: 550< ***.. "*---**-*'" ---*~----*--**** -* * **--*---*--..**..--****-"*..***-*-*-*----*-*-- -*---..*-
0 u 540-+-_.,__.._ ..._ . ....._
. _, ----...i'-"--..--.'""- ' ~' --i..*----.----..:.._...----.,
0 50 100 150 200 TIME (SECONDS) 6..,.-------------~---------------------------~----.
1+------~----__,P----~....i-.-.i-_,_--.. ____~--------~------.~
0 50 100 150 200 TIME {SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-2b REACTOR COOLANT LOOP DELTA T TRANSIENT CORE AVERAGE TEMPERATURE TRANSIENT AND DNBR TRANSIENT FOR FEEDWATER CONTROL VALVE MALFUNCTION
REVISION 15 DECEMBER 2014 1.2
,,........
0 a::
w*- c
- .
~E 0.8'"' ...
00 0..Z ..*
O! ....... 0.6- ::-
l5 ~
~o U+:i
- > 0 0.4 ... ~
z2 LL.. 0.2- ...
""""'
0 ' ' .
J
.I I
- .
I 400 I
'
(~~~ONDS) 0 100 200 500 600 TIME 2500 LU a:: 2400
=:.
(/)
Ul w 2300
~
a.a a:: .iii 2200 -
~.!:; 2100 0::
~
tn 2000 .. **
ff] , ..,
'"
a:: 1900 a...
1800 0 100 200 400 500 600 TIME (affioNDS) w 1200
~
_.
- )
~ 1100 Ct:
w
~...-... 1000 3:!3 a::;~
w 900 N
0::
- >
en 800 Ul w
a:: ,
a.. 700 400 0 100 200 TIME (a~ONDS) 500 600 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-3 TEN PERCENT STEP LOAD INCREASE MINIMUM REACTIVITY FEEDBACK MANUAL REACTOR CONTROL
REVISION 15 DECEMBER 2014 620--------------------------------------------------~
~ 610
~ 600 f5 a...
~'t 590 W'C"
~ IU
~~580 w
~
w 570 a:::
0
(..)
560 5~:..._ _________,._________ ~~---------+----...._ __. . . . .________ ~
0 120 240 360 480 600 TIME (SECONDS)
J 2.a--
2.6 .. ~
2.......
~
m z.i- -
z Q
2 ~
LB--:-
l.6-"'
1.4 I
l .
f ' I I
. ' .
I 0 100 200 300 400 500 000 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1 -4 TEN PERCENT STEP LOAD INCREASE MINIMUM REACTIVITY FEEDBACK MANUAL REACTOR CONTROL
REVISION 15 DECEMBER 2014 1.2
,........,
Ct:
cc: 1
~*e o.s 00 a..Z
~~
o.s
~~
_, 0
(.J!;j
- > 0.4 '
ze 0
.,
~ 0.2 0
0 50 100
(~~ONOS) 200 300 25(}
l\ME 2500 :
w .*
tt 2400-:-
- l
(/)
Ul w 2300-~
0:: ~
a."O 2200- ....
w *-Ul 0::.
N8 210Q- -
er:
- )
Ul 2000-.:
(f) w ,...
er: 1900"'""
0...
1800
-*
'* t I
t I I .
I
.
I
'
.
200 300
(~t~ONDS) 0 50 100 250 TIME w 1200
~
30
> !100 0::
UJ
~ ......... \000
~!:J ffi~
N 900 a=
- l ..
ei 800 ..
Lt.J 0::
~ 100-+--------""f-------....,_----__...,..._______- t -_______.,________ ~
0 too TIME {~~ONDS) 200 250 300 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-5 TEN PERCENT STEP LOAD INCREASE MAXIMUM REACTIVITY FEEDBACK MANUAL REACTOR CONTROL
REVISION 15 DECEMBER 2014 s20-,-. - - - - - - - - - - - - - - - - - - - - - - - - -
w 510--
c::
- J
~ 600- ~
w
~ ~
wh:' 590-- -----------------------~
t- I LLJ 0\
(!) Q)
~~ sso-~
~ 570--
w a::
0
(..)
550~-
550 0
I I
50
.
t 100
.150* .*
200
.
I 250 300 TIME (SECONDS) 3.......--------------------------------------------------
a::
~ 2-~
~---------~------------------~......J l ...------*--..:------*__,.!--
- -----....
- -------*-+!-- ; ----.-+!-~1 ------.
0 50 !00 150 200 300 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1 -6 TEN PERCENT STEP LOAD INCREASE MAXIMUM REACTIVITY FEEDBACK MANUAL REACTOR CONTROL
REVISION 15 DECEMBER 2014 t1--------------------------
o.....-~'-*-----~:_._* --....--~*~:r---~-*--*-:.....-----* -*-.....-:*--*--*--+:--------1 O 50 lOO TIME (~~ONDS) 200 2500----------------------------------~----~-------
LIJ
~ 2400 (fJ f.:.1 2300
~
0..0 2200 ffi 'iii
!:::!~ 2100 0::
~
(/}
2000 LU g: 1900 i i : ~
1800 0 50 100 TrME (~~ONDS) 200 w 1200
~
3
~ 1100 Q:;
1.cJ
~~ 1000 -
?;:~
~.:!::,.
w N
900 a;:
~
(/)
800 1.cJ er::
~ 700.,._______....._______..._______.........,______-+-_ _ _ _ _ __,.__ _ _ _ _ _..,.
0 50 too TIME (~ONOS) 200 250 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1 - 7 TEN PERCENT STEP LOAD INCREASE MINIMUM REACTIVITY FEEDBACK AUTOMATIC REACTOR CONTROL
REVISION 15 DECEMBER 2014
~o----__.. ________________________________________,,_,
.......
550_..,_____..._._...,!_....__.._.-+-!--------~-*------*- : * -- * ---*~!~-------t 0 50 100 150 200 250 300 TIME (SECONDS)
Q:'.*
m z
o, 2
t5
\-+-__._...__.,.________..,.________ +--------------~~---- .....
0 50 fOO ?50 200 250 300 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1 -8 TEN PERCENT STEP LOAD INCREASE MINIMUM REACTIVITY FEEDBACK AUTOMATIC REACTOR CONTROL
REVISION 15 DECEMBER 2014 1.2
-c-c 0:::, c:
~*E (t8 oo o...Z 0::: ...... 0.6
~g U:p
- - 0 0.4
- ze LL
'-"' 0.2 0
250 0 50 100 TIME (~~ONDS) 200 300 2500 w
0::: 2400
=i
~
LtJ 2300 0:: "'
o...- 2200 ffi *~
t::!~ 2100 a:::
- >
U1 2000*
tn w
a:: \900 a..
1800 0 100 TlME (~~ONOS) 200 250 300 w 1200
~
3
~ 1100--
~<..-... moo-=.
~!!]
o:~
U..I 900"' -
N Lr
=> soo--
.
- wm
...
er 0.. 700
- I 50 I
f .
f
.
I f
'
JOO 0 100 TIME (~ONDS) 200 250 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-9 TEN PERCENT STEP LOAD INCREASE MAXIMUM REACTIVITY FEEDBACK AUTOMATIC REACTOR CONTROL
REVISION 15 DECEMBER 2014
~0----------------------------------------------..
~ 610
~ ~
~ y--.___________________________________________~
a{L' 590
..._ I La.I if
~~580 3(
w 570 a:
0
<..>
560 550....,_______________..,______..........,.....____________.--4_ _ _ _ _ __ _
0 50 IOO 150 200 250 300 TIME {SECONDS) 3------------------------------------------------
2.5- ..
ct:
Ql z
0
, ; t
"
I
" .
I
- . I . l
.
I 0 50 100 150 200 250 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1-10 TEN PERCENT STEP LOAD INCREASE MAXIMUM REACTIVITY FEEDBACK AUTOMATIC REACTOR CONTROL
REVISION 15 DECEMBER 2014 1.04
~
...__, 1.03
~
'-
0
........,
<..> 1.02
~
c: 1.01 0
~
0
.u 1 .
0..
- t=;
- s
- E 0.99 -
0.98 100 200 300 400 500 600 Coolant Average Temperoture (deg-F)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1- 11 KEFF VS COOLANT AVERAGE TEMPERATURE
B/B-UFSAR Figure 15.1-lla has been deleted intentionally.
REVISION 10 - DECEMBER 2004
B/B -UFSAR Figures 15.1 - 12 and 15.1 - 13 have been deleted intentionally .
REVISION 7 - DECEMBER 1998
REVISION 15 DECEMBER 2014
--e-2000....-------------------------------------------------~
u 0.
'-""'
0 0 10 20 30 40 50 Core Power (%)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1 -1 4 DOPPLER POWER FEEDBACK
B/B -UFSAR Figures 15.1 - 14a through 15.1 -** 17 have been deleted intentionally .
REVISION 7 - DECEMBER 1998
REVISION 15 DECEMBER 2014
-
Ci
- ~
0-2 0
z 0 0.15 c.
.Q u0
.._..
Lt: 0.1 x
~
j o.SE-01 u
~
50 JOO 158 200 Time (seconds) t:'
I jsso
- !:? 500 30 8.450
~400 QJ eJSO
~
cu JOO 8 250--+--------------~------------------------~
0 50 too Time (seconds) foul ltd Loop
- - - -
- Intact I.oops
~ o.8
- i
- z
"'a c:
°"
0 1~
e
~
(;: 0.2 e
! 9...-----------------------------.
0 50 100 150 200 Time (seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-18 1.4 FT 2 STEAMLINE BREAK OFFSITE POWER AVAILABLE UNIT 2
REVISION 15 DECEMBER 2014 0
0 50 100 150 200 Time (seconds) o......
0
--_............____......------------r------------------------1 50 100 150 200 Time (seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-19 1.4 FT 2 STEAMLINE BREAK OFFSITE POWER AVAILABLE UNIT 2
REVISION 15 DECEMBER 2014 c::
0
- .;::;
0
.....,
~
c:
Q)
CJ c:
0 u
c:
e 0
CD 0 50 100 l5D 200 Time (seconds)
JOO()
- E u
..__,,
0..
2000 1000
..
.c;.. 0
- s;
- J ..
(,.)
0 (I)
-1000 et:::
-2000
-3000 0 50 100 150 200 Time {seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-20 1.4 FT 2 STEAMLINE BREAK OFFSITE POWER AVAILABLE UNIT 2
B/B ~ UFSAR Figures 15 . 1-21 through 15.1 -26 have been deleted intentionally .
REVISION 7 - DECEMBER 1998
REVISION 15 DECEMBER 2014 1.4
_.....,
0 c:
.E 1..2 0
z
...._
0 Q. .....~..i.....-.+ _ _...__._ _....,_._ _ _ _._.._ _.._._ _.......,._____._.._ _........____
0 5 w 15 ~ 25 Time (Seconds) 1.4
.--..
0 1.2 c:
.E 0
z 0
c:
0 o.a
- .;:::;
--
<..>
0 l..L. 06 x
- =J G: o.4
.......
0 QJ
- c 0.2 0
0 5 10 15 20 25 Time (Seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-27 STEAM SYSTEM PIPING FAILURE AT FULL POWER - 0.95 FT 2 BREAK UNIT 1
REVISION 15 DECEMBER 2014 2D
--
.E fl)
_.....
Q..
2200
-
Ql
~ 2100
-
fl)
Q,)
a..
0 5 10 15 20 25 Time (Seconds) 5 10 15 20 Time (Seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.1-28 STEAM SYSTEM PIPING FAILURE AT FULL POWER - 0.95 FT2 BREAK UNIT 1
REVISION 15 DECEMBER 2014 600
--.
LI..
O'\
Q) 0
...........
e
- E
- t Q) c..
E
..,.._
Q)
Q) l-0 0\
520
.._
4.>
0
(._)
500 0 5 10 15 20 Time (Seconds)
- - - foul led SG
- - - - - totoc:C SC 1000----------~--------------------------------
1000
--..
0
- u; 950
- a..
.._
Q.)
- t 900 en
-
UJ Q) 0 ..
e
~
2
'--'
800
(.f) 150 700....___...__..._.,,___.___.._._______..,_____.._..____.__.......__.__.........,.
0 5 10 t5 20 25 Time (Seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15. 1-29 STEAM SYSTEM PIPING FAILURE AT FULL POWER - 0.95 FT 2 BREAK UNIT 1
B/B-UFSAR Figures 15.1 30 through 15.1 32 have been deleted intentionally.
REVISION 7 - DECEMBER 1998
REVISION 15 DECEMBER 2014
,..... \.Z
~} 1 0 fi .8 a,.Z
~l>
.6 ol
~!
A
.2 e.
0 20 .co 60 80 100 TIME (SECONDS)
- -
- RCP OuUel Pressure Pnsnurizer Pressatt 2800...--------------------------------------------
~ 2600 t:
-~ 2CXJO 1800-+-_______..,._...._..___- - t_________.,.__ _ _ _ _ __...,_ _ _ __....___
40 . 80
. 11.£ (SECQNOS)
L&J-22 11DJ-.---------------------------------------------
~ 110>
i~ 1400 P.irA 12t0 Bo*
a:::
- ..,_...__.._-""--+--""'______,,_..._,..__.__....__.__._~---'---'--..___,
1000 I~
a..- . 40 60 80 100 11t.£ (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-1 LOSS OF LOAD/TURBINE TRIP UNIT 1 - RCS PRESSURE CASE
REVISION 15 DECEMBER 2014
-g c 0.8
....
CJ 0.6
~
0 0....
lo..
0 D.4 Q) u::I z
0.2 0
0 20 40 60 80 100 120 Ttme (s) 2500
,.......... 2400 c
- u;
..e 2300 d>
""'
- ,
U'J Cl) 2200 o_
....
Q.l i; 2100 N
- c:
- i Ul In 2000 Cl>
d:
1900 1800 0 20 40 60 80 100 120 Time {s) 1500 g 1400 QJ E
- ,
g 1200 (5
0 3::
.... uoo O,.l N
- c 1000
- i t/'J en e
a.. 900 000 Q 20 40 5G 8(1 100 120 T\me {s)
BYRON/BRA! DWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-10 LOSS OF LOAD/TURBINE TRIP UNIT 1 - MSS PRESSURE CASE
REVISION 15 DECEMBER 2014 600------------------------------------------~
540 530..._...._...__._.,._________...,____ ~-..------~..--.------.
20 l1ME (SECONDS) l;!
I :.,
~'='
ik :
~
~.
570 m :.....________..,.________ - t -________....._ _ _ _ _ _---t..,._____- - . .
~ 40 60 BO TIME (SECONOS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-2 LOSS OF LOAD/TURBINE TRIP UNIT 1 - RCS PRESSURE CASE
REVISION 15 DECEMBER 2014
- - - Tio Tov9 620 '
,,.-......
Li-
,, \
ar s10
-..z
"'O E
600 /
I I
I \
\
'\
0 \
""""
Q) 0...
E 590
\
.
~
°§580 ._,,..,.... __
0
'-'
0 ----------
VJ 570 560*..,._...__________________________________________,,.
2E 0 20 60 80 100 Time (s) 1JSQ-------------------------------------------
- 0
--~ 1300 ,
1050+-.._._________,..,___....-._________.__.__.................._______,.
0 20 60 100 120 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-20 LOSS OF LOAD/TURBINE TRIP UNIT 1 - MSS PRESSURE CASE
REVISION 15 DECEMBER 2014 0.8
...........
c::
~
.....
cu Q.6
~
0..
0 0.4 cu u:::I z
0.2 0
0 20 40 60 80 100 120 Time (s) 2500 e::>
~ 2
~
Cl..
li; 21 N
- c
- I
~ 2000 Q) 0:
1900 1800 0 20 40 60 ~ 100 120 Time {s) 1700 g 1600 E ,soo .
'::l 0
- 1400 cu 0
- = 1300
.....
tU N
"§ 1200 Ul C/'l
~
Cl.. 1100 1000 0 20 40 00 00 100 120 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-3 LOSS OF LOAD/TURBINE TRIP UNIT 1 - DNB CASE
REVISION 15
- - - Tin DECEMBER 2014
- - - - - Tovg
-
u...
2' 610 '
,,. \
-
"'O
~600 ,'
I I \
\
\
,B I '
e cu 590 .
./ '
a_
E
~580 - ...... ________ ..,._
550....-.............--.__.--.....________________...._____________...._.
0 20 40 60 100 120 Time {s) s--------~---------------------------------
J 2
o,.._---..._,..------..------......... ..-.~ ....................._.___........_._~
0 20 -
60 100 120 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-4 LOSS OF LOAD/TURBINE TRIP UNIT 1 - DNB CASE
REVISION 15 DECEMBER 2014 1.2
~
D
~
~I c:
'
.I rzf 11..Z
.&
.4
.2 b
0 0 20 40 60 ID IOO TIME (SECONDS)
RCP Outfet Pre111Jre Prenurizer Prnsure 2100 fl D: 2&00
- t eI Q,. ......
.,.,
uoo 2200 u
QC 2800 -- ------~---111111m-...,,....~~ .............. ~ .....
1800 21 40 60 80 t 0 TIME (SECONDS)
..........
2"" 1800 3~
~w z 1&00
~::>
"'w i~
1'00
.
w j 1200 NC iiiE 1000
- lO 133 a:~ 800 CL- 0 20 40 60 80 too TIM£ (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-5 LOSS OF LOAD/TURBINE TRIP RCS PRESSURE CASE UNIT 2
REVISION 15 DECEMBER 2014 o.a
...........
-
c:
..£
.....
Q.)
0.6 ir; 0
a..
~
0 0.4 IU u::J
- z 0-2 0
0 20 4{J 60 8C 100 120 Time (s) 2500 '
2400
--...
0
~ 2l00 Q,)
5 2200
~
VJ Cl) ct 2100 .
~
q)
-~ 2000 -
- l en fil L.
1900 .
CL t800 1700 0 20 -40 60 80 100 120 Time (s)
-g cu 1400 E
- >
0
- >
'--
2 1200
~
~
.~
en 1000 (I)
QJ
~
0..
eoo 0 20 60 80 100 120
"° Time {s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-50 LOSS OF LOAD/TURBINE TRIP UNIT 2 - MSS PRESSURE CASE
REVISION 15 DECEMBER 2014
....8¥ 120
- >
a a.
- &,,,,..,
610 IOO wL&. 590 t-b S!.
a; 580 570
.;.;,
510
=
~ 550 0 20 40 60 10 100 TIME (SECONDS)
~
no 5 110
""
a..
600
~~
wl 590 c°' 580
~.g w- 570
~
_, 560 w
en U) 550
!::! 0 20 4' 0 6G BO 100 TIME (SECONDS) w a::
- J UOD ti 1.300
~
IL 1200 es ai:.-.
~I.
....
~
- 1100 IOOO 900
- &
- i 800 ..
f;; 0 20 ,0 60 ao 110 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-6 LOSS OF LOAD/TURBINE TRIP RCS PRESSURE CASE UNIT 2
REVISION 15 DECEMBER 2014
- - - 1 ; f'!
Ta v 9
-
Li...
en sw-r---~~~~-------------------------~---~--__,
I I \
\
--8 610
'
I
\
\
\
' \
'
.......
c:: 580 0
0
............ __ ..__.._,_...., __ _
0 u
c.n 570 u
a::::
560 0 20 60 BO 100 \20 Time (s) 1350
-- 0
- u; tlOO 0..
f:! l2SO
- i en VJ E
- a. !200 '
......
0
,.._J e
Q.)
1150 c::
QJ
(.,!)
1100 E
0
(\)
V1 \050 1000 0 40 60 80 \(JO 120 lime (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-60 LOSS OF LOAD/TURBINE TRIP UNIT 2 - MSS PRESSURE CASE
REVISION 15 DECEMBER 2014 0.8
...-
c:;
~
Q.6 Q;
- &\:
0 a...
....
0 0.4 Q,)
u~
- z 0.2 0
0 40 60 80 100 120 Time (s) 2500 .
Q)
$
~ 2 cu
"-
CL a:; 2100
.!::!!
~
~ 21XXJ c;i
'-
CL IQOO 1800 0 20 40 60 Ir) 100 120 Time (s) 1700
~ 1600 .
...__
QJ E 1500
=i g
L...
1400
..s0 3C 1300 L...
OJ N
'c::
- i t200 .
tl'I U1 cu ct 1100 1000 D 20 40 60 30 100 120 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-7 LOSS OF LOAD/TURBINE TRIP UNIT 2 - DNB CASE
REVISION 15 DECEMBER 2014
- - - t* n
- - - - - T ov9
,,
6ro....--------------------~-------------------...
I \.
I \
I \
\
I \
~ ~ I \
B I 0 .. /
'-
~ 590 *-
E
-
~ 580 .
c:
0 0 570 .
0
.. ~---~------
u
~560 0:::
5~.+-.--i ........~----,.L._.--...-..._...,--..i..-..;._, ____......._...................__........- t 0 60 100 1.20 Time {s)
~
CD z
0 2
0...-------..-------........ ...._--~-------..,.._. ____._,,________
0 60 80 100 120 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-8 LOSS OF LOAD/TURBINE TRIP UNIT 2 - DNB CASE
B/B-UFSAR Figure 15.2-9 has been deleted intentionally.
REVISION 16 - DECEMBER 2016
REVISION 16 DECEMBER 2016 2800
- i
! 2600
~
- > 2400 ffi a:
a.. 2200 a:
~
- > 2000
"'
f.3 0::
El. 1800 1600 G 1 2 3 4 10 10 \0 111 11l TIME (SECONDS)
~ 2000
- i 1800
--
tJ w
- E 1600 3 1400
~ 1200 a:
w I- 1000
~ 800 a::
w tl! 600 a::
- ,
Cl.I 400 ti.I w 200 a:
a.
0 0 , 2 3
- 10 10 10 10 10 TIME (SECONDS)
BYRON STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-9a PRESSURI~ER PRESSURE AND WATER VOLUME TRANSIENTS FOR LOSS OF OFFSITE POWER
REVISION 16 DECEMBER 2016 2800
- 2600
- ~
a..
............
w 0:::: 2400
- >
~
LL.I 0::::
0... 2200 a:::
w
~
- > 2000
(/")
r.n LL.I a:::
a.. 1800 1600 0 1 2 3 4 10 10 10 10 10 TIME (SECONDS) 2000
- ct
- i
(..;) 1600 1800
............
w
- IE 1400
- >
g 1200 a:::
1..1.J 1000
!<
31:: 800 a::::
w N 600 C2
- >
Vl 4{]0 l:fl O:::'.
0... 200 0
0 t 2 3 4 10 10 10 10 10 TIME {SECONDS)
BRAIDWOOD STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-9a PRESSURilER PRESSURE AND WATER VOLUME TRANSIENTS FOR LOSS OF OFFSITE POWER
B/B-UFSAR Figure 15.2-10 has been deleted intentionally.
REVISION 16 - DECEMBER 2016
REVISION 16 DECEMBER 2016 ii:' 680 SAT e.i> 660 w 640
~
- )
620
~
UJ 600 HOT Q.
- E 580 w
...a.
""" 560 8
..J 540 COLD 520 500 0 l 2 3 4 10 10 10 10 10 TIME (SECONDS)
.......
ftl 1600
'iii
......
Cl.
1400 w
0::
- > 1200
~
w a::
Cl.
1000 ct:
g 800
~
w 600 zw
(!) 400
- IE
~
en 200 0 t . 3 0 1 4 10 10 10 10 10 TIME (SECONDS)
BYRON STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-10a LOOP TEMPERATURES AND STEAM GENERATOR PRESSURE FOR LOSS OF OFFSITE POWER
REVISION 16 DECEMBER 2016 680
'W:.' s60 L~SA:~iT'- - - -- - - -""'
e.. °" 640 cu LLI D::: 620
"""\~
~
WOOD JifOT
~-
_,, ____ .,, .. ___ ,.,, ... '------ ---- ----------
,,...,, -~- --
0...
a5 580
, ___ _______ __ ___ _
1--
(
560 I 0...
g MO OOl.D 520 500
()
10 10 f
10 2
10 J
- 10 TIME (SECONDS) 1600 0.(ii 1400 a...
.............
w D:::
1200
- >
U")
~ 1000 D:::
0...
a::: 800 0
~
- z 600 LLI
(.!)
400
~
tn 200 0
10 0
10 I
10 2 3 10
- 10 TIME (SECONDS)
BRAIDWOOD STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-10a LOOP TEMPERATURES AND STEAM GENERATOR PRESSURE FOR LOSS OF OFFSITE POWER
REVISION 10 DECEMBER 2004 2600 I 2400 II!
I 2200 1-1800 1800+---.._--~~...,_ _ _............._..............._~~~.................-_._~.........................,____._.._.........,.........~
1.E-01 1.E+OO 1.E+01 1.E+02 1.e+o3 1.E+04 TJlllE (SECONDS) 1800 1800
- ~ 1400 1* 1200 .r-----------..__--~
. m 1000 I
I: 400 200 O+.--.._....._......._......_....,__ _,_........._._........,....,__~_._ ........................_ _._~~......._~,____._._.......,_.........~
1.E-01 1.E+OO
- 1.E+01 1.E+02 1.E+03 1.E+04 TIME 18ECOND8J BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-11 PRESSURIZER PRESSURE AND WATER VOLUMETRANSIENTS FOR LOSS OF NORMAL FEEDWATER
REVISION 10 DECEMBER 2004 700 880 SAT 880 G:' 640 t
iii 820 HOT
~
1800
..
~
580 9 580 COLD 540 1120 800 1.E..01 1.E+OO 1.E+01 1.E+02 1.E+03 1.E+04 TIME (BECONDS) 1200 i 1000 ll!
I 800 I
m 800
! 400 200 O+-~~~~~...,_~~~~~ ........~~---~~........,---__..~__...~..........+--~'--.._.~~~
1.E*01 1.E+OO 1.E+01 1.E+02 1.E+04 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-12 LOOP TEMPERATURES AND STEAM GENERATOR PRESSURE FOR LOSS OF NORMAL FEED WATER
REVISION 9 DECEMBER 2002
-
IX .c L&J-31:
0 E
1.4 1.2 oo 1 a,.Z
.8 IX-
~:
o-2
- 6.
- 4
=>u Zf .2
-~
0 10
_,
10 0
10 I
10 2
10 3
10 '
TIME (SECONDS)
.6 ....-----------~~--------------~-----------------,
- 4
.2 0
-.2
-.4
-.6 4----L-J.....J...L..U...U~-L..~"-L..J...u.i.+---'-...........1..U..u+--~...........u..u-'+---'---'-...L-L.J...U.Li
-I D I 2 3 4 10 10 10 10 10 10 TIME (SECONDS) 1000 ....---------------------------------- -------------,
~ 0 -'--------------.
2::~
._......._
~..:it-1000 L&J I a:~ -2000
..J Ill
~c b- -3000
._
-I 0 1 2 3 10 10 10 10 10 10' TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED rlNAL SAFE:TY ANALYSIS REPORT FIGURE 15.2-13 NUCLEAR POWER, TOTAL CORE REACTIVITY AND FEEDLINE BREAK FLOW TRANSIENTS FOR MAIN FEEDLINE BREAK WITH OffSITE POWER AVAILABLE
REVISION 9 DECEMBER 2002
....
0:
2600
- >
Cl) 2400
....
Cl) 0::
D.-
2200 0:: .2
--
La.I N
0:
- >
Cl) fl) rn D..
2000 1800 1600 La.I 0::
Q. 1400
-I 0 I 2 3 10 10 10 10 10 1 0' TIME {SECONDS) 120 Lt..
La.I
- J 100 La.I
. o::-c;- 80 0: CD La.I rn N ....... 60 02 E
- > .0 40 en.:::-
Cl)
La.I 0:
20 a..
0
-1 0 1 2 3 4 10 10 10 10 10 10 TIME {SECONDS)
....
- E
- >
..... 2000 0
>
IX 1500 L&J
~- 1000
==
a::-
!:?
~
Lo.I N
02 500
- >
Cl)
(/)
Lo.I 0 IX -1 0 I 2 3 4 a.. 10 10 10 10 10 10 TIME {SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT rlGURE 15.2-14 PRESSURIZER PRESSURE, PRESSURIZER RELIEF AND PRESSURIZER WATER VOLUME TRANSIENTS FOR MAIN FEEDLINE BREAK WITH OFrSITE POWER AVAILABLE
REVISION 9 DECEMBER 2_002
- - - Thot
- - - Tcold
-*-*-* Tsot
-1 0 1 2 3 10 10 10 10 10 10 '
TIME (SECONDS)
- - - Thot Tcold Tsot 700....------------------------------------------------.
- -*-*-*-*-*-*-*-*-*-*-*-*-*-*, .._,..,*- .......... _'
-1 1 2 10 0
10 10 10 10 3
10' TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED rlNAL SAFETY ANALYSIS REPORT FIGURE 15.2-15 FAULTEO ANO INTACT LOOP COOLANT TEMPERATURE TRANSIENTS FOR MAIN FEEDLINE BREAK WITH OffSITE POWER AVAILABLE
REVISION 9 DECEMBER 2002
- 1. 4 1*2 Xe
- >-
-
'"6
...J E a.o
,_z .8 c(.._
LLI 0
- c c LLI 0
- 6 IX:;:::
OU oe
-
la.
- 4
.2 0
-t 0 I 2 3 4 10 10 10 10 10 10 TIME (SECONDS)
FAULTED
- - - INTACT 1400 LLI 0:: 120 0 r--*****
- ) I
(/)
(/') I LLI 0::
1000 /
Q.. I a::_ 800 Oo t--
c(
IX Q.
z.....
LLI
(!)
-fl) 600 400
~
c(
.....
..... 200
(/)
0
-1 0 I 2 3 4 10 10 10 10 10 10 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED rlNAL SAFETY ANALYSIS REPORT FIGURE 15.2-16 CORE HEAT fLUX AND STEAM GENERATOR PRESSURE TRANSIENTS FOR MAIN FEEDLINE BREAK WITH OFFSITE POWER AVAILABLE
REVISION 9 DECEMBER 2002 1 .4 .....-~~~~~~~~~~~~~~~~~~~~~~--,
1.2
- 8
.6
.4
.2
_, 0 3 10 10 10 I
10 2
10 10 '
TIME {SECONDS)
-
311:0 o.E
- 6
.4
-'E
.... 0
~z
.2
~-
IX 0 0
me
-.2 c -0 La.lb
....... -.4
-
La.ID
.....
-.6
-I 0 I 2 3 4 10 10 10 10 10 10 TIME {SECONDS)
~
>-
,
....U_x-1000
--X i5 I o::,E -2000
...J ID c:( 0 b-
.... -3000
-4000 -J--L-L...W..LI.J..l-l---'-...1.....1....W...L..Ll.J-~..L..-L...J.-L...U-Uf---L-1.....L....L..U...U-f---L-.....l-J...J...J...i..u..t
-t 0 I 2 3 4 10 10 10 10 10 10 TIME {SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-17 NUCLEAR POWER, TOTAL CORE REACTIVITY AND FEEDLINE BREAK FLOW TRANSIENTS FOR MAIN FEEDLINE BREAK WITHOUT OFFSITE POWER AVAILABLE
REVISION 9 DECEMBER 2002
...... 2600 a::
- >
Cl)
Cl) 2400
......
a:: 2200 a..-
a:: .2. 2000
--
...... 0 Na.
a:::
- >
Cl)
Cl)
LI.I 1800 1600 a:: 1400 a..
10
-I 10 D
10 I
10 2
10 3 ..
10 TIME (SECONDS)
..... 120
......
- 100
......
a:::- u 80 0:: CD l&J rn N' 60
-e a:: .a
- > _
en- 40 Cl)
LI.I a:: 20 a..
0 10
-I 10 D
10 I
10 2
10 3
10
..
TIME. {SECONDS)
......
- E
- > 2000
-'
0
>
0:: 1500
......
I-
<-
31:~ .._ 1000 o::-
......
N iX 500
- >
VJ VJ
...... 0 a:: -I 0 I 2 3 4 ll..
10 10 10 10 10 10 TIME {SECONDS)
- BYRON/BRAIDWOOD STATIONS UPDATED f'INAL SAFETY ANALYSIS REPORT FIGURE 15.2-18 PRESSURIZER PRESSURE, PRESSURIZER RELIEF ANO PRESSURIZER WATER VOLUME TRANSIENTS FOR MAIN FEEDLINE BREAK WITHOUT OFFSITE POWER AVAILABLE
REVISION 9 DECEMBER 2002
- - - Thot.
Tcold Tsot 650 *-*-*-*-*-*-*-*-*-*-*-*-*- \.
- - *-*-- . ..._./ .
/*-*-*-*-., ,,_ ..
/
/
600 550
_, I 2 l 10 10 10 10 10 '
TIME (SECONDS)
- - - Thot
- - - Tcold
-*-*-* Tsot 700------------------------------------------------~
....,
0::
~ *-*-*-*-
....,
650 \. *...._*-*-.. . ....__/* / ' *r**
ll.
...., ....
~
.,_I 600 ll. GI 0 "tJ o"-'
_J Cl
,,
....
0 550
~
~
-I I 2 3 10 10 0
10 10 10 10' TIME {SECONDS}
BYRON/BRAIDWOOD STATIONS UPDATED f'INAL SAF'ETY ANALYSIS REPORT f'IGURE 15.2-19 FAULTED AND INTACT LOOP COOLANT TEMPERATURE TRANSIENTS FOR MAIN FEEDLINE BREAK WITHOUT
. OffSITE POWER AVAILABLE
REVISION 9 DECEMBER 2002
- 1. 4
- 1. 2
. -::::-
D
>< c
- >-
......
a...o E
.,_z .8
~
- c
...
0 c *6 a::=
l.&I 0 OU o~
-La.
- 4
.2 0
-1 0 I 2 3 10 10 10 10 10 1 0' TIME (SECONDS)
FAULTED
- - - INTACT 1400 LI.I Q:
- > 1200 I en I en LI.I 1000 I Q:
a.. I oi:_ 800 I Oo /
~u; Q: Q.
1.1.1- 60 0 zLI.I C) 400
- IE
~
l.&I t- 200 en 0
-1 0 1 z 3 4 10 10 10 10 10 10 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.2-20 CORE HEAT FLUX AND STEAM GENERATOR PRESSURE TRANSIENTS FOR MAIN FEEDLINE BREAK WITHOUT OFFSITE POWER AVAILABLE
REVISION 15 DECEMBER 2014
,--...
.E 2350 c.n
- c...
..._
~ 2250
.\::
- l en en Q) ct 2200 2150...,_._._____________________________,,..-----------------.1--t 0 2 ... 6 8 10 Time (seconds)
- - - c o r e Flow
~faulted Loop f l ow
- 1.2....--------------------------
0
.Ec 0
-
- z 0
c 0.8 .........
0
- ..::; .... _
u 0 o.s -- - - -- ---
-
\...
LI....
-.
3:
0 o.4 G:
.......
c::
0 Q.2 0
0 u
2 4 6 8 to Time (seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-1 PARTIAL LOSS OF FORCED REACTOR COOLANT FLOW
REVISION 15 DECEMBER 2014
-.E 0 c:
1.2 0
z
'+--
0 0.8 c:
0
- .;::::;
u 0
L.. o.s
...._...
I..&-
L Q) 3': D.4 0
0....
L..
c (J) 0.2 u
=>
z 0
0 2 4 6 B 10 Time {seconds)
J 1.5 1
0 2 3 .. 5 6 7 Time (seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-2 PARTIAL LOSS OF FORCED REACTOR COOLANT FLOW
REVISION 15 DECEMBER 2014 2200....__,__.._...____________._...__.....____,___ -&,._.....__,,__.a.-..._..._.~
0 2 4 6 8 10 Time (seconds) 1.2
~
cc:
- e0 z
- o.a 0
c:
0
- .;:::;
v
._ 0.6 0
-
l.&-.
3:
0 OA w::
_,.,
c:
0 0 0.2 0
CJ 0
0 2 4 6 8 10 Time {seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-3 COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW
REVISION 15 DECEMBER 2014 1.2
- ...........,
0 c:
- e 0
-z:
0 c::
0 o.s
- .;::::;
(.)
0 Q.6
~
'- ~
Q)
0 04 a...
'-
u;::,~ 0.2
.:z 0
0 2 4 6 8 lO Time (seconds)
J-- ---------------------------------------------.
c::
~ 2 0
1.S 1
0 2 3 4 5 7 Time (seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-4 COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW
REVISION 9 DECEMBER 2002
- - - Neutron Flux
- - - Hot Channel Heat Flux
-0 c
e 0
><z .8
- > .... ......
......
...J 0 L&.c
.... .....
-se u
- 6 .... ....
-
-..._ .4
- 2
--
0 0 2 4 6 8 10 TIME {SECONDS) 2000 La.I 0:::
- >
~
....,...._
o:::-
Q.. O>
- E ID 1500
..,.._
LIJQ Q
- su 1000 0 2 4 6 8 10 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPOATEO flNAL SAFETY ANALYSIS REPORT f"IGURE 15.3-5 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR
REVISION 9 DECEMBER 2002 2800 2700
.....
_
0:::
- >. 2600 en
.....
en 0::: 0 ID..- 2500 lll en c.
u-0:::
~ 2400
- 2 2300 2200 0 2 4 6 8 10 TIME (SECONDS)
Core Flow
- - - f culled Loop flow
-c 31= .!:
oE
_, 0 I.&.. z .5
.___
- z 0 \
- 5 c 0.2 o-U u
e
~
........ -.5
' '
------ - - - - -
0 2 4 6 8 10 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.3-6 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR
REVISION 9 DECEMBER 2002 BYRON I BRAJDWOOQ STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE JS.4-1 NEUTRON FLUX TRANSIENT FOR UNCONTROllED ROD WITIIDRAWAL FROM A SUBCRITICAL CONDmON
REVISION 9 DECEMBER 2002
---:---
E 0 .J
-
z 0
_....,:.
<..>
-
0
~
l...L_
.2
><
- >
....._
_J
.1 25
- r:
0 5 10 15 20 25 Time (s)
BYRON I BRAIDWOOD STATION UPDATED FINAL SAFETY ANALYSIS REPORT AGURE 15.4-2 THERMAL A..UX TRANSIENT FOR UNCONTROLLED ROD WITHDRAW AL FROM A SUBCRITICAL CONDITION
REVISION 9 DECEMBER 2002
'G:'
cn 700
-a:
Q)
Cl
- .:2:
t::=! 650 0:::
w
- z:
- z:
600
- s Cl
{_)
550-+-____ _...._...--+__.~..__..__...._..,.__.__._......._._--+-_....__....__....._~..,_..._......__.__._-I 0 5 10 15 20 25 Time (s)
BYRON I BRAIDWOOD STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE IS.4-3 FUEL AND CLAD TEMPERATURE TRANSIENT~
FORUNCONTROLLED ROD WITHDRAW AL FROM A SUBCRITICAL CONDffiON (SHEET 1 OF 2)
- REVISION 9 DECEMBER 2002 2200
..-- 2000 LL-O'l (l) 1800 0
--: 1600 CL
- E LLJ I-1400 LLJ
(.!) 1200
<:
0:::
LLJ 1000
~
_J LLJ 800
=>
LL-600 400 0 5 10 15 20 25 Time (s)
BYRON I BRAIDWOOD STATION UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-3 FUEL AND CLAD TEMPERATURE TRANSIENT~
FORUNCONTROLI.ED ROD WITHDRAW AL FROM A SUBCRITICAL CONDITION (SHEET 2 OF 2)
REVISION 15 DECEMBER 2014 1.4---------------------------.
.__
0 c:
0
- _;:j u
o.s e
u...
.._.... 0.6
"'-
Q)
£~ 0.4 . . . . . . . . . . . . . . . . . . . . . . . . . . * . . .. .. - . - ' .
o-+-----------------------------------------------------~
0 2 4 6 10 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-4 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVllY FEEDBACK
{80 PCM/SEC WITHDRAWAL RATE)
(SHEET 1 OF 3)
REVISION 15 DECEMBER 2014 t.4....---------------------------......
-::=--
g 1.2 E
0 z
1 c::
0 1:> 0.8
~
lJ_
...............
x 0.6
- 3 CL:
.,...,_
0 Q,) 0.4
- c 0.2 ........_ _ _.............,._ __.._........,_.--_________-.-____________________..
0 2 4 6 8 10 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-4 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (80 PCM/SEC WITHDRAWAL RATE)
(SHEET 2 OF 3)
REVISION 15 DECEMBER 2014
...--... 620 ..
u..
I . .
~ 610"" . .. *..* . ... : ... . ... . .... . ' . . . . . . . . . . . . . . . . . .... ... ...... . .
,__
01
<I)
-0 . . . . *
............... 600"' . . * . .. . . . . . ' : . . . . . . . . . . . : . . . . . . . . . . . : . . . . . . . . * . * : . . . . . . . * . . .
~
.:::l : _ _..;.----*-----... : :
15 ,__ 590--...-....................~ . ... .. .. . . . .. . * ~ . - . . . . .. .*... . . . ... .. . . .
.
<1) * * -
- CL * . . .
~ 580- *.. .. ..
1--
Q.)
O' .*
0 570- ~
. . . . . . . . . . . .. . . . . . . . . . . . .. . . .. - . . . . - - .. . . . . . . . . . . . .. . . . . . . . . . . .
ti.- * * * .
(1) "
~ 560- "' . . . . .. . . . .. : ....... . . .. : . . . . . ...... : . . . . . . . . . . . : .... .. .. . . .
~
0 u
550-+---------~,....-~ ' --~~ * --,.--,--......._-----T,----------.,----------t 0 2 4 6 8 10 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-4 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK
{80 PCM/SEC WITHDRAWAL RATE)
{SHEET 3 OF 3)
REVISION 15 DECEMBER 2014 2600
~
0
. ~ 2400 ........ ' . . . . . . . . . . . . . ' .. . ... ' . ' ... . .. . .. ' . . . . .... ' .... .. .
....__,
,_
(l)
N
- c::
- s (I) 8l 2000 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......... '
""-
a....
1aoo....---------.
0
. . 2. . -----------.----------..-.-------------------_.
4 6 8 10 Time {s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-5 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (80 PCM/SEC WITHDRAWAL RATE)
(SHEET 1 OF 3)
REVISION 15 DECEMBER 2014
.--.. 1400..---------------------~-----------------------------~
..,.._>
Q)
<J) w...
(..)
- .0: l 1200- .*...... .. . : . ....... , . . . : . ' ...... . .. : ... . .... . . . : ..... . .. . .
~
u
.............
Q.)
E
...2
.g 1000-
""--
~ ~- , ....... , , . . . . . . ......... . . .. . .. . ........... ... .. . . ... . ... .
- c:
- J
(/)
(./)
Q.)
5..-
a..
600-+---'-.......-----...-.~*--~'----*.-----~'----~* ...--:-------...----------1 0 2 4 6 8 10 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-5 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK
{80 PCM/SEC WITHDRAWAL RATE)
(SHEET 2 OF 3}
REVISION 15 DECEMBER 2014 5------------------------------------------------------..
4 -*** * * ** ****** * ***** * ** * *** * . * *********** * **** * **** * - * *** **
a:::
~ 3 .. . .. .. . . . . . .. . . . . . .. . . . . ... . . .. . . . ' .. .. . - . . . .... .. ...... .
a 1-+-_.________,_______________.....______..,__________-......_________.
0 2 4 6 8 10 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-5 UNCONTROLLED RCCA BANK WITHDRAWAL FRO~
FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (80 PCM/SEC WITHDRAWAL RATE)
(SHEET 3 OF 3)
REVISION 15 DECEMBER 2014
~- 1.2 0
c::
E 0 1
- z
"+-
0 c: 0.8 0
~
(..)
0
'- 0.6
...__..
LI-
'-
<l.)
3= 0.4 0
a_
'-
0
<L>
0.2
(.)
- >
z 0 0 10 20 30 40 50 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-6 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (3.0 PCM/SEC WITHDRAWAL RATE)
{SHEET 1 OF 3)
REVISION 15 DECEMBER 2014
...---..
c c: 1.2 E
0
- z
.....__ 1 J:__--.-.-~--:-:~~:-:---:-:~~~~~~-- . ' . . . . . . . . , . . . . . . . . . . . .
0 c::
0
~
.+J
(.)
o.s 0
~
..___...
x
- J o.s LL-
..........,
~ o.4 * * * * * * * * * * * * * * . * * ~ * * * * * * , * * * * * * * * * * * * * * * * * * * * *
- r::
0.2..,._.......__......_...__..............,._ _ _ _.......,._ _- - i_ ___,r--......_......_......_...,_...._______...,.
0 10 20 30 50 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-6 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (3.0 PCM/SEC WITHDRAWAL RATE)
{SHEET 2 OF 3)
REVISION 15 DECEMBER 2014 620
..--..
LL.
t 610 . . . . . . . . . . . . . . . . . . ' . . . . . . . . . . . . . . . . . . - . . . . . . . . . . .
~
Q)
....__...
-0
<J.) 600
"-
- )
0 _____;__.---
~ 590-t-..:.*...........----- -*...--:"". . . * . . . . * . . * . . . . . . . . . . . . . . * * . . . . . . . .
- CL E
~ 500 ... .. ............... . .... . . . . .... . . . . .... ... ... .
<J.)
C°)
~ 570 . . . . . . . -. . -.. . . . . .. .. . . . . .. . . . . . . .. .. . ' . . . . . . . . . .
>
<
~ 560 ...* . .. . . . ..... . .... .. .. . .. . .. . ..... . . .. . ... .....
0
(._)
550-+------------------------.------.
0 10 20
. . . . . --------------.----------1 30 40 50 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-6 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (3.0 PCM/SEC WITHDRAWAL RATE)
{SHEET 3 OF 3)
REVISION 15 DECEMBER 2014 2600
.....--
0
.R
...__.-
2400 ******** ********************. *********-**********
~
- J en en cu ct 2200 1aoo------------.---------------------r---------.._..
20 30
______ --t 0 10 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-7 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVIlY FEEDBACK (3.0 PCM/SEC WITHDRAWAL RATE)
(SHEET 1 OF 3)
REVISION 15 DECEMBER 2014
_.--.... 1800
.........
Q)
<J.)
LI,_
u 1600
.0
- s u
"--""
Q) 1400 E
- s
.g 1200 L------------------------
- * * * .. * ' .. .. .. . .. .. ... * * * * * ' * * * * * * "' ' . *
- 4 * * * * * - * * * " .. '
- 111 ... * .. * "
80() * *** * * * * * * * * * * * * * * ** * * * * * * * * * * * ** ** * * * * * * * ** * * * *
- 600 0 10 20 30 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4- 7 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK (3.0 PCM/SEC WITHDRAWAL RATE)
(SHEET 2 OF 3)
REVISION 15 DECEMBER 2014 5
4 ******** * - * * * ***** * ** * * *** * ** ** * -**** * * * * * * * * * * * **
. . . ..
2~ * . . . ..
.. -
. : . .... .. .. : *. ..... . .. .: . ... .. *. " . : ...
....... ~--.i,;,- - . * : ..
j* . . .. .
- -:----~ :
- : * ~~----
1 0 10 20 30 40 50 Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4- 7 UNCONTROLLED RCCA BANK WITHDRAWAL FROM FULL POWER WITH MINIMUM REACTIVITY FEEDBACK
( 3.0 PCM/SEC WITHDRAWAL RATE)
(SHEET 3 OF 3)
REVISION 15 DECEMBER 2014
"""~- -.----- --~--;ww::s.; :;:;tu_,,,.- ---- ~ .....x ....,.. -~,..--=:::::::;*;;;c;...,..,,k....,- =- '""'*'~:-,--~ ~""
-Maximum ResctMty Feedback
- - ........... Reacti\iy Feedback OTATlrip 1.7 --*--**
a:
16
. ~-
'
I
! 1.5
- &
z i
1.4 .
'
' *\./* /
1.3 1.2 ------+--___________,....__________,.
0.1 1 10 100 1000 lnsarlon Ratlt (pcm/ second)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-8 UNCONTROLLED RCCA BANK WITHORAWL FROM 100% POWER MINIMUM DNBR VS REACTIVITY INSERTION RATE
REVISION 15 DECEMBER 2014
- ----- --* *--r~. =~--~~ ~==="'-----, ------ -~---- **-*-- *--*-* . ------***--** *1 2.2 ---------*----------~--...-- - -- -- - ..
SG v.Jves open before trip OTATtrip 1.6 ** *-- ~- ***-- .. _, ...... . . .... 4
'
..... - -* ___..
igb Flux trip 1.4 ...---------+-_______,,,_______,_________...
0.1 1 10 100 1000 Insertion Rate (pcm/ second)
, ___...,.,,'
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-9 UNCONTROLLED RCCA BANK WITHDRAWAL FROM 60% POWER MINIMUM DNBR VS REACTIVITY INSERTION RATE
REVISION 15 DECEMBER 2014
- 1**~,*:~~;;;:~~;1-- --- --- ----- ---- *1 OTAT trip i SO Valves opeo
! bctm 1r1p .................... J 2.8 ~~-~ i ~------+---+----- I I I I 11 r4 I
~,
I
!
- E z
2.2 - -
I J
I
...........,...--ii.--___,.........,__.___ ---=-1 f i 2 - ..-. .. ... *---* *-* .. " . ... ~ - . . ***- .... ~
i 1 .8 * ~ * ""' * * --- ~ - ~i - * ---l
!
l 1.6
- eai;; open \ ..a-_ igh Flux trip J I
1.4 *-~------------------------'
- 0. 1 1 10 100 1000 L------~- ---*~ ..._---.----- -- *-~* . . ~-- * . -
tnsertlon Rat. (pcm/ second)
.. - -_- . .--- - - - --- -------- -- *--~---. -- ---*-_,_
J BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-10 UNCONTROLLED RCCA BANK WITHDRAWAL FROM 100% POWER MINIMUM ONBR VS REACTIVITY INSERTION RATE
B/B -UFSAR Figures 15.4 - 11 and 15 . 4-12 have been deleted intentionally.
1.2
-_,
<(
1.1 a:~ 1.0 w ::E 3=o
~z 0.9 a: u.
<o wz 0.8
-'o
(,,,)
- >
zU
-I-
<( 0.7 a:
u.
0.6 0.5 1.2
- 1.1 x~ 1.0
- J.-
_, ::E u.o 1-Z 0.9
<( u.
we
- c wz 0.8 a:2 ct-(J (J
<( 0.7 a:
u.
0.6 0.5 0 50 100 150 200 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4*12a TYPICAL TRANSIENT RESPONSE TO A DROPPED RCCA (OR RCCAS) IN AUTOMATIC CONTROL (SHEET 1 OF 2)
600
.,.._ 580 zu.
c:c<2-
..J w 0 a: 560 o::::>
(,,) ....
W<(
e,, a:
<(W 540 a: a.
w::E
>W
<( ....
520 500 2400
<
~ 2300
-
w a:
- > 2200
~
w a:
- a. 2100 a:
w N 2000 a:
- >
~
w 1900 a:
a.
1800 0 50 100 150 200 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4*12a TYPICAL TRANSIENT RESPONSE TO A DROPPED RCCA (OR RCCAS) IN AUTOMATIC CONTROL (SHEET 2 OF 2)
B/B-UFSAR Figures 15.4-13 thru 15.4-17 have been deleted intentionally.
REVISION 15 - DECEMBER 2014
REVISION 9 DECEMBER 2002
. 1. 4
- 1. 2
'::J" C(
c::Z LJ-3: :I!
~~ .....
IXO .8 C(
!:jZ u2
- >I-z~ *6 ai:::
.....
........
- 4
- 2 D 2 4 6 8 10 TIME ~SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-18 NUCLEAR POWER TRANSIENT.
SOL HFP ROD EJECTION ACCIDENT
REVISION 9 DECEMBER 2002 FUEL CENTER TEMPERATURE f"UEL AVERAGE TEMPERATURE CLAD OUTER TEMPERATURE
/
/
3000 I I
I
.,,,*-*- .........
2000
/
/ -*- -. -*- *---
/
/ -*-*- *-*-*- *-*- *-*-. -*-.
I 1000 I I
0 2 4 6 8 10 TIME {SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-19 HOT SPOT FUEL AND CLAD TEMPERATURES VERSUS TIME BOL HFP ROD EJECTION ACCIDENT
REVISION 9 DECEMBER 2002 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.4-20 NUCLEAR POWER TRANSIENT.
EOL HFP ROD EJECTION ACCIDENT
. REVISION *9 DECEMBER 2002 FUEL CENTER TEMPERATURE
- - - FUEL AVERAGE TEMPERATURE
- * - * -
- CLAD OUTER TEMPERATURE 5000 L&J 0:::
- >
<
a:
L&J 4000
/ .......
0.. / ......
......
- IE / ..... .....
...........
L&J- /
QC,,
- 5~
u-3000 I
I I
- ....... -.
0 z ,,.. ...-*-*-*--.
-. -.
<
-'
L&J 2000
/
/
/
-*- *- --- *-*-*-*
- >
...... / -*- -*-*- -*-.
I 1000 I 0 *2 .. 6 8 I0 TIME (SECONDS)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFElY ANALYSIS REPORT FIGURE 15.4-21 HOT SPOT FUEL AND CLAD TEMPERATURES VERSUS TIME EOL HFP ROD EJECTION ACCIDENT
REVISION 15 DECEMBER 2014
-.E 0
c:
1.2 0
z
.......
0
().8 c
0
+:;
(.,1 e
...._ Q.6
~
\......
<1' 3: OA 0
a..
._
0 Cl) 0.2 0
- J
- z:
0 0 100 200 JOO 400 500 600 Time( seconds)
~ 580
......,
- )
0 a>
a..
570 E
~560 Cl)
~
e Q) 550
~
~ 540 5~...,______________________________________________.,.
0 u
0 100 200 JOO 400 500 600 Time(seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.5-1 NUCLEAR POWER TRANSIENT CORE AVERAGE WATER TEMPERATURE TRANSIENT FOR INADVERTENT ACTUATION OF ECCS DURING POWER OPERATION
REVISION 15 DECEMBER 2014 2&00-----------------------------------------------
1600_,___________________..._.________________________......
0 too 200 300 400 500 600 Time(seconds) 2000
..............
~
t:
..__..
<I.> 1500 E::;)
0
>
""-
Cb
......... 1000 0
3:
,_
Cl>
N
- c
- l (I')
500 en (I.)
'-
0...
0 100 200 JOO 400 500 600 Time( seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.5-2 PRESSURIZER PRESSURE TRANSIENT PRESSURIZER WATER VOLUME TRANSIENT FOR INADVERTENT ACTUATION OF ECCS DURING POWER OPERATION
REVISION 15 DECEMBER 2014 7
6 5
0:::: 4 CD
- z 0
3 2
1 0
0 100 200 JOO 400 500 600 Time(seconds)
-c; 1.2------------------------
c:
- e0
- z:
'O
~ O.B 0
e
.......
-c: 0.6
~ (..)
~ 0.4
~
~
E 0.2 0
2 VJ 0---...........----...-------.~--------------------------..-....
0 100 200 JOO 400 500 Time{seconds}
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.5-3 DNBR AND STEAM FLOW TRANSIENTS FOR INADVERTENT ACTUATION OF ECCS DURING POWER OPERATION
REVISION 15 DECEMBER 2014 1-2
,.......
cc:
- e 0 1
- c:
0 c:
0 Q.8
- .;::::;;
u
- E 0.6
"'-'"
~
Q.)
- a:
0 (M
~
....
0 v
u::I 0.2
- z 0
0 10 20 30 40 50 60 Time (seconds) 4 a:::
~ 3 0
1 0 10 20 30 Time (seconds) "° BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFElY ANALYSIS REPORT FIGURE 15.6-1 INADVERTENT OPENING OF A PRESSURIZER SAFETY VALVE
REVISION 15 DECEMBER 2014
&00--------~---------------------~-----------------.
-c::-
1
- E 590~--------------
e:?
l
=:II 580
~
~ 570
~
"11 560
- i!
10 20 30 40 Time (seconds) 2400-----------~----------~---------------------------
-c;- 2200
~
~ 2000
- I o_~ 1800
~
- c::
=s 1600
~
~
G:l-1400 1200 0 10 20 "° Time {seconds)
-40 50 60 l*-
f ,_, -
J 1400 j
-1 *200 a.. 1000+-"'----------------~------...,.-..._i._ _________ __.~--.----__.
0 10 20 30 40 50 60 Time (seconds)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6- 2 INADVERTENT OPENING OF A PRESSURIZER SAFETY VALVE
BREAK OCCURS t REACTOR TRIP (COMPENSATED PRESSURIZER PRESSURE!
PUMPED SAFETY INJECTION SIGNAL (Hf*I CONT. PRESS. OR LO PRESSURIZER PRESS.I z PUMPED SAFETY INJECTION BEGINS (ASSUMING OFFSITE POWER AVAILABLE)
~
0 3= ACCUMULATOR INJECTION 0
-'
II) CONTAINMENT HEAT REMOVAL SYS. INITIATION (ASSUMING OFFSITE POWER AVAIL.)
i END OF BYPASS
.1 -'
-'
END OF SLOWDOWN PUMPED SAFETY INJECTION BEGINS (ASSUMING LOSS OF OFFSITE POWERI I:
w a:
I BOTTOM OF CORE RECOVERY CONTAINMENT HEAT REMOVAL SYS. INITIATION (ASSUMING LOSS OF OFFSITE POWERI Q
8-' ACCUMULATORS EMPTY
~
w a:
--'
CORE QUENCHED SWITCH TO COLD LEG RECIR . ON RWST LOW LEVEL ALARM (SEMI-AUTOMATIC) c,,
z
8 ~~
(,)
SWITCH TO LONG-TERM RECIRCULATION (MANUAL ACTION)
~
a:
w
~
c,,
z 0
-'
l BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3 SEQUENCE OF EVENTS FOR LARGE BREAK LOCA ANALYSIS
REVISION 15 DECEMBER 2014
--RCS
- - - -
- I ntact SG
- -* ~**-*
- RsCllVt c d SC
' ... ... .__. ,,,,,.- -,
o,+---....----------~-......,.--~-----------1 0 l080 M.,.------------------~~-----------
so
~
E
.J:J
.::::;::.. ,J) cu 15 0:::
- it
~ 20
(/)
en 0
~
IO 0
- 3T'ame (;)'° BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3A PRESSURIZER AND STEAM GENERATOR PRESSURE AND RUPTURED TUBE FLOW RATE SGTR RESULTS {OFFSITE DOSE CASE)-UNIT 1
REVISION 15
_____.. K~~\**cd ~G DECEMBER 2014
- - - -
- Toto* *ntocl SCs ato----------~------------~--~--.
1 2000
-g
~
E
I I
f I
' '
I t
- '
- t
'
.,"
\
.'
Q~
\
.... --* .. ------:
.............~.._
I ..,. ~
-., ,,.., __ -
...................,.......__._...__............____...-___ ~
I
- o.......__...___ __..__ _ _ _ _ _ _ _.........,.........
- JOO()
Time (sJ
- BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-38 STEAM GENERATOR RELEASE RATES ANO MASS RETAINED SGTR RESULTS {OFFSITE DOSE CASE)-UNIT 1
REVISION 15 DECEMBER 2014
-**--RCS
- - - - * ! 1st oc; I SG
- Rup t1H eel Sir
. .\
r--- - - - ... - .........\
I
's
..
.
I I
.....,;
'
'"-*.... /~ \
-- ...**-..........:' \\
"
' ' ...
-.._.,,,,,,_,.._.._..., __
Cl
-
o-+-._._...._.__ __,_..____,.~---.----__...--......,_
JllO Twe {sJ
- ___,
~----------~--~---------------~---
__...___________________________________.....,
0 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3C PRESSURIZER AND STEAM GENERATOR PRESSURE ANO RUPTURED TUBE FLOW RATE SGTR RESULTS {OFFSITE DOSE CASE)-UNIT 2
REVISION 15 DECEMBER 2014
- - R11p&ure4 SG
- - - -
- lotiH t11hc:t SCa I
'(ii"
""'-
e
,'*I g *
..!!!
~ MIGO *
f
~ .,'
0 c;:
U)
Cit
,,
0
- E I
~
____ .,...
\
.\
\
t .......
.. ___ _
UID 20llO
.. . '. ' ---
SOCIO 6000 7000
- *150CXD
.0 E
c-
. . . . .--------.,..-.--...._. . . . .
0 HIGO
._..~--...-----------r BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3D STEAM GENERATOR RELEASE RATES AND MASS RETAINED SGTR RESULTS {OffSITE DOSE CASE)-UNIT 2
REVISION 15 DECEMBER 2014
'
- -*--.. "s
.. U~ 4 ~\ S~
- ****---* 'h'\ "*c c *G 2!i0 -------------
- ~~'"""""'"
c 5lll
_____
1000
_,_.........._...__.....,._
- ~ ,. 1.!C>>
.........,._._~
lOOO -
T:-if (s)
~ ---~---------------~-----
- . 1*.
- . r . . . ._
. * ~- '\
\. \
l \
l; \
~} \
i .
.
"YI. ~ ~ noo ~ JtJIJC 111e (S i BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3E PRESSURIZER ANO STEAM GENERATOR PRESSURE AND RUPTURED TUBE FLOW RATE SGTR RESULTS (MARGIN TO OVERFILL CASE)-UNIT 1
REVISION 15 11J $C DECEMBER 2014
" . 11t sc:,
J5C
.m 7.ie 200 I t.1 *
Q r:;:
~
£ f
- ~ f
~
0
- a t@
t
~
- '
~
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3F STEAM GENERATOR RELEASE RATES AND MASS RETAINED SGTR RESULTS (MARGIN TO OVERFILL CASE)-UNIT
REVISION 15 DECEMBER 2014
--RCS
- - - -
- l*het SG
- --*---* 1t11,tur1ti SG
\
\
\
\
... ', _____ .... ,
°";'*""'"' ',._. __
0----.. . . . . .---....-.--..------.__...----_......... . . . .
0 so I
- 500 1580 fne (sJ
.,
-
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3G PRESSURI2ER AND STEAM GENERATOR PRESSURE AND RUPTURED TUBE FLOW RATE SGTR RESULTS (MARGIN TO OVERFILL CASE)-UNIT 2
REVISION 15
- *- - Ruoluecl SG DECEMBER 2014
- - - - ' 1 o t OI I n I act SG s 250..-------------------------~----------i
,,',,
- 'I \
I I
' '\
\
\
\
'
'\ I\
t '\
' ' .. ' '
I '\'
'
I I l I ' '
I
- .. ,-
0
- .*
lime
~
(SJ
-
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3H STEAM GENERATOR RELEASE RATES AND MASS RETAINED SGTR RESULTS (MARGIN TO OVERFILL CASE)-UNIT 2
REVISION 15 DECEMBER 2014 u-----------------------------~-
015
--
,,4-.;....,;.-.--.-------------~-----------t e 3COO ...
Time (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-31 BREAK FLOW FLASHING FRACTION SGTR RESULTS (OffSITE DOSE CASE) - UNIT 1
REVISION 15 DECEMBER 2014 0-2-----------------
c::
0
(,,) 0.1 e
.......
oa:-ot o...__..___...._........_--.....________..._________.......__"'-'
8 .... 5000 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-3J BREAK FLOW FLASHING FRACTION SGTR RESULTS {OFFSITE DOSE CASE) - UNIT 2
B/B-UFSAR Figure 15.6-4 has been deleted intentionally.
REVISION 14 - DECEMBER 2012
REVISION 1 DECEMBER 1989 N L 0 0 T c R T CORE PRESSURE, CORE FLOW, u MIXTURE LEVEL, AND A
FUEL ROD POWER M HISTORY p 0 <TIME < CORE COVERED
' --
. .
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-5 CODE INTERFACE DESCRIPTION FOR SMALL BREAK MODEL M2148.004 07-80
B/B-UFSAR Figure 15. 6-6 has been deleted intentionally .
REVISION 6 - DECEMBER 1996
B/B -UFSAR Figure 15.6 -6a has been deleted intentionally .
REVISION 7 - DECEMBER 1998
REVISION 14 DECEMBER 2012 0.6 0.5
.40 0.4
- 0.3
.27 0.2 0.1 0.2 0.25 0.3 0.35 0.4 0.45 0.5 PMID PBOT = integrated power fraction In the bottom third of the core PMID =integrated power fraction in the middle third of the core BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6- 7 Pbot/Pmid OPERATING LIMITS
REVISION 14 DECEMBER 2012 1800~-------------------'
1600 1400 1200 1000 800 600+--...:..----,-----'---,-----'-r----,----.:----i o 100 200 300 400 500 After Break (s) 4G5547171 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-80 HOT ROD PEAK CLADDING TEMPERATURE FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012
\ '/ '\ f 60000,----------------------.
50000 40000 10000
.... ,dol o
-10000 +-----,.--'--'--.------,'--'--'----,----'---1 o 100 200 300 400 500 After Break (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8b BREAK FLOW ON VESSEL SIDE OF BROKEN COLD LEG FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012 30000,..---------------------,
25000 20000
- 5000 10000
\
o
-5000-1-----,----..,......;----,---'---..,..---:----1 o 50 100 200 250 Time After 105547171 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8c BREAK FLOW ON LOOP SIDE OF BROKEN COLD LEG FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012 r P ,\
- - A r!
1c 14
/
I 08 06 I I I I
i 04 I 02 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8d VOID FRACTION AT THE INTACT AND BROKEN LOOP PUMP INLET FOR ECCS LB LOCA PCT LIMITING CASE-UNIT 1
REVISION 14 DECEMBER 2012
'01TT"-------------------.
25 3G 41)5547171 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8e VAPOR FLOW RATE AT TOP OF CORE AVERAGE CHANNEL DURING BLOWDOWN FOR ECCS LB LOCA PCT LIMITING CASE-UNIT 1
REVISION 14 DECEMBER 2012 I; 1 '/ A L 1 ;
10 o
-5
_1O+-'--'---r-~~-r---'-----,--~-..,...--'---,----..i.-'-I 25 30 4C554?171 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8f VAPOR FLOW RATE AT TOP OF HOT ASSEMBLY CHANNEL DURING BLOWDOWN FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012 12~------------------'
10
-
O+----~-------r--~--~----~------~
a 1()O BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8g COLLAPSED LIQUID LEVEL IN LOWER PLENUM FOR ECCS LB LOCA PCT LIMITING CASE-UNIT 1
REVISION 14 DECEMBER 2012 J;"
2500...---------------------,
r\
1500
\ \
1000 i
I 500 I
\
-500+----..,...-----'-..,...-'-----,.---..,..------!
o 100 400 500 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8h ACCUMULATOR MASS FLOW RATE VAPOR FLOW RATE FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012 1200-r---------------------.
"'.
200 o
o 100 400 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 1S.6-8i TOTAL SI MASS FLOW RATE FOR ONE INTACT LOOP FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012 Ll \Gt'jL~N 12-r---------------------,
10 o~~~~~~~~~----~--~-r--~~
o 100 200 300 400 500 Time After Break (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8j COLLAPSED LIQUID LEVEL IN CORE AVERAGE CHANNEL FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012 i \' 1 ! (jWN JO~----------------------------------~
20 15 10 5
o 100 200 400 soc Time After Break 405547171 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8k COLLAPSED LIQUID LEVEL IN DOWNCOMER CHANNEL FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
REVISION 14 DECEMBER 2012 250000-r--------------------,
LCOOOO 150000 100000 r
50000 o 100 200 400 500 After Break
- 05547171 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-81 VESSEL FLUID MASS FOR ECCS LB LOCA PCT LIMITING CASE-UNIT 1
REVISION 14 DECEMBER 2012
<J! !
12-,...----------.-------------.
JfL-,.--------'r 10 0+------,~--~~--~---,----~~~--~
o 100 400 500 4-05541171 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-8m HOT ROD PEAK CLADDING TEMPERATURE LOCATION FOR ECCS LB LOCA PCT LIMITING CASE-UNIT
B/B - [JFSAR Figures 15.6-8n through 15.6-8p have been deleted intentionally.
REVISION 6 - DECEMBER 1996
REVISION 14 DECEMBER 2012
'.; I _ C 1800-r-------------------,
1000 600 400 200+-----,----,.------,----'---'--,---'-""'---1 100 200 300 400 500 Time After Break (5)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-90 HOT ROD PEAK CLADDING TEMPERATURE rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012 60000-.---------------------.
50000 40000 30000 20000 10000
-10000+----'--....,.----'--,-----,.--"""--"-,....----1 o 100 400 500 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9b BREAK FLOW ON VESSEL SIDE OF BROKEN COLD LEG rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012 2 L w 30000,----------------------,
25000 20000 15000 I 10000
-5000+-------,----,----,.------,-----1 o 50 200 250 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9c BREAK FLOW ON LOOP SIDE OF BROKEN COLD LEG rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012
'I
'\'1 14 12
/
I 08 1 I I I !
04 02 o~j------r-----~--~--~--~--~--~--~
o 20 80 100 1906')l651)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9d VOID FRACTION AT THE INTACT AND BROKEN LOOP PUMP INLET rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012 L H 1 \
- o".-,-~------------------.
-IO-+---,.-------,------r-~-,.--~.....,._-__1 o 25 ,,0 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-ge VAPOR FLOW RATE AT TOP OF CORE AVERAGE CHANNEL DURING BLOWDOWN rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012
-/ :\
f5~--------------------'
v o
-5
-!O+---.,...---..,.---,...---.,..-'---,-----'--l 25 30 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9f VAPOR FLOW RATE AT TOP OF HOT ASSEMBLY CHANNEL DURING BLOWDOWN rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012
'; 1 1£ 10 o 100 200 400 500 Time BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9g COLLAPSED LIQUID LEVEL IN LOWER PLENUM rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012
" r 2000~----------------""""""
1000 I 500 o
-500+------,-----'-..,..------r----r-.:...--~
o iOO 200 400 500 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9h ACCUMULATOR MASS FLOW RATE VAPOR FLOW RATE rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012 i r 1200~-----------------""'"
....
1000 800 I
600
(/)
- ,;')
0 I~
2:
400 200 o+-~~~~----~~~--~----~~--~~
o 100 400 500 "90651 ['.511 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 1S.6-9i TOTAL SI MASS FLOW RATE FOR ONE INTACT LOOP rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012
--!It,,;T 2 Llc! ')E L! ! l'lt Iq 12-.--------------------,
10 o 100 200 300 400 500 Time After Break (s)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9j COLLAPSED LIQUID LEVEL IN CORE AVERAGE CHANNEL rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012
--UNi t:\;EL I lei ,N A,Cr L 30-r---------------------,
20 15 10 o 100 200 300 400 500 Time flJter Break (s)
"}06'516511 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9k COLLAPSED LIQUID LEVEL IN DOWNCOMER CHANNEL rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012 I : U i 200000 150000 100000 50000 O+-~--~------~----~------r-----4 o 100 400 500 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-91 VESSEL FLUID MASS rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
REVISION 14 DECEMBER 2012 00 1 K CL AT i ,')
IL 10 1111 LJJ 8
6 4
a 100 200 300 400 500 48664576 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-9m HOT ROD PEAK CLADDING TEMPERATURE LOCATION rOR ECCS LB LOCA PCT LIMITING CASE-UNIT 2
B/B-I1FSAR Figures 15.6 9n through 15.6-12p have been deleted intentionally.
REVISION 6 - DECEMBER 1996
REVISION '9 DECEMBER 2002 100 90 80 10
.-.
u 60 Q)
......
tn' E
-
.c 50
-.-.
-u.
~
0 40 30 20 10 ~
o o 500 1000 1500 2000 2500 Pressure (psi)
Byron /Braid wood Stations Updated Final Safety Analy sis Repo rt Intact Loop Safety Injection Flow Rates Used in Small Break Figure 15.6*13a
REVISION 9 DECEMBER 2002 40 30 u
~
Q)
-...
til E
20 "Q
~
-.
u 0
10 o
o 500 1000 1500 2000 2500 Pressure (psi)
Byron /Braid wood Statio ns Updated Final Safety Analy sis Report Broken Loop Safety Injectio n Flow Rates Used in SmaU Break Figure 15.6-13b
REVISION 9 DECEMBER 2002 14 12 10 8
6 4
2 o
o 2 4 6 8 10 Elevation (ft)
Byron/13raldwood Stations Updated Final Safety Analysis Re,port Core Power Sbape Used in Small Break Figure 15.S*13c
REVISION 12 DECEMBER 210108 600 500
-E Q.
DI
- 400
==
o u:
!
-
- J S
( ,)
300
.5 200 100 0+-----.,....------,...........----,...-------,-------,.------,-----1 o 250 500 750 1000 1250 1500 Reactor Pressure (psia)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-13d RECIRCULATION PHASE INTACT LINE ECCS FLOW RATES USED IN SMALL BREAK LOCA
REVISION 12 DECEMBER 2008 1800 1600 1400
-e 1200 Co S1000
~
u:
=
'a BOO U) 600 400 200 O+----,------,.----.,------,..----.,.----.....,-----r----..,...----J o 200 400 600 800 1000 1200 1400 1600 Reactor Pressure (psia)
BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.6-13e RECIRCULATION PHASE SPILL LINE ECCS FLOW RATES USED IN SMALL BREAK LOCA
REVISION 9 DECEMBER 2002 740 .......- - - - - - - - - - - - - - - - - - - - - - - . . . ,
720 700 680 Q) 660
'-
........
- 3 e
2i. 640 E
Q) t--
620 600 580 560 4-.J..-.L-J,--r-i-..J-..L.-.,..-L.-J.--J..--,.......J.-.J.--~r___L--J..-L-.,_...L-.l.-J-,.-J.......I-....J.-.I 16000 18000 20000 . 22000 24000 26000 28000 30000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit 1 (I.Sloch, Reduced T AVG)
Figure 15.6-14a
REVISION 9 DECEMBER 2002 600 ~----------------------..,
595 590
....
Q)
..a 585 e
Q)
Q..
E
~
580 575 570 -J-.l-.Ii..-L.....,..."":-_J-,r--'--J--L-.,...-I...-L.-I-....,-..l-...I.-L.--r--l--L-...I---r---J"--I.--i--I 16000 18000 20000 22000 24000 26000 28000 30000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysi,s Report Fluid Temperature - Unit 1 (1.5 Inch, Reduced TAve)
Figure 15.6-14b
REVISION 9 DECEMBER 2002 5
10 - - - - - - - - - - - - - - - - - - - - - - - - - - ,
4 to
--Q) 0
--e U
'- 2 Q) 10 en c:
I--
-+-J 0
Q.)
- c 1 10 18000 20000 22000 24000 26000 28000 30000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coefficient - Unit 1 (1.5 locb, Reduced T AVG)
Figure 15.6-14c
REVISION 9 DECEMBER 2002 2400 2200 -
--
2000 -
10
'iii Cl..
"--""
eu 1800 -
~
- J en tf)
~
0..
1600 -
,
1400 -
.
La II ,
1200 20000 25000 30000 0 5000 10000 15000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit 1 (1,5 Inch, Reduced T AVC)
Figure 15.6*14d
REVISION 9 DECEMBER 2002 36.....-------------------------,
34 32 Q)
>
..3 28
--
(1)
I-
- J X
~ 26 24 22 20 +-J-J....J--'-.,-.J....,.J~_..1_~..I...-JL.-l.-..1-..,..-J.-J--J,.-l.-""T"-~l.-J.-.I..-r-..l--L-.L--t..-4 10000 15000 20000 25000 30000 o 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis' Report Core Mbture Level- Unit 1 (1.5 Inch, Reduced TAVG)
Figure 15.6-14&
REVISION 9 DECEMBER 2002 160 - - - - - - - - - - - - - - - - - - - - - - - - .
140 120 cn100
'-
E
..0
-=::::,...
80
......
Q) o 0:::::
- 60 o
G:
en en o 40
- a 20 o
J-J_J.--J...~r_l_....J-..I-J....,......I-J-J.-'--r---J--J-...L-J--r--'--....L..-L-J.-r-....L.-.JI--L....I.-f o 5000 10000 15000. 20000 25000 30000 Time (s)
Byron/Braidwood Stations Updated*;;Final Safety Analysis Report Core Steam Flow Rate - Unit 1 (l.S Incb, Reduced TAve)
Figure 15.6-141
REVISION 9 DECEMBER 2002 500 - - - - - - - - - - - - - - - - - - - - - - - - - ,
400 100 O+-J....J,--I..-'--r-l.....J,.-l...~r-1....J..-J.-.,-..I.-..J-....I..-J'-T-J-...I- ............,..........~I-..l.-l o 5000 10000 ,15000 20000 25000 30000 Time (s)
Byron/Braidwood Stations Updated Final: Safety Analysis Report Break Flow - Unit 1 (1.5 loeb, Reduced T .wc)
Figure 15.6.14g
REVISION 9 DECEMBER 2002 800------------------------,
750 600 550 J...J,-J-_l-J,.-I--I.-I.-J-..,...L..J,,-I-....-L..-I...I-~....L......l--r--lI.-l--l-....-L..-J,.~
14000 16000 18000 20000 22000 24000 30000 Time (5)
Byron/Braidwood Stations Updated Final Safety,Analysis Report .
Peak Clad Temperature. Unit 1 (1.5 Incb, Nominal T,we)
Figure 15.6-14h
REVISION 9 DECEMBER 2002 605 ~---------------------.....,
600 595
--.a
'G:"
....
Q) 590 e
Q) 0..585 E
Q)
I-580 575 16000 18000 20000 22000 24000 26000 28000 30000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analy~is Report ..
Fluid Temperature - Unit]
(1.5 Ineb, Nominal T AVC)
Figure 15.6-141
REVISION 9 DECEMBER 2002 5
10 ----------------------~....,
4 10
-
'1D 0
c....:>
2
<1> 10
~
....-
en c:
--
0 l.-
I-0 Q.)
- x: 1 10 10o J-J.-I-_.L...J.~_r_J.-.j,--'-....-J..-l,--L..-r-I..-L...J-,....-I:........I-..J-,....-I........J-..,..-.I'--I......I--f 14000 16000 18000 20000 22000 24000 26000 28000 30000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coefficient - Unit 1 (l.Slnch, Nominal T AVC)
Figure 15.6-14j
REVISION 9 DECEMBER 2002 2400 2200 -
--
2000 0
.c;;
---
0..
CD
"-
1800-
- s' en en
~
....
Q) 1600 -
1400 -
I 1200 25000 0 5000 10000 15000 20000 30000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit 1 (1.5 IDcb, Nominal TAvcl Figure 1S.6-14k
REVISION 9 DECEMBER 2002 36 - - - - - - - - - - - - - - - - - - - -
---,
34 32
~
.3 28
~
........
- 1 X
~ 26 24 22 20 +-L.-J.- J.-.J.-.. --J.-L-- "-"'--T- 4-....L- .l..-lo-. ,...................-4--r"'"
o 5000 10000 15000 20000
.J--L-.a....l.--r-I--J-....I-~
25000 30000 Time (5)
Byron /Braid wood Statio ns Updated Final Safety Analy sis Report Core Mbtur e Level - Unit 1 (J.5lnc b, Nominal T AVG)
Figure 15.6-141
REVISION 9 DECEMBER 2002 600------------------------.
500
.......
(1) o 300 0::
- s:
o G:
V>
V>
~ 200 100 O.......J.....L~~...._J--L.....J-..J.-r-J.-J.......L-.l-r-J......J.-J.....L--,-..J-...i...-I~-,-....L-.l..-J,~-l o 5000 10000 15000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Steam Flow R.ate - Unit 1 (1.5 ]ncb, Nominal T AVG)
Figure 15.6-14m
REVISION 9 DECEMBER 2002 400 - - - - - - - - - - - - - - - - - - - - - - - - . . ,
350 300 en
~
.............
E 250
..0
~
Q) o 200
~
0:::
- =
o CL:
en 150 en o
- !E 100 50 O-l-J.-I..-1-..l-...--J....,J.....J.,...l-..,...J.-I..-l-.......-r-L.-l.-l-...l-~I.-J,.-l-...l-T""'""L.-I.-l-...I-f o 5000 10000 15000 20000 25000 30000 Time (8)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit I (1.5lnc:h, Nominal T Avc)
Figure 15.6-14"
REVISION 9 DECEMBER 2002 1800 - - - - - - - - - - - - - - - - - - - - - - -
---.
1600 1400 1200 e
- l
.........
o....
E 1000 a>
Q)
I-800 600 400 -+--.l--J._I--..L--r---l.---Jl....-..!--....L--r--1.-1--..L-~-'-r--J-.J--....L--
-L--;
2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety:Analysis Report Peak Clad Tempe rature. Unit J (2 inch, Reduce d TAVG)
Figure 15.6-15a
REVISION 9 DECEMBER 2002 1600 - - - - - - - - - - - - - - - - - - - - - - - -
--,
1400 1200
~
~ 1000 L-Q.)
0-E
~
800 600 400 -+-.....l-....J..--I_J--.,.---L----..i---l-J...-~....I---L.--L-I--r---'--..L---L
..I 2000 3000 4000 5000 6000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Fluid Tempe rature - Unit J (2 loch, Reduced T AVG)
Figure 15.6-15b
REVISION 9 DECEMBER 2002 105 . . . . - - - - - - - - - - - - - - - - - : - - - - -
-- ---,
4 10
-
""4-Cl>
o
<..>
s.... 2
..::£ 10 -
en c
t-E
.......
o Cl>
- 1::\
10 100 +--J.._1.-~_J.._~...L-.-J--L.--..1--r----'I....-..J.I-~....L..--,~...L
---.L.--I-..L---l 2000 3000 4000 5000 6000 Time (s)
Byron/Br~ldwood Stations Updat~d Final Safety Analysis Report Heat Transfe r Coefficient - Unit 1 (1 Incb, Reduced T AVG)
Figure 15.6-15c
REVISION 9 DECEMBER 2002 2400 2200 -
2000 -
,-... 1800 0
"<ii
...---
0..
Q)
I-1600
- )
en en Q)
'V\
I-0.- 1400 1
1200 1000
',800 I I I' I 0 1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressur e - Unit 1 (2 Inch, Reduced TAVG)
Figure 15.6*15d
REVISION 9 DECEMBER 2002 4O...-- --- --- --- --- --- --- --- -a 35 20 t5 -l-~I._J,.......J__.,_...J-..L-.L--II.--r~-l-...J.-..L......,....-L--L.-L-...I---r--J.,....J...-J..--'--r-..I-.a.,.....l
--'-~
. 0 1000 2000 3000 4000 5000 6000 Time (5)
Byron/Braidwood Stations Upda ted Fine" Safety Analy sis Report Core Mixtur e Level - Unit 1 (2 Inch, Reduce d T AVC)
Figure 15.6*15e
REVISION 9 DECEMBER 2002 800 600 -
..--.
en
.............
E
...0 400
..::::::.-
......
<V 0
~
~
0 LL. 200 en V
en
..
0 &II .....
- iE r~ .1, .1
.... '"I'"nl ("'
0
-200 0 1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Steam Flow Rate - Unit 1 (2 Inch, Reduced T AVG)
Figure 15.6-15'
REVISION 9 DECEMBER 2002 800 - . - - - - - - - - - - - - - - - - - - - -
.
700
......
~
~ 1000
~
Cl>
0.
E Cl>
t-800 600 400 ~.J-,J:....J.--l--.--.!--..L-l-...J-.,......J.-I,~-L-.,.-I-..L-l-...l--~~.-.......l.-,.--J---I--l-...J-..f 2000 2500 3000 3500 4000 4500 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit 1 (2 Incb, Nominal TAve )
Figure 15.6*15h
REVISION 9 DECEMBER 2002 1600 -------------------------~
1400 1200
. 800 600 400 4-1.-L-l-~.,.....L-..L.-1-.J.-...-J-~...l.-I~~.J.-L-J.--r--..-.L-J,--I-..,.....l.-J:.-.L......L....i 2000 2500 3000 3500 4000 4500 5000 Time (5)
Byron/Braidwood Stations Updated Fina) Safety Analysis
'Report Fluid Temperature - Unit 1 (Z Incb, NoRlinal T AVG)
Figure 15.6-15/
REVISION 9 DECEMBER 2002 5
10
-
I-4 10 -
...--
-
l.L.
I
.....,
N I
...r=
~
--
........... 3
- s 10 -
.....,
CD
<l>
0
<.....>
2
.....<l>
L...
10 (I) c:
E t-oo+J
\..
_A
-
,
0
<l>
- z:
10 0
10 2000 2500 30'00 3500 40'00 4500 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis .'
Report Heat Transfe r Coefficient - Unit 1 (2 locb, Nominal TAvc)
Figure 15.6-15j
REVISION 9 DECEMBER 2002 2400 2200 -
2000 -
--
....-..1800-0 "enc..
e::::s 1600-en en e
CL 1400 1200 -
1
-
1000 -
800 40'00 5000 0 1000 2000 3000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit I (lInch, Nominal T AVC)
Figure 1S.6-15k
REVISION 9 DECEMBER 2002 40 - - - - - - - - - - - - - - - - - - - - - - - - . . ,
35 20 15 -lO--..l.--L.-J-...J-1...,OOr-O...J..-..I.-J.-J..2-000...-J-.J.--L....4-3-000r---l.....,l......L.~-4000,.-.L..-.,.J'---t---L-j5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Mixture Level - UoJt J (2 Incb,NonUnaITAvc)
Figure 15.6-151
REVISION 9 DECEMBER 2002 800 - - - - - - - - - - - - - - - - - - - - - - - - - ,
700 600 en
~
'- .
E 500
..0
...::::::::;..
.....,
Q) o 400 c::
3:
o G:
en 300 en o
- ?E 200 100 o 5000 o 1000 2000 JOOO Time (s)
Byron/Br~idwood Stations Updated Final Safety Analysis Report Core Steam Flow Rate - Unit 1 (2 Inch, Nominal TAve)
Figure 15.6-15m
REVISION 9 DECEMBER 2002 700 . . , . . - - - - - - - - - - - - - - - - - - - - - - ,
600 500
--...
C/)
.............
E
...0
.:::::.- 400
....-
Q) o c::::
~
o 300 G:
(/)
C/)
o
- a::
200 100 o-t-..l.-..L.-J...-L......,r---"~--L.--'-__r_-L....J......J-..I.-_r_-'-- .....I..-.I..,----"--'----.--.._t o 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit 1 (2 Incb, Nominal T AVC)
Figure 15.6-15n
REVISION 9 DECEMBER 2002 1600 - - - - - - - - - - - - - - - - - - - - - - - - ,
1400 1200
<1>
\-
~ 1000
\-
<1>
0-E
<1>
I--
800 600 400 -J-....L...-I---L......J,-..,r--J..-...L.-..J.-...l---r--l---I..~--I-r_.l..-..L-...J--l---t 1000 2000 3000 4000 5000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit 1 (3Incb, Reduced TAvc)
Figure 15.6-16a
REVISION 9 DECEMBER 2002 1400 ..-p----------------------,
1200
-.......
~ 1000 BOO 600 400 -l-~-.L..--L.--L.__r---l-L.-.L.-.l.--_r_....l.-..J-,-I.-~_r--L.--I.--...J_.......,
1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated. Final Safety Analysis Report Fluid Temperature - Unit 1 (3 Inch, Reduced T Ave)
Figure 15.6*16b
REVISION 9 DECEMBER 2002 5
10 l-I-
I-
.
---
u...
, 10 -
-
I N
~
~
I
....t=
3
-
~10
~
CD
' +-
'+-
Q) 0 U
~ 2
'+-
Q) 10 -
en c:
~
~
- c 0
Q) 1
- ---
10 0
10 I
1000 2000 :J:
3000 4000 5000 Time (s)
Byron'Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coefficient - Unit I (3 Incb, Reduced TAVC)
Figure 15.6.16c:
REVISION 9 DECEMBER 2002 2500 - - - - - - - - - - - - - - - - - - - - - - - - - ,
2000
---*en o
a..
-...-
CD I-1500
~
en en CD l-n.
1000 500 ~..l.-J........I..-l.--.-~....L......l-...l--r-.l--II.....-J,--I.--r--l-...J.-..J-..J.-.-L..-l--..L-..J..-1 o 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit j (3 inch, Reduced T AVC)
Figure 15.6-16d
REVISION 9 DECEMBER 2002 40.....---------------------,
35
-
- +-'
~ 30 20 Byron/Braidwood Stations Updated Final Safety Analysis Report Core Mixture Level- Unit 1 (3 Incb, Reduced T AVC)
Figure 15.6-16e
REVISION 9 DECEMBER 2002 250 - - - - - - - - - - - - - - - - - - - - - . ; . . - - - - - ,
200
..--...
(I)
-........
E
=9 150
-...-
.......
Q.)
o
~
3=
o u.. 100 (I)
(I) o
- 2E 50 o 1000 2000 3000 4000 5000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Steam FJow Rate - Unit I (3 Inch, Reduced TAve )
figure 15.6-16f
REVISION 9 DECEMBER 2002 1800 1600 1400
--
-en 1200..;
E
...0
..:::::;..
Q) 1000 -
-+-'
0 0::::
3=
0 800 c:::
en en 600 -
~
0
- 2 400 -
200 -
"""'-- ..... j 0 I 0 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit 1 (3 Incb, Reduced T AVC)
Figure 15.6.16 9
REVISION 9 DECEMBER 2002 1600 -r------------------------,
1400 1200 Cl.>
-5
~
1000
~
Cl.>
0-E Cl.>
I--
800 600 400 +-...l--.L..J--I.-r-l.---L..-..l.-~r__I_-.l---l-...I-.,.._J..-L...-I.---l.._r.....l-....l.--'--J.....-f o 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysl$
Report Peak Clad Temperature - Unit 1 (3 Inch, Nominal T AVC)
Figure 15.6*16h
REVISION 9 DECEMBER 2002 1400....-----------------------,
1200
..--....
~ 1000 600 400 ....l..-..L.-J.--.l.---..--J__L.-J.-..l.~-l-...l-....l..-.J...--r_L..-J--L--L-;-J--.l---L-...l..--j o 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Fluid Temperature - Unit 1 (3lncb. Nominal T AVC) figure 15.6-161
REVISION 9 DECEMBER 2002 5
10 4
10 -
,...--..,
-
I...J-I N
........
~
I
...s::=
--.... 3
=' 10 -
........
co
- 0.>
0
-
U
~ 2 0.> 10 -
t f)
C 0
~
t-- \-/""--
........
0 0.>
- r: 1 10 0 I 10 I
0 1000 2000 3000 4000 5000 Time (s)
- Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coefficient - Unit 1 (3 Incb, Nominal TAVC)
Figure 15.6-16j
REVISION 9 DECEMBER 2002 2500 - - - - - - - - - - - - - - - - - - - - - - - - ,
200P
-
-
o
-e;:;
a..
~ 1500
- 3 en en Q) l.-
e...
1000 500 -+-J.-l...-L--J--.,-...J-...J_~~,.....J.--I...--L......J-_r__J-..J:..-.J.--..&.._,_.....l---...L--"__I o 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit 1 (3 Incb, Nominal T AVG)*
Figure 15.6-16k
REVISION 9 DECEMBER 2002 40.,....-------------------:------,
35 o 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final S;ilfety Analysis Report Core Mixture Level- Unit J (3 Inch, Nominal TAVG)
Figure 15.6-161
REVISION 9 DECEMBER 2002 2500...,.-----------------------,
2000 -
--....
~
(/)
E
..0 1500-
~
........
Q.)
o 0:::
3=
o G:: 1000-en
(/)
o
- a=
500 -
\ I
--, -,.-.,..,...."- -=-====-==r=-~=--,,,,=o,,-~""""""=_4
- Byron/Braidwood Stations Updated Final Safety Analysis Report Core Steam Flow Rate - Unit I (3 locb, Nomina) T AVC)
Figure 15.6-16m
REVISION 9 DECEMBER 2002 1600 1400 -
1200 -
en E 1000
..a
.::::::.-
..--
C1>
~
0 800 V
- e 0
Ci:
en
(/J 600 0
- E 400
-
--
,
200
-
II.-
~ j
- I I 0 I 5000 a 1000 2000 3000 4000 Time (s)
Byron/Braidwood'Stations Updated'Final Safety Analysis Report Break Flow - Unit 1 (3 Inch, Nominal T AVG)
Figure 15.6-16n
REVISION 9 DECEMBER 2002 1400 1200 I
.!::::.- 1000 -
e
.......
- J 0
V, l....
(1) a..
E (1) 800 I--
V\
600 -
,.j 400 500 10'00 '500 2000 2500 3000
- 3500
-
4000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Tempe rature - Unit 1 (4 Inch, Reduce d T ,wc)
Figure 15.6.17a
REVISION 9 DECEMBER 2002 1200-1100 -
)
1000 -
---
LL..
'--'
900 -
Q)
'-
- J
....-J 0
800 -
'Q)
-
0..
E cv V
I-- 700 -
600
-
500 -
""'-
400 500 1000 1500 2000 2500 3000 3500 4000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analyslsi.
Report Fluid Temperature - Unit 1 (4 Inch, Reduced T Ave)
Figure 15.6-17b
REVISION 9 DECEMBER 2002 5
10
---
4 to l.&-
I
.......
N
'+-
~
I
..c
-....... 3
....... 10 -
---
~
CD
'+-
'+-
Q) 0 U
2 10 -
~
Q)
Jf\J
'+-
en c:
0
~
I-
....... ~
0 Q)
- r:: 1 10 -
0 10 2500 30'00 3500 4000 500 1000 1500 2000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coefficient - Unit 1 (4IDcb, Reduced T Ave)
Figure 15.6-1i'c
REVISION 9 DECEMBER 2002 2500 - , . . - - - - - - - - - - - - - - - - - - - - - - - ,
2000
-
"01500 "Ci5 0..
Q)
'-
- s en en
~ 1000 0-500 O+--l--l--I---'---r--L-...JI..--.L--.l--,..--..J--..J.-...L-....l--r--'-.-Jo..-..l---.J---t o 1000 2000 3000 4000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure* Unit 1 (4 Incb, Reduced T AVC )
Figure 15.S-11d
REVISION 9 DECEMBER 2002 4O-r-----------------------,
35 20 15 -'"O----'"--..L..-..L.--l--OrOO-l..--L---O!...-.J..-20TO-O~-l-.-.L---L-3..,OO~O-J........j,.---L-l.-..j4000 t
Time (5)
Byron/Braidwood Stations Updated Final-Safely Analysis Report Core Mixture Level - Unit 1 (4 Inch, Reduced T AVC)
Figure 15.6-17e
REVISION 9 DECEMBER 2002 350 - - - - - - - - - - - - - - - - - - - - - - ,
300 250
--..
-......
en E
..0
-==--200
<I,)
o
-+oJ 0:::
3:
o 150 i:i:
en en o
~
100 50 o ~OL.Ji--J..-.i-..:.-l-000r--.1.-..1-..L--L--2000r-.L.-..1-.-I--.l--JooTa--L.~-J-..L..--14000 Time (5)
Byron/Braidwood Stations Updated 'Final Safety'Analysis Report Core Steam Flow Rate, Unit 1 (4 Inch, Reduced TAvc)
Figure 15.6*17f
REVISION 9 DECEMBER 2002 3000 2500 -
....--...
en 2000-
...........
---
E
£2
........
(l) a::: 1500 -
0
~
0 lL.
en en
~
0 1000 -
- E 500 -
0
....
- ALi ..............
LolII&&.....
~.~
0 1000 2000 3000 4000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit J (4 Incb, Reduced TAve)
Figure 15.6-17g
REVISION 9 DECEMBER 2002 1600
~
1400 -
1200 -
..---.
L&-.
.....
Q)
.z.....
0 1000-Q) 0-E Q)
I
~
800 -
600 -
~ \
,
.-
400 4000 0 toDD 2000 3000 Time (s)
Byron/Br~;dwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit]
(4 Inch, Nominal T AVC)
Figure 15.6-17h
REVISION 9 DECEMBER 2002 1400 1200 - )
--
-!:::::.- 1000 -
Q)
'- ..
...--
- J
"
c
'Q)-
Cl..
E Q) 800 -
~
600 l(1
~
-~--
400 3000 4000 0 1000 2000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Fluid Temperature - Unit 1 (4Iocb, Nominal TAve)
Figure 15.6-171
REVISION 9 DECEMBER 2002 s
10 4
10 -
'G:"
I
.......
C"I
..-
I
..c
~
__ 3'
--
- , 10
.......
CO Q) 0
-
U 2
10 *
~
Q.)
(/)
c
-
0
~
~
t-0 Q)
- r: t 10 -
0 ,
10
)
0 10'00 2000 3000 4000 Time (s)
Byron /Braid wood Statio ns Updated Final Safety Analy sis Repor t Heat Transfe r Coefficient - Unit II (4 IDtb, Nominal TAvc)
FIgure 15.6-17}
-
REVISION 9 DECEMBER 2002 2500 - - - - - - - - - - : - - - - - - - - - - - - - - - ,
2000 a>
10-
='
en en e
0-1000 500 o..J--l.--I---L---J----..---l--'t.--J...-..L--.,.-..L--J--J.-~_r--L---L..---.l.-.l~
o 1000 2000 3000 4000 Time (5)
Byron/Braidwood Stations
.Updated Final Safety Analysis Report Res Pressure - Unit 1 (4 Inch, Nominal T AVC)
Figure 15.6-17k
REVISION 9 DECEMBER 2002 40-----------------------,
35 15 -lo-L..-.l---'-.....l--1..,.OO-O..J---l.---l.--I-20rOO--J.--I-J...-...J.-3-:l000c:-.L..-...l.---l...-J.-4~OOO Time (5)
Byron/Braidwood Stations Upd~ted Fin'al Safety Analysis Report Core Mixture Level - Unit 1 (4Incb, Nominal TAVd Figure 15.6-171
REVISION 9 DECEMBER 2002 2500 f-2000 -
---(/)
---
E
~ 1500-
........
Q) 0 0::::
3:
0 G: 1000-(/)
(/)
0
- 2:
500 La.. II..... T a..l I 0
. 1 r "1 ....'" "
0 10'00 2000 3000 4000 Time (s)
Byron/Braidwood Stations Updated *Fin~1 Safety Analysis Report Core Steam Flow Rate - Unit i (4 Inch, Nominal T Avc)
Figure i5.6-17m
REVISION 9 DECEMBER 2002 3000
....
2500 -
Z2000 E
..0
..::::::.-
.......
Q) a:: 1500 0
3:
0 c:::
V en en Z
(;)
1000 500
- ~ .
0 4000 0 1000 2000 3000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - 'Unit 1 (4 Inch, Nominal T AVC)
Figure 15.6-17"
REVISION 9 DECEMBER 2002 900 - - - - - - - - - - - - - - - - - - - - - - - - - ,
850 800 750
~
- J
.........
~
Q.)
0.. 700 E
Q.)
I--
650 600 550 -I-.L-J--J..~_r_J--L....l--'---r--L.....L....J-J......,--J......l-.J.-,i4.....J-.,..L-J....I,--r...J-...L.....IL.-L-I 10000 15000 20000 25000 30000 35000 40000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit 2 (1.5 Inch, Reduced T AVC)
Figure 15.6-18a
REVISION 9 DECEMBER 2002 700 - y - - - - - - - - - - - - - - - - - - - - - - - - - - ,
680 660
.....
Q).
~.....
Q) 0- 620 E
tE 600 580 560 +-J.-J--l......L-...,........L....JL.....l.--I.--y--~--'-~..,.._L.....l.~ ......."T'"'"'"L.....L.--'-.....,..._l--..l.-I.........l.__oI 10000 15000 20000 25000 30000 35000 40000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis
- .. Report Fluid Temperature. Unit 2 (15 Inch, Reduced T AVC)
Figure 15.6-18b
REVISION 9 DECEMBER 2002 10 5
-r----------------------.....,
-
4 10 l...L-
-
I
.......
C'I I
L..
...c
1
~10
......
co
<1>
-
0 U
L.. 2
<1> 10 en c
0 r-L..
.......
0
<1.>
- r:: 1 10 15000 20000 25000 30000 35000 40000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coefficient - Unit 2 (1.5 Incb, Reduced T Avc)
Figure 15.6*18c
REVISION 9 DECEMBER 2002 2400 2200 -
2000
.....-
---
0
.Ci5 Q..
Q) 1800 *
~
- ,
(/)
(/)
Q)
~
0-1600 1400 ~
1200 0
I --- 10000 20000 30000 40000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure* Unit :2
(].*5 incb, Reduced T AVC)
Figure 'i5.6-18d
REVISION 9 DECEMBER 2002 36...,...------------------------,
34 32
---
30
...--
'+-
'-'"
Q) 28
>
Q.)
.-oJ Q.)
~
26
...--
- J X
- a:
24 22 20 18 --L-.-L---L.--L_,..-...L--...J.........J---I.---,--.J.---J._L..-..J..--y-..J--.L..--L.---J.-;
o 10000 20000 30000 40000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Mixture Level - Unit 2 (1.5lncb, Reduced T Ave )
Figure 15.6-188
REVISION 9 DECEMBER 2002 160 - - - - - - - - - - - - - - - - - - - - - - - ,
140 120
---.
en
'E- 100
.CJ
.::::::.-
.23 o
a::: 80 3:
o G:
en 60.
en o
~
40 20 o 10000 Byron/Braidwood Stations Updated'Fina' Safety Analysis Report Core Steam Flow Rate - Uolt 2 (l.5 Inch, Reduced T Ave)
Figure 15.6-18f
REVISION 9 DECEMBER 2002 500 - . - - - - - - - - - - - - - - - - - - - - - - - - - ,
400
-
~
E
...0
.;::::::...
300 Q.')
o a::::
3=
o G: 200 en en o
- z:
100 04---L..---l._L..-...L_--r~_..L.._...L__l___,..---'-.s.- ......--'-___r-.l.--~-.l---"__f o 10000 20000 30000 40000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit 2 (I.Slnc:h, Reduc:ed T AVC)
Figure 15.6-18g
REVISION 9 DECEMBER 2002 950 _---------------------.--0,)
900 850
-
'G:" 800
~
..3 o 750
~
Q.)
0-E Q.)
t- 700 650 600 550 -I-J.-L,....l-I....r...l-~...L.._,r_i_ .........L.....I-..,.....JL.....I-...L....I.....,.........~~--'-.l--L--'-'T"""""--.l-..I-J.~
5000 10000 15000 20000 25000 30000 35000 40000 Time (s)
Byron/Braidwood StatIons Updated Final Safety Analysis Report Peak Clad Temperature - Unit 2 (J.S loeh, Nominal TAve)
Figure 15.6-18h*
REVISION 9 DECEMBER 2002 750 - . - - - - - - - - - - - - - - - - - - - - - - - - . . . ,
700
~
.z o 650 L-a>
0-E
~
600 550 4-J1..-L...J-J-,....J-J"'-L--L-,r---L-...L-JI..-L...,-L-J.-..l.-/.-r-.J.,...L...l-~~.L-L...J-r..L...J-.L-..L.~
5000 WOOD 15000 20000 25000 .30000 35000 40000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Fluid Temperature - Unit 2 (l.5 Inch, Nominal T AVG).
Figure 15.6-181
REVISION 9 DECEMBER 2002 5
10 - - - - - - - - - - - - - - - - - - - - - - - - - - - - .
10
- C I
C"J c::::
....I
-c 3
-
~10
-+J CD
~
~
Q) 0 U
\0- 2 Cl> 10 en
~
C 0
\0-J--
-+J 0
Cl>
- I: 1 10 10000 15000 20000 25000 30000 35000 40000 Time (s)
Byron/Braidwood Stations Updated Firial Safety Analysis Report Heat Transfer Coefficient - Unit 2 (J.S Inch, Nominal T Ave)
Figure 15.6-18j
REVISION 9 DECEMBER 2002 2400 2200 -
--
2000 -
-
_2 en c...
Q.) 1800
'-
~
en en Q.)
'-
CL 1600 1400
.t - .......
1200 30000 40000 0 10000 20000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit 2 (J.Slnch, Nominal TAvc)
Figure 15.6-18k
REVISION 9 DECEMBER 2002' 36....----------------------,
34 32
-
30
--..
........
'-"
Q) 28
>
Q)
--'
Q)
L-
=:J 26
........
x
- 2:
24 22 20 18 -I-~.....I---I..---L____,r__J..-..l--..J-....l-__r--'----l.-l..-..L--.,.........1---'---L.---..I.---t o 10000 20000 30000 40000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis.
Report Core Mixture Level - Unit 2 (l.5 Incb, Nominal TAvc )
Figure 15.6-181
REVISION 9 DECEMBER 2002 700 - - - - - - - - - - - - - - - - - - - - - - - - ,
600 500
.--
en
~
E
...0
~400 a>
-+-I o
a:::
3::
o 300 Ci:
en en o
- E 200 100 o-l-...l---J--.:...~___r-J.--JI--L.-..J.---r--..J.-.....L--1-.....J...--r-.....J---J..---l.--JL...-t o 10000 20000 40000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Steam Flow Rate - Unit 2 (151nc:b, Nominal T Ave)
Figure is.6-18m
REVISION 9 DECEMBER 2002 400 - - - - - - - - - - - - - - - - - - - - - - . . ,
350 300
,--...,
en
-.......
E 250
~
-....-
Q)
-+-'
"
0::: 200
==
o G:
(J') 150 en
"
- E 100 50 O~-'---'---J---L---.~J..-..l..-...L--'----r---L-.....J.,--J.---I~r-..J....-...l:-...J-....J--i o 10000 20000 30000 40000 Time (s)
Byron/BraidWood Stations Updated Final Safety Analysis Report Break Flow* Unit 2 (1.5Incb, Nominal 'fAVG)
Figure 15.6-18n
REVISION 9 DECEMBER 2002 1800 - r - - - - - - - - - - - - - - - - - - - - - - - - - ,
1600 1400 1200 Q)
~
- l
.........
E:
E 1000 Q) .
Q)
J--
800 600 400 +-...J-....L-...L-...L-.,.--L--L--L-....I--r-....I-...l-...l-...l---r--.L--.L--.L---I...--r-J..-.J.--.J.----J--1 1000 2000 3000 4000 5000 6000 Time (8)
Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit 2 (2 Inch, Reduced T AVC)
Figure 15.6-19a
REVISION 9 DECEMBER 2002 1600 - - - - - - - - - - - - - - - - - - - - - - -
---.
1400 1200 800 600 400 -+-.-L--. ....l-.....l -.....l--r- ...J-....L- ....L---I. ---..---I.-
-'--'-..........--..-..........--J--.l-- .l---r--' ---'----1 .---L--/
1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Upd~ted Final Safety Analysis Report Fluid Tempe rature - Unit 2 (2 Inch, Reduced T AVG)
Figure 15.6*19b
REVISION 9 DECEMBER 2002 10 5
-r- --- --- --- --- --- --- --- --- --- ,
I-4 10 ~
"- 2
~ 10 ~
en c
--
.~
l--
........
a (1)
Xl 10 10o I....I-.....-_.l...-I...L-...I--....JII..-.r--.l...-..L
-J--L.-I .-_..L-
-...I--I-.:-.,..-~....J..-...L-1-1'-.- -.1...-
'...I.-... I.-.-I--j 1000 2000 30'00 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfe r Coefficient - Unit 2 (2 Inch, Reduced T AVe>
Figure 15.6-19 c
REVISION 9 DECEMBER 2002 2400 2200 -
2000
---.. 1800
---
0 "en 0..
CD
~
1600
='
en
(/)
<1>
VL
~
0-1400 -
.-
1 1200 -
1000 - ...--==
i 800 J.
0 1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Antilysis Report Res Pressure - Unit 2 (2 loth, Reduted T AVG)
Figure 15.6-19d
REVISION 9 DECEMBER 2002 40 .... .-- --- --- --- --- --- --- --- --. .... ,
35 20 15 -+-...l.-.l.....J~__r_-l-...I-..J.-..L..-,.........-..I--J-~.....-.l..-L.-I......J---,-.....J....~..J-~-r-I...
-1......J---I--1 o 1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Mixtur e IJevel - Unit 2 (2 Inch, Reduced TAVd Figure 15.6-1ge
REVISION 9 DECEMBER 2002 200 - - - - - - - - - - - - - - - - - - - - - - - - - -
---.
100
...--..
(J)
............
E
..0
......-
Q)
-+-'
0 0:::
0
- =
0 c:
(J) en 0
~
-100
-200 -+-.J..-I-.J,~_r_~..I--'--""--r-l--"--.l--L-r-'-...I.--'---'L..-..,---'--l--'-.J..-...-,~-
-'-....L-~
o 1000 2000 3000 4000 sooo 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Steam Flow Rate - Unit 2 (2 Inch, Reduced TAvc)
Figure 15.6-19f
REVISION 9 DECEMBER 2002 800 - . - - - - - - - - - - - - - - - - - - - - -
- ------.
700 600
---en
-.......
E 500
...a
.......
(I) .
~ 400 5:
o G:
en 300 en o
- 2 200 100 O-f-.J..-l..-l--l---...-....L--'--..l--.......,,-L-"---'--'-.,.-"--'--'--"-~..J--
..l-- ..........-..,.--'--- '-....l-..J -.-.I o 1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit 2 (2 Inch, Reduced T A VG)
Figure 15.6-19g
REVISION 9 DECEMBER 2002 1100....---------------------
---,
1000 900
'G:"
......-
--
<I.>
"-
- J c 800
"-
<I.>
a..
E
<I.>
I-700 600 500 +'-.J.-.L -J--.L... .,..--J.-. ..J.......l .-.J.-y-- L.....J.-- L.....J-.. ,.-""--l..
.--L.....J ....,...-L -..l.-.J.- .L..-..j 1000 2000 3000 4000 5000 6000 Time (s)
Byron /Braid wood Statio ns Upda ted Final Safet y Analy sis Repo rt Peak Clad Tempe rature - Unit :1 (2 Inch, Nominal T AVC)
Figure 15.6*19h
REVISION 9 DECEMBER 2002 900......-----------------------,
850 800
___ 750
--.-
LL.
e
.ae 700 Q,)
c....
E Q,)
.- 650 600 550 500 .+-.J..-J.....,J......J--,.-J.--I--J-....l-..,-.J..-J..-L...-I-,.--.l..--1--L--J-.,-..L-.J.-.L-L-....j 1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final-Safety Analysis Report Fluid Temperature - Unit 2 (2lncb, NominaJ TAVG )
Figure 15.6-191
REVISION 9 DECEMBER 2002 5
10 4
10 . F-
....-
-
W-I l-
....--
N I-I
...c.
........... 3
~ 10 .
....--
CD
- Q) 0
-
U I- 2 Q) 10 t f) c 0
l-I--
....--
0 Q)
"
- c 1 10 .
D 10 4000 5000 6000 1000 2000 3000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Beat Transfer Coefficient - Unit 2 (2 Inch, Nominal TAve) figure 15.6-19j
REVISION 9 DECEMBER 2002 2400 2200 -
2000
---
--
0
. en 1800-0..
Q.>
~
- 3 en en 1600 -
Q.>
~
0-1400 '"
L
'--, ~
a 1200 -
1000 0 1000 2000 3000 4000 5000 6000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit 2 (2 lncb, Nomina) T Ave)
Figure 15.6.19k
REVISION 9 DECEMBER 2002 40-------------------------,
35 20 4000 6000 o 1000 2000 3000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core MistuTe Level - Unit 1 (1 Inch, Nominal T AVC)
Figure 15.6-19\
REVISION 9 DECEMBER 2002 800 - - - - - - - - - - - - - - - - - - - - - - - - - ,
600
....--
-.....
en
E
- 2
<l>
a::: 400 0
s::
0 G:
en en 0
~
200 o -IlA...L.-l-1-l.1--'-..,..._.-.....i'--L-r....L....I..-.L-l.--r-l-..J....~l.._r.......J--L-...L--.I...__,r___J_.......J--J......I-..l o tOOO 2000 3000 4000 5000 6000 lime (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core S'eam Flow Rate* Unit 2 (2 Inch, Nominal T,WG)
Figure 15.6-19m
REVISION 9 DECEMBER 2002 700 -v------------------------,
600 500
---
Q) 0 0:::
- 0== 300 G:
en cn 0
- a:
200 100 o'+-J--L-J....J-r---L.-L......I....J--r-...J-J....-L-J-,.-L-L..-L-..J...-,-.L.....J-J--J.-,-...L-J-J.....,
o 1000 2000 3000 4000 5000 6000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Vnit :1 (2Incb, Nominal T AVC)
Figure 15.6.19n
REVISION 9-DECEMBER 2002 1600......------------------------,
1400 1200 e
..31000 e
Q) 0-E
~
600 600 400 -I-..J.-..L.-.L--L......,--L--J,-l.--L......,.......L-..l-..J.-~r__'l.--l.--l.---L..,.--'---'--J......L.....1 o 1000 2000 3000 4000 5000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit 2 (3 Inch, Reduced T AVC)
Figure 1S.6-20a
REVISION 9 DECEMBER 2002 1400......-----------------------,
1200
...-
.!:::.. 1000
--
Q)
L-
- J oL-Q) a..
~ 800 J--
600 400 ~O--'--.L.~-'-,..,.OOO--L......J-~...J..-200-r-O .L--I..-I.--'--3000,.-I..-I.--I.-..L.-:4OQOr-!..--l--L.--J.,,-:-lsooo Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Fluid Temperature - Unit 2 (3 Incb, Reduced T AVC)
Figure 15.6*20b
REVISION 9 DECEMBER 2002 5
10
--
4 10
-
W-I N
........
~
I
.J::
- }
~10
........
-
CD
'--'"
OJ 0
-
U 2
10
~
OJ en c
0
~
I- .....
........
0 OJ
- c 1 10 -
0 10 4000 5000 0 1000 2000 3000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coefficient - Unit l (3 Incb, Reduced T Ave)
Figure 15.6-20c
REVISION 9 DECEMBER 2002 2500 - - - - - - - - - - - - - - - - - - - - - - - ,
2000 1000 500 +-J--L.--J..--L..-,--L-...l-..I.-J..-,--l---L..-L-...J.-.-J....-J.-.Io--J..--r-..l-o..d.-..r....-J..-f o 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Finial Safety Analysis Report Res Pressure - Unit 2 (3lncb. Reduced T AVG)
Figure 15.6*20d
REVISION 9 DECEMBER 2002 40.,-----------------------..,
35 20 o 1000 2000 3000 Time (s)
Byron/Braidwood Stations Updated Final Safety Art"alysls Report Core Mixture Level - Uoit 2 (3 ID~b, Redu~ed TAVC)
Figure 15.6*20e
REVISION 9 DECEMBER 2002 350 - - - - - - - - - - - - - - - - - - - - -
---,
300 250
....--.
~
..0 E
.;::::;.. 200
......
<l.>
o 0:::
3:
o 150 G:
(I) en o
- 2:
100 50 0-JL..l..-.l....-J...-~,.._l---l.--L-.J..--r"-l.-..J-~..L---r-...-."--L-J--r---'---!..--a..--'--t o 1000 2000 3000 4000 5000 Time (s)
Byron /Braid wood Statio ns Updated Final Safet y Analy sis Repo rt Core Steam Flow Rate* Unit 2 (3 Incb, Reduced T AVC)
Figure 15.6-20f
REVISION 9 DECEMBER 2002 1800 1600 -
1400 -
~ 1200 E
---
.J::j Q)
-+oJ' 1000 -
0 0::::
3: 800 0
G:
0 en en 0
- E 600-l 400-200 -
~ .AI I 0
0 1000 2000 3000 4000 5000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Uolt 2 (3 loch, Reduced T AVC)
Figure 15.6.20g
REVISION 9 DECEMBER 2002 ZIRLO CLAD FUEL ZIRC4 CLAD FUEL WITH 6K BURN-UP 1800 ----------------------~
1600 1400 1200
.....
Q)
E
~
Q) 0.. 1000 E
~
800 600 400 ~..I.-J--.J...-J...-"......J~--J.--J,___._--.L..--I--L-...J..._r_...I-."""--....I-..J.-,.__..L.--.l..--'---JI.._f o 1000 2000 3000 4000 5000 Time (5)
'Byron/Braidwood Stations Updated Final Safety Analysis Report Peak Clad Temperature - Unit 2 (3 Incb, Nominal T AVC)
Figure 15.6.20h
REVISION 9 DECEMBER 2002 ZIRLO CLAD FUEL
ZIRC4 CLAD FUEt. WITH 6K BURN-UP 1600
_....
I f l
1400-
-
l .L-
-...-
1200 -
,,
Q)
~
- J I b 1000-
~
Q) e::t..
E Q)
~
~
BOO*
1 600
~
400 0 1000 2000 3000 4000 5000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Fluid Temperature - Unit 2 (3 Inch, Nominal T AVC)
Figure 15.6-201
REVISION 9 DECEMBER 2002 ZIRLO CLAD FUEL
ZIRC4 CLAD FUEL WITH 6K BURN-UP 5
10
___ 104 *
-
l...L..
I N
--
oofooJ I
'-
..c::
3
.3 10 -
-....
CD
'+-
'+-
(J,)
0
<..:>
lo...
2 10 -
JJ (J,)
'+-
en it c:
0 I-lo..."
~
-o-J 0
(J,)
1
- r:: 10 -
0 , I 10 I
10'00 2000 3000 4000 5000 0
Time (8)
Byron/Braidwood Stations Updated Final Safety Analysis Report Hellit Transfer Coefficient - Unit 1 (3Inc:b, Nominal T Ave)
Figure 15.6-20j
REVISION 9 DECEMBER 2002 2500 2000
---.
0
"(;5
.... 1500 -
0..
Q.)
- J en en
....
Q.)
C-1000
~
500 0 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated final Safety Analysrs Report
,
Res Pressure - Unit 2 (3Iocb, Nominal T AVC)
Figure 15.6-20k
REVISION 9 DECEMBER 2002 40------------------------,
35 Q)
>
Q)
-J Q)
'-
.......
- J
. ~ 25
- E 20 t5 -l-.J.-....L..-.L-..L..---"!--J--t...-J.-r'....J---J--'-..J--r-....J-...L---'--J-,r-'---'~~-(
o t 000 2000 3000 4000 5000 TIme (s)
Byron/Braidwood Stations .
Updated* Final* Safety Analysis Report Core Mixture Level- Unit 2 (3 Incb, Nominal T Ave)
Figure 15.6*201
REVISION 9 DECEMBER 2002 2500 2000 -
--en
""-
E
---
- e
~
<J) 1500-0 0::::
3:
0 c:: 1000 en en 0
- 2 500 -
LJ. .
.,.....,." , I
,-
0 a 1000 2000 3000 4000 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Steam Flow Rate - Unit 2 (3 Inch, Nominal T AVC)
Figure 15.6*20m
REVISION 9 DECEMBER 2002 1600 1400 -
1200
.....
en E 1000
..0
.::::::::.-
Q>
-+-'
I:'
0:::
800-3=
0 c:::
en en I:'
- aE 600 -
if 400 -
200
.L.-.l
~
.. ... i .... ",WL ...
~
0 0 lOho 2000 3000 .w'o0 5000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit 2 (3 Inch, Nominal T AVC)
Figure 15.6*20n
REVISION 9 DECEMBER 2002 1400 . . . - - - - - - - - - - - - - - - - - - - - - - - ,
1200 I
---
~ 1000-800 -
600-400 "'0_.l--...J.-.-l--l---"_I.-L--.J.-...I--L---r-J._I--.J.-...l--'tOO1rO.-..l._I--...J---'---t 1000 2000 v 4000 ftme*fs1 Byron/Braidwood Stations Updated Final 'Safety Analysis Report Peak Clad Temperature - Unit 2 (4 loch, Reduced T AVC)
Figure 15.6-21 a i
REVISION 9 DECEMBER 2002 1200 I-1100 J
1000 -
I..L.
900 -
CD
'-
- ,
e 800 -
-oJ CD 0-E V
~ 700 600 "J
500
....
I 400 3000 4000 0 1000 2000 Time (5)
Byron/Braidwood Stations Updated' Final Safety Analysis Report Fluid Temperature - Unit 2 (4 Inch, Reduced T AVC)
Figure 15.6-21b
REVISION 9 DECEMBER 2002 s
10
-.
,:.
4 10 - ""'--
.....-
l..I-I
.......
......
N I
'-
.r::
"- 3
~ 10 -
.......
CD
......."
......
'+-
Q) 0 c..:>
'- 2
......
Q) 10 en c:
0 L...
t--
....... ~
0 Q)
- c 1 10 -
0 I 10 3000 4000 0 1000 2000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Heat Transfer Coerncient - Unit :2 (4Intb. Reduced T AVC)
Figure 15.6-21 c
REVISION 9 DECEMBER 2002 2500
. . ..
2000 -
-01500 -
'w a..
-....-..
-~
Q)
~
- J en en
~ 1000 -
CL 500 -
......
0 0 1000 2000 3000 4000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit 2 (4 Inch, Reduced TAvc)
Figure 15.6-21d
REVISION 9 DECEMBER 2002 40-----------------------,
35 20 15 -lO-L.--.l.--..L-.....L..--O""'OO--l---.l.--J1-"--20'O-O...I-..J---I---l-"""":30:JOr:O---~--I.-~4000 1
Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Core Mixture Level - Unit 2 (4 Inch, Reduced TAvc)
Figure 15.6-21e
REVISION 9 DECEMBER 2002 300 - - - - - - - - - - - - - - - - - - - - - - - ,
250 en
'"'""'-
200 E
...0
.:::::.-
.......
Q.)
o 150 0:::
3:
o u-(J)
.~ 100
- E 50 o 1000 Byron/Braidwood Stations Updated 'Final Safety Analysis Report Core Steam Flow Rate - Unit 2 (4 Incb, Reduced T AVC)
Figure 15.6.21f
REVISION 9 DECEMBER 2002 3000 - - - - - - - - - - - - - - - - - - - - - - - - ,
2500
~2000 E
.J:l
.;::::..
---o Q.)
0:::
1500
~
o c;:
(f)
(f)
~ 1000 500 04--l-~__I.___'t.__~..J--...J.-....L---J--r---l-.L..-..J---L..-.....,....-.l----'_I--..l-o--I o 1000 2000 3000 4000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Unit 2 (4 lnch, Reduced TAVC>
Figure 15.6-21g
REVISION 9 DECEMBER 2002 1600 1400 -
!'
--
1200 -
.--
u...
e
~ 1000-
"-
CD 0-E
~
800 600
~
400 3000
- 4000 0 10'00 2000 Time (5)
Byron/Braidwood Stations <
Updated Final Safety Analysis Report Peak Clad Temperature - Unit 2 (4Inc:b, Nominal T AVC)
Figure 15.6-21h I
REVISION 9 DECEMBER 2002 1400
~
1200
--....
.!:::.- 1000 -
Q.)
L..
- s
-t-'
2 Q.)
a..
E Q.) 800 -
I-600 -
~
~
400 0 1000 2000 3000
- 4000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report lFJuid Tempe rature - Unit :2 (4 IDeb, NomJuaJ TAVG)
Figure 15.6-211
REVISION 9 DECEMBER 2002 5
10 4
10 ~ rw-..
--.
lL-I C"J
-:=
~
I
...c:
............ J
- s 10
....... ~
co
--
.....,.......
'+-
Q) 0
'-'
2 10 -
~
Q)
'+-
In c:
C
~ }
L..
I-
.......
C
- c Q)
,
10 0 ,
10 I 0 1000 2000 30'00 ~OOO Time (s)
Byron /Braid wood Statio ns Updated Final Safety Analy sis Report Heat Transf er Coefficient ~ Unit 2 (4 Inch, Nominal T AVG)
Figure 15.6-21)
REVISION 9 DECEMBER 2002 2500 2000 01500-
'en 0-
~
Q) t.-
- ,
(J)
(J)
Q)
O-t.- 1000 -
500 -
0 I 20'00 30'00 4000 0 1000 Time (s)
Byron/Braidwood Stations Updated Final Safety Analysis Report Res Pressure - Unit 2 (4 lotb, Nominal TAvc)
Figure 15.6-21k
REVISION 9 DECEMBER 2002 40 -- -- -- -- -- -- -- -- -- -- -- -,
35 20 15 -l0-'-- .J...-.. J-..J--or-OO--l----L.......JL.-..J...-20TOO---L-L-.J--..L--3..,OOr-O--l 1
---l---J,-~4000 Time (s)
Byron /Braid wood Stations Updated Final Safety Analy sis Report Core Mixtur e Level- Unit 2 (4 Inch, Nominal T AVG)
Figure 15.6-211
REVISION 9 DECEMBER 2002 2500 2000 cn1500-
...........
E
..0
.::::::--
........
<D a::: 1000-0 3=
0 u-(f)
(f) 0 500 -
.:a:
W'(T
.All ....
- 0- 1
-500 . I 0 1000 2000 3000 4000 Time (5)
Byron/Braidwood Stations Updated Final Safety Anaiysis Report Core Steam Flow Rate* Uoit 2 (4 loch, Nominal T AVG)
Figure 15.6-21 m a
REVISION 9 DECEMBER 2002 3000 - - - - - - - - - - - - - - - - - - - - - - - - - ,
2500
-u> 2000
-......
E
..0
~
Q) '
~ 1500 3=
o G:
en en
~ 1000 500 o-l---'--J---J..--L.---y-.L....-~.L.-..J-_r_-l-~~--"__,r_.l--~-l..--I.-_f o 1000 2000 3000 4000 Time (5)
Byron/Braidwood Stations Updated Final Safety Analysis Report Break Flow - Vnlt 2 (4Ioc:h, Nominal T AVC)
Figure 15.6-21n
DECEMBER 2010 Security - Related Information Figure Withheld Under 10 CFR 2.390 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPOR T FIGURE 15,7*1 SPENT FUEL STORAGE POOL PLAN
DECEMBER 2010 Security - Related Information Figure Withheld Under 10 CFR 2.390 BYRON/BRAIDWOOD STATIONS UPDATED FINAL SAFETY ANALYSIS REPORT FIGURE 15.7-2 SPEN T FUE L STORAGE POO L SECT I ON