DCL-17-038, Diablo Canyon Power Plant, Units 1 & 2, Revised Updated Final Safety Analysis Report, Rev. 23, Chapter 15, Accident Analyses
Text
DCPP UNITS 1 &
2 FSAR UPDATE i Revision 23 December 2016 CHAPTER 15 ACCIDENT ANALYSES CONTENTS Section Title
Page 15 ACCIDENT ANALYSES 15-1
15.1 CONDITION I - NORMAL OPERATION AND OPERATIONAL TRANSIENTS (INITIAL CONDITIONS) 15.1-1
15.
1.1 INTRODUCTION
15.1-1
15.1.2 COMPUTER CODES UTILIZED 15.1-2 15.1.2.1 FACTRAN 15.1-2 15.1.2.2 LOFTRAN 15.1-3 15.1.2.3 PHOENIX-P 15.1-4 15.1.2.4 ANC 15.1-4 15.1.2.5 TWINKLE 15.1-4 15.1.2.6 THINC 15.1-5 15.1.2.7 RETRAN-02 15.1-5 15.1.2.8 RETRAN-02W 15.1-5 15.1.2.9 NOTRUMP 15.1-6 15.1.2.10 SBLOCTA (LOCTA-IV) 15.1-6 15.1.2.11 WCOBRA/TRAC 15.1-6 15.1.2.12 HOTSPOT 15.1-6 15.1.2.13 MONTECF 15.1-7 15.1.2.14 COCO 15.1-7 15.1.3 OPTIMIZATION OF CONTROL SYSTEMS 15.1-7 15.1.4 INITIAL POWER CONDITIONS ASSUMED IN ACCIDENT
ANALYSES 15.1-8 15.1.4.1 Power Rating 15.1-8 15.1.4.2 Initial Conditions 15.1-9 15.1.4.3 Power Distribution 15.1-9 15.1.5 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN
ACCIDENT ANALYSES 15.1-11 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page ii Revision 23 December 2016 15.1.6 CALORIMETRIC ERRORS - POWER RANGE NEUTRON FLUX 15.1-12 15.1.7 ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS 15.1-12
15.1.8 REACTIVITY COEFFICIENTS 15.1-13
15.1.9 FISSION PRODUCT INVENTORIES 15.1-13
15.1.10 RESIDUAL DECAY HEAT 15.1-13 15.1.10.1 Fission Product Decay 15.1-14 15.1.10.2 Decay of U-238 Capture Products 15.1-14 15.1.10.3 Residual Fissions 15.1-15 15.1.10.4 Distribution of Decay Heat Following Loss-of-Coolant Accident 15.1-15
15.1.11 REFERENCES 15.1.16
15.2 CONDITION II - FAULTS OF MODERATE FREQUENCY 15.2-1
15.2.1 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL FROM A SUBCRITICAL CONDITION 15.2-2 15.2.1.1 Acceptance Criteria 15.2-2 15.2.1.2 Identification of Causes and Accident Description 15.2-2 15.2.1.3 Analysis of Effects and Consequences 15.2-4 15.2.1.4 Results 15.2-5 15.2.1.5 Conclusions 15.2-5
15.2.2 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL AT POWER 15.2-6 15.2.2.1 Acceptance Criteria 15.2-6 15.2.2.2 Identification of Causes and Accident Description 15.2-6 15.2.2.3 Analysis of Effects and Consequences 15.2-7 15.2.2.4 Results 15.2-8 15.2.2.5 Conclusions 15.2-10
15.2.3 ROD CLUSTER CONTROL ASSEMBLY MISOPERATION 15.2-11 15.2.3.1 Acceptance Criteria 15.2-11 15.2.3.2 Identification of Causes and Accident Description 15.2-11 15.2.3.3 Analysis of Effects and Consequences 15.2-13 15.2.3.4 Results 15.2-14 15.2.3.5 Conclusions 15.2-16 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page iii Revision 23 December 2016 15.2.4 UNCONTROLLED BORON DILUTION 15.2-16 15.2.4.1 Acceptance Criteria 15.2-16 15.2.4.2 Identification of Causes and Accident Description 15.2-16 15.2.4.3 Analysis of Effects and Consequences 15.2-17 15.2.4.4 Conclusions 15.2-21
15.2.5 PARTIAL LOSS OF FORCED REACTOR COOLANT FLOW 15.2-23 15.2.5.1 Acceptance Criteria (Historical) 15.2-23 15.2.5.2 Identification of Causes and Accident Description (Historical) 15.2-23 15.2.5.3 Analysis of Effects and Consequences (Historical) 15.2-24 15.2.5.4 Results (Historical) 15.2-25 15.2.5.5 Conclusions (Historical) 15.2.25
15.2.6 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP 15.2-25 15.2.6.1 Identification of Causes and Accident Description (Historical) 15.2-25 15.2.6.2 Analysis of Effects and Consequences (Historical) 15.2-26 15.2.6.3 Results (Historical) 15.2-27 15.2.6.4 Conclusions (Historical) 15.2-27
15.2.7 LOSS OF EXTERNAL ELECT RICAL LOAD AND/OR TURBINE TRIP 15.2-27 15.2.7.1 Acceptance Criteria 15.2-27 15.2.7.2 Identification of Causes and Accident Description 15.2-28 15.2.7.3 Analysis of Effects and Consequences 15.2-29 15.2.7.4 Results 15.2-32 15.2.7.5 Conclusions 15.2-33
15.2.8 LOSS OF NORMAL FEEDWATER 15.2-34 15.2.8.1 Acceptance Criteria 15.2-34 15.2.8.2 Identification of Causes and Accident Description 15.2-34 15.2.8.3 Analysis of Effects and Consequences 15.2-35 15.2.8.4 Results 15.2-36 15.2.8.5 Conclusions 15.2-37
15.2.9 LOSS OF OFFSITE POWER TO THE STATION AUXILIARIES 15.2-37 15.2.9.1 Acceptance Criteria 15.2-37 15.2.9.2 Identification of Causes and Accident Description 15.2-37 15.2.9.3 Analysis of Effects and Consequences 15.2-38 15.2.9.4 Results 15.2-39 15.2.9.5 Conclusions 15.2-39
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page iv Revision 23 December 2016 15.2.10 EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTIONS 15.2-40 15.2.10.1 Acceptance Criteria 15.2-40 15.2.10.2 Identification of Causes and Accident Description 15.2-40 15.2.10.3 Analysis of Effects and Consequences 15.2-40 15.2.10.4 Results 15.2-41 15.2.10.5 Conclusions 15.2-42
15.2.11 SUDDEN FEEDWATER TEMPERATURE REDUCTION 15.2-42 15.2.11.1 Acceptance Criteria 15.2-42 15.2.11.2 Identification of Causes and Accident Description 15.2-42 15.2.11.3 Analysis of Effects and Consequences 15.2-43 15.2.11.4 Results 15.2-44 15.2.11.5 Conclusions 15.2.44
15.2.12 EXCESSIVE LOAD INCREASE INCIDENT 15.2-45 15.2.12.1 Acceptance Criteria 15.2-45 15.2.12.2 Identification of Causes and Accident Description 15.2-45 15.2.12.3 Analysis of Effects and Consequences 15.2-45 15.2.12.4 Results 15.2-46 15.2.12.5 Conclusions 15.2-47
15.2.13 ACCIDENTAL DEP RESSURIZATION OF THE REACTOR COOLANT SYSTEM 15.2-48 15.2.13.1 Acceptance Criteria 15.2-48 15.2.13.2 Identification of Causes and Accident Description 15.2-48 15.2.13.3 Analysis of Effects and Consequences 15.2-48 15.2.13.4 Results 15.2-49 15.2.13.5 Conclusions 15.2-49
15.2.14 ACCIDENTAL DEP RESSURIZATION OF THE MAIN STEAM SYSTEM 15.2-49 15.2.14.1 Acceptance Criteria 15.2-49 15.2.14.2 Identification of Causes and Accident Description 15.2-50 15.2.14.3 Analysis of Effects and Consequences 15.2-50 15.2.14.4 Conclusions 15.2-51
15.2.15 SPURIOUS OPERATION OF THE SAFETY INJECTION SYSTEM AT POWER 15.2-51 15.2.15.1 Acceptance Criteria 15.2-51 15.2.15.2 Spurious Safety Injection (SSI) DNBR Analysis 15.2-51 15.2.15.3 Spurious Safety Injection (SSI) Pressurizer Overfill Analysis 15.2-54 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page v Revision 23 December 2016 15.2.16 REFERENCES 15.2-59
15.3 CONDITION III - INFREQUENT FAULTS 15.3-1
15.3.1 LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES THAT ACTUATE EMERGENCY CORE COOLING SYSTEM 15.3-1 15.3.1.1 Acceptance Criteria 15.3-1 15.3.1.2 Identification of Causes and Accident Description 15.3-2 15.3.1.3 Analysis of Effects and Consequences 15.3-3 15.3.1.4 Results 15.3-4 15.3.1.5 Conclusions 15.3-5
15.3.2 MINOR SECONDARY SYSTEM PIPE BREAKS 15.3-6 15.3.2.1 Acceptance Criteria 15.3-6 15.3.2.2 Identification of Causes and Accident Description 15.3-6 15.3.2.3 Analysis of Effects and Consequences 15.3-6 15.3.2.4 Conclusions 15.3-7
15.3.3 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO AN IMPROPER POSITION 15.3-7 15.3.3.1 Acceptance Criteria 15.3-7 15.3.3.2 Identification of Causes and Accident Description 15.3-7 15.3.3.3 Analysis of Effects and Consequences 15.3-8 15.3.3.4 Results 15.3-8 15.3.3.5 Conclusions 15.3-9
15.3.4 COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.3-9 15.3.4.1 Acceptance Criteria 15.3-9 15.3.4.2 Identification of Causes and Accident Description 15.3-9 15.3.4.3 Analysis of Effects and Consequences 15.3-10 15.3.4.4 Results 15.3-11 15.3.4.5 Conclusions 15.3-11
15.3.5 SINGLE ROD CLUSTER CONT ROL ASSEMBLY WITHDRAWAL AT FULL POWER 15.3-12 15.3.5.1 Acceptance Criteria 15.3-12 15.3.5.2 Identification of Causes and Accident Description 15.3-12 15.3.5.3 Analysis of Effects and Consequences 15.3-12 15.3.5.4 Results 15.3-13 15.3.5.5 Conclusions 15.3-13 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page vi Revision 23 December 2016 15.
3.6 REFERENCES
15.3-14
15.4 CONDITION IV - LIMITING FAULTS 15.4-1
15.4.1 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOCA)15.4-2 15.4.1.1 Acceptance Criteria 15.4-2 15.4.1.2 Background of Best Estimate Large Break LOCA 15.4-3 15.4.1.3 WCOBRA/TRAC Thermal-hydraulic Computer Code 15.4-4 15.4.1.4 Thermal Analysis 15.4-6 15.4.1.4A Unit 1 Best Estimate Large Break LOCA Evaluation Model 15.4-10 15.4.1.5A Unit 1 Containment Backpressure 15.4-13 15.4.1.6A Unit 1 Reference Transient Description 15.4-13 15.4.1.7A Unit 1 Sensitivity Studies 15.4-14 15.4.1.8A Unit 1Additional Evaluations 15.4-16 15.4.1.9A Unit 1 10 CFR 50.46 Results 15.4-16 15.4.1.10A Unit 1 Plant Operating Range 15.4-17 15.4.1.4B Unit 2 Best Estimate Large Break LOCA Evaluation Model 15.4-18 15.4.1.5B Unit 2 Containment Backpressure 15.4-19 15.4.1.6B Unit 2 Confirmatory Studies 15.4-20 15.4.1.7B Unit 2 Uncertainty Evaluation 15.4-20 15.4.1.8B Unit 2 Limiting PCT Transient Description 15.4-21 15.4.1.9B Unit 2 10 CFR 50.46 Results 15.4-21 15.4.1.10B Unit 2 Plant Operating Range 15.4-22 15.4.1.11 Conclusions (Common) 15.4-22
15.4.2 MAJOR SECONDARY SYSTEM PIPE RUPTURE 15.4-23 15.4.2.1 Rupture of a Main Steam Line at Hot Zero Power 15.4-24 15.4.2.2 Major Rupture of a Main Feedwater Pipe 15.4-30 15.4.2.3 Rupture of a Main Steam Line at Full Power 15.4-35 15.4.2.4 Major Rupture of a Main Feedwater Pipe for Pressurizer Filling 15.4-39
15.4.3 STEAM GENERATOR TUBE RUPTURE (SGTR) 15.4-44 15.4.3.1 Acceptance Criteria 15.4-44 15.4.3.2 Identification of Causes and Accident Description 15.4-44 15.4.3.3 Analysis of Effects and Consequences 15.4-48 15.4.3.4 Conclusions 15.4-55
15.4.4 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR 15.4-55 15.4.4.1 Identification of Causes and Accident Description 15.4-55 15.4.4.2 Analysis of Effects and Consequences 15.4-55 15.4.4.3 Results 15.4-58 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page vii Revision 23 December 2016 15.4.4.4 Conclusions 15.4-58 15.4.5 FUEL HANDLING ACCIDENT 15.4-58 15.4.5.1 Acceptance Criteria 15.4-58 15.4.5.2 Identification of Causes and Accident Description 15.4-58 15.4.5.3 Results 15.4-62 15.4.5.4 Conclusions 15.4-63 15.4.6 RUPTURE OF A CONTROL ROD DRIVE MECHANISM
HOUSING (ROD CLUSTER CONTROL ASSEMBLY EJECTION) 15.4-63 15.4.6.1 Acceptance Criteria 15.4-63 15.4.6.2 Identification of Causes and Accident Description 15.4-63 15.4.6.3 Analysis of Effects and Consequences 15.4-66 15.4.6.4 Results 15.4-70 15.4.6.5 Conclusions 15.4-71 15.4.7 RUPTURE OF A WASTE GAS DECAY TANK 15.4-72 15.4.8 RUPTURE OF A LIQUID HOLDUP TANK 15.4-72 15.4.9 RUPTURE OF VOLUME CONTROL TANK 15.4-72 15.4.10 REFERENCES 15.4-72 15.5 RADIOLOGICAL CONSEQUENCES OF PLANT
ACCIDENTS 15.5-1 15.5.1 DESIGN BASES 15.5-2 15.5.2 APPROACH TO ANALYSIS OF RADIOLOGICAL EFFECTS OF
ACCIDENTS 15.5-5 15.5.3 ACTIVITY INVENTORIES IN THE PLANT PRIOR TO ACCIDENTS 15.5-6
15.5.4 EFFECTS OF PLUTONIUM INVENTORY ON POTENTIAL
ACCIDENT DOSES (Historical) 15.5-8 15.5.5 POST-ACCIDENT METEOROLOGICAL CONDITIONS 15.5-9 15.5.6 RATES OF ISOTOPE INHALATION 15.5-15 15.5.7 POPULATION DISTRIBUTION 15.5-15 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page viii Revision 23 December 2016 15.5.8 RADIOLOGICAL ANALYSIS PROGRAMS 15.5-15 15.5.8.1 Description of the EMERALD (Revision 1) and EMERALD-NORMAL Program 15.5-15 15.5.8.2 Description of the LOCADOSE Program 15.5-16 15.5.8.3 Description of the ORIGEN-2 Program 15.5-17 15.5.8.4 Description of the ISOSHLD Program 15.5-17 15.5.8.5 Description of the ISOSHLD II Program 15.5-17 15.5.8.6 Description of the RADTRAD Program 15.5-17
15.5.9 (Deleted) 15.5-18
15.5.10 RADIOLOGICAL CONSEQUENCES OF CONDITION II FAULTS 15.5-18 15.5.10.1 Acceptance Criteria 15.5-18 15.5.10.2 Identification of Causes and Accident Description 15.5-18 15.5.10.3 Conclusions 15.5-20
15.5.11 RADIOLOGICAL CONSEQUENCES OF A SMALL-BREAK LOCA 15.5-20 15.5.11.1 Acceptance Criteria 15.5-20 15.5.11.2 Identification of Causes and Accident Description 15.5-21 15.5.11.3 Conclusions 15.5-22
15.5.12 RADIOLOGICAL CONSEQUENCES OF MINOR SECONDARY SYSTEM PIPE BREAKS 15.5-23 15.5.12.1 Acceptance Criteria 15.5-23 15.5.12.2 Identification of Causes and Accident Description 15.5-23 15.5.12.3 Conclusions 15.5-24
15.5.13 RADIOLOGICAL CONSEQUENCES OF INADVERTENT LOADING OF A FUEL ASSEMBLY INTO AN IMPROPER POSITION 15.5-24 15.5.13.1 Acceptance Criteria 15.5-24 15.5.13.2 Identification of Causes and Accident Description 15.5-24 15.5.13.3 Conclusions 15.5-24
15.5.14 RADIOLOGICAL CONSEQUENCES OF COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.5-25 15.5.14.1 Acceptance Criteria 15.5-25 15.5.14.2 Identification of Causes and Accident Description 15.5-25 15.5.14.3 Conclusions 15.5-25
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page ix Revision 23 December 2016 15.5.15 RADIOLOGICAL CONSEQUENCES OF AN UNDERFREQUENCY ACCIDENT 15.5-26 15.5.15.1 Acceptance Criteria 15.5-26 15.5.15.2 Identification of Causes and Accident Description 15.5-26 15.5.15.3 Conclusions 15.5-27
15.5.16 RADIOLOGICAL CONSEQUENCES OF A SINGLE ROD CLUSTER CONTROL ASSEMBL Y WITHDRAWAL AT FULL POWER 15.5-27 15.5.16.1 Acceptance Criteria 15.5-27 15.5.16.2 Identification of Causes and Accident Description 15.5-27 15.5.16.3 Conclusions 15.5-28
15.5.17 RADIOLOGICAL CONSEQUENCES OF MAJOR RUPTURE OF PRIMARY COOLANT PIPES 15.5-28 15.5.17.1 Acceptance Criteria 15.5-28 15.5.17.2 Identification of Causes and Accident Description 15.5-29 15.5.17.3 Conclusions 15.5-45
15.5.18 RADIOLOGICAL CONSEQUENCES OF A MAJOR STEAM PIPE RUPTURE 15.5-46 15.5.18.1 Acceptance Criteria 15.5-46 15.5.18.2 Identification of Causes and Accident Description 15.5-46 15.5.18.3 Conclusions 15.5-51
15.5.19 RADIOLOGICAL CONSEQUENCES OF A MAJOR RUPTURE OF A MAIN FEEDWATER PIPE 15.5-52 15.5.19.1 Acceptance Criteria 15.5-52 15.5.19.2 Identification of Causes and Accident Description 15.5-52 15.5.19.3 Conclusions 15.5-53
15.5.20 RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE (SGTR) 15.5-53 15.5.20.1 Acceptance Criteria 15.5-53 15.5.20.2 Identification of Causes and Accident Description 15.5-54 15.5.20.3 Conclusions 15.5-61
15.5.21 RADIOLOGICAL CONSEQUENCES OF A LOCKED ROTOR ACCIDENT 15.5-62 15.5.21.1 Acceptance Criteria 15.5-62 15.5.21.2 Identification of Causes and Accident Description 15.5-62 15.5.21.3 Conclusions 15.5-64 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 CONTENTS (Continued)
Section Title Page x Revision 23 December 2016 15.5.22 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT 15.5-65 15.5.22.1 Fuel Handling Accident in the Fuel Handling Area 15.5-65 15.5.22.2 Fuel Handling Accident Inside Containment 15.5-68 15.5.22.3 Conclusion, Fuel Handling Accidents 15.5-71
15.5.23 RADIOLOGICAL CONSEQUENCES OF A ROD EJECTION ACCIDENT 15.5-71 15.5.23.1 Acceptance Criteria 15.5-71 15.5.23.2 Identification of Causes and Accident Description 15.5-72 15.5.23.3 Conclusion 15.5-73
15.5.24 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A WASTE GAS DECAY TANK 15.5-74 15.5.24.1 Acceptance Criteria 15.5-74 15.5.24.2 Identification of Causes and Accident Description 15.5-74 15.5.24.3 Conclusion 15.5-76
15.5.25 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A LIQUID
HOLDUP TANK 15.5-76 15.5.25.1 Acceptance Criteria 15.5-76 15.5.25.2 Identification of Causes and Accident Description 15.5-77 15.5.25.3 Conclusions 15.5-78 15.5.26 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A VOLUME CONTROL TANK 15.5-78 15.5.26.1 Acceptance Criteria 15.5-78 15.5.26.2 Identification of Causes and Accident Description 15.5-78 15.5.26.3 Conclusions 15.5-80 15.5.27 REFERENCES 15.5-80
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 TABLES Table Title xi Revision 23 December 2016 15.0-1 Regulatory Guide 1.70, Revision 1, Applicability Matrix
15.1-1 Nuclear Steam Supply System Power Ratings
15.1-2 Trip Points and Time Delays to Trip Assumed in Accident Analyses
15.1-3 Deleted in Revision 10
15.1-4 Summary of Initial Conditions and Computer Codes Used
15.2-1 Time Sequence of Events for Condition II Events
15.2-2 Deleted in Revision 6
15.3-1 Time Sequence of Events - Small Break LOCA
15.3-2 Fuel Cladding Results - Small Break LOCA
15.3-3 Time Sequence of Events for Condition III Events
15.4-A Deleted in Revision 12
15.4-B Deleted in Revision 12
15.4.1-1A Unit 1 Best Estimate Large Break LOCA Time Sequence of Events for the Reference Transient
15.4.1-1B Unit 2 Best Estimate Large Break Sequence of Events for Limiting PCT
Case
15.4.1-2A Unit 1 Best Estimate Large Break LOCA Analysis Results
15.4.1-2B Unit 2 Best Estimate Large Break LOCA Analysis Results
15.4.1-3A Unit 1 Key Best Estimate Large Break LOCA Parameters and Reference Transient Assumptions
15.4.1-3B Unit 2 Key Best Estimate Large Break LOCA Parameters and Initial Transient Assumptions
15.4.1-4A Unit 1 Sample of Best Estimate Sensitivity Analysis Results for Original Analysis DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 TABLES (Continued)
Table Title xii Revision 23 December 2016 15.4.1-4B Unit 2 Results From Confirmatory Studies
15.4.1-5A Unit 1 Containment Back Pressure Analysis Input Parameters Used for Best Estimate LOCA Analysis
15.4.1-5B Unit 2 Containment Back Pressure Analysis Input Parameters Used for Best Estimate LBLOCA Analysis
15.4-6 Deleted in Revision 18
15.4-7 Deleted in Revision 18
15.4.1-7A Unit 1 Plant Operating Range Allowed by the Best-Estimate Large Break LOCA Analysis
15.4.1-7B Unit 2 Plant Operating Range Allowed by the Best-Estimate Large Break LOCA Analysis
15.4-8 Time Sequence of Events for Major Secondary System Pipe Ruptures
15.4-8A Deleted in Revision 19
15.4-9 Deleted in Revision 19
15.4-10 Summary of Results for Locked Rotor Transient
15.4-11 Typical Parameters Used in the VANTAGE 5 Reload Analysis of the Rod Cluster Control Assembly Ejection Accident
15.4-12 Operator Action Times for Design Basis SGTR Analysis
15.4-13 Deleted in Revision 20
15.4-13A Timed Sequence of Events - SGTR MTO Analysis
15.4-13B Timed Sequence of Events - SGTR Dose Analysis
15.4-14 Mass Release Results -
SGTR Dose Input Analysis
15.4-14A Deleted in Revision 19
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 TABLES (Continued)
Table Title xiii Revision 23 December 2016 15.5-1 Reactor Coolant Fission and Corrosion Product Activities During Steady State Operation and Plant Shutdown Operation 15.5-2 Results of Study of Effects of Plutonium on Accident Doses (Historical)
15.5-3 Design Basis Post-accident Atmospheric Dilution Factors
15.5-4 Expected Post-accident Atmospheric Dilution Factors
15.5-5 Atmospheric Dilution Factors 15.5-6 Assumed Onsite Atmospheric Dilution Factors for the Control Room
15.5-7 Breathing Rates Assumed in Analysis
15.5-8 Population Distribution
15.5-9 Summary of Offsite Doses from Loss of Electrical Load
15.5-10 Summary of Offsite Doses from a Small Loss-of-Coolant Accident
15.5-11 Summary of Offsite Doses from an Underfrequency Accident
15.5-12 Summary of Offsite Doses from a Single Rod Cluster Control Assembly Withdrawal
15.5-13 Calculated Activity Releases from LOCA - Expected Case
15.5-14 Calculated Activity Releases from LOCA - Design Basis Case
15.5-15 Thyroid Dose, 2-hour, Containme nt Leakage, Expected Case
15.5-16 Deleted in Revision 22
15.5-17 Thyroid Dose, 30-day, Containme nt Leakage, Expected Case
15.5-18 Deleted in Revision 22
15.5-19 Whole Body Dose, 2-hour, Containment Leakage, Expected Case
15.5-20 Deleted in Revision 22
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 TABLES (Continued)
Table Title xiv Revision 23 December 2016 15.5-21 Whole Body Dose, 30-day, Containment Leakage, Expected Case
15.5-22 Deleted in Revision 22
15.5-23 Summary of Exposure from Containment Leakage
15.5-24 Assumptions Used to Calculate Offsite Exposures from Post-LOCA Circulation Loop Leakage in the Auxiliary Building
15.5-25 Deleted in Revision 22 15.5-26 Percentage Occurrence of Wind Direction and Calm Winds Expressed as Percentage of Total Hourly Observations Within Each Season at the Site
(250-Foot Level)
15.5-27 Diablo Canyon Power Plant Site Probability of Persistence Offshore Wind Direction Sectors (250-Foot Level)
15.5-28 Assumptions Used to Calcu-late Onshore Controlled Containment Venting
15.5-29 Onshore Controlled Containment Venting Exposures
15.5-30 Atmospheric Dispersion Factors for Onshore Controlled Containment Venting (Stability Category D)
15.5-31 Control Room Infiltration Assumed for Radiol ogical Exposure Calculations
15.5-32 Assumptions Used to Calculate P ost-accident Control Room Radiological Exposures
15.5-33 Estimated Post-accident Exposure to Control Room Personnel
15.5-34 Steam Releases Following a Major Steam Line Break
15.5-35 Deleted in Revision 16
15.5-36 Deleted in Revision 16
15.5-37 Deleted in Revision 7
15.5-38 Deleted in Revision 7
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 TABLES (Continued)
Table Title xv Revision 23 December 2016 15.5-39 Deleted in Revision 7
15.5-40 Long-term Activity Release Fractions for Fuel Failure Accidents (Historical)
15.5-41 Activity Releases Following a Locked Rotor Accident (Curies)
15.5-42 Summary of Offsite Doses from a Locked Rotor Accident
15.5-43 Deleted in Revision 16
15.5-44 Composite Source Term for Fuel Handling Accident in the Fuel Handling Building 15.5-45 Assumptions for Fuel Handling Accident in the Fuel Handling Area
15.5-46 Deleted in Revision 16
15.5-47 Summary of Doses from Fuel Handling Accident in the Fuel Handling Area
15.5-48 Design Inputs and Assumptions for Fuel Handling Accidents Inside Containment
15.5-49 Activity Releases from Fuel Handling Accident Inside Containment (Curies) 15.5-50 Summary of Offsite Doses from Fuel Handling Accident Inside Containment
15.5-51 Activity Releases Following A Rod Ejection Accident (Curies)
15.5-52 Summary of Offsite Doses from a Rod Ejection Accident
15.5-53 Summary of Offsite Doses from a Rupture of a Gas Decay Tank
15.5-54 Deleted in Revision 11
15.5-55 Deleted in Revision 11
15.5-56 Summary of Offsite Doses from Rupture of a Liquid Holdup Tank
15.5-57 Summary of Offsite Doses from Rupture of a Volume Control Tank
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 TABLES (Continued)
Table Title xvi Revision 23 December 2016 15.5-58 Deleted in Revision 16
15.5-59 Deleted in Revision 16
15.5-60 Deleted in Revision 16
15.5-61 Deleted in Revision 22
15.5-62 Deleted in Revision 22
15.5-63 Post-LOCA Doses with Margin Recirculation Loop Leakage
15.5-64 Parameters Used in Evaluating Radio logical Consequences For SGTR Analysis 15.5-65 Iodine Specific Activities in the Primary and Secondary Coolant -
SGTR Analysis
15.5-66 Iodine Spike Appearance Rates - SGTR Analysis (Curies/Second)
15.5-67 Noble Gas Specific Activities in the Reactor Coolant Based on 1% Fuel Defects - SGTR Analysis
15.5-68 Atmospheric Dispersion Factors and Breathing Rates - SGTR Analysis
15.5-69 Thyroid Dose Conversion Factors - SGTR Analysis
15.5-70 Average Gamma and Beta Energy for Noble Gases - SGTR Analysis
15.5-71 Offsite Radiation Doses from SGTR Accident
15.5-72 Control Room Parameters Used in Evalu ating Radiolo gical Consequences for SGTR Analysis
15.5-73 Deleted in Revision 16
15.5-74 Control Room Radiation Doses from Airborne Activity in SGTR Accident
15.5-75 Summary of Post-LOCA Doses from Various Pathways (DF of 100)
15.5-76 Whole Body Dose Conversion Factors Dose Equivalent XE-133 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES Figure Title xvii Revision 23 December 2016 15.1-1 Illustration of Overpower and Overtemperature T Protection 15.1-2 Rod Position Versus Time on Reactor Trip
15.1-3 Normalized RCCA Reactivity Worth Versus Percent Insertion 15.1-4 Normalized RCCA Bank Reactivity Worth Versus Time After Trip
15.1-5 Doppler Power Coefficient Used in Accident Analysis
15.1-6 Residual Decay Heat (Best Estimate LBLOCA 1979 ANS Decay Heat)
15.1-7 1979 ANS Decay Heat Curve (Used for Non-LOCA Analyses)
15.1-8 Fuel Rod Cross Section
15.2-1 Deleted in Revision 6
15.2-2 Deleted in Revision 6
15.2-3 Deleted in Revision 6
15.2-4 Deleted in Revision 6
15.2-5 Deleted in Revision 6
15.2-6 Deleted in Revision 3
15.2-7 Deleted in Revision 3
15.2-8 Deleted in Revision 3
15.2-9 Deleted in Revision 3
15.2-10 Deleted in Revision 3
15.2-11 Deleted in Revision 6
15.2-12 Deleted in Revision 6
15.2-13 Deleted in Revision 6
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xviii Revision 23 December 2016 15.2-14 Deleted in Revision 6
15.2-15 Deleted in Revision 6
15.2-16 Deleted in Revision 6
15.2-17 Deleted in Revision 6
15.2-18 Deleted in Revision 6
15.2-19 Deleted in Revision 6
15.2-20 Deleted in Revision 3
15.2-21 Deleted in Revision 3
15.2-22 Deleted in Revision 3
15.2-23 Deleted in Revision 3
15.2-24 Deleted in Revision 3
15.2-25 Deleted in Revision 3
15.2-26 Deleted in Revision 3
15.2-27 Deleted in Revision 3
15.2-28 Deleted in Revision 3
15.2-29 Deleted in Revision 6
15.2-30 Deleted in Revision 6
15.2-31 Deleted in Revision 6
15.2-32 Deleted in Revision 6
15.2-33 Deleted in Revision 6
15.2-34 Deleted in Revision 6
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xix Revision 23 December 2016 15.2-35 Deleted in Revision 6
15.2-36 Deleted in Revision 6
15.2-37 Deleted in Revision 6
15.2-38 Deleted in Revision 3
15.2-39 Deleted in Revision 3
15.2-40 Deleted in Revision 3
15.2-41 Deleted in Revision 6
15.2-42 Deleted in Revision 6
15.2-43 Deleted in Revision 6
15.2-44 Deleted in Revision 6
15.2-45 Deleted in Revision 6
15.2-46 Deleted in Revision 6
15.2-47 Deleted in Revision 3
15.2-48 Deleted in Revision 6
15.2-49 Deleted in Revision 6
15.2-50 Deleted in Revision 6
15.2-51 Deleted in Revision 6
15.2-52 Deleted in Revision 6
15.2-53 Deleted in Revision 6
15.2-54 Deleted in Revision 6
15.2-55 Deleted in Revision 6
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xx Revision 23 December 2016 15.2-56 Deleted in Revision 6
15.2-57 Deleted in Revision 6
15.2-58 Deleted in Revision 6
15.2-59 Deleted in Revision 6
15.2-60 Deleted in Revision 6
15.2-61 Deleted in Revision 6
15.2-62 Deleted in Revision 6
15.2-63 Deleted in Revision 6
15.2-64 Deleted in Revision 6
15.2-65 Deleted in Revision 6
15.2-66 Deleted in Revision 6
15.2-67 Deleted in Revision 6
15.2-68 Deleted in Revision 6
15.2-69 Deleted in Revision 6
15.2-70 Deleted in Revision 6
15.2-71 Deleted in Revision 6
15.2.1-1 Uncontrolled Rod Withdrawal from a Subcritical Condition - Neutron Flux Versus Time
15.2.1-2 Uncontrolled Rod Withdrawal from a Subcritical Condition - Average Channel Thermal Flux Versus Time
15.2.1-3 Uncontrolled Rod Withdrawal from a Subcritical Condition - Temperature Versus Time, Reactivity Insertion Rate 75 x 10
-5K/sec DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxi Revision 23 December 2016 15.2.2-1 Rod Withdrawal at Power - Minimum Feedback, 75 pcm/sec Insertion Rate - Pressurizer Pressure and Neutron Flux Versus Time 15.2.2-2 Rod Withdrawal at Power - Minimum Feedback, 75 pcm/sec Insertion Rate - DNBR and Tavg Versus Time
15.2.2-3 Rod Withdrawal at Power - Minimum Feedback, 3 pcm/sec Insertion Rate - Pressurizer Pressure and Neutron Flux Versus Time
15.2.2-4 Rod Withdrawal at Power - Minimum Feedback, 3 pcm/sec Insertion Rate - DNBR and T avg Versus Time
15.2.2-5 Rod Withdrawal at Power - Reactivity Insertion Rate vs. DNBR for 100%
Power Cases
15.2.2-6 Rod Withdrawal at Power - Reactivity Insertion Rate vs. DNBR for 60%
Power Cases
15.2.2-7 Rod Withdrawal at Power - Reactivity Insertion Rate vs. DNBR for 10%
Power Cases
15.2.3-1 Transient Response to Dropped Rod Cluster Control Assembly, Nuclear Power and Core Heat Flux Versus Time
15.2.3-2 Transient Response to Dropped Rod Cluster Control Assembly, Average Coolant Temperature and Pressurizer Pressure Versus Time
15.2.4-1 Variation in Reactivity Insertion Rate with Initial Boron Concentration for a Dilution Rate of 262 gpm
15.2.5-1 All Loops Operating, Two Loops Coasting Down - Core Flow Versus Time (Historical)
15.2.5-2 All Loops Operating, Two Loops Coasting Down - Failed Loop Flow Versus Time (Historical)
15.2.5-3 All Loops Operating, Two Loops Coasting Down - Heat Flux Versus Time (Historical)
15.2.5-4 All Loops Operating, Two Loops Coasting Down - Nuclear Power Versus Time (Historical)
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxii Revision 23 December 2016 15.2.5-5 All Loops Operating, Two Loops Coasting Down, DNBR Versus Time (Historical) 15.2.6-1 Nuclear Power Transient During Startup of an Inactive Loop
15.2.6-2 Average and Hot Channel Heat Fl ux Transients During Startup of an Inactive Loop
15.2.6-3 Core Flow During Startup of an Inactive Loop
15.2.6-4 Pressurizer Pressure Transient and Core Average Temperature Transient During Startup of an Inactive Loop 15.2.6-5 DNBR Transient During Startup of an Inactive Loop
15.2.7-1 Loss of Load With Pressurizer Spray and Power Operated Relief Valve for DNB Concern at Beginning of Life
- DNBR and Nuclear Power Versus Time 15.2.7-2 Loss of Load With Pressurizer Spray and Power Operated Relief Valve for DNB Concern at Beginning of Life - Average Core Temperature and
Pressurizer Water Volume Versus Time
15.2.7-3 Loss of Load With Pressurizer Spray and Power Operated Relief Valve for DNB Concern at End of Life - DNBR, Steam Temperature, Pressurizer
Pressure, and Nuclear Power Versus Time
15.2.7-4 Loss of Load With Pressurizer Spray and Power Operated Relief Valve for DNB Concern at End of Life - Average Core Temperature and Pressurizer
Water Volume Versus Time
15.2.7-5 Deleted in Revision 16
15.2.7-6 Deleted in Revision 16
15.2.7-7 Deleted in Revision 16
15.2.7-8 Deleted in Revision 16
15.2.7-9 Loss of Load Without Pressurizer Spray and Power Operated Relief Valves for Overpressure Concern at Beginning of Life - Reactor Power, Pressurizer Pressure, and Lower Plenum Pressure Versus Time DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxiii Revision 23 December 2016 15.2.7-10 Loss of Load Without Pressurizer Spray and Power Operated Relief Valves for Overpressure Concern at Beginning of Life - Steam Generator
Steam and Water Pressure Versus Time
15.2.7-11 Loss of Load With Pressurizer Spray and Power Operated Relief Valves for Overpressure Concern at Beginning of Life - Reactor Power, Pressurizer Pressure, and Lower Plenum Pressure Versus Time
15.2.7-12 Loss of Load With Pressurizer Spray and Power Operated Relief Valves for Overpressure Concern at Beginning of Life - Steam Generator Steam
and Water Pressure Versus Time
15.2.8A-1 Deleted in Revision 19
15.2.8-1 Loss of Normal Feedwater - RCS Temperatures and Steam Generator Mass Transients
15.2.8A-2 Deleted in Revision 19
15.2.8-2 Loss of Normal Feedwater - Pressurizer Water Volume and Pressurizer Pressure Transients
15.2.8A-3 Deleted in Revision 19
15.2.8-3 Loss of Normal Feedwater - Nuclear Power and Steam Generator Pressure Transients
15.2.9-1 Loss of Offsite Power RCS Temperatures and Steam Generator Mass Transients
15.2.9-2 Loss of Offsite Power Pressurizer Water Volume and Pressurizer Pressure Transients
15.2.9-3 Loss of Offsite Power Nuclear Power and Steam Generator Pressure Transients
15.2.10A-1 Deleted in Revision 19
15.2.10-1 Main Feedwater Regulating Valv e Malfunction - Full Power, Manual Rod Control, Nuclear Power and Average Chann el Core Heat Flux Transients
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxiv Revision 23 December 2016 15.2.10A-2 Deleted in Revision 19
15.2.10-2 Main Feedwater Regulating Valv e Malfunction - Full Power, Manual Rod Control, Pressurizer Pressure and Faulted Loop Delta-T Transients
15.2.10-3 Main Feedwater Regulating Valv e Malfunction - Full Power, Manual Rod Control, Core Average Temperature and DNBR Transients
15.2.11-1 Deleted in Revision 22
15.2.11-2 Deleted in Revision 22
15.2.11-3 Deleted in Revision 22
15.2.11-4 Deleted in Revision 22
15.2.11-5 Deleted in Revision 22
15.2.11-6 Deleted in Revision 22
15.2.11-7 Deleted in Revision 22
15.2.11-8 Deleted in Revision 22
15.2.12-1 Excessive Load Increase Without Control Action at Beginning of Life, (MTC), Minimum Feedback, T and T avg as a Function of Time
15.2.12-2 Excessive Load Increase Without Control Action at Beginning of Life, (MTC), Minimum Feedback, DNBR, Nuclear Power and Pressurizer
Pressure as a Function of Time
15.2.12-3 Excessive Load Increase Without Control Action at End of Life, (MTC), Maximum Feedback, T and T avg as a Function of Time
15.2.12-4 Excessive Load Increase Without Control Action at End of Life, (MTC), Maximum Feedback, DNBR, Nuclear Power and Pressurizer Pressure as
a Function of Time
15.2.12-5 Excessive Load Increase With Reactor Control at Beginning of Life, (MTC), Minimum Feedback, T and T avg as a Function of Time DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxv Revision 23 December 2016 15.2.12-6 Excessive Load Increase With Reactor Control at Beginning of Life, (MTC), Minimum Feedback, DNBR, Nuclear Power and Pressurizer
Pressure as a Function of Time
15.2.12-7 Excessive Load Increase With Reactor Control at End of Life, (MTC), Maximum Feedback, T and T avg as a Function of Time
15.2.12-8 Excessive Load Increase With Reactor Control at End of Life, (MTC), Maximum Feedback, DNBR, Nuclear Power and Pressurizer Pressure as
a Function of Time
15.2.13-1 Nuclear Power and DNBR Transients for Accidental Depressurization of the Reactor Coolant System
15.2.13-2 Pressurizer Pressure and Core Average Temperature Transients for Accidental Depressurization of the Reactor Coolant System
15.2.13-3 Deleted in Revision 17.
15.2.14-1 Deleted in Revision 16.
15.2.14-2 Deleted in Revision 16.
15.2.15-1 Spurious Actuation of Safety Injection System at Power DNBR Analysis -
Pressurizer Water Volume and Pressurizer Pressure Versus Time
15.2.15-2 Spurious Actuation of Safety Injection System at Power DNBR Analysis -
Nuclear Power, Steam Flow, and Core Water Temperature Versus Time
15.2.15-3 SSI Pressurizer Overfill Analysis - Typical Pressurizer Pressure Response 15.2.15-4 SSI Pressurizer Overfill Analysis -
Typical Pressurizer Liquid Volume Response 15.2.15-5 SSI Pressurizer Overfill Analysis - Typical RCS Average Temperature Response 15.3-1 Safety Injection Flow Rate for Small Break LOCA
15.3-2 RCS Depressurization 4-inch Cold Leg Break DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxvi Revision 23 December 2016 15.3-3 Core Mixture Elevation 4-inch Cold Leg Break
15.3-4 Cladding Temperature Transient 4-inch Cold Leg Break
15.3-5 Deleted in Revision 13.
15.3-6 Deleted in Revision 13.
15.3-7 Deleted in Revision 13.
15.3-8 LOCA Core Power Transient
15.3-9 RCS Depressurization 3-inch Cold Leg Break
15.3-10 Deleted in Revision 13.
15.3-11 Core Mixture Elevation 3-inch Cold Leg Break
15.3-12 Deleted in Revision 13.
15.3-13 Clad Temperature Transient 3 inch Cold Leg Break
15.3-14 Deleted in Revision 13.
15.3-14a Deleted in Revision 13.
15.3-14b Deleted in Revision 13.
15.3-14c Deleted in Revision 13.
15.3-14d Deleted in Revision 13.
15.3-14e Deleted in Revision 13.
15.3-14f Deleted in Revision 13.
15.3-15 Interchange Between Region 1 and Region 3 Assembly
15.3-16 Interchange Between Region 1 and Region 2 Assembly - Burnable Poison Rods Being Retained by the Region 2 Assembly
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxvii Revision 23 December 2016 15.3-17 Interchange Between Region 1 and Region 2 Assembly - Burnable Poison Rods Being Transferred to the Region 1 Assembly 15.3-18 Enrichment Error - A Region 2 Assembly Loaded into the Core Central Position 15.3-19 Loading a Region 2 A ssembly into a Region 1 Position Near Core Periphery
15.3-20 Deleted in Revision 3 15.3-21 Deleted in Revision 3
15.3-22 Deleted in Revision 3
15.3-23 Deleted in Revision 3
15.3-24 Deleted in Revision 3
15.3-25 Deleted in Revision 3
15.3-26 Deleted in Revision 6
15.3-27 Deleted in Revision 6
15.3-28 Deleted in Revision 6
15.3-29 Deleted in Revision 6
15.3-30 Deleted in Revision 6
15.3-31 Deleted in Revision 6
15.3-32 Deleted in Revision 6
15.3-33 Top Core Node Vapor Temperature 3-inch Cold Leg Break
15.3-34 Rod Film Coefficient 3-inch Cold Leg Break
15.3-35 Hot Spot Fluid Temperature 3 inch Cold Leg Break
15.3-36 Break Mass Flow 3-inch Cold Leg Break DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxviii Revision 23 December 2016 15.3-37 RCS Depressurization 2-inch Cold Leg Break 15.3-38 Core Mixture Elevation 2-inch Cold Leg Break
15-3-39 Cladding Temperature Transient 2-inch Cold Leg Break
15-3-40 RCS Depressurization 6-inch Cold Leg Break
15-3-41 Core Mixture Elevation 6-inch Cold Leg Break 15.3.4-1 Complete Loss of Forced Reactor Coolant Flow - All Loops Operating, All Loops Coasting Down - Flow Coastdown Versus Time
15.3.4-2 Complete Loss of Forced Reactor Coolant Flow - All Loops Operating, All Loops Coasting Down - Heat Flux Versus Time
15.3.4-3 Complete Loss of Forced Reactor Coolant Flow - All Loops Operating, All Loops Coasting Down - Nuclear Power Versus Time
15.3.4-4 Complete Loss of Forced Reactor Coolant Flow - All Loops Operating, All Loops Coasting Down - DNBR Versus Time
15.4.1-1A Unit 1 Reference Transient PCT and PCT Location
15.4.1-1B Unit 2 Limiting PCT Case and PCT Location
15.4.1-2A Unit 1 Reference Transient Vessel Side Break Flow
15.4.1-2B Unit 2 Limiting PCT Case Vessel Side Break Flow
15.4.1-3A Unit 1 Reference Transient Loop Side Break Flow
15.4.1-3B Unit 2 Limiting PCT Case Loop Side Break Flow
15.4.1-4A Unit 1 Reference Transient Broken and Intact Loop Pump Void Fraction
15.4.1-4B Unit 2 Limiting PCT Case Broken and Intact Loop Pump Void Fraction
15.4.1-5A Unit 1 Reference Transient Hot Assembly/Top of Core Vapor Flow
15.4.1-5B Unit 2 Limiting PCT Case Hot Assembly/Top of Core Vapor Flow DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxix Revision 23 December 2016 15.4.1-6A Unit 1 Reference Transient Pressurizer Pressure
15.4.1-6B Unit 2 Limiting PCT Case Pressurizer Pressure
15.4.1-7A Unit 1 Reference Transient Lower Plenum Collapsed Liquid Level
15.4.1-7B Unit 2 Limiting PCT Case Lower Plenum Collapsed Liquid Level
15.4.1-8A Unit 1 Reference Transient Vessel Water Mass 15.4-1-8B Unit 2 Limiting PCT Case Vessel Fluid Mass
15.4.1-9A Unit 1 Reference Transient Loop 1 Accumulator Flow
15.4.1-9B Unit 2 Limiting PCT Case Loop 1 Accumulator Flow
15.4.1-10A Unit 1 Reference Transient Loop 1 Safety Injection Flow
15.4.1-10B Unit 2 Limiting PCT Case Loop 1 Safety Injection Flow
15.4.1-11A Unit 1 Reference Transient Core Average Channe l Collapsed Liquid Level
15.4.1-11B Unit 2 Limiting PCT Case Core Average Ch annel Collapsed Liquid Level
15.4.1-12A Unit 1 Reference Transient Loop 1 Downcomer Collapsed Liquid Level
15.4.1-12B Unit 2 Limiting PCT Case Loop 1 Downcom er Collapsed Liquid Level
15.4.1-13A Unit 1 Total ECCS Flow (3 Lines Injecting)
15.4.1-13B Unit 2 Total ECCS Flow (3 Lines Injecting)
15.4.1-14A Unit 1 Reference Transient Pressure Transient
15.4.1-14B Unit 2 Lower Bound COCO Containment Pressure Transient
15.4.1-15A Unit 1 Axial Power Distribution Limits
15.4.1-15B Unit 2 Axial Power Distribution Limits
15.4-2 Deleted in Revision 18 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxx Revision 23 December 2016 15.4-3 Deleted in Revision 18
15.4-4 Deleted in Revision 18
15.4-5 Deleted in Revision 18
15.4-5A Deleted in Revision 12
15.4-5B Deleted in Revision 12 15.4-6 Deleted in Revision 18
15.4-7 Deleted in Revision 18
15.4-8 Deleted in Revision 18
15.4-9 Deleted in Revision 18
15.4-9A Deleted in Revision 12
15.4-9B Deleted in Revision 12
15.4-10 Deleted in Revision 18
15.4-11 Deleted in Revision 18
15.4-12 Deleted in Revision 18
15.4-13 Deleted in Revision 18
15.4-13A Deleted in Revision 12
15.4-13B Deleted in Revision 12
15.4-14 Deleted in Revision 18
15.4-15 Deleted in Revision 18
15.4-16 Deleted in Revision 12
15.4-17 Deleted in Revision 12 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxi Revision 23 December 2016 15.4-17A Deleted in Revision 12
15.4-18 Deleted in Revision 12
15.4-19 Deleted in Revision 12
15.4-20 Deleted in Revision 12
15.4-21 Deleted in Revision 12 15.4-21A Deleted in Revision 12
15.4-22 Deleted in Revision 12
15.4-23 Deleted in Revision 12
15.4-24 Deleted in Revision 12
15.4-25 Deleted in Revision 12
15.4-25A Deleted in Revision 12
15.4-26 Deleted in Revision 12
15.4-27 Deleted in Revision 12
15.4-28 Deleted in Revision 12
15.4-29 Deleted in Revision 12
15.4-29A Deleted in Revision 12
15.4-29B Deleted in Revision 12
15.4-30 Deleted in Revision 12
15.4-31 Deleted in Revision 12
15.4-32 Deleted in Revision 12
15.4-33 Deleted in Revision 12 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxii Revision 23 December 2016 15.4-33A Deleted in Revision 12
15.4-33B Deleted in Revision 12
15.4-34 Deleted in Revision 12
15.4-35 Deleted in Revision 12
15.4-36 Deleted in Revision 12 15.4-37 Deleted in Revision 12
15.4-37A Deleted in Revision 12
15.4-38 Deleted in Revision 12
15.4-39 Deleted in Revision 12
15.4-40 Deleted in Revision 12
15.4-41 Deleted in Revision 12
15.4-41A Deleted in Revision 12
15.4-42 Deleted in Revision 12
15.4-43 Deleted in Revision 12
15.4-44 Deleted in Revision 12
15.4-45 Deleted in Revision 12
15.4-45A Deleted in Revision 12
15.4-46 Deleted in Revision 12
15.4-47 Deleted in Revision 12
15.4-48 Deleted in Revision 12
15.4-49 Deleted in Revision 12 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxiii Revision 23 December 2016 15.4-49A Deleted in Revision 12
15.4-50 Deleted in Revision 12
15.4-51 Deleted in Revision 12
15.4-51A Deleted in Revision 12
15.4-52 Deleted in Revision 12 15.4-53 Deleted in Revision 12
15.4-53A Deleted in Revision 12
15.4-54 Deleted in Revision 12
15.4-55 Deleted in Revision 12
15.4-56 Deleted in Revision 12
15.4-57 Deleted in Revision 12
15.4-57A Deleted in Revision 12
15.4-58 Deleted in Revision 12
15.4-59 Deleted in Revision 12
15.4-59A Deleted in Revision 12
15.4-60 Deleted in Revision 12
15.4-61 Deleted in Revision 12
15.4-61A Deleted in Revision 12
15.4-62 Deleted in Revision 12
15.4-63 Deleted in Revision 6
15.4-64 Deleted in Revision 6 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxiv Revision 23 December 2016 15.4-65 Deleted in Revision 6
15.4-66 Deleted in Revision 6
15.4-67 Deleted in Revision 6
15.4-68 Deleted in Revision 6
15.4-69 Deleted in Revision 6 15.4-70 Deleted in Revision 6
15.4-71 Deleted in Revision 6
15.4-72 Deleted in Revision 6
15.4-73 Deleted in Revision 6
15.4-74 Deleted in Revision 6
15.4-75 Deleted in Revision 2
15.4-75a Deleted in Revision 3
15.4-75b Deleted in Revision 3
15.4-75c Deleted in Revision 3
15.4-75d Deleted in Revision 3
15.4-75e Deleted in Revision 3
15.4-75f Deleted in Revision 3
15.4-75g Deleted in Revision 3
15.4-75h Deleted in Revision 3
15.4-75i Deleted in Revision 6
15.4-75j Deleted in Revision 6 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxv Revision 23 December 2016 15.4-75k Deleted in Revision 6
15.4-75l Deleted in Revision 6
15.4-75m Deleted in Revision 6
15.4-75n Deleted in Revision 6
15.4-75o Deleted in Revision 6 15.4-75p Deleted in Revision 6
15.4-76 Deleted in Revision 7
15.4-77 Deleted in Revision 7
15.4-78 Deleted in Revision 3
15.4-79 Deleted in Revision 3
15.4-80 Deleted in Revision 3
15.4-81 Deleted in Revision 3
15.4-82 Deleted in Revision 3
15.4-83 Deleted in Revision 3
15.4-84 Deleted in Revision 3
15.4-85 Deleted in Revision 3
15.4-86 Deleted in Revision 3
15.4-87 Deleted in Revision 3
15.4-88 Deleted in Revision 3
15.4-89 Deleted in Revision 6
15.4-90 Deleted in Revision 6 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxvi Revision 23 December 2016 15.4-91 Deleted in Revision 6
15.4-92 Deleted in Revision 6
15.4-93 Deleted in Revision 6
15.4-94 Deleted in Revision 6
15.4-95 Deleted in Revision 6 15.4-96 Deleted in Revision 6
15.4-97 Deleted in Revision 6
15.4-98 Deleted in Revision 6
15.4-99 Deleted in Revision 16
15.4-100 Deleted in Revision 16
15.4-101 Deleted in Revision 16
15.4-102 Deleted in Revision 16
15.4-103 Deleted in Revision 16
15.4-104 Deleted In Revision 16
15.4-105 Deleted in Revision 16
15.4-106 Deleted in Revision 16
15.4-107 Deleted in Revision 16
15.4-108 Deleted in Revision 16
15.4-109 Deleted in Revision 16
15.4.2A-1 Deleted in Revision 19
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxvii Revision 23 December 2016 15.4.2-1 Rupture of a Main Steam Line - Variation of Reactivity with Power at Constant Core Average Temperature 15.4.2A-2 Deleted in Revision 19
15.4.2-2 Rupture of a Main Steam Line - Variation of K eff with Core Average Temperature
15.4.2A-3 Deleted in Revision 19
15.4.2-3 Rupture of a Main Steam Line - Safety Injection Curve
15.4.2A-4A Deleted in Revision 19
15.4.2A-4B Deleted in Revision 19
15.4.2A-4C Deleted in Revision 19
15.4.2A-4D Deleted in Revision 19
15.4.2-4 Rupture of a Main Steam Line with Offsite Power Available - Core Heat Flux and Steam Flow Transients
15.4.2A-5 Deleted in Revision 19
15.4.2-5 Rupture of a Main Steam Line with Offsite Power Available - Loop Average Temperature and reactor Coolant Pressure Transients
15.4.2-6 Rupture of a Main Steam Line with Offsite Power Available - Reactivity and Core Boron Transients
15.4.2A-7 Deleted in Revision 19
15.4.2-7 Rupture of a Main Steam Line without Offsite Power Available - Core Heat Flux and Steam Flow Transients
15.4.2A-8 Deleted in Revision 19
15.4.2-8 Rupture of a Main Steam Line without Offsite Power Available - Loop Average Temperature and Reactor Coolant Pressure Transients
15.4.2A-9 Deleted in Revision 19 DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxviii Revision 23 December 2016 15.4.2-9 Rupture of a Main Steam Line without Offsite Power Available - Reactivity and Core Boron Transients
15.4.2A-10 Deleted in Revision 19
15.4.2-10 Main Feedline Rupture with Offsite Power Available - Nuclear Power and Core Heat Flux Transients
15.4.2A-11 Deleted in Revision 19 15.4.2-11 Main Feedline Rupture with Offsite Power Available - Pressurizer Pressure and Core Water Volume Transients
15.4.2A-12 Deleted in Revision 19
15.4.2-12 Main Feedline Rupture with Offsite Power Available - Reactor Coolant Temperature Transients for the Faulted and Intact Loops
15.4.2A-13 Deleted in Revision 19
15.4.2-13 Main Feedline Rupture with Offsite Power Available - Steam Generator Pressure and Total mass Transients
15.4.2A-14 Deleted in Revision 19
15.4.2-14 Main Feedline Rupture without Offsite Power Available - Nuclear Power and Core Heat Flux Transients
15.4.2A-15 Deleted in Revision 19
15.4.2-15 Main Feedline Rupture without Offsite Power Available - Pressurizer Pressure and Water Volume Transients
15.4.2A-16 Deleted in Revision 19
15.4.2-16 Main Feedline Rupture without Offsite Power Available - Reactor Coolant Temperature Transients for the Faulted and Intact Loops
15.4.2A-17 Deleted in Revision 19
DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xxxix Revision 23 December 2016 15.4.2-17 Main Feedline Rupture without Offsite Power Available - Steam Generator Pressure and Total Mass Transients 15.4.2A-18 Deleted in Revision 19 15.4.2-18 Main Steam Line Rupture at Full Power, 0.49 ft 2 Break - Nuclear Power and Core Heat Flux Transients 15.4.2-19 Main Steam Line Rupture at Full Power, 0.49 ft 2 Break - Pressurizer Pressure and Water Volume Transients
15.4.2-20 Main Steam Line Rupture at Full Power, 0.49 ft 2 Break - Reactor Vessel Inlet Temperature and Loop Average Temperature Transients
15.4.2-21 Main Steam Line Rupture at Full Power, 0.49 ft 2 Break - Total Steam Flow and Steam Pressure Transients
15.4.2-22 Main Feedline Rupture for Pressurizer Filling (Unblock Pressurizer PORV)- Pressurizer Pressure and Water Volume Transients
15.4.2-23 Main Feedline Rupture for Pressurizer Filling (Unblock Pressurizer PORV) - PSV Relief Flow Rate and Enthalpy Transients
15.4.2-24 Main Feedline Rupture for Pressurizer Filling (Isolate Charging and Stop RCP Seal Injection Flow) - Pressurizer Pressure and Water Volume
15.4.2-25 Main Feedline Rupture for Pressurizer Filling (Isolate Charging and Stop RCP Seal Injection Flow) - PORV and PSV Relief Flow Rate Transients
15.4.2-26 Main Feedline Rupture for Pressurizer Filling (Isolate Charging and Stop RCP Seal Injection Flow) - PORV/PSV Rel ief Flow Enthalpy and Total Number of PORV Cycles Transients
15.4.2-27 Main Feedline Rupture for Pressurizer Filling (Isolate Charging and Stop RCP Seal Injection Flow) - Cold Leg Injection Flow Rate Transient
15.4.3A-1 Deleted in Revision 19
15.4.3-1 Deleted in Revision 20
15.4.3-1A Pressurizer Level - SGTR MTO Analysis DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xl Revision 23 December 2016 15.4.3-1B Pressurizer Level - SGTR Dose Analysis
15.4.3A-2 Deleted in Revision 19
15.4.3-2 Deleted in Revision 20
15.4.3-2A Pressurizer Pressure - SGTR MTO Analysis
15.4.3-2B Pressurizer Pressure - SGTR Dose Input Analysis 15.4.3A-3 Deleted in Revision 19
15.4.3-3 Deleted in Revision 20
15.4.3-3A Secondary Pressure - SGTR MTO Analysis
15.4.3-3B Secondary Pressure - SGTR Dose Input Analysis
15.4.3A-4 Deleted in Revision 19
15.4.3-4 Deleted in Revision 20
15.4.3-4A Intact Loop Hot and Cold Leg RCS Temperatures - SGTR MTO Analysis
15.4.3-4B Intact Loop Hot and Cold Leg RCS Temperatures - SGTR Dose Input Analysis 15.4.3A-5 Deleted in Revision 19
15.4.3-5 Deleted in Revision 20
15.4.3-5B Ruptured Loop Hot and Cold Leg RCS Temperatures - SGTR Dose Input Analysis 15.4.3A-6 Deleted in Revision 19
15.4.3-6 Deleted in Revision 20
15.4.3-6A Primary to Secondary Break Flow Rate - SGTR MTO Analysis
15.4.3-6B Primary to Secondary Break Flow Rate - SGTR Dose Input Analysis DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xli Revision 23 December 2016 15.4.3A-7 Deleted in Revision 19
15.4.3-7A Ruptured SG Water Volume - SGTR Margin-to-Overfill Analysis
15.4.3-7B Ruptured SG Water Volume - SGTR Dose Input Analysis
15.4.3A-8 Deleted in Revision 19
15.4.3-8 Deleted in Revision 20 15.4.3-8A Ruptured Steam Generator Water Mass - SGTR MTO Analysis 15.4.3-8B Ruptured Steam Generator Water Mass - SGTR Dose Input Analysis
15.4.3A-9 Deleted in Revision 19
15.4.3-9 Ruptured SG Mass Release Rate to the Atmosphere - SGTR Dose Input Analysis 15.4.3A-10 Deleted in Revision 19
15.4.3-10 Intact SGs Mass Release Rate to the Atmosphere - SGTR Dose Input Analysis 15.4.3A-11 Deleted in Revision 19
15.4.3-11 Total Flashed Break Flow - SGTR Dose Input Analysis
15.4.4-1 Single Reactor Coolant Pump Locked Rotor - Maximum RCS Pressure Versus Time
15.4.4-2 Single Reactor Coolant Pump Locked Rotor - Clad Average Temperature Versus Time
15.4.4-3 Single Reactor Coolant Pump Locked Rotor - Core Flow Versus Time
15.4.4-4 Single Reactor Coolant Pump Locked Rotor - Hot Channel Heat Flux Versus Time
15.4.4-5 Single Reactor Coolant Pump Locked Rotor - Nuclear Power Versus Time
15.4.6-1 Nuclear Power Transient, BOL HZP, Rod Ejection Accident DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xlii Revision 23 December 2016 15.4.6-2 Hot Spot Fuel and Clad Temperature Versus Time BOL, HZP, Rod Ejection Accident
15.4.6-3 Nuclear Power Transient, EOL, HFP, Rod Ejection Accident
15.4.6-4 Hot Spot Fuel and Clad Temperatures Versus Time, EOL, HZP, Rod Ejection Accident
15.5-1 Ratio of Short-Term Release Concentration to Continuous Release Concentration Versus Release Duration
15.5-2 Thyroid Dose at 800 Meters Versus Weight of Steam Dumped to Atmosphere (Design Basis Case Assumptions)
15.5-3 Thyroid Dose at 10,000 Meters Versus Weight of Steam Dumped to Atmosphere (Design Basis Case Assumptions)
15.5-4 Thyroid Dose at 10,000 Meters Versus Weight of Steam Dumped to Atmosphere (Expected Case Assumptions)
15.5-5 Thyroid Dose at 800 Meters Versus Weight of Steam Dumped to Atmosphere (Expected Case Assumptions) 15.5-6 Thyroid Exposures for 15% Nonremovable Iodine (Historical)
15.5-7 DBA Two-hour 800-meter Thyroid Exposures Versus Spray Removal Constant and Percent Nonremovable Iodine (Historical)
15.5-8 DBA Thirty-hour 800-meter Thyroid Exposures Versus Spray Removal Constant and Percent Nonremovable Iodine (Historical) 15.5-9 Containment Recirculation Sump Activity Pathway to the Atmosphere for Small Leak Case 15.5-10 Containment Recirculation Sump Activity Pathway to the Atmosphere for Large Leak Case
15.5-11 Equilibrium Elemental Iodine Parti tion and Decontamination Factors for the Expected Case - Large Circulation L oop Leakage in the Auxiliary
Building DCPP UNITS 1 &
2 FSAR UPDATE CHAPTER 15 FIGURES (Continued)
Figure Title xliii Revision 23 December 2016 15.5-12 Equilibrium Elemental Iodine Parti tion and Decontamination Factors for the DBA Case - Large Circulation Loop Leak age in the Auxil iary Building 15.5-13 Deleted in Revision 7
15.5-14 Potential Radiation Exposures as a Result of Accidents Involving Failure of Fuel Cladding (Design Basis Case Assumptions)
15.5-15 Potential Radiation Exposures as a Result of Accidents Involving Failure of Fuel Cladding (Expected Case Assumptions)
15.5-16 Incremental Long-term Doses from Accidents Involving Failure of Fuel Cladding (Historical)
15.5-17 Deleted in Revision 16
15.5-18 Deleted in Revision 16
15.5-19 Deleted in Revision 19
15.5-20 Deleted in Revision 16
15.5-21 Deleted in Revision 16
15.5-22 Deleted in Revision 16
DCPP UNITS 1 &
2 FSAR UPDATE 15-1 Revision 22 May 2015 Chapter 15 ACCIDENT ANALYSES Since 1970, the ANS classification of plant conditions has been used to divide plant conditions into four categories in accordance with anticipated frequency of occurrence
and potential radiological consequences to t he public. The four categories are as follows: (1) Condition I: Normal Operation and Operational Transients (Initial Conditions)
(2) Condition II: Faults of Moderate Frequency (3) Condition III: Infrequent Faults (4) Condition IV: Limiting Faults The basic principle applied in relating design requirements to each of the conditions is that the most frequent occurrences must yield little or no 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
safety features functioning is assumed, to the extent allowed by considerations such as
the single failure criterion, in fulfilling this principle.
In the evaluation of the radiological consequ ences associated with initiation of a spectrum of accident conditions, numerous assumptions must be postulated. In many
instances these assumptions are a product of extremely conservative judgments. This
is due to the fact that many physical pheno mena, in particular fission product transport under accident conditions, are not understood to the extent that accurate predictions
can be made. Therefore, the set of assumptions postulated would predominantly
determine the accident classification.
The specific accident sequences analyzed in this chapter include those required by Revision 1 of Regulatory Guide 1.70, Standard Format and Content of Safety Analysis
Reports for Nuclear Power Plants, and others considered significant for the Diablo
Canyon Power Plant (DCPP). Refer to UFSAR Table 15.0-1 for a comparison between Regulatory Guide 1.70, Revision 1, October 1972, Table 15-1 and the corresponding section(s) where the conditions are discussed. Because the DCPP design differs from other plants, some of the representative types of events identified in Table 15-1 of Regulatory Guide 1.70, Revision 1, October 1972 are not applicable to this plant. In addition, some events are analyzed or discussed in separate UFSAR chapters. The location of the analysis for each event or reason the event is not applicable to DCPP is provided in UFSAR Table 15.0-1.
DCPP UNITS 1 &
2 FSAR UPDATE 15-2 Revision 22 May 2015 This section of the FSAR describes the acceptance criteria, input assumptions, analysis techniques, equipment performance, and analysis results of the required accident analysis but does not include details on the set points, capacity or capabilities of mitigating equipment or operational limitations that determine the initial conditions for each analysis. For details of required reactor operational limitations and of the performance capabilities of the emergency equipment not covered in Chapter 15, refer to the following chapters of the UFSAR:
- Reactor coefficients, power distribution, reactivity controls, Refer to Chapter 4 - Reactor coolant flow, Refer to Chapter 5 - ECCS, Auxiliary feed water, Containment systems, Refer to Chapter 6 - Reactor trips and permissives, ESF actuation, Refer to Chapter 7 - Boration capabilities, Refer to Chapter 9 Additionally the availability, testing and perfor mance criteria of the operational limits and mitigating systems are administratively controlled by the plant Technical Specifications described in Chapter 16 and Appendix A of the Diablo Canyon Power Plant Unit 1 and Unit 2 Operating Licenses.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-1 Revision 23 December 2016 15.1 CONDITION I - NORMAL OPERATION AND OPERATIONAL TRANSIENTS (INITIAL CONDITIONS) 15.
1.1 INTRODUCTION
Condition I occurrences are those that 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 wou ld require either automatic or manual protective action. Since 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 is generally based on a conservative set of initial conditions corresponding to the most adverse set
of conditions that can occur during Condition I operation.
Typical Condition I events are shown below:
(1) Steady state and shutdown operations Mode 1 - Power operation (greater than 5 percent of rated thermal power)
Mode 2 - Startup (k eff 0.99, less than or equal to 5 percent of rated thermal power)
Mode 3 - Hot standby (k eff less than 0.99, T avg greater than or equal to 350°F) Mode 4 - Hot shutdown (subcritical, residual heat removal system in operation, k eff less than 0.99, 200
°F less than T avg less than 350°F) Mode 5 - Cold shutdown (subcritical, residual heat removal system in operation, k eff less than 0.99, T avg less than or equal to 200
°F) Mode 6 - Refueling (k eff less than or equal to 0.95, T avg less than or equal to 140°F) (2) Operation with permissible deviations Various deviations that may occur during continued operation as permitted
by the plant Technical Specifications (Reference 1) must be considered in
conjunction with other operational modes. These include:
(a) Operation with components or systems out of service (b) Leakage from fuel with cladding defects DCPP UNITS 1 &
2 FSAR UPDATE 15.1-2 Revision 23 December 2016 (c) Activity in the reactor coolant
- 1. Fission products
- 2. Corrosion products
- 3. Tritium (d) Operation with steam generator leaks up to the maximum allowed by the Technical Specifications (3) Normal Operational transients Normal design transients which do not result in a reactor trip are listed below. Refer to Section 5.2.2.1.5.1 for additional details on these transients.
(a) Plant heatup and cooldown (b) Step load changes (up to plus or minus 10 percent between 15 percent load and full load)
(c) Ramp load changes (up to 5 percent per minute between 15 percent load and full load)
(d) Turbine load reduction up to and including a 50 percent load rejection from full power (e) Steady state fluctuations of the reactor coolant average temperature, for purposes of design, is assumed to increase or
decrease at a maximum rate of 6°F in 1 minute.
15.1.2 COMPUTER CODES UTILIZED
Summaries of some of the principal computer codes used in transient analyses are given below. Other codes, in particular, ver y specialized codes in which the modeling has been developed to simulate one given accident, such as the NOTRUMP code used
in the analysis of the RCS small pi pe rupture (Section 15.3.1), and which consequently have a direct bearing on the analysis of the accident itself, are summarized in their
respective accident analyses sections. The codes used in the analyses of each
transient event are listed in Table 15.1-4.
15.1.2.1 FACTRAN FACTRAN calculates the transient temperature distribution in a cross section of a
metalclad UO 2 fuel rod (refer to Figure 15.1-8) and the transient heat flux at the surface DCPP UNITS 1 &
2 FSAR UPDATE 15.1-3 Revision 23 December 2016 of the cladding using as input the nuclear power and the time-dependent coolant parameters (pressure, flow, temperature, and density).
The code uses a fuel model that exh ibits the following features simultaneously:
(1) A sufficiently large number of finite difference radial space increments to handle fast transients such as rod ejection accidents (2) Material properties that are functions of temperature and a sophisticated fuel-to-cladding gap heat transfer calculation (3) The necessary calculations to handle post-DNB transients: film boiling heat transfer correlations, zirconium-water reaction, and partial melting of
the materials
The gap heat transfer coefficient is calculated according to an elastic pellet model. The
thermal expansion of the pellet is calculated as the sum of the radial (one-dimensional) expansions of the rings. Each ring is assumed to expand freely. The cladding diameter
is calculated based on thermal expansion and internal and external pressures.
If the outside radius of the expanded pellet is smaller than the inside radius of the expanded cladding, there is no fu el-cladding contact and the gap conductance is calculated on the basis of the thermal conductivity of the gas contained in the gap. If
the pellet outside radius so calculated is larg er than the cladding inside radius (negative gap), the pellet and the cladding are pictured as exerting upon each other a pressure
sufficient to reduce the gap to zero by elastic deformation of both. This contact pressure determines the heat transfer coefficient.
FACTRAN is further discussed in the licensing topical report, Section 1.6.1, Item 44.
15.1.2.2 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 modeling the reactor core and vessel, hot and cold leg piping, steam generator (tube
and shell-sides), pressurizer, and reactor coo lant pumps, with up to four reactor coolant loops. 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 neutron flux, overpower and overtemperature reactor coolant 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 safety injection system (SIS), including the accumulators, is also modeled.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-4 Revision 23 December 2016 LOFTRAN is a versatile program that 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 DNB based on the input from the core limits illustrated in
Figure 15.1-1. The core limits represent the minimum value of DNBR as calculated for
a typical or thimble cell. LOFTRAN is further discussed in the licensing topical report, Section 1.6.1, Item 47.
15.1.2.3 PHOENIX- P The PHOENIX-P computer code is a two-dimensional, multi-group, transport based
lattice code and is capable of providing all necessary data for PWR analysis. Being a
dimensional lattice code, PHOENIX-P does not rely on pre-determined spatial/spectral
interaction assumptions for a heterogeneous fuel lattice. The PHOENIX-P computer
code is approved by the NRC as the lattice code for generating macroscopic and
microscopic few group cross sections for PWR analysis.
The PHOENIX-P computer code is described in more detail in Section 4.3.3.10.2 and is further discussed in the licensing topical report, Section 1.6.1, Item 60.
15.1.2.4 ANC With the advent of VANTAGE 5 fuel and axial features such as axial blankets and part
length burnable absorbers, the three dimensiona l nodal codes ANC (Advanced Nodal Code) has replaced the previous two group X-Y TURTLE code. The three dimensional
nature of the nodal codes provides both the radial and axial po wer distributions, and also determines the critical boron concentrations and power distributions. The moderator coefficient is evaluated by varying the inlet temperature in the same calculations used for power distri bution and reactivity predictions.
Axial calculations are used to determine differential control rod worth curves (reactivity
versus rod insertion) and axial power shapes during steady state and transient xenon conditions. Group constants are obtained from three-dimensional nodal calculations homogenized by flux volume weighting.
The ANC computer code is described in more detail in Section 4.3.3.10.3 and is further discussed in the licensing topical reports, Section 1.6.1, Items 60 and 61.
15.1.2.5 TWINKLE The TWINKLE program is a multidimensional 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 multiregion fuel-cladding-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 DCPP UNITS 1 &
2 FSAR UPDATE 15.1-5 Revision 23 December 2016 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 provide channelwise
power, axial offset, enthalpy, volumetric surge, pointwise power, fuel temperatures, and so on.
The TWINKLE code is used to predict the kinetic behavior of a reactor for transients that
cause a major perturbation in the spatial neutron flux distribution.
TWINKLE is further described in the licensing topical report, Section 1.6.1, Item 50.
15.1.2.6 THINC The Steady state and transient analysis using the THINC code (THINC-I, THINC-III and
THINC-IV) is described in Section 4.4.3. THINC is further described in the licensing
topical reports, Section 1.6.1, Item 28.
15.1.2.7 RETRAN-02 The Electric Power Research Institute (EPRI) RETRAN-02 program is used to perform
the best-estimate thermal-hydraulic analysis of operational and accident transients for
light water reactor systems. The program is constructed with a highly flexible modeling technique that provides the RETRAN-02 program the capability to model the actual
performance of the plant systems and equipment.
The main features of the RETRAN-02 program are:
(1) A one-dimensional, homogeneou s equilibrium mixture thermal-hydraulic model for the reactor cooling system (2) A point neutron kinetics model for the reactor core (3) Special auxiliary or component models (such as non-equilibrium pressurizer temperature transport delay)
(4) Control system models (5) A consistent steady st ate initialization technique The RETRAN-02 program is further discussed in Reference 21.
15.1.2.8 RETRAN-02W The RETRAN-02W program is the Westinghouse version of the RETRAN-02 program.
RETRAN-02W is used to determine plant transient response to selected accidents, as
described in Sections 15.2 and 15.4.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-6 Revision 23 December 2016 RETRAN-02W is further described in the licensing topical report, Section 1.6.1, Item 58.
15.1.2.9 NOTRUMP The NOTRUMP computer code is a state-of-the-art, one-dimensional general network code consisting of a number of advanced features. Among these features is the calculation of thermal nonequil ibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter current flow limitations, mixture level tracking logic in
multiple-stacked fluid nodes, and regime-dependent heat transfer correlations.
Additional features of the code are condens ation heat transfer model applied in the steam generator region, loop seal model, core reflux model, flow regime mapping, etc.
NOTRUMP is used to model the thermal-hydraulic behavior of the system and thereby
obtain time-dependent values of various core region parameters, such as system
pressure, temperature, fluid levels and flow rates, etc.
Small-Break LOCA (SBLOCA) analysis performed using the NOTRUMP code is further
described in Section 15.3 and in the licensing topical reports, Section 1.6.1, Items 63
and 64.
15.1.2.10 SBLOCTA (LOCTA-IV)
The NOTRUMP topical report WCAP-10054-P-A makes reference to the LOCTA-IV
code (WCAP-8301) and provides modifications to the LOCTA-IV code for use in small
break LOCA analyses (i.e., Small Break LOCTA). Further modifications for an annular
fuel pellet model were submitted and approved by the NRC in WCAP-14710-P-A, which
states, the revised model has been installed in the SBLOCTA code, which is one of a series of codes descended from the original LOCTA-IV code, and is specific to
analyzing small-break LOCA transients. So, SBLOCTA is the actual computer code
name, with base references of WCAP-8301 and WCAP-10054-P-A.
Small-Break LOCA analysis performed using the LOCTA-IV code is further described in
Section 15.3 and listed as Reference 4 in that section.
15.1.2.11 WCOBRA/TRAC The thermal-hydraulic computer code (WCOBRA/TRAC, Version Mod 7A, Revision 1) that was reviewed and approved for the calculation of fluid and thermal conditions in the
PWR during a large break LOCA in WCAP-12945-P-A, Volumes I through V is
described in Section 15.4.1.3 and in the licensing topical report, Section 1.6.1, Item 62.
15.1.2.12 HOTSPOT The use of HOTSPOT along with WCOBRA/TRAC to examine Unit 2 uncertainty using the ASTRUM methodology is discussed in Section 15.4.1.7B.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-7 Revision 23 December 2016 15.1.2.13 MONTECF Unit 2 uncertainty evaluation calculations usi ng the ASTRUM methodology was performed by applying a direct, random Monte Carlo sampling to generate the input for
the WCOBRA/TRAC and HOTSPOT computer codes as discussed in Section 15.4.1.7B.
15.1.2.14 COCO Containment pressure is calculated using the COCO code (WCAP-8327 and WCAP-
8326) as discussed in Section 15.4.1.3 and listed as Reference 61 in that section.
15.1.3 OPTIMIZATION OF CONTROL SYSTEMS Prior to initial startup, a setpoint study (Reference 2) was performed in order to simulate
performance of the reactor control and protection systems. Emphasis was placed on the development of a control system that will automatically maintain prescribed
conditions in the plant even under the most 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 was
determined. In areas where the resultant setpoints were different, compromises based on the optimum overall performance were made and verified. A consistent set of control
system parameters was derived satisfying plant operational requirements throughout
the core life and for power levels between 15 and 100 percent. The study contained an
analysis of the following control systems:
rod cluster assembly control, steam dump, steam generator level, pressurizer pressure, and pressurizer level.
Since initial startup, setpoints and control system components have been maintained to optimize performance. Plant operability margin-to-trip analyses are performed on the
NSSS control systems for DCPP Units 1 and 2. The purpose of these analyses is to
demonstrate that the margin to relevant reactor trip and Engineered Safety Features
Actuation System (ESFAS) setpoints is adequate. The NSSS control systems setpoints
and time constants are analyzed to provide stable plant response during and following
the operational (Condition I) transients:
- 50 percent load rejection from 100 percent power
- 10 percent step-load decrease from 100 percent power
- 10 percent step-load increase from 90 percent power
- Turbine trip without reactor trip from permissive P-9 setpoint When changes are made, the accident analyses are reviewed and revised as necessary. The impact of maintaining pressurizer level between 22% and 35% during a it was determined that there is no adverse impact on any accident analyses (Reference 31). The impact of maintaining pressurizer level greater than or equal to DCPP UNITS 1 &
2 FSAR UPDATE 15.1-8 Revision 23 December 2016 22 percent and less than or equal to 90 percent in Modes 3, 4, and 5 has been evaluated as acceptable because there is no adverse impact on any accident analyses (References 28 and 29).
The analysis for the 50 percent load reduction (References 33 and 35) shows that the DCPP control system is capable of controlling the RSG water level so that a reactor trip on steam generator low-low level or turbine trip / feedwater isolation on steam generator high-high level does not occur. Specific analysis results show that the steam generator level is maintained within +/-20 percent of the nominal setpoint and all control system responses are smooth and have no sustained oscillations or divergence. To ensure that a load reduction transient presents no hazard to the integrity of the RCS or the main steam system, the Condition II analysis presented in Section 15.2.7 continues to assume a total loss of external electrical load without an immediate reactor trip.
15.1.4 INITIAL POWER CONDITIONS ASSUMED IN ACCIDENT ANALYSES Reactor power-related initial conditions assumed in the accident analyses presented in
this chapter are described in this section.
15.1.4.1 Power Rating Table 15.1-1 lists the principal power rating values that are assumed in analyses
performed in this section. Two ratings are given:
(1) The rated thermal power (RTP) output. The RTP is the total reactor core heat transfer rate to the reactor c oolant of 3411 MWt for each unit.
(2) The nuclear steam supply system (NSSS) thermal power output. This power output includes the RTP plus the thermal power generated by the
reactor coolant pumps.
(3) The engineered safety features (ESF) design rating. The Westinghouse-supplied ESFs are designed for a thermal power higher than the NSSS value in order not to preclude realization of future potential power
capability. This higher thermal power value is designated as the ESF
design rating.
Where initial power operating conditions ar e assumed in accident analyses, the NSSS or core rated thermal power output (plus allowance for errors in steady state power
determination for some accidents) is assumed. Where demonstration of the adequacy
of the ESF is concerned, the ESF design rating plus allowance for error is assumed.
The thermal power values for each transient analyzed are given in Table 15.1-4.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-9 Revision 23 December 2016 15.1.4.2 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 Reference 3. This
procedure is known as the "Improved Thermal Design Procedure" (ITDP) and these
accidents utilize the WRB-1 and WRB-2 DNB correlations (References 4 and 5). ITDP
allowances may be more restrictive than non-ITDP allowances. The initial conditions for
other key parameters are selected in such a manner to maximize the impact on DNBR.
Minimum measured flow is used in all ITDP transients. The allowances on power, temperature, pressure, and flow that were evaluated for their effect on the ITDP analyses for a 24-month fuel cycle are reported in Reference 22. These allowances are
conservatively applicable for shorter fuel cycle lengths.
For accident evaluations that are not DNB li mited, or for which the Improved Thermal Design Procedure is not employed, the initial conditions are obtained by adding maximum steady state errors to rated values.
The following steady state errors are considered:
(1) Core power Plus or minus 2 percent allowance calorimetric error (2) Average RCS Plus or minus 4.7
°F allowance for deadband and measurement error temperature
(3) Pressurizer pressure Plus or minus 38 psi or plus or minus 60 psi allowance for steady state fluctuations and measurement error (see Note)
Note: Pressurizer pressure uncertainty is plus or minus 38 psi in analyses performed prior to 1993; however, NSAL 92-005 (Reference 17) indicates plus or minus 60 psi is a conservative value for future analyses.
Reference 18 evaluates the acceptability of existing analyses, which use plus or minus 38 psi.
For some accident evaluations, an additiona l allowance has been conservatively added to the measurement error for the average RCS temperatures to account for steam generator fouling.
DCPP Units 1 and 2 are expected to operate at a Reactor Coolant System vessel
average temperature (Tavg) over a range from 565
ºF to 577.3/577.6 ºF (Unit 1/Unit 2).
15.1.4.3 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 DCPP UNITS 1 &
2 FSAR UPDATE 15.1-10 Revision 23 December 2016 distribution through the placement of fuel assemblies, control rods, and by operation instructions. The power distribution may be characterized by the radial peaking factor FH and the total peaking factor F
- q. The peaking factor limits are given in the Technical Specifications.
For transients that may be DNB-limited, the radia l 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 15.1-1. 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 4.4.3.13.
For transients that may be overpower-limited, the total peaking factor F q is of importance. The value of F q may increase with decreasing power level so that the full power hot spot heat flux is not exceeded, i.e., F q x Power = design hot spot heat flux.
All transients that may be overpower-limited are assumed to begin with a value of F q consistent with the initial power level as defin ed in the Technical Specifications.
The value of peak kW/ft can be directly related to fuel temperature as illustrated in
Figures 4.4-1 and 4.4-2. For transients that are slow with respect to the fuel rod thermal time constant (approximately 5 seconds), the fuel temperatures are illustrated in
Figures 4.4-1 and 4.4-2. For transients that are fast with respect to the fuel rod thermal time constant, (for example, rod ejection), a detailed heat transfer calculation is made.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-11 Revision 23 December 2016 15.1.5 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 mechanism to release the rod cluster control assemblies (RCCAs), which then fall
by gravity into the core. There a re 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 15.1-2. Reference is made in that table to the overtemperature and overpower T trip shown in Figure 15.1-1. This figure presents the allowable reactor coolant loop average temperature and T for the design flow and the NSSS Design Thermal Power distribution 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 a 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 safety analysis limit values (1.68 and 1.71 for V-5 thimble cell and
typical cells, respectively) for analyses using the ITDP.
All points below and to the left
of a DNB line for a given pressure have a DNBR greater than the limit values. 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 current fuel cycles for the DCPP Unit 1 and Unit 2 only use the Vantage 5 (V-5) fuel
assembly type. However, the safety analyses performed in support of the transition to
Vantage-5 fuel also considered the presence of the Standard type fuel assemblies. The DNBR values and transient results presented in the UFSAR continue to reflect the
Standard limits, since they are limiting with respect to DNB margin in comparison to the
Vantage-5 limits. Analyses performed subsequent to the transition to a full Vantage-5
core reflect only the Vantage-5 limits as described in Sections 15.2, 15.3, and 15.5.
The area of permissible operation (power, pressure and temperature) is bounded by the
combination of reactor trips: high pressurizer pressure (fixed setpoint); low pressurizer pressure (fixed setpoint); overpower and overtemperature T (variable setpoints); and by a line defining conditions at which the steam generator safety valves open.
The limit values, which were used as the DNBR limits for all accidents analyzed with the
ITDP are conservative compared to the actual desig n DNBR values required to meet the DNB design basis.
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.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-12 Revision 23 December 2016 During startup tests, it is demonstrated that actual instrument errors and time delays are equal to or less than the assumed values.
Accident analyses that assume the steam generator low-low water level to initiate
protection functions may be affected by the trip time delay (TTD) (Reference 19) that
was developed to reduce the incidence of unnecessary feedwater related reactor trips.
Refer to Section 7.2.2.1.5 for a discussion about the low-low steam generator water
level trip, including the TTD.
15.1.6 CALORIMETRIC ERRORS - POWER RANGE NEUTRON FLUX 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 these measurements with accuracy
tolerances much tighter than those that would be required to control feedwater flow.
15.1.7 ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS The negative reactivity insertion following a reactor trip is a function of the acceleration
of the RCCA and the variation in rod worth as a function of rod position.
With respect to accident analyses, the c ritical parameter is the time of insertion up to the dashpot entry or approximately 85 percent of the rod cluster travel. For accident analyses, the insertion time to dashpot entry is conservatively taken as 2.7 seconds.
The RCCA position versus time assumed in accident analyses is shown in
Figure 15.1-2.
Figure 15.1-3 shows the fraction of total negative reactivity insertion for a core where
the axial distribution is skewed to the lower region of the core. This curve is used as
input to all point kinetics core models used in transient analyses.
There is inherent conservatism in the use of this curve in that it is based on a skewed
axial power distribution that would ex ist relatively infrequently. For cases other than those associated with xenon oscillations, significant negative reactivity would have been
inserted due to the more favorable axial p ower distribution existing prior to trip.
The normalized RCCA negative reactivity insertion versus time is shown in
Figure 15.1-4. The curve shown in this figure was obtained from Figures 15.1-2 and 15.1-3. A total negative reactivity insertion following a trip of 4 percent 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 Tables 4.3-2 and 4.3-3.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-13 Revision 23 December 2016 The normalized RCCA negative reactivity insertion versus time after trip curve for an
axial power distribution skewed to the bottom (Figure 15.1-4) is used in transient analyses.
Where special analyses require the use of three-dimensional or axial one-dimensional core models, the negative reactivity insertion resulting from reactor trip is calculated
directly by the reactor kinetic code and is not separable from other reactivity feedback effects. In this case, the RCCA position versus time of Figure 15.1-2 is used as code input.
15.1.8 REACTIVITY COEFFICIENTS The transient response of the reactor coolant 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.
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 ruptures in the RCS, do not depend on reactivity feedback
effects. The values used are given in Table 15.1-4; reference is made in that table to
Figure 15.1-5 that shows the upper and lower 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 is discussed on an event-by-event basis.
15.1.9 FISSION PRODUCT INVENTORIES The fission product inventories existing in the core and fuel rod gaps are described in
Section 15.5.3. The description of the models used for calculating fuel gap activities is
included in Section 15.5.3.
15.1.10 RESIDUAL DECA Y HEAT Residual heat in a subcritical core consists of:
(1) Fission product decay energy (2) Decay of neutron capture products (3) Residual fissions due to the effect of delayed neutrons
These constituents are discussed separately in the following paragraphs.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-14 Revision 23 December 2016 15.1.10.1 Fission Product Decay The heat generation rates from radioactive decay of fission products that have been
assumed in the small break LOCA (SBLOCA) accident analyses are equal to 1.2 times
the values for infinite operating time in the 1971 Draft ANS-5 Standard. (Reference 30)
The decay heat curve used for the Best Estimate large break LOCA (LBLOCA) analysis
is based on the 1979 ANS decay heat curve as described in Section 8 of Reference 23.
This curve with the 20 percent factor included is shown in Figure 15.1-6.The 1979 ANS
decay heat curve (Reference 11) is used for the non-LOCA analyses. Figure 15.1-7
presents this curve as a function of time after shutdown.
15.1.10.2 Decay of U-238 Capture Products Betas and gammas from the decay of U-239 (23.5-minute half-life) and Np-239
(2.35-day half-life) contribute significantly to the heat generation after shutdown. The
cross sections for production of these isotopes and their decay schemes are relatively
well known. For long irradiation times their contribution can be written as:
watts/wattt eMeV200(1c)E(EP/P 1 1 101++= (15.1-1) watts/wattte)t et (eMeV200(1c)E(EP/P 2 1 221 2 2 202+++= (15.1-2) where:
P 1/P 0 is the energy from U-239 decay P 2/P 0 is the energy from Np-239 decay t is the time after shutdown (seconds) c(1+) is the ratio of U-238 captures to total fissions = 0.6 (1 + 0.2) 1 = the decay constant of U-239 = 4.91 x 10
-4 per second 2 = the decay constant of Np-239 decay = 3.41 x 10
-6 per second E1 = total -ray energy from U-239 decay = 0.06 MeV E2 = total -ray energy from Np-239 decay = 0.30 MeV E1 = total -ray energy from U-239 decay = 1/3 (a) x 1.18 MeV E2 = total -ray energy from Np-239 decay = 1/3 (a) x 0.43 MeV (a) Two-thirds of the potential -energy is assumed to escape by the accompanying neutrinos.
For the SBLOCA, based on conservative modeling of the ratio of U-238 captures to total
fissions, heavy element decay heat is calculated without applying further uncertainty
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-15 Revision 23 December 2016 correction (Reference 24). For the Best Estimate LOCA analysis, the heat from the radioactive decay of U-239 and Np-239 is calculated as described in Section 8 of Reference 23. The decay of other isotopes, produced by neutron reactions other than
fission, is neglected. For the non-LOCA analysis, the decay of U-238 capture products
is included as an integral part of the 1979 decay heat curve presented as Figure 15.1-7.
15.1.10.3 Residual Fissions The time dependence of residual fission power after shutdown depends on core
properties throughout a transient under consideration. Core average conditions are
more conservative for the calculation of reactivity and power level than actual local
conditions as they would exist in hot areas of the core. Thus, unless otherwise stated in the text, static power shapes have been assumed in the analysis and these are factored
by the time behavior of core average fission power calculated by a point kinetics model
calculation with six delayed neutron groups.
For the purpose of illustration, only one delayed neutron group calculation, with a constant shutdown reactivity of -4 percent k is shown in Figure 15.1-6.
15.1.10.4 Distribution of Decay Heat Following Loss-of-Coolant Accident During an SBLOCA the core is rapidly shut down by void formation or RCCA insertion, or both, and long-term shutdown is assured by the borated ECCS water. 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 that are impo rtant for the neutron dependent part of the
heat generation do not apply to the gamma ray source contribution. The steady state factor of 97.4 percent that represents the fraction of heat generated within the cladding and pellet drops to 95 percent for the hot rod in a LOCA.
For example, 1/2 second after the rupture about 30 percent of the heat generated in the
fuel rods is from gamma ray absorption. The gamma power shape is less peaked than
the steady state fission 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 percent of the gamma ray c ontribution or 3 percent of the total. Since the water density is considerably reduced at this time, an average of 98 percent of the
available heat is deposited in the fuel rods, the remaining 2 percent 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.
For the Best Estimate LOCA analysis, the energy deposition modeling is performed as
described in Section 8 of Reference 23.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-16 Revision 23 December 2016 15.1.11 REFERENCES
- 1. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.
- 3. H. Chelemer, et al., Improved Thermal Design Procedure, WCAP-8567-P-A (Proprietary) and WCAP-8568-A (Non-Proprietary), February 1989.
- 4. F. E. Motley, et al., New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids, WCAP-8762-P-A and WCAP-8763-A, July 1984.
- 5. S. L. Davidson, and W. R. Kramer; (Ed.) Reference Core Report VANTAGE 5 Fuel Assembly, WCAP-10444-P-A (Proprietary) and WCAP-10445-NP-A (Non-Proprietary), Appendix A.2.0, September 1985.
- 6. Deleted in Revision 22.
- 7. Deleted in Revision 22.
- 8. Deleted in Revision 22.
- 9. Deleted in Revision 22.
- 10. Deleted in Revision 22.
- 11. ANSI/ANS-5.1-1979, Decay Heat Power In Light Water Reactors, August 29, 1979.
- 12. Deleted in Revision 22.
- 13. Deleted in Revision 22.
- 14. Deleted in Revision 22.
- 15. Deleted in Revision 22.
- 16. Deleted in Revision 22.
- 17. Diablo Canyon Pressurizer Pressure Controller Uncertainty, Westinghouse Nuclear Safety Advisory Letter (NSAL)92-005, September 22, 1992.
DCPP UNITS 1 &
2 FSAR UPDATE 15.1-17 Revision 23 December 2016
- 18. PG&E Nuclear Plant, Diablo Canyon Units 1 and 2, Pressurizer Pressure Control System Uncertainty Safety Assessment, Westinghouse Letter PGE-93-659, November 18, 1993.
- 19. S. Miranda, et al., Steam Generator Low Water Level Protection System Modifications to Reduce Feedwater Related Trips, WCAP-11325-P-A, Rev. 1, February 1988.
- 20. Deleted in Revision 13.
- 21. RETRAN-02 -- A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, Volume 1: Theory and Numerics, (Revision 5), EPRI NP-1850-CCM-A, March 1992.
- 22. Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Methodology, Diablo Canyon Units 1 and 2, 24-Month Fuel Cycle Evaluation, WCAP-11594, Revision 2, January 1997.
- 23. S. M. Bajorek, et al., C ode Qualification Document for Best Estimate LOCA Analysis, Volume I: Models and Correlations, WCAP-12945-P-A, Volume I, Revision 2, March 1998.
- 24. NUREG-0800, Standard Review Plan, Bran ch Technical Position ASB 9-2, Residual Decay Energy for Light-Water Reactors for Long Term Cooling, July 1981.
- 25. Deleted in Revision 22.
- 26. Deleted in Revision 22.
- 27. Deleted in Revision 22.
- 28. Westinghouse Letter PGE 53, "Transmittal of LBIE to Address the Increase in Pressurizer Level in Modes 3, 4, & 5, September 23, 2010.
- 29. Diablo Canyon Units 1 and 2 Tavg and Tfeed Ranges Program NSSS Engineering Report, WCAP-16985-P, Revi sion 2 (Proprietary), April 2009.
- 30. Proposed American Nuclear Society Standard"Dec ay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors." Approved by Subcommittee ANS-5, ANS Standards Committee, October 1971.
- 31. Westinghouse Letter PGE-12-25, Pressurizer Level Increase up to 35% Span at 20% Power to Mode 3 Revised Final Engineering Report, March 7, 2012.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-1 Revision 23 December 2016 15.2 CONDITION II - FAULTS OF MO DERATE FREQUENCY These faults result at worst in the reactor shutdown with the plant capable of returning to operation. By definition, these faults (or events) do not propagate to cause a more
serious fault, i.e., a Condition III or IV fault. In addition, Condition II events are not
expected to result in fuel rod failures, reactor coolant system (RCS) overpressurization, or main steam system (MSS) overpressurization.
For the purposes of this report the following faults have been grouped into these categories:
(1) Uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition (2) Uncontrolled rod cluster control assembly bank withdrawal at power (3) Rod cluster control assembly misoperation (4) Uncontrolled boron dilution (5) Partial loss of forced reactor coolant flow (6) Startup of an inactive reactor coolant loop (Historical)
(7) Loss of external electrical load and/or turbine trip (8) Loss of normal feedwater (9) Loss of offsite power to the station auxiliaries (10) Excessive heat removal due to feedwater system malfunctions (11) Sudden feedwater temperature reduction (12) Excessive load increase incident (13) Accidental depressurization of the reactor coolant system (14) Accidental depressurization of the main steam system (15) Spurious operation of the safety injection system at power
Each of these faults of moderate frequency are analyzed in this section. In general, each analysis includes acceptance criteria, an identification of causes and description of the accident, an analysis of effects and consequences, a presentation of results, and
relevant conclusions.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-2 Revision 23 December 2016 An evaluation of the reliability of the reactor protection system actuation following initiation of Condition II events has been completed and is presented in Reference 1 for the relay protection logic. Standard reliabi lity engineering techniques were used to assess the likelihood of the trip failure due to random component failures.
Common-mode failures were also qualitatively investigated. It was concluded from the
evaluation that the likelihood of no trip following the initiation of Condition II events is extremely small (2 x 10
-7 derived for random component failures). The solid-state protection system design has been evaluated by the same methods as used for the
relay system and the same order of magnitude of reliability is provided.
Hence, because of the high reliability of the protection system, no special provision is
included in the design to cope with the consequences of Condition II events without trip.
The time sequence of events corresponding to the respective Condition II fault is shown
in Table 15.2-1.
15.2.1 UNCONTROLLED ROD CLUST ER CONTROL ASSEMBLY BANK WITHDRAWAL FROM A SUBCRITICAL CONDITION
15.2.1.1 Acceptance Criteria The following is the relevant specific acceptance criterion.
(1) Minimum DNBR is not less than the appropriate limit value at any time during the transient.
15.2.1.2 Identification of Causes and Accident Description
A rod cluster control assembly (RCCA) withdrawal accident is defined as an
uncontrolled increase in reactivity in the reactor core caused by withdrawal of RCCAs
resulting in a power excursion. Such a transient could be caused by a malfunction of
the reactor control or control rod drive systems. The Section 15.2.1 event occurs with
the reactor at hot zero power (i.e., subcritical). The at-power case is discussed in
Section 15.2.2.
Although the reactor can be brought to power from a subcritical condition by means of
RCCA withdrawal, startup procedures following refuelin g also permit boron dilution.
The maximum rate of reactivity increase in the case of boron dilution is less than that
assumed in this analysis (Refer to Section 15.2.4).
The RCCA drive mechanisms are wired into preselected bank configurations that are
not altered during core reactor life. These circuits prevent the assemblies from being
withdrawn in other than their respective banks. Power supplied to the banks is
controlled so that no more than two banks can be withdrawn at the same time. The
RCCA drive mechanisms are of the magneti c latch type and coil actuation is sequenced to provide variable speed travel. The maximum reactivity insertion rate analyzed in the DCPP UNITS 1 &
2 FSAR UPDATE 15.2-3 Revision 23 December 2016 detailed plant analysis is that occurring with the simultaneous withdrawal of the two control banks having the maximum combined worth at maximum speed.
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 burst 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.
15.2.1.2.1 Source Range High Neutron Flux Reactor Trip The source range high neutron flux reactor trip is actuated when either of two
independent source range channels indicat es a neutron flux level above a preselected manually adjustable setpoint. This trip function may be manually bypassed when either 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.
15.2.1.2.2 Intermediate Range High Neutron Flux Reactor Trip The intermediate range high neutron flux reactor trip is 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 when two of
the four power range channels give reading s above approximately 10 percent of full power and is automatically reinstated when three of the four channels indicate a power below this value.
15.2.1.2.3 Power Range High Neutron Flux Reactor Trip (Low Setting)
The power range high neutron flux trip (low setting) is actuated when two-out-of-four
power range channels indicate a power level above approximately 25 percent of full power. This trip function may be manually bypassed when two of the four power range
channels indicate a power level above appro ximately 10 percent of full power and is automatically reinstated when three of the four channels indicate a power level below
10 percent.
15.2.1.2.4 Power Range High Neutron Flux Reactor Trip (High Setting)
The power range high neutron flux reactor trip (high setting) is actuated when
two-out-of-four power range channels indicate a power level above a preset setpoint.
This trip function is always active.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-4 Revision 23 December 2016 15.2.1.2.5 Power Range High Positive Neutron Flux Rate Trip The power range high positive neutron flux rate trip is actuated when the rate of change
in power on two-out-of-four power range channels exceeds the preset setpoint. This trip
function is always active.
15.2.1.3 Analysis of Effects and Consequences This transient is analyzed by three digital computer codes. The TWINKLE (Reference 2) code is used to calculate the reactivity transient and hence the nuclear
power transient. The FACTRAN (Reference 3) code is then used to calculate the
thermal heat flux transient based on the nuclear power transient calculated by the
TWINKLE code. FACTRAN also calculates the fuel, cladding, and coolant
temperatures. A detailed thermal and hydraulic computer code, THINC (refer to
Section 1.6.1, Item 28 and Section 4.4.3) (Reference 9) is used to calculate the DNB.
The event is not analyzed with the Improved Thermal Design Procedure since it is
analyzed with reduced flow.
In order to give conservative results for a startup accident, the following assumptions are made concerning the initial reactor conditions:
(1) Since the magnitude of the power peak reached during the initial part of the transient for any given rate of reactivity insertion is strongly dependent
on the Doppler coefficient, conservative values (low absolute magnitude)
as a function of power are used. Refer to Section 15.1.6 and Table 15.1-4.
(2) Contribution of the moderator reactivity coeff icient 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. The
conservative value, given in Table 15.1-4, is used in the analysis to yield
the maximum peak heat flux.
(3) The reactor is assumed to be at hot zero power. This assumption is more conservative than that of a lower initial system temperature. The higher
initial system temperature yields a larger 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 since this results in
maximum neutron flux peaking.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-5 Revision 23 December 2016 (4) 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 RCCA release, is taken into account. A 10 percent increase is assumed for the power range
flux trip setpoint, raising it from the nominal value of 25 to 35 percent.
Previous results, however, show that the rise in neutron flux is so rapid
that the effect of error on this 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 RCCA is
stuck in its fully withdrawn position.
Refer to Section 15.1.5 for RCCA insertion characteristics.
(5) The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the combination of the two control banks having the greatest combined worth at maximum speed
(45 inches/minute). Control rod drive mechanism design is discussed in
Section 4.2.3.
(6) The initial power level is assumed to be below the power level expected for any shutdown condition. The combination of highest reactivity
insertion rate and lowest initial power produces the highest peak heat flux.
15.2.1.4 Results Figures 15.2.1-1 through 15.2.1-3 show the transient behavior for the indicated
reactivity insertion rate with the accident terminated by reactor trip at 35 percent nominal power. This insertion rate is greater than that for the two highest worth control banks, both assumed to be in their highest incremental worth region.
Figure 15.2.1-1 shows the neutron flux transient. The neutron flux overshoots the full
power nominal value but this occurs for only a very short time period. Hence, the
energy release and the fuel temperature increase are relatively small. The thermal flux
response, of interest for departure from nucleate boiling (DNB) considerations, is shown
in Figure 15.2.1-2. The beneficial effect on the inherent thermal lag in the fuel is
evidenced by a peak heat flux less than the full power nominal value. The DNBR
remains above the applicable safety analysis limit value at all times.
Figure 15.2.1-3 shows the response of the average fuel, cladding, and coolant
temperatures at the hot spot.
15.2.1.5 Conclusions
The analysis demonstrates that the acceptance criterion is met as follows:
(1) Minimum DNBR remains above the appropriate limit value at any time during the transient.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-6 Revision 23 December 2016 In the event of an RCCA withdrawal accide nt from the subcritical condition, the core and the RCS are not adversely affected since the combination of thermal power and the
coolant temperature result in a DNBR above the limiting value.
15.2.2 UNCONTROLLED ROD CLUST ER CONTROL ASSEMBLY BANK WITHDRAWAL AT POWER
15.2.2.1 Acceptance Criteria The following are the relevant specific acceptance criteria.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient.
(2) The peak core average power (heat flux) does not exceed a value that would cause fuel centerline melt at any time during the transient (refer to
Section 4.4.3.2.7).
(3) The RCS pressure does not exceed 110% of design pressure (2,750 psia) at any time during the transient.
(4) The pressurizer does not go water solid at any time during the transient.
15.2.2.2 Identification of Causes and Accident Description Uncontrolled RCCA bank withdrawal at power results in an increase in the core heat
flux. Since the heat extraction from the steam generator (SG) 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 would eventually result in DNB, an RCS overpressure condition, or the pressurizer
filled with liquid. Therefore, the reactor protection system is designed to terminate any
such transient before the DNBR falls below the safety analysis limit values, the RCS
pressure exceeds 110 percent of the design value, or the pressurizer becomes filled
with liquid.
The automatic features of the reactor protection system that ensure these limits are not
exceeded following the postulated accident include the following:
(1) The power range neutron flux instrumentation actuates a reactor trip if two-out-of-four channels exceed a high flux or a positive flux rate high
setpoint.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-7 Revision 23 December 2016 (2) The reactor trip is actuated if any two-out-of-four T channels exceed an overtemperature T setpoint.
(3) The reactor trip is actuated if any two-out-of-four T channels exceed an overpower T setpoint.
(4) A high pressurizer pressure reactor trip actuated from any two-out-of-four pressure channels that are set at a fixed point.
(5) A high pressurizer water level reactor trip actuated from any two-out-of-three level channels that are set at a fixed point.
The positive flux rate trip provides adequate protection to ensure that the most limiting RCCA bank withdrawal event does not result in the peak RCS pressure exceeding 110 percent of the design limit. The positive flux rate trip setpoint and response time that are credited in the evaluation of this event are listed in Table 15.1-2. Various reactor trips (e.g. High Neutron Flux) may also be credited to prevent RCS overpressurization during an RCCA bank withdrawal event.
Reference 18 documents a generic and cons ervatively bounding evaluation that has been performed to ensure that pressurizer overfill conditions are not a concern for this event. The evaluation demonstrates that the pressurizer water level high trip prevents a pressurizer overfill condition for those RCCA bank withdrawal events that are very slow and do not generate any other automatic protection signal. The pressurizer water level high trip response time is listed as N/A with the note indicating that the evaluation results are insensitive to the assumed response time.
The manner in which the combination of overpower and overtemperature T trips provide fuel cladding protection over the full range of RCS conditions is described in Chapter 7 and Section 15.1.3.
15.2.2.3 Analysis of Effects and Consequences This transient is analyzed by the LOFTRAN (Reference 4) code. This code simulates the neutron kinetics, RCS, 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 core
limits as illustrated in Figure 15.1-1 are used as input to LOFTRAN to determine the
minimum DNBR during the transient.
This accident is analyzed with the Improved Thermal Design Procedure and the initial
condition uncertainties are included in the li mit DNBR as described in Reference 5.
Therefore, initial conditions of nominal core power, nominal reactor coolant average
temperatures (including 2.5°F for steam generator fouling) and nominal reactor coolant
pressure are assumed.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-8 Revision 23 December 2016 In order to obtain conservative results, the following assumptions are made:
(1) Reactivity Coefficients - two cases are analyzed:
(a) Minimum reactivity feedback. A positive moderator coefficient of reactivity of +5 pcm/°F is assumed. A variable Doppler power
coefficient with core power is used in the analysis. A conservatively
small (in absolute magnitude) value is assumed.
(b) Maximum reactivity feedback. A conservatively large positive A large (in absolute magnitude) negative Doppler power coefficient is
assumed. (2) The reactor trip on high neutron flux is assumed to be actuated at a conservative value of 118 percent of nominal full power. The T trips include all adverse instrumentation and setpoint errors, while the delays
for the trip signal actuation are assumed at their maximum values.
(3) The RCCA trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.
(4) The maximum positive reactivity insertion rate is greater than that which would be obtained from the simultaneous withdrawal of the two control rod
banks having the maximum combined worth at maximum speed.
The effect of RCCA movement on the axial core power distribution is accounted for by
margin to DNB.
15.2.2.4 Results Figures 15.2.2-1 and 15.2.2-2 show the response of neutron flux, pressure, average
coolant temperature, and DNBR (thimble cell) due to a rapid RCCA withdrawal 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 T avg and pressure result and a large margin to DNB is maintained.
The response of neutron flux, pressure, average coolant temperature, and DNBR (thimble cell) for a slow control rod assembly withdrawal from full power is shown in Figures 15.2.2-3 and 15.2.2-4. 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 never less than the safety
analysis limit values.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-9 Revision 23 December 2016 Figure 15.2.2-5 shows the minimum DNBR (thimble cell) as a function of reactivity insertion rate from initial full power operation for the minimum and for the maximum
reactivity feedbacks. 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
analysis limit values.
Figures 15.2.2-6 and 15.2.2-7 show the minimum DNBR (thimble cell) as a function of
reactivity insertion rate for RCCA withdrawal incidents starting at 60 and 10 percent
power, respectively. The results are similar to the 100 percent power case, except that
effective is increased. In neither case does the DNBR fall below the safety analysis limit values.
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 15.2.2-7, for example, it is noted that:
(1) For reactivity insertion rates above 30 pcm/sec reactor trip is initiated by the high neutron flux trip for the minimum reactivity feedback cases. The
neutron flux level in the core rises rapidly for these insertion rates while
core heat flux and coolant system temperatur e lag behind due to the thermal capacity of the fuel and coolant system fluid. Thus, the reactor is
tripped prior to significant increase in heat flux or water temperature with
resultant high minimum DNB ratios during the transient. As reactivity
insertion rate decreases, core heat flux and coolant temperatures can remain more nearly in equilibrium with the neutron flux. Minimum DNBR during the transient thus decreases with decreasing insertion rate.
(2) trip circuit initiates a reactor trip when
average temperature and pressure. It is important to note that the average temperature contribution to the circuit is lead-lag compensated in
order to decrease the effect of th e thermal capacity of the RCS in response to power increase.
(3) For reactivity insertion rate below trip terminates the transient.
For reactivity insertion rates between 30 pcm/sec and 7 pcm/sec the trip increases (in terms of
increased minimum DNBR) 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.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-10 Revision 23 December 2016 (4) For reactivity insertion rates less than 7 pcm/sec, the rise in the reactor coolant temperature is sufficiently high so that the steam generator safety valve setpoint is reached prior to trip. Opening of these valves, which act
as an additional heat load on the RCS, sharply decreases the rate of
increase of RCS average temperature. This decrease in rate of increase
of the average RCS temperature during the transient is accentuated by
to be reached later with a resulting lower minimum DNBR.
Figures 15.2.2-5, 15.2.2-6, and 15.2.2-7 illustrate minimum DNBRs calculated for
minimum and maximum reactivity feedback.
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 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 percent 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 cor e power distribution, the peak fuel centerline temperature will still remain below the fuel melting temperature.
For slow reactivity insertion rates, the core heat flux remains more nearly in equilibrium
reactor trip before a DNB condition is reached. The peak heat flux again is maintained
below 118 percent of its nominal value. Taking into account the effect of the RCCA
withdrawal on the axial core power distribution, the peak fuel centerline temperature will remain below the fuel melting temperature.
Since DNB is not predicted to 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 in Table 15.2-1. 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.
15.2.2.5 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell). The accompanying DNBR figures for this event (Figure
15.2.2-2, and Figures 15.2.2-4 through 15.2.2-7) reflect the results for the
more limiting Standard fuel (limit 1.48/1.44) previously in the core.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-11 Revision 23 December 2016 (2) The core heat flux is maintained below 118 percent of its nominal value.
Thus the peak fuel centerline temperature will remain below the fuel
melting temperature (refer to Section 4.4.3.2.7).
(3) The RCS pressure does not exceed 110 percent of design pressure (2,750 psia) at any time during the transient.
(4) The pressurizer does not become water solid during the event.
protection over the entire range of possible reactivity insertion rates; i.e., the minimum
value of DNBR is always larger than the safety analysis limit values.
15.2.3 ROD CLUSTER CONTROL ASSEMBLY MISOPERATION This section discusses RCCA misoperation that can result either from system
malfunction or operator error.
15.2.3.1 Acceptance Criteria The following is the relevant specific acceptance criterion.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient.
15.2.3.2 Identification of Causes and Accident Description RCCA misoperation accidents include:
(1) One or more dropped RCCAs within the same group (2) A dropped RCCA bank (3) Statically misaligned RCCA
Each RCCA has a position indicator channel that displays the 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.
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 DCPP UNITS 1 &
2 FSAR UPDATE 15.2-12 Revision 23 December 2016 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 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 that would cause rod withdrawal would affect a minimum of one group. Mechanical failures are in the direction of insertion, or immobility.
A dropped RCCA, or RCCA bank, is detected by:
(1) A sudden drop in the core power level as seen by the nuclear instrumentation system (2) Asymmetric power distribution as seen on out-of-core neutron detectors or core-exit thermocouples (3) Rod at bottom signal (4) Rod deviation alarm (5) Rod position indication
Misaligned RCCAs are detected by:
(1) Asymmetric power distribution as seen on out-of-core neutron detectors or core-exit thermocouples (2) Rod deviation alarm (3) Rod position indicators
The deviation alarm alerts the operator whenever an individual rod position signal
deviates from the other rods in the bank by a preset limit.
During time intervals when the Rod Position Deviation Monitor is inoperable:
(1) Each rod position indicator is determined to be operable by verifying that the Demand Position Indication System and the Digital Rod Position
Indication System agree within 12 steps at least once per four hours.
During time intervals when the rod insertion limit monitor is inoperable, the individual rod
positions are verified to be within insertion limits at least once per four hours.
If one or more rod position indicator channels should be out of service, detailed
operating instructions are followed to ensure the alignment of the nonindicated RCCAs.
The operator is also required to take action as required by the Technical Specifications (TS).
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-13 Revision 23 December 2016 15.2.3.3 Analysis of Effects and Consequences The accident is analyzed with the Improved Thermal Design Procedure and the initial condition uncertainties are included in the limit DNBR as described in Reference 5.
Therefore, initial conditions of nominal core power, nominal reactor coolant average
temperature and nominal reactor coolant pressure are assumed.
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, RCS, 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 conditio ns from the transient and the hot channel factor from the nuclear analysis, the DNB design basis is
shown to be met using the THINC code (refer to Section 1.6.1, Item 28
and Section 4.4.3). The transient response, nuclear peaking factor
analysis, and DNB design basis confirmation are performed in accordance with the methodology described in Reference 10.
(2) Dropped RCCA Bank A dropped RCCA bank results in a symmetric power change in the core.
As discussed in Reference 10, assumptions made for the dropped RCCA(s) analysis provide a bounding analysis for the dropped RCCA bank. (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 used as input to
the THINC code (refer to Section 1.6.1, Item 28 and Section 4.4.3) to
calculate the DNBR. The analysis examines the case of the worst rod withdrawn from control bank D inserted at the insertion limit with the
reactor initially at full power. The analys is assumes this incident to occur at beginning of life or the time in core life which this results in the minimum
value of moderator temperature coefficient. This assumption maximizes DCPP UNITS 1 &
2 FSAR UPDATE 15.2-14 Revision 23 December 2016 the power rise and minimizes the tendency of increased moderator temperature to flatten the power distribution.
15.2.3.4 Results (1) One or More Dropped RCCAs Single or multiple dropped RCCAs within the same group result in a negative reactivity insertion. The core is not adversely affected during this
period since power is decreasing rapidly.
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 equi librium 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. Figures 15.2.3-1 and 15.2.3-2
show a typical transient response to a dropped RCCA(s) in automatic
control. Uncertainties in the initial conditions are included in the DNB
evaluation as described in Reference 10. In all cases, the minimum DNBR remains above the safety analysis limit value.
Following plant stabilization, the operator may manually retrieve the
RCCA(s) by following approved operating procedures.
(2) Dropped RCCA Bank A dropped RCCA bank typically results in a reactivity insertion of greater
than 500 pcm. The core is not adversely affected during the insertion
period since power is decreasing rapidly. The dropped RCCA bank
transient will proceed as described in the previous section for one or more
dropped RCCA(s), except the return to power will be less due to the
greater worth of the entire bank. The power transient for a dropped RCCA
bank is symmetric. Following plant stabilization, normal procedures are
followed.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-15 Revision 23 December 2016 (3) Statically Misaligned RCCA The most severe misalignment situations with respect to DNBR at significant power levels arise from cases in which one RCCA 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. The full power insertion
limits on control bank D must be chosen to be above that position which
meets the minimum DNBR and peaking factor limits. The full power
insertion limits is usually dictated by other criteria. Detailed results will
vary from cycle to cycle depending on fuel arrangements.
For this RCCA misalignment, with Bank D inserted to its full power
insertion limit and one RCCA fully withdraw n, DNBR does not fall below the safety analysis limit value. This case is 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.
For RCCA misalignments with one RCCA fully inserted, the DNBR does
not fall below the safety analysis limit value. This case is 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.
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 a xial power distribution. The resulting linear heat generation is well below that which would cause fuel
melting.
Following the identification of an RCCA group misalignment condition by
the operator, the operator is required to take action as required by the
plant Technical Specifications and operating instructions.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-16 Revision 23 December 2016 15.2.3.5 Conclusions The analysis demonstrates that the acceptance criterion is met as follows:
(1) For all cases of RCCA misoperation, the DNBR remains greater than the Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell); therefore, the
DNB design criterion is met.
15.2.4 UNCONTROLLED BORON DILUTION 15.2.4.1 Acceptance Criteria (1) There is ample/adequate time for the operator to mitigate a boron dilution event. 15.2.4.2 Identification of Causes and Accident Description Reactivity can be added to the core by feeding unborated water into the RCS via the
reactor makeup portion of the chemical and volume control system (CVCS). Boron
dilution is a manual operation under strict administrative controls with procedures calling
for a limit on the rate and duration of dilution. A boric acid blend system is provided to
permit the operator to match the boron concentration of reactor coolant makeup water to that in the RCS during normal makeup injec tion. The CVCS is designed to limit, even under various postulated failure modes, the potential rate of dilution to a value, which
after indication through alarms and instrumentation, provides the operator with sufficient
time to correct the situation in a safe and orderly manner.
The opening of the primary water makeup control valves provides makeup to the RCS
that can dilute the reactor coolant. Inadvert ent dilution from this source can be readily terminated by closing the control valve. In order for makeup water to be added to the
RCS at pressure, at least one charging pump must be running in addition to a primary
makeup water pump.
The rate of addition of unborated makeup water to the RCS when it is not at pressure is limited by the capacity of the primary water supply pumps. The maximum net addition rate in this case is 200 gpm, which is based on a conservative evaluation of two primary water pumps operating in parallel through a common flow path.
The boric acid from the boric acid tank is blended with primary grade water in the
blender and the composition is determined by the preset flowrates of boric acid and
primary grade water on the control board. In order to dilute, two separate operations
are required:
(1) The operator must change from the automatic makeup mode to the dilute mode DCPP UNITS 1 &
2 FSAR UPDATE 15.2-17 Revision 23 December 2016 (2) The operator must select start to initiate system start Excluding either step would prevent dilution.
Information on the status of the reactor coolant makeup is continuously available to the operator. Lights are provided on the control board to indicate the operating condition of
the pumps in the CVCS. Alarms are actuated to warn the operator if boric acid or
demineralized water flowrates deviate from preset values as a result of system
malfunction.
Consistent with the DCPP licensing basis, the acceptance criteria of meeting a minimum time before loss of SDM from the start of dilution for Modes 2-5 is 15 minutes; for Mode 6, it is 30 minutes.
In order to meet the acceptance criteria for Mode 4 on RHR and Mode 5 (for both filled and mid-loop operation), the analysis defines a required minimum critical boron concentration ratio (Cbi/Cbc) that must be confirmed on a reload basis. This is a ratio of the initial boron concentration (Cbi) in the reactor coolant system to the boron concentration at which shutdown margin is lost (Cbc). Consistent with this approach, a minimum Cbi/Cbc ratio is also defined for Mode 3 and Mode 4 with one RCP in operation, in order to provide additional margin. These limits are evaluated for the core reloads of both units as part of the normal Restart Safety Analysis Checklist process. If analysis shows that these ratios will be violated for future reload cycles, administrative and/or operating procedures will need to be revised to ensure that these limits are maintained. In such case, the Core Operating Limits Report must be revised at that time to specify either an increased SDM requirement or the required minimum Cbi/Cbc ratio(s) directly.
15.2.4.3 Analysis of Effects and Consequences 15.2.4.3.1 Method of Analysis To cover all phases of plant operation, boron dilution during refueling, cold shutdown, hot shutdown, hot standby, startup, and power operation is considered in this analysis.
Table 15.2-1 contains the time sequence of events for this accident.
15.2.4.3.2 Dilution during Refueling During refueling the following conditions exist:
(1) One residual heat removal (RHR) pump is operating to ensure continuous mixing in the reactor vessel.
(2) The seal injection water supply to the reactor coolant pumps (RCPs) is typically isolated for the purpose of performing RCP maintenance.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-18 Revision 23 December 2016 (3) Boric acid supply to the suction of the charging pumps is available for the addition of boric acid to the RCS. Alternatively, boric acid supply may be lined up to the suction of the safety injection pumps when all the reactor
vessel head bolts are fully detensioned.
(4) The boron concentration in the refueling water is greater than or equal to 2000 ppm, corresponding to a shutdown margin of at least 5 percent k with all RCCAs in; periodic sampling ensures that this concentration is maintained.
(5) Neutron sources are installed in the core and the source range detectors outside the reactor vessel are active and provide an audible count rate.
During initial core loading, BF 3 detectors are installed inside the reactor vessel and are connected to instrumentation giving audible count rates to provide direct monitoring of the core. (Historical)
(6) A minimum water volume in the RCS of 3462 cubic feet is considered.
This corresponds to the volume necessary to fill the reactor vessel above the nozzles to ensure mixing via the RHR loop.
(7) A maximum dilution flow of 200 gpm and uniform mixing are assumed.
The operator has prompt and definite indication of any boron dilution from the audible count rate instrumentation and the high flux at shutdown alarm in the control room.
Count rate will increase with the subcritical multipl ication factor during boron dilution.
If a safety injection pump is used for boration, it is aligned to take suction from the refueling water storage tank (RWST) and discharge to the cold legs of the RCS, and the boundary valves from the CVCS to the safety injection system (SIS) are closed. These
requirements ensure no new dilution flowpath s are introduced when using the SIS boration flowpath.
15.2.4.3.3 Dilution during Cold Startup In this mode, the plant is being taken into or out of refueling or hot shutdown. Typically, the plant is maintained in the cold shutdown mode when reduced RCS inventory is necessary, ambient temperatures are required to address various plant issues, or as a result of a Technical Specification action statement. The water level can be dropped to the mid-plane of the hot leg for maintenance work that requires the steam generators to be drained. The plant is maintained in cold shutdown at the beginning of the cycle for start-up testing of certain systems and components. Conditions used for the analysis are as follows:
(1) A maximum dilution flow of 200 gpm, limited by the capacity of two primary water makeup pumps operating in parallel through a common flow path, is considered.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-19 Revision 23 December 2016 (2) A minimum RCS water volume of 4690 cubic feet is used, corresponding to the active RCS volume excluding the pressurizer and the reactor vessel upper head. A minimum RCS water volume of 3462 cubic feet is also considered for mid-loop operation.
(3) The required operator action time for this mode is 15 minutes. To meet this requirement, a minimum critical boron concentration ratio must be maintained to ensure sufficient time is available from the initiation of the dilution event for the operators to act to prevent a loss of shutdown margin and preclude criticality.
The analysis determined that the minimum critical boron concentration ratio to meet the required operator action is 1.128 for the most limiting case, assuming mid-loop operation.
15.2.4.3.4 Dilution during Hot Shutdown In this mode, the plant is being taken into or out of cold shutdown or hot standby. The plant is maintained in this mode at the beginning of the cycle for startup testing of certain systems and components. Throughout the cycle, the plant may enter hot shutdown if plant issues arise requiring a plant shutdown or as a result of a Technical Specification action statement. In hot shutdown, mixing of the RCS is provided by either the RHR system or a single RCP, de pending on system pressure and temperature. Conditions used for the analysis are as follows:
(1) A maximum dilution flow of 200 gpm, limited by the capacity of two primary water makeup pumps operating in parallel through a common flow path, is considered.
(2) A minimum RCS water volume of 9365 cubic feet is used, corresponding to the active RCS volume excluding the pressurizer and the reactor vessel upper head with one RCP in operation. A reduced minimum RCS water volume of 4690 cubic feet is considered for the case with no RCPs operating and mixing provided by the RHR system.
(3) The required operator action time for this mode is 15 minutes. To meet this requirement, a minimum critical boron concentration ratio must be maintained to ensure sufficient time is available from the initiation of the dilution event for the operators to act to prevent a loss of shutdown margin and preclude criticality.
The analysis determined that the minimum critical boron concentration ratio to meet the required operator action time is 1.101 for the most limiting case, assuming RHR system operation only.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-20 Revision 23 December 2016 15.2.4.3.5 Dilution during Hot Standby In this mode, the plant is being taken into or out of hot shutdown or startup. The plant is maintained in hot standby at the beginning of cycle for startup testing of certain systems and components, and to achieve plant heatup before entering the startup mode and going critical. During cycle operation, the plant will enter this mode following a reactor trip or as a result of a Technical Specification action statement. During hot standby, not all RCPs may be in operation. Rod control is in manual and the rods may be partially or completely withdrawn. The more limiting hot standby dilution scenario is with the control rods not withdrawn and the reactor shut down by boron to the Technical Specifications minimum requirement for this mode. Conditions used for the analysis are as follows:
(1) A maximum dilution flow of 200 gpm, limited by the capacity of two primary water makeup pumps operating in parallel through a common flow path, is considered.
(2) A minimum RCS water volume of 9365 cubic feet is used, corresponding to the active RCS volume excluding the pressurizer and the reactor vessel upper head with at least one RCP in operation.
(3) The required operator action time for this mode is 15 minutes. To meet this requirement, a minimum critical boron concentration ratio must be maintained to ensure sufficient time is available from the initiation of the dilution event for the operators to act to prevent a loss of shutdown margin and preclude criticality.
The analysis determined that the minimum critical boron concentration ratio to meet the required operator action time is 1.059.
15.2.4.3.6 Dilution during Startup In this mode, the plant is being taken into or out of hot standby or power operation. The RCS is filled with borated water from the blender during vacuum fill. Conditions used for the analysis are as follows:
(1) The initial boron concentration is modeled as 2000 ppm boron, which is conservative.
(2) Core monitoring is by external BF 3 detectors.
(3) Mixing of the reactor coolant is accomplished by operation of all four reactor coolant pumps.
(4) The High Flux at Shutdown, NIS Source Range High Flux 1/2 Reactor Trip, and Reactor Trip Initiated alarms are available to wan the operator of the transient.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-21 Revision 23 December 2016 (5) A maximum dilution flow of 200 gpm, limited by the capacity of the two primary water makeup pumps operating in parallel through a common flow path, is considered.
(6) The volume of the reactor coolant is 9883 cubic feet, which is the minimum active volume of the RCS excluding the pressurizer.
The analysis determined that to maintain a minimum critical boron concentration ratio (initial boron concentration at the most reactive time in core life to the maximum critical boron concentration) of 1.25, the operator action time is 61.6 minutes.
15.2.4.3.7 Dilution at Power With the unit at power and the RCS at pressure, the dilution rate is limited by the capacity of the charging pumps. The effective reactivity addition rate for the reactor at
full power and for a boron dilution flow of 262 gpm is shown as a function of RCS boron
concentration in Figure 15.2.4-1. This figure includes the effect of increasing boron worth with dilution. The reactivity rate used in the analysis is 1.752 x 10
-5 k/sec based on a conservatively high value for the expected boron concentration (1600 ppm) at power.
15.2.4.4 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
For dilution during refueling and startup, the a nalysis assumes the following.
In refueling, cold shutdown, hot shutdown, hot standby, and startup, the reactor operators are relied upon to detect and recover from an inadvertent boron dilution event. Numerous alarms from the chemical and volume control system, the reactor makeup water system, and the nuclear instrumentation system are available to provide assistance to the reactor operators in the detection of an inadvertent boron dilution event. Analysis has demonstrated that the reactor operators have at least 15 minutes from initiation of the dilution event in cold shutdown, hot shutdown, hot standby, and startup, and at least 30 minutes in refueling, to terminate the dilution event and initiate boration of the RCS prior to the loss of the available shutdown margin.
For dilution during full power operation:
(1) With the reactor in automatic control at full power, the power and temperature increase from boron dilution results in the insertion of the
RCCAs and a decrease in shutdown margin. Continuation of dilution and
RCCA insertion would cause the assemblies to reach the minimum limit of
the rod insertion monitor in approximately 4.7 minutes, assuming the
RCCAs to be initially at a position providing the maximum operational DCPP UNITS 1 &
2 FSAR UPDATE 15.2-22 Revision 23 December 2016 maneuvering band consistent with maintaining a minimum control band incremental rod worth. Before reaching this point, however, two alarms
would be actuated to warn the operator of the accident condition. The first
of these, the low insertion limit alarm, alerts the operator to initiate normal
boration.
The other, the low-low insertion limit alarm, alerts the operator to follow emergency boration procedures. The low alarm is set sufficiently above
the low-low alarm to allow normal boration without the need for emergency
procedures. If dilution continues after reaching the low-low alarm, it takes
approximately 15.0 minutes after the low-low alarm before the total
shutdown margin (assuming 1.6 percent, consistent with the Technical
Specifications) is lost due to dilution. Therefore, adequate time is
available following the alarms for the operator to determine the cause, isolate the primary grade water source, and initiate boration.
(2) With the reactor in manual control and if no operator action is taken, the power and temperature rise will cause the reactor to reach the high
accident in this case is essentially identical to a RCCA withdrawal accident
at power. The maximum reactivity insertion rate for boron dilution is
shown in Figure 15.2.4-1 and is seen to be within the range of insertion
rates analyzed for a RCCA withdrawal accident. Reactor trip will occur
approximately 40 seconds after event initiation. If dilution were to continue
after the reactor trip, there would still be approximately 14.5 minutes left
after a reactor trip for the operator to determine the cause of dilution, isolate the primary grade water sources, and initiate reboration before the
reactor can return to criticality assuming a 1.6 percent shutdown margin at
the beginning of dilution. Therefore, there is ample time available (approximately 40 seconds to reactor trip plus 14.5 minutes after a reactor
trip).
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-23 Revision 23 December 2016 15.2.5 PARTIAL LOSS OF FORCED REACTOR COOLANT FLOW HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED During review of this accident, it was identified that the complete lo ss of flow reanalysis (Westinghouse calculation note CN-TA-12-29) without undervoltage and underfrequency reactor trips credits the reactor coolant loop (RCL) low flow reactor trip as protection. This reactor trip is also credited in the partial loss of flow event. Since the complete loss of flow reanalysis assu mes all four reactor coolant pumps (RCPs) coasting down and the partial loss of flow analysis assumes two out of four RCPs coast down, the complete loss of flow accident, as discussed in UFSAR Section 15.3.4 is bounding.
15.2.5.1 Acceptance Criteria
The following is the relevant specific acceptance criterion.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refe r to Section 4.4.4.1) at any time during the transient.
15.2.5.2 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. 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.
The necessary protection against a partial loss of coolant flow accident is provided by
the low primary coolant flow reactor trip that is actuated by two-out-of-three low flow signals in any reactor coolant loop.
Above approximately 35 percent power (Permissive 8), low flow in any loop will actuate a reactor trip. Between approximately 10 and 35 percent power (Permissive 7 and Permissive 8), low flow in any two loops will
actuate a reactor trip. Reactor trip on low flow is blocked below Permissive 7.
A reactor trip on reactor coolant pump breakers open is provided as a backup to the low
flow signals. Above Permissive 7, a breaker open signal from any two pumps will actuate a reactor trip. Reactor trip on reactor coolant pump breakers open is blocked
below Permissive 7.
Normal power for the reactor coolant pumps is supplied through buses connected through transformers to the generator. Two reactor coolant pumps are on each bus.
When a generator trip occurs, the buses are automatically transferred to a power source supplied from external power lines, and the pumps will continue to supply coolant flow
to the core. Following any turbine trip where there are no electrical or mechanical faults DCPP UNITS 1 &
2 FSAR UPDATE 15.2-24 Revision 23 December 2016 that require immediate tripping of the generator from the network, the generator remains connected to the network for approximately 30 seconds. The reactor coolant pumps
remain connected to the generator thus ensuring full flow for approximately 30 seconds
after the reactor trip before any transfer is made.
15.2.5.3 Analysis of Effects and Consequences
15.2.5.3.1 Method of Analysis
The following case has been analyzed:
(1) All loops operating, two loops coasting down.
This transient is analyzed by three digital computer codes. First the
LOFTRAN code is used to calculate the loop and core flow during the
transient. The LOFTRAN code is also used to calculate the time of
reactor trip, based on the calculated flows and the nuclear power transient
following reactor trip. The FACTRAN code is then used to calculate the heat flux transient based on the nuclear power and flow from LOFTRAN.
Finally, the THINC code (refer to Section 1.6.1, Item 28 and Section 4.4.3) is used to calculate the minimum DNBR during the transient based on the heat flux from FACTRAN and the flow from LOFTRAN. The DNBR transient presented represents the minimum of the typical and thimble
cells for Standard fuel, which bound VANTAGE 5 fuel.
15.2.5.3.2 Initial Conditions The accident is analyzed using the Improved Therm al Design Procedure and the initial condition uncertainties are included in the lim it DNBR as described in Reference 5.
Therefore, initial conditions of nominal core power, nominal reactor coolant average temperature (including 2.5 °F for steam generator fouling) and nominal reactor coolant pressure are assumed.
15.2.5.3.3 Reactivity Coefficients
A conservatively large absolute value of the Doppler-only power coefficient is used (refer to Table 15.1-4). The total i ntegrated Doppler reactivity from 0 to 100 percent
The most positive moderator temperature coefficient (+5 pcm/°F) is assumed since this results in the maximum hot spot heat flux during the initial part of the transient when the minimum DNBR is reached.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-25 Revision 23 December 2016 15.2.5.3.4 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 mom entum balance, and the pump characteristics and is based on high estimates of system pressure losses.
15.2.5.4 Results
The calculated sequence of events is shown in Table 15.2-1. Figures 15.2.5-1 through
15.2.5-4 show the core flow coastdown, the loop flow coastdown, the heat flux
coastdown, and the nuclear power coastdown. The minimum DNBR is not less than the
safety analysis limit value. A plot of DNBR v
- s. time is given in Figure 15.2.5-5 for the most limiting thimble cell for Standard fuel, which boun ds VANTAGE 5 fuel.
15.2.5.5 Conclusions
The analysis demonstrates that the acceptance criterion is met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell). The accompanying DNBR figure for this event (Figure 15.2.5-5) reflects the results for the more limiting Standard fuel (limit 1.48/1.44) previously in the core.
The analysis shows that the DNBR will not decrease below the safety analysis limiting values at any time during the transient. Thus no core safety limit is violated.
15.2.6 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED
In accordance with the TS, DCPP operation during startup and power operation with less than four loops is not permitted. This analysis is presented for completeness.
15.2.6.1 Identification of Causes and Accident Description
If a plant is operating with one pump out of service, there is reverse flow through the
loop due to the pressure difference across the reactor vessel. The cold leg temperature
in an inactive loop is identical to the cold leg temperature of the active loops (the reactor
core inlet temperature). If the reactor is operated at power, and assuming the secondary side of the steam generator in the inactive loop is not isolated, there is a
temperature drop across the steam generator in the inactive loop and, with the reverse
flow, the hot leg temperature of the inactive l oop is lower than the reactor core inlet
temperature.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-26 Revision 23 December 2016 Starting of an idle reactor coolant pump w ithout bringing the inactive loop hot leg temperature close to the core inlet temperature would result in the injection of cold water into the core, which causes a rapid reactivity insertion and subsequent power
increase.
This event is classified as an ANS Condition II incident (an incident of moderate
frequency) as defined at the beginning of this chapter.
Should the startup of an inactive reactor coolant pump at an incorrect temperature
occur, the transient will be terminated autom atically by a reactor trip on low coolant loop flow when the power range neutron flux (two-out-of-four channels) exceeds the
P-8 setpoint, which has been previously reset for three-loop operation.
15.2.6.2 Analysis of Effects and Consequences
This transient is analyzed by three digital computer codes. The LOFTRAN Code (Reference 4) is used to calculate the loop and core flow, nuclear power and core
pressure and temperature transients following the startup of an idle pump. FACTRAN (Reference 3) is used to calculate the core heat flux transient based on core flow and
nuclear power from LOFTRAN. The THINC Code (Reference 9) is then used to
calculate the DNBR during the transient based on system conditions (pressure, temperature, and flow) calculated by LOFTRAN and heat flux as calculated by
FACTRAN.
In order to obtain conservative results for the startup of an inactive pump accident, the following assumptions are made:
(1) Initial conditions of maximum core power and reactor coolant average temperatures and minimum reactor coolant pressure resulting in minimum
initial margin to DNB. A 25 percent maximum steady state power level
including appropriate allowances for calibrat ion and instrument errors is assumed, however DCPP is not allowed to be at power with an inactive
loop. The high initial power gives the greatest temperature difference
between the core inlet temperature and the inactive loop hot leg
temperature.
(2) Following the start of the idle pump, the inactive loop flow reverses and accelerates to its nominal full flow value.
(3) A conservatively large (absolute value) negative moderator coefficient associated with the end of life.
(4) A conservatively low (absolute value) negative Doppler power coefficient is used.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-27 Revision 23 December 2016 (5) The initial reactor coolant loop flows are at the appropriate values for one pump out of service.
(6) The reactor trip is assumed to occur on low coolant flow when the power range neutron flux exceeds the P-8 setpoint, which has been reset for N-1 loop operation. The P-8 setpoint is conservatively assumed to be
84 percent of rated power, which corresponds to the nominal N-1 loop
operation setpoint plus 9 percent for nuclear instrumentation errors.
15.2.6.3 Results
The results following the startup of an idle pump with the above listed assumptions are shown in Figures 15.2.6-1 through 15.2.6-5. As shown in these curves, during the first
part of the transient, the increase in core flow with cooler water results in an increase in nuclear power and a decrease in core average temperature. The minimum DNBR during the transient is considerably greater than the safety analysis limit values.
Reactivity addition for the inactive loop startup accident is due to the decrease in core
water temperature. During the transient, this decrease is due both to a) the increase in
reactor coolant flow, and b) as the inactive loop flow reverses, to the colder water
entering the core from the hot leg side (colder temperature side prior to the start of the
transient) of the steam generator in the inactive loop. Thus, the reactivity insertion rate
for this transient changes with time.
The resultant core nuclear power transient, computed with consideration of both mo derator and Doppler reactivity feedback effects, is shown in Figure 15.2.6-1.
The calculated sequence of events for this accident is shown in Table 15.2-1. The transient results illustrated in Figures 15.2.6-1 through 15.2.6-5 indicate that a stabilized
plant condition, with the reactor tripped, is approached rapidly. Plant cooldown may subsequently be achieved by following normal shutdown procedures.
15.2.6.4 Conclusions
The transient results show that the core is not adversely affected. There is considerable
margin to the safety analysis DNBR limit val ues; thus, no fuel or cladding damage is predicted.
15.2.7 LOSS OF EXTERNAL ELEC TRICAL LOAD AND/OR TURBINE TRIP
15.2.7.1 Acceptance Criteria The following are the relevant specific acceptance criteria.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-28 Revision 23 December 2016 (2) The RCS pressure does not exceed 110 percent of design pressure (2,750 psia) at any time during the transient.
(3) The Main Steam System (MSS) pressure does not exceed 110 percent of design pressure (1,210 psia) at any time during the transient.
15.2.7.2 Identification of Causes and Accident Description A major load loss on the plant can result from either a loss of external electrical load or
from a turbine trip. For either case, offsite power is available for the continued operation
of plant components such as the reactor coolant pumps. The case of loss of offsite
power is analyzed in Section 15.2.9.
For a turbine trip, the reactor would be tripped directly (unless it is below the P-9
setpoint) from a signal derived from the turbine autostop oil pressure and turbine stop
valves. The automatic steam dump system accommodates 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 were not available, the excess steam generation
would be dumped to the atmosphere. Additionally, main feedwater flow would be lost if
the turbine condenser were not available. For this situation, steam generator level
would be maintained by the auxiliary feedwater (AFW) system.
For a loss of external electrical load without subsequent turbine trip, no direct reactor
trip signal would be generated. A continued steam load of approximately 5 percent would exist after total loss of external electrical load because of the electrical demand of plant auxiliaries.
In the event the 10 percent atmospheric 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
coolant temperatures will increase rapidly. The pressurizer safety valves (PSVS) and
steam generator safety valves are, however, sized to protect the 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, automatic RCCA control, or direct reactor trip on turbine trip.
The steam generator safety valve capacity is sized to remove the steam flow at the
engineered safeguards design rating (105 percent of steam flow at rated power) from
the steam generator without exceeding 110 percent of the steam system design pressure. The pressurizer safety valve capacity is sized based on a complete loss of
heat sink with the plant initially operating at the maximum calculated turbine load along
with operation of the steam generator safety valves. The pressurizer safety valves are DCPP UNITS 1 &
2 FSAR UPDATE 15.2-29 Revision 23 December 2016 then able to maintain the RCS pressure within 110 percent of the RCS design pressure without direct or immediate reactor trip action.
A more complete discussion of overpressure protection can be found in Reference 8.
15.2.7.3 Analysis of Effects and Consequences In this analysis, the behavior of the unit is evaluated for a complete loss of steam load
from full power without a direct reactor trip. This is done to show the adequacy of the pressure-relieving devices and to demonstrate core protection margins. The reactor is
not tripped until conditions in the RCS result in a trip. The turbine is assumed to trip
without actuating all the turbine stop valve limit switches. This assumption delays
reactor trip until conditions in the RCS result in a trip due to other signals. Thus, the
analysis assumes a worst case transient. In addition, no credit is taken for steam dump
actuation. Main feedwater flow is terminated at the time of turbine trip, with no credit taken for auxiliary feedwater (except for long-term recovery) to mitigate the consequences of the transient.
Total loss-of-load transients are analyzed for DNB and overpressure concerns. The
LOFTRAN computer program (refer to Section 15.1) is used to analyze the total loss of load transients for the DNB concern. The RETRAN-02 computer program (refer to Section 15.1) is used to analyze the transients for the overpressure concern. Both
programs simulate the neutron kinetics, RCS, pressurizer, pressurizer relief and safety
valves, pressurizer spray, steam generator, and steam generator safety valves. The
programs compute pertinent variables, including temperatures, pressures, and power
level. The following assumptions are used in the LOFTRAN analysis for the DNB concern.
(1) Initial Operating Conditions The accident is analyzed using the Improved Thermal Design Procedure and the initial condition uncertainties are included in the limit DNBR as
described in Reference 5. Therefore, initial conditions of nominal core
power, nominal reactor coolant average temperature (including 2.5°F for
steam generator tube fouling) and nominal reactor coolant pressure are
assumed. (2) Moderator and Doppler Coefficients of Reactivity The turbine trip is analyzed with both maximum and minimum reactivity
feedback. The maximum feedback for end of life (EOL) cases assume a
large negative moderator temperature coefficient and the most negative
Doppler power coefficient. The minimum feedback for beginning of life (BOL) cases assume a minimum moderator temperature coefficient and
the least negative Doppler coefficient.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-30 Revision 23 December 2016 (3) Reactor Control From the standpoint of the maximum pressures attained, it is conservative to assume that the reactor is in manual contr ol. If the reactor were in automatic control, the control rod banks would move prior to trip and
reduce the severity of the transient.
(4) Steam Release No credit is taken for the operation of the steam dump system or steam generator power-operated relief valves. The steam generator pressure
rises to the safety valve setpoint where steam release through safety
valves limits secondary steam pressure at the setpoint value.
(5) Pressurizer Spray and Power-Operated Relief Valves Two cases for both BOL and EOL are analyzed using the LOFTRAN
computer program.
(a) 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 operable.
(b) No credit is taken for t he effect of pressurizer spray and power-operated relief valves in reducing or limiting the coolant pressure. Safety valves are operable.
(6) 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; however, the auxiliary feedwater
pumps would be expected to start on a trip of the main feedwater pumps.
The auxiliary feedwater flow would remove core decay heat following plant
stabilization.
The following assumptions are used in the RETRAN-02 analysis for the overpressure
concern only.
(1) Initial Operating Conditions
The accident analysis assumes: maximum core power; maximum Tavg, and minimum operating RCS pressure 2189.7 psia is used in the analysis, which includes 60 psi uncertainty.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-31 Revision 23 December 2016 (2) Moderator and Doppler Coefficients of Reactivity BOL minimum reactivity feedback is modeled assuming the most positive
moderator temperature coefficient and the least negative Doppler
temperature coefficient.
(3) Steam Release
No credit is taken for secondary h eat removal from the steam dump system. Only the pressurizer safety valves and the main steam safety
valves are credited for overpressure protection.
(4) Pressurizer Pressure Control
Since the total loss of load overpressure transients result in higher peak
RCS and steam generator pressures at BOL, two cases are analyzed
using the RETRAN-02 computer program for BOL only.
(a) For the peak secondary side pressure case assumes pressurizer pressure control. This delays the reactor trip and maximizes the
heat transfer to the steam generators. Safety valves are also
(b) For the peak RCS pressure case, no credit is taken for pressurizer pressure control, which maximizes the peak RCS pressure for the event. Safety valves are operable.
(5) Feedwater Flow
The turbine stop valves and feedwater control valves are assumed to
close instantaneously at the initiation of the event to maximize the duration
of the primary to secondary heat imbalance. The auxiliary feedwater
system is conservatively not credited and assumed unavailable for decay
heat removal during the event.
(6) Main Steam Safety Valves The MSSV setpoints are assumed to be at their maximum 3 percent drift
values. The RETRAN MSSVs are modeled to provide zero to full flow as a
linear function from the lift setpoint to the full open 3 percent accumulation
value.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-32 Revision 23 December 2016 (7) Pressurizer Safety Valve Water Loop Seal All pressurizer safety valves have been converted to a steam-seat design
and condensate in the loop is now continuously drained back to the
pressurizer, thereby eliminating the water loop seal. Even though the
water loop seal has been eliminated, the resulting benefit is not credited in
the analysis.
The presence of a water loop seal delays the opening of the pressurizer safety valve. The loop seal water starts to leak out from the
safety valve when the safety valve setpoint is reached. However, no
pressure is relieved from the pressurizer until the loop seal water is
completely purged, after which the safety valve pops full open in less than
0.1 second. The loop seal water purge time of 1.272 seconds was used in
the analysis.
(8) Maximize Reactor Power It is conservative to maximize the reactor power. Therefore, the reactor trip due to high neutron flux is not credited in the analysis.
In all cases for DNB and overpressure concerns reactor trip is actuated by the first
reactor protection system trip setpoint reached with no credit taken for the direct reactor
trip on the turbine trip.
15.2.7.4 Results The transient responses for a total loss of load from full power operation are shown for four cases the DNB concern is evaluated at BOL and EOL with pressure control since this condition bounds the cases without pressure control and overpressure concern is
evaluated at BOL with and without pressure control. Refer to Figures 15.2.7-1 through
15.2.7-4 and Figures 15.2.7-9 through 15.2.7-12.
Figures 15.2.7-1 and 15.2.7-2 show the transient responses for the total loss of steam
load at BOL, for the DNB concern, assuming full credit for the pressurizer spray and
pressurizer power-operated relief valves. No credit is taken for the steam dump. The
reactor is tripped by the high pressurizer pressure trip channel. The minimum DNBR is (thimble cell) well above the limit value.
Figures 15.2.7-3 and 15.2.7-4 show the responses for the total loss of load at EOL, for
the DNB concern, assuming a large (absolute value) negative moderator temperature
coefficient. All other plant parameters are the same as in the above case. As a result
of the maximum reactivity feedback at EOL, no reactor protection system trip setpoint is reached. Because main feedwater is assumed to be lost, the reactor is tripped by the
low-low steam generator water level trip channel. The DNBR (thimble cell) increases
throughout the transient and never drops below its initial value. The pressurizer safety
valves are not actuated in these transients.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-33 Revision 23 December 2016 Figures 15.2.7-9 and 15.2.7-10 show the typical transient responses for the total loss of load at BOL for the RCS overpressure concern. No credit is taken for the pressurizer
spray, pressurizer power-operated relief valves, or steam dump. The pressurizer and
main steam safety valves are modeled as described in assumptions 6 and 7. The initial
pressurizer pressure includes the pressurizer pressure uncertainty to maximize the
peak pressure. The reactor is tripped on the high pressurizer pressure signal. This
case results in the highest RCS peak pressure among all cases. The peak RCS
pressure is below 110 percent of the design value.
Figures 15.2.7-11 and 15.2.7-12 show the typical transient responses for the total loss
of load at BOL for the secondary side overpressure concern, assuming full credit for the
pressurizer spray and the pressurizer power-operated relief valves. No credit is taken
for the steam dump. The models for the pressurizer and main steam safety valves and the initial pressurizer pressure are the same as those used in the above case. The
reactor trip due to high neutron flux is not credited in order to maximize the peak steam
generator pressure. The reactor is tripped on the high pressurizer pressure signal. This
case results in the highest steam generator peak pressure among all cases. The peak
steam generator pressure is below 110 percent of the design value.
Reference 8 presents additional results for a complete loss of heat sink including loss of
main feedwater. This report shows the overp ressure protection that is afforded by the pressurizer and steam generator safety valves.
Technical Specification 3.7.1 establishes reduced plant operating power limits for off
normal conditions when one or more MSSVs are inoperable to ensure a loss of load
event does not result in overpressurization of the steam generators. When two or more MSSVs are inoperable per steam generator loop, the reduced power limits are established using a conservative energy balance algorithm established in the
Westinghouse Nuclear Safety Advisory Letter NSAL-94-001 as documented in
Reference 21. To evaluate off normal plant operation with a single inoperable MSSV on
one or more steam generator loops, an additional spectrum of loss of load analyses are
performed as documented in Reference 22. These analyses use the RETRAN-02W
code to analyze the BOL loss of load overpressure case as discussed in this section
and which represents the limiting case for challenging the steam generator peak
pressure limit. These analysis results, as summarized in the Technical Specification
rip to demonstrate that the specified reduced operating power limit ensures that t he available relief capacity with one inoperable MSSV per loop maintains the peak steam generator pressure below 110
percent of the design value.
15.2.7.5 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell). The accompanying DNBR figures for this event DCPP UNITS 1 &
2 FSAR UPDATE 15.2-34 Revision 23 December 2016 (Figures 15.2.7-1 and 15.2.7-3) reflect the results for the more limiting Standard fuel (limit 1.48/1.44) previously in the core.
(2) The calculated RCS pressure (2723 psia) does not exceed 110 percent of design pressure (2,750 psia) at any time during the transient.
(3) The calculated MSS peak pressure (1203 psia) does not exceed 110 percent of design pressure (1,210 psia) at any time during the
Results of the analyses, including those in Reference 8, show that the plant design is
such that a total loss of external electrical load without a direct or immediate reactor 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 safety analysis limit values. Thus, no core safety limit will be violated.
15.2.8 LOSS OF NORMAL FEEDWATER 15.2.8.1 Acceptance Criteria The following is the relevant specific acceptance criterion.
(1) The pressurizer does not go water solid at any time during the transient.
15.2.8.2 Identification of Causes and Accident Description A loss of normal feedwater (resulting from pump failures, valve malfunctions, or loss of
offsite ac power) results in a reduction in the abil ity of the secondary system to remove the heat generated in the reactor core. If the reactor were not tripped during this accident, core damage would possibly occur from a sudden loss of heat sink. If an
alternative supply of feedwater were not supplied to the SGs, residual heat following
reactor trip would heat the primary system water to the point where water relief from the pressurizer would occur. A significant loss of water from the RCS could conceivably
lead to core
damage. Since the plant is tripped well before the steam generator heat transfer
capability is reduced, the primary system conditions never approach a DNB condition.
The following provide the necessary protection against a loss of normal feedwater:
(1) Reactor trip on low-low water SG level in any steam generator (2) Two motor-driven AFW pumps that are started on:
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-35 Revision 23 December 2016 (a) Low-low SG water level in any steam generator (b) Trip of both main feedwater pumps (c) Any safety injection signal (d) Loss of offsite power (automatic transfer to diesel generators)
(e) Manual actuation (3) One turbine-driven auxiliary feedwater pump that is started on:
(a) Low-low SG water level in any two steam generators (b) Undervoltage on both reactor coolant pump buses (c) Manual actuation The motor-driven AFW pumps are connected to vital buses and are supplied by the
diesels if a loss of offsite power occurs. The turbine-driven pump utilizes steam from
the secondary system and exhausts it to the atmosphere. The controls are designed to start both types of pumps within 1 minute even if a loss of all AC power occurs
simultaneously with loss of normal feedwater.
The AFW pumps take suction from the condensate storage tank for delivery to the steam generators.
The analysis shows that following a loss of normal feedwater, the AFW system is capable of removing the stored energy and residua l decay heat, and RCP heat thus preventing either overpressurization of the RCS or liquid relief through the pressurizer
power operated relief valves (PORVs) or safety valves.
15.2.8.3 Analysis of Effects and Consequences A detailed analysis using the RETRAN-02W code (Reference 19) is performed in order
to determine the plant transient following a loss of normal feedwater. The code
describes the plant neutron kinetics, RCS including factors that influence the natural
circulation, pressurizer, steam generators, and feedwater system, and compute
pertinent variables, including the pressurizer pressure, pressurizer water level, and
reactor coolant average temperature.
Major assumptions are:
(1) Reactor trip occurs on steam generator low-low level at 8 percent of narrow range span.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-36 Revision 23 December 2016 (2) The plant is initially operating at 102 percent of the nuclear steam supply system (NSSS) rating, including a conservatively large RCP heat of 20 MWt. (3) Conservative core residual heat generation based on long-term operation at the initial power level preceding the trip is assumed. The ANSI/ANS-5.1-1979 + 2 was used for calculation of residual decay heat levels.
(4) The auxiliary feedwater system is actuated by the low-low steam generator water level signal.
(5) The limiting single failure in the auxiliary feedwater system occurs (turbine-driven pump failure). The auxiliary feedwater system is assumed
to supply a total of 600 gpm to all four SGs from the motor-driven pumps.
(6) The pressurizer sprays and heaters are assumed operable. This maximizes the peak transient pressurizer water volume. Sensitivity
analyses determined that it is conservative to assume that the PORVs are
inoperable (Reference 20).
(7) Secondary system steam relief is achieved through the self-actuated safety valves. The main steam safety valves are assumed to begin to lift
3 percent above the set pressure with a 5 psi accumulation to full open.
Note that steam relief will, in fact, be through the 10 percent atmospheric dump valves or 40 percent condenser dump valves for most cases of loss of normal feedwater. However, for the sake of analysis these have been
assumed unavailable.
(8) The initial reactor coolant average temperature is 5.5°F lower than the nominal value. The initial pressurizer pressure is 60 psi above the
nominal value.
(9) The minimum steam generator tube plugging (SGTP) of 0 percent was assumed. (10) The initial feedwater temperature is assumed to be 435°F.
15.2.8.4 Results Figures 15.2.8-1 through 15.2.8-3 show plant parameters following a loss of normal
feedwater at the conditions associated with Unit 2, which were determined to be limiting
when compared to Unit 1. Figure 15.2.8-2 shows the pressurizer pressure as a function
of time.
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 DCPP UNITS 1 &
2 FSAR UPDATE 15.2-37 Revision 23 December 2016 steam flow through the safety valves continues to dissipate the stored and generated heat. One minute following the initiation of th e low-low SG level trip, the motor-driven AFW pumps are automatically started, reducing the rate of water level decrease.
The capacity of the motor-driven AFW pumps combined with the available secondary inventory is capable of dissipating the core residual heat without liquid water relief from
the RCS PORVs or safety valves.
From Figure 15.2.8-2 it can be seen that at no time is there liquid relief from the
pressurizer. If the AFW delivered is greater than that of two motor-driven pumps, the initial reactor power is less than 102 percent of the NSSS rating, or the steam generator
water level in one or more steam generators is above the low-low level trip point at the
time of trip, then the results for this transient will be less limiting.
The calculated sequence of events for this accident is listed in Table 15.2-1. As shown in Figures 15.2.8-1 through 15.2.8-3, the plant approaches a stabilized condition
following reactor trip and AFW initiation.
Plant procedures may be followed to further cool down the plant.
15.2.8.5 Conclusions The analysis demonstrates that the acceptance criterion is met as follows:
(1) The pressurizer does not become water solid during the event as shown in Figure 15.2.8-2.
Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system, since the AFW capacity is such that the
pressurizer does not become water solid, which ultimately precludes reactor coolant
liquid relief from the pressurizer relief or safety valves. This ensures a Condition III or
IV event will not be generated.
15.2.9 LOSS OF OFFSITE POWER TO THE STATION AUXILIARIES 15.2.9.1 Acceptance Criteria The following is the relevant specific acceptance criterion.
(1) The pressurizer does not go water solid at any time during the transient.
15.2.9.2 Identification of Causes and Accident Description During a complete loss of offsite power and a turbine trip there will be loss of power to
the plant auxiliaries, i.e., the reactor coolant pumps, condensate pumps, etc.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-38 Revision 23 December 2016 The events following a loss of AC power with turbine and reactor trip are described in the sequence listed below:
(1) Plant vital instruments are supplied by emergency power sources.
(2) As the steam system pressure rises following the trip, the steam system power-operated relief valves are automatically opened to the atmosphere.
Steam dump to the condenser is assumed not to be available. If the
power-operated relief valves are not available, the steam generator
self-actuated safety valves may lift to dissipate the sensible heat of the
fuel and coolant plus the residual heat produced in the reactor.
(3) As the no-load temperature is approached, the steam system power-operated relief valves (or the self-actuated safety valves, if the
power-operated relief valves are not available) are used to dissipate the
residual heat and to maintain the plant at the hot standby condition.
(4) The emergency diesel generators started on loss of voltage on the plant emergency buses begin to supply plant vital loads.
The AFW system is started automatically as discussed in the loss of normal feedwater
analysis. The steam-driven auxiliary feedwater pump utilizes steam from the secondary
system and exhausts to the atmosphere. The motor-driven AFW pumps are supplied
by power from the diesel generators. The pumps take suction directly from the
condensate storage tank for delivery to the steam generators.
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.
15.2.9.3 Analysis of Effects and Consequences A detailed analysis using the RETRAN-02W code (Reference 19) is performed in order
to determine the plant transient following loss of offsite power. The code describes the plant neutron kinetics, RCS including factors that influence the natural circulation, pressurizer, steam generators, and feedwater system, and computes pertinent
variables, including the pressurizer pressure, pressurizer water level, and reactor
coolant average temperature.
Major assumptions differing from those in a loss of normal feedwater are:
(1) No credit is taken for immediate response of control rod drive mechanisms caused by a loss of offsite power.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-39 Revision 23 December 2016 (2) RCP coastdown to natural circulation conditions is assumed after reactor trip (i.e., rod motion), which is more limiting for long-term heat removal capability.
(3) The initial feedwater temperature is assumed to be 425°F.
(4) A nominal reactor coolant pump heat input of 14 MWt.
15.2.9.4 Results The time sequence of events for the accident at the conditions associated with Unit 2, which were determined to be limiting, is given in Table 15.2-1. This event is bounded
by the complete-loss-of-flow analysis (Section 15.3.4), in terms of minimum DNBR (Reference 23). Therefore, this event is not analyzed for DNB concerns, but rather, for the long-term heat removal capability. Afte r the reactor trip, stored and residual heat must be removed to prevent damage to either the RCS or the core. The RETRAN-02W
code results show that the natural circulation flow available is sufficient to provide
adequate core decay heat removal following reactor trip and RCP coastdown.
Figures 15.2.9-1 through 15.2.9-3 show plant parameters following a loss of offsite power at the conditions associated with Unit 2, which were determined to be limiting.
Figure 15.2.9-2 shows the pressurizer water volume as a function of time.
15.2.9.5 Conclusions
The analysis demonstrates that the acceptance criterion is met as follows:
(1) The pressurizer does not become water solid during the event as shown in Figure 15.2.9-2.
Results of the analysis show that, for the loss of offsite power to the station auxiliaries
event, all safety criteria are met. Since the DNBR remains above the safety analysis limit, the core is not adversely affected. AFW capacity is sufficient to prevent the
pressurizer from becoming water solid, which ultimately precludes reactor coolant liquid
relief from the pressurizer relief and safety valves; this assures that the RCS is not
overpressurized. This ensures that a Condition III or IV event will not be generated.
Analysis of the natural circulation capability of the RCS demonstrates that sufficient
long-term heat removal capability exists following reactor coolant pump coastdown to
prevent fuel or cladding damage.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-40 Revision 23 December 2016 15.2.10 EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTIONS 15.2.10.1 Acceptance Criteria The following are the relevant specific acceptance criteria:
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient.
(2) The peak linear heat generation rate does not exceed a value which would cause fuel centerline melt at any time during the transient (refer to
Section 4.4.3.2.7).
15.2.10.2 Identification of Causes and Accident Description Reductions in feedwater temperature or excessive feedwater additions are means of
increasing core power above full power. Such transients are attenuated by the thermal
capacity of the secondary plant and of the RCS. The overpower-overtemperature protection (neutron high flux, overtemperature T, and overpower T trips) prevent any power increase that could lead to a DNBR that is less than the DNBR limit.
One example of excessive feedwater flow would be a full opening of a main feedwater regulating valve (MFRV) 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 generator.
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.
Continuous excessive feedwater addition is prevented by the steam generator high-high
level feedwater isolation and turbine trip.
15.2.10.3 Analysis of Effects and Consequences The excessive heat removal due to a feedwater system malfunction transient is
analyzed with the RETRAN-02W code. 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 evaluate plant behavior in the event of a feedwater system
malfunction. The accident is analyzed with the Improved Thermal Design Procedure
and initial condition uncertainties are include d in the limit DNBR as described in Reference 5. Therefore, initial conditions of nominal core power, nominal reactor
coolant average temperatures and nominal reactor coolant pressure are assumed.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-41 Revision 23 December 2016 Excessive feedwater addition due to a control system malfunction or operator error that allows a MFRV to open fully is considered. Two conditions are evaluated as follows:
(1) Accidental opening of one MFRV with the reactor just critical at zero load conditions. The feedwater flow increase event at hot zero power
conditions is not limiting with respect to departure from nucleate boiling
concerns and is bounded by the full power event; therefore, the event has
not been explicitly analyzed.
(2) Accidental opening of one MFRV at full power (with automatic and manual rod control).
The reactivity insertion rate following a feedwater system malfunction is calculated with
the following assumptions:
(1) For the MFRV accident at full power, one MFRV is assumed to malfunction resulting in a step increase to 250 percent of nominal
feedwater flow to one steam generator.
(2) Coincident with the feedwater flow increase in the faulted loop, the feedwater temperature in all loops decreases approximately 23°F from the
nominal full power value. This accounts for the effect of the feedwater
passing through the heaters at a higher velocity.
(3) The initial water level in all the steam generators is at a conservatively low level. (4) No credit is taken for the heat capacity of the RCS and steam generator thick metal in attenuating the resulting plant cooldown.
(5) The feedwater flow resulting from a fully open control valve is terminated by the steam generator high-high level signal that closes all MFRVs, closes all feedwater bypass valves, trips the main feedwater pumps, and
shuts the main feedwater isolation valves (MFIVs).
15.2.10.4 Results The full power case (EOL, with manual rod control) gives the largest reactivity feedback
and results in the greatest power increase. A turbine trip and reactor trip is actuated
when the steam generator level reaches the high-high level setpoint. Although turbine
trip and subsequent reactor trip are assumed, the results show that the DNBR remains
relatively constant prior to the time of reactor trip. This demonstrates that a reactor trip on turbine trip is not needed to protect against DNB, but is assumed as a means to
terminate the transient.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-42 Revision 23 December 2016 Transient results (refer to Figures 15.2.10-1 through 15.2.10-3) show the core heat flux, pressurizer pressure, core T avg , and DNBR (thimble cell), as well as the increase in nuclear power and loop T associated with the increased thermal load on the reactor.
Steam generator level rises until the feedwater is terminated as a result of the high-high
steam generator level trip. The DNBR does not drop below the limit safety analysis DNBR.
15.2.10.5 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) as shown on Figure 15.2.10-3.
(2) Figure 15.2.10-1 shows the total power is maintained below 118 percent of its nominal value. Thus the peak fuel centerline temperature will remain
below the fuel melting temperature (refer to Section 4.4.3.2.7).
An excessive feedwater addition at no-load conditio ns is bounded by the analysis at full power. The DNBRs encountered for excessive feedwater addition at power are well
above the safety analysis limit DNBR values.
15.2.11 SUDDEN FEEDWATER TEMPERATURE REDUCTION
15.2.11.1 Acceptance Criteria The following are the relevant specific acceptance criteria.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient.
(2) The peak linear heat generation rate does not exceed a value that would cause fuel centerline melt at any time during the transient (refer to Section 4.4.3.2.7).
15.2.11.2 Identification of Causes and Accident Description A concern was raised during the Unit 1 power ascension test program that an
inadvertent actuation of the load transient bypass relay (LTBR) might initiate a transient
that exceeds analyzed reactor operating li mits. An evaluation performed showed that since the expected feedwater temperature decrease due to inadvertent actuation of the
LTBR was significantly less than that of the net load trip, the consequences and events
of inadvertent actuation of the LTBR were bounded by the feedwater temperature
decrease event.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-43 Revision 23 December 2016 The automatic load transient bypass (LTB) feature has been eliminated for Units 1 and 2. Control of the feedwater heater bypass valve has been changed to manual only.
A reduction in feedwater temperature may be caused by an inadvertent manual opening
of the feedwater heater bypass valve. This would divert flow around the low pressure
feedwater heaters. A consequent maximum 70°F reduction in feedwater temperature to
the steam generators would occur.
Feedwater temperature may also be reduced during a load rejection trip. The feedwater
transient data taken from a 100 percent net load trip test with LTB active showed that a
maximum feedwater temperature decrease of 230°F occurred over a 400-second time
period. The temperature decrease without LTB is significantly less.
Reductions in temperature of feedwater entering the steam generators, if not
accompanied by a corresponding reduction in steam flow, would result in an increase in
core power and create a greater load demand on the RCS. The net effect on the RCS
of a reduction in reactor coolant temperature is similar to the effect of increasing
secondary steam flow. Such transients are attenuated by the thermal capacity of the
less than the limit value. The reactor may reach a new equilibrium condition at a power
results in only a small increase in reactor power and do es not result in a reactor trip. A larger temperature reduction produces a larger increase in reactor power and may
cause a power/temperature mismatch and a reactor trip.
15.2.11.3 Analysis of Effects and Consequences The accident is analyzed with the Improved Thermal Design Procedure and initial condition uncertainties are included in the limit DNBR as described in Reference 5.
Therefore, initial conditions of nominal core power, nominal reactor coolant average
temperatures and nominal reactor coolant pressure are assumed.
15.2.11.3.1 Temperature Drops Less than 73
°F
The protection available to mitigate the consequences of a decrease in feedwater
temperature is the same as that for an excessive increase in steam flow event, as
discussed in Section 15.2.12. A step load increase of 10 percent from full load was
analyzed, and the minimum DNBR for this event was found to be above the safety
analysis limit values.
The increase in heat load resulting from a 10 percent increase in load is equivalent to a
73°F drop in feedwater temperature at the steam generator inlet. Thus a feedwater
temperature transient that results in a feedwater temper ature drop of 73°F or less at the steam generator inlets is less severe than the excessive load increase incident
presented in Section 15.2.12 and as such does not exceed any safety limits.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-44 Revision 23 December 2016 15.2.11.3.2 Temperature Drops Greater than 73
°F To address feedwater temperature reductions that exceed 73°F, analyses were
performed assuming instantaneous temperature drops of 175°F and 250°F at the steam generator, with corresponding steam load reductions of 50 percent and 100 percent, respectively. The maximum temperature drop of 250°F was chosen to bound the
temperature decrease of 230°F experienced during the net load trip test when the LTBR
was actuated in response to a load reduction.
In this test, feedwater temperature dropped approximately 230°F over a time period of 400 seconds, which is significantly
less severe than the instantaneous drop of 250°F assumed in the analysis. Since LTB
has been eliminated, the feedwater temperature drop will be significantly less and is
bounded by the instantaneous drop of 250°F assumed in the analysis.
15.2.11.4 Results Both a 175°F feedwater temperature reduction concurrent with a 50 percent load
reduction and a 250°F feedwater temperature reduction concurrent with a full
(100 percent) load reduction were analyzed. The analysis shows that the cooldown
effects of the large feedwater reduction are more than counteracted by the reduced heat
removal resulting from the turbine load reduction, such that the transient causes a
heatup of the RCS. As a result, the core power decreases and the DNBR increases
during the transient. These cases do not challenge core thermal limits.
15.2.11.5 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell).
(2) The total power is maintained below 118 percent of its nominal value.
Thus the peak fuel centerline temperature will remain below the fuel
melting temperature (refer to Section 4.4.3.2.7).
All safety criteria are met for credible scenarios of sudden feedwater temperature
reduction. Instantaneous feedwater temperature reductions up to 73°F result in an RCS
cooldown that is bounded by the analysis of an excessive load increase incident presented in Section 15.2.12. This bounds the maximum feedwater temperature
decrease of 70°F that could result from the inadvertent opening of a feedwater heater
bypass valve. For feedwater temperature reductions during a load reduction transient, analyses conclude that these cases result in a net RCS heatup and core power
decrease, with no significant challenge to the core thermal limits.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-45 Revision 23 December 2016 15.2.12 EXCESSIVE LOAD INCRE ASE INCIDENT 15.2.12.1 Acceptance Criteria The following is the relevant specific acceptance criterion.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient.
15.2.12.2 Identification of Causes and Accident Description An excessive load increase inci dent is defined as a rapid increase in the 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 percent
step-load increase or a 5 percent per minute ramp load increase in the range of 15 to
100 percent of full power. Any loading rate in excess of these values may cause a
reactor trip actuated by the reactor protection system.
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.
Protection against an excessive load increase accident is provided by the following reactor protection system signals:
(1) Overpower (2) Overtemperature (3) Power range high neutron flux 15.2.12.3 Analysis of Effects and Consequences This accident is analyzed using the LOFTRAN code. The code simulates the neutron
kinetics, RCS, 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.
Four cases are analyzed to demonstrate the plant behavior following a 10 percent step
load increase from rated load. These cases are as follows:
(1) Reactor control in manual with BOL minimum moderator reactivity feedback (2) Reactor control in manual with EOL maximum moderator reactivity feedback DCPP UNITS 1 &
2 FSAR UPDATE 15.2-46 Revision 23 December 2016 (3) Reactor control in automatic with BOL minimum moderator reactivity feedback (4) Reactor control in automatic with EOL maximum moderator reactivity feedback For the BOL minimum moderator feedback cases, the core has the least negative
moderator temperature coefficient of reactivity and the least negative Doppler only
power coefficient curve; therefore the least inherent transient response capability. For
the EOL maximum moderator feedback cases, the moderator temperature coefficient of
reactivity has its highest absolute value and the most negative Doppler only power
coefficient curve. 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 being taken for pressurizer heaters.
The accident is analyzed using the Improved Thermal Design Procedure and the initial condition uncertainties are included in the limit DNBR as described in Reference 5.
Therefore, initial conditions of nominal core power, reactor coolant average temperature (plus 2.5°F for steam generator fouling) and nominal reactor coolant pressure are
assumed.
Plant characteristics and initial conditions ar e further discussed in Section 15.1.
Normal reactor control systems and engineered safety systems are not required to function. The reactor protection system is a ssumed to be operable; however, reactor trip is not encountered for most 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 function is not required.
15.2.12.4 Results The calculated sequence of events for the excessive load increase incident is shown in Table 15.2-1.
Figures 15.2.12-1 through 15.2.12-4 illustrate the transient with the reactor in the manual
control mode. As expected, for the BOL minimum moderator 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 EOL maximum moderator feedback manually controlled case, there is a much larger increase in reactor DCPP UNITS 1 &
2 FSAR UPDATE 15.2-47 Revision 23 December 2016 power due to the moderator feedback. A re duction in DNBR is experienced but DNBR remains above the limit value.
Figures 15.2.12-5 through 15.2.12-8 illustrate the transient assuming the reactor is in
the automatic control mode. Both the BOL minimum and EOL maximum moderator
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.
For all cases, the plant rapidly reaches a stabilized condition at the higher power level.
Normal plant operating procedures would then be followed to reduce power.
The excessive load increase incident is an overpower transient for which the fuel
temperatures will rise. Reactor trip does not occur for any 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 DNB is not predicted to occur 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.
15.2.12.5 Conclusions The analysis demonstrates that the acceptance criterion is met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell). The accompanying DNBR figures for this event (Figures 15.2.12-2, 15.2.12-4, 15.2.12-6 and 15.2.12-8) reflect the results for the more limiting Standard fuel (limit 1.48/1.44) previously in the core.
The figures represent the thimble cell results.
The analysis presented above shows that for a 10 percent step load increase, the
DNBR remains above the safety analysis li mit values, thereby precluding fuel or cladding damage. The plant reaches a stabil ized condition rapidly, following the load increase.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-48 Revision 23 December 2016 15.2.13 ACCIDENTAL DEPR ESSURIZATION OF THE REACTOR COOLANT SYSTEM 15.2.13.1 Acceptance Criteria The following is the relevant specific acceptance criterion.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient.
15.2.13.2 Identification of Causes and Accident Description An accidental depressurization of the RCS 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 flowrate 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 RCS are associated with an inadvertent
opening of a pressurizer safety valve. Initially, the event results in a rapidly decreasing
RCS pressure, which could reach the hot leg saturation pressure if a reactor trip does
not occur. The pressure continues to decrease throughout the transient. The effect of the pressure decrease is to decrease the neutron flux via the moderator density
feedback, but the reactor control system (if in the automatic mode) functions to maintain
the power and average coolant temperature essentially constant until the reactor trip
occurs. Pressurizer level increases initia lly due to expansion caused by depressurization and then decreases following reactor trip.
The reactor will be tripped by the following reactor protection system signals:
(1) Pressurizer low pressure (2) Overtemperature 15.2.13.3 Analysis of Effects and Consequences The accidental depressurization transient is analyzed with the LOFTRAN code. The
code simulates the neutron kinetics, RCS, 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 Improved Thermal Design Procedure and the initial
condition uncertainties are included in the li mit DNBR as described in Reference 5.
Therefore, initial conditions of nominal core power, nominal reactor coolant average
temperature (plus 2.5 °F for steam generator fouling) and nominal reactor coolant
pressure are assumed.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-49 Revision 23 December 2016 In order to obtain conservative results, the following assumptions are made:
(1) A positive moderator temperature coefficient of reactivity for (+
7 pcm/°F)
BOL operation is assumed in order to provide a conservatively high
amount of positive reactivity feedback due to changes in moderator
temperature. The spatial effect of voids due to local or subcooled boiling
is not considered in the analysis with respect to reactivity feedback or core
power shape. These voids would tend to flatten the core power
distribution.
(2) A low (absolute value) Doppler-only power coefficient of reactivity such that the resultant amount of negative feedback is conservatively low in
order to maximize any power increase due to moderator reactivity
feedback.
15.2.13.4 Results Figure 15.2.13-1 illustrates the nuclear power transient following the RCS
depressurization accident. The nuclear power increases until the time reactor trip
The time of reactor trip is shown in Table 15.2-1. The pressure decay transient
following the accident is given in F igure 15.2.13-2. The resulting DNBR (thimble cell) never goes below the safety analysis limit value as shown in Figure 15.2.13-1.
15.2.13.5 Conclusions The analysis demonstrates that the acceptance criterion is met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) shown on Figure 15.2.13-1.
signals provide adequate protection against this accident, and the minimum DNBR
remains in excess of the safety analysis limit value.
15.2.14 ACCIDENTAL DEPR ESSURIZATION OF THE MAIN STEAM SYSTEM
15.2.14.1 Acceptance Criteria The following is the relevant specific acceptance criterion.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-50 Revision 23 December 2016 (1) Minimum DNBR is not less than the applicable Safety Analysis Limit of 1.45 (W-3 DNB correlation, for coolant pressure less than 1,000 psia) at any time during the transient.
15.2.14.2 Identification of Causes and Accident Description The most severe core conditions resulting from an accidental depressurization of the
main steam system are associated with an inadvertent opening of a single steam dump, relief, or safety valve. The analyses, assuming a rupture of a main steam pipe, are discussed in Section 15.4.
The steam released as a consequence of this accident results in an initial increase in
steam flow that 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 a
reduction of core shutdown margin.
The analysis is performed to demonstrate that the following criterion is satisfied:
Assuming a stuck RCCA and a single failure in the engineered safety features (ESF)
the limit DNBR value will be met after reactor t rip for a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, relief, or safety valve.
The following systems provide the necessary mitigation of an accidental
depressurization of the main steam system.
(1) SIS actuation from any of the following:
(a) Two-out-of-four low pressurizer pressure signals
(b) Two-out-of-three low steam line pressure signals on any one loop (2) The overpower reactor trips (neutron flux and reactor trip, and the reactor trip occurring in conjunction with receipt of the safety injection signal.
(3) Redundant isolation of the main feedwater lines: Sustained high feedwater flow would cause additiona l cooldown. Therefore, a safety injection signal will rapidly close all MFRVs, trip the main feedwater pumps, and close the MFIVs.
15.2.14.3 Analysis of Effects and Consequences Due to the size of the break and the assumed initial conditions, an Accidental Depressurization of the Main Steam System event is bounded by the Main Steam Line
Rupture accident analyzed in Section 15.4.2.1. As such, no explicit analysis is DCPP UNITS 1 &
2 FSAR UPDATE 15.2-51 Revision 23 December 2016 performed for the Accidental Depressurization of the Main Steam System. All applicable acceptance criteria are shown to be met via the results and conclusions in
Section 15.4.2.1.
15.2.14.4 Conclusions The applicable acceptance criterion is shown to be met via the results and conclusions in Section 15.4.2.1. The Rupture of a Main Steam Line at Hot Zero Power is a
Condition IV event that meets the minimum DNBR criterion of the Condition II Accidental Depressurization of the Main Steam System. The Condition IV event models
a large double-ended rupture, with more limiting results than the inadvertent opening of
a SG PORV or dump valve, regardless of any differences in safety system actuations.
Therefore, demonstrating that the Condition IV event meets the same Condition II
criterion for minimum DNBR, as the calculated value is never below the limit value of
1.45 at any time during the transient, satisfies the acceptance criterion for the
Condition II event discussed in Section 15.2.14.
15.2.15 SPURIOUS OPERATION OF THE SAFETY INJECTION SYSTEM AT POWER 1 5.2.15.1 Acceptance Criteria The following are the relevant specific acceptance criteria.
(1) The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell) (refer to Section 4.4.4.1) at any time during the transient. This criterion is discussed in Section 15.2.15.2.
(2) The SSI event is terminated prior to exceeding the assumed maximum pressurizer PORV cycles supported by the backup nitrogen accumulators
and prior to challenging the PSV l iquid relief capability. This criterion is discussed in Section 15.2.15.3.
The RCS overpressure concern is bounded by the Section 15.2.7 event.
15.2.15.2 Spurious Safety Injection (SSI) DNBR Analysis 15.2.15.2.1 Identification of Causes and Accident Description Spurious SIS operation at power could be caused by operator error or a false electrical
actuating signal. A spurious signal may originate from any of the safety injection
actuation channels. Refer to Section 7.2 for a description of the actuation system.
Following the actuation signal, the suction of the coolant charging pumps is diverted
from the volume control tank to the RWST. The charging injection valves between the
charging pumps and the injection header open automatically. The charging pumps then DCPP UNITS 1 &
2 FSAR UPDATE 15.2-52 Revision 23 December 2016 pump RWST water through the header and injection line and into the cold legs of each loop. The safety injection pumps also start automatically but provide no flow when the
RCS is at normal pressure. The passive injection system and the RHR system also
provide no flow at normal RCS pressure.
The analyses of the potential for DNB, loss of fuel integrity, and excessive cooldown are
presented in the discussions herein.
An SIS signal normally results in a reactor tri p followed by a turbine trip. However, it cannot be assumed that any single fault that actuates the SIS will also produce a
reactor trip. Therefore, two different courses of events are considered.
Case A: Trip occurs at the same time spurious injection starts.
Case B: The reactor protection system produces a trip later in the transient.
For Case A, the operator should determine if the spurious signal was transient or steady state in nature, i.e., an occasional occurrence or a definite fault. The operator will
determine this by following approved procedures.
In the transient case, the operator would stop the safety injection. If the SIS must be disabled for repair, boration should
continue and the plant brought to cold shutdown.
For Case B, the reactor protection system does not produce an immediate trip and the
reactor experiences a negative reactivity excursion causing a decrease in the reactor
power. The power unbalance causes a drop in T avg and consequent coolant shrinkage, and pressurizer pressure and level drop. Load will decrease due to the effect of reduced steam pressure on load if the electrohydraulic governor fully opens the turbine throttle valve. If automatic rod control is used, these effects will be lessened until the
rods have moved out of the core. The transient is eventually terminated by the reactor
protection system low-pressure trip or by manual trip.
The time to trip is affected by initial operating conditions includ ing core burnup history that affects initial boron concentration, rate of change of boron concentration, and
Doppler and moderator coefficients.
Recovery from this incident for Case B is in the same manner as for Case A. The only
difference is the lower T avg and pressure associated with the power imbalance during this transient. The time at which reactor trip occurs is of no concern for this accident.
At lighter loads coolant contraction will be slower resulting in a longer time to trip.
15.2.15.2.2 Analysis of Effects and Consequences The spurious operation of the SIS is analyzed for DNBR with the LOFTRAN program.
The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety
valves, pressurizer spray, steam generator, steam generator safety valves, and the DCPP UNITS 1 &
2 FSAR UPDATE 15.2-53 Revision 23 December 2016 effect of the SIS. The program computes pertinent plant variables including temperatures, pressures, and power level.
Because of the power and temperature reduction during the transient, operating
conditions do not approach the core limits. Analyses of several cases show that the
results are relatively independent of time to trip.
A typical transient is considered representing conditions at BOL. Results at EOL are
similar except that moderator feedback effects result in a slower transient.
The accident is analyzed using the Improved Thermal Design Procedure and the initial
condition uncertainties are included in the li mit DNBR as described in Reference 5.
Therefore, initial conditions of nominal core power, nominal reactor coolant average
temperature (plus 2.5°F for steam generator fouling) and nominal reactor coolant
pressure are assumed.
The assumptions used in the analysis are:
(1) Moderator and Doppler Coefficients of Reactivity A positive BOL moderator temperature coefficient was used. A low
absolute value Doppler power coefficient was assumed.
(2) Reactor Control The reactor was assumed to be in manual control.
(3) Pressurizer Heaters Pressurizer heaters were assumed to be inoperative in order to increase
the rate of pressure drop.
(4) Boron Injection At time zero, two charging pumps (CCP1 and CCP2) begin injection and
pump borated water through the SIS and into the cold leg of each loop.
(5) Turbine Load Turbine load was assumed constant until the electrohydraulic governor
drives the throttle valve wide open. Once the throttle valve is wide open,
turbine load drops as the steam pressure decreases.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-54 Revision 23 December 2016 (6) Reactor Trip Reactor trip was initiated by low pressure. The trip was conservatively assumed to be delayed until the pressure reached 1860 psia.
15.2.15.2.3 Results The transient response for the minimum feedback case is shown in Figures 15.2.15-1
through 15.2.15-2. Nuclear power starts decreasing immediately due to boron injection, but steam flow does not decrease until 25 seconds into the transient when the turbine
throttle valve goes wide open. The mismatch between load and nuclear power causes
Tavg, pressurizer water level, and pressurizer pressure to drop. The low-pressure trip
setpoint is reached at 23 seconds and rods start moving into the core at 25 seconds.
15.2.15.2.4 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) There is margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell).
Results of the DNBR analysis show that spurious safety injection with or without
immediate reactor trip presents no hazard to the integrity of the RCS.
DNBR is never less than the initial value. Thus, there will be no cladding damage and
no release of fission products to the RCS.
If the reactor does not trip immediately, the low-pressure reactor trip will be actuated.
This trips the turbine and prevents excess cooldown thereby expediting recovery from
the incident.
15.2.15.3 Spurious Safety Injection (SSI)
Pressurizer Overfill Analysis 15.2.15.3.1 Identification of Causes and Accident Description The causes and accident description are essentially identical for the SSI DNBR analysis
discussed in Section 15.2.15.2 and the SSI pressurizer overfill evaluation in this section.
The pressurizer overfill cases model the long term plant response and the operator
actions taken to terminate the event before the liqu id relief capability of the PSV is challenged. The operator recovery actions for SSI mitigation at power are provided in
the plant emergency operating procedures (EOPs). These operator actions, including
making a pressurizer PORV available, stopping all but one centrifugal charging pump (CCP1 or CCP2), throttling the charging flow, and establishing RCS letdown flow, are
discussed below.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-55 Revision 23 December 2016 (a) Make Pressurizer PORV Available One of the first recovery actions that the EOPs describe is to verify a pressurizer PORV is available for pressure relief. The operator is directed
to open an associated isolation valve as necessary to make a PORV
available. The pressurizer overfill evaluation assumes that the operator
makes a PORV available within 11 minutes of the initiation of the event.
(b) Stop All But One CCP (CCP1 or CCP2)
The EOPs provide direction, that in the event of a reactor trip or safety
injection, the nonsafety-related centrifugal charging pump CCP3 is not
needed and is secured. The pressurizer overfill evaluation conservatively
assumes that CCP3 is operating when the SSI event occurs, since this
maximizes the pressurizer fill rate. The operators stop the CCP3 within
9 minutes of the event initiation.
Once the operators have identified that the SI is unnecessary, the EOPs
direct the operators to stop all but one CCP (CCP1 or CCP2), and throttle
the CCP (CCP1 or CCP2) flow as necessary to minimize the potential for
pressurizer overfill while maintaining adequate RCP seal injection flow.
The operators are assumed to stop all but one CCP (CCP1 or CCP2)
within 14 minutes, and require one additional minute to throttle the
charging flow.
(c) Restore Instrument Air and Establish RCS Letdown The SI signal causes a Phase A containment isolation and a loss of
instrument air to containment. In order to establish RCS letdown and terminate the SSI event, the EOPs direct the operators to restore
instrument air to containment. The operators are assumed to restore
instrument air to containment within 21 minutes of the event. The EOPs
then direct the operators through a series of steps, which allow them to
establish RCS letdown and stabilize the pressurizer level. The operators
are able to establish RCS letdown and terminate the SSI event within
26 minutes.
There are three different cases analyzed to bound the potential impact of the plant
control systems operation on the SSI event and the potential for pressurizer overfill.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-56 Revision 23 December 2016 Case 1 Case 1 assumes that the pressurizer pressure control system malfunctions such that the sprays, backup heaters, and proportional heaters all remain on during the event.
Both Class 1 PORVs are unavailable. Case 1 establishes the maximum time available for the operators to open a pressurizer PORV block valve and make a PORV available, before the liquid relief capability of the PSV is challenged. The PSV capability is defined as a maximum of 3 openings under liquid relief conditions with the liquid
temperature remaining greater than 613°F as established in Reference 16.
Case 2 Case 2 assumes that the pressurizer pressure control system malfunctions such that the sprays, backup heaters, and proportional heaters all remain on during the event.
This case causes the earliest filling of the pressurizer and the earliest initiation of liquid
relief through the pressurizer PORV. This case evaluates that the minimum capacity of
the backup nitrogen accumulators is adequate to allow termination of the SSI event
without challenging the liquid relief capability of the PSV.
Case 3 Case 3 assumes there is a loss of instrument air such that the pressurizer sprays are not operable. The pressurizer heaters remain on during the event. This case causes the earliest pressure increase to the PORV lift setpoint. The analyses of Cases 2 and 3
establish the bounding conditions for evaluating the potential impact of the pressurizer
control systems on the time at which the pressurizer fills and the relative number of steam relief and liquid relief PORV cycles which occur during an SSI event.
15.2.15.3.2 Analysis of Effects and Consequences The SSI event is analyzed for pressurizer overfill conditions with the RETRAN-02
program. The code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, steam generator safety valves, and the effect of the SSI. The program computes pertinent plant variables including
temperatures, pressures, and power level.
The assumptions are:
(1) Initial Operating Conditions The initial pressurizer pressure is assumed to be at 2,190 psia, which is
60 psi lower than the nominal value. The pressurizer pressure control
system is also assumed to control to a reduced setpoint of 2,190 psia
when it is operable. This lower RCS pressure results in increased
emergency core cooling system (ECCS) injection flow during the transient
and maximizes the challenges to the PSVs and PORVs.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-57 Revision 23 December 2016 The initial RCS Tavg is assumed to be at the minimum pressurizer program level corresponding to 560
°F, which includes a bounding maximum RCS temperature uncertainty of 5
°F. This conservatively maximizes the initial RCS mass, and minimizes the RCS volumetric shrinkage after the reactor trip. For this initial Tavg, the corresponding
programmed initial pressurizer level is 51.2 percent, which bounds the
pressurizer level uncertainty of 6.1 percent.
(2) Pressurizer Heaters Both the backup and proportional pressurizer heaters are assumed to
remain on even after the normal control setpoint is reached to
conservatively maximize the pressurizer liquid volume and decrease the
time to fill the pressurizer with liquid.
(3) Reactor Trip / Turbine Load The reactor trip occurs coincident with the SI actuation, which results in an immediate turbine trip. There is no credit for heat removal from the steam dump system to the condenser or atmosphere. Only the main steam
safety valves are assumed to be operable, with a 3 percent setpoint drift
and 3 percent accumulation. Main feedwater is lost coincident with the
reactor/turbine trip. The analysis conservatively assumes that one motor-
driven auxiliary feedwater (MDAFW) pump delivers the minimum flow of
390 gpm to four steam generators although a MDAFW is aligned to only
two steam generators. The AFW fluid temperature is a maximum value of
100°F. The minimum heat transfer from the primary coolant loop to the secondary system leads to a conservatively early pressurizer fill condition and challenge to the pressurizer overfill condition.
(4) Moderator and Doppler Coefficients of Reactivity Similar to the DNBR analysis, the pressurizer overfill analysis assumes a
positive BOL moderator temperature coefficient and low absolute value
Doppler power coefficient. Since the reactor trip occurs immediately for
the pressurizer overfill case, these reactivity coefficients have a negligible
impact on the results.
(5) Reactor Decay Heat Conservative core residual heat generation is assumed based on long-
term operation at the initial power level preceding the trip. The 1973
decay heat ANSI x 1.2 was used for calculation of residual decay heat
levels.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-58 Revision 23 December 2016 (6) Pressurizer PORVs The pressurizer PORV lift setpoint is assumed to be a minimum of 2,298 psia. The pressurizer PORV delay and stroke time are minimized.
The PORV opens with a delay time of 0.589 second and a stroke time of
0.416 second, and closes with a delay time of 0.825 second and a stroke
time of 0.819 second. The PORV valve area is assumed to
increase/decrease linearly as the valve strokes open and closed. These
assumptions conservatively maximize the number of PORV open cycles
during the SSI event. The backup nitrogen accumulators can provide for
more than 100 PORV cycles in the event of a loss of instrument air (refer
to Section 9.3.1.6).
(7) ECCS Injection Flow Two trains of ECCS pumps are assumed to provide the maximum
injection flow versus RCS pressure. The RWST fluid temperature is
assumed to be 35
°F to maximize the ECCS fluid density.
15.2.15.3.3 Results The sequence of events for the 3 pressurizer overfill cases is listed in Table 15.2.-1.
Typical transient responses are shown in Figures 15.2.15-3 through 15.2.15-5.
Case 1 The spurious safety injection signal occurs at one second. This generates a concurrent
reactor trip signal from full power conditions followed by a turbine trip signal one second later. The pressurizer pressure and pressurizer level initially decrease as the RCS
power and temperature reduce from full power conditions to hot no load conditions. The
initiation of the ECCS injection flow halts the post trip pressure decrease and then
rapidly increases the pressure until the pressurizer spray valves open enough to
maintain the pressurizer pressure relatively constant. The pressurizer level continues to
increase due to ECCS injection flow until the pressurizer fills. The water solid RCS then
experiences a rapid pressure increase to the pressurizer safety valve lift setpoint. The
Case 1 analysis evaluation is considered complete when the fourth liquid relief of the PSV begins at a minimum of 720 seconds. This establishes the minimum time available for the operators to unblock a pressurizer PORV to prevent challenging the liquid relief
capability of the PSV.
Case 2 The first part of each SSI case is essential ly identical as the plant experiences the spurious safety injection, reactor trip, and turbine trip from full power conditions. The
plant response for Case 2 is identical to Case 1 including up to the time that the pressurizer becomes water solid. For Case 2, the RCS pressure increases only to the DCPP UNITS 1 &
2 FSAR UPDATE 15.2-59 Revision 23 December 2016 pressurizer PORV lift setpoint where it is mai ntained relatively constant as the PORV continues to cycle and relieve liquid. By the time the SSI event is terminated at 26 minutes, the PORV has cycled a maximum of 50 times.
Case 3 In Case 3, the pressurizer sprays are not availab le such that after the reactor trip the RCS pressure continues increasing to the pressurizer PORV lift setpoint. The pressurizer PORV continues to cycle and relieve steam as the pressurizer level
increases due to the ECCS injection flow. Without the pressurizer sprays, the RCS
pressure is maintained near the PORV setpoint such that the pressurizer fills later than
Case 1. Once the pressurizer becomes water solid, the PORV begins relieving liquid.
By the time the SSI event is terminated at 26 minutes, the PORV has cycled a
maximum of 93 times.
15.2.15.3.4 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) The operators have adequate time to terminate the event prior to exceeding the assumed maximum pressurizer PORV cycles supported by the backup nitrogen accumulators. The mitigation function of the Class I PORVs ensures that the SSI event can be terminated prior to challenging the PSV liquid relief capability.
Case 1 establishes that for the limiting SSI event, the operators have a minimum time of about 720 seconds or 12 minutes to make a pressurizer PORV available to prevent challenging the PSV li quid relief capability. These results conservatively bound the 11 minutes assumed for the operators to manually unblock a pressurizer PORV.
Cases 2 and 3 establish that with the worst-case control system operation, the
operators have adequate time to terminate an SSI event prior to exceeding the capacity
of pressurizer PORV cycles provided by the backup nitrogen accumulators. The
mitigation function of the Class I PORVs ensures that the SSI event can be terminated
prior to challenging the PSV liquid relief capability.
15.2.16 REFERENCES
- 1. W. C. Gangloff, An Evaluation of Anticipated Operational Transients in Westinghouse Pressurized Water Reactors, WCAP-7486, May 1971.
- 2. D. H. Risher, Jr. and R. F.
Barry, TWINKLE-A Multi-Dimensional Neutron Kinetics Computer Code, WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non-Proprietary), January 1975.
- 3. H. G. Hargrove, FACTRAN, A Fortran IV Code for Thermal Transients in UO 2 Fuel Rod, WCAP-7908-A, December 1989.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-60 Revision 23 December 2016
- 4. T. W. T. Burnett, et al., LOFTRAN Code Description, WCAP-7907-A, April 1984.
- 5. H. Chelemer, et al., Improved Thermal Design Procedure, WCAP-8567-P-A (Proprietary) and WCAP-8568-A (Non-Proprietary), February 1989.
- 6. Deleted in Revision 22.
- 7. Deleted in Revision 22.
- 8. K. Cooper, et al., Overpressure Protection for Westinghouse Pressurized Water Reactor, WCAP-7769, Rev. 1, June 1972.
- 9. J. S. Shefcheck, Application of the THINC Program to PWR Design, WCAP-7359-L, August 1969 (Proprietary), and WCAP-7838 (Non-Proprietary),
January 1972.
- 10. R. L. Haessler, et al., Methodology for the Analysis of the Dropped Rod Event, WCAP-11394 (Proprietary) and WCAP-11395 (Non-Proprietary), April 1987.
- 11. Deleted in Revision 16.
- 12. Deleted in Revision 16.
- 13. Deleted in Revision 16.
- 14. Deleted in Revision 18.
- 15. Deleted in Revision 18.
- 16. Westinghouse letter PGE-98-503, Diablo Canyon Units 1 & 2 Inadvertent ECCS Actuation at Power Analysis - PSV Operability Issue, January 13, 1998.
- 17. Westinghouse Letter NSAL-02-11, Reactor Protection System Response Time Requirements, July 29, 2002
- 18. Westinghouse Letter PGE-02-072, Diablo Canyon Units 1 & 2 Evaluation of Reactor Trip Functions for Uncontrolled RCCA Withdrawal at Power, December 13, 2002.
- 19. RETRAN-02 Modeling and Qualifi cation for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, WCAP-14882-P-A (Proprietary), April 1999, and WCAP-15234-A (Non-Proprietary), May 1999.
- 20. Westinghouse Letter NSAL-07-10, Loss-of-Normal Feedwater/Loss-of-Offsite AC Power Analysis PORV Modeling Assumptions, November 7, 2007.
DCPP UNITS 1 &
2 FSAR UPDATE 15.2-61 Revision 23 December 2016
- 21. PG&E Design Calculation N-115, Reduce d Power Levels for a Number of MSSVs Inoperable, March 14, 1994.
- 22. Westinghouse Letter PGE-10-43, Diablo Canyon Units 1 & 2 Loss of Load /
Turbine Trip Analysis with One Inoperable MSSV per Steam Generator, September 2, 2010.
- 23. Diablo Canyon Units 1 and 2 Replacement Steam Generator NSSS Licensing Report, WCAP-16638-P, Revision 1 (Proprietary), January 2008.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-1 Revision 23 December 2016 15.3 CONDITION III - INFREQUENT FAULTS By definition, Condition III occurrences are faults that may occur very infrequently during
the life of the plant. They will be accompanied with the failure of only a small fraction of
the fuel rods although sufficient fuel damage might occur to preclude resumption of
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 its elf, generate a Condition IV fault or result in a consequential loss of function of the reactor coolant system (RCS) or containment
barriers. For the purposes of this report the following faults have been grouped into this
category:
(1) Loss of reactor coolant, from sma ll ruptured pipes or from cracks in large pipes, that actuates the emergency core cooling system (ECCS).
(2) Minor secondary system pipe breaks.
(3) Inadvertent loading of a fuel assembly into an improper position.
(4) Complete loss of forced reactor coolant flow.
(5) Single rod cluster control assembly (RCCA) withdrawal at full power.
Each of these infrequent faults is analyzed in this section. In general, each analysis includes acceptance criteria, an identification of causes and description of the accident, an analysis of effects and consequences, a presentation of results, and relevant conclusions.
The time sequences of events during four Condition III faults of type (1) above, small-
break loss-of-coolant accident (SBLOCA), are shown in Table 15.3-1.
15.3.1 LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES THA T ACTUATES EMERGENCY CORE COOLING SYSTEM
15.3.1.1 Acceptance Criteria
15.3.1.1.1 10 CFR Part 50, Section 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors (1) Peak cladding temperature.
The calculated maximum fuel element cladding temperature shall not exceed 2200°F.
(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-2 Revision 23 December 2016 (3) Maximum hydrogen generation. 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.
This reduces the potential for explosive hydrogen/oxygen mixtures inside containment.
(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
(5) Long-term cooling. After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.
15.3.1.1.2 Radiological Criteria
The radiological consequences of a SBLOCA are within the applicable guidelines and limits specified in 10 CFR Part 100 detailed in Section 15.5.11.
15.3.1.2 Identification of Causes and Accident Description A LOCA is defined as a rupture of the RCS piping or of any line connected to the
system. This includes small pipe breaks, typically a 3/8-inch diameter opening (0.11 square inch), up to and including a break size of 1.0 square foot that results in
flow that is greater than the makeup flow rate from either CCP1 or CCP2 (refer to Section 6.3.3.6.2.2). Refer to Section 3.6 fo r a more detailed description of the LOCA boundary limits. The coolant that would be released to the containment contains fission products.
The maximum break size for which the normal makeup system can maintain the
pressurizer level is obtained by comparing the calculated flow from the RCS through the
postulated break against the charging system flow capability when aligned for maximum
charging at normal RCS pressure.
Should a larger break occur, depressurization of the RCS causes fluid to flow to the
RCS from the pressurizer resulting in a pressure and level decrease in the pressurizer.
Reactor trip occurs when the pressurizer low-pressure trip setpoint is reached. The
safety injection system (SIS) is actuated when the appropriate pressurizer low-pressure
setpoint is reached. Reactor trip and SIS actuation are also initiated by a high
containment pressure signal. The consequences of the accident are limited in
two ways:
(1) Reactor trip and borated water injection complement void formation in causing rapid reduction of nuclear power to a residual level corresponding
to the delayed fission and fission product decay DCPP UNITS 1 &
2 FSAR UPDATE 15.3-3 Revision 23 December 2016 (2) Injection of borated water ensures sufficient flooding of the core to prevent excessive cladding temperatures Before the break occurs, the plant is in an equilibrium condition; i.e., the heat generated
in the core is being removed via the secondary system. During blowdown, heat from decay, hot internals, and the vessel continues to be transferred to the RCS. The heat
transfer between the RCS and the secondary system may be in either direction
depending on the relative temperatures. In t he case of continued heat addition to the secondary system, system pressure increases and steam dump may occur. Makeup to
the secondary side is automatically provided by the auxiliary feedwater (AFW) pumps.
The safety injection signal stops normal feedwater flow by closing the main feedwater
line isolation valves and initiates emergency feedwater flow by starting AFW pumps.
The secondary flow aids in the reduction of RCS pressure. When the RCS
depressurizes to below approximately 600 psia, the accumulators begin to inject water
into the reactor coolant loops. The reactor coolant pumps are assumed to be tripped at
the beginning of the accident and the ef fects of pump coastdown are included in the blowdown analyses.
15.3.1.3 Analysis of Effects and Consequences For loss-of-coolant accidents due to small breaks less than 1 square foot, the
NOTRUMP (Reference 12) computer code is used to calculate the transient
depressurization of the RCS as well as to describe the mass and enthalpy of flow
through the break. The NOTRUMP computer code is a one-dimensional general
network code with a number of features. Among these features are the calculation of thermal nonequilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, and regime-dependent heat transfer correlations. The NOTRUMP
SBLOCA emergency core cooling system (ECC S) evaluation model was developed to determine the RCS response to design basis SBLOCAs 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 nodalized into volumes interconnected by flowpaths. The
broken loop is modeled explicitly, with the intact loops lumped into a second loop. The transient behavior of the system is determined from the governing conservation
equations of mass, energy, and momentum applied throughout the system. A detailed
description of the NOTRUMP code is provided in References 12 and 13.
The use of NOTRUMP in the analysis involves, among other things, the representation
of the reactor core as heated control volumes with the associated bubble rise model to
permit a transient mixture height calculation.
The multinode capability of the program enables an explicit and detailed s patial representation of various system components.
In particular, it enables a proper calculation of the behavior of the loop seal during a
loss-of-coolant transient.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-4 Revision 23 December 2016 Safety injection flowrate to the RCS as a function of the system pressure is used as part of the input. The SIS was assumed to be delivering water to the RCS 27 seconds after
the generation of a safety injection signal.
For the analysis, the SIS delivery considers pumped injection flow that is depicted in
Figure 15.3-1 as a function of RCS pressure. This figure represents injection flow from
the SIS pumps based on performance curves degraded 5 percent from the design head.
The 27-second delay includes time required for diesel startup and loading of the safety
injection pumps onto the emergency buses. The effect of residual heat removal (RHR)
pump flow is not considered here since their shutoff head is lower than RCS pressure
during the time portion of the transient considered here. Also, minimum safeguards
ECCS capability and operability have been assumed in these analyses.
Peak cladding temperature analyses are performed with the LOCTA IV (Reference 4)
code that determines the RCS pressure, fuel rod power history, steam flow past the
uncovered part to the core, and mixture height history.
15.3.1.4 Results 15.3.1.4.1 Reactor Coolant Sy stem Pipe Breaks This section presents the results of a spectrum of small break sizes analyzed. The
small break analysis was performed at 102 percent of the Rated Core Power (3411
MWt), a Total Peaking Factor (FQT) of 2.70, a Thermal Design Flow of 87,700 / 88,500 gpm/loop (Unit 1 / Unit 2) and a steam generator tube plugging level of 10 percent. For
Unit 1, the small-break analysis was performed for the Replacement Steam Generator (RSG). For Unit 2, the small break analysis was performed for the upflow core
barrel/baffle configuration, upper head temperature reduction and RSG.
The limiting small break size was shown to be a 3-inch diameter break in the cold leg.
In the analysis of this limiting break, an RCS T avg window of 577.3 / 577.6°F, +5°F, -4°F (Unit 1 / Unit 2) was considered. The high T avg cases were shown to be more limiting than the Low T avg cases and therefore are the subject of the remaining discussion. The time sequence of events and the fuel cladding results for the breaks analyzed are
shown in Tables 15.3-1 and 15.3-2.
During the earlier part of the small break transient, the effect of the break flow is not
strong enough to overcome the flow maintained by the reactor coolant pumps through
the core as they are coasting down follo wing reactor trip. Therefore, upward flow through the core is maintained. The resultant heat transfer cools the fuel rods and
cladding to very near the coolant temperature as long as the core remains covered by a
two-phase mixture. This effect is evident in the accompanying figures.
The depressurization transients for the limiting 3-inch breaks are shown in
Figure 15.3-9. The extent to which the core is uncovered for these breaks is presented
in Figure 15.3-11. The maximum hot spot cladding temperature reached during the DCPP UNITS 1 &
2 FSAR UPDATE 15.3-5 Revision 23 December 2016 transient, including the effects of fuel densification as described in Reference 3, is 1391 / 1288°F (Unit 1 / Unit 2). The peak cladding temperature transients for the 3-inch
breaks are shown in Figure 15.3-13. The top core node vapor temperatures for the 3-inch breaks are shown in Figure 15.3-33. When the mixture level drops below the top of the core, the top core node vapor temperature increases as the steam superheats
along the exposed portion of the fuel. The rod film coefficients for this phase of the
transients are given in Figure 15.3-34. The hot spot fluid temperatures are shown in
Figure 15.3-35 and the break mass flows are shown in Figure 15.3-36.
The core power (dimensionless) transient following the accident (relative to reactor
scram time) is shown in Figure 15.3-8. The reactor shutdown time (4.7 sec) is equal to
the reactor trip signal processing time (2.0 seconds) plus 2.7 seconds for complete rod
insertion. During this rod insertion period, the reactor is conservatively assumed to
operate at 102 percent rated power. The small break analyses considered 17x17 Vantage 5 fuel with IFMs, ZIRLO cladding, and an axial blanket. Fully enriched annular
pellets, as part of an axial blanket core design, were modeled explicitly in this analysis.
The results when modeling the enriched annular pellets were not significantly different than the results from solid pellet modeling.
Several figures are also presented for the additional break sizes analyzed.
Figures 15.3-37, 15.3-2, and 15.3-40 present the RCS pressure transient for the 2-, 4-,
and 6-inch breaks, respectively. Figures 15.3-38, 15.3-3, and 15.3-41 present the core
mixture height plots for 2-, 4-, and 6-inch breaks, respectively. The peak cladding
temperature transients for the 2-inch breaks are shown in Figure 15.3-39. The peak
cladding temperature transients for the 4-inch breaks are shown in Figure 15.3-4.
These results are not available for the 6-inch break because the core did not uncover for this transient.
The small break analysis was performed with the Westinghouse ECCS Small Break
Evaluation Model (References 12 and 4) approved for this use by the Nuclear
Regulatory Commission in May 1985. An approved cold leg SI condensation model, COSI (Reference 26), was utilized as part of the Evaluation Model.
15.3.1.5 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
15.3.1.5.1 10 CFR Part 50, Section 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors (1) Peak cladding temperature. The calculated maximum fuel element cladding temperature does not exceed 2200°F, as shown in Table 15.3-2.
(2) Maximum cladding oxidation. The calculated total oxidation of the cladding nowhere exceeds 0.17 times the total cladding thickness before oxidation, as shown in Table 15.3-2.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-6 Revision 23 December 2016 (3) Maximum hydrogen generation. Table 15.3-2 shows that the average cladding oxidation is less than 0.01 times the cladding thickness. Thus
the calculated total amount of hydrogen generated from the chemical
reaction of the cladding with water or steam does 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.
(4) & (5)
Coolable Geometry and Long Term Cooling. The results associated with the SBLOCA analysis performed with the NOTRUMP Evaluation Model
explicitly demonstrate compliance with Criteria 1 through 3. Because of
the fuel rod burst and blockage models used in the LOCTA code, and
modeling of the cold leg recirculation phase in NOTRUMP, SBLOCA
analysis results also support the coolable geometry and long term cooling criteria. Since Criteria 1 through 3 are explicitly met, Criteria 4 and 5 are
met as well. The SBLOCA phenomena and results are therefore in
compliance with 10 CFR 50.46 acceptance criteria.
15.3.1.5.2 Radiological The radiological consequences of a SBLOCA are within the applicable guidelines and limits specified in 10 CFR Part 100 detailed in Section 15.5.11.
15.3.2 MINOR SECONDARY SYSTEM PIPE BREAKS 15.3.2.1 Acceptance Criteria (1) The minimum departure from nucleate boiling ratio (DNBR) does not go below the safety analysis limit (refer to Section 15.4.2.1.1 and 15.4.2.3.1) at any time during the transient to ensure that the core remains geometrically intact with no loss of core cooling capability.
(2) Any activity release must be such that the calculated doses at the site boundary are a small fraction of the applicable guideli nes and limits specified in 10 CFR Part 100 as detailed in Section 15.5.12.
15.3.2.2 Identification of Causes and Accident Description Included in this grouping are ruptures of sec ondary system lines which would result in steam release rates equivalent to a 6-inch diameter break or smaller.
15.3.2.3 Analysis of Effects and Consequences Minor secondary system pipe breaks must not result in more than the failure of only a small fraction of the fuel elements in the reactor. Since the results of analysis presented in Section 15.4.2 for a major secondary system pipe rupture also meet these criteria, separate analyses for minor secondary system pipe breaks is not required.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-7 Revision 23 December 2016 The analyses of the more probable accidenta l opening of a secondary system steam dump, relief, or safety valve is presented in Section 15.2.14. These analyses are illustrative of a pipe break equivalent in size to a single valve opening.
15.3.2.4 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) The analysis presented in Section 15.4.2 demonstrates that the consequences of a minor secondary system pipe break are acceptable
because a DNBR of less than the design basis values does not occur
even for a more critical major secondary system pipe break.
(2) Section 15.5.12 demonstrates the potential radiological exposures to the public following a minor secondary system pipe rupture per the applicable guidelines and limits specified in 10 CFR Part 100 are met.
15.3.3 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO AN IMPROPER POSITION 15.3.3.1 Acceptance Criteria (1) In the event of a fuel loading error not identified until normal operation, the offsite dose consequences should be a small fraction of the applicable guidelines and limits specified in 10 CFR Part 100 as detailed in
Section 15.5.1.
15.3.3.2 Identification of Causes and Accident Description Fuel and core loading errors such as inadvertently loading one or more fuel assemblies
into improper positions, loading a fuel rod during manufacture with one or more pellets
of the wrong enrichment, or loading a full fuel assembly during manufacture with pellets
of the wrong enrichment will lead to increased heat fluxes if the error results in placing
fuel in core positions calling for fuel of lesser enrichment. The inadvertent loading of
one or more fuel assemblies requiring burnable poison rods into a new core without
burnable poison rods is also included among possible core loading errors.
Any error in enrichment, beyond the normal manufacturing tolerances, can cause power
shapes that are more peaked than those calculated with the correct enrichments. The
incore system of movable neutron flux detectors that is used to verify power shapes at
the start of life is capable of revealin g any assembly enrichment error or loading error that causes power shapes to be peaked in excess of the design value.
To reduce the probability of core loading errors, each fuel assembly is marked with an
identification number and loaded in accordance with a core loading diagram. For each
core loading, the identification number is checked to ensure proper core configuration.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-8 Revision 23 December 2016 The power distortion due to any combination of misplaced fuel assemblies would significantly raise peaking factors and would be readily observable with movable incore neutron flux detectors. In addition to the flux detectors, thermocouples are located at
the outlet of about one-third of the fuel assemblies in the core. There is a high
probability that these thermocouples would also indicate any abnormally high coolant enthalpy rise. Incore flux measurements are taken during the startup subsequent to
every refueling operation. A more detailed discussion of the flux detection capabilities may be found in Section 4.3.3.2.
15.3.3.3 Analysis of Effects and Consequences Steady state power distributions in the x-y plane of the core are calculated with the
TURTLE code (refer to Section 1.6.1, Item 49, and Section 4.3.3.8.5, based on macroscopic cross sections calculated by the LEOPARD code (refer to Section 1.6.1, Item 48, and Section 4.3.3.10.2). A discrete representation is used wherein each individual fuel rod is described by a mesh interval. The power distributions in the x-y plane for a correctly loaded core assembly are given in Chapter 4 based on
enrichments given in that section.
For each core loading error case analyzed, the percent deviations from detector
readings for a normally loaded core are shown at all incore detector locations (refer to Figures 15.3-15 through 15.3-19).
15.3.3.4 Results The following core loading error cases have been analyzed:
(1) Case A The case in which a Region 1 assembly is interchanged with a Region 3 assembly. The particular case considered was the interchange of two
adjacent assemblies near the periphery of the core (refer to Figure 15.3-15). (2) Case B The case in which a Region 1 assembly is interchanged with a neighboring Region 2 fuel assembly. Two analyses have been performed
for this case (refer to Figures 15.3-16 and 15.3-17).
In Case B-1, the interchange is assumed to take place with the burnable poison rods
transferred with the Region 2 assembly mistakenly loaded into Region 1.
In Case B-2, the interchange is assumed to take place closer to core center and with
burnable poison rods located in the correct Region 2 position but in a Region 1
assembly mistakenly loaded into the Region 2 position.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-9 Revision 23 December 2016 (3) Case C Enrichment error: the case in which a Region 2 fuel assembly is loaded in the core central position (refer to Figure 15.3-18).
(4) Case D The case in which a Region 2 fuel assembly instead of a Region 1 assembly is loaded near the core periphery (refer to Figure 15.3-19).
15.3.3.5 Conclusions In the event that a single rod or pellet has a higher enrichment than the nominal value, the consequences in terms of reduced DNBR and increased fuel and cladding
temperatures will be limited to the incorrectly loaded rod or rods.
Fuel assembly loading errors are prevented by administrative procedures implemented
during core loading. In the unlikely event that a loading error occurs, analyses in this
section confirm that resulting power distribution effects will either be readily detected by the incore movable detector system or will cause a sufficiently small perturbation to be
acceptable within the uncertainties allowed be tween nominal and design power shapes.
The analysis demonstrates the acceptance criterion is met as follows:
(1) No events leading to environmental radiological consequences are expected as a result of loading errors (refer to Section 15.5.13).
15.3.4 COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.3.4.1 Acceptance Criteria (1) Maintain the minimum DNBR greater than the safety analysis limit for fuel (refer to Section 4.4.3.1).
15.3.4.2 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 departure from nucleate boiling (DNB) with subsequent fuel damage if the reactor were not tripped promptly. The
following reactor trips provide necessary protection against a loss of coolant flow
accident:
(1) Low reactor coolant loop (RCL) flow (primary protection)
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-10 Revision 23 December 2016 (2) Undervoltage or underfrequency on reactor coolant pump power supply buses (backup protection to low RCL flow trip)
(3) Pump circuit breaker opening (also backup to low reactor coolant loop flow trip) The reactor trip on low RCL flow is provided as primary protection against loss-of-flow conditions. This function is generated by two-out-of-three low-flow signals per reactor coolant loop. Above approximately 35 percent power (Permissive 8), low flow in any
loop will actuate a reactor trip.
Between approximately 10 and 35 percent power (Permissive 7 and Permissive 8), low-flow in any two loops will actuate a reactor trip.
The reactor trip on reactor coolant pump bus undervoltage is provided as protection against conditions that can cause a loss of voltage to all reactor coolant pumps, i.e., loss of offsite power. This function serves as backup protection to the low RCL flow trip and is blocked below approximately 10 percent 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 major power grid. Underfrequency also opens the reactor coolant pump breakers that disengage the reactor coolant pumps from the power grid so that the pumps flywheel kinetic energy is available for full coastdown. This function also serves as backup protection to the low RCL flow trip.
A reactor trip from opened pump breakers is also provided as a backup to the low-flow signals. Above Permissive 7 a breaker open signal from any 2 of 4 pumps will actuate a reactor trip. Reactor trip on reactor coolant pump breakers open is blocked below Permissive 7.
Normal power for the reactor coolant pumps is supplied through buses from a
transformer connected to the generator. Two pumps are on each bus. When a
generator trip occurs, the buses are automatically transferred to a transformer supplied from external power lines, and the pumps will continue to supply coolant flow to the
core. Following any turbine trip, where there are no electrical or mechanical faults
which require immediate tripping of the generator from the network, the generator
remains connected to the network for approximately 30 seconds. The reactor coolant
pumps remain connected to the generator thus ensuring full flow for 30 seconds after
the reactor trip before any transfer is made.
15.3.4.3 Analysis of Effects and Consequences This transient is analyzed by three digital computer codes. First the LOFTRAN (Reference 8) code is used to calculate the loop and core flow during the transient. The
LOFTRAN code is used to calculate the nuclear power transient. The FACTRAN (Reference 9) code is then used to calculate the heat flux transient based on the DCPP UNITS 1 &
2 FSAR UPDATE 15.3-11 Revision 23 December 2016 nuclear power and flow from LOFTRAN. Finally, the THINC-IV code (refer to Section 4.4.3.4) is used to calculate the minimum DNBR during the transient based on the heat flux from FACTRAN and flow from LOFTRAN. The transients presented represent the minimum of the typical and thimble cells.
The following cases have been analyzed:
(1) Four of four loops coasting down.
(2) Reactor coolant pumps power supply frequency decay at a maximum constant 3 Hz/sec rate (underfrequency).
The method of analysis and the assumptions made regarding initial operating conditions and reactivity coefficients are identical to those discussed in Section 15.2 and Table 15.1-4, except that following the loss of electrical supply to all pumps at power, the reactor trip is actuated by low RCL flow rather than bus undervoltage or bus underfrequency.
15.3.4.4 Results The calculated sequence of events is shown in Table 15.3-3. Figures 15.3.4-1 through
15.3.4-3 show the flow coastdown, the heat flux response, and the nuclear power response for the limiting complete loss of flow event(four-loop coastdown). For analysis purposes, the reactor is assumed to trip on the low RCL flow signal, as this trip signal is PG&E Design Class I, while the undervoltag e and underfrequency trips are PG&E Design Class II. Because the low RCL flow trip responds to an actual condition (while the undervoltage/underfrequency trips are anticipatory), the reactor trip is delayed and the transient is more DNBR limiting. A plot of DNBR versus time is given in Figure 15.3.4-4. This plot represents the limiting (thimble) cell for the four-loop coastdown.
Figures 15.3.4-1 through 15.3.4-4 present only the Unit 1 results; however, the Unit 2 results are nearly identical. Therefore, the figures are representative of Unit 2 response.
15.3.4.5 Conclusions The safety analysis results described in Sect ion 15.3.4.4 have demonstrated that for the complete loss of forced reactor coolant flow, the minimum DNBR is above the safety
analysis limit values of 1.71/1.68 (typical cell/thimble cell) during the transient; therefore, no core safety limit is violated.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-12 Revision 23 December 2016 15.3.5 SINGLE ROD CLUSTER CONTROL ASSE MBLY WITHDRAWAL AT FULL POWER 15.3.5.1 Acceptance Criteria (1) No more than 5 percent of the fuel rods experience a DNBR less than the limit value.
15.3.5.2 Identification of Causes and Accident Description By design, no single electrical or mechanical failure in the rod control system could
cause the accidental withdrawal of a single RCCA from the inserted bank at full power
operation. The operator could deliberately withdraw a single RCCA in the control bank;
this feature is necessary in order to retrieve an assembly should one be accidentally
dropped. In the extremely unlikely event of simultaneous electrical failures that could
result in single RCCA withdrawal, rod deviation and control rod urgent failure may be
displayed on the plant annunciator, and the rod position indicators would indicate the relative positions of the assemblies 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.
Each bank of RCCAs in the system is divided into two groups of four mechanisms each (except Group 2 of Bank D which consists of five mechanisms). 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 and deactuation of the stationary gripper, movable gripper, and lift coils of a mechanism is required to withdraw the
RCCA attached to the mechanism. Since the four stationary grippers, movable
grippers, and lift coils associated with the four RCCAs of a rod group are driven in
parallel, any single failure that would cause rod withdrawal would affect a minimum of one group, or four RCCAs. Mechanical failures are either in the direction of insertion or
immobility.
In the unlikely event of multiple failures that result in continuous withdrawal of a single
RCCA, it is not possible, in all cases, to provide assurance of automatic reactor trip so
that core safety limits are not violated. 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 covered by the RCCA.
15.3.5.3 Analysis of Effects and Consequences Power distributions within the core are calculated by the ANC code based on
macroscopic cross sections generated by PHOENIX-P (refer to Section 4.3.3.10.2).
The peaking factors calculated by ANC (refer to Section 4.3.3.10.3) are then used by THINC (refer to Section 1.6.1, Item 28, and Section 4.4.3) to calculate the minimum DCPP UNITS 1 &
2 FSAR UPDATE 15.3-13 Revision 23 December 2016 DNBR for the event. The plant was analyz ed for the case of the worst rod withdrawn from Control Bank D inserted at the insertion limit, with the reactor initially at full power.
15.3.5.4 Results Two cases have been considered as follows:
(1) If the reactor is in the automatic control mode, 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 2
described below. For such cases as above, a trip will ultimately ensue, although not sufficiently fast in all cases to prevent a minimum DNBR in
the core of less than the safety limit. (2) 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 failed RCCA. In terms of the overall system response, this case is
similar to those presented in Section 15.2; however, the increased local
power peaking in the area of the withdrawn RCCA results in lower
minimum DNBR 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 safety limit value. Evaluation of this case to determine
the most limiting DNBR condition, which would occur at the power and coolant condition 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 safety limit value is 5 percent.
15.3.5.5 Conclusions The analysis demonstrates the acceptance criterion is met as follows:
(1) For both cases of one RCCA fully withdrawn, with the reactor in either the automatic or manual control mode and initially operating at full power with
Bank D at the insertion limit, 5 percent or less of the total fuel rods in the
core will go below the minimum DNBR safety analysis limit.
For both cases discussed, the indicators and alarms mentioned would function to alert
the operator to the malfunction before any DNB could occur. For Case 2 discussed
above, the insertion limit alarms (low and low-low alarms) would also serve in this
regard. However, operator action is not requ ired to meet the acceptance criteria.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-14 Revision 23 December 2016 15.
3.6 REFERENCES
- 1. Deleted in Revision 3.
- 2. Deleted in Revision 3.
- 3. J. M. Hellman, Fuel Densification Experimental Results and Model for Reactor Application, WCAP-8218-P-A, March 1975 (Westinghouse Proprietary) and WCAP-8219-A, March 1975.
- 5. Deleted in Revision 15.
- 6. Deleted in Revision 15.
- 7. Deleted in Revision 3.
- 8. T. W. T. Burnett, et al., LOFTRAN Code Description, WCAP-7907-A, April 1984.
- 9. H. G. Hargrove, FACTRAN, A Fortran-IV Code for Thermal Transients in UO 2 Fuel Rods, WCAP-7908-A, December 1989.
- 10. Deleted in Revision 21.
- 11. Deleted in Revision 6.
- 12. P. E. Meyer, NOTRUMP, A Nodal Transient Small Break and General Network Code, WCAP-10079-P-A, August 1985.
- 13. H. Lee, S. D. Rupprecht, W. D. Tauche, W. R. Schwarz, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, WCAP-10054-P-A, August 1985.
- 14. Deleted in Revision 13.
- 15. Deleted in Revision 13.
- 16. Deleted in Revision 13.
- 17. Deleted in Revision 12.
- 18. Deleted in Revision 12.
DCPP UNITS 1 &
2 FSAR UPDATE 15.3-15 Revision 23 December 2016
- 19. Deleted in Revision 12.
- 20. Deleted in Revision 13.
- 21. Deleted in Revision 12.
- 22. Deleted in Revision 12.
- 23. Deleted in Revision 13.
- 24. Deleted in Revision 13.
- 25. Deleted in Revision 13.
- 26. Thompson, C. M., et al., Addendum to the Westinghouse Small Break LOCA Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken
Loop and the COSI Condensation Model, WCAP-10054-P-A, Addendum 2, Rev. 1, (proprietary), October 1995.
- 27. T. Q. Nguyen, et al., Qualification of the PHOENIX-P/ANC Nuclear Design System for Pressurized Water Reactor Cores, WCAP-11596-P-A, June 1988.
- 28. S. L. Davidson, (Ed), et al., ANC: Westinghouse Advanced Nodal Computer Code, WCAP-10965-P-A, September 1986.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-1 Revision 23 December 2016 15.4 CONDITION IV - LIMITING FAULTS Condition IV occurrences are faults that are not expected to take place, but are
postulated because their consequences would include the potential for the release of
significant amounts of radioactive material. These are the most drastic occurrences that
must be designed against and represent limiting design cases. Condition IV faults shall
not cause a fission product release to the environment resulting in an undue risk to
public health and safety in excess of guideline values of 10 CFR Part 100. A single Condition IV fault shall not cause a consequential loss of required functions of systems
needed to cope with the fault including those of the emergency core cooling system (ECCS) and the containment. For the purposes of this report the following faults have
been classified in this category:
(1) Major rupture of pipes containing reactor coolant up to and including double-ended rupture of the largest pipe in the reactor coolant system (RCS), i.e., loss-of-coolant-accident (LOCA)
(2) Major secondary system pipe ruptures (3) Steam generator tube rupture (4) Single reactor coolant pump (RCP) locked rotor (5) Fuel handling accident (6) Rupture of a control rod mechanism housing (rod cluster control assembly (RCCA) ejection)
(7) Rupture of a gas decay tank (8) Rupture of a liquid holdup tank (9) Rupture of a volume control tank
Each of these nine limiting faults is analyzed in this section. In general, each analysis
includes acceptance criteria, an identification of causes and description of the accident, an analysis of effects and consequences, a presentation of results, and relevant
conclusions.
The analyses of thyroid and whole body doses, resulting from events leading to fission
product release, are presented in Section 15.5. The fission product inventories that
form a basis for these calculations are presented in Chapter 11 and Section 15.5. Also included is a discussion of system interdependency contributing to limiting fission product leakages from the containment following a Condition IV occurrence.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-2 Revision 23 December 2016 The large break LOCA analysis contained in Section 15.4.1 has been revised to incorporate separate Best Estimate LOCA analyses for Units 1 and 2. The general
discussion of the Best Estimate LOCA transient in Sections 15.4.1.2, 15.4.1.3, and
15.4.1.4 are applicable to Units 1 and 2. However, the statistical treatment
methodologies are slightly different for Units 1 and 2. Statistical treatment
methodologies for Units 1 and 2 are discussed in Sections 15.4.1.4A and 15.4.1.4B
respectively.
15.4.1 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOCA) 15.4.1.1 Acceptance Criteria
15.4.1.1.1 10 CFR Part 50, Section 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors
It must be demonstrated that there is a high l evel of probability that the limits set forth in 10 CFR 50.46 are met. The acceptance criteria are listed below:
(1) Peak cladding temperature. The calculated maximum fuel element cladding temperature shall not exceed 2200 °F.
(2) Maximum cladding oxidation. The calculated total oxidation of the cladding shall nowhere exceed 0.
17 times the total cladding thickness before oxidation.
(3) Maximum hydrogen generation.
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.
(4) Coolable geometry. Calculated changes in core geometry shall be such that the core remains amenable to cooling.
(5) Long-term cooling. After any calculated successful initial operation of the
ECCS, the calculated core temperature shall be maintained at an
acceptably low value and decay heat shall be removed for the extended
period of time required by the long-lived radioactivity remaining in the
core. 15.4.1.1.2 Radiological Criteria (1) The resulting potential exposures to individual members of the public and to the general population shall be lower than the applicable guidelines and limits specified in 10 CFR Part 100.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-3 Revision 23 December 2016 15.4.1.2 Background of Best Estimate Large Break LOCA The analysis performed to comply with the requirements of 10 CFR 50.46 (Reference 1), and Revisions to the Acceptance Criteria (Reference 54) is presented in
this section.
In 1988, the NRC Staff amended the requirements of 10 CFR 50.46 and Appendix K, ECCS Evaluation Models, 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 55).
A LOCA evaluation methodology for three-and four-loop 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. The methodology is documented in WCAP-12945, Code Qualification Document (CQD) for Best Estimate LOCA Analysis (Reference 56).
The time sequence of events during a nominal large double-ended cold leg guillotine (DECLG) break LOCA is shown in Tables 15.41-1A and 15.4.1-1B. The results of the large break LOCA analysis are shown in Tables 15.4.1-2A and 15.4.1-2B and show compliance with the acceptance criteria. The analytical techniques used for the large
break LOCA analysis are in compliance with 10 CFR 50.46 (Reference 1) as amended
in Reference 54, and are described in Reference 56. Due to the significant differences
between the Unit 1 and Unit 2 reactor vessel internals, plant-specific vessel models
were developed and evaluated. The significant differences between the units are summarized below:
Unit 1 Unit 2 Top Hat-Upper Support Plate Flat Upper Support Plate Domed Lower Support Plate Flat Lower Support Plate Thermal Shield Neutron Pads Diffuser Plate No Diffuser Plate
An analysis of each unit was performed and a comparison determined that the Unit 1
vessel model resulted in more limiting PCT values. As a result, the Best Estimate base
Large Break LOCA analysis (Reference 60) results were based on Unit 1 and were
considered bounding for both Unit 1 and Unit
- 2. Recently, the Unit 1 Best Estimate LOCA was reanalyzed for Unit 1 using the approved reanalysis methodology established in Reference 56. In the process of performing the Unit 1 reanalysis DCPP UNITS 1 &
2 FSAR UPDATE 15.4-4 Revision 23 December 2016 (Reference 67), it was determined that the Unit 1 vessel model no longer consistently resulted in the limiting PCTs, and could not be considered bounding for Unit 2.
Therefore, the reanalysis methodology (Reference 56) was only applied to Unit 1, and a new and separate Best Estimate Large Break LOCA analysis was performed for Unit 2
using an updated and slightly different methodology as described in Reference 69.
Both Unit 1 and Unit 2 use the base Best Estimate Large Break LOCA analysis
methodology and computer code as described in Reference 60 and described in
Section 15.4.1.2, which is applicable to Units 1 and 2. Separate subsequent
subsections describe the Unit 1 reanalysis methodology (Reference 67) and the Unit 2
analysis methodology, and the respective results.
15.4.1.3 WCOBRA/TRAC Thermal-hydraulic Computer Code
The thermal-hydraulic computer code that was reviewed and approved for the calculation of fluid and thermal conditions in the PWR during a large break LOCA is
WCOBRA/TRAC, Version Mod 7A, Revision 1 (Reference 56). 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 codes
ability to predict key physical phenomena in the PWR large break LOCA. Slightly
different revisions to this computer code were used for the Unit 1 reanalysis and the
separate Unit 2 analysis as described in later sections.
WCOBRA/TRAC combines two-fluid, three-field, multi-di mensional 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, and momentum transfer in different flow regimes (4) Ability to represent important rea ctor components such as fuel rods, steam generators, reactor coolant pumps, etc.
The two-fluid formulation uses a separate set of conservation equations and constitutive
relations for each phase. The effects of one phase on another are accounted for by
interfacial friction and heat and mass transfer interaction terms in the equations. The
conservation equations have the same form for each phase; only the constitutive
relations and physical properties differ. Dividing the liquid phase into two fields is a convenient and physically accurate way of handling flows where the liquid can appear in
both film and droplet form. The droplet field permits more accurate modeling of DCPP UNITS 1 &
2 FSAR UPDATE 15.4-5 Revision 23 December 2016 thermal-hydraulic phenomena, such as entrainment, de-entrainment, fallback, liquid pooling, and flooding.
WCOBRA/TRAC also features a two-phase, one-dimensional hydrodynamics formulation. In this model, the effect of phase slip is modeled indirectly via a
constitutive relationship that provides the phase relative velocity as a function of fluid
conditions. Separate mass and energy conservation equations exist for the two-phase
mixture and for the vapor.
The reactor vessel is modeled with the three-dimensional, three field model, while the
loop, major loop components, and safety injection points are modeled with the
one-dimensional model.
All geometries modeled using the three-dimensional model are represented as a matrix
of cells. The number of mesh cells used depends on the degree of detail required to
resolve the flow field, the phenomena being modeled, and practical restrictions such as
computing costs and core storage limitations.
The equations for the flow field in the three-dimensional model are solved using a
staggered difference scheme on the Eulerian mesh. The velocities are obtained at
mesh cell faces, and the state variables (e.g., pressure, density, enthalpy, and phasic
volume fractions) are obtained at the cell center. This cell is the control volume for the
scalar continuity and energy equations. The momentum equations are solved on a
staggered mesh with the momentum cell centered on the scalar cell face.
The basic building block for the mesh is the channel, a vertical stack of single mesh cells. Several channels can be connected together by gaps to model a region of the reactor vessel. Regions that occupy the same level form a section of the vessel.
Vessel sections are connected axially to complete the vessel mesh by specifying
channel connections between sections. Heat transfer surfaces and solid structures that
interact significantly with the fluid can be modeled with rods and unheated conductors.
One-dimensional components are connected to the vessel. The basic scheme used
also employs the staggered mesh cell. Special purpose components exist to model
specific components such as the steam generator and pump.
A typical calculation using WCOBRA/TRA C begins with the establishment of a steady-state, initial condition with all loops intact. The input parameters and initial
conditions for this steady-state calculation are discussed in the next section.
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 61) and
mass and energy releases from t he WCOBRA/TRAC calculation.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-6 Revision 23 December 2016 15.4.1.4 Thermal Analysis 15.4.1.4.1 Westinghouse Performance Criteria for ECCS The reactor is designed to withstand thermal effects caused by a LOCA including the
double-ended severance of the largest RCS pipe. The reactor core and internals
together with the ECCS are designed so that the reactor can be safely shut down and
the essential heat transfer geometry of the core preserved following the accident.
The ECCS, even when operating during the injection mode with the most severe single
active failure, is designed to meet the acceptance criteria of 10 CFR 50.46.
15.4.1.4.2 Sequence of Events and Systems Operations The sequence of events following a nominal large DECLG break LOCA is presented in
Tables 15.4.1-1A and 15.4.1-1B for Units 1 and 2, respectively. Should a major break
occur, depressurization of the RCS results in a pressure decrease in the pressurizer.
The reactor trip signal subsequently occurs when the pressurizer low pressure trip
setpoint is reached. A safety injection signal is generated when the appropriate setpoint
is reached. These countermeasures will limit the consequences of the accident in two
ways:
(1) Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat. No credit is taken during the LOCA transient
for negative reactivity due to the boron concentration of the injection water. However, an average RCS/sump mixed boron concentration is calculated to ensure that the post-LOCA core remains subcritical. In
addition, the insertion of control rods to shut down the reactor is not
assumed in the large break analysis.
(2) Injection of borated water provides the fluid medium for heat transfer from the core and prevents excessive cladding temperatures.
For the present Westinghouse PWR design, the limiting single failure assumed for a
large break LOCA is the loss of one train of ECCS pumps (one charging pump (CCP1
or CCP2), one high-head safety injection (SI) pump, and one residual heat removal
pump). One ECCS train delivers flow through the injection lines to each loop, with the
least resistant branch injection line spilling to containment backpressure (Figures
15.4.1-14A and 15.4.1-14B and Tables 15.4.1-7A and 15.4.1-7B). All emergency diesel
generators (EDGs) are assumed to start in the modeling of the containment fan coolers
and spray pumps. Modeling full operation of the containment heat removal system is
required by Branch Technical Position CSB 6-1, and is a conservative assumption for
the large break LOCA analysis.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-7 Revision 23 December 2016 15.4.1.4.3 Description of a Large Break LOCA Transient Before the break occurs, the RCS is assumed to be operating normally at full power in
an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. A large DECLG break is assumed to open almost instantaneously
in one of the main RCS pipes. Calcul ations have demonstrated that the most severe transient results occur for a DECLG break between the pump and the reactor vessel.
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 blowdow n 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 15.4.1-1A to 15.4.1-12A for Unit 1 and
Figures 15.4.1-1B to 15.4.1-12B for Unit 2.
(1) Critical Heat Flux (CHF) Phase Immediately following the cold leg rupture, the break discharge rate is subcooled and high. 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 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. 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 15.4.1-4A and
15.4.1-4B show the void fraction for one intact loop pump and the broken
loop pump for Units 1 and 2, respectively. The figures show that the intact
loop remains in single-phase liquid 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.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-8 Revision 23 December 2016 (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, as shown in
Figures 15.4.1-5A and 15.4.1-5B for Units 1 and 2, respectively, best
illustrate 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. 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.
(4) Refill Period As the refill period begins, the core begins a period of heatup and the
vessel begins to fill with ECCS water. 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. 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 repressurization.
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 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 accumul ators rapidly discharge borated cooling water into the RCS, filling the lower plenum and contributing to the filling of
the downcomer. The pumped ECCS water aids in the filling of the
downcomer and subsequently supplies wate r to maintain a full downcomer and complete the reflood period. As the quench front progresses up the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-9 Revision 23 December 2016 core, the PCT location moves higher into the top core region. As the vessel continues to fill, the PCT location is cooled and the early reflood
period is terminated.
A second cladding heatup transient may oc cur 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 15.4.1-12A and 15.4.1-12B show only a
slight reduction in downcomer level which indicates that a late reflood
heatup does not occur for either Unit. However, the Unit 1 reanalysis
methodology (Reference 67) still requires that both the early and late
reflood PCT periods be considered, while the Unit 2 updated analysis
methodology (Reference 69) has eliminated the need to evaluate the late
reflood period for PCT. For the Unit 1 reanalysis, the first reflood peak is
considered to be the maximum PCT, which occurs after the beginning of
reflood, and before the beginning of gravity driven reflood. In Unit 1
Figure 15.4.1-1A, this corresponds to the maximum PCT between about 35 and 50 seconds after the break. The second reflood peak is then
considered to be the maximum PCT, which occurs after the beginning of gravity driven reflood. This terminology for first and second reflood PCTs is only used in the further discussions of the Unit 1 Best Estimate LBLOCA
reanalysis.
Continued operation of the ECCS pumps suppli es water during the long-term cooling period. Core temperatures have been reduced to long-term steady state levels
associated with dissipation of residual heat generation. When low level is reached in
the refueling water storage tank (RWST), switchover to the recirculation phase is
initiated. The residual heat removal (RHR) pumps are tripped, and the operator
manually aligns the charging (CCP1 or CCP
- 2) and safety injection (SI) pumps to the RHR pump discharge. Once the alig nment is completed, all ECCS pumps recirculate containment recirculation sump water. The containment spray pumps continue to draw
suction from the RWST until the low-low level is reached, at which time the containment
spray pumps are tripped. If two RHR pumps are running, the containment spray valves
can be aligned so that an RHR pump can be utilized to deliver recirculation water to the containment spray ring headers and spray nozzles for continued containment spray
system post-accident operation.
Approximately 7.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after initiation of the LOCA, the ECCS is realigned to supply
water to the RCS hot legs in order to control the boric acid concentration in the reactor DCPP UNITS 1 &
2 FSAR UPDATE 15.4-10 Revision 23 December 2016 vessel. Long-term cooling also includes long-term criticality control. To achieve long-term criticality control, a mixed-mean sump boron concentration is determined and
verified against core design margins to ensure core subcriticality, without credit for
RCCA insertion. A mixed-mean sump boron concentration is calculated based on minimum volumes for boron sources and maximum volumes for dilution sources. The
calculated mixed-mean sump boron concentration is verified against available core
design margins on a cycle-specific basis.
15.4.1.4A Unit 1 Best Estimate Large Break LOCA Evaluation Model The thermal-hydraulic computer code that was reviewed and approved for the
calculation of fluid and thermal conditions in the PWR during a large break LOCA is
WCOBRA/TRAC, Version MOD7A Rev. 1 (Reference 56). Modeling of the PWR introduces additional uncertainties that are identified and quantified for the plant-specific Unit 1 analysis (Reference 60). The final step of the best estimate analysis
methodology is to combine all the uncertainties related to the code and plant parameters, and estimate the PCT at 95 percent probability. The steps taken to derive
the PCT uncertainty estimate are summarized below
(1) Plant Model Development In this step, a WCOBRA/TRAC model of the plant is developed. A high level of noding detail is used in order to provide an accurate simulation of
the transient. However, specific guidelin es 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. 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. The most limiting input conditions, based on these confirmatory runs, are then combined into a single transient, which is then
called the reference transient.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-11 Revision 23 December 2016 (3) PWR Sensitivity Calculations A series of PWR transients is performed in which the initial fluid conditions and boundary conditions are ranged around the nominal condition used in
the reference transient. The results of these calculations for DCPP form the basis for the determination of the initial condition bias and uncertainty
discussed in Section 6 of Reference 60.
Next, a series of transients is performed that vary the power distribution, taking into account all possible po wer distributions during normal plant operation. The results of these calculations for DCPP form the basis for
the determination of the power distribution bias and uncertainty discussed
in Section 7 of Reference 60.
Finally, a series of transients is performed that vary parameters that affect the overall system response (global parameters) and local fuel rod
response (local parameters). The results of these calculations for DCPP
form the basis for the determination of the model bias and uncertainty
discussed in Section 8 of Reference 60.
(4) Response Surface Calculations Regression analyses are performed to derive PCT response surfaces from
the results of the power distribution run matrix and the global model run
matrix. The results of the initial conditions run matrix are used to generate
a PCT uncertainty distribution.
(5) Uncertainty Evaluation The total PCT uncertainty from the initial conditions, power distribution, and model calculations is derived using the approved methodology (Reference 56). The uncertainty calculations assume certain plant
operating ranges that may be varied depending on the results obtained.
These uncertainties are then combined to determine the initial estimate of
the total PCT uncertainty distribution for the DECLG and split breaks. The
results of these initial estimates of the total PCT uncertainty are compared
to determine the limiting break type.
If the split break is limiting, an additional set of split transients is performed that vary overall system
response (global parameters) and local fuel rod response
(local parameters). Finally, an additional series of runs is made to
quantify the bias and uncertainty due to assuming that the above three
uncertainty categories are independent. The final PCT uncertainty
distribution is then calculated for the limiting break type, and the
95th percentile PCT is determined.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-12 Revision 23 December 2016 (6) 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 established in step 2, or
may be narrower for some parameters to gain additional margin.
There are three major uncertainty categories or elements:
(1) Initial condition bias and uncertainty (2) Power distribution bias and uncertainty (3) Model bias and uncertainty
Conceptually, these elements may be assumed to affect the reference transient PCT as
shown below.
PCT i = PCT REFi + PCT ICi + PCT PDi + PCTMODi (15.4.1-1) where, PCTREFi = Reference transient PCT: The reference transient PCT is calculated using WCOBRA/TRAC at the nominal conditions identified in Table 15.4.1-3A, for blowdown (i=1), first reflood (i=2), and second reflood (i=3).
PCTICi = Initial condition bias and uncertain ty: This bias is the difference between the reference transient PCT, which assumes several
nominal or average initial conditions, and the average PCT taking
into account all possible values of the initial conditions. This bias takes into account plant variations that have a relatively small effect
on PCT. The elements that make up this bias and its uncertainty
are plant specific.
PCTPDi = Power distribution bias and uncertainty: This bias is the difference between the reference transient PCT, which assumes a nominal
power distribution, and the average PCT taking into account all
possible power distributions during normal plant operation.
Elements that contribute to the uncertainty of this bias are
calculational uncertainties, and variations due to transient operation
of the reactor.
PCTMODi = Model bias and uncertainty: This component accounts for uncertainties in the ability of the WCOBRA/TRAC code to accurately predict important phenomena that affect the overall system DCPP UNITS 1 &
2 FSAR UPDATE 15.4-13 Revision 23 December 2016 response (global parameters) and the local fuel rod response (local parameters). The code and model bias is the difference
between the reference transient PCT, which assumes nominal
values for the global and local parameters, and the average PCT
taking into account all possible values of global and local
parameters.
The separability of the uncertainty components in the manner described above is an approximation since the parameters in each element may be affected by parameters in
other elements. The bias and uncertainty associated with this assumption are
quantified as part of the overall uncertainty methodology and included in the final estimates of the 95-percentile PCT ( PCT 95%).
The application of the reanalysis methodology to Unit 1 first determines a new reference
transient PCT. The bias and uncertainty associated with the initial conditions, power
distributions, and models are assumed to remain unchanged. This assumption is
assessed to determine that the fundamental LOCA transient characteristics remain
unchanged from the new reference transient to that of the original analysis. If applicable, the uncertainty in applying the reanalysis methodology is determined when the superposition assumption is requantified (i.e., the assumption that the major uncertainty elements are independent), and the new bias and new uncertainty is
calculated.
15.4.1.5A Unit 1 Containment Backpressure A conservatively bounding minimum containment back pressure (Figure 15.4.1-14A) is calculated using the methods and assumptions described in Reference 2, Appendix A.
Containment back pressure is calculated using the COCO code (Reference 61) and
mass and energy releases from the WCOBRA/TRA C calculation. Input parameters used for the Unit 1 containment backpressure calculation are presented in
Table 15.4.1-5A. This minimum containment back pressure is modeled using a time
dependent pressure table as a boundary condition for the Best Estimate Large Break
LOCA analysis.
15.4.1.6A Unit 1 Reference Transient Description A series of WCOBRA/TRAC calculations is performed to determine the PCT effect of variations in key LOCA parameters. An initial transient calculation is performed in which
several parameters are set at their assumed bounding (most limiting) values in order to
calculate a conservative PCT response to a large break LOCA. The results of these
confirmatory runs, as well as the limiting pl ant determination runs, are incorporated into a final calculation that is referred to as the reference transient. The Unit 1 reference transient models a DECLG break that assumed the conditions listed in Table 15.4.1-3A
and includes the Loss of Offsite P ower (LOOP) assumption that was shown to produce
more limiting PCT results than the offsite power available assumption. The reference
transient calculation was performed with several parameters set at their bounding DCPP UNITS 1 &
2 FSAR UPDATE 15.4-14 Revision 23 December 2016 values in order to calculate a relatively high PCT. Single parameter variation studies based on the reference transient were performed to assess which parameters have a
significant effect on the PCT results. The results of these studies are presented in
Section 15.4.1.7A. The reference transient is the basis for the uncertainty calculations
necessary to establish the Unit 1 PCT95%. 15.4.1.7A Unit 1 Sensitivity Studies A large number of single parameter sensitivity calculations of key LOCA parameters
was performed to determine the PCT effect on the LBLOCA transient. These
calculations are required as part of the approved Best Estimate LOCA methodology (Reference 56) to develop data for use in the uncertainty evaluation. For each
sensitivity study, a comparison between the reference transient results and the
sensitivity transient results was made. These single parameter sensitivity calculations
were determined to remain applicable for the Unit 1 reanalysis methodology, as applied (Reference 67).
The results of a small sample of these sensitivity studies performed for the original
analysis (Reference 60) are summarized in Table 15.4.1-4A. The results of the entire
array of sensitivity studies are included in Reference 60. The Unit 1 reanalysis is
documented in Reference 67. The conclusions of the confirmatory cases were
determined to remain the same (i.e., limiting direction of conservatism).
15.4.1.7A.1 Unit 1 Initial Condition Sensitivity Studies Several calculations were performed to evaluate the PCT effect of changes in the initial conditions on the LBLOCA transient. These calculations modeled single parameter variations in key initial plant conditions over the expected ranges of operation, including
T AVG , RCS pressure, and ECCS temperatures, pressures, and volumes. The results of these studies are presented in Section 6 of Reference 60.
The results of these sensitivity studies were used to develop uncertainty distributions for
the blowdown, first, and second reflood peaks. The uncertainty distributions resulting from the initial conditions, PCT ICi , are used in the overall PCT uncertainty evaluation to determine the final estimate of PCT 95%. 15.4.1.7A.2 Unit 1 Power Distribution Sensitivity Studies Several calculations were performed to evaluate the PCT effect of changes in power
distributions on the LBLOCA transient. The approved methodology was used to
develop a run matrix of peak linear heat rate relative to the core average, maximum
relative rod power, relative power in the bottom third of the core, and relative power in
the middle third of the core, as the power distribution parameters to be considered.
These calculations modeled single parameter variations as well as multiple parameter variations. The results of these studies indicate that power distributions with peak DCPP UNITS 1 &
2 FSAR UPDATE 15.4-15 Revision 23 December 2016 powers skewed to the top of the core produced the most limiting PCTs. These results are presented in Section 7 of Reference 60.
The results of these sensitivity studies were used to develop response surfaces, which are used to predict the PCT due to changes in power distributions for the blowdown, first, and second reflood peaks. The uncertainty distributions resulting from the power
distributions, PCT PDi , are used in the overall PCT uncertainty evaluation to determine the final estimate of PCT 95%. 15.4.1.7A.3 Unit 1 Global Model Sensitivity Studies Several calculations were performed to evaluate the PCT effect of changes in global
models on the LBLOCA transient. Reference 56 provides a run matrix of break
discharge coefficient, broken cold leg resistance, and condensation rate as the global
models to be considered for the double-ended guillotine break.
These calculations modeled single parameter variations as well as multiple parameter variations. The
limiting split break size was also i dentified using the approved methodology (Reference 56). These results are presented in Section 8 of Reference 60.
The results of these sensitivity studies were used to develop response surfaces, which are used to predict the PCT due to changes in global models for the DECLG blowdown, first, and second reflood peaks. The uncertainty distribution resulting from
the global models, PCTMODi , is used in the overall PCT uncertainty evaluation to determine the final estimate of PCT 95%. These single parameter sensitivity calculations were determined to remain applicable
for the Unit 1 reanalysis methodology, as applied (Reference 67).
15.4.1.7A.4 Unit 1 Overall PCT Uncertainty Evaluation and Results
The equation used to initially estimate the 95 percentile PCT (PCT i of Equation 15.4.1-1) was presented in Section 15.4.1.4A. Each of the uncertainty elements (PCTICi , PCT PDi , PCTMODi) is considered to be independent of each other.
Each element includes a correction or bias, which is added to PCT REFi to move it closer to the expected, or average, PCT.
The bias from each element has an uncertainty associated with the methods used to derive the bias.
Each bias component of the uncertainty elements is considered a random variable, whose uncertainty distribution is obtained directly, or is obtained from the uncertainty of
the parameters of which the bias is a function. Since PCTi is the sum of these biases, it
also becomes a random variable. Separate initial PCT frequency distributions are
constructed as follows for the DECLG break and the limiting split break:
(1) Generate a random value of each uncertainty element (PCT IC , PCT PD , PCTMOD)
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-16 Revision 23 December 2016 (2) Calculate the resulting PCT using Equation 15.4.1-1 (3) Repeat the process many times to generate a histogram of PCTs The results of this assessment showed the DECLG break to be the limiting break type.
A final verification step is performed to quantify the bias and uncertainty resulting from
the superposition assumption (i.e., the assumption that the major uncertainty elements
are independent). Several additional WCOBRA/TRAC calculations are performed in which variations in parameters from each of the three uncertainty elements are modeled
for the DECLG break. These predictions are compared to the predictions based on
Equation 15.4.1-1, and additional biases and uncertainties are applied where appropriate.
The superposition assumption verification step was performed for the Unit 1 reanalysis (Reference 67). These calculations resulted in an adjustment of the bias and
uncertainty that is required for the reanalysis methodology.
The estimate of the PCT at 95 percent probability is determined by finding that PCT
below which 95 percent of the calculated PCTs reside. This estimate is the licensing basis PCT, under the revised ECCS rule (10 CFR 50.46). The results of the Best Estimate LBLOCA analysis are presented in Table 15.4.1-2A. The difference between
the 95 percentile PCT and the average PCT increases with each subsequent PCT
period, due to propagation of uncertainties.
15.4.1.8A Unit 1 Additional Evaluations Zircaloy Clad Fuel: An evaluation of Zircaloy clad fuel has shown that the Zircaloy clad
fuel is bounded by the results of ZIRLO clad fuel analysis.
IFBA Fuel: An evaluation of IFBA fuel has shown that the IFBA fuel is bounded by the
results of the non-IFBA fuel analysis.
T AVG Coastdown: An end-of-cycle, full power T AVG coastdown at 565
°F evaluation was performed and concluded that there would be no adverse effect on the Best Estimate
LBLOCA analysis as a T AVG window between 565
°F and 577.3
°F was explicitly modeled in the Best Estimate LBLOCA analysis.
These evaluations have been shown to continue to apply for the Unit 1 reanalysis (Reference 67).
15.4.1.9A Unit 1 10 CFR 50.46 Results It must be demonstrated that there is a high l evel of probability that the limits set forth in 10 CFR 50.46 are met. The demonstration that these limits are met is as follows:
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-17 Revision 23 December 2016 (1) There is a high level of probability that the PCT shall not exceed 2200°F.
The 95 th percentile PCT results presented in Table 15.4.1-2A indicate that this regulatory limit has been met.
(2) The local maximum oxidation (LMO) calculated in the original BELOCA analysis results (Reference 60) is based on a limiting PCT transient that is
in excess of the Unit 1 reanalysis 95 percentile PCT and remains
bounding for Unit 1. Based on this original conservative PCT transient, a
LMO of 11 percent was calculated, which meets the 10 CFR 50.46
acceptance criterion (b)(2), i.e., Local Maximum Oxidation of the cladding
less than 17 percent, remains bounding for Unit 1, and is presented as an
upper bound in Table 15.4.1-2A.
(3) The maximum core wide oxidation (CWO) determined in the original BELOCA analysis results (Reference 60) was based on limiting fuel
temperatures that exceed those in the Unit 1 reanalysis and remain
bounding for Unit 1. Based on these original conservative fuel
temperatures, the total amount of hydrogen generated (i.e., CWO) , is
0.0089 times (0.89 percent) the maximum theoretical amount, which
meets the 10 CFR 50.46 acceptance criterion (b)(3), i.e., Core-Wide Oxidation less than 1 percent, remains bounding for Unit 1, and is presented as an upper bound in Table 15.4.1-2A.
(4) Criterion (b)(4) 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. The approved methodology (Reference 56) specifies that effects of LOCA and seismic loads on core geometry do not need to be considered unless grid crushing
extends beyond the assemblies in the low-power channel as defined in the
DCPP WCOBRA/TRAC model. This situati on has not been calculated to occur for DCPP Unit 1. Therefore, acceptance criterion (b)(4) is satisfied.
(5) The approved Westinghouse position on criterion (b)(5) is that this requirement is satisfied if a coolable geometry is maintained, and the core
remains subcritical following the LOCA (Reference 56). This position is
independent from and unaffected by the use of best-estimate LOCA
methodology.
15.4.1.10A Unit 1 Plant Operating Range The expected PCT and associated uncertainty presented above for Unit 1 are valid for a range of plant operating conditions. Many parameters in the reference transient
calculation are at nominal values. The range of variation of the operating parameters
has been accounted for in the estimated PCT uncertainty. Table 15.4.1-7A summarizes
the operating ranges for Unit 1. Note that Figure 15.4.1-15A illustrates the axial power
distribution limits that were analyzed and are verified on a cycle-specific basis.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-18 Revision 23 December 2016 Table 15.4.1-5A summarizes the LBLOCA containment data used for calculating containment back pressure. If plant operation is maintained within the plant operating
ranges presented in Table 15.4.1-7A, the LOCA analyses presented in this section are
considered to be valid.
15.4.1.4B Unit 2 Best Estimate Large Break LOCA Evaluation Model The thermal-hydraulic computer code, which was reviewed and approved for the
calculation of fluid and thermal conditions in a PWR during a large break LOCA, is
WCOBRA/TRAC Version MOD7A, Rev. 1 (Reference 56). Westinghouse has since developed an alternative uncertainty methodology called ASTRUM, which stands for Automated Statistical Treatment of Uncertainty Method (Reference 69). This method is still based on the "Code Qualification Document" (CQD) methodology (Reference 56).
The ASTRUM methodology replaces the response surface technique with a statistical
sampling method where the uncertainty parameters are simultaneously sampled for each case. The ASTRUM methodology has received NRC approval for referencing in
licensing calculations (SER appended to Reference 69). The WCOBRA/TRAC MOD7A, Revision 6, is an evolution of Revision 1 that includes logic to facilitate the automation
aspects of ASTRUM, user conveniences, and error corrections. WCOBRA/TRAC MOD7A, Revision 6, is documented in Reference 69.
A detailed assessment of the computer code WCOBRA/TRAC was made through comparisons with experimental data. These assessments were used to develop
quantitative estimates of the codes ability to predict key physical phenomena in a PWR large break LOCA. Modeling of a PWR introduces additional uncertainties that are identified and quantified in the plant-specific analysis.
The final step in application of the best-estimate methodology for Unit 2, in which all
uncertainties of the LOCA parameters are accounted for to estimate a PCT, local maximum oxidation (LMO), and core-wide oxidatio n (CWO) at 95-percent probability, is described below.
(1) Plant Model Development In this step, a WCOBRA/TRAC model of the plant is developed. A high level of noding detail is used in order to provide an accurate simulation of
the transient. However, specific guidelin es 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 DCPP UNITS 1 &
2 FSAR UPDATE 15.4-19 Revision 23 December 2016 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. 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 uncertainties.
(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 69.
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.
15.4.1.5B Unit 2 Containment Backpressure A conservatively bounding minimum containment back pressure (Figure 15.4.1-14B) is
calculated using the methods and assumptions described in Reference 2, Appendix A.
Containment back pressure is calculated using the COCO code (Reference 61), the
input parameters presented in Table 15.4.1-5B, mass and energy releases from the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-20 Revision 23 December 2016 WCOBRA/TRAC calculation, and the stru ctural heat sinks presented in Table 15.4.1-5A. Input parameters used for the Unit 2 containment backpressure calculation are presented in Table 15.4.1-5B. This minimum containment back pressure is modeled
using a time dependent pressure table as a boundary condition for the Best Estimate
Large Break LOCA analysis.
15.4.1.6B Unit 2 Confirmatory Studies A few confirmatory studies were performed to establish the limiting conditions for the
uncertainty evaluation. In the confirmatory studies performed, key LOCA parameters
are varied over a range and the impact on the peak cladding temperature is assessed.
The results for the confirmatory studies are summarized in Table 15.4.1-4B. In
summary, the limiting conditions for the plant at the time the design basis accident is
postulated to occur are reflected in the final reference transient. These limiting
conditions are:
(1) Loss of offsite power (2) High RCS average temperature (3) High steam generator tube plugging of 15 percent (4) High average power fraction in the assemblies on the core periphery (fraction of power in outer assemblies = 0.8)
15.4.1.7B Unit 2 Uncertainty Evaluation The ASTRUM methodology (Reference 69) differs from the previously approved
Westinghouse Best-Estimate methodology (Reference 56) primarily in the statistical
technique used to make a singular probabilistic statement with regard to the
conformance of the system under analysis to the regulatory requirement of
The ASTRUM methodology applies a non-parametric statistical technique to generate output e.g., PCT, LMO, and CWO from a combination of WCOBRA/TRAC and HOTSPOT (Reference 68) calculations. These calculations are performed by applying
a direct, random Monte Carlo sampling to generate the input for the WCOBRA/TRAC and HOTSPOT computer codes.
This approach allows the formulation of a simple singular statement of uncertainty in the
form of a tolerance interval for the numerical acceptance criteria of 10 CFR 50.46.
Based on the non-parametric statistical approach, the number of Monte Carlo runs is
only a function of the tolerance interval and associated confidence level required to
meet the desired level of safety.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-21 Revision 23 December 2016 15.4.1.8B Unit 2 Limiting PCT Transient Description The DCPP Unit 2 PCT-limiting transient is a DECLG break which analyzes conditions that fall within those listed in Table 15.4.1-7B. The sequence of events following is
presented in Table 15.4.1-1B. The PCT-limiting case was chosen to show a
conservative representation of the response to a large break LOCA.
15.4.1.9B Unit 2 10 CFR 50.46 Results It must be demonstrated that there is a high l evel of probability that the limits set forth in 10 CFR 50.46 are met. The demonstration that these limits are met is as follows:
(1) Because the resulting PCT for the limiting case is 1872 °F, which represents a bounding estimate of the 95 th percentile PCT at the 95-percent confidence level, the analysis confirms that 10 CFR 50.46
acceptance criterion (b)(1), i.e., P eak Cladding Temperature less than 2200 °F, is met. The results are shown in Table 15.4.1-2B.
(2) Because the resulting local maximum oxidation (LMO) for the limiting case is 1.64 percent, which represents a bounding estimate of the 95 th percentile LMO at the 95-percent confidence level, the analysis confirms
that 10 CFR 50.46 acceptance criterion (b)(2), i.e., Local Maximum
Oxidation of the cladding less than 17 percent, is met. The results are
shown in Table 15.4.1-2B.
(3) The limiting hot fuel assembly rod has a calculated maximum oxidation of 0.17 percent. Because this is the hottest fuel rod within the core, the
calculated maximum oxidation for any other fuel rod would be less than
this value. For the low power peripheral fuel assemblies, the calculated
oxidation would be significantly le ss than this maximum value. The core wide oxidation (CWO) is essentially the sum of all calculated maximum oxidation values for all of the fuel rods within the core. Therefore, a detailed CWO calculation is not needed because the calculated sum will
always be less than 0.17 percent. Because the resulting CWO is
conservatively assumed to be 0.17 percent, which represents a bounding
estimate of the 95 th percentile CWO at the 95-percent confidence level, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3), i.e.,
Core-Wide Oxidation less than 1 percent, is met. The results are shown in Table 15.4.1-2B.
(4) Criterion (b)(4) 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. The approved
methodology (Reference 56) specifies that effects of LOCA and seismic
loads on core geometry do not need to be considered unless grid crushing
extends beyond the assemblies in the low-power channel as defined in the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-22 Revision 23 December 2016 DCPP WCOBRA/TRAC model. This situati on has not been calculated to occur for DCPP Unit 2. Therefore, acceptance criterion (b)(4) is satisfied.
(5) The approved Westinghouse position on Criterion (b)(5) is that this requirement is satisfied if a coolable geometry is maintained, and the core remains subcritical following the LOCA (Reference 56). This position is
independent from and unaffected by the use of best-estimate LOCA
methodology.
15.4.1.10B Unit 2 Plant Operating Range The accepted PCT and its uncertainty developed previously are valid for a range of
Unit 2 plant operating conditions. The range of variation of the operating parameters
has been accounted for in the uncertainty evaluation. Table 15.4.1-7B summarizes the
operating ranges for DCPP Unit 2 as defined for the proposed operating conditions, which are supported by the Best-Estimate LBLOCA analysis. Table 15.4.1-5B
summarizes the LBLOCA containment data used for calculating containment back
pressure. It should be noted that other non-LBLOCA analyses may not support these
ranges. If operation is maintained within these ranges, the LBLOCA results developed in this report 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 (T avg). 15.4.1.11 Conclusions (Common)
15.4.1.11.1 10 CFR 50.46 Acceptance Criteria It must be demonstrated that there is a high l evel of probability that the limits set forth in 10 CFR 50.46 are met. The demonstration that these limits are met is as follows:
(1) The limiting PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level such that the
analysis confirms that 10 CFR 50.46 acceptance criterion (b)(1), i.e.,
Peak Cladding Temperature less than 2200 ºF, is demonstrated.
(2) 10 CFR 50.46 acceptance criterion (b)(2), requires that the maximum calculated reduction in fuel cladding thickness at any location in the core
due to the zirconium and water (Zr-H 2 O) reaction shall be less than 17 percent of the original cladding thickness. Because the Zr-H 2 O reaction essentially oxidizes the fuel cladding and generates hydrogen as
a by-product, the reduction in cladding thickness is evaluated based on
the amount of H 2 generated (i.e., oxidation) at a given core location. The BELOCA methodology calculates the local maximum oxidation (LMO), which corresponds to a bounding estimate of the 95th percentile LMO at
the 95-percent confidence level such that the analysis confirms that the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-23 Revision 23 December 2016 10 CFR 50.46 acceptance criterion (b)(2), i.e., Local Maximum Oxidation of the Cladding Less than 17 percent, is demonstrated.
(3) 10 CFR 50.46 acceptance criterion (b)(3) requires that the total quantity of fuel cladding oxidize d due to the Zr-H 2 O reaction shall be less than 1 percent, which is verified by ensuring the total calculated amount of H 2 generated is less than 1 percent of the theoretical maximum possible if all
of the fuel cladding in the core was oxidized. The BELOCA methodology
calculates the limiting core wide oxi dation (CWO) which corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level such that the analysis confirms that 10 CFR 50.46
acceptance criterion (b)(3), i.e., Core-Wide Oxidation Less than 1
percent, is demonstrated.
(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. The
approved methodology (Reference 56) specifies that effects of LOCA and
seismic loads on core geometry do not need to be considered unless fuel
grid crushing extends beyond the assemblies representing the low-power
channel.
(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.
The approved Westinghouse position on this criterion is that this
requirement is satisfied if a coolable geometry is maintained, and the core
remains subcritical following the LOCA (Reference 56). This position is
independent from and unaffected by the use of best-estimate LOCA
methodology.
15.4.1.11.2 Radiological
Section 15.5.17.11 concludes that the resulting potentia l exposures have been found to be lower than the applicable guidelines and limits specified in 10 CFR Part 100.
15.4.2 MAJOR SECONDARY SYST EM PIPE RUPTURE Three major secondary system pipe ruptures are analyzed in this section: rupture of a
main steam line at hot zero power, rupture of a main feedwater pipe, and rupture of a
main steam line at power. The time sequence of events for each of these events is
provided in Table 15.4-8.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-24 Revision 23 December 2016 15.4.2.1 Rupture of a Main Steam Line at Hot Zero Power 15.4.2.1.1 Acceptance Criteria The following limiting criteria are applicable for a main steam line rupture at hot zero power: 15.4.2.1.1.1 Fuel Damage Criteria Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capability. This is conservatively demonstrated by meeting the following criteria:
(1) DNB will not occur on the lead rod with at least a 95 percent probability at a 95 percent confidence level. The minimum DNBR must not go below
the applicable limit value of 1.45 at any time during the transient.
15.4.2.1.1.2 Radiological Criteria (1) The resulting potential exposures to individual members of the public and to the general population shall be lower than the applicable guidelines and
limits specified in 10 CFR Part 100.
15.4.2.1.2 Identification of Causes and Accident Description The steam release from a rupture of a main steam pipe would result in an initial
increase in steam flow that 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 a positive reactivity insertion and subsequent reduction of core
shutdown margin. If the most reactive 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 steam pipe rupture is a
potential problem mainly because of the high power peaking factors that 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 SIS and accumulators.
In order to allow for routine plant heatups an d cooldowns, plant procedures allow the SIS to be blocked per permissive P-11, provided that the RCS boron concentration is
maintained at a value greater than or equal to the cold shutdown margin requirement.
As discussed in Reference 63, this additional shutdown margin ensures that there
would be no return to power for a steam pipe rupture such that the analysis of a rupture of a steam line at hot zero power remains bounding.
The analysis of a main steam pipe rupture is performed to demonst rate that the following criteria are satisfied:
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-25 Revision 23 December 2016 (1) Assuming a stuck RCCA, with or without offsite power, and assuming a single failure in the engineered safety features (ESF) there is no consequential damage to the primary system and the core remains in
place and intact.
(2) Energy release to containment from the worst steam pipe break does not cause failure of the containment structure (refer to Appendix 6.2D).
Although DNB and possible cladding perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis, in fact, shows that the DNB design
basis is met for any rupture assuming the most reactive assembly stuck in its fully
withdrawn position.
The following functions provide protection for a steam line rupture:
(1) SIS actuation from any of the following:
(a) Two-out-of-four low pressurizer pressure signals (b) Two-out-of-three low steam line pressure signals in any one loop (c) Two-out-of-three high containment pressure signals (2) The overpower reactor trips (neutron reactor trip, and the reactor trip occurring in conjunction with receipt of the safety injection signal.
(3) Redundant isolation of the main feedwater lines: sustained high feedwater flow would cause additiona l cooldown. Therefore, a safety injection signal will rapidly close all MFRVs, trip the main feedwater pumps, and close the MFIVs that backup the control valves.
(4) Closure of the fast acting main steam line isolation valves on: (refer to Figure 7.2-1 and the Technical Specifications (Reference 30))
(a) Two-out-of-three low steam line pressure signals in any one loop (b) Two-out-of-four high-high containment pressure (c) Two-out-of-three high negative steam line pressure rate signals in any one loop (used only during cooldown and heatup operations)
The fast-acting isolation valves are provided in each main steam line and will fully close
within 10 seconds of a large steam line break. 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 blow down even if one of the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-26 Revision 23 December 2016 isolation valves fails to close. A description of steam line isolation is included in Chapter 10.
The effective throat area of the integral flow restrictors in the steam generators is
1.388 ft 2 , which is considerably smaller than the area of the main steam pipe. These restrictors serve to limit the maximum stea m flow for any break at any location.
15.4.2.1.3 Analysis of Effects and Consequences The analysis of the steam pipe rupture has been performed to determine:
(1) The plant transient conditions, including core heat flux and RCS temperature and pressure resulting from the cooldown following the steam
line break. The RETRAN-02W code (Reference 70) has been used.
(2) The thermal and hydraulic behavior of the core following a steam line break. A detailed thermal and hydraulic digital-computer code, THINC (refer to Section 1.6.1, Item 28, and Section 4.4.3), has been used to determine if DNB occurs for the core conditions computed in (1) above.
The following conditions were assumed to exist at the time of a main steam line break
accident.
(1) End of life (EOL) shutdown margin at no-load, equilibrium xenon conditions, and the most reactive assembly 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 steam line break accident will not lead to a more adverse condition than the case
analyzed.
(2) The negative moderator coefficient corresponds to the EOL rodded core with the most reactive rod in the fully withdrawn position. The variation of
the coefficient with temperature and pressure has been included. The k eff versus temperature at 1050 psia corresponding to the negative moderator
temperature coefficient, plus the Doppler temperature effect used is
shown in Figure 15.4.2-2. The effect of power generation in the core on
overall reactivity is shown in Figure 15.4.2-1.
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. To verify the conservatism of this method, the reactivity as well as the power distribution was checked with the advanced
nodal code core model (refer to Section 4.3.3.10.3). These core analyses considered the Doppler reactivity from the high fuel temperature near the
stuck RCCA, moderator feedback from the high water enthalpy near the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-27 Revision 23 December 2016 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 confirmed that the reactivity feedback model employed in the RETRAN-
02W kinetics analysis was consistent with the core analysis and the
overall analysis is conservative.
(3) The modeling of the SIS in RETRAN-02W is described in Reference 70.
The minimum boric acid solution concentration of 2300 ppm in the RWST
is assumed. The SIS piping downstream of the RWST isolation valves is assumed to contain no boron (0 ppm), which delays the delivery of boron to the reactor coolant loops from the RWST water. With this conservative
assumption, the SIS and accumulators combine to limit the return to
power. Cases were examined for both minimum and maximum SIS flow
rates. For the minimum SIS flow rate cases the most restrictive single failure in
the SIS is considered. The SIS flow assumed conservatively corresponds
to that delivered by only one high-head charging pump delivering full flow to the cold leg header. The charging pump (CCP1 or CCP2) is assumed
to begin providing flow to the RCS at 25 seconds after receipt of the SI
signal for the case in which offsite power is assumed available, and at
35 seconds for the case where offsite power is not available; the additional
10-second delay is assumed to start the diesels and load the necessary
safety injection equipment onto them.
For the maximum SIS flow rate cases, a flow profile was assumed that
bounds the maximum flow from two high-head charging pumps (CCP1
and CCP2) plus two intermediate-head SI pumps plus the nonsafety-
related CVCS charging pump (CCP3). A 2-second signal delay was
assumed. For this analysis, it was determined that the maximum SIS flow rate
assumption is conservative for the more limiting case with offsite power
available, due to the effect of higher SIS flow on the timing of cold leg
accumulator actuation. The cold leg accumulators provide an additional
source of borated water to the core when the RCS pressure decreases
below the actuation setpoint. The minimum accumulator boron
concentration of 2200 ppm is assumed, along with a conservatively low
actuation setpoint of 577.2 psia. Actuation of the accumulators causes a
significant influx of boron, which rapidly shuts down the reactor. Assuming
the maximum SIS flow rate slows down the rate of the RCS pressure
decrease and thus delays the accumulator actuation. If the most reactive
RCCA is assumed stuck in its fully withdrawn position after a reactor trip, there is an increased possibility that the core will become critical and
return to power. A return to power following a steam pipe rupture is a DCPP UNITS 1 &
2 FSAR UPDATE 15.4-28 Revision 23 December 2016 potential problem mainly because of the high power peaking factors that would exist assuming the most-reactive RCCA to be stuck in its fully
withdrawn position. Therefore, the limiting case presented herein
conservatively assumes a maximum SIS flow rate.
(4) Because the steam generators are equipped with integral flow restrictors with a 1.388 ft 2 throat area, any rupture with a break greater than this size, regardless of the location, would have the same effect on the reactor as a
1.388 ft 2 break. The following two cases have been considered in determining the core power and RCS transients:
(a) Complete severance of a pipe with the plant initially at no-load conditions and with offsite power available. Full reactor coolant
flow is maintained.
(b) Complete severance of a pipe with the plant initially at no-load conditions and with offsite power unavailable. Loss of offsite
power results in reactor coolant pump coastdown.
(5) Power peaking factors corresponding to one stuck RCCA and non-uniform core inlet coolant temperatures are determined at EOL. 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 steam line break. This void in conjunction with the large
negative moderator coefficient partially offsets the effect of the stuck assembly. The power peaking factors depend on the core power, operating history, temperature, pressure, and flow.
All the cases above assume initial hot shutdown conditions at time zero, because this represents the most limiting initial condition. Should the reactor be just critical or operating at power at the time of a steam line 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 steam line 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.
However, because the initial steam generator water inventory is greatest
at no-load, the magnitude and duration of the RCS cooldown are less for
steam line breaks occurring at power.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-29 Revision 23 December 2016 (6) In computing the steam flow during a steam line break, the Moody Curve (Reference 16) for fl/D = 0 is used.
(7) Perfect moisture separation in the steam generator is assumed. This assumption leads to conservative results because, in fact, considerable water would be discharged. Water carryover would reduce the magnitude
of the temperature decrease in the core.
(8) To maximize the primary-to-secondary heat transfer rate, 0 percent steam generator tube plugging is assumed.
(9) All main and auxiliary feedwater pumps are assumed to be operating at full capacity when the rupture occurs. This assumption maximizes the
cooldown. A conservatively high auxiliary feedwater flow rate of
1700 gpm at a minimum temperature of 60ºF is assumed to be delivered
to the affected steam generator. Main feedwater is isolated 64 seconds
following the SI signal by closure of the MFIVs. No credit is taken for the faster-closing MFRVs. Auxiliary feedwater continues for the duration of the transient.
(10) The effect of heat transferred from thick metal in the reactor coolant system and the steam generators is not included in the cases analyzed.
The heat transferred from these sources would be a net benefit because it
would slow the cooldown of the RCS.
15.4.2.1.4 Results The double-ended rupture of a main steam line at zero power was analyzed for both
Units 1 and 2; however, only the results from the slightly more limiting Unit 1 cases are
presented. Unit 2 results are similar. The time sequence of events, both with and
without offsite power available for Unit 1, are presented in Table 15.4-8.
Figures 15.4.2-4 through 15.4.2-6 show the plant response following a main steam pipe
rupture. Offsite power is assumed to be available such that full reactor coolant flow
exists. The transient shown assumes an uncontrolled steam release from only one
Figures 15.4.2-7 through 15.4.2-9 show the plant response for the case with a loss of
offsite power. This assumption results in a coastdown of the reactor coolant pumps. In
this case, the core power increases at a slower rate and reaches a lower peak value
than in the 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.
It should be noted that following a steam line break only one steam generator blows
down completely. Thus, the remaining steam generators are still available for DCPP UNITS 1 &
2 FSAR UPDATE 15.4-30 Revision 23 December 2016 dissipation of decay heat after the initial transient is over. In the case with loss of offsite power, this heat would be removed to the atmosphere via the main steam safety valves.
15.4.2.1.5 Conclusions The analysis demonstrates the acceptance criteria are met as follows:
15.4.2.1.5.1 Fuel Limits Based on the results of the analysis, the core will remain in place and intact with no loss
of core cooling capability.
A DNB analysis was performed for the limiting steam line break case with offsite power
available as described above. The analysis demonstrated that the minimum DNBR
remains well above the limit value of 1.45.
Therefore, the DNB design basis is met for the steam line break event initiated from zero power.
15.4.2.1.5.2 Radiological Section 15.5.18 concludes that potential expo sures from major steam line ruptures will be well below the guideline levels specified in 10 CFR Part 100.
15.4.2.2 Major Rupture of a Main Feedwater Pipe
15.4.2.2.1 Acceptance Criteria
The following limiting criteria are applicabl e for a main feedwater pipe rupture:
15.4.2.2.1.1 Fuel Damage Criteria Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capability. This is
conservatively demonstrated by meeting the following criteria:
(1) With respect to fuel damage due to dryout where the water level in the vessel drops below the top of the core, criterion that no bulk boiling occurs in the primary coolant system prior to event turnaround is applied.
Turnaround is defined as the point when the heat removal capability of the
steam generators, being fed by auxiliary feedwater (AFW), exceeds NSSS
heat generation.
15.4.2.2.1.2 Maximum RCS and Main Steam System Pressure Requirements:
The maximum pressure in the RCS and main steam system should be maintained
below 110 percent of the design value, 2748.5 psia and 1208.5 psia, respectively.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-31 Revision 23 December 2016 15.4.2.2.1.3 Radiological Criteria The resulting potential exposures to individu al members of the public and to the general population shall be lower than the applicabl e guidelines and limits specified in 10 CFR Part 100.
15.4.2.2.2 Identification of Causes and Accident Description A major feedwater line rupture is defined as a break in a feedwater pipe 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. Further, a break in this location could preclude
the subsequent addition of AFW to the af fected steam generator. (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 Section 15.2.8).
Depending on the size of the break and the plant operating conditions at the time of the
break, the break could cause either an RCS cooldown (by excessive energy discharge
through the break), or an RCS heatup. Potential RCS cooldown resulting from a
secondary pipe rupture is evaluated in Section 15.4.2.1. Therefore, only the RCS
heatup effects are evaluated for a feedline rupture.
A feedline rupture reduces the ability to remove heat generated by the core from the
RCS for the following reasons:
(1) Feedwater to the steam generators is reduced. Since feedwater is subcooled, its loss may cause reactor coolant temperatures to increase
prior to reactor trip
(2) Liquid in the steam generator may be discharged through the break, and would then not be available for decay heat removal after trip
(3) The break may be large enough to prevent the addition of any main feedwater after trip
The following provide the necessary protection against a main feedwater line rupture:
(1) A reactor trip on any of the following conditions:
(a) High pressurizer pressure (b) Overtemperature (c) Low-low steam generator water level in any steam generator (d) Safety injection signals from any of the following:
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-32 Revision 23 December 2016
- Low steam line pressure
- High containment pressure (Refer to Chapter 7 for a description of the actuation system.)
(2) An AFW system to provide an assured source of feedwater to the steam generators for decay heat removal (Refer to Chapter 6 for a description of the AFW system.)
15.4.2.2.3 Analysis of Effects and Consequences The feedline break transient is analyzed using the RETRAN-02W computer code
described in Reference 70. The RETRAN-02W model simulates the reactor coolant
system, neutron kinetics, pressurizer, pressu rizer relief and safety valves, pressurizer heaters, pressurizer spray, steam generators, feedwater system, and main steam safety valves. The code computes pertinent plant variables including steam generator mass, pressurizer water volume, reactor coolant average temperature, reactor coolant system
pressure, and steam generator pressure.
The feedline rupture analysis methodology presented in Section 15.4.2.2 is not intended to minimize the predicted time to pressurizer filling, as this scenario is evaluated in Section 15.4.2.4.
Major assumptions are:
(1) The plant is initially operating at 102 percent of the NSSS rating, including a conservatively large RCP heat of 20 MWt for the case with offsite power available and a nominal (minimum guaranteed) RCP heat of 14 M W t for the case without offsite power available. These assumptions maximize
the primary side heat that must be removed for each case.
(2) Initial reactor coolant average temperature is 5.0°F above the nominal value, and the initial pressurizer pressure is 60 psi above its nominal
value. (3) The initial pressurizer level is set to the nominal full power programmed level plus an uncertainty of +5.7 percent span for Diablo Canyon Units 1
and 2, resulting in an initial pressurizer level of 66.4 percent span and
66.8 percent span, respectively. Initial steam generator water level is at
75 percent narrow range span (NRS) in the faulted steam generator, and
at 55 percent NRS in the intact steam generators.
(4) No credit is taken for the pressurizer power-operated relief valves or pressurizer spray.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-33 Revision 23 December 2016 (5) No credit is taken for the high pressurizer pressure reactor trip.
(6) Main feed to all steam generators is assumed to stop at the time the break occurs (all main feedwater spills out through the break).
(7) The break discharge quality is calculated by RETRAN-02W as a function of pressure and temperature.
(8) Reactor trip is assumed to be initiated when the low-low level trip setpoint in the ruptured steam generator is reached. A low-low level setpoint of 0 percent NRS is assumed.
(9) A double-ended break area of 0.5184 ft 2 is assumed. A break area of 0.5184 ft 2 corresponds to the flow area of the reducer leading to the feedring, and is the largest effective area of flow out of the steam
generators for the feedline break event. This minimizes the steam
generator fluid inventory available for removal of long-term decay heat and
stored energy following reactor trip, and thereby maximizes the resultant
heatup of the reactor coolant.
(10) No credit is taken for heat energy deposited in RCS metal during the RCS heatup. (11) No credit is taken for charging or letdown.
(12) The steam generator heat transfer correlation for the steam generator tubes is automatically adjusted by RETRAN-02W as the shell-side
inventory decreases.
(13) Conservative core residual heat generation based on the 1979 ANS 5.1 (Reference 32) decay heat standard plus uncertainty was used for
calculation of residual decay heat levels.
(14) The AFW is assumed to be initiated 10 minutes after the trip with a feed rate of 390 gpm
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-34 Revision 23 December 2016 15.4.2.2.4 Results Analyses were performed for both Units 1 and 2 separately; the most limiting case with
offsite power and the corresponding case without offsite power are presented.
Results for two feedline break cases are presented. Results for a case in which offsite
power is assumed to be available are presented in Section 15.4.2.2.4.1. Results for a
case in which offsite power is assumed to be lost following reactor trip are presented in
Section 15.4.2.2.4.2. The calcula ted sequence of events for both cases is listed in Table 15.4-8.
15.4.2.2.4.1 Feedline Rupture with Offsite Power Available The system response following a feedwater line rupture, assuming offsite power is
available, is presented in Figures 15.4.2-10 through 15.4.2-13. Results presented in
Figures 15.4.2-11 and 15.4.2-13 show that pressures in the RCS and main steam
system remain below 110 percent of the design pressures, 2748.5 psia and
1208.5 psia, respectively. Pressurizer pressure decreases after reactor trip on low-low
steam generator water level due to the reduction of heat input. Following this initial
decrease, pressurizer pressure increases to the pressurizer safety valve setpoint. This
increase in pressure is the result of coolant expansion caused by the reduction in heat
transfer capability in the steam generators. Figure 15.4.2-11 indicates a pressurizer
water volume equivalent to a water-solid condition; however, this is not an acceptance
criteria for the analysis. Pressurizer filling during a main feedwater pipe rupture event is evaluated in Section 15.4.2.4. At approximately 5900 seconds, decay heat generation decreases to a level such that the total RCS heat generation (decay heat plus pump heat) is less than auxiliary feedwater heat removal capability, and RCS pressure and temperature begin to decrease.
The results show that the core remains covered at all times and that no boiling occurs in
the reactor coolant loops.
15.4.2.2.4.2 Feedline Rupture with Offsite Power Unavailable The system response following a feedwater line rupture without offsite power available
is similar to the case with offsite power available. However, as a result of the loss of
offsite power (assumed to occur at reactor tr ip), the reactor coolant pumps coast down.
This results in a reduction in total RCS heat generation by the amount produced by
pump operation.
The reduction in total RCS heat generation produces a milder transient than in the case
where offsite power is available. Results presented in Figures 15.4.2-14 through
15.4.2-17 show that pressure in the RCS and main steam system remain below
110 percent of the design pressures, 2748.5 psia and 1208.5 psia, respectively.
Pressurizer pressure decreases after reactor trip on low-low steam generator water
level due to the reduction of heat input. Following this initial decrease, pressurizer DCPP UNITS 1 &
2 FSAR UPDATE 15.4-35 Revision 23 December 2016 pressure increases to a peak pressure of 2426 psia at 106 seconds. This increase in pressure is the result of coolant expansion caused by the reduction in heat transfer
capability in the steam generators. Figure 15.4.2-15 shows that the water volume in the
pressurizer increases in response to the heatup, but does not fill the pressurizer. At
approximately 2200 seconds, decay heat generation de creases to a level less than the auxiliary feedwater heat removal capability, and RCS temperature begins to decrease.
The results show that the core remains covered at all times and that no boiling occurs in
the reactor coolant loops.
15.4.2.2.5 Conclusions Results of the analysis show that for the postulated feedline rupture, the assumed AFW
system capacity is adequate to remove decay heat, to prevent overpressurizing the
RCS, and to prevent uncovering the reactor core. The analysis documents that the
acceptance criteria for a postulated feedline rupture are met as follows:
15.4.2.2.5.1 Fuel Damage Any fuel damage calculated to occur is of sufficiently limited extent that the core will
remain in place and intact with no loss of core cooling capability. This is conservatively
demonstrated by Figures 15.4.2-12 and 15.4.2-16 that show no bulk boiling occurs in
the primary coolant system prior to event turnaround.
15.4.2.2.5.2 Maximum RCS and Main Steam System Pressure As shown in Figures 15.4.2-11 and 15.4.2-13, the maximum pressure in the RCS and main steam system is maintained below 110 percent of the design value, 2748.5 psia and 1208.5 psia, respectively.
15.4.2.2.5.3 Radiological Section 15.5.19 concludes that potential ex posures from major feedwater line ruptures will be well below the guide line levels specified in 10 CFR Part 100, and that the occurrence of such ruptures would not result in undue risk to the public.
15.4.2.3 Rupture of a Main Steam Line at Full Power 15.4.2.3.1 Acceptance Criteria The following limiting criteria are applicable for a main steam line rupture at full power:
15.4.2.3.1.1 Fuel Damage Criteria Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capability. This is conservatively demonstrated by meeting the following criteria:
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-36 Revision 23 December 2016 (1) DNB will not occur on the lead rod with at least a 95 percent probability at a 95 percent confidence level. The minimum DNBR must not go below the DNBR Safety Analysis Limit of 1.68/1.71 (refer to Section 4.4.4.1) at any time during the transient.
(2) The peak linear heat generation rate will not exceed a 22 kW/ft (refer to Section 4.4.4.2 and Figure 4.4-2) which would cause fuel centerline melt.
15.4.2.3.1.2 Radiological Criteria The resulting potential exposures to individu al members of the public and to the general population shall be lower than the applicabl e guidelines and limits specified in 10 CFR Part 100.
15.4.2.3.2 Identification of Causes and Accident Description A rupture in the main steam system piping from an at-power condition creates an
increased steam load, which extracts an increased amount of heat from the reactor coolant system via the steam generators. This results in a reduction in reactor coolant
system temperature and pressure. In the presence of a strong negative moderator temperature coefficient, typical of end-of-cycle conditions, the colder core inlet coolant temperature causes the core power to increase from its initial level due to the positive reactivity insertion. The power approaches a level equal to the total steam flow.
Depending on the break size, a reactor trip may occur due to overpower conditions or
as a result of a steam line break protection function actuation.
The steam system piping failure accident analysis, described in Section 15.4.2.1, is
performed assuming a hot zero power initial condition with the control rods inserted in
the core, except for the most reactive rod, w hich remains fully withdrawn out of the core.
This condition could occur while the reactor is at hot shutdown at the minimum required
shutdown margin, or after the plant has been tripped manually, or by the reactor
protection system following a steam line break from an at-power condition. For an at-
power break, the FSAR Update Section 15.4.2.1 analysis represents the limiting
condition with respect to core protection for the time period following reactor trip. The analysis of a main steam pipe rupture at po wer is performed to demonstrate that the following criteria are satisfied:
(1) Assuming a stuck RCCA and a single failure in the engineered safety features, there is no damage to the primary system and the core remains
in place and intact.
(2) Core protection is maintained prior to, and immediately following, a reactor trip, if one is required, such that the DNBR remains above the applicable
limit value for any rupture assuming the most reactive assembly stuck in
its fully withdrawn position.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-37 Revision 23 December 2016 Depending on the size of the break, this event is classified as either an ANS Condition III (infrequent fault) or Condition IV (limiting fault) event. The main steam pipe
rupture at power is protected by the same reactor protection and ESF functions as the
main steam pipe rupture at hot zero power. Although DNB and possible clad
perforation following a steam pipe rupture are not necessarily unacceptable, the
analysis shows that the calculated DNBR re mains above the applicable DNBR limit value.
15.4.2.3.3 Analysis of Effects and Consequences The analysis of the steam line rupture is performed in the following stages:
(1) The RETRAN-02W code (Reference 70) is used to calculate the nuclear power, core heat flux, and RCS temperature and pressure transients
resulting from the cooldown following the steam line break.
(2) The core radial and axi al peaking factors are determined using the thermal-hydraulic conditions from the transient analysis as input to the
nuclear core models. The THINC-IV code (refer to Section 4.4.3) is then used to calculate the DNBR for the limiting time during the transient.
This accident is analyzed with the Improved Thermal Design Procedure as described in
Reference 62.
To give conservative results in calculating the DNBR during the transient, the following
assumptions are made:
(1) Initial Conditions - The initial core power, reactor coolant temperature, and RCS pressure are assumed to be at their nominal full-power values. The full power condition is more limiting than part-power with respect to DNBR.
Uncertainties in initial conditions are include d in the DNBR limit value, as described in Reference 62.
(2) Break size - A spectrum of break sizes is analyzed. Small breaks do not result in a reactor trip; in this case core power stabilizes at an increased
level corresponding to the increased steam flow. Intermediate-size breaks may result in a reactor trip on overpower T as a result of the increasing core power. Larger break sizes result in a reactor trip soon after the break from the safety injection signal actuated by low steam line pressure, which
includes lead/lag dynamic compensation.
(3) Break flow - The steam flow out the pipe break is calculated using the Moody curve for an fL/D value of 0 (Reference 16).
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-38 Revision 23 December 2016 (4) Reactivity Coefficients - The analysis assumes maximum EOL moderator reactivity feedback and minimum Doppler-only power reactivity feedback in order to maximize the power increase following the break.
(5) Protection System - The analysis only models those reactor protection system features that would be credited for at power conditions and up to
the time a reactor trip is initiated. Section 15.4.2.1, presents the analysis
of the bounding transient following reactor trip, where engineered safety
features are actuated to mitigate the effects of a steam line break.
(6) Control Systems - The results of a main steam pipe rupture at power would be made less severe as a result of control system actuation.
Therefore, the mitigation effects of control systems have been ignored in
the analysis.
15.4.2.3.4 Results A spectrum of steam line break sizes was analyzed for each unit. The results show that
for break sizes up to 0.49 ft 2 (Unit 1) and 0.50 ft 2 (Unit 2) a reactor trip is not generated.
In this case, the event is similar to an excessive load increase event as described in
Section 15.2.12. The core reaches a new equilibrium condition at a higher power equivalent to the increased steam flow. For break sizes larger than those noted above, a reactor trip is generated within a few seconds of the break on the safety injection
signal from low steam line pressure.
The limiting case for demonstrating DNB protection is the 0.49 ft 2 (Unit 1) break, the largest break size that does not result in an early trip on low steam pressure SI
actuation. The peak linear heat rate (kW/
ft) remains below a value corresponding to fuel centerline melting. The time sequence of events for this case is shown in
Table 15.4-8. Figures 15.4.2-18 through 15.4.2-21 show the transient response.
15.4.2.3.5 Conclusions The analysis demonstrates the acceptance criteria are met as follows:
15.4.2.3.5.1 Fuel Damage Any fuel damage calculated to occur is of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capability. This is conservatively
demonstrated by the following:
(1) The analysis demonstrates that there is a large margin to the DNBR Safety Analysis Limit of 1.71/1.68 (typical cell/thimble cell).
(2) The analysis calculates that the maximum linear power meets the fuel centerline melt limit of 22.0 kW/ft.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-39 Revision 23 December 2016 The analysis concludes that the DNB and fue l centerline design bases are met for the limiting case. Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable and not precluded by the criteria, the above analysis shows that the minimum DNBR remains above the safety analysis limit.
15.4.2.3.5.2 Radiological Section 15.5.18 concludes that potential expo sures from main steam line ruptures at full power will be well below the guideline levels specified in 10 CFR Part 100, and that the occurrence of such ruptures would not result in undue risk to the public.
15.4.2.4 Major Rupture of a Main Feed water Pipe for Pressurizer Filling 15.4.2.4.1 Acceptance Criteria The acceptance criterion is to ensure the major rupture of a main feedwater pipe (hereinafter referred to as feedwater line break or FLB) for pressurizer filling event does not result in liquid water (hereinafter referred to as water) relief through the pressurizer safety valves (PSVs) in order to prevent an unisolable reactor coolant pressure boundary breach due to a PSV failing op en. This can be accomplished through appropriate operator actions and equipment design/response that mitigate the consequences of the event before water relief through the PSVs occurs.
15.4.2.4.2 Identification of Causes and Accident Description The causes and accident description for the pressurizer filling analysis of the main feedwater pipe rupture described in this section are discussed generally in Section 15.4.2.2.2. The aspects that relate specifically to pressurizer filling follow.
Following a FLB accident, secondary water level decreases in the steam generators (SGs) until auxiliary feedwater (AFW) flow is initiated, after which level will begin to recover in the SGs being fed with AFW flow (i.e., the intact SGs). Depending on the AFW flow available, there is the potential for an increase in reactor coolant temperatures in the early part of the post-trip transient, along with an increase in reactor coolant volume due to thermal expansion. Also, following initiation of the FLB accident, a low steam line pressure setpoint will be reached in the faulted loop, causing actuation of the safety injection (SI) signal and start of the two PG&E Design Class I charging pumps. The reactor coolant inventory addition from the charging flow and reactor coolant system (RCS) thermal expansion contributes to pressurizer filling.
If pressurizer filling occurs, the pressurizer power-operated relief valves (PORVs) are available to relieve water inventory from the RCS, as long as an air supply is available from instrument air to containme nt or from the PG&E Design Class I backup nitrogen accumulators. Also, since Technical Specifications define a PORV as operable with its block valve closed if the PORV can be made available for automatic pressure relief, operators may need to take action to open the block valve to enable the PORV to DCPP UNITS 1 &
2 FSAR UPDATE 15.4-40 Revision 23 December 2016 provide water relief. Using the PORVs to relieve water from the RCS precludes water relief through the PSVs, which could render the PSVs inoperable.
Mitigation of the pressurizer filling condition is complete when (1) the heat removal capability of the SGs being fed by AFW exceeds NSSS heat generation and stops thermal expansion of the RCS and (2) operator actions are taken to isolate charging flow, and subsequently stop reactor coolant pump (RCP) seal injection flow, which terminates all remaining reactor coolant inventory addition.
The pressurizer filling analysis models the long term plant response to a FLB to demonstrate that operator actions, if taken in a timely manner, preclude water relief through the PSVs. The operator actions for mitigation of a FLB accident are included in the plant Emergency Operating Procedures (EOPs).
15.4.2.4.3 Analysis of Effects and Consequences The FLB transient is analyzed for pressurizer filling in accordance with the NRC approved methodology for a 4-loop plant (Reference 70) using the Westinghouse version of the RETRAN-02 computer code (RETRAN-02W), which is also used for the analysis of the FLB transient described in Section 15.4.2.2.3.
Separate cases to accommodate different limiting assumptions were analyzed to determine the time by which the operators would need to ensure a PG&E Design Class I PORV is available and the times by which the operators would need to isolate charging flow and subsequently stop RCP seal injection flow. Cases were also analyzed with and without offsite power available to determine the more limiting condition.
The assumptions for the pressurizer filling analysis are conservatively chosen to minimize the time to reach a water-solid condition and maximize the number of pressurizer PORV relief open/close cycles predicted. Sensitivity studies were performed for a number of parameters to determine the appropriate conservative assumptions. Major assumptions are the same as described in Section 15.4.2.2.3 with the following changes:
(1) Initial reactor coolant average temperature is 5.5°F below the nominal value, and the initial pressurizer pressure is 60 psi below its nominal value.
(2) No credit is taken for relief through the PORV that is actuated on a compensated pressurizer pressure deviation signal (i.e., the non-safety-grade PORV).
However, relief through the PORVs that are actuated on the indicated (measured) pressurizer pressure signal (i.e., the safety-grade, PG&E Design Class I PORVs) has been modeled with assumptions that maximize the number of PORV opening cycles experienced. The number of safety-grade PORVs available for relief (i.e., either one or both of the PG&E Design Class I PORVs) depends on the single failure being considered.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-41 Revision 23 December 2016 Also, since an SI signal causes Phase A containment isolation and the instrument air is a PG&E Design Class II (non-safety-grade) system, there is a loss of instrument air to containment due to this signal. Accordingly, the PG&E Design Class I backup nitrogen accumulators are needed to maintain functionality of the PG&E Design Class I PORVs. The backup nitrogen accumulators are each sized and leak tested to ensure at least 300 PORV cycles before the backup nitrogen supply is depleted, after which the PORV would be unavailable. Therefore, transient mitigation must be demonstrated to occur before 300 PORV cycles is exceeded.
(3) No credit is taken for normal charging or letdown flow.
(4) The turbine-driven auxiliary feedwater pump (TDAFWP) is aligned to all four SGs, whereas the motor-driven auxiliary feedwater pumps (MDAFWPs) are each independently aligned to two of the four SGs.
The AFW flow for each case analyzed depends on the assumed single failure.
With the single failure of the TDAFWP considered, it is assumed that 390 gpm total AFW flow will be delivered to two of the intact SGs at 1 minute after the trip and an additional 195 gpm of AFW flow will be delivered to the third intact SG at 10 minutes after the trip. The AFW flow initiated at 1 minute after the trip is delivered from the MDAFWP that is alig ned to two intact SGs. All flow from the other MDAFWP aligned to both the third intact SG and the faulted SG is initially assumed to spill out the break. Subsequently, a TCOA is taken within 10 minutes to isolate the faulted SG and direct AFW flow from this MDAFWP to the third intact SG.
With the single failure of a PG&E Design Class I pressurizer PORV considered, the AFW flow from the MDAFWPs is the same as described above. However, with this scenario it is also assumed the TDAFWP will deliver an additional total 585 gpm of AFW flow to the three intact SGs at 10 minutes after the trip when the TCOA is taken to isolate the faulted SG.
(5) Maximum SI flow rates were conservatively modeled with a flow profile that bounds the maximum flow from the two PG&E Design Class I high-head centrifugal charging pumps (CCP1 and CCP2), plus the non-safety-related CVCS charging pump (CCP3), plus two intermediate-head SI pumps. Full SI flow was conservatively assumed to occur immediately after the SI actuation signal. The maximum SI flow profile, which includes RCP seal injection flow, is modeled until the TCOA is taken to isolate charging flow. Note that no flow is actually injected from the intermediate-head SI pumps, since RCS pressure remains above the shutoff head of these pumps during the transient.
(6) Maximum RCP seal injection flow was conservatively modeled until the TCOA is taken to stop it. A limiting FLB inside containment may cause the high-high DCPP UNITS 1 &
2 FSAR UPDATE 15.4-42 Revision 23 December 2016 containment pressure setpoint to be reached, resulting in Phase B isolation and a loss of component cooling water (CCW) to the RCPs. Accordingly, RCP seal injection flow must be maintained to ensure RCP cooling until operator action can be taken to reset the Phase B containment isolation and restore CCW flow to the RCPs. (7) The air-operated pressurizer spray valves are assumed to be inoperable, since instrument air to containment is lost on an SI signal and normal pressurizer spray flow is unavailable followin g coastdown of the RCPs. There are auxiliary spray flow lines that are equipped with backup nitrogen if the spray valves are unavailable; however, aux iliary spray requires a manual alignment that would not be completed until after the TCOAs necessary to mitigate this transient are complete.
(8) For the cases with a loss of offsite power, the pressurizer heaters are assumed to be inoperable, since they are not automatically loaded onto an emergency diesel generator (EDG) bus and will not be manually loaded onto the EDGs until after the TCOAs to mitigate this transient are complete. For cases with offsite power available, the pressurizer heaters are assumed to be operable.
(9) For cases with loss of offsite power, the RCPs are assumed to trip automatically following reactor trip. For cases with offsite power, the RCPs continue to operate unless manually tripped by the operators. The EOPs direct the operators to trip the RCPs within 5 minutes following the Phase B containment isolation (to protect the RCP motors, which are cooled by CCW). For the case to determine time by which the operators need to ensure a pressurizer PORV is available, it is assumed the operators manually trip the RCPs at greater than 90 seconds after FLB initiation, because Phase B containment isolation from high-high containment pressure would not occur before this time. However, for the case analyzed to determine the times by which the operators would need to isolate charging flow and subsequently stop RCP seal injection flow, it was conservatively assumed that the RCPs are manually tripped following reactor trip. 15.4.2.4.4 Results The results for Unit 2 were more limiting than those calculated for Unit 1.
With respect to the single failure scenarios, it was found the failure of the TDAFWP is limiting for the calculation of minimum time to pressurizer filling, unless one of the two PG&E Design Class I PORVs is blocked at the start of the transient. If a PORV is blocked, the failure of the other PG&E Design Class I PORV is limiting for pressurizer filling. The failure of a Class I PORV is also limiting for the calculation of the operator action times required to ensure that transient mitigation is complete before the maximum number of PORV cycles is reached.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-43 Revision 23 December 2016 For cases with offsite power available, pressurizer pressure is maintained after the RCP seal injection flow is stopped, since the pressurizer heaters (specifically, the backup heaters, actuated on high pressurizer level deviation) continue to operate. However, as a steam bubble forms again in the pressurizer and the pressurizer water volume begins decreasing, relief flow switches from water to steam. Because there is no longer a concern relative to water relief through the PSVs, transient mitigation is complete. For cases with a loss of offsite power, pressurizer pressure decreases after the RCP seal injection flow is stopped. Because the pressurizer PORV and PSV setpoints are no longer challenged, transient mitigation is complete for these cases.
The results of the FLB analysis for pressurizer filling demonstrate that, if the pressurizer fills, the following TCOAs preclude water relief through the PSVs:
(1) Ensure a pressurizer PORV is available within 8.6 minutes If no pressurizer PORV relief is available at the start of the transient because of a failure of one of the PG&E Design Class I PORVs and isolation of the other by its respective block valve, operator action is required to ensure a PG&E Design Class I PORV is available in time to prevent water relief through the PSVs. The analysis determined that the minimum time to pressurizer filling is 8.3 minutes and the minimum time to subsequently lift the PSVs is 8.6 minutes; therefore, the operators must ensure a PG&E Class I PORV is available within 8.6 minutes of event initiation.
The system response for the limiting case with no pressurizer PORV relief available at the start of the transient is presented in Figures 15.4.2-22 and 15.4.2-23. The calculated sequence of events is listed in Table 15.4-8.
(2) Isolate the faulted SG within 10 minutes Similar to the main feedwater pipe rupture analysis discussed in Section 15.4.2.2, the operators are assumed to isolate the faulted SG within 10 minutes after the low-low SG water level setpoint is reached in accordance with operating procedures. This directs all available AFW flow to the intact SGs.
(3) Isolate charging flow within 25 minutes and (4) Stop RCP seal injection flow within 45 minutes The results of the limiting case determined that in order for transient mitigation to occur before the maximum number of PORV cycles is reached, the operators must isolate charging flow within 25 minutes after the low-low SG water level setpoint is reached and subsequently stop RCP seal injection flow within 45 minutes after the low-low SG water level setpoint is reached. These actions ensure that a steam bubble is formed in the pressurizer and the pressurizer water volume begins to decrease, causing relief flow to switch from water to steam before the capacity of the backup nitrogen accumulators is DCPP UNITS 1 &
2 FSAR UPDATE 15.4-44 Revision 23 December 2016 depleted. Once this occurs, there is no longer a concern relative to water relief through the PSVs and transient mitigation is complete for these cases.
The system response is presented in Figures 15.4.2-24 through 15.4.2-27.
Table 15.4-8, "Sequence of Events," indicates that a steam bubble forms again in the pressurizer at 6723 seconds. This occurs at cycle 295 before the maximum number of 300 PORV cycles is reached at 7137.6 seconds.
Summary of TCOAs The TCOAs established for mitigation of the FLB event are summarized below. All TCOA times are from event initiation.
- 1. Ensure a PG&E Design Class I pressurizer PORV is available within 8.6 minutes 2. Isolate the faulted SG within 10 minutes 3. Isolate charging flow within 25 minutes 4. Stop RCP seal injection flow within 45 minutes 15.4.2.4.5 Conclusion The results of the FLB analysis for pressurizer filling show that operator actions, when taken in a timely manner, will preclude water relief through the PSVs. Thus, the reactor coolant pressure boundary integrity is maintained.
15.4.3 STEAM GENERATOR TUBE RUPTURE (SGTR) 15.4.3.1 Acceptance Criteria The following limiting criteria are applicable for a SGTR:
(1) The resulting potential exposures to individual members of the public and to the general population shall be lower than the applicable guidelines and limits specified in Section 15.5.20.
(2) There are no regulatory acceptance criteria associated with a SGTR margin-to-overfill transient analysis. However, it will be demonstrated that
there is sufficient margin to prevent overfill of the SG during an SGTR
event. Overfill of the SG may result in significantly increased offsite dose
consequences, along with damage to secondary components such as the
turbine and the main steam line.
15.4.3.2 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 power with the reactor coolant contaminated DCPP UNITS 1 &
2 FSAR UPDATE 15.4-45 Revision 23 December 2016 with fission products corresponding to continuous operation with a limited amount of defective fuel rods. The accident leads to an increase in contamination of the
secondary system due to leakage of radioactive coolant from the reactor coolant system (RCS). In the event of a coincident loss of offsite power, or failure of the condenser
steam dump system, discharge of activity to the atmosphere takes place via the steam generator power-operated relief valves (and safety valves if their setpoint is reached).
Although the steam generator tube material is thermally treated Inconel 690, a highly
ductile material, it is assumed that complete severance could occur. The more probable
mode of tube failure would be one or more minor leaks of undetermined origin. Activity in the steam and power conversion system is subject to continual surveillance and an
accumulation of minor leaks that exceeds the limits established in the Technical Specifications (Reference 30) is not permitted during the unit operation.
The operator is expected to determine that a steam generator tube rupture has
occurred, to identify and isolate the ruptured steam generator, and to complete the
required recovery actions to stabilize the plant and terminate the primary to secondary
break flow. These actions should be performed on a restricted time scale in order to
minimize contamination of the secondary system and ensure termination of radioactive
release to the atmosphere from the ruptured unit. Consideration of the indications
provided at the control board, together with the magnitude of the break flow, leads to the
conclusion that the recovery procedure can be carried out on a time scale that ensures
that break flow to the secondary system is terminated before water level in the affected
steam generator rises into the main steam pipe. Sufficient indications and controls are
provided to enable the operator to carry out these functions satisfactorily.
Assuming normal operation of the various pl ant control systems, the following sequence of events is initiated by a tube rupture:
(1) Pressurizer low pressure and low-level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the
secondary side there is a steam flow/feedwater flow mismatch before trip
as feedwater flow to the affected steam generator is reduced due to the
break flow that is now being supplied to that unit.
(2) The main steam line radiation monitors, the air ejector radiation monitor and/or the steam generator blowdown radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system, and
steam generator blowdown will be automatically terminated.
(3) Continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure or overtemperature T. An SI signal, initiated by low pressurizer pressure, follows soon after the reactor trip. The SI signal automatically terminates normal feedwater supply and
initiates AFW addition.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-46 Revision 23 December 2016 (4) The reactor trip automatically trips the turbine and, if offsite power is available, the 40 percent condenser dump valves open permitting steam dump to the condenser. In the event of a coincident loss of offsite power, the 40 percent condenser dump valves would automatically close to protect the condenser. The stea m generator pressure would rapidly increase resulting in steam discharge to the atmosphere through the steam generator power-operated relief valves (PORVs) and safety valves
if their setpoint is reached.
(5) Following reactor trip and SI actuation, the continued action of AFW supply and borated SI flow (supplied from the refueling water storage
tank) provides a heat sink that absorbs some of the decay heat. This
reduces the amount of steam bypass to the condenser, or in the case of
loss of offsite power, steam relief to the atmosphere.
(6) SI flow results in stabilization of the RCS pressure and pressurizer water level, and the RCS pressure trends toward the equilibrium value where the
SI flow rate equals the break flow rate.
In the event of an SGTR, the plant operators must diagnose the SGTR and perform the
required recovery actions to stabilize the plant and terminate the primary to secondary
leakage. The operator actions for SGTR recovery are provided in the Emergency
Operating Procedures (Reference 42). The major operator actions include identification
and isolation of the ruptured steam generator, cooldown and depressurization of the
RCS to restore inventory, and termination of SI to stop primary to secondary leakage.
These operator actions are described below:
(1) Identify the ruptured steam generator. High secondary side activity, as indicated by the main steam line radiation monitors, the air ejector radiation monitor, or steam generator blowdown
radiation monitor typically will provide the first indication of an SGTR
event. The ruptured steam generator can be identified by an unexpected
increase in steam generator level, or a high radiation indication on the
corresponding main steam line monitor, or from a radiation survey of the
main steam lines. For an SGTR that results in a reactor trip at high power, the steam generator water level may decrease off-scale on the narrow
range for all of the steam generators. The AFW flow will begin to refill the
steam generators, distributing approximately equal flow to each of the
steam generators. Since primary to secondary leakage adds additional
liquid inventory to the ruptured steam generator, the water level will return
to the narrow range earlier in that steam generator and will continue to increase more rapidly. This response, as indicated by the steam
generator water level instrumentation, provides confirmation of an SGTR
event and also identifies the ruptured steam generator.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-47 Revision 23 December 2016 (2) Isolate the ruptured steam generator from the intact steam generators and isolate feedwater to the ruptured steam generator. Once a tube rupture has been identified, recovery actions begin by isolating steam flow from and stopping feedwater flow to the ruptured
steam generator. In addition to minimizing radiological releases, this also
reduces the possibility of overfilling the ruptured steam generator with
water by (a) minimizing the accumulation of feedwater flow and (b)
enabling the operator to establish a pressure differential between the
ruptured and intact steam generators as a necessary step toward
terminating primary to secondary leakage.
(3) Cool down the RCS using the intact steam generators. After isolation of the ruptured steam generator, the RCS is cooled as
rapidly as possible to less than the saturation temperature corresponding
to the ruptured steam generator pressure by dumping steam from only the intact steam generators. This ensures adequate subcooling in the RCS
after depressurization to the ruptured steam generator pressure in
subsequent actions. If offsite power is available, the normal steam dump
system to the condenser can be used to perform this cooldown. However, if offsite power is lost, the RCS is cooled using the PORVs on the intact
(4) Depressurize the RCS to restore reactor coolant inventory. When the cooldown is completed, SI flow will increase RCS pressure until
break flow matches SI flow. Consequently, SI flow must be terminated to
stop primary to secondary leakage. However, adequate reactor coolant
inventory must first be assured. This includes both sufficient reactor
coolant subcooling and pressurizer inventory to maintain a reliable
pressurizer level indication after SI flow is stopped. Since leakage from
the primary side will continue after SI flow is stopped until the RCS and
ruptured steam generator pressures equalize, an "excess" amount of
inventory is needed to ensure pressurizer level remains on span. The "excess" amount required depends on RCS pressure and reduces to zero
when RCS pressure equals the pressure in the ruptured steam generator.
The RCS depressurization is performed using normal pressurizer spray if
the reactor coolant pumps (RCPs) are runnin
- g. However, if offsite power is lost or the RCPs are not running for some other reason, normal
pressurizer spray is not available. In this event, RCS depressurization can
be performed using a pressurizer PORV or auxiliary pressurizer spray.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-48 Revision 23 December 2016 (5) Terminate SI to stop primary to secondary leakage. The previous actions will have e stablished adequate RCS subcooling, a secondary side heat sink, and sufficient reactor coolant inventory to ensure that SI flow is no longer needed. When these actions have been
completed, SI flow must be stopped to terminate primary to secondary
leakage. Primary to secondary leakage will continue after SI flow is
stopped until the RCS and ruptured steam generator pressures equalize.
Charging flow, letdown, and pressurizer heaters will then be controlled to
prevent repressurization of the RCS and reinitiation of leakage into the
ruptured steam generator.
Following SI termination, the plant conditions will be stabilized, the primary to secondary break flow will be terminated and all immediate safety concerns will have been
addressed. At this time a series of operator actions are performed to prepare the plant for cooldown to cold shutdown conditions. Subsequently, actions are performed to
cooldown and depressurize the RCS to cold shutdown conditions and to depressurize the ruptured steam generator.
15.4.3.3 Analysis of Effects and Consequences
15.4.3.3.1 SGTR Margin to Overfill (MTO) Analysis An SGTR results in the leakage of contaminated reactor coolant into the secondary
system and subsequent release of a portion of the activity to the atmosphere.
Therefore, an analysis must be performed to assure that the radiological consequences resulting from an SGTR are within allowable guidelines. Another concern for SGTR consequences is the possibility of steam gen erator overfill because this could potentially result in a significant increase in the radiolog ical consequences. Overfill could result in water entering the main steam line. If water continues to leak into the main steam lines, the release of liquid through the steam generator safety valves could result in an
increase in radiological doses. Therefore, an analysis was performed to demonstrate
margin to steam generator overfill, assuming the limiting single failure relative to overfill.
The results of this analysis demonstrate that t here is margin to steam generator overfill
for DCPP.
The overfill analysis is presented in Reference 72 and the major assumptions include:
(1) Complete severance of a single tube located at the top of the tube sheet on the outlet side of the steam generator, resulting in double ended flow (2) Initiation of the event from full power (3) A loss of offsite power coincident with reactor trip (4) Failure of an AFW control valve to close (limiting single failure)
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-49 Revision 23 December 2016 (5) The PORVs on all three intact steam generators are fully opened during the RCS cooldown (6) Operator actions are consistent with the times shown in Table 15.4-12 The SGTR MTO analysis acceptance criterion is to maintain a positive margin to overfill
when the event is terminated. The limiting margin to overfill analysis presented in
Reference 72 demonstrates that the steam generator liquid volume is 30 cubic feet less
than the total steam generator volume of 5800 cubic feet when the SGTR event is
terminated. The SGTR MTO analysis sequence of events is listed in Table 15.4-13A
and the transient responses are presented in Figures 15.4.3-1A to 15.4.3-4A and
Figures 15.4.3-6A to 15.4.3-8A.
An analysis was also performed to determine the transient thermal hydraulic data for input into the radiological consequences analysis, assuming the limiting single failure relative to doses without steam generator overfill (as opposed to one that is relative to
overfill). Because steam generator overfill does not occur, the radiation consequences (refer to Section 15.5.20) calculated using the results of this analysis represent the limiting consequences for an SGTR for DCPP. The thermal hydraulic results used by
the radiological consequences (Dose) analysis are discussed below.
15.4.3.3.2 SGTR Dose Input Analysis
A thermal and hydraulic analysis was performed to determine the plant response for a design basis SGTR, and to determine the integrated primary to secondary break flow and the mass releases from the ruptured and intact steam generators to the condenser and to the atmosphere. This information was then used to calculate the quantity of
radioactivity released to the environment and the resulting radiological consequences.
The thermal and hydraulic analysis discussed in this section is presented in
Reference 41 and the results of the radiological consequences analysis are discussed
in Section 15.5.20.
The plant response following an SGTR was analyzed with the RETRAN-02W program until the primary to secondary break flow is terminated. The reactor protection system
and the automatic actuation of the engineered safeguards systems were modeled in the analysis. The major operator actions which are required to terminate the break flow for
an SGTR were also simulated in the analysis.
Analysis Assumptions The accident modeled is a double-ended break of one steam generator tube located at
the top of the tube sheet on the outlet (cold leg) side of the steam generator. However, as indicated subsequently, the break flow flashing fraction was conservatively
calculated assuming that all of the break flow comes from the hot leg side of the steam
generator. The combination of these conservative assumptions regarding the break DCPP UNITS 1 &
2 FSAR UPDATE 15.4-50 Revision 23 December 2016 flow location results in a very conservative calculation of the radiation doses. It was assumed that the reactor is operating at full power at the time of the accident and the secondary mass was assumed to correspond to operation at the steam generator
nominal level with an allowance for uncertaint ies. It was also assumed that a loss of
offsite power occurs at the time of reactor trip and the highest worth control assembly
was assumed to be stuck in its fully withdrawn position at reactor trip.
The limiting single failure was assumed to be the failure of the PORV on the ruptured
steam generator. Failure of this PORV in the open position will cause an uncontrolled depressurization of the ruptured steam generator which will increase primary to
secondary leakage and the mass release to the atmosphere. It was assumed that the
ruptured steam generator PORV fails open when the ruptured steam generator is
isolated, and that the PORV was isolated by locally closing the associated block valve.
The major operator actions required for the recovery from an SGTR are discussed in Section 15.4.3.2 and these operator actions were simulated in the analysis. The operator action times which were used for the analysis are presented in Table 15.4-12.
It is noted that the PORV on the ruptured steam generator was assumed to fail open at
the time the ruptured steam generator was isolated. It was assumed that the operators
isolate the failed open PORV by locally cl osing the associated block valve to complete the isolation of the ruptured steam generator before proceeding with the subsequent
recovery operations. It was assumed that the ruptured steam generator PORV was isolated at 30 minutes after the valve was assumed to fail open. After the ruptured steam generator PORV was isolated, an additional delay time of 5 minutes (Table 15.4-12) was assumed for the operator action time to initiate the RCS cooldown.
Transient Description The RETRAN-02W (Reference 70) analysis results are described below. The sequence
of events for this transient is presented in Table 15.4-13B.
Following the tube rupture, reactor coolant flows from the primary into the secondary side of the ruptured steam generator since the primary pressure is greater than the steam generator pressure. In response to t his loss of reactor coolant, pressurizer level decreases as shown in Figure 15.4.3-1B. The pressurizer pressure also decreases as
shown in Figure 15.4.3-2B as the steam bubble in the pressurizer expands. As the
RCS pressure decreases due to the continued primary to secondary leakage, automatic
After reactor trip, core power rapidly decreases to decay heat levels. The turbine stop
valves close and steam flow to the turbine is terminated. The steam dump system is
designed to actuate following reactor trip to limit the increase in secondary pressure, but
the 40 percent condenser dump valves remain closed due to the loss of condenser vacuum resulting from the assumed loss of offsite power at the time of reactor trip.
Thus, the energy transfer from the primary system causes the secondary side pressure
to increase rapidly after reactor trip until the steam generator PORVs (and safety valves DCPP UNITS 1 &
2 FSAR UPDATE 15.4-51 Revision 23 December 2016 if their setpoints are reached) lift to dissipate the energy, as shown in Figure 15.4.3-3B.
The main feedwater flow will be terminated and AFW flow will be automatically initiated
following reactor trip and the loss of offsite power.
The RCS pressure decreases more rapidly after reactor trip as energy transfer to the secondary shrinks the reactor coolant and the tube rupture break flow continues to
deplete primary inventory. Pressurizer level also decreases more rapidly following reactor trip. The decrease in RCS inventory results in a low pressurizer pressure SI
signal. After SI actuation, the SI flow rate maintains the reactor coolant inventory and the pressurizer level begins to stabilize. The RCS pressure also trends toward the
equilibrium value where the SI flow rate equals the break flow rate.
Because offsite power was assumed lost at reactor trip, the RCPs trip and a gradual
transition to natural circulation flow occurs. I mmediately following reactor trip the temperature differential across the core decreases as core power decays (refer to Figures 15.4.3-4B and 15.4.3-5B), however, the temperature differential subsequently
increases as natural circulation flow develops. The cold leg temperatures trend toward
the steam generator temperature as the fluid residence time in the tube region increases. The intact steam generator loop temperatures slowly decrease due to the
continued AFW flow until operator actions are taken to control the AFW flow to maintain
the specified level in the intact steam generators. The ruptured steam generator loop
temperatures also continue to slowly decrease until the ruptured steam generator is
isolated, at which time the PORV is assumed to fail open.
Major Operator Actions (1) Identify and Isolate the Ruptured Steam Generator As indicated in Table 15.4-12, it was assumed that the ruptured steam generator is identified and isolated at 10 minutes after the initiation of the
SGTR or when the narrow range level reaches 38 percent, whichever time
is longer. Since the time to reach 38 percent narrow range level was 953
seconds, it was assumed that the actions to isolate the ruptured steam
generator are performed at this time.
The ruptured steam generator PORV was also assumed to fail open at
this time, and the failure was simulated at 953 seconds. The failure
causes the ruptured steam generator to rapidly depressurize, which
results in an increase in primary to secondary leakage. The
depressurization of the ruptured steam generator increases the break flow
and energy transfer from primary to secondary which results in a decrease
in the ruptured loop temperatures as shown in Figure 15.4.3-5B. As noted
previously, the intact steam generator loop temperatures also decrease, as shown in Figure 15.4.3-4B, until the AFW flow to the intact steam
generators is throttled. These effects result in a decrease in the RCS
pressure and pressurizer level, until the failed open PORV is isolated.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-52 Revision 23 December 2016 It was assumed that the time requ ired for the operator to identify that the ruptured steam generator PORV is open and to locally close the associated block valve is 30 minutes. Thus, the isolation of the ruptured steam generator was completed at 2753 seconds, and the
depressurization of the ruptured steam generator was terminated. At this time, the ruptured steam generator pressure increases rapidly and the
primary to secondary break flow begins to decrease.
(2) Cool Down the RCS to establish Subcooling Margin After the ruptured steam generator PORV block valve was closed, a
5 minute operator action time was imposed prior to initiation of cooldown.
The depressurization of the ruptured steam generator affects the RCS cooldown target temperature because the temperature is dependent upon
the pressure in the ruptured steam generator. Since offsite power was
lost, the RCS was cooled by dumping steam to the atmosphere using the
intact steam generator PORVs. The cooldo wn was continued until RCS was subcooled 36°F including an allowance for instrument uncertainty.
Because the pressure in the ruptured steam generator continued to
decrease during the cooldown, the associated temperature the RCS was
less than the initial target temperature, which had the net effect of
extending the time for cooldown. The cooldown was initiated at 3053 seconds and was completed at 4424 seconds.
The reduction in the intact steam generator pressures required to accomplish the cooldown is shown in Figure 15.4.3-3B, and the effect of the cooldown on the RCS temperature is shown in Figure 15.4.3-4B. The
pressurizer level and pressurizer pressure also decrease during this
cooldown process due to shrinkage of the reactor coolant, as shown in
Figures 15.4.3-1B and 15.4.3-2B, respectively.
(3) Depressurize to Restore Inventory
After the RCS cooldown, a 4 minute operator action time was included
prior to depressurization. The RCS depressurization was initiated at
4664 seconds to assure adequate coolant inventory prior to terminating SI
flow. With the RCPs stopped, normal pressurizer spray is not available
and thus the RCS was depressurized by opening a pressurizer PORV.
The depressurization was continued until a ny of the following conditions
are satisfied: RCS pressure is less than the ruptured steam generator
pressure and pressurizer level is greater than the allowance of 12 percent
for pressurizer level uncertainty, or pressurizer level is greater than
74 percent, or RCS subcooling is less than the 20°F allowance for
subcooling uncertainty. The RCS depressurization reduces the break flow DCPP UNITS 1 &
2 FSAR UPDATE 15.4-53 Revision 23 December 2016 as shown in Figure 15.4.3-6B, and increases SI flow to refill the pressurizer as shown in Figure 15.4.3-1B.
(4) Terminate SI to Stop Primary to Secondary Leakage The previous actions have established ad equate RCS subcooling, verified a secondary side heat sink, and restored the reactor coolant inventory to
ensure that SI flow is no longer needed. When these actions have been
completed, the SI flow must be stopped to prevent repressurization of the
RCS and to terminate primary to secondary leakage. The SI flow is
terminated after a delay to allow for operator response if RCS subcooling
is greater than the 20°F allowance for uncertainty, minimum AFW flow is
available or at least one intact steam generator level is in the narrow
range, the RCS pressure is stable or increasing, and the pressurizer level
is greater than the 12 percent allowance for uncertainty.
After depressurization was completed, an operator action time of 2 minutes was
assumed prior to SI termination. Since the above requirements are satisfied, SI termination was performed at this time.
After SI termination, the pressurizer pressure decreases as shown in Figure 15.4.3-2B. Figure 15.4.3-6B shows that the primary to
secondary leakage continues after the SI flow was stopped until the RCS and ruptured
steam generator pressures equalize.
The ruptured steam generator water volume for the radiological conseque nces analysis is shown in Figure 15.4.3-7B. The mass of water in the ruptured steam generator is
also shown as a function of time in Figure 15.4.3-8B.
Mass Releases The mass releases were determined for use in evaluating the exclusion area boundary
and low population zone radiatio n exposure. The steam releases from the ruptured and intact steam generators, the feedwater flows to the ruptured and intact steam
generators, and primary to secondary break flow into the ruptured steam generator
were determined for the period from accident initiation until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the accident
and from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident.
The releases for 0-2 hours were used to calculate the radiation doses at the exclusion area boundary for a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exposure, and
the releases for 0-8 hours were used to calculate the radiation doses at the low
population zone for the duration of the accident.
The operator actions for the SGTR recovery up to the termination of primary to secondary leakage were simulated in the RETRAN-02W analysis. Thus, the steam
releases from the ruptured and intact steam generators, the feedwater flows to the
ruptured and intact steam generators, and the primary to secondary leakage into the
ruptured steam generator were determined from the RETRAN-02W results for the period from the initiation of the accident until the leakage was terminated.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-54 Revision 23 December 2016 Following the termination of leakage, it was assumed that the actions are taken to cool down the plant to cold shutdown conditions.
The PORVs for the intact steam generators were assumed to be used to cool down the RCS to the RHR system
operating temperature of 350°F, at the maximum allow able cooldown rate of 100°F/hr.
The steam releases and the feedwater flows for the intact steam generator for the
period from leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> were determined from a mass and energy
balance using the calculated RCS and intact steam generator conditions at the time of
leakage termination and at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The RCS cooldown was assumed to be continued
after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> until the RHR system in-service temperature of 350°F is reached.
Depressurization of the ruptured steam generator was then assumed to be performed to
the RHR in-service pressure of 405 psia via steam release from the ruptured steam generator PORV. The RCS pressure was also assumed to be reduced concurrently as
the ruptured steam generator is depressurized. It was assumed that the continuation of
the RCS cooldown and depressurization to RHR operating conditions are completed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident since there is ample time to complete the operations
during this time period. The steam releases and feedwater flows from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> were
determined for the intact steam generators from a mass and energy balance using
conditions at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at the RHR syste m in-service conditions. The steam released from the ruptured steam generator from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> was determined based on a mass and energy balance for the ruptured steam generator using the conditions at the time of
leakage termination and saturated conditions at the RHR in-service pressure.
After 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, it was assumed that further plant cooldown to cold shut down as well as
long-term cooling is provided by the RHR system. Therefore, the steam releases to the
atmosphere are terminated after RHR in-service conditions are assumed to be reached
at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
During the time period from initiation of the accident until leakage termination, the
releases were determined from the RETRAN-02W results for the ti me prior to reactor trip and following reactor trip. Since the condenser is in service until reactor trip, any
radioactivity released to the atmosphere prior to reactor trip would be through the
condenser air ejector and/or the condenser vacuum pump exhaust (if in operation).
After reactor trip, the releases to the atmosphere were assumed to be via the steam
generator PORVs. The mass release rates to the atmosphere from the RETRAN-02W
analysis are presented in Figures 15.4.3-9 and 15.4.3-10 for the ruptured and intact
steam generators, respectively, for the time period until leakage termination. The total flashed break flow from the RETRAN-02W analysis is presented in Figure 15.4.3-11.
The mass releases calculated from the time of leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and
from 2-8 hours were also assumed to be released to the atmosphere via the steam
generator PORVs. The mass releases for the SGTR event for the 0-2 hour and
2-8 hour time intervals are presented in Table 15.4-14.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-55 Revision 23 December 2016 15.4.3.4 Conclusions The analysis demonstrates the acceptance criteria are met as follows:
15.4.3.4.1 Overfill Analysis The SGTR MTO analysis acceptance criteria are to maintain a positive margin to overfill
when the event is terminated. Therefore, the limiting margin to overfill analysis
demonstrates that the steam generator liquid volume is less than the total steam
generator volume of 5800 cubic feet when the SGTR event is terminated.
15.4.3.4.2 Radiological Section 15.5.20 demonstrates that the acceptance criteria for Dose Consequences of a
SGTR are met. Table 15.5-71 provides offsite radiation doses from SGTR accident.
Table 15.5-74 provides control room radiation doses from airborne activity in SGTR
accident.
15.4.4 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR 15.4.4.1 Identification of Causes and Accident Description The accident postulated is an instantaneous seizure of an RCP rotor.
Following initiation of the reactor trip, heat stored in the fuel rods continues to be
transferred to the coolant causing the coolant to expa nd. At the same time, heat transfer to the shell-side of the s team generators 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 reduced heat transfer in the steam generators causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge
into the pressurizer compresses the steam volume, actuates the automatic spray
system, opens the power-operated relief valves, and opens the pressurizer safety
valves in that sequence. The three power-operated relief valves are designed for
reliable operation and would be e xpected 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 the analysis.
15.4.4.2 Analysis of Effects and Consequences Three digital computer codes are used to analyze this transient. The LOFTRAN (Reference 26) code is used to calculate the resulting loop and core coolant flow
following the pump seizure. The LOFTRAN code is also used to calculate the time of reactor trip based on the calculated flow, the nuclear power following reactor trip, and to
determine the peak pressure. The thermal behavior of the fuel located at the core hot DCPP UNITS 1 &
2 FSAR UPDATE 15.4-56 Revision 23 December 2016 spot is investigated using the FACTRAN (Ref erence 17) code, using the core flow and the nuclear power calculated by LOFTRAN. The FACTRAN code includes the use of a film boiling heat transfer coefficient. The THINC (Reference 31) code (refer to Section 4.4.3) is used to calculate the DNBR during the transient based on flow calculated by LOFTRAN and heat flux calculat ed by FACTRAN.
At the beginning of the postulated locked rotor accident, i.e., at the time the shaft in one of the RCPs is assumed to seize, for the DNB evaluation the plant is assumed to be under steady state operating conditions consistent with use of the Improved Thermal Design Procedure (Reference 62).
When the peak pressure is evaluated, the initial power is assumed as 2 percent above nominal full power, the initial coolant average temperature is assumed 5°F above nominal, and the initial pressure is conservatively assumed as 60 psi above nominal pressure (2250 psia) to allow for uncertainties.
This is done to obtain the highest possible rise in the coolant pressure during the transient. The pressure response for the point in the RCS having the maximum pressure is shown in Figure 15.4.4-1.
The analysis accounts for the potential effect of asymmetric steam generator tube plugging, which results in a loop-to-loop flow asymmetry. The loop with the locked rotor is assumed to have the highest initial flow rate, which conservatively minimizes the core flow during the transient.
15.4.4.2.1 Evaluation of the Pressure Transient After pump seizure and reactor trip, the neutron flux is rapidly reduced by the effect of control rod insertion. Rod motion is assumed to begin 1 second after the flow in the
affected loop reaches 85 percent of nominal flow. No credit is 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.
The pressurizer safety valve model includes a +3 percent opening tolerance plus 5 psi accumulation above the nominal setpoint of 2500 psia. A purge delay of 1.272 seconds was also included to account for the presence of water-filled loop seals (see Reference 75). The analysis conservatively assumes an additional +1% shift in the opening setpoint. Note that all pressurizer safety valves have been concerted to a steam seat design and condensate in the loop is now continuously drained back to the pressurizer, thereby eliminating the water loop seal. Even though the water loop seal has been eliminated, the resulting benefit is not credited in the analysis.
15.4.4.2.2 Evaluation of the Effects of DNB in the Core During the Accident For this accident, DNB is assumed to occur in the core and, therefore, an evaluation of
the consequences with respect to fuel rod thermal transients is performed. Results DCPP UNITS 1 &
2 FSAR UPDATE 15.4-57 Revision 23 December 2016 obtained from analysis of this hot spot condition represent the upper limit with respect to cladding temperature and zirconium-water reaction.
In the evaluation, the rod power at the hot spot is conservatively assumed to be greater
than or equal to 2.7 (i.e., F Q
15.4.4.2.3 Film Boiling Coefficient The film boiling coefficient is calculated in the FACTRAN code using the
Bishop-Sandberg-Tong film boiling correlation. 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 on the actual heat transfer conditions at the
time. The neutron flux, and mass flowrate as a function of time are used as program input.
For this analysis, the initial values of the pressure and the bulk density are used
throughout the transient since they are the most conservative with respect to cladding
temperature response. For conservatism, DNB was assumed to start at the beginning of the accident.
15.4.4.2.4 Fuel Cladding Gap Coefficient The magnitude and time dependence of the heat transfer coefficient between fuel and
cladding (gap coefficient) has a pronounced i nfluence on the thermal results. The larger the value of the gap coefficient, the more heat is transferred between pellet and
cladding. Based on investigations on the effect of the gap coefficient upon the maximum cladding temperature during the transient, the gap coefficient was assumed
to increase from a steady state value consis tent with the initial fuel temperature to 10,000 Btu/hr-ft 2-°F within 0.5 seconds after the initiation of the transient. This assumption causes energy stored in the fuel to be released to the cladding at the
initiation of the transient and maximizes the cladding temperature during the transient.
15.4.4.2.5 Zirconium-steam Reaction The zirconium-steam reaction can become significant above 1800°F (cladding temperature). The Baker-Just parabolic rate equation shown below is used to define
the rate of the zirconium-steam reaction.
x=1.986T 45,500 exp 61033.3 dt)2(w d (15.4-1) where:
w = amount reacted, mg/cm 2 t = time, sec T = temperature, °K DCPP UNITS 1 &
2 FSAR UPDATE 15.4-58 Revision 23 December 2016 and the reaction heat is 1510 cal/gm.
15.4.4.3 Results Transient plots of maximum RCS pressure, flow coastdown, hot channel heat flux, and neutron flux are shown in Figure 15.4.4-1 and Figures 15.4.4-3 through 15.4.4-5.
Maximum RCS pressure, maximum cladding temperature, and amount of
zirconium-water reaction are contained in Table 15.4-10. Figure 15.4.4-2 shows the
cladding temperature transient for the worst case.
15.4.4.4 Conclusions (1) Because the peak RCS pressure reached during any of the transients is less than that which would cause stresses to exceed the faulted condition stress limits, the integrity of the p rimary coolant system is maintained. For the main steam system, the maximum pressure is bounded by the analysis of the loss of external el ectrical load/turbine trip event (Section 15.2.7). (2) Because the peak cladding average temperature calculated for the hot spot during the worst transient remains considerably less than 2700°F and
the amount of zirconium-water reaction is small, the core will remain in
place and intact with no consequential loss of core cooling capability.
(3) The results of the transient analysis show that less than 10 percent of the fuel rods will have DNBRs below the safety analysis limit values.
15.4.5 FUEL HANDLING ACCIDENT 15.4.5.1 Acceptance Criteria The following limiting criterion is applicabl e for a fuel handling accident:
(1) The resulting potential exposures to individual members of the public and to the general population shall be lower than the applicable guidelines and
limits specified in 10 CFR Part 100.
15.4.5.2 Identification of Causes and Accident Description 15.4.5.2.1 Fuel Handling Procedures One major task that must be performed routinely as part of the operation of a nuclear power plant is the handling of the reactor fuel. The bulk of this fuel handling occurs during refueling outages, which occur every one to two years, and all of these operations are carried out with the fuel under water. A typical refueling outage would
include the following major operations:
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-59 Revision 23 December 2016 (1) Shutdown of the reactor and cooldown to ambient conditions (2) Removal and storage of pressure vessel head (3) Filling of refueling cavity above the pressure vessel with water to provide shielding from radioactive fuel (4) Transfer of the reactor fuel assemblies from the reactor itself to underwater storage racks in the spent fuel pool (5) Performance of outage tasks appropriate to the core off-load window (6) Return of the appropriate number of partially burned and new fuel assemblies to the reactor Fuel handling operations within the containment build ing and the fuel handling area are accomplished with overhead cranes, speciall y designed fuel grapples, and miscellaneous other equipment. To facilitate the transfer of the fuel between the two buildings, an underwater penetration called the transfer tube is provided through the
walls where the buildings adjoin. A conveyer cart is used to transport the fuel from one
building to the other through this penetration. A more detailed description of the
equipment used in fuel handling operations can be found in Chapter 9.
Spent fuel remains in storage in the spent fuel pool until placed in a cask for transport to
the Diablo Canyon Independent Spent Fuel Storage Installation (ISFSI) or for shipment from the site.
15.4.5.2.2 Probability of Activity Release In the above operations, there exists the remote possib ility that one or more fuel
assemblies will sustain some mechanical damage. There exists an even more remote
possibility that this damage will be severe enough to breach the cladding and release
some of the radioactive fission products contained therein.
Both the fuel handling procedure and the fuel handl ing equipment design adhere to the following safety criteria:
(1) Fuel handling operations must not commence before short-lived core activity has decayed, leaving only relatively long-lived activity. Equipment
Control Guidelines for refueling operations specify the minimum waiting
time. (2) Fuel handling operations must preclude any critical configuration of the core, spent fuel, or new fuel.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-60 Revision 23 December 2016 (3) The fuel handling system design must ensure an adequate water depth for radiation shielding of operating personnel.
(4) Active components of the fuel handling systems must be designed such that loss-of-function failures will terminate in stable modes.
(5) The design of fuel handling equipment must minimize the possibility of accidental impact of a moving fuel assembly with any structure.
(6) The design of fuel handling equipment and procedures must minimize the possibility of any massive object damaging a stationary fuel assembly.
(7) Fuel assembly design must minimize the possibility of damage in the event that portable or hand tools come into contact with a fuel assembly.
(8) The design of structures around the fuel handling system must minimize the possibility of the structures themselves failing in the event of a design earthquake (DE), double design earthquake (DDE), or Hosgri earthquake (HE). Furthermore, the structures must minimize the possibility of any external missile from reaching fuel assemblies.
(9) Fuel handling equipment must be capable of supporting maximum loads under seismic conditions. Furthermore, fuel handling equipment must not generate missiles during seismic conditions. The earthquake loading of the fuel handling equipment is evaluated in accordance with the seismic
considerations addressed in Sections 9.1.4.3.1 and 9.1.4.3.9.
Implementation of the above safety criteria into the fuel handling system design is
discussed in greater detail in Chapter 9.
Because of the above design, the probability of breaching the fuel cladding and releasing radioactive fission products is very small.
15.4.5.2.3 Accident Description In order to assess the probable extent of fuel cladd ing damage from a fuel handling accident, it is necessary to look more closely at specific fuel handling accidents that
might realistically occur.
Multiple assemblies are loaded into the multi-purpose canister (MPC)/transfer cask
assembly for movement to the ISFSI, as des cribed in Section 9.1.4.2.6. The MPC is subsequently drained, evacuated, backfilled with helium, and sealed. However, extensive design and analysis along with application of the ISFSI Technical
Specifications ensure temperatures remain within the design basis and no fuel cladding
damage occurs.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-61 Revision 23 December 2016 The possibility of damaging fuel claddin g by overheating during fuel handling operations was considered. Because irradiated fuel is always handled under water, overheating would require draining either the refueling cavity or the spent fuel pool while irradiated fuel was located within them. Consideration has been given in design of the cavity and
pool to prevent either of these possibilities. The probability of losing coolant while an assembly is in the transfer tube is also extremely small in view of the fact that the tube
is open on one end to the reactor cavity and on the other end to the pool. There is no
realistic occurrence that would simultaneously block off both ends of the tube.
Therefore, it is expected that there will be no radiological consequences over the lifetime of the plant that results from overheating during fuel handling operations.
The possibility of dropping a foreign object of sufficient size to produce cladding rupture
onto irradiated fuel located either in the reactor or the pool is extremely remote because
the design of the plant is such that only rarely are objects of this size transported over
locations containing irradiated fuel. The three large objects that are routinely handled in
the vicinity of irradiated fuel are the reactor head, upper internals package, which must
be removed and reinstalled from the pressure vessel at each refueling outage, and the
spent fuel shipment cask, which must be placed in the pool for loading. As discussed in
Section 9.1.4.2.5, load drop analyses were performed for the reactor head and upper
internals and are summarized in the PG&E NUREG-0612 submittal. It is not necessary
to lift the cask over the fuel racks in moving it to or from the pool. Protection of nuclear fuel assemblies from overhead load handling is a key element of the Control of Heavy Loads Program described in Section 9.1.4.3.10.
The possibility has also been considered of one of the bridge cranes falling into the reactor or the pool as a result of an earthquake. However, both of these cranes are seismically qualified for the DE, DDE, and HE. Therefore, it is expected that there will be no radiological consequences over the lifetime of the plant that result from dropping
objects onto radiated fuel.
If a fuel assembly were to strike an object, it is possible that the object might damage
the fuel rods with which it comes into contact.
If a fuel assembly were to strike against a flat, plane-like object or a linear, edge-like object, impact loads would be distributed across several fuel rods, and no cladding damage woul d be expected. If a fuel assembly were to strike against a sharp, corner-like object, impact loads would be
concentrated, and cladding damage might occur. Thus, there is a very remote
possibility that impact loads would be severe enough to rupture fuel cladding.
Analyses have been made by Westinghouse of the effects that would result from
dropping a fuel assembly from an initial vertical orientation onto a flat surface, the core, or a loaded fuel rack. Westinghouse has also analyzed the case where an assembly in the holder on the conveyor car falls from the vertical to the horizontal position. The
results of these analyses indicate there is only a very remote possibility of fuel cladding
rupture.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-62 Revision 23 December 2016 The above discussion indicates that the unlikely event of a fuel cladding integrity failure would most likely result from a fuel assembly striking a sharp object or dropping a fuel
assembly.
15.4.5.3 Results 15.4.5.3.1 Containment Building Accident During fuel handling operations, the contain ment ventilation penetrations to the outside atmosphere are maintained in a closed or automatically isolable condition. Isolation is
automatically actuated if either of the Containment Purge Exhaust (CPE) monitors, RM-44A or RM-44B, alarms due to a concentration of radioactivity in the containment
purge exhaust duct that exceeds the alarm setpoint. However, these penetrations are
also allowed to be open under administrative controls, which provide the capability of
closure within approximately 30 minutes.
Other containment penetrations, such as the personnel airlock and equipment hatch are
allowed to be open during fuel handli ng operations. These penetrations are capable of manual closure and will be closed in accordance with plant procedures should a fuel
handling accident occur.
In addition to the functions of the above mentioned monitors, fixed area radiation monitors are located in the containment. Should a fuel assembly be dropped and
release activity above a prescribed level, the area monitors would sound an audible
alarm. Personnel would exit the containment and containment closure would be
initiated immediately per administrative procedures.
Because of containment isolation and closur e capabilities, activity released from damaged fuel rods will be managed such that both the onsite and offsite exposures are
minimized. The containment iodine removal system (refer to Section 9.4.5) can be used to remove any radioactive iodine from the containment atmosphere, but is not credited for iodine removal in the radiological analysi s (refer to Section 15.5.22), and controlled containment venting can be initiated with offshore winds.
Thus, there is a reasonable probability that only limited onshore exposures will result from a containment fuel handling accident.
15.4.5.3.2 Fuel Handling Area Accident
A fuel assembly could be damaged in the transfer canal or the spent fuel pit in the fuel
handling area. Supply air for the spent fuel pit area is swept across the fuel pit and
transfer canal and exhausted through the vent. An area radiation monitor is located on
the bridge over the spent fuel pit. Doors in the fuel handling area are closed to maintain
controlled leakage characteristics in the spent fuel pit region during refueling operations
involving irradiated fuel. Should a fuel assembly be damaged in the canal or in the pit
and release radioactivity above a prescribed level, the radiation monitors sound an
alarm and the spent fuel pit ventilation exhaust through charcoal filters will remove most DCPP UNITS 1 &
2 FSAR UPDATE 15.4-63 Revision 23 December 2016 of the halogens prior to discharging it to the atmosphere. If the discharge is greater than the prescribed levels, an alarm sounds and the supply and exhaust ventilation systems servicing the spent fuel pit area can be manually shut down from the control
room, limiting the leakage to the atmosphere.
The analysis of the radiological effects of this accident is contained in Section 15.5.22.1.
15.4.5.4 Conclusions The analysis demonstrates the acceptance criteria are met as follows:
(1) Section 15.5.22 concludes that all potential exposures from a fuel handling accident will be well below the guidelin e levels specified in 10 CFR Part 100, and that the occurrence of such accidents would not result in undue risk to the public. Table 15.5-47 provides a summary of
doses from a fuel handling accident in the fuel handling area.
Table 15.5-50 provides a summary of o ffsite doses from a fuel handling
accident inside containment.
15.4.6 RUPTURE OF A CONTROL ROD DRIVE MECHANISM HOUSING (ROD CLUSTER CONTROL ASSEMBLY EJECTION) 15.4.6.1 Acceptance Criteria Conservative 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:
15.4.6.1.1 Fuel Damage Criteria (1) Average fuel pellet enthalpy at the hot spot below 225 cal/gm for unirradiated fuel and 200 cal/gm for irradiated fuel
(2) Average cladding temperature at the hot spot below the temperature at which cladding embrittlement may be expected (2700
°F)
(3) Fuel melting will be limited to less than 10 percent of the fuel volume at the hot spot even if the average fuel pellet enthalpy is below the limits of
Criterion (1) above
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-64 Revision 23 December 2016 15.4.6.1.2 Maximum RCS Pressure Criteria
(1) Peak reactor coolant pressure less than that which would cause stresses to exceed the faulted condition stress limits
15.4.6.1.3 Radiological Criteria (1) The resulting potential exposures to individual members of the public and to the general population shall be lower than the applicable guidelines and limits specified in 10 CFR Part 100.
15.4.6.2 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 an RCCA and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion and system depressurization
together with an adverse core power distribution, possib ly leading to localized fuel rod damage.
15.4.6.2.1 Design Precautions and Protection Certain features of the DCPP are intended to preclude the possibility of a 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 that lessens the potential ejection worth of RCCAs and minimizes the number of assemblies inserted at high power levels.
15.4.6.2.2 Mechanical Design The mechanical design is discussed in Section 4.2. Mechanical design and quality control procedures intended to preclude the possibility of an RCCA drive mechanism
housing failure are listed below:
(1) Each full length control rod drive mechanism housing is completely assembled and shop tested at 3107 psig.
(2) Pressure housings were individually hydrotested. The lower latch housing to nozzle connection is hydrotested during hydrotest of the completed
replacement RVCH.
(3) 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 DE, DDE, or HE can be accepted within the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-65 Revision 23 December 2016 allowable primary working stress range specified by the ASME Code,Section III, for Class I components.
(4) The latch mechanism housing and rod travel housing are each a single length of forged Type-304 stainless steel. T his material exhibits excellent
notch toughness at all temperatures that will be encountered.
(5) The CRDM housing plug is an integral part of the rod travel housing.
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 j oints between the latch mechanism housing and rod travel housing are threaded joints reinforced by canopy-type rod welds. Administrative
regulations require periodic inspections of these (and other) welds.
15.4.6.2.3 Nuclear Design Even if a rupture of an 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 rod
banks are selected during the nuclear design to lessen the severity of an RCCA ejection
accident. Therefore, should an 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. There are low and low-low level insertion
monitors with visual and audio signals. Operating instructions require boration at
low-level alarm and emergency boration at the low-low alarm.
15.4.6.2.4 Reactor Protection The reactor protection in the event of a rod ejection accident has been described in
Reference 18. The protection for this accident is provided by the power range high
neutron flux trip (high and low setting) and high rate of neutron flux increase trip. These
protection functions are described in detail in Section 7.2.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-66 Revision 23 December 2016 15.4.6.2.5 Effects on Adjacent Housings Disregarding the remote possibility of the occurrence of an RCCA mechanism housing failure, investigations have shown that failure of a housing due to either longitudinal or circumferential cracking is not expected to c ause damage to adjacent housings leading
to increased severity of the initial accident.
15.4.6.2.6 Limiting Criteria Due to the extremely low probability of an RCCA ejection accident, limited fuel damage is considered an acceptable consequence.
Comprehensive studies of the threshold of fuel failure and of the threshold of significant
conversion of the fuel thermal energy to mechanica l energy have been carried out as part of the SPERT project by the Idaho Nuc lear Corporation (Reference 19). Extensive tests of zirconium-clad UO 2 fuel rods representative of those in PWR-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 20) results, which indicated a failure
threshold of 280 cal/gm. Limited results have indicated that this threshold decreases by
about 10 percent with fuel burnup. The cladding failure mechanism appears to be
melting for zero burnup rods and brittle fracture for ir radiated 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) even for irradiated rods, did not occur below 300 cal/gm.
15.4.6.3 Analysis of Effects and Consequences The analysis of the RCCA ejection accident is performed in two stages: (a) an average
core nuclear power transient calculation and (b) a hot spot heat transfer 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 21.
15.4.6.3.1 Average Core Analysis The spatial kinetics computer code, TWINKLE (refer to Section 1.6.1 item 50 and Section 15.1.9.5) is used for the average core transient analysis. This code solves the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-67 Revision 23 December 2016 two group neutron diffusion theory kinetic equations 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 multi-region, transient fuel-clad-coolant heat transfer model for calculating pointwise Doppler, and moderator feedback effects.
In this analysis, the code is used as a one-di mensional axial kinetics code since it allows a more realistic representation of the spatial effects of axial moderator feedback
and RCCA movement and the elimination of axial feedback weighting factors.
However, since the radial dimension is missing, it is still necessary to employ very
conservative methods (described below) of calculating the ejected rod worth and hot
channel factor. A further description of TWINKLE appears in Section 15.1.9.5.
15.4.6.3.2 Hot Spot Analysis The average core energy addition, calculated as described above, is multiplied by the appropriate hot channel factors, and the hot spot analysis is performed using the
detailed fuel and cladding transient heat tra nsfer computer code, FACTRAN. This computer code calculates the transient temperature distribution in a cross section of a
metalclad UO 2 fuel rod, and the heat flux at the surface of the rod, using as input the nuclear power versus time and the local cool ant conditions. The zirconium and water (Zr-H 2 O) reaction is explicitly represented, and all material properties are represented as functions of temperature. A parabolic radial power generation is used within the fuel
rod.
FACTRAN uses the Dittus-Boelter (Reference 28) or Jens-Lottes (Reference 29) correlation to determine the film heat transfer before DNB, and the Bishop-Sandberg-Tong correlation (Reference 23) to determine the film boiling coefficient after DNB. The
DNB heat flux 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 that predicted by design fuel heat transfer codes.
For full power cases, the design initial hot channel factor (F Q) is input to the code. The hot channel factor during the transient is assumed to increase from the steady state
design value to the maximum transient value in 0.1 seconds, and remain at the
maximum for the duration of the transient. This is conservative, since detailed spatial
kinetics models show that the hot channel factor decreases shortly after the nuclear
power peak due to power flattening caused by preferential feedback in the hot channel.
Further description of FACTRAN appears in Section 15.1.8.
15.4.6.3.3 System Overpressu re 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 DCPP UNITS 1 &
2 FSAR UPDATE 15.4-68 Revision 23 December 2016 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 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 system, heat transfer to the
steam generators, and the action of the pressurizer spray and pressure relief valves.
No credit is taken for the possible pressure reduction caused by the assumed failure of
the control rod pressure housing (Reference 21).
15.4.6.3.4 Calculation of Basic Parameters Input parameters for the analysis are conser vatively selected on the basis of calculated values for this type of core. The more important parameters are discussed below.
Table 15.4-11 presents the parameters used in this analysis. A summary of the values
used in the reload analysis process is also provided in Table 15.4-11.
15.4.6.3.5 Ejected Rod Worths and Hot Channel Factors The values for ejected rod worths and hot channel factors are calculated using three-
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 ins ertion at a given power level as determined by the rod insertion limits.
Adverse xenon distributions are considered in the calculations.
The total transient hot channel factor F Q is then obtained by combining the axial and radial factors.
15.4.6.3.6 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 regions is
dependent on flux, these regions have high weights. This means that the reactivity
feedback is larger than that indicated by a simple single channel analysis. Physics
calculations were 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 that, 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 used. 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 DCPP UNITS 1 &
2 FSAR UPDATE 15.4-69 Revision 23 December 2016 for the missing spatial dimension. These weighting factors were shown to be conservative compared to three-dimensional analysis.
15.4.6.3.7 Moderator and Doppler Coefficient The critical boron concentrations at the beginning of life (BOL) and end-of-life (EOL) 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.
The Doppler reactivity defect is determined as a function of power level using the
one-dimensional steady state computer code with a Doppler weighting factor of 1. The
resulting curve is conservative compared to design predictions for this plant. The
Doppler weighting factor should be larger than 1 (approximately 1.3), just to make the present calculation agree with design predictions before ejection. This weighting factor will increase under accident conditions, as discussed above.
15.4.6.3.8 Delayed Neutron Fraction Calculations of the effective delayed neutron fraction (eff) typically yield values of 0.70 percent at BOL and 0.50 percent at EOL for the first cycle. The accident is sensitive to eff if the ejected rod worth is nearly equal to or greater than eff as in zero power transients. In order to allow for future fuel cycles, pessimistic estimates of 0.55 percent at beginning of cycle and 0.44 percent at end of cycle were used in the
analysis.
15.4.6.3.9 Trip Reactivity Insertion
The trip reactivity insertion assumed is given in Table 15.4-11 and includes the effect of one stuck rod. 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. The analyses presented are applicable for a rod insertion time of 2.7 seconds from coil release to entrance to the dashpot, although measurements
indicate that this value should be closer to 1.8 seconds.
The choice of such a conservative insertion rate means that there is over 1 second after
the trip point is reached before significant shutdown reactivity is inserted into the core.
This is particularly important conservatism for hot full power accidents.
The rod insertion versus time is described in Section 15.1.4.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-70 Revision 23 December 2016 15.4.6.4 Results Typical reload values of the parameters used in the VANTAGE 5 analysis, as well as
the results of the analysis, are presented in Table 15.4-11 and discussed below. Actual
values vary slightly from reload to reload.
15.4.6.4.1 Beginning of Cycle, Full Power Control Bank D was assumed to be inserted to its insertion limit. The worst ejected rod
6.70, respectively. The peak hot spot cladding average temperature was 2434°F. The
peak hot spot fuel center temperature exceeded the BOL melting temperature of
4900°F; however, melting was restricted to less than 10 percent of the pellet.
15.4.6.4.2 Beginning of Cycle, Zero Power For this condition, control Bank D was assumed to be fully inserted and C was at its
insertion limit. The worst ejected rod is located in control Bank D and was
of 13. The peak hot spot cladding average temperature reached only 2660°F.
15.4.6.4.3 End of Cycle, Full Power Control Bank D was assumed to be inserted to its insertion limit. The ejected rod worth
respectively. This resulted in an average PCT of 2218°F. The peak hot spot fuel center temperature exceeded the EOL melting temperature of 4800°F. However, melting was restricted to less than 10 percent of the pellet.
15.4.6.4.4 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 Bank C at its insertion limit. The results were
temperatures were 2632°F and 3849°F, respectively.
A summary of the cases presented above is given in Table 15.4-11. The nuclear power
and hot spot fuel cladding temperature transients for these representative BOL full
power and EOL zero power cases are presented in Figures 15.4.6-1 through 15.4.6-4.
15.4.6.4.5 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 percent of the rods entered DNB based on a
detailed three-dimensional THINC analysis. Although limited fuel melting at the hot spot was predicted for the full power cases, in practice melting is not expected since the DCPP UNITS 1 &
2 FSAR UPDATE 15.4-71 Revision 23 December 2016 analysis conservatively assumed that the hot spots before and after ejection were coincident.
15.4.6.4.6 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, diff erential expansion between separate rods cannot produce distortion. However, the temperature gradients across individual rods
may produce a force tending to bow the midpoint of the rods toward the hot spot.
Physics calculations indicate that the net result of this would be a negative reactivity insertion. 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 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 analyses.
15.4.6.5 Conclusions Even on a pessimistic basis, the analyses in dicate 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
show that less than 10 percent of the fuel rods enter DNB. Even in the portion of the
core which does reach DNB, there will be no excessive release of fission product activity if the limiting hot channel factors are not exceeded (Reference 21).
The analysis shows the acceptance criteria for a RCCA Ejection Accident has been met
as follows:
15.4.6.5.1 Fuel Damage (1) Table 15.4-11 shows the average fuel pellet enthalpy at the hot spot (Maximum fuel stored energy) below 225 cal/gm for non-irradiated fuel and 200 cal/gm (360 Btu/lb) for irradiated fuel.
(2) Table 15.4-11 shows the average clad temperature at the hot spot (Maximum cladding average temperature) below 2700°F, the temperature
above which clad embrittlement may be expected.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-72 Revision 23 December 2016 (3) Table 15.4-11 shows the fuel melting limited to less than the innermost 10 percent of the fuel pellet at the hot spot.
15.4.6.5.2 Maximum RCS Pressure (1) A detailed calculation of the pressure surge for an ejection worth of one dollar reactivity insertion at BOL, hot full power, indicates that the peak pressure does not exceed that which would cause stress to exceed the
faulted condition stress limits. Because the severity of the present analysis
does not exceed this worst case analysis, the accident for this plant will not
result in an excessive pressure rise or further damage to the RCS.
15.4.6.5.3 Radiological (1) Section 15.5.23 concludes that offsite exposures from a RCCA ejection accident will be well below the guideline levels specified in 10 CFR Part 100, and that the occurrence of such accidents would not result in undue risk to the public. Table 15.5-52 provides a summary of
offsite doses from a rod ejection accident.
15.4.7 RUPTURE OF A WASTE GAS DECAY TANK Refer to Section 15.5.24 for the description of this event.
15.4.8 RUPTURE OF A LIQUID HOLDUP TANK Refer to Section 15.5.25 for the description of this event.
15.4.9 RUPTURE OF VOLUME CONTROL TANK Refer to Section 15.5.26 for the description of this event.
15.4.10 REFERENCES
- 1. "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.
- 3. Deleted in Revision 12.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-73 Revision 23 December 2016
- 4. Deleted in Revision 12.
- 5. Deleted in Revision 12.
- 6. Deleted in Revision 12.
- 7. Deleted in Revision 12.
- 8. Deleted in Revision 12.
- 9. Deleted in Revision 22.
- 10. Deleted in Revision 12.
- 11. Deleted in Revision 12.
- 12. Deleted in Revision 12.
- 13. Deleted in Revision 12.
- 14. Deleted in Revision 12.
- 15. Deleted in Revision 12.
- 16. F. S. Moody, "Transactions of the ASME," Journal of Heat Transfer, February 1965, Figure 3, page 134. 17. H. G. Hargrove, FACTRAN, A Fortran IV Code for Thermal Transients in a UO 2 Fuel Rod, WCAP-7908-A, December 1989.
- 18. T. W. T. Burnett, Reactor Protection System Diversity in Westinghouse Pressurized Water Reactor, WCAP-7306, April 1969.
- 19. T. G. Taxelius, ed. "Annual Report - SPERT Project, October 1968 September 1969," Idaho Nuclear Corporation IN-1370, June 1970.
- 20. R. C. Liimatainen and F. J. Testa, Studies in TREAT of Zircaloy-2-Clad, UO 2 Core Simulated Fuel Elements, ANL-7225, January - June 1966, p. 177, November 1966.
- 21. D. H. Risher, Jr., An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods, WCAP-7588, Revision 1-A, January 1975.
- 22. Deleted in Revision 21.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-74 Revision 23 December 2016
- 23. A. A. Bishop, et al., "Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31, August 1965.
- 24. Deleted in Revision 12.
- 25. Deleted in Revision 12.
- 26. T. W. T. Burnett, et al., LOFTRAN Code Description, WCAP-7907-P-A (Proprietary), WCAP-7907-A (Nonproprietary), April 1984.
- 27. Deleted in Revision 13.
- 28. F. W. Dittus and L. M. K. Boelter, University of California (Berkeley), Publs. Eng., 2,433, 1930.
- 29. W. H. Jens and P. A. Lottes, Analysis of Heat Transfer, Burnout, Pressure Drop, and Density Data for High Pressure Water, USAEC Report ANL-4627, 1951.
- 30. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.
- 31. J.S. Shefcheck, Application of the THINC Program to PWR Design, WCAP-7359-L, August 1969 (Westinghouse Proprietary) and WCAP-7838, January
1972. 32. ANSI/ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors, 1979.
- 33. Deleted in Revision 12.
- 34. Deleted in Revision 12.
- 35. Deleted in Revision 12.
- 36. Deleted in Revision 12
- 37. Deleted in Revision 12.
- 38. Deleted in Revision 12.
- 39. Deleted in Revision 12.
- 40. Deleted in Revision 18.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-75 Revision 23 December 2016
- 41. Diablo Canyon Units 1 and 2 Replacement Steam Generator Program - NSSS Licensing Report, WCAP-16638 (Proprietary), Revision 1, January 2008.
- 42. Plant Manual, Volume 3A, Emergency Operating Procedures, Diablo Canyon Power Plant Units 1 and 2.
- 43. Deleted in Revision 12.
- 44. Deleted in Revision 12.
- 45. Deleted in Revision 12.
- 46. Deleted in Revision 16.
- 47. Deleted in Revision 18.
- 48. Deleted in Revision 12.
- 49. Deleted in Revision 12.
- 50. Deleted in Revision 12.
- 51. Deleted in Revision 12.
- 52. PG&E Calculation N-160, Liquid Hold up Tank Rupture Doses, Revision 0, October 11, 1994.
- 53. Deleted in Revision 18.
- 54. Emergency Core Cooling Systems: Revisions to Acceptance Criteria, Federal Register, V53, N180, pp. 35996-36005, September 16, 1988.
- 55. Regulatory Guide 1.157, Best-Estimate Calculations of Emergency Core Cooling System Performance, USNRC, May 1989. 56. Westinghouse Code Qualification Document for Best Estimate Loss of Coolant Accident Analysis, WCAP-12945-P (Proprietary), Volumes I-V.
- 57. Deleted in Revision 18.
- 58. Deleted in Revision 21.
- 59. Deleted in Revision 18.
- 60. Best Estimate Analysis of the Large Break Loss of Coolant Accident for Diablo Canyon Power Plant Units 1 and 2 to Support 24-Month Fuel Cycles and Unit 1 Uprating, WCAP-14775, January 1997.
DCPP UNITS 1 &
2 FSAR UPDATE 15.4-76 Revision 23 December 2016
- 61. Containment Pressure Analysis Code (COCO), WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974.
- 62. H. Chelemer, et al., Improved thermal Design Procedure, WCAP-8567-P-A (Proprietary) and WCAP-8568-A (Non-Proprietary), February 1989.
- 63. Nuclear Safety Advisory Letter NSAL-02-04, Steam Line Break During Mode 3, October 30, 2002.
- 64. WCAP-11677, Pressurizer Safety Relief Valve Operation for Water Discharge During Feedwater Line Break, January 1988.
- 65. Deleted in Revision 21.
- 66. Deleted in Revision 21.
- 67. Letter from W.R. Rice (Westinghouse Electric Company) to J. Ballard (PG&E), Diablo Canyon Unit1, BELOCA Reanalysis Final En gineering Report, PGE-03-33, June 6, 2003.
- 68. SECY-83-472, Information Report from W.J. Dircks to the Commission, Emergency Core Cooling System Analysis Methods, November 17, 1983.
- 69. Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Method (ASTRUM), WCAP-16009-P-A, (Proprietary), January 2005.
- 70. RETRAN-02 Modeling and Qualifi cation for Westinghouse Pressurized Water Reactor Non-LOCA Safety Analyses, WCAP-14882-P-A (Proprietary), April 1999, and WCAP-15234-A (Non-Proprietary), May 1999.
- 71. Deleted in Revision 21.
- 72. PGE-10-56, PG&E Diablo Canyon Units 1 and 2, Steam Generator Tube Rupture Margin to Overfill Analysis (CN-CR A-10-45 Rev. 0), October 18, 2010
- 73. WCAP-16443-P , Rev. 1, Diablo Canyon Unit 2 ASTRUM BE-LBLOCA Engineering Report, November 2005
- 74. Barrett, G.O., et al., Pressurizer Safety Valve Set Pressure Shift, WCAP-12910, Rev 1-A (Proprietary), May 1993.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-1 Revision 22 May 2015 15.5 RADIOLOGICAL CONSEQUENCES OF PLANT ACCIDENTS The purposes of this section are: (a) to identify accidental events that could cause radiological consequences, (b) to provide an assessment of the consequences of these accidents, and (c) to demonstrate that the potential consequences of these occurrences
are within the limits, guidelines, and regulations established by the NRC.
An accident is an unexpected chain of events
- that is, a process, rather than a single event. In the analyses reported in this section, the basic events involved in various
possible plant accidents are identified and studied with regard to the performance of the
engineered safety features (ESF). The full spectrum of plant conditions has been divided into four categories in accordance with their anticipated frequency of occurrence
and risk to the public. The four categories as defined above 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 these conditions is that the most frequent occurrences must yield little or no 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.
These categories and principles were developed by the American Nuclear Society (Reference 1). Similar, though not identical, categories have been defined in the guide
to the Preparation of Environmental Reports (Reference 3). While some differences
exist in the manner of sorting the different accidents into categories in these documents, the basic principles are the same.
It should also be noted that the range of plant operating parameters included in the Condition I category, and some of those in the Condition II category, fall in the range of normal operation. For this reason, the radioa ctive releases and radiological exposures associated with these conditions are analyzed in Chapter 11 and are not discussed
separately in this chapter. The analyses of the variations in system parameters
associated with Condition I occurrences or operating modes are discussed in Chapter 7
since these states are not accident conditions. In addition, some of the events identified as potential accidents in Regulatory Guide 1.70, Revision 1 (Reference 2), have no significant radiological consequences, or result in minor releases within the range of normal releases, and are thus not analyzed separately in this chapter.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-2 Revision 22 May 2015 15.5.1 DESIGN BASES The following regulatory requirements, including Code of Federal Regulations (CFR) 10 CFR Part 100, General Design Criteria (GDC), Safety Guides, and Regulatory Guides are applicable to the DCPP radiologi cal consequence analyses presented in this Chapter. They form the bases of the acceptance criteria and methodologies as described in the following Sections:
(1) 10 CFR Part 100, Reactor Site Criteria (2) 10 CFR 50.67, Accident Source Term (3) General Design Criterion 19, 1971 Control Room (4) Regulatory Guide 1.4, Revision 1, Assumptions Used for Evaluating the Potential Radiological Conseq uences of a Loss of Coolant Accident for Pressurized Water Reactors (5) Safety Guide 7, March 1971, Control of Combustible Gas Concentrations in Containment (6) Safety Guide 24, March 1972, Assumptions Used for Evaluating the Potential Radiological Consequ ences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure (7) Safety Guide 25, March 1972, Assumptions Used for Evaluating the Potential Radiological Consequ ences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (8) Regulatory Guide 1.183, July 2000, Alternative Radiological Source Terms for Evaluating Design Basis Accident s at Nuclear Power Reactors (9) Regulatory Guide 1.195, May 2003, Methods and Assumptions for Evaluating Radiological Consequ ences of Design Basis Accidents at Light-Water Nuclear Power Reactors The following table summarizes the accident events that have been evaluated for radiological consequences. The table identif ies the applicable UFSAR Section describing the analysis and results for each event, the offsite/onsite locations and applicable dose limits, and the radiologica l analysis and isotopic core inventory codes used.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-3 Revision 22 May 2015 Accident Event FSAR Section Boundary Dose Limit Radiological Analysis Code(s) Isotopic Core Inventory Code(s) CONDITION II Loss of Electrical Load 15.5.10 EAB and LPZ Thyroid Whole Body 300 rem 25 rem EMERALD EMERALD CONDITION III Small Break LOCA 15.5.11 EAB and LPZ Thyroid Whole Body 300 rem 25 rem EMERALD EMERALD Minor Secondary System Pipe Breaks 15.5.12 EAB and LPZ Thyroid Whole Body 300 rem 25 rem N/A Refer to Section 15.5.12 N/A Refer to Section 15.5.12 Inadvertent Loading of a Fuel Assembly 15.5.13 EAB and LPZ Thyroid Whole Body 300 rem 25 rem N/A Refer to Section 15.5.13 N/A Refer to Section 15.5.13 Complete Loss of Forced Reactor Coolant Flow 15.5.14 EAB and LPZ Thyroid Whole Body 300 rem 25 rem N/A Refer to Section 15.5.14 N/A Refer to Section 15.5.14 Under-Frequency 15.5.15 EAB and LPZ Thyroid Whole Body 300 rem 25 rem EMERALD EMERALD Single Rod Cluster Control Assembly Withdrawal 15.5.16 EAB and LPZ Thyroid Whole Body 300 rem 25 rem EMERALD EMERALD CONDITION IV Large Break LOCA 15.5.17 EAB and LPZ Thyroid Whole Body Control Room Thyroid Whole Body 300 rem 25 rem 30 rem 5 rem EMERALD LOCADOSE EMERALD ORIGEN-2 DCPP UNITS 1 &
2 FSAR UPDATE 15.5-4 Revision 22 May 2015 Accident Event FSAR Section Boundary Dose Limit Radiological Analysis Code(s) Isotopic Core Inventory Code(s) Main Steam Line Break 15.5.18 EAB and LPZ Pre-Accident Iodine Spike Thyroid Whole Body Accident-initiated Iodine Spike Thyroid Whole Body Control Room Thyroid Whole Body 300 rem 25 rem 30 rem 2.5 rem 30 rem 5 rem LOCADOSE ORIGEN-2 Main Feedwater Line Break 15.5.19 EAB and LPZ Thyroid Whole Body 300 rem 25 rem N/A Refer to Section 15.5.19 N/A Refer to Section 15.5.19 Steam Generator Tube Rupture 15.5.20 EAB and LPZ Pre-Accident Iodine Spike Thyroid Whole Body Accident-initiated Iodine Spike Thyroid Whole Body Control Room Thyroid Whole Body 300 rem 25 rem 30 rem 2.5 rem 30 rem 5 rem RADTRAD EMERALD-NORMAL Locked Rotor 15.5.21 EAB and LPZ Thyroid Whole Body Control Room Thyroid Whole Body 300 rem 25 rem 30 rem 5 rem EMERALD EMERALD Fuel Handling- Fuel Handling Area 15.5.22.1 EAB and LPZ Control Room 0.063 Sv TEDE (6.3 rem) 0.05 Sv TEDE (5 rem) LOCADOSE ORIGEN-2 DCPP UNITS 1 &
2 FSAR UPDATE 15.5-5 Revision 22 May 2015 Accident Event FSAR Section Boundary Dose Limit Radiological Analysis Code(s) Isotopic Core Inventory Code(s) Fuel Handling-Inside Containment 15.5.22.2 EAB and LPZ Thyroid Whole Body Control Room Thyroid Whole Body 75 rem 6 rem 30 rem 5 rem LOCADOSE ORIGEN-2 Control Rod Ejection 15.5.23 EAB and LPZ Thyroid Whole Body Control Room Thyroid Whole Body 300 rem 25 rem 30 rem 5 rem EMERALD EMERALD Waste Gas Decay Tank Rupture 15.5.24 EAB and LPZ Thyroid Whole Body 300 rem 25 rem EMERALD EMERALD Liquid Holdup Tank Rupture 15.5.25 EAB and LPZ Thyroid Whole Body 300 rem 25 rem LOCADOSE ORIGEN-2 Volume Control Tank Rupture 15.5.26 EAB and LPZ Thyroid Whole Body 300 rem 25 rem EMERALD EMERALD 15.5.2 APPROACH TO ANALYSE S OF RADIOLOGICAL EFFECTS OF ACCIDENTS
The potential radiological effects of plant accidents are analyzed by the evaluation of all physical factors involved in each chain of events which might result in radiation exposures to humans. These factors include the meteorological conditions existing at the time of the accident, the radionucli de uptake rates, exposure times and distances, as well as the many factors which depend on the plant design and mode of operation.
In these analyses, the factors affecting the consequences of each accident are
identified and evaluated, and uncertainties in their values are discussed. Because
some degree of uncertainty always exists in the prediction of these factors, it has
become general practice to assume conservative values in making calculated estimates
of radiation doses. For example, it is customarily assumed that the accident occurs at a
time when very unfavorable weather conditions exist, and that the performance of the
plant engineered safety systems is degraded by unexpected failures. The use of these
unfavorable values for the various factors involved in the analysis provides assurance
that each safety system has been designed adequately; that is, with sufficient capacity
to cover the full range of effects to which each system could be subjected. For this
reason, these conservative values for each factor have been called design basis values.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-6 Revision 22 May 2015 In a similar way, the specific chain of events in which all unfavorable factors are coincidentally assumed to occur has been called a design basis accident (DBA). In the process of safety review and licensing, the radiation exposure levels calculated for
the DBA are compared to the guideline values established in 10 CFR 100.11 and 10 CFR 50.67, and if these calculated expos ures fall below the guideline levels, the plant safety systems are judged to be adequate.
The calculated exposures resulting from a DBA are generally far in excess of what would be expected and do not provide a realistic means of assessing the expected radiological effects of real plant accidents.
For this reason, the original licensing basis included two evaluations, or cases, for each accident. The first case, called the expected case, used values, for each factor involved in the accident, which are estimates of the actual values expected to occur if the
accident took place. The resulting doses were close to the doses expected to result from an accident of this type. The second case, the DBA, used the customary conservative assumptions. The calculated doses for the DBA, while not a realistic
estimate of expected doses, can provide a basis for determination of the design
adequacy of the plant safety systems.
The specific values of all important parameters, data, and assumptions used in the
radiological exposure calculations are listed in the following sections. The details of the implementation of the equations, models, and parameters for accidents evaluated using the original licensing basis computer code EMERALD are described in the description of the EMERALD computer program (Reference 4) and the EMERALD-NORMAL
computer program (Reference 5), which are described briefly in Section 15.5.8.1.
As discussed earlier, some of the radiological source terms for accidents and some of
the releases resulting from Condition I and Condition II events have been included in Chapter 11.
15.5.3 ACTIVITY INVENTORIES IN THE PLAN T PRIOR TO ACCIDENTS The fission product inventories in the reactor core, the fuel rod gaps, and the primary
coolant prior to an accident have been calculated using the same assumptions, models, and physical data described in Section 11.1, but for different core and plant operating conditions. The pre-accident inventories were calculated using the EMERALD computer code and are similar to those calculated for Tables 11.1-1 through 11.1-12 by the EMERALD-NORMAL code, except for slight differences in some nuclides due to different initial core inventories and irradiation times in the accident calculation.
The steam system operating conditions assumed for the calculation of pre-accident secondary system inventories are listed in Table 11.1-23. It should be noted that these
steam system flowrates and masses are approximate lumped values, used for activity
balances only, and assume gross lumping of feedwater system component flows and
masses. While these values are adequate for activity balances, they should not be DCPP UNITS 1 &
2 FSAR UPDATE 15.5-7 Revision 22 May 2015 used in the context of actual plant flow and energy balances. The activity inventories and concentrations existing in the secondary system are listed in Table 11.1-26.
Activity inventories in various radwaste system tanks are also listed in sections of
Chapters 11 and 12 and will be cross-referenced in the sections of this chapter dealing
with accidental releases from these tanks.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
Refueling shutdown studies at operating Westinghouse PWRs indicate that, during cooldown and depressurization of the RCS, a release of activated corrosion products
and fission products from defective fuel has been found to increase the coolant activity
level above that experienced during steady state operation. An increased core activity release of this sort, commonly referred to as "spiking," could be expected to occur
during the depressurization of the RCS as the result of an accident, and should
therefore be taken into account in the calculation of post-accident releases of primary coolant to the environment.
Table 15.5-1 illustrates the anticipated coolant activity increases of several isotopes for DCPP during shutdown. This table lists the expected activities during steady state
operation and anticipated peak activities during plant cooldown operations. These data are based on measurements from an operating PWR that is similar in design to the DCPP and has operated with significant fuel defects. The measured activity levels for
the operating plant are also included in Table 15.5-1.
The dominant nongaseous fission product released to the coolant during system depressurization is I-131. The activity level in the coolant was observed to be higher
than the normal operating level for nearly a week following initial plant shutdown with the system purification rate varying between approximately 1 x 10
-5 and 3 x 10-5 per second. Although lesser in magnitude, the other fission product particulates (cesium isotopes) exhibited a simi lar pattern of release and removal by purification. It is reasonable to project these data to the DCPP since the purification constants are
similar, and it is standard operating procedure to purify the coolant through the
demineralizers during plant cooldown.
Fission gas data from operating plants indicate a maximum increase of approximately
1.5 over the normal coolant gas activity concentration. However, system
degassification procedures are implem ented prior to and during shutdowns, and have proven to be an effective means for reducing the gaseous activity concentration and
controlling the activity to levels lower than the steady state value during the entire
cooldown and depressurization procedure. Although a steady state Xe-133 concentration of 127
µCi/gm was observed p rior to degassification procedures (see Table 15.5-1), the ma ximum coolant con centration during the reactor depressurization was 65
µCi/gm. Further, the coolant activity was then reduced to approximately 1
µCi/gm in less than two days of degassification.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-8 Revision 22 May 2015 The corrosion product activity releases have been determined to be predominantly dissolved Co-58. From Table 15.5-1, it is noted that this contribution is less than
1 percent of the total expected coolant activity and is, therefore, considered to be a
minor contribution.
For the calculation of the effect of spiking on accidental plant releases, the original licensing basis assumed the dominant isotopes were iodines, and all others were neglected. Using the measured I-131 concentrations given in Table 15.5-1 and a
primary purification rate of 1 x 10
-5 per second, effective I-131 fuel escape rate to the reactor coolant during a spike of 30 times the normal equilibrium value was calculated.
This value was then applied to all iodine isotopes. Subsequent analyses assumed an accident-initiated spike of 500 times (for MSLB, refer to Section 15.5.18) and 335 times (for SGTR, refer to Section 15.5.20) the normal equilibrium value, as described in the appropriate subsections that follow, to be consistent with Technical Specification 3.4.16.
The duration of the spike was assumed to be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This assumption can be justified
by examining graphs of I-131 coolant concentration versus time during shutdowns for
operating BWR plants (Reference 14). The assumption that the fuel escape rate
continues at the elevated rates discussed above for the full 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the spike is conservative. The effect of iodine spiking was included in all accidents that involved leakage of primary coolant directly or indirectly to the environment.
15.5.4 EFFECTS OF PLUTONIUM INVENTORY ON POTENTIAL ACCIDENT DOSES HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
Because of the somewhat higher fission yields of some isotopes associated with thermal fissions in Pu-239, a sensitivity study was conducted to determine the possible influence of this effect on potential accident doses.
This study demonstrates that accident doses are only slightly affected by the
incorporation of Pu-239 fission yields into total core fission yields, even using the EOL
plutonium inventories. The resulting differences, listed in Table 15.5-2, indicate that
thyroid doses generally increase from 4 to 6 percent, and whole body doses generally
decreased, from 2 to 5 percent, assuming the accident occurred at EOL.
In this study, total core fission yields were calculated by a mass weighting of U-235 fission yields and Pu-239 fission yields. Because the core mass of U-235 is
considerably greater than the core mass of Pu-239, total core fission yields are close to
U-235 fission yields. The masses of U-238 and Pu-24 1 that fission are extre mely small, and thus U-238 and Pu-241 have essentially no effect on the total core fission yields.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-9 Revision 22 May 2015 15.5.5 POST-ACCIDENT METEOROLOGICAL CONDITIONS For the analyses of offsite doses from the DBAs, the rare and unfavorable set of
atmospheric dilution factors assumed in the NRC Regulatory Guide 1.4, Revision 1 (Reference 6) was used. On the basis of meteorological data collected at the DCPP site, these unfavorable dilution factors, assumed for the design bases cases, are not expected to exist for onshore wind directions more than 5 percent of the time. The
particular values used for this site are given in Table 15.5-3.
For the analyses of offsite doses from the expected case accidents, the assumed
atmospheric dilution factors are listed in Table 15.5-4. For these cases, 10 percent of
the design basis case numbers were used. On the basis of study of the site data at
DCPP, this assumption will result in calculat ed exposures higher than would be expected.
Because of the low probability of occurrence associated with these assumed dilution factors, significant downwind decay, variable shifts in population distribution due to
possible emergency evacuation, and large variations in concentrations due to downwind topographical characteristics, appropriate assumptions for population exposure (man-rem) estimates following a significant accidenta l release are difficult to select. It is clear that using the same factors of conservatism established for individual exposures at
locations near the site (the regulatory guide dilution factors) would yield calculated population exposures much higher than could physica lly occur. For these reasons, the population exposures (man-rem) for the expected cases have been calculated using the
long-term dilution factors given in Table 15.5-5, and ten times these values have been
assumed for the DBA cases.
Effects of release duration on downw ind ground level concentration have been measured directly and determined theoretically from knowledge of the horizontal and
vertical spectrum of turbulence. Both the observations and theory generally agree that
only the horizontal components of turbulence near ground level contain any significant
amount of energy in periods longer than a few seconds. As a result, only the lateral
dimension of the cloud need be modified for concentration estimates for noncontinuous
releases. Slade (Reference 7) using the approach recommended by Cramer, gives a time-dependent adjustment of the lateral component of turbulence to be:
(T) = (T o) (T/T o)0.2 (15.5-1) where:
(T) = lateral intensity of turbulence of a time period T, where T is a value less than 10 minutes (T o) = lateral intensity of turbulence measured over a time period T o , where T o is on the order of 10 minutes
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-10 Revision 22 May 2015 Near a source there is a direct linear relationship between and the plume crosswind dimension y so that the y versus distance curves presented by Slade can be directly scaled by the factor (T/T o)0.2 to provide estimates of a reference y at about 100 meters downwind from the source. Beyond this distance, the lateral expansion rates for continuous and noncontinuous point source releases are approximately the same, and
thus the ratio of short-term release concentration to continuous release concentration
for point sources is independent of stability class, downwind distance, or windspeed.
For distances less than a few thousand meters the ratio approaches unity as the volume
of the source increases.
Using the above scaling concept, the dilution equation in Regulatory Guide 1.4, Revision 1, and the cloud dimension curves given by Slade, the ratio of short-term release concentration to continuous release concentration was calculated for several
different release durations (Figure 15.5-1).
For a10-second duration, the short-term dilution factor is only 2.3 higher than the continuous release dilution factor, and thus the
appropriate short-term release correction is within the uncertainty limits of the
continuous release dilution factor.
The various plant accidents considered in Sections 15.2, 15.3, and 15.4 may result in
activity release through various pathways: containment leakage, secondary steam
dumping, ventilation discharge, and radioactive waste system discharge.
Post-accident containment leakage is a slow continuous process, and thus continuous release dilution factors apply for these cases.
Because of secondary loop isolation capabilities and because significant activity release
is accompanied by large steam release, secondary steam dumping accidents release
significant quantities of activity only through relief valves. Relief valve flow limitations
combined with large steam release result in activity releases of long duration. Thus continuous release dilution factors apply for these cases.
The approximate duration of a ventilation discharge activity release can be estimated by
dividing the volume of contaminated air by the discharge flowrate. Because estimates of release duration for liquid holdup tank rupture, gas decay tank rupture, volume
control tank rupture, and fuel handling area accident are all over in 10 minutes, continuous release dilution factors apply for these cases.
Continuous release dilution factors have been appli ed to all Conditions II, III, and IV accidents discussed in Chapter 15 for the following reasons:
(1) Almost all Conditions II, III, and IV releases are definitely long-term releases (2) Releases that might be considered short-term releases result in exposures well within 10 CFR Part 100 limits
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-11 Revision 22 May 2015 (3) Short-term release dilution factors are only about twice as high as continuous release dilution factors (4) The appropriate short-term release corrections are within the range of the uncertainties in the continuous release dilution factors Furthermore, the above reasons indicate that a more sophisticated or complex
short-term release dilution model is not justified.
The atmospheric dispersion factors for pressurization and infiltration air flows to the control room are analyzed using the modified Halitsky /Q methodology, which is discussed below.
As a result of the TMI accident, the NRC, in NUREG-0737 Section III.D.3.4, asked all nuclear power plants to review their post-LOCA control room habitability designs using
the guidance of Standard Review Plan (SRP) 6.4 and the 1974 Murphy-Campe (M-C) paper (Reference 17). These reviews concluded that the atmospheric dispersion factor
(/Q) methodology recommended in the M-C paper was overly conservative and inappropriate for most of the plant designs. The M-C equations are based primarily on the Halitsky data for round-topped EBR-II (PWR type) containments and are valid only
for intake locations at least a half containment diameter from the containment wall. In
most cases, however, the intake locations are closer to the building causing the wake.
Thus, review of recent literature on building wake /Qs, models, wind tunnel tests, and field measurements resulted in the modified Halitsky /Q model.
Historically, the preliminary work on building wake /Qs was based on a series of wind tunnel tests by James Halitsky et al. Halitsky summarized these results in Meteorology and Atomic Energy 1968, D. H. Slade, Editor (Reference 7). In 1974 K. Murphy and K. Campe of the NRC published their paper based on a survey of existing data. This
/Q methodology, which presented equations without derivation or justification, was adopted as the interim methodology in SRP 6.4 in 1975. Since that time, a series of
actual building wake /Q measurements have been conducted at Rancho Seco (Reference 25), and several other papers have been published documenting the results of additional wind tunnel tests (see References 26 through 31).
The Diablo Canyon plant complex is composed of square-edged buildings and two cylindrical containment buildi ngs. Infiltration air into the control room would come from the auxiliary building, which has air intakes slightly above the control room. This intake
of air will be subject to building wake caused by the portion of the containment building above the highest roof elevation of the auxiliary building. Pressurization air for the
control room is provided from intakes on the turbine building. The intake will be subject
to building wake caused by a portion of the containment building above the turbine building roof and a portion of the turbine buil ding wall facing west and the wall facing north.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-12 Revision 22 May 2015 J. Halitsky's efforts, summarized in Reference 7, present the basic equation as follows:
uAKQ//= (15.5-2) where A = cross sectional area, m 2 orthogonal to u u = wind speed, m/s K = isopleth (concentration coefficient - dimensionless)
It is found in many cases that the above Ha litsky equation still provides a reasonable estimate of /Q. The following correction factors can be applied to this equation to account for situation and plant-specific features:
- Stream line flows are used in most wind tunnel tests
- Release points are generally much higher than 10 meters above ground
- Null wind velocity is observed at certain periods of time
- Isothermal temperatures are used in wind tunnel tests
- Buoyancy and jet momentum effects are ignored Typical 1 hr field tests account for plume meander effects, while 3 to 5 minute wind
tunnel tests do not.
A modified Halitsky /Q methodology, formulated by R. Bhatia, et al (Reference 32), is presented below.
/Q = K x f 1 x f 2 x f 3 x f 4 x f 5 x f 6 (sec/m 3) (15.5-3) A u This modified Halitsky methodology is inherently conservative because the wind is
assumed to be blowing towards the control room during the first or worst part of the
accident, and because 5 percent wind speeds are used rather than 50 percent. In
addition, the adjustment factors are always biased towards the minimum reduction that the data justifies.
As a test of the modified Halitsky method, calculated values of /Q, without using factors f 4 and f 5 due to their uncertainty, were compared to the 1-hour field test /Q data from Rancho Seco. Only one /Q was found to be higher than the calculated value.
This was due to an external wake influence caused by wind chann eling between the DCPP UNITS 1 &
2 FSAR UPDATE 15.5-13 Revision 22 May 2015 nearby cooling towers. The wind channeling prevented the normal wake turbulence and variation effects over time, which normally spread the plume over a wide area. In most cases the modified Halitsky /Q was found to be a conservative estimate of the measured /Q; in some cases it was significantly higher.
The choice of K factors and the suggested modifying factors, f 1 , f 2 , etc., are discussed below. K factors:
The choice of an appropriate K factor from the wind tunnel test data is critical for the /Q estimate to be valid. Halitsky in Reference 7 has several sets of K isopleths for round-topped containments (f or PWRs) and block buildings (for BWRs). Multiple building complexes must be simulated by single equivalent structures. The effluent velocity to wind spe ed ratio of approximately 1 is valid for most power coolant systems. Various angles of wind incidence are shown to account for vortexing that could result in worse conditions than a wind normal to the building face. K factors should be estimated for various combinations of wind
incidence angle and the appropria te effective building cross-sectional area causing the wake (not just the containment area) to determine the peak value, as
was done by Walker (Reference 26).
The K factors were determined from Figure 5.29c in Reference 7, based on a
conservative analysis of the locations for infiltration and pressurization intake
airflows and the appropriate dimensions relative to the containment. A single
pressurization intake nearest the containment was assumed. The selected
K factors and appropriate building cross-sectional areas used for the base /Q values are given below. The 5 percent wind speed was derived from an analysis of Diablo Canyon meteorological data over a period of 10 years.
Case K u (meter/second) A(m
- 2) Base /Q (sec/m 3) Pressurization 4 1 3690 1.084x10
-3 Infiltration 5 1 1661 3.01x10
-3 u , wind speed:
Halitsky's K values are based on wind speeds measured at the top of the containment or building. Therefore, the M-C 5 percent wind speed at a
10 meter height should be adjusted to the actual speed at the top of containment or release point. The 5 percent wind speed is adjusted using the formulation
presented by Wilson (Reference 30) as follows.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-14 Revision 22 May 2015
.23 Z Zuu Ref T= (15.5-4) where:
T u = wind speed at height Z Ref Z = 10 meters (5 percent wind speed reference height) f 1 , wind speed change factor/ f 2 , wind direction change factor:
The factors shown below were used. They are based on Diablo Canyon meteorological data for a 10 year period of record.
Time Periods f 1 f 2 0 - 8 hrs 1.0 1.0 8 - 24 hrs 0.83 0.92 24 - 96 hrs 0.66 0.84 96 - 720 hrs 0.48 0.67 f 3 , wind turbulence effect:
Wilson in Reference 30 and field tests confirm Halitsky's statement that his K isopleths are a factor of 5 to 10 too conservative due to not accounting for
random fluctuations of the wind approaching the building. Therefore, a factor of 0.2 was used for f
- 3. f 4 , elevated release effect:
Bouwmeester et al. (Reference 31) indicate that there are up to 10 null wind speed conditions during an hour of data collection. During these periods the effects of jet momentum, plume rise and buoyancy would result in the radioactive
effluent being discharged above the effective wake boundary and thus not
entering the wake cavity. A reduction factor of 1 was used.
F 5 , time averaging effects:
Wind speed variations and wind direction meandering effects are not modeled in wind tunnel tests to account for this effect.
Reference 31 indicates the use of the following equation:
1/2 m t p t m C p C= (15.5-5)
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-15 Revision 22 May 2015 where: p C= prototype concentration m C= model concentration p t = prototype sampling time m t= model equivalent sampling time Normal wind tunnel data is taken for 3 to 10 minute samples. Thus, for a 1-hour
field test, p C = 0.22 to 0.41 m C , and for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> field test, p C= 0.08 to 0.14.
m C A value of 0.5 was conservatively assumed for f
- 5. f 6 , adjustments to top of containment:
To account for wind speed at the top of containment, instead of the M-C 5 percent wind speed at 10 meter height, the factor f 6 = u/u T was included. The f 6 value equals 0.65.
Table 15.5-6 presents the resultant atmospheric dispersion factors (/Q) calculated using the modified Halitsky /Q methodology. These dispersion factors do not take credit for dual pressurization inlets and do not include the control room occupancy factors.
15.5.6 RATES OF ISOTOPE INHALATION The breathing rates used in the calculations of inhalation doses are listed in
Table 15.5-7. These values are based on the average daily breathing rates assumed
in ICRP Publication 2 (Reference 8) which are also used in Regulatory Guide 1.4, Revision 1. The active breathing rates are used for all onsite dose calculations, which are based on expected exposure times.
15.5.7 POPULATION DISTRIBUTION The distribution of population surrounding the plant site, which was used for the
population exposure calculations, is discussed in Section 2.1, and the population distribution used is listed in Table 15.5-8. The actual post-accident population distribution could be significantly lower if any evacuation plan were implemented.
15.5.8 RADIOLOGICAL ANALYSIS PROGRAMS 15.5.8.1 Description of the EMERAL D (Revision I) and EMERALD-NORMAL Program The EMERALD program (Reference 4) is designed for the calculation of radiation releases and exposures resulting from abnormal operation of a large PWR. The DCPP UNITS 1 &
2 FSAR UPDATE 15.5-16 Revision 22 May 2015 approach used in EMERALD is similar to an analog simulation of a real system. Each component or volume in the plant that contains a radioactive material is represented by a subroutine, which keeps track of the production, transfer, decay, and absorption of
radioactivity in that volume. During the course of the analysis of an accident, activity is
transferred from subroutine to subroutine in the program as it would be transferred from
place to place in the plant. For example, in the calculation of the doses resulting from a
LOCA, the program first calculates the activity built up in the fuel before the accident, then releases some of this activity to the containment volume. Some of this activity is
then released to the atmosphere. The rates of transfer, leakage, production, cleanup, decay, and release are read in as input to the program.
Subroutines are also included that calculate the onsite and offsite radiation exposures at various distances for individual isotopes and sums of isotopes. The program contains a
library of physical data for 25 isotopes of most interest in licensing calculations, and
other isotopes can be added or substituted. Because of the flexible nature of the
simulation approach, the EMERALD program can be used for most calculations
involving the production and relea se of radioactive materials, including design, operational and licensing studies. The complete description of the program, including
models and equations, is contained in Reference 4.
The EMERALD-NORMAL program (Reference 5) is a program incorporating the features of EMERALD, but designed specifically for releases from normal and near-normal operating conditions. It contains an expa nded library of isotopes, including all those of interest in gaseous and liquid environmental exposures. Models for a radwaste system are included, using the specific configuration of radwaste system components in
the DCPP. The program contains a subroutine for doses via liquid release pathways developed by the Bechtel Corporation and a tritium subroutine. The code calculates activity inventories in various radwaste tanks and plant components which are used for
the initial conditions for accidents involving th ese tasks. In addition, it is used in some near-normal plant conditions classified in this document as Condition I and Condition II
and discussed in Chapter 11.
15.5.8.2 Description of the LOCADOSE Program The LOCADOSE program (Reference 47) is designed to calculate radionuclide activities, integrated activities, and releases from a number of arbitrarily specified regions. One region is specified as the environment. Doses and dose rates for five organs (thyroid, lung, bone, beta skin, and whole body) can be calculated for each region, and for a number of offsite locations with specified atmospheric dispersion factors. The control room can be specified as a special region for convenience in modeling airborne doses to the control room operators.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-17 Revision 22 May 2015 15.5.8.3 Description of the ORIGEN-2 Program The core inventory and gamma ray energy spectra of post-accident fission products for selected accidents (See Section 15.5.1) were computed using the ORIGEN-2 computer program.
ORIGEN-2 (Reference 50) is a versatile point depletion and decay computer code for use in simulating nuclear fuel cycles and calculating the nuclide compositions of materials contained therein. This code represents a revision and update of the original ORIGEN computer code which has been distributed world-wide beginning in the early 1970s. Included in it are provisions for incorporating data generated by more sophisticated reactor physics codes, free-format input, the ability to simulate a wide variety of fuel cycle flowsheets, and more fl exible and controllable output features.
15.5.8.4 Description of the ISOSHLD Program ISOSHLD (Reference 9) is a computer code used to perform gamma ray shielding
calculations for isotope sources in a wide variety of source and shield configurations.
Attenuation calculations are performed by point kernel integration; for most geometries
this is done by Simpson's rule numerical integration. Source strength in uniform or
exponential distribution (where applicable) may be calculated by the linked fission product inventory code RIBD or by other options as desired. Buildup factors are
calculated by the code based on the number of mean free paths of material between the
source and detector points, the effective atomic number of a particular shield region (the
last unless otherwise chosen), and the point isotropic Nuclear Development Associates (NDA) buildup data available as Taylor coefficients in the effective atomic number range of 4 to 82. Other data needed to solve most isotope shielding problems of practical interest are linked to ISOSHLD in various libraries.
15.5.8.5 Description of the ISOSHLD II Program ISOSHLD II (Reference 11) is a shielding co de that is principally intended for use in calculating the radiation dose, at a field point, from bremsstrahlung and/or decay gamma rays emitted by radioisotope sources. This program, with the newly-added bremsstrahlung mode, is an extension of the earlier version (ISOSHLD). Five shield regions can be handled with up to twenty materials per shield; the source is considered to be the first shield region, i.e., bremsstrahl ung and decay gamma rays are produced only in the source. Point kernel integration (over the source region) is used to calculate the radiation dose at a field point.
15.5.8.6 Description of the RADTRAD Program RADTRAD (Reference 52) uses a combinatio n of tables and numerical models of source term reduction phenomena to determine the time-dependent dose at user-specified locations for a given accident scena rio. It also provides the inventory, decay chain, and dose conversion factor tables needed for the dose calculation. The DCPP UNITS 1 &
2 FSAR UPDATE 15.5-18 Revision 22 May 2015 RADTRAD code can be used to assess occupational radiation exposure, typically in the control room, as well as site boundary doses, and to estimate the dose attenuation due to modification of a facility or accident sequence.
15.5.9 (DELETED)
The information previously in this section has been moved to Section 15.5.8.4.
15.5.10 RADIOLOGICAL CONSEQUENCES OF CONDITION II FAULTS 15.5.10.1 Acceptance Criteria The radiological consequences of accidents analyzed in Section 15.2 (or from other events involving insignificant core damage, but requiring atmospheric steam releases) shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
15.5.10.2 Identification of Causes and Accident Description As reported in Section 15.2, Condition II faults are not expected to cause breach of any of the fission product barriers, thus preventing fission product release from the core or plant. Under some conditions, however, small amounts of radioactive isotopes could be
released to the atmosphere following Condi tion II events as a result of atmospheric steam dumps required for plant cooldown.
The particular Condition II events that are
expected to result in some atmospheric steam release are:
(1) Loss of electrical load and/or turbine trip (2) Loss of normal feedwater (3) Loss of offsite power to the station auxiliaries (4) Accidental depressurization of the main steam system
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-19 Revision 22 May 2015 The amount of steam released followin g these events depends on the time relief valves remain open and the availability of condenser bypass cooling capacity.
The amount of radioactive iodine released depends on the amount of steam released and the iodine concentration in the steam generator water prior to the accident. An
analysis of potential thyroid doses has been made over the full range of possible values of these two key parameters; the results are presented in Figures 15.5-2 through 15.5-5. As shown on the figures, the potential thyroid doses are higher with increasing
steam releases and iodine concentrations. Figures 15.5-2 and 15.5-3 are results that
assume Regulatory Guide 1.4, Revision 1, assumptions for post-accident meteorology and breathing rates (Design Basis Case Assumptions). As shown in Figure 15.5-2, approximately 1.6 x 10 6 lbm of steam is the maximum steam release expected for a full cooldown without any condenser availability, and a steam release of approximately 1 x 10 5 lbm would result from releasing only the contents of one steam generator due to a safety valve release or steam line break with condenser cooling available.
Figures 15.5-2 through 15.5-5 illustrate the range of possible thyroid doses from
Condition II events. The highest anticipated doses would result from an event such as
loss of electrical load, and the potential thyroid and whole body doses from this
particular event have been analyzed using the EMERALD program. For both the design
basis case and the expected case, it was assumed that 656,000 lbm of steam would be
released to the atmosphere during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and an additional 1,035,000 lbm
would be released during the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for a limiting total release of about
1.7E+06 lbm (see Table 6.4.2-1 of Reference 49 for a summary of OSG and RSG
Condition II event steam releases). The assumptions used for meteorology, breathing
rates, population density, and other common factors were described in earlier paragraphs. Note that the preceding steam release quantities are associated with the original steam generator (OSG) loss of load (LOL) analysis which provides the basis for
the dose analysis of record. These values are greater than the replacement steam
generator (RSG) LOL with Tavg and Tfeed Range analysis releases (651,000 Ibm and 1,023,000 Ibm, respectively) and are therefore bounding since total dose is proportional
to total steam release.
For the design basis case, it was assumed that the plant had been operating
continuously with 1 percent fuel cladding defects and 1 gpm primary-to-secondary
leakage. For the expected case calculation, operation at 0.2 percent defects and
20 gallons per day to the secondary was assumed. In both cases, leakage of water
from primary to secondary was assumed to continue during cooldown at 75 percent of
the pre-accident rate during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at 50 percent of the pre-accident rate during the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These values were derived from primary-to-secondary
pressure differentials during cooldown.
It was also conservatively assumed for both cases that the iodine partition factor in the
steam generators releasing steam was 0.01, on a mas s basis. In addition, to account for the effect of iodine spiking, fuel escape rate coefficients for iodines of 30 times the normal operation values given in Table 11.1-8 were used for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> DCPP UNITS 1 &
2 FSAR UPDATE 15.5-20 Revision 22 May 2015 following the start of the accident.
Other detailed and less significant modeling assumptions are presented in Reference 4.
The resulting potential exposures from this type of accident are summarized in Table 15.5-9 and are consistent with the parametric analyses presented in
Figures 15.5-2 through 15.5-5.
15.5.10.3 Conclusions It can be concluded from the results discussed that the occurrence of any of the events analyzed in Section 15.2 (or from other events involving insignificant core damage, but requiring atmospheric steam releases) will result in insignificant radiation exposures.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are insignificant as shown in Table 15.5-9.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are insignificant as shown in Table 15.5-9.
15.5.11 RADIOLOGICAL CONSEQUENCES OF A SMALL-BREAK LOCA 15.5.11.1 Acceptance Criteria (1) The radiological consequences of a small-break loss-of-coolant-accident (SBLOCA) shall not exceed the dose limits of 10 CFR 100.11 as outlined below: i. An individual located at any point on the boundary of the exclusion area for the two hours immediately foll owing the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-21 Revision 22 May 2015 (2) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident.
15.5.11.2 Identification of Causes and Accident Description As discussed in Section 15.3.1, a SBLOCA is not expected to cause fuel cladding failure. For this reason, the only activity release to the containment will be the dissolved
noble gases and iodine in the reactor coolant water expelled from the pipe rupture.
Some of this activity could be released to the containment atmosphere as the water
flashes, and some of this amount could leak from the containment as a result of a rise in containment pressure.
The detailed description of the models used in calculating the potential exposures from
a small LOCA is contained in Reference 4, and a general description is contained in
Section 15.5.17 of this report. The specific assumptions used in the analysis are as
follows:
(1) The fission product inventories, meteorological data, breathing rates, and population data are described in Sections 15.5.3, 15.5.5, 15.5.6, and 15.5.7, respectively. Other common assumptions are described in the previous sections of 15.5.
(2) It has been assumed that all of the water contained in the RCS is released to the containment. For the design basis case, the reactor coolant activities associated with 1 percent defective cladding were used; and for the expected case, the reactor coolant activities associated with 0.2
percent defective cladding were used. These activities and concentrations
are listed in Tables 11.1-11 and 11.1-12, and all models and assumptions
used in determining these values are described in Section 11.1.
(3) Of the amounts of noble gases contained in the primary coolant, 100 percent is assumed to be released to the containment atmosphere at
the time of the accident. For the iodines, it is assumed that only 10 percent of the dissolved iodine in the coolant is released to the containment
atmosphere, due to the solubility of the iodine. It is assumed that the
amounts of iodine in chemical forms that are not affected by the spray system are negligible. These release fractions are used for both the design
basis case and the expected case.
(4) In addition, to account for the ef fect of iodine spiking, all of the activity released from the fuel up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident is assumed to be
released to the containment. Of the amounts of noble gases released to
the containment, 100 percent is assumed to be released to the
containment atmosphere. For the iodines, it is assumed that only DCPP UNITS 1 &
2 FSAR UPDATE 15.5-22 Revision 22 May 2015 10 percent of the iodines released to the containment are released to the containment atmosphere.
(5) The spray removal rates for the SBLOCA are assumed to be the same as those applicable for the large break LOCA as described in
Section 15.5.17.
(6) The containment leakage rates in this analysis are also assumed to be the same as for the large break LOCA and are discussed in Section 15.5.17.
The resulting potential exposures are listed in Table 15.5-10 and demonstrate that all calculated doses are well below the guideline values specified in 10 CFR 100.11. Since the activity releases from this type of event will be significantly lower than those from a large break LOCA, any control room exposure which might occur would be well within
the established criteria discussed in Section 15.5.17. In addition, because of significantly
lower fission product releases to the sump and the absence of any zirconium-water
reaction, the amounts of free hydrogen produced by sump radiolysis following a small
LOCA would not be of concern.
15.5.11.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are well within the dose limits of 10 CFR 100.11 as shown in Table 15.5-10.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are well within the dose limits of 10 CFR 100.11 as shown in Table 15.5-10.
(3) Since the activity releases from the SBLOCA are less than those from a large-break LOCA (LBLOCA), any control room dose which might occur would be well within the establish ed criteria discussed in Section 15.5.17.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-23 Revision 22 May 2015 15.5.12 RADIOLOGICAL CONSEQUENCES OF MI NOR SECONDARY SYSTEM PIPE BREAKS 15.5.12.1 Acceptance Criteria The radiological consequences of accidents analyze d in Section 15.3 such as minor secondary system pipe breaks shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
15.5.12.2 Identification of Causes and Accident Description The effects on the core of sudden depressu rization of the secondary system caused by an accidental opening of a steam dump, relief or safety valve were described in Section 15.2 and apply also to the case of minor secondary system pipe breaks. As shown in that analysis, no core damage or fuel rod failure is expected to occur. In
Section 15.5.18, analyses are presented that show the effects on the core of a major
steam line break, and, in this case also, no fuel rod failures are expected to occur.
The analyses presented in Section 15.2 demonstrate that a departure from nucleate
boiling ratio (DNBR) of less than the safety a nalysis limit will not occur anywhere in the
core in the event of a minor secondary system pipe rupture. The possible radiological consequences of this event, due to the release of some steam that might contain radioactive iodines, are discussed in Section 15.5.10. The resulting thyroid doses are presented parametrically in Figures 15.5-2 through 15.5-5 as a function of quantity of
steam released and secondary system activity. In the event that a complete plant
cooldown without condenser coo ling capacity is necessary following the break, the potential exposures would be the same as those reported in Table 15.5-9 for loss of
electrical load.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-24 Revision 22 May 2015 15.5.12.3 Conclusions On the basis of the discussed results, it can be concluded that the potential exposures following a minor secondary system pipe rupture would be insignificant.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
- The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release are insignificant as shown in Table 15.5-9.
- The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are insignificant as shown in Table 15.5-9.
15.5.13 RADIOLOGICAL CONSEQUENCES OF INADVERTENT LOADING OF A FUEL ASSEMBLY INTO AN IMPROPER POSITION 15.5.13.1 Acceptance Criteria Fuel assembly loading errors shall be prevented by administrative procedures implemented during core loading. In the unlikely event that a loading error occurs, analyses supporting Section 15.3.3 shall confirm that no events leading to radiological consequences shall occur as a result of loading errors.
15.5.13.2 Identification of Causes and Accident Description Fuel and core loading errors such as inadvertently loading one or more fuel assemblies
into improper positions, loading a fuel rod during manufacture with one or more pellets
of the wrong enrichment, or loading a full fuel assembly during manufacture with pellets
of the wrong enrichment will lead to increased heat fluxes if the error results in placing
fuel in core positions calling for fuel of lesser enrichment. The inadvertent loading of
one or more fuel assemblies requiring burnable poison rods into a new core without
burnable poison rods is also included among possible core loading errors. Because of
margins present, as discussed in detail in Section 15.3.3, no events leading to
radiological consequences are expected as a result of loading errors.
15.5.13.3 Conclusions Because of margins present, as discussed in detail in Section 15.3.3, no events leading to radiological consequences are expected as a result of loading errors.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-25 Revision 22 May 2015 15.5.14 RADIOLOGICAL CONSEQUENCES OF COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.5.14.1 Acceptance Criteria The radiological consequences of small amounts of radioactive isotopes that could be released to the atmosphere as a result of atmospheric steam dumping required for plant cooldown following a complete loss of forced reactor coolant flow shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
15.5.14.2 Identification of Causes and Accident Description As discussed in Section 15.3.4, a complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical s upplies to all reactor coolant pumps (RCPs). 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.
The analysis performed and reported in Section 15.3.4 has demonstrated that for the
complete loss of forced reactor coolant flow, the DNBR does not decrease below the
safety analysis limit during the transient, and thus there is no cladding damage or
release of fission products to the RCS. For this reason, this accident has no significant
radiological effects.
15.5.14.3 Conclusions The analysis described in Section 15.3.4 demonstrates that there are no significant environmental effects of the Complete Loss of Forced Reactor Coolant Flow event.
Therefore, the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release are insignificant.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-26 Revision 22 May 2015 (2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are insignificant.
15.5.15 RADIOLOGICAL CONSEQUENCES OF AN UNDERFREQUENCY ACCIDENT 15.5.15.1 Acceptance Criteria The radiological consequences of small amounts of radioactive isotopes that could be released to the atmosphere as a result of atmospheric steam dumping required for plant cooldown following an underfrequency accident shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
15.5.15.2 Identification of Causes and Accident Description A transient analysis for this unlikely event has been carried out. The analysis
demonstrates that for an underfrequency accident, the DNBR does not decrease below
the safety analysis limit during the transient, and thus there is no cladding damage or release of fission products to the RCS.
However, small amounts of radioactive isotopes could be released to the atmosphere as a result of atmospheric steam dumping
required for plant cooldown.
A detailed discussion of the potential radio logical consequences of accidents involving atmospheric steam dumping is presented in Section 15.5.10. From the parametric
analyses presented in that section, the potential exposures from an underfrequency
accident are given in Table 15.5-11. On the basis of these potential exposures, it can
be concluded that, although very unlikely, the occurrence of this accident would not
cause undue risk to the health and safety of the public.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-27 Revision 22 May 2015 15.5.15.3 Conclusions On the basis of the potential exposures discussed, it can be concluded that, although very unlikely, the occurrence of this accident woul d not cause undue risk to the health and safety of the public.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release are insignificant as shown in Table 15.5-11.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are insignificant as shown in Table 15.5-11.
15.5.16 RADIOLOGICAL CONSEQUENCES OF A SINGLE ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL AT FULL POWER 15.5.16.1 Acceptance Criteria The radiological consequences of a single rod cluster control assembly withdrawal shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
15.5.16.2 Identification of Causes and Accident Description A complete transient analysis of this accident is presented in Section 15.3.5. For the
condition of one rod cluster control assembly (RCCA) fully withdrawn with the rest of the
bank fully inserted, at full power, an upper bound of the number of fuel rods DCPP UNITS 1 &
2 FSAR UPDATE 15.5-28 Revision 22 May 2015 experiencing DNBR less than the safety analysis limit is 5 percent of the total fuel rods in the core.
A detailed discussion of the potential radio logical consequences of accidents involving small amounts of fuel rod failure is included in Section 15.5.21. From the parametric analyses presented in that section, the potential e xposures from an RCCA withdrawal at
full power resulting in 5 percent fuel failure are given in Table 15.5-12.
15.5.16.3 Conclusions On the basis of the potential exposures discussed, it can be concluded that the occurrence of this accident would not cause undue risk to the health and safety of the public. Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are insignificant as shown in Table 15.5-12.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are insignificant as shown in Table 15.5-12.
15.5.17 RADIOLOGICAL CONSEQUENCES OF MAJOR RUPTURE OF PRIMARY COOLANT PIPES Various aspects of the radiological consequences of a large-break loss-of-coolant-accident (LBLOCA) are presented in this section.
15.5.17.1 Acceptance Criteria (1) The radiological consequences of a major rupture of primary coolant pipes shall take into consideration fission product releases due to leakage from the containment, post-LOCA recirculation Loop leakag e in the Auxiliary Building (inclusive of a residual heat removal (RHR) pump seal failure resulting in a 50 gpm leak for 30 minutes starting at T=24 hrs post-LOCA), and containment shine. (2) The radiological consequences of a major rupture of primary coolant pipes shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-29 Revision 22 May 2015
- i. An individual located at any point on the boundary of the exclusion area for the two hours immediately foll owing the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(3) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident.
(4) In the event controlled venting of the containment is implemented post-LOCA using the containment hydrogen purge system (serves as a back-up capability for hydrogen control to the hydrogen recombiners), an individual located at any point on the boundary of the exclusion area, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 0.5 rem/year in accordance with 10 CFR Part 20.
15.5.17.2 Identification of Causes and Accident Description 15.5.17.2.1 Basic Events and Release Fractions The accidental rupture of a main coolant pipe is the event assumed to initiate a LBLOCA. Analyses of the response of the reactor system, including the emergency core cooling system (ECCS), to r uptures of various sizes have been presented in
Sections 15.3.1 and 15.4.1. As demonstrated in these analyses, the ECCS, using emergency power, is designed to keep cladding temperatures well below melting and to
limit zirconium-water reactions to an insignificant level. As a result of the increase in
cladding temperature and the rapid depressurization of the core, however, some cladding failure may occur in the hottest regions of the core. Following the cladding
failure, some activity would be released to the primary coolant and subsequently to the inside of the containment building. Because of the pressurization of the containment
building by the primary coolant water escaping from the pipe break, some of the volatile radioactive iodines and noble gases could le ak from the containment building to the atmosphere.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-30 Revision 22 May 2015 It is not expected that a significant amount of organic iodine would be liberated from the fuel as a result of a LBLOCA. This conclusion is based on the results of fuel meltdown experiments conducted by the Oak Ridge National Laboratory. The fraction of the total iodine that is released in organic forms is expected to be on the order of 0.2 percent, or less, since the rate of thermal radiolytic decomposition would exceed the rate of
production.
Organic compounds of iodine can be formed by reaction of absorbed elemental iodine
on surfaces of the containment vessel. Experiments have shown that the rate of
formation is dependent on specific conditions such as the concentration of iodine, concentration of impurities, radiation level, pressure, temperature, and relative humidity.
The rate of conversion of airborne iodine is proportional to the surface-to-volume ratio of
the enclosure, whether the process is limited to diffusion to the surface or by the
reaction rate of the absorbed iodine. The observed yields of organic iodine as a
function of aging time in various test enclosures, with various volume-to-surface area
ratios, were extrapolated to determine the values for the DCPP containment vessel.
The iodine conversion rates predicted in this manner did not exceed 0.0005 percent of the atmospheric iodine per hour.
The potential exposures follow ing the postulated sequence of events in LBLOCAs have been analyzed for two cases. In the expected case, it has been assumed that the entire inventory of volatile fission products contained in the pellet-cladding gap spaces is
released to the coolant during the time the core is being flooded by the ECCS. Of this
gap inventory, 25 percent of the iodines and 100 percent of the noble gases are considered to be released to the containment atmosphere immediately following the
pipe rupture. In this respect, the expected case does contain some degree of conservatism since the ECCS is designed to prevent gross cladding damage. In accordance with the experimental data reported in the previous paragraph, the fraction
of iodine that is released in organic form is assumed to be 0.2 percent, and the
production rate of organic forms is considered negligible. The iodine plateout rates are negligible (Reference 10) compared to the spray washout rates and are assumed to be zero. The particulate fraction of iodine is also assumed to be zero for the expected
case since this fraction is small and the spray removal rates for particulates is large as
shown in Reference 10.
For the design basis LOCA, it has been assumed that 25 percent of the equilibrium
radioactive iodine inventory in the core is immediately available for leakage from the
reactor containment. Ninety-one percent of this 25 percent is assumed to be in the form
of elemental iodine, 4 percent of this 25 percent is in the form of organic iodides, and
5 percent of this 25 percent is in the form of particulate iodine. In addition, 100 percent
of the noble gas inventory in the core is assumed to be immediately released to the
containment building. As discussed in earlier paragraphs, releases of these magnitudes
are not expected to occur, even if the ECCS does not perform as expected. An analysis
using these assumptions is presented because these values are considered acceptable
for a design basis analysis in Regulatory Guide 1.4, Revision 1.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-31 Revision 22 May 2015 15.5.17.2.2 Spray System Iodine Removal Rates The containment spray system (CSS) is described in detail, along with a performance
analysis, in Sections 6.2.2 and 6.2.3. The performance analysis includes the representation of the spatial distribution of droplets and iodine in the containment, as
well as drop coalescence and other effects.
For the expected case analyses, the CSS is assumed to function with both spray pumps
operating, giving an effective elemental iodine removal coefficient of 92 per hour. On
the basis of experiments at Battelle, as described in Reference 10, the spray removal
rate for organic iodides was assumed to be 0.058 per hour.
For the design basis case, it is assumed that one of the two spray pumps fails to
operate, and the elemental iodine removal coefficient is reduced to 31 per hour. This
assumption is consistent with the value of 32 per hour used in the PSAR analysis. It has also been assumed, for the design basis case, that the CSS has no effect on the organic and particulate iodines.
Although a subsequent safety evaluation showed that the Design Case coefficient of
31 per hour (for 2600 gpm spray header flow) should be reduced to approximately
29 per hour (for 2466 gpm spray header flow), the potential offsite dose increase due to
this change is extremely small and can be considered insignificant (Reference 39).
15.5.17.2.3 Offsite Exposures from Containment Leakage As a result of the pressurization of the containment following a LOCA, there is a possibility of containment leakage during the time that the containment pressure is
above atmospheric. For the design basis case, the leakage rate has been assumed to
be 0.1 percent 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 /> following the accident, and 0.05 percent per
day after the first day. These assumed rates are consistent with the Technical
Specifications (Reference 22) limit, the assumed rates considered acceptable in Regulatory Guide 1.4, Revision 1, and the values assumed in the PSAR analyses.
For the expected case, the containment leakage rates used are 0.05 percent per day for
the first day and 0.025 percent per day for t he periods after 1 day. These rates were determined from averages of the actual predicted containment pressures presented in
previous sections, with the assumption that s ome of the heat removal systems do not function at full capacity.
In this regard, the leakage rates assumed for the expected case analysis retain some
degree of conservatism since the containment heat removal systems are designed to
reduce the containment pressure to atmospheric following the initial pressure rise, thus
terminating the leakage.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-32 Revision 22 May 2015 15.5.17.2.4 Containment Leakage Exposure Sensitivity Study HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
Sensitivity studies were performed to illustrate the dependence of the thyroid exposures on the spray system removal constant and the fraction of nonremovable iodines present
in the containment. The results of these studies are shown in Figures 15.5-6, 15.5-7, and 15.5-8. The thyroid exposures, normalized to the exposure for zero spray removal
constant, are shown as a function of spray constant in Figure 15.5-6, for a fixed fraction
of nonremovable iodine forms of 15 percent. In Figures 15.5-7 and 15.5-8 thyroid
exposures are plotted as a function of two parameters: the spray removal constant and
the percent of nonremovable iodines. To determine an absolute exposure (rem) from Figure 15.5-7, the normalized exposure should be multiplied by 940.9 rem, which is the reference 2 hour-800 meter exposure for the design basis case with a zero spray
constant and a zero nonremovable fraction.
To determine an absolute exposure (rem) from Figure 15.5-8, the normalized exposure should be multiplied by 197.4 rem, which is the reference 30-day-10,000 meter exposure for the design basis case with a zero
spray constant and a zero nonremovable fraction. As shown in these figures, combinations of these parameters that result in normalized exposures below the
criterion line would result in a calculated absolute exposure less than the 300 rem
guideline level specified in 10 CFR Part 100.
15.5.17.2.5 Radiological Consequences with DF of 100
The design basis LOCA was reviewed to evaluate potential differences in the offsite
radiological dose consequences using a containment decontamination factor of 100 and
a containment mixing flowrate of 94,000 cfm.
A containment mixing rate of 94,000 cfm corresponds with our current minimum design
basis operation of two containment fan cooler units (CFCU). Calculations were based
on reload fuel. A containment spray delay of 80 seconds was used. The radionuclide
inventory source terms for the various fuel conditions were calculated using the
ORIGEN-2 computer code with a power level of 3580 MWt. The radionuclide atmospheric releases and offsite doses were calculated with the LOCADOSE computer code.
Calculations were made relative to 10 CFR 100.11 requirements for offsite doses, at the 800 meter exclusion area boundary (EAB) at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the 10,000 meter low population zone (LPZ) at 30 days, from post-LOCA containment leakage.
Table 15.5-75 presents the calculated offsite dose consequences from post-LOCA from
various pathways. The limiting dose is the thyroid at the EAB. For the containment leakage pathway, the maximum thyroid dose of 107.06 rem exceeds the original design basis LOCA thyroid dose of 95.9 rem in Table 15.5-23. For the pre-existing small
leakage, the EAB thyroid dose is 8.22 rem.
These doses are comparable with the corresponding original design basis LOCA large leakage and small leakage cases
doses.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-33 Revision 22 May 2015 All doses are within the 10 CFR 100.11 guidelines.
15.5.17.2.6 Offsite Exposures from Containment Shine The site boundary 30-day DBA exposure from direct containment gamma radiation (containment shine) is estimated to be 0.0048 rem. Containment shine is a function of
the activity present in the containment atmosphere. The EMERALD computer code was
used to calculate the post-accident containment activity time-history, and the ISOSHLD II computer code was then used to calculate the containment shine exposure.
The shine exposure model assumes a cylindrical radiation source having the same
radius and height as the containment structure with a 3.5-foot-thick concrete shield
surrounding it. The site boundary receptor point is assumed to be 800 meters from the
containment structure.
15.5.17.2.7 Offsite Population Exposur es from Containment Leakage The calculated population exposures for the design basis case assumptions, and for the expected case, are summarized in Table 15.5-23. These whole body population
exposures do not include the effects of any population redistribution due to evacuation.
These exposures were calculated using the EMERALD computer code. The
atmospheric dilution factors and population distribution utilized in the population
exposure calculations are discussed in Section 15.5.5.
15.5.17.2.8 Offsite Exposures from Post-LOCA Recirculation Loop Leakage in the Auxiliary Building Reactor coolant water that collects in the containment recirculation sump after a LOCA would contain radioactive fission products.
Because containment recirculation sump water is circulated outside the containment, problems of potential exposure due to
post-LOCA operation of external circulation loops with leakage have been evaluated.
Reactor coolant water, ECCS injection water, and containment spray water accumulate
in the containment recirculation sump following a LOCA. Containment recirculation
sump water is circulated by the RHR pumps, cooled via the RHR heat exchangers, returned to the containment via the RHR system piping and the CSS piping (if
recirculation spray is used), passed through the RCS and the containment spray
nozzles (if recirculation spray is used), and finally returned to the containment
recirculation sump. In the event of circulation loop leakage in the aux iliary building, post-LOCA activity has a pathway to the atmosphere.
An illustration of this pathway for a small leak is given in Figure 15.5-9. For the small leakage situation, fission products in the leakage water are exposed to auxili ary building ventilation air flow for a long period of time.
Thus, for the small leakage situation, all
activity released to the auxiliary building would be released to the aux iliary building air, i.e., no credit for liquid-gas partitioning.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-34 Revision 22 May 2015 An illustration of post-LOCA activity pathway for a large leak is given in Figure 15.5-10.
For the large leakage situation, fission products in the leakage water are exposed to
auxiliary building ventil ation air flow for a short period of time.
Thus, most of the activity released to the auxiliary build ing would be transferred to the floor drain receiver tank, i.e., credit for liquid-gas partitioning.
The complete RHR system and CSS descriptions; including estimates of leakage, detection of leakage, equipment isolation, and corrective maintenance, are contained in
Sections 5.5.6 and 6.2.2, respectively.
Post-LOCA auxiliary building lo op leakage exposures we re calculated for four different leakage cases:
(1) Expected small leakage case (2) Expected large leakage case (3) DBA small leakage case (4) DBA large leakage case
Assumptions and numerical values used to calculate loop leakage exposures are listed in Table 15.5-24. Table 15.5-63 shows the results of the calculations based on these assumptions. Because an insignificant amount of noble gases would be in the containment recirculation sump water, the whole body exposures are negligible.
One possible approach to the evaluatio n of offsite exposures from post-LOCA recirculation loop leakage would include the following assumptions:
(1) A LOCA as an initiating event (2) Failure of two ECCS trains resulting in gross fuel damage: Release of 50 percent of core iodine inventory and 100 percent of core noble gas
inventory to the containment (3) Failure of an RHR pump seal, resulting in the release of a significant amount of the above containment activity to the auxiliary building (4) Failure of the passive auxiliary bu ilding charcoal filters resulting in the unfiltered release of iodine fission products to the environment
The assumption of this sequence of failures for analysis of offsite exposures, however, would be requiring plant design features in excess of the current guides and regulations, and in particular the requirements of ANS Standard N18.2, Nuclear Safety Criteria for
the Design of Stationary Pressurized Water Power Plants. (See proposed addendum to
ANS Standard N18.2, Single Failure Criteria for Fluid Systems (Reference 16)).
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-35 Revision 22 May 2015 Applying the proposed standard to post-LOCA recirculation loop leakage the LBLOCA was assumed as the initiating event:
"The unit shall be designed to tolerate an initiating event which may be a single active or passive failure in any system intended for use during normal operation."
The ECCS was assumed to function properly, as required by the ECCS acceptance
criteria, preventing gross fuel damage. Although meeting these criteria is expected to
preclude gross cladding damage, it was assumed for this analysis that 100 percent of
the gap iodine and noble gas inv entories were released to the containment recirculation sump.
For the large leakage cases, failure of an RHR pump seal was assumed as the single
failure and can be tolerated without loss of the required functioning of the RHR system, as required by the following clauses in the proposed addendum to the ANS Standard N18.2:
"Fluid systems provided to mitigate the consequences of Condition III and Condition IV events shall be designed to tolerate a single failure in addition to the incident which requires their function, withou t loss of the function to the unit.
"A single failure is an occurrence which results in the loss of capability of a
component to perform its intended safety functions when called upon. Multiple failures resulting from a single occurrence are consider ed to be a single failure.
Fluid and electrical systems are considered to be designed against a single
failure if neither (a) a single fail ure of any active component (assuming passive components function properly); nor (b) a single failure of a passive component (assuming active components function properly) results in a loss of the safety
function to the nuclear steam electric generating unit.
"An active failure is a malfunction, excluding passive failures, of a component
which relies on mechanical movement to complete its intended function upon demand. "Examples of active failures include the failure of a valve or a check valve to
move to its correct position, or the failure of a pump, fan or diesel generator to
start. "Spurious action of a powered component originating within its actuation system
shall be regarded as an active failure unless specific design features or operating
restrictions preclude such spurious action.
"A passive failure is a breach of the fluid pressure boundary or blockage of a
process flowpath."
For the expected and DBA large leakage cases, the failure of auxiliary building charcoal filters, a second failure, was not assumed, in accordance with the standard.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-36 Revision 22 May 2015 For the expected and DBA small leakage cases, failure of auxiliary building charcoal
filters was assumed as the single failure and can be tolerated without loss of the
required function of the auxiliary building ventilation system, which provides cooling for ECCS components.
For the long-term small leakage cases, the charcoal filters are not needed to reduce
exposures below the guideline values given in 10 CFR Part 100. In any case, the fans in the ventilation system are redundant, and only the passive charcoal beds themselves
are not redundant.
For the expected small and large leakage cases, it was assumed that two ECCS trains, five fan coolers, and two containment spray trains functioned. For the DBA small and
large leakage cases, it is assumed that two ECC S trains, two fan coolers, and one containment spray train functioned. The DBA assumptions result in high containment
recirculation sump water temperatures and minimum containment recirculation sump
water pHs.
For all four circulation loop leakage cases it was assumed that 100 percent of the gap
iodine inventory was deposited in containment recirculation sump water.
For the expected small and large leakage cas es, the assumed gap iodine inventories are listed in Table 11.1-7. The expected case gap iodine was assumed to be only
elemental iodine. For the DBA small and large leakage cases, the assumed gap iodine
inventories are based on release fractions given in Safety Guide 25, March 1972 (Reference 23). The DBA case gap iodine was assumed to be 99.75 percent elemental iodine and 0.25 percent organic iodine per Safety Guide 25, March 1972.
Radiological decay of activity in the contain ment recirculation sump was assumed for all leakage cases for both the time periods before and during loop leakage. No credit was taken for cleanup of activity in the containment recirculation sump.
Reactor coolant water, accumulator water, and refueling water storage tank (RWST)
water make up the total volume of water in which activity is deposited. Consideration of
emergency core cooling injection flowrates and containment spray injection flowrates
yields the volume of RWST water (Chapter 6). Table 15.5-24 lists the assumed volume
of water in which activity is deposited for the four leakage cases. For the large leakage
cases, the volume of diluting water was taken as the volume when the leakage began.
No credit was taken for the extra diluting water added from the RWST during the
30-minute leakage period.
Sodium hydroxide spray additive wil l provide for an increased pH in the containment recirculation sump water. Consideration of emergency core cooling injection flowrates
and containment spray injection flowrates yields the pH of the containment recirculation
sump water (Chapter 6). Table 15.5-24 lists the assumed pH of recirculation loop
leakage water for the four leakage cases.
For the large leakage cases, the pH was DCPP UNITS 1 &
2 FSAR UPDATE 15.5-37 Revision 22 May 2015 taken as the pH when the leakage began. No credit was taken for the extra sodium hydroxide in the spray water added during the 30-minute leakage period.
The design evaluation conducted for the containment functional design yields the
temperature of containment recirculation sump water as a function of time (Chapter 6).
Table 15.5-24 lists the assumed temperature of recirculation loop leakage water for the
four leakage cases. For the large leakage cases, the water temperature was taken as
the temperature when the leakage began. No credit was taken for the decrease of
water temperature during the 30-minute leakage period.
A review of the equipment in the RHR system loop and the CSS loop indicates that the
largest leakage would result from the failure of an RHR pump seal. Evaluation of RHR
pump seal leakage rate, assuming only the presence of a seal retention ring around the
pump shaft, shows that flows less than 50 gpm would result (Chapter 6). Circulation
loop piping leaks, valve packing leaks, and flange gasket leaks are much smaller and less severe than an RHR pump seal failure leak. Leakage from these components
during normal post-LOCA operation of the RHR syste m loop and the CSS loop is estimated to be 1910 cc/hr (Chapter 6).
On this basis, a 50 gpm leakrate was assumed for both the expected large leakage case and the DBA large leakage case, and a 1910 cc/hr leakrate was assumed for both the expected small leakage case and the
DBA small leakage case.
For the DBA large leakage case, recirculation loop leakage was assumed to commence
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the start of the LBLOCA. This assumption is consistent with the discussion in Sections 3.1.1.1 and 6.3.3.5.3, and with the guidance in Standard Review Plan 15.6.5, Appendix B. In this context, the limiting recirculation loop long term passive failure is 50 gpm leakage at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the start of the LBLOCA.
Evaluation of an RHR pump seal failure shows that the failure could be detected and
the pump isolated well within 30 minutes (Chapter 6). A leakage duration of 30 minutes
is conservatively assumed for both the expected and DBA large leakage cases. A
leakage duration of 30 days is assumed for both the expected and DBA small leakage cases. As discussed earlier, the aux iliary building DF is a function of the Partition Factor (PF) for a particular isotope (Equation 15.5-7).
For both the expected and DBA large leakage cases, it was assumed that leakage
water was pumped away to the floor drain receiver tank. Iodine in the leakage water
was assumed to be exposed to auxiliary building ventilation air flow for a short period of time (0.05-0.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />), and thus, liquid-gas partitioning was assumed for elemental
iodine isotopes.
The large leakage case elemental iodi ne PFs were calculated using the previously presented PF expression. Because the circulation water will be above 212
°F (Chapter 6), a flashing process must be considered. For heat energy conservation on
the basis of 1 lb:
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-38 Revision 22 May 2015 xhx)(1h=h gff0+ (15.5-11)
Rearranging yields fgff0h-hh -h x= (15.5-12) where:
h f0 = initial enthalpy of liquid, Btu/lbm h f = final enthalpy of liquid, Btu/lbm h g = final enthalpy of vapor, Btu/lbm x = fraction of initial mass that became vapor
The end point of the flashing process is 212
°F, and thus the final enthalpies are based on this temperature. The mass fraction, x, is the ratio of the final mass of vapor to the total initial mass of water, so the mass ratio at the end of the flashing process becomes:
x1 x M M liquid vapor= (15.5-13)
Figures 15.5-11 and 15.5-12 present the expected and DBA large leakage case
elemental iodine PFs as a function of both temperature and pH. For small PFs, the DF (see Equation 15.5-7) is approximately equal to the reciprocal of the PF.
Figures 15.5-11 and 15.5-12 illustrate that auxi liary building iodine PFs and resulting DFs are relatively insensitive to water temperature, but much more sensitive to pH.
Table 15.5-24 lists the assumed temperatures and pHs along with the resulting
elemental iodine PFs and auxi liary building decontamination factors for both the expected and DBA large leakage cases.
For both the expected and DBA small leakage cases, it was assumed that leakage
water was not pumped away. Elemental iodine in the leakage water was assumed to
be exposed to auxiliary building ventilation air flow for a long period of time (100-150 hours), and thus, liquid-gas partitioning for elemental iodine isotopes was not
assumed. For the small leakage case all e lemental iodine activity released to the auxiliary building was assumed to be released to the auxiliary building atmosphere, i.e., a DF of 1.
Liquid-gas partitioning for organic iodine isotopes was not assumed for any of the four leakage cases. All organic iodine activity released to the auxiliary building was assumed to be released to the auxiliary building atmosphere, i.e., a decontamination
factor of 1.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-39 Revision 22 May 2015 For all four loop leakage cases, no credit was taken for auxiliary building radiological decay or fission product plateout.
For the expected and DBA large cases, credit for auxiliary building charcoal filters was
taken, and for the expected and DBA small leakage cases, no credit for auxiliary
building charcoal filters was taken (as previously discussed with reference to ANS
Standard N18.2 single failure criteria). Table 15.5-24 lists the assumed iodine filter
efficiencies for each loop leakage case.
From the calculated DBA case offsite expo sures from post-LOCA recirculation loop leakage in the auxiliary building listed in Table 15.5-63, it can be concluded that any exposures that occur via this combination of unlikely events would be well below the guideline levels in 10 CFR Part 100. In addition, even if no consideration is given to the effectiveness of the auxiliary building charc oal filters for the DBA leakage cases, the calculated exposures would still be belo w guideline levels specified in 10 CFR Part 100.
15.5.17.2.8.1 Maximum Allowable Leakage From Post-LOCA Recirculation Loop Calculations have been performed to determine the maximum allowable leakage from recirculation loop components that could occur during post-LOCA recirculation
operations before offsite and control room operator design basis radiation doses would
exceed regulatory limits. A computer code (LOCADOSE) was used to determine design
basis EAB and low population zone outer boundary (LPZ) offsite radiation doses and control room operator airborne radiation dose from post-LOCA containment leakage, RHR pump seal leakage, and pre-existing leakage from recirculation loop components outside containment. The calculations determined the amount of pre-existing
recirculation leakage which could exist befo re offsite exposures would exceed 10 CFR 100.11 limits or control room operator exposures would exceed GDC 19, 1971 limits, if a LOCA were to simultaneously occur.
Table 15.5-63 shows the results of the calculations based on the above assumptions
which determined that the maximum allowable leakage (in addition to the RHR pump seal leakage) from the recirculation loop at post-LOCA conditions of pressure and temperature was 1.85 gpm where the airborne activity is filtered by charcoal filters or
0.186 gpm where the airborne activity is unfiltered. The limitation is the GDC 19, 1971 allowable dose for the control room.
15.5.17.2.9 Offsite Exposures from Controlled Post-accident Containment Venting Because of the potential release of significant amounts of hydrogen to the containment
atmosphere following a LBLOCA, it is necessary to provide means of monitoring and controlling the post-accident concentration of hydrogen in the containment atmosphere.
Redundant thermal hydrogen recombiners are the primary means of post-accident hydrogen control. As a backup, controlled containment venting (via the containment
hydrogen purge system) with offshore flow, wind directions from northwest through
east-southeast measured clockwise, provides hydrogen control with a high probability of DCPP UNITS 1 &
2 FSAR UPDATE 15.5-40 Revision 22 May 2015 no inland exposures. As shown in Table 15.5-26, offshore wind directions occur over 50 percent of the time regardless of the season and, as shown in Table 15.5-27, have a high degree of persistence. The large time period (312 hours0.00361 days <br />0.0867 hours <br />5.15873e-4 weeks <br />1.18716e-4 months <br /> for DBA case) between
the proposed hydrogen venting level (3.5 v/o) and the hydrogen flammability level
(4.0 v/o) is much greater than the longest recorded period (37 consecutive hours) of onshore winds in any 22.5
° sector. These data ensure a very high probability that venting can be carried out during the occurrence of offshore winds.
Even though there is a high probability that containment venting can be carried out
when the wind is blowing offshore, if necessary at all, an evaluation is presented in the
following paragraphs to determine potential exposures if venting were carried out during
onshore winds.
Section 6.2.5 contains the analysis of post-accident hydrogen production and accumulation in the containment atmosphere and its control. Containment venting is
also described in Section 6.2.5. The purge stream is withdrawn from the containment
through one of two penetration lines. The stream is routed through a flow-measuring
device, charcoal filters, exhaust fans, the radiation monitors, and finally to the plant
vent. Post-accident containment venting activity releases are calculated with the following equation:
dtt)I(e)I(ACxVENRAT VOLUME (I)]60 0.01FILEFF
[1.0 ACT(I))2(T)1(T= (15.5.14) where:
ACT(I) = activity of isotope I released to the atmosphere, Ci AC(I) = activity of isotope I released to the containment atmosphere, Ci VOLUME = volume of containment atmosphere, cu ft VENRAT = venting rate, cfm (I) = removal constant for isotope I, hr
-1 FILEFF(I) = filter efficiency for isotope I, % T(1) = time after LOCA that containment venting begins, hr T(2) = time after LOCA that containment venting ends, hr t = time, hr 60 = minutes per hr
The above equation considers radiological d ecay during the time period prior to containment venting and the time period during containment venting. It also assumes
that the LBLOCA activity released to the containment atmosphere is homogeneously dispersed throughout the containment atmospheric volume. Exposures from activity
released to the atmosphere were calculated using the EMERALD computer code. EMERALD assumes there is no radiological decay during the atmospheric dispersion.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-41 Revision 22 May 2015 Containment venting exposures were calculated for both the expected case and the DBA case. Assumptions and numerical values used to calculate venting exposures are
itemized in Table 15.5-28. Onshore controlled containment venting thyroid and whole
body exposures are listed in Table 15.5-29.
Post-accident containment venting schedules are evaluated in Section 6.2.5. Assuming the venting system will operate an average 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per day, the system flowrates during
short venting periods are 120 cfm (expected) and 300 cfm (DBA). Equivalent
continuous venting rates, 10 cfm and 25 cfm, were used to calculate venting activity
releases.
In the event containment venting should be required during periods with onshore flow, the venting would be limited to those periods when Pasquill Stability Category D exists.
Therefore, ground-level centerline atmospheric dispers ion factors for Pasquill Stability Category D and an elevated release height of 70 meters were evaluated using a conventional Gaussian plume model and are listed in Table 15.5-30. The meteorological input parameters utilized were determined from onsite measurements, given in References 18, 19, and 20. Because an individual is assumed to be located on
the plume centerline for the entire venting duration, exposures are centerline exposures
and represent worst case conditions. The probability of an individual being located on
the plume centerline for a 2-hour period is very small, and thus centerline exposures
listed in Table 15.5-29 are very conservative.
During the time period prior to venting, activity released to the containment atmosphere
is significantly reduced by both radiologic al decay and functioning of the safety features systems. The main contributors of radioactivity several hundred hours after the accident are the noble gases: Kr-85, Xe-133, and, to some extent, Xe-131m. Because Kr-85 has a half-life of 10.6 years, the exposures resulting from containment venting
would not be significantly reduced if the venting could be further delayed for many
months.
It can be concluded from the results presented in Table 15.5-29, along with the
consideration of the very high probability of opportunities for offshore venting and the
other favorable factors associated with the DCPP design and site, that, as a backup to
the internal hydrogen recombiner system, controlled venting using the containment
hydrogen purge system is an acceptable contingency method of post-accident hydrogen control for this plant. In addition, it can be concluded that the expected exposures due
to venting, even using the assumptions in Safety Guide 7, will not exceed the annual dose limits of 10 CFR Part 20.
15.5.17.2.10 Post-accident Control Room Exposures The design basis for control room ventilation, shielding, and administration is to permit
access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any DCPP UNITS 1 &
2 FSAR UPDATE 15.5-42 Revision 22 May 2015 part of the body, for the duration of the most severe design basis accident. This basis is consistent with GDC 19, 1971.
The control room shielding, described in Section 12.1 is designed to attenuate gamma radiation from post-accident sources to levels consistent with the requirements of GDC 19, 1971.
The control room ventilation system is described in Section 9.4.1. It is designed to limit the concentration of post-accident activity in the control room air to levels consistent with requirements of GDC 19, 1971.
The control room post-accident administration is described in the DCPP Manual. It is to limit post-accident control room personnel exposures to levels consistent with requirements of GDC 19, 1971.
Exposures to control room personnel have been estimated for a design basis LOCA to
evaluate the adequacy of the control room shiel ding, the adequacy of the control room ventilation system, and the adequacy of the control room administration in limiting
exposures to the specified limits. Exposures have also been calculated for the expected case LBLOCA to obtain a more realistic estimate of exposure to control room personnel.
Radiation exposures to personnel in the control room could result from the following sources:
(1) Airborne activity, which infiltrates into the control room (2) Direct gamma radiation to the co ntrol room from activity in the containment structure (3) Direct gamma radiation to the co ntrol room from activity in the containment leakage plume.
The control room ventilation system is designed to minimize infiltration of post-accident airborne activity into the control room complex. Mode 4 operation of the ventilation
system provides zone isolation with filtered positive pressurization and filtered
recirculation. Mode 4 operation of the ventilation system is initiated automatically and
the least contaminated positive pressurization inlet is selected manually as described in
Chapter 9.4.1. Both the pressurization and partial recirculation air flow pass through high-efficiency particulate air (HEPA) and charcoal filters.
In addition to positive pressurization, there are vestibules on control room doors that will
minimize infiltration. Table 15.5-31 identifies infiltration pathways and flowrates that have been used in the calculation of post-accident control room radiological exposures.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-43 Revision 22 May 2015 Airborne radiation doses inside the control room were evaluated for a DBA LOCA.
Regulatory Guide 1.4, Revision 1 was used to determine activity levels in the containment. Activity releases are based on a containment leakage of 0.1 percent/day
for the first day and 0.05 percent/day thereafter.
The containment leakage was assumed to be released unfiltered from the containment
building to the atmosphere. Recirculation loop leaka ges, assumed to be from an RHR pump seal, will pass through charcoal filters and be released to the atmosphere through
the main vent at the top of the containment.
Radioactivity from the atmosphere would ent er the control room through two pathways:
(1) via the pressurization air intakes through charcoal filters (2) via infiltration of air inleakage
The flow rate of pressurization air into the control room is 2100 cfm. The flow rate of
recirculated control room air through the charcoal filters is 2100 cfm. Previous analyses
had not taken credit for recirculation of control room air. This was an unnecessary
conservatism in that a passive failure had already been assumed to occur (RHR pump
seal leak) and a second failure is not required.
A 10 CFM inleakage rate per Standard Review Plan, Section 6.4, was conservatively
assumed in the analysis due to the possible pathway through the single doors from the
equipment condensing unit areas to the HVAC equipment room. Additionally, an assumed 10-second delay in closure of the CRV S outside air isolation dampers results in 2110 cfm of control room infiltration for the first 10 seconds following the design basis
LOCA.
Table 15.5-32 presents a summary of the parameters used in the analysis.
The control room shielding is designed to minimize direct gamma radiation (containment shine). Control room exposures resulting from containment shine were estimated using
ISOSHLD II. The control room receptor poin t is 27 feet from the containment structure and protected by an additional 2.5-foot-thick concrete shield. A further contribution to
control room direct gamma radiation results from the atmospheric activity cloud external
to the control room. Control room exposures resulting from plume shine were estimated using ISOSHLD II. The shine exp osure model assumes a parallelepiped radiation source located directly above the control room. The control room receptor point is
protected by a 1.5-foot-thick concrete shield.
Radiation exposures to personnel during egr ess and ingress could result from the following sources:
(1) Airborne activity in the containment leakage plume
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-44 Revision 22 May 2015 (2) Direct gamma radiation from fission products in the containment structure Post-accident egress-ingress exposures are based on 27 outbound excursions, from the control room to the site boundary, and 26 inbound excursions, from the site
boundary to the control room. It was estimated that each excursion would take 5 minutes, and no credit was taken for breathing apparatus or special whole body
shielding.
Egress-ingress thyroid and whole body exposures from airborne activity are functions of
containment activity, containment leakage, atmospheric dispersion, and excursion time.
The EMERALD computer code was used to calculate the airborne activity concentrations, and then conventional exposure equations were used to calculate
gamma, beta, and thyroid exposures (Reference 6).
The exposure from betas is
calculated on the basis of an infinite uniform cloud, and exposure from gammas is
calculated on the basis of a semi-infinite cloud.
Because of the containment shielding and short excursion time, egress-ingress
containment shine exposures are small. Egress-ingress containment shine exposures
were calculated using ISOSHLD-II. The shine model assumes a cylindrical radiation
source having the same radius and height as the containment structure with a
3.5-foot-thick concrete shield surrounding it. The receptor point is assumed to be a
distance of 10 meters.
Estimates of post-accident control room exposures and egress-ingress exposures are listed in Table 15.5-33. The sum of the DBA case exposures are within the specified criteria, and the expected case exposures demonstrate the conservatism of the DBA case exposures.
15.5.17.2.11 Summary In the preceding sections, the potential expo sures from a major primary system pipe rupture have been calculated for various possible mechanisms:
(1) Containment leakage (2) RHR recirculation loop leakage (3) Controlled post-accident containment venting (4) Containment shine
The analyses have been carried out using the models and assumptions specified in
regulations 10 CFR Part 100, 10 CFR Part 50, and the safety and regulatory guides. In all analyses, the resulting potential exposur es to plant personnel, to individual members of the public, and to the general population have been found to be lower than the DCPP UNITS 1 &
2 FSAR UPDATE 15.5-45 Revision 22 May 2015 applicable guidelines and limits specified in 10 CFR Part 100, 10 CFR Part 50, and 10 CFR Part 20.
15.5.17.3 Conclusions Based on the results discussed, the occurrence of a major pipe rupture in the primary system of a DCPP unit would not constitute an undue risk to the health and safety of the public. In addition, the ESF provided for the mitigation of the consequences of a LBLOCA are adequately designed.
Finally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiological consequences of a major rupture of primary coolant pipes shall take into consideration fission product releases due to leakage from the containment, post-LOCA recirculation loop l eakage in the Auxiliary Building (inclusive of a RHR pump seal failure resulting in a 50 gpm leak for 30 minutes starting at T=24 hrs post-LOCA), and containment shine as shown in Section 15.5.17.2.11.
(2) The radiological consequences of a major rupture of primary coolant pipes shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
- i. An individual located at any point on the boundary of the exclusion area for the two hours immediately foll owing the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure as shown by the EAB whole body dose reported for containment shine in Section 15.5.17.2.6, and the remaining doses presented in Table 15.5-75.
ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure as shown by the EAB whole body dose reported for containment shine in Section 15.5.17.2.6 (conservative when applied to the LPZ), and the remaining doses presented in Table 15.5-75.
(3) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident as shown in Table 15.5-33.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-46 Revision 22 May 2015 (4) In the event controlled venting of the containment is implemented post-LOCA using the containment hydrogen purge system (serves as a back-up capability for hydrogen control to the hydrogen recombiners), an individual located at any point on the boundary of the exclusion area, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of the annual dose limit of 10 CFR Part 20 as shown in Table 15.5-29.
15.5.18 RADIOLOGICAL CONSEQUENCES OF A MAJOR STEAM PIPE RUPTURE 15.5.18.1 Acceptance Criteria (1) The radiological consequences of a major steam pipe rupture shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
- i. An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case. ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case.
(2) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident for both the pre-accident and the accident- initiated iodine spike cases.
15.5.18.2 Identification of Causes and Accident Description As reported in Section 15.4.2, a major steam line rupture is not expected to cause
cladding damage, and thus no release of fission products to the coolant is expected following this accident. If significant radioactivity exists in the secondary system prior to
the accident, however, some of this activity will be released to the environment with the
steam escaping from the pipe rupture. In addition, if an atmospheric steam dump from the unaffected steam generators is necessitated by unavailability of condenser capacity, DCPP UNITS 1 &
2 FSAR UPDATE 15.5-47 Revision 22 May 2015 additional activity will be released. Section 15.5.18.2.1 discusses the main steam line break (MSLB) dose analysis of record which is based on the OSGs. The OSG MSLB dose analysis is bounding for the RSGs as discussed in the following section. (See
Table 6.4.2-1 of Reference 49 for a summary of OSG and RSG MSLB steam releases.)
15.5.18.2.1 Radiological Assessment for Accident-Induced Leakage Because tubes in the faulted steam generator encounter a higher differential pressure during steam line rupture conditions than normal operating conditions, there is a
potential for primary-to-secondary leakage in degraded tubing to increase to a rate that
is higher than that during normal operation. This leakage is referred to as accident-
induced leakage. This section provides the updated licensing basis description and
radiological consequence analysis for a major steam line rupture analysis using an
accident-induced leak rate of 10.5 gpm (at room temperature conditions), which is higher than the operational leakage limit in the Technical Specifications. The NRC approved this analysis in a letter to PG&E dated February 20, 2003, Issuance of
Amendment: RE: Revision to Technica l Specification 1.1, Definitions, Dose Equivalent I-131, and Revised Steam Generator Tube Rupture and Main Steam Line Break
Analyses. Application of this accident-induced leak rate is governed by SG Program
accident-induced leakage performance criteria documented in the Technical
Specifications.
The methodology selected for performing the radiological assessment follows NRC
SRP 15.1.5, Steam System Piping Failures Inside and Outside of Containment (PWR),
Revision 2, 1981. Using an accident-induced leak rate of 10.5 gpm (at room
temperature conditions) in the faulted SG, calculations using the LOCADOSE computer program demonstrate that the offsite doses are within 10 percent of 10 CFR 100.11 limits and control room doses are within GDC 19, 1971, limits.
The resultant doses from the MSLB event using an accident-induced leak rate of
10.5 gpm are listed below. The limiting case is the accident initiated iodine spike as the
thyroid dose at the EAB is at the 30 rem limit.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-48 Revision 22 May 2015 Dose (rem) Location Thyroid CDE Beta Skin SDE Whole Body DDE Case 1: Accident-Initiated Spike EAB (0-2 hr) 30.0 1.50E-1 9.40E-2 LPZ (30 days) 6.49 1.92E-2 1.18E-2
- Dose Limit (10% of 10 CFR 100.11) 30.0 2.5 2.5 Control Room (30 days) 6.68E-1 7.10E-3 1.50E-4
- Dose Limit (GDC 19) 30 5 5 Case 2: Pre-Existing Spike EAB (0-2 hr) 46.4 1.37E-1 8.26E-2 LPZ (30 days) 3.69 9.70E-3 5.72E-3
- Dose Limit (10 CFR 100.11) 300 25 25 Control Room (30 days) 4.61E-1 5.56E-3 1.10E-4
- Dose Limit (GDC 19) 30 5 5 The input parameters for the dose analysis are summarized below.
(1) The operational (pre-MSLB) primary-to-secondary leak rate was assumed to be 1 gpm to yield a conservatively high isotopic concentration in the
secondary system. Use of 1 gpm is more conservative than the Technical Specifications operational leak rate limit of 150 gpd per SG.
(2) During the accident, the primary-to-secondary leak rate in the faulted steam generator is assumed at the maximum rate of 10.5 gpm. The primary-to-secondary leak rate in each intact SG was assumed to be at
the Technical Specifications operational l eak rate limit of 150 gpd; therefore, the total leakage is 450 gpd.
(3) The MSLB occurred in the section of piping between the containment building and the main steam line isolation v alves (MSIVs). Prior to control room isolation and pressurization, the control HVAC intake /Q is the unfiltered /Q utilized for the LOCA dose consequences analysis for control room inleakage.
(4) Loss of offsite power is assumed to occur coincident with MSLB accident.
(5) Conservatively, based on the Technical Specifications requirements for the safety injection signal and containment Phase A isolation, the control
room will be isolated well within 35 seconds. To add more conservatism
in this calculation, the control room is assumed to be isolated in 2 minutes.
(6) All releases were assumed to end after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, when the plant is placed on the RHR system.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-49 Revision 22 May 2015 (7) For a pre-existing iodine spike, the activity in the reactor coolant is based upon an iodine spike that has raised the reactor coolant concentration to 60 µCi/g of I-131 DEC, based on the Technical Specifications. The secondary coolant activity is 0.1
µCi/g of I-131 DEC, based on the Technical Specifications. Noble gas activity is based on 651
µCi/g of Xe-133 DEC associated with 1 percent failed fuel, which bounds the value in the Technical Specifications. The calculation of Xe-133 DEC ignores the
contribution from Kr-83m, Kr-85, Kr-89, Xe-131m, and Xe-137 due to low
concentration, short half life, or small dose conversion factor.
(8) For an accident-initiated (concurrent) iodine spike, the accident initiates an iodine spike in the reactor coolant system (RCS) that increases the iodine
release rate from the fuel to a value 500 times greater than the release rate corresponding to an RCS concentration of 1
µCi/g of I-131 DEC. The 1 µCi/g I-131 DEC is based on the Technical Specifications. The iodine activity released to the RCS for the duration of the accident is conservatively assumed to mix instantaneously and uniformly in the RCS.
Noble gas activity is based on 651
µCi/g of Xe-133 DEC associated with 1 percent failed fuel, which bounds the value in the Technical Specifications. To maximize the accident-initiated iodine spiking, a RCS letdown rate of 143 gpm with 100 percent iodine removal through the
filters in the demineralizers is assumed.
(9) The thyroid dose conversion factors are based on ICRP Publication 30 (Reference 21) as documented in Federal Guidance Report (FGR) 11 and
FGR 12 (References 41 and 42). The noble gas whole body dose
conversion factors are based on those documented in FGR 12, Table III.1.
The following table summarizes these conversion factors.
Isotope Dose Conversion Factor I-131 1.08E+06 (Rem/Ci)
I-132 6.44E+03 (Rem/Ci} I-133 1.80E+05 (Rem/Ci} I-134 1.07E+03 (Rem/Ci} I-135 3.13E+04 (Rem/Ci} Kr-85m 7.48E-15 (sv m 3/bq s) Kr-87 4.12E-14 (sv m 3/bq s) Kr-88 1.02E-13 (sv m 3/bq s) Xe-133m 1.37E-15 (sv m 3/bq s) Xe-133 1.56E-15 (sv m 3/bq s) Xe-135m 2.04E-14 (sv m 3/bq s) Xe-135 1.19E-14 (sv m 3/bq s) Xe-138 5.77E-14 (sv m 3/bq s)
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-50 Revision 22 May 2015 (10) Following the pipe rupture, auxilia ry feedwater to the faulted loop is isolated and the SG is allowed to steam dry. The iodine partition factor for the faulted SG is assumed to be 1.0. Also, the iodine partition factor for the intact SG is conservatively assumed to be 1.0; i.e., no credit is taken for iodine partition.
(11) All activity in the SGs is released to the atmosphere in accordance with the release rates in Table 15.5-34, with added releases from primary-to-
secondary leaks in the faulted loop and intact loops.
Atmospheric steam releases (not including primary-to-secondary leaks):
Ruptured loop 162,784 lb at 45.0 lb/ft 3 (0-2 hr) 0 lb (2-8 hr)
Intact loops 393,464 lb at 45.0 lb/ft 3 (0-2hr) 915,000 lb at 50.0 lb/ft 3 (2-8 hr)
The above steam releases are for the OSG MSLB. The RSG MSLB steam
releases are shown in Table 15.5-34.
As noted above, the limiting dose for MSLB is the EAB thyroid dose for the
accident initiated iodine spike case and is based on steam releases in the
first two hours of the accident. The OSG MSLB dose calculation assumes an accident-induced SG tube leak rate of 10.5 gpm using the Alternate
Repair Criteria (ARC) methodology. The RSGs can not credit ARC and
are required to maintain a much lower assumed SG tube leakage
subsequent to a MSLB. Note that although the zero to two hour RSG
ruptured loop release of 171,100 Ib is slightly greater than the equivalent
OSG release of 162,784 Ib, the OSG MSLB dose analysis bounds the
RSG MSLB releases since the assumed ARC tube leakage impact on
dose is the dominant factor in the assessment of post-accident
radiological consequences.
(12) The source term is based on a composite source term of 3.5 percent and 4.5 percent fuel enrichment. An evaluation has been performed and
concluded that the current source term bounds the 5 percent enrichment
fuel up to 50,000 MWD/MTU for a 21-month operating cycle.
(13) Atmospheric Dispersion Factors (sec/m3)
(Reference Tables 15.5-3 and 15.5-6)
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-51 Revision 22 May 2015 Time EAB LPZ Control Room Pressurized Infiltration 0-2 hr 5.29E-4 2.20E-5 7.05E-5 1.96E-4 2-8 hr 2.20E-5 7.05E-5 1.96E-4 8-24 hr 4.75E-6 5.38E-5 1.49E-4 24-96 hr 1.54E-6 3.91E-5 1.08E-4 96-720 hr 3.40E-7 2.27E-5 6.29E-5 (14) Control Room HVAC Flow Rates and Filtration Efficiencies:
Filtered Intake Flow 2100 cfm Unfiltered Intake Flow 10 cfm (2110 cfm for t=0 to 10 sec.)
Exhaust Flow 2110 cfm Filtered Recirculation Flow 2100 cfm Charcoal Filter Iodine Removal Efficiency Elemental 95 percent Organic 95 percent Particulate 95 percent (15) RCS and Secondary Water Volume and Water Mass RCS water volume 94,000 gallons RCS water mass 566,000 pounds Water in SGs 6735.54 ft 3 at 45.0 lb/ft 3 (0-2 hr) and 50.0 lb/ft 3 (2-8 hr) Loop 1 1683.88 ft 3 Loops 2, 3, 4 5051.65 ft 3 Water in Condensers 27243.59 ft 3 at 62.4 lb/ft 3 Water in SGs and Condensers 33979.13 ft 3 15.5.18.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroi d for the pre-existing iodine spike case and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case as shown in Section 15.5.18.2.1.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not DCPP UNITS 1 &
2 FSAR UPDATE 15.5-52 Revision 22 May 2015 receive a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case as shown in Section 15.5.18.2.1.
(3) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident for both the pre-existing and the accident-initiated iodine spike cases as shown in Section 15.5.18.2.1.
As noted in Section 15.5.18.2.1, the above dose estimates reflect the OSGs and an accident induced leak rate of 10.5 gpm. The limiting case is a thyroid dose at the EAB which corresponds to the dose limit of 30 rem for an accident-initiated iodine spike.
These dose estimates bound the doses with the RSGs which cannot credit Alternate Repair Criteria (ARC) for the steam generator tubes as the OSGs do.
15.5.19 RADIOLOGICAL CONSEQUENCES OF A MAJOR RUPTURE OF A MAIN FEEDWATER PIPE 15.5.19.1 Acceptance Criteria The radiological consequences of a major rupture of a main feedwater pipe shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure 15.5.19.2 Identification of Causes and Accident Description As reported in Section 15.4.2, a major feedwater line rupture is not expected to cause
cladding damage, and thus no release of fission products to the coolant is expected following this accident. If significant radioactivity exists in the secondary system prior to
the accident, however, some of this activity will be released to the environment with the
feedwater escaping from the pipe rupture. In addition, if an atmospheric steam dump DCPP UNITS 1 &
2 FSAR UPDATE 15.5-53 Revision 22 May 2015 from the unaffected steam generators is necessitated by unavailability of condenser capacity, additional activity will be released.
As discussed in Section 15.5.18, about 1.47E+06 lbm of secondary coolant is the limiting Condition IV event release expected for a full cooldown without any condenser availability.
The radiological consequences of about 1.47E+06 lbm of secondary coolant release have been discussed in Section 15.5.18.
15.5.19.3 Conclusions Based on the results discussed, it can be concluded that potential exposures from major feedwater line ruptures will be well below the guideline levels specified in 10 CFR 100.11, and that the occurrence of such ruptures would not result in undue risk to the public.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are insignificant as shown in Table 15.5-9.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are insignificant as shown in Table 15.5-9.
15.5.20 RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE (SGTR) 15.5.20.1 Acceptance Criteria (1) The radiological consequences of a steam generator tube rupture shall not exceed the dose guidelines of SRP, Section 15.6.3, Revision 2, as outlined below i. An individual located at any point on the boundary of the exclusion area for the two hours immediately foll owing the onset of the postulated fission product release shall not receive a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case, and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-54 Revision 22 May 2015 ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case, and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case.
(2) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident for both the pre-accident and the accident- initiated iodine spike cases.
15.5.20.2 Identification of Causes and Accident Description The SGTR accident is reanalyzed for RSGs and is discussed in Section 15.4.3, and the
thermal and hydraulic analysis presented in Section 15.4.3.3 provides the basis for the
evaluation of radiological consequences discussed in this section.
15.5.20.2.1 Offsite Exposures The evaluation of the radiological consequ ences of a steam generator tube rupture
event assumes that the reactor has been operating at the maximum allowable Technical
Specification (Reference 22) limits for primary coolant activity and 1 gpm primary to
secondary leakage for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant. Radionuclides from the primary coolant enter the steam generator via the ruptured tube and primary to secondary leakage, and are released to the atmosphere through the steam generator
PORVs (and safety valves) and via the condenser air ejector exhaust and/or the vacuum pump exhaust (if in operation).
The quantity of radioactivity released to the environment, due to an SGTR, depends
upon primary and secondary coolant activity, iodine spiking effects, primary to
secondary break flow flashing fractions, attenuation of iodine carried by the flashed
portion of the break flow, partitioning of iodine between the liquid and steam phases, the
mass of fluid released from the generator, and liquid-vapor partitioning in the turbine
condenser hot well.
(1) Design Basis Analytica l Assumptions The major assumptions and parameters used in the analysis are itemized
in Table 15.5-64.
(2) Source Term Calculations
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-55 Revision 22 May 2015 The radionuclide concentrations in the primary and secondary system, prior to and following the SGTR are determined as follows:
(a) The iodine concentrations in the reactor coolant will be based upon pre-accident and accident initiated iodine spikes.
(i) Accident Initiated Spike - The initial primary coolant iodine concentration is 1
µCi/gm of Dose Equivalent (DE) I-131.
Following the primary system depressurization associated with the SGTR, an iodine spike is initiated in the primary system
which increases the iodine release rate from the fuel to the
coolant to a value 335 times greater than the release rate
corresponding to the initial primary system iodine
concentration. The initial appearance rate can be written as
follows: P i = A i i (15.5-15) where: P i = Equilibrium appearance rate for iodine nuclide i A i = equilibrium RCS inventory of iodine nuclide i corresponding to 1
µCi/gm of DE I-131 i = removal coefficient for iodine nuclide i (j) Pre-accident Spike - A reactor transient has occurred prior to the SGTR and has raised the primary coolant iodine concentration from 1 to 60
(b) The initial secondary coolant iodine concentration is 0.1
(c) The chemical form of iodine in the primary and secondary coolant is assumed to be elemental.
(d) The initial noble gas concentrations in the reactor coolant are based upon 651
µCi/g of Xe-133 DEC for the noble gasses Kr-85m, Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135m, Xe-135, and Xe-138, using noble gas whole body dose conversion factors documented
in FGR 12 (Reference 42) Table III.1, associated with 1 percent fuel
defects. The calculation of Xe-133 DEC ignores the contribution
from Kr-85 and Xe-131m due to low concentration and small dose
conversion factor.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-56 Revision 22 May 2015 (3) Radioactivity Transport Analysis The iodine transport analysis considers break flow flashing, steaming, and partitioning. The analysis assumes that a fraction of the iodine carried by
the break flow becomes airborne immediately due to flashing and
atomization. The analysis conservatively took no credit for scrubbing of
iodine contained in the atomized coolant droplets. The fraction of primary
coolant iodine which is not assumed to become airborne immediately
mixes with the secondary water and is assumed to become airborne at a
rate proportional to the steaming rate and the iodine partition coefficient.
This analysis conservatively assumes an iodine partition coefficient of 100
between the steam generator liquid and steam phases. Droplet removal
by the dryers is conservatively assumed to be negligible.
The following assumptions and parameters were used to calculate the
activity released to the atmosphere and the offsite doses following a
SGTR. (a) The mass of reactor coolant discharged into the secondary system through the rupture and the mass of steam released from the
ruptured and intact steam generators to the atmosphere are
presented in Table 15.4-14.
(b) The mass of break flow that flashes to steam and is immediately released to the environment is contained in Table 15.4-14 and is
presented in Figure 15.4.3-11. The break flow flashing fraction was conservatively calculated assuming that 100 percent of the break flow is from the hot leg side of the steam generator, whereas the
break flow actually consists of flow from both the hot leg and cold
leg sides of the steam generator.
(c) No iodine scrubbing is credited for the break flow that flashes in the analysis and the iodine scrubbing efficiency is assumed to be
0 percent. Thus the location of the tube rupture is not significant
for the radiological consequences. However, as discussed in
Section 15.4.3.3, in the thermal and hydraulic analysis the tube
rupture break flow is calculated conservatively assuming that the
break is at the top of the tube sheet.
(d) The rupture (or leakage) site is assumed to be always covered with secondary water based on Reference 33, which concluded the
effect of tube uncovery is essentially neg ligible for the radiological consequences for the limiting SGTR transient.
(e) The total primary to secondary le ak rate for the 3 intact steam generators is assumed to be 1.0 gpm. The leakage to the intact DCPP UNITS 1 &
2 FSAR UPDATE 15.5-57 Revision 22 May 2015 steam generators is assumed to persist for the duration of the accident.
(f) The iodine partition coefficient between the liquid and steam of the ruptured steam generator is assumed to be 100 for non-flashed
flow and 1 for flashed flow. The iodine partition coefficient between
the liquid and steam of the intact steam generator is assumed to be
100. (g) No credit was taken for radioactive decay during release and transport, or for cloud depletion by ground deposition during
transport to the site boundary or outer boundary of the low
population zone.
(h) Short-term atmospheric dispersion factors (/Qs) for accident analysis and breathing rates are provided in Table 15.5-68. The breathing rates were obtained from NRC Regulatory Guide 1.4, Revision 2 (Reference 35). (Note: Although revision 2 was referenced in the analysis, the breathing rates are the same as those in Revision 1, which is the DCPP licensing basis).
(i) The noble gases in the break flow and primary to secondary leakage are assumed to be transferred instantly out of the steam generator to the atmosphere. The whole body gamma doses are
calculated combining the dose from the released noble gases with
the dose from the iodine releases.
(j) For the accident initiated iodine spike case, an iodine spiking factor of 335, obtained from Regulatory Guide 1.195, May 2003 (Reference 44) is assumed.
(4) Offsite Dose Calculation In equations 15.5-17 and 15.5-18, no credit is taken for a cloud depletion
by ground deposition or by radioactive decay during transport to the
exclusion area bounda ry or to the outer boundary of the low population zone. Offsite thyroid doses are calculated using the equation:
()
=ij j)Q/(j)BR(ij)IAR(i DCF Th D (15.5-17)
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-58 Revision 22 May 2015 where: ij)IAR( = integrated activity of iodine nuclide i released during the time interval j in Ci j)BR( = breathing rate during time interval j in meter 3/ second (Table 15.5-68) j)Q/( = atmospheric dispersion factor during time interval j in seconds/meter 3 (Table 15.5-68)
()i DCF = thyroid dose conversion factor via inhalation for iodine nuclide i in rem/Ci (Table 15.5-69)
Th D = thyroid dose via inhalation in rem Offsite whole-body gamma doses are calculated using the equation:
=ij j)Q/(ij)IAR(iE25.0 D (15.5-18) where: ij)IAR( = integrated activity of noble gas nuclide i released during time interval j in Ci j)Q/( = atmospheric dispersion factor during time interval j in seconds/m 3 i E = average gamma energy for noble gas nuclide i in MeV/dis (Table 15.5-70)
D = whole body gamma dose due to immersion in rem
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-59 Revision 22 May 2015 (5) Offsite Dose Results Thyroid and whole-body gamma doses at the Exclusion Area Boundary and the outer boundary of the Low Population Zone are presented in Table 15.5-71. All of these RSG doses are within the allo wable guidelines as specified by the SRP, Revision 2 (Section 15.6.3).
The SGTR dose analysis of record is based on the RSGs and all doses are within 10 CFR 100.11 limits. The limiting dose for the SGTR analysis accepted by the NRC based on the OSGs is the EAB zero to two hour thyroid dose of 30.5 rem for the accident initiated iodine spike analysis
case. This dose exceeds the SRP 15.6.3 allowable guideline value of 30 rem by 0.5 rem. However, the NRC found the 30.5 rem value acceptable in a letter to PG&E, dated February 20,2003, "Issuance of
Amendment: RE: Revision to Technical Specification 1.1, 'Definitions, Dose Equivalent 1-131,' and Revi sed Steam Generator Tube Rupture and Main Steam Line Break Analyses." 15.5.20.2.2 Control Room Exposures Additional analyses were performed to determine the airborne doses to the control room operators from an SGTR. These calculations used the atmospheric releases of radioactivity determined in the analysis discussed in Section 15.5.20.2.1 and Reference 46. The control room is modeled as a discrete volume. The atmospheric dispersion factors calculated for the transfer of activity to the control room intake contained in Table 15.5-68 are used to determine the activity available at the control room intake. The inflow (filtered and unfiltered) to the control room and the control room
filtered recirculation flow are used to calculate the concentration of activity in the control room. Control room parameters used in the analysis are presented in Table 15.5-72.
The control room occupancy factors assumed were taken from Table 15.5-32.
Thyroid, whole body gamma, and beta skin doses are calculated for 30 days in the control room. Although all releases are terminated when the RHR system is put in
service, the calculation is continued to account for additional doses due to continued occupancy.
The total primary to secondary leak rate is assumed to be 1.0 gpm. The leakage to the intact steam generators is assumed to persist for the duration of the accident.
The calculations determine the thyroid doses based on a pre-accident iodine spike and
based on an accident initiated iodine spike with a spiking factor of 335. Both spike
assumptions consider 0.1
µCi/gm D.E. I-131 secondary activity. The whole body doses are calculated combining the dose from the r eleased noble gases with the dose from
the iodine releases.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-60 Revision 22 May 2015 Control room thyroid doses are calculated using the following equation:
()=ij j ij i ThBR*Conc DCF D
(15.5-19) where:
D Th = thyroid dose via inhalation (Rem)
DCF i = thyroid dose conversion factor via inhalation for isotope i (Rem/Ci) (Table 15.5-69)
Conc ij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleak age, filtered recirculation and filtered inflow (Ci-sec/m
- 3) (BR)j = breathing rate during time interval j (m 3/sec) (Table 15.5-68)
Control room whole body doses are calculate d using the following equation:
=ij ij i WB ConcE*GF 1*25.0D (15.5-20) where:
D WB = whole body dose via cloud immersion (Rem) GF = geometry factor, calculated based on Reference 17, using the equation 0.338 V 1173 GF= where V is the control room volume in ft 3 Ei = average gamma disintegration energy for isotope i (MeV/dis)
(Table 15.5-70)
Conc ij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleak age, filtered recirculation and filtered inflow (Ci-sec/m
- 3) Control room skin doses are calculated using the following equation:
=ij ij i ConcE*23.0D
(15.5-21) where D = whole body dose via cloud immersion (Rem)
Ei = average beta disintegration energy for isotope i (MeV/dis) (Table 15.5-70)
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-61 Revision 22 May 2015 Conc i j = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleak age, filtered recirculation and filtered inflow (Ci-sec/m
- 3) Table 15.5-74 presents the resulting airborne doses to the control room operators. The
resultant doses are well below the guideline s of GDC 19, 1971, and are below the corresponding post-LOCA control room exposures presented in Table 15.5-33.
15.5.20.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose in excess of dose guidelines of SRP, Section 15.6.3, Revision 2 (i.e., the 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case) as shown in Table 15.5-71.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of the dose guidelines of SRP, Section 15.6.3, Revision 2 (i.e., 10 CFR 100.11 dose limits for the whole body and the thyroid for the pre-existing iodine spike case and 10 percent of the 10 CFR 100.11 dose limits for the whole body and the thyroid for the accident-initiated iodine spike case) as shown in Table 15.5-71.
(3) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident for both the pre-existing and the accident-initiated iodine spike cases as shown in Table 15.5-74.
As noted in Section 15.5.20.2, the above dose estimates reflect the RSGs and are within 10 CFR 100.11 limits. The SGTR analysis accepted by the NRC based on OSGs is the EAB zero to two hour thyroid dose of 30.5 rem for the accident-initiated iodine spike analysis case. This dose exceeds the SRP 15.6.3 allowable guide line value of 30 rem by 0.5 rem. However, the NRC found the 30.5 rem value acceptable in a letter to PG&E, dated February 20,2003, "Issuance of Amendment: RE: Revision to Technical Specification 1.1, 'Definitions, Dose Equivalent 1-131,' and Revised Steam Generator Tube Rupture and Main Steam Line Break Analyses.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-62 Revision 22 May 2015 15.5.21 RADIOLOGICAL CONSEQUENCES OF A LOCKED ROTOR ACCIDENT 15.5.21.1 Acceptance Criteria (1) The radiological consequences of a locked rotor accident shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
- i. An individual located at any point on the boundary of the exclusion area for the two hours immediately foll owing the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident.
15.5.21.2 Identification of Causes and Accident Description Under adverse circumstances, a locked rotor accident could cause small amounts of fuel cladding failure in the core. If this occurs, some fission products will enter the
coolant and will mostly remain in the coolant until cleaned up by the primary coolant
demineralizers, or in the case of noble gases, until stripped from the coolant. Following such an incident, there are several possible modes of release of some of this activity to
the environment.
In the short-term, if the accident occurs at a time when significant primary-to-secondary leakage exists, some of the additional activit y entering the coolant will leak into the secondary system. The noble gases will be discharged to the atmosphere via the air
ejectors or by way of atmospheric steam dump. The iodines will remain mostly in the
liquid form and be picked up by the blowdow n treatment system. Some fraction of the iodines, however, will be released via the air ejectors or by way of atmospheric steam
dump. In addition, if an atmospheric steam dump is necessary, some of the activity
contained in the secondary system prior to the accident will be released.
The amounts of steam released depend on the time relief valves remain open and the
availability of condenser bypass cooling capacity.
The amounts of radioactive iodine released depend on the amounts of steam released, the amount of activity contained in
the secondary system prior to the accident, and the amount contained in the primary DCPP UNITS 1 &
2 FSAR UPDATE 15.5-63 Revision 22 May 2015 coolant which leaks into the secondary system. As discussed in Section 15.5.10, the amount of steam released following the locked rotor accident, if no condenser cooling is
available, would not exceed approximately 1.7E+06 Ibm. In the analysis of both the design basis case and the expected case, this amount of steam was assumed to be
released.
For the design basis case, it was assumed that the plant had been operating
continuously with 1 percent fuel cladding defects and 1 gpm primary-to-secondary
leakage. For the expected case calculation, operation at 0.2 percent defects and
20 gallons per day to the secondary was assumed. In both cases, leakage of water
from primary to secondary was assumed to continue during cooldown at 75 percent of
the pre-accident rate during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at 50 percent of the pre-accident rate during the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These values were derived from primary-to-secondary
pressure differentials during cooldown. It was also conservatively assumed for both
cases that the iodine Partition Factor in the steam generators releasing steam was 0.01 on a mass basis (Reference 15). In addition, to account for the effect of iodine spiking, fuel escape rate coefficients for iodines of 30 times the normal operation values given in
Table 11.1-9 were used for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the start of the accident. Other
detailed and less significant modelin g assumptions are presented in Reference 4.
The assumptions used for meteorology, breathing rates, population density and other
common factors were also described earlier. Both the primary and secondary coolant
activities prior to the accident are discussed in Section 15.5.2.
In order to determine the primary coolant activities immediately after the accident, it was
assumed that less than 10 percent of the total activity contained in the fuel rod gaps would be immediately released to the coolant and mixed uniformly in the coolant system volume. The gap inventories used are listed in Table 11.1-7.
All of the data and assumptions listed above were used with the EMERALD computer
program to calculate the activity releases and potential doses following the accident.
The calculated activity releases are listed in T able 15.5-41. The potential doses are given in Table 15.5-42. The exposures are also shown in Figures 15.5-14 and 15.5-15
as a function of the amount of fuel failure that occurs. On the left boundary of these
graphs, in the region of negligible fuel failures, the exposures are just the component
resulting from the activity already present in the secondary system, or which leaks
through the steam generators at pre-accident primary coolant levels. These exposures correspond to those shown in Figures 15.5-2 through 15.5-5.
HISTORICAL INFORMATION IN ITALICS BELOW NOT REQUIRED TO BE REVISED.
Another mode of release following a locked rotor accident, or any accident involving significant fuel failure, is the long-term release by way of cleanup and leakage from the primary coolant system. The activity going through these pathways, principally Kr-85, would result in some increm ental long-term dose beyond the normal yearly releases.
This pathway of release has been evaluated, and the results are presented in DCPP UNITS 1 &
2 FSAR UPDATE 15.5-64 Revision 22 May 2015 Figure 15.5-16. Since the activity released in this way would reach the environment over a long term, the annual average at mospheric dilution factors (Table 15.5-5) and breathing rates have been used. The amounts of activity released were determined by
multiplying the activities released from the ga ps following the accident by the release fractions listed in Table 15.5-40.
These long-term release fractions were deter mined from the normal radioactivity transport analysis carried out for Chapter 11, for the anticipated operational occurrences
case. In essence, these fractions are the fractions of a curie reaching the environment
per curie released to the coolant, for each isotope. The pathways included are primary
cleanup, leakage to the containment, and leakage to the auxiliary building. As shown in
Table 15.5-40, essentially all of the Kr-85 released to the coolant is eventually released
to the environment, as would be physically expected, and lower fractions of the other
isotopes are released, depending on their respective overall cleanup, leakage, and decay factors in the plant. It can be conclud ed by comparing these exposures to the short-term exposures in Figure 15.5-12 that t he incremental long-term exposures are negligible additions to the radiological consequences of accidents of this kind.
In addition, it can be concluded that accidents of this kind would not result in significant additions to the annual doses expected from normal plant operation.
From these short-term and long-term analyses, it can also be concluded that all potential exposures from a locked rotor accident will be well below the guideline levels
specified in 10 CFR 100.11, and that the occurrence of such accidents would not result in undue risk to the public. A detailed eva luation of potential exposures to control room personnel was made in Section 15.5.17, for conditions following a LBLOCA. The containment shine contribution to control room dose would not be applicable following a
locked rotor accident.
15.5.21.3 Conclusions By comparing the activity releases following a locked rotor accident, given in Table 15.5-41, with the activity releases calculated for a LBLOCA, given in Tables 15.5-13 and 15.5-14, it can be concluded that any control room exposures following a locked rotor accident will be well below the GDC 19, 1971, criterion level.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-42.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release DCPP UNITS 1 &
2 FSAR UPDATE 15.5-65 Revision 22 May 2015 (during the entire period of its passage), are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-42.
(3) Since the activity releases from the locked rotor accident given in Table 15.5-41 are less than those from a LBLOCA (see Table 15.5-13 and 15.5-14), any control room dose which might occur would be well within the established criteria of GDC 19, 1971 and discussed in Section 15.5.17.
15.5.22 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT The procedures used in handling fuel in the containment and fuel handling area are described in detail in Section 15.4.5. In addition, design and procedural measures
provided to prevent fuel handling accidents a re also described in that section, along with a discussion of past experience in fuel handling operations. The basic events that could
be involved in a fuel handling accident are discussed in that section, and the following discussion evaluates the potential radiological consequences of such an accident.
15.5.22.1 Fuel Handling Accident In The Fuel Handling Area 15.5.22.1.1 Acceptance Criteria The radiological consequences of a fuel handling accident in the fuel handling area shall not exceed the dose limits of 10 CFR 50.67 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for any two hour period following the onset of the postulated fission product release shall not receive a total radiation dose in excess of 0.063 Sv (6.3 rem) total effective dose equivalent (TEDE).
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.063 Sv (6.3 rem) total effective dose equivalent (TEDE).
(3) The dose to the control room operator under accident conditions shall not be in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
15.5.22.1.2 Identification of Causes and Accident Description The radiological consequences of a fuel handling accident in the fuel handling area
were analyzed using the LOCADOSE computer code.
The values assumed for individual fission product inventories are calculated for a
source term assuming approximately 105 percent full power operation (3580 MW DCPP UNITS 1 &
2 FSAR UPDATE 15.5-66 Revision 22 May 2015 thermal) immediately preceding shutdown. The accident is assumed to occur 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown. This latter interval represents approximately the minimum time required
to prepare (cooldown, head and internals removal, cavity flooding, etc.) the core for
refueling and is therefore somewhat conservative in that it would require that the
accident occur during handling of the first few fuel assemblies.
The source term is conservatively assumed to be a composite of the highest fission
product activity totals for various combinations of burnup and enrichment. The
ORIGEN-2 computer code was used to calculate these worst-case fission product
inventories. The DBA gap activity inventory is based on NRC Safety Guide 25, March 1972, assumptions: radial peaking factor of 1.65, gap fraction of 10 percent for noble gases other than Kr-85, gap fraction of 30 percent for Kr-85, and gap fraction of 10 percent for iodines.
The assumption is made for both cases that 100 percent of the activity (consisting
principally of fission product isotopes of the elements xenon, krypton, and iodine)
present in the gap between the fuel pellets and the cladding in the damaged rods is
immediately released to the pool or cavity water. This assumption is conservative for
elemental iodine because the low cladding and gap temperatures would result in a large fraction of it being condensed and temporarily retained within the cladding.
The analysis assumes that the fission product release occurs at a water depth of
23 feet, which is the minimum water depth above the top of the fuel as required by
Technical Specifications. The spent fuel pool, where handling operations are most
likely to result in fuel damage, has a water depth of about 38 feet. Using a depth of
23 feet accounts for cases in which the release occurs from the top of an assembly that is resting vertically on the floor, and for releases that occur near the top of the storage
racks. Finally, consistent with Safety Guide 25, March 1972 the analysis assumes that all activity that escapes from the pool to the fuel handling area air spaces is released
from the area within a 2-hour time period.
Of the activity reaching the water, 100 percen t of the noble gases, xenon and krypton, are assumed to be immediately released to the fuel handling area air spaces. However, the ability of the pool water to scrub iodine from the gas bubbles as they rise to the
surface has been considered. The pool DFs for the inorganic and organic species are
500 and 1, respectively, giving an overall effective DF of 200 (i.e., 99.5 percent of the
total released from the damaged rods is retained by the pool water). This difference in
DFs for inorganic and organic iodine species results in the iodine above the fuel pool
being composed of 75 percent inorganic and 25 percent organic species. These
assumptions are consistent with those suggested in NRC Regulatory Guide 1.183, July 2000. Table 15.5-44 itemizes the gap activity available for release from the FHB atmosphere to the environment.
Table 15.5-45 itemizes the assumptions and numerical values used to calculate the fuel
handling accident radiological exposures. The potential releases of activity to the atmosphere are listed in Table 15.5-44. The exposures resulting from the postulated DCPP UNITS 1 &
2 FSAR UPDATE 15.5-67 Revision 22 May 2015 fuel handling accident inside the fuel handlin g area are presented in Table 15.5-47.
These exposures are well below the Regulatory Guide 1.183, July 2000 limits and demonstrate the adequacy of the fuel handling safety systems.
In the very unlikely event of a serious fuel handling accident and in combination with the conservative assumptions discussed above, containment building or fuel handling area
activity concentrations may be quite high. High activity concentrations necessitate the
evacuation of fuel handling areas in order to limit exp osures to fuel handling personnel.
Upon indication of a serious fuel handling accident, the fuel handling area will be evacuated until the extent of the fuel damage and activity levels in the area can be
determined. Any serious fuel handling accident would be both visually and audibly
detectable via radiation monitors in the fuel handl ing areas that locally alarm in the event of high activity levels and would alert personnel to evacuate.
Although conservatively neglected for this analysis, the fuel handling area has the additional safety feature of ventilation air flow that sweeps the surface of the spent fuel
pool carrying any activity away from fuel handling personnel. This sweeping of the
spent fuel pool is expected to considerably lower activity levels in the fuel handling area in the event of a serious fuel handling accident.
After charcoal filter cleanup (another design feature conservatively neglected in this analysis), fuel handling area post-accident ventilation air exhausts through the plant
vent at a height of 70 meters. Site meteorology is such that it is very unlikely that any airborne activity will enter the control room ventilation system.
Spent fuel cask accidents in the fuel handl ing area causing fuel damage are precluded due to crane travel limits and design and operating features as described in
Sections 9.1.4.3.9 and 9.1.4.2.6. Spent fuel handling accidents in the fuel handling area would not jeopardize the health and safety of the public.
15.5.22.1.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) An individual located at any point on the boundary of the exclusion area for any two hour period following the onset of the postulated fission product release shall not receive a total radiation dose in excess of 0.063 Sv (6.3 rem) total effective dose equivalent (TEDE) as shown in Table 15.5-47.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose in excess of 0.063 Sv (6.3 rem) total effective dose equivalent (TEDE) as shown in Table 15.5-47.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-68 Revision 22 May 2015 (3) The dose to the control room operator under accident conditions shall not be in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident as shown in Table 15.5-47.
15.5.22.2 Fuel Handling Accident Inside Containment 15.5.22.2.1 Acceptance Criteria (1) The radiological consequences of a fuel handling accident inside containment shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
- i. An individual located at any point on the boundary of the exclusion area for the two hours immediately foll owing the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 6 rem or a total radiat ion dose in excess of 75 rem to the thyroid from iodine exposure.
ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 6 rem, or a total radiation dose in excess of 75 rem to the thyroid from iodine exposure.
(2) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident.
15.5.22.2.2 Identification of Causes and Accident Description The offsite radiological consequences of a p ostulated fuel handling accident inside the containment are mitigated by containment closure. The following evaluation shows that in all cases the calculated exposures would be well below limits specified in
During fuel handling operations, containment closure is not required. Generally, the containment ventilation purge system is operational and exhausts air from the containment through two 48-inch containment isolation valves. These two valves are
connected in series. This flow of air from the containment is discharged to the
environment via the plant vent.
This exhaust stream is monitored for activity by monitors in the plant vent. In the event of a postulated fuel handling accident, the plant vent monitors will alarm and result in
the automatic closure of containment ventilation isolation valves. This activity release
may result in offsite radiological exposures.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-69 Revision 22 May 2015 Containment penetrations are allowed to be open during fuel handling operations. The most prominent of these penetrations are the equipment hatch and the personnel
airlock. Closure of these penetrations is achieved by manual means as discussed in
Section 15.4.5. The closure of these penetrations is not credited in the design-basis
fuel handling accident inside containment.
The FHA analysis assumes that the control room ventilation system of each unit
remains in the normal mode of operation following the FHA. Thus, the design basis
FHA does not credit charcoal filtration of the control room atmosphere intake flow or recirculation flow.
The evaluation of potential offsite exposures was performed for a design basis case, assuming plant parameters as limited by Technical Specifications. The assumptions of
Safety Guide 25, March 1972, were used as guidance with the exceptions detailed below.
15.5.22.2.2.1 Activity Released to Containment Atmosphere The assumptions made in determining the quantity of activity available for release from
the containment refueling pool following the postulated accident are identical to those
discussed in Section 15.5.22.1. For the DBA case, these assumptions are consistent
with those in Safety Guides 25, March 1972, and 1.183, July 2000.
Consistent with the guidance of Safety Guide 25, March 1972, it was assumed that all the gap activity in the damaged rods is released and consists of 10 percent of the total noble gases other than Kr-85, 30 percent of the Kr-85, and 10 percent of the total radioactive iodine in the rods at the time of the accident.
An effective DF of 200 for the iodines was assumed for the water in the refueling cavity.
This DF is consistent with the current guidance provided in Regulatory Guide 1.183, July 2000.
The dose conversion factors used are from ICRP Publication 30 (Reference 45). The use of these dose conversion factors is consistent with the current guidance provided in
Regulatory Guide 1.183, July 2000.
15.5.22.2.2.2 Containment Closure
Following the postulated accident, airborne activity evolves from the surface of the pool
where it mixes with air above the pool. Airborne activity is then assumed to be
discharged to the environment via the open penetrations. The duration of the release
was assumed to be within two seconds.
In addition to radiation monitor indications, a fuel handling accident would immediately
be known to refueling personnel at the scene of the accident. These personnel would DCPP UNITS 1 &
2 FSAR UPDATE 15.5-70 Revision 22 May 2015 initiate containment closure actions and are r equired by an Equipment Control Guideline to be in constant communication with control room personnel. The plant intercom system is described in Section 9.5.2.
15.5.22.2.2.3 Activity Released to Environment The containment refueling pool is approxim ately rectangular in shape with approximate dimensions of 25 by 70 feet. The pool has a surface area of about 1750 square feet.
It was assumed that activity evolved from the pool was instantaneously mixed and
retained within the approximately 33,600 cubic foot rectangular parallelepiped formed by the 25- by 70-foot pool and the 40-foot-high steam generators. Where the steam
generators do not surround the pool, the radioactivity would actually be dispersed into a
larger volume of air which would have the effect of reducing the dose. However, for
conservatism, it was assumed that all the radioactivity remained within this
33,600-cubic-foot volume and was then transported to the environment within a two second time period through the open equipment hatch.
15.5.22.2.2.4 Offsite Exposures
The integrated release of activity to the environment and the resulting offsite radiological exposures were calculated for the postulated fuel handling accident inside containment using the LOCADOSE computer program.
Table 15.5-48 itemizes the DBA assumptions and numerical values used to calculate
fuel handling accident radiological exposures.
The calculated releases of activity to the atmosphere are listed in Table 15.5-49. The DBA exposures resulting from the postulated fuel handling accident inside conta inment are presented in Table 15.5-50.
These exposures are well within the 10 CFR 100.11 limits.
15.5.22.2.2.5 Action Following Containment Isolation Following manual containment closure after the fuel handling accident, activity can be removed from the containment atmosphere by the redundant PG&E Design Class II Iodine Removal System (two trains at 12,000 cfm per train), which consists of HEPA/charcoal filters. This system is described in Section 9.4.5. There are no Technical Specification requirements for this filtration system.
The containment can also be purged to the atmosphere at a controlled rate of up to 300 cfm per train through the HEPA/charcoal filters of the hydrogen purge system. This
system is described in Section 6.2.5.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-71 Revision 22 May 2015 15.5.22.2.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-50.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-50.
(3) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem thyroid and beta skin, Reference 51) for the duration of the accident as shown in Table 15.5-50.
15.5.22.3 Conclusion, Fuel Handling Accidents In the preceding sections the potential offsite exposures from major fuel handling
accidents have been calculated. The analyses have been carried out using the models
and assumptions specified in pertinent regulatory guides. In all analyses the resulting potential exposures to individual members of the public and the general population have been found to be lower than the applic able guidelines and limits specified in 10 CFR 100.11. (FHA in Containment) and 10 CFR 50.67 (FHA in FHB).
On this basis, it can be concluded that the occurrence of a major fuel handling accident
in a DCPP unit would not constitute an undue risk to the health and safety of the public.
Additionally, it can be concluded that the ESF provided for the mitigation of the consequences of a major fuel handling accident are adequate.
15.5.23 RADIOLOGICAL CONSEQUENCES OF A ROD EJECTION ACCIDENT 15.5.23.1 Acceptance Criteria (1) The radiological consequences of a rod ejection accident shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
- i. An individual located at any point on the boundary of the exclusion area for the two hours immediately foll owing the onset of the postulated fission product release shall not receive a total radiation dose to the whole body DCPP UNITS 1 &
2 FSAR UPDATE 15.5-72 Revision 22 May 2015 in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
ii. An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) In accordance with the requirements of GDC 19, 1971, the dose to the control room operator under accident conditions shall not be in excess of 5 rem whole body or its equivalent to any part of the body (i.e., 30 rem) for the duration of the accident.
15.5.23.2 Identification of Causes and Accident Description As discussed in Section 15.4.6, under adverse combinations of circumstances, some
fuel cladding failures could occur following a rod ejection accident. In this case, some of
the activity in the fuel rod gaps would be released to the coolant and in turn to the inside
of the containment building. As a result of pressurization of the containment, some of this activity could leak to the environment.
For the design basis case, it was assumed that the plant had been operating
continuously with 1 percent fuel cladding defects and 1 gpm primary-to-secondary
leakage. For the expected case calculation, operation at 0.2 percent defects and
20 gallons per day to the secondary was assumed.
Following a postulated rod ejection accident, activity released from the fuel
pellet-cladding gap due to failure of 10 percent of the fuel rods is assumed to be instantaneously released to the primary coolant. Releases to the primary coolant are
assumed to be immediately and uniformly mixed throughout the coolant.
The activity released to the containment from the primary coolant through the ruptured
control rod mechanism pressure housing is assumed to be mixed instantaneously throughout the containment and is available for leakage to the atmosphere.
It has been assumed for both the design basis and expected cases that 10 percent of
the elemental iodine leaked to the coolant is released to the containment atmosphere as a result of flashing of some of the primary coolant water. Of the amounts of noble gases released to the primary coolant, 100 percent is assumed to be released to the
containment atmosphere at the time of the accident. It is assumed that the amount of
iodine in chemical forms that are not affected by the spray system are negligible. These release fractions are used for both the design basis case and the expected case.
Following the release to the containment, the fission products are assumed to leak from
the containment at the same rates assumed for the LBLOCA, discussed in DCPP UNITS 1 &
2 FSAR UPDATE 15.5-73 Revision 22 May 2015 Section 15.5.17. In addition, the spray system is assumed to be in operation and acts to remove the iodines from the containment atmosphere at the same rates assumed for the LBLOCA.
The assumptions used for meteorology, breathing rates, population density, and other
common factors were also described in earlier sections. Both the primary and
secondary coolant activities prior to the accident are given in Section 15.5.3. The gap activities are listed in Table 11.1-7.
All of the data and assumptions listed above were used with the EMERALD computer
program to calculate the activity releases and potential doses following the accident.
The calculated activity releases are listed in T able 15.5-51, and the potential doses are given in Table 15.5-52. Thyroid doses that would result from secondary steam releases can be determined from Figures 15.5-2 and 15.5-3 for the DBA conditions and Figures 15.5-4 and 15.5-5 for the expected conditions.
If atmospheric steam releases occur followin g this accident, there will be some additional exposures via this pathway.
The detailed assumptions used in estimating mode of exposure are described in Section 15.5-21. The results are given
parametrically in Figures 15.5-14 and 15.5-15. It should be noted that these figures are based on the assumptions of a full plant cooldo wn with no condenser capacity available, a condition that would not be expected to occur following a rod ejection accident.
From these analyses, it can be concluded that offsite exposures from this accident will
be well below the guidelin e levels specified in 10 CFR 100.11, and that the occurrence of such accidents would not result in undue risk to the public. A detailed evaluation of potential exposures to control room personnel is made in Section 15.5.17 for conditions following a LOCA.
15.5.23.3 Conclusions By comparing the activity releases following a rod ejection accident, given in Table 15.5-51, with the activity releases calculated for a LOCA, given in Tables 15.5-13 and 15.5-14, it can be concluded that any control room exposures following a rod ejection accident will be wel l below the GDC 19, 1971, criterion level.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-52.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release DCPP UNITS 1 &
2 FSAR UPDATE 15.5-74 Revision 22 May 2015 (during the entire period of its passage), are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-52.
(3) Since the activity releases from the rod ejection accident given in Table 15.5-51 are less than those from a LBLOCA (see Table 15.5-13 and 15.5-14), any control room dose which might occur would be well below the established criteria of GDC 19, 1971, and discussed in Section 15.5.17.
15.5.24 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A WASTE GAS DECAY TANK 15.5.24.1 Acceptance Criteria The radiological consequences of a rupture of a waste gas decay tank shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
15.5.24.2 Identification of Causes and Accident Description Radioactive waste gas decay tanks are used to permit decay of radioactive gases as a means of reducing or preventing the release of radioactive materials to the atmosphere.
This system is discussed in detail in Section 11.3.
Three gas decay tanks are provided for each unit to afford operating flexibility and allow one or more tanks to be isolated from the rest of the system fo r an extended period of time. Most of the gas stored in the decay tanks is nitrogen cover gas displaced from the liquid waste holdup tanks. The radioactive components are principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of halogens.
A number of combinations of inadvertent operator errors and equipment malfunctions or failures could be identified that might result in a release of some or all of the activity stored in these tanks. In general, the amounts of activity that could be released by any such combination of events are limited in the following ways:
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-75 Revision 22 May 2015 Plant Feature Function Limits on primary coolant activity Restricts total curies present in volume control tank and gas decay tanks Radiation monitor Allows early detection of release, allowing operator action to terminate release Limits on tank size Restricts total curies present in any one tank Isolation valves Allows operator to terminate release Operating procedures Reduces probability of releases In the evaluation of the waste gas decay tank failure accident, the fission product accumulation and release assumptions for the DBA case are consistent with those of
NRC Safety Guide 24, March 1972 (Reference 24). These assumptions are:
(1) The reactor has been operating at full power with 1 percent defective fuel and a shutdown to cold condition has been conducted at the end of an
equilibrium core cycle.
(2) All noble gases have been removed from the primary cooling system and transferred to the gas decay tank that is assumed to fail. No radioactive decay is assumed during transfer.
(3) The failure occurs immediately on completion of the waste gas transfer, releasing the entire maximum contents of the tank to the auxiliary building.
The assumption of the release of the noble gas inventory from only a single tank is based on a design that allows all gas decay tanks to be
isolated from each other when they are in use.
(4) All of the gases are exhausted from the auxiliary building at ground level over a 2-hour time period. There is no decay in the auxiliary building.
The evaluation of the radiation doses resulting from the design basis case accident is
based on the maximum gas decay tank inventories given in Table 11.3-5.
The fission product accumulation and release assumptions used for the expected case
are identical with those used for the DBA basis case, except that the tank inventories
are based on operation with 0.2 percent defective fuel. The radiation doses resulting
from the expected case accident are calculated from the maximum gas decay
inventories given in Table 11.3-6.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-76 Revision 22 May 2015 The whole body doses resulting from the rupture of a gas decay tank were calculated for the time period 0-2 hours using the semi-infinite cloud submersion model.
Atmospheric dispersion factors used in the analysis are given in Tables 15.5-3 and
15.5-4, and the breathing rates used are given in Table 15.5-7. Due to the presence of
only trace amounts of iodine in the waste gas tanks, inhalation thyroid doses are
negligible.
15.5.24.3 Conclusions The resulting approximate radiation exposures from the rupture of a gas decay tank are presented in Table 15.5-53. As shown in the table, the individual doses are all well below the guideline doses of 10 CFR 100.11.
The analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-53.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-53.
15.5.25 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A LIQUID HOLDUP TANK 15.5.25.1 Acceptance Criteria The radiological consequences of a rupture of a liquid hold tank shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-77 Revision 22 May 2015 15.5.25.2 Identification of Causes and Accident Description Radioactive liquid waste holdup tanks are used as part of the chemical and volume control system (CVCS) to collect and permit decay of radioactive liquids drawn from the reactor primary coolant for reactivity control.
The CVCS is described in detail in Section 9.3.4.
Five liquid holdup tanks are provided for the two units to afford operating flexibility and allow one or more tanks to be isolated from the rest of the system for extended periods of time. The liquid processed through the holdup tanks contains dissolved fission and activation products, as well as radioactive no ble gases mixed with nitrogen cover gas used in the tanks.
The liquid holdup tanks are located in vaults which are Design Class I structures, so that in the event of a rupture or spill all liquids are retained in the vaults. The volume of holdup tank vaults is sufficient to contain the full contents of the holdup tank without spillage from the vaults. Any gases released from the liquid holdup tanks are collected by the auxiliary building ventilation system and discharged via the auxiliary building vent. In the evaluation of the liquid waste holdup tank rupture accident, the following fission
product accumulation and release assumptions are used for the design basis case:
(1) The reactor has been operating at full power with 1 percent defective fuel for an equilibrium core cycle.
(2) A liquid holdup tank has been filled with primary coolant at a rate of 120 gpm, with credit for decay as the tank is filling.
(3) The failure occurs immediately upon completion of the liquid transfer, releasing the entire contents of the tank to t he auxiliary build ing vault. The assumption of the release of the contents of only a single tank is based on
a design that allows all liquid holdup tanks to be isolated from each other when they are in use.
(4) All of the noble gases and varying amounts of the iodines are released from the auxiliary building vault to the auxiliary building atmosphere.
These effluents are exhausted from the auxiliary building at ground level.
There is no decay in the auxiliary building.
No liquids escape from the vaults during the accident.
The whole body radiation doses resulting from the rupture of a liquid holdup tank were
calculated for the time period 0-2 hours using the semi-infinite cloud submersion model, and the inhalation thyroid doses were calculated using the models discussed earlier.
Atmospheric dispersion factors used in the analysis are given in Tables 15.5-3 and
15.5-4, and the breathing rates used are given in Table 15.5-7.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-78 Revision 22 May 2015 15.5.25.3 Conclusions The resulting radiation exposures from the rupture of a liquid holdup tank are listed in Table 15.5-56. As shown in the table, the indi vidual doses are well below the guideline doses of 10 CFR 100.11.
Additionally, the analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-56.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-56.
15.5.26 RADIOLOGICAL CONSEQUENCES OF A RUPTURE OF A VOLUME CONTROL TANK 15.5.26.1 Acceptance Criteria The radiological consequences of a rupture of a volume control tank shall not exceed the dose limits of 10 CFR 100.11 as outlined below:
(1) An individual located at any point on the boundary of the exclusion area for the two hours immediately following the onset of the postulated fission product release shall not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), shall not receive a total radiation dose to the whole body in excess of 25 rem, or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
15.5.26.2 Identification of Causes and Accident Description The volume control tank is used as part of the CVCS to collect the excess water released from the RCS during modes 1 through 6, which is not accommodated by the pressurizer. For a complete description of the CVCS and the volume control tank in all modes of operation refer to Section 9.3.4.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-79 Revision 22 May 2015 The liquid processed through the volume control tank contains dissolved fission and activation products, as well as undissolved radioactiv e noble gases. A spray nozzle located inside the tank on the inlet line strips part of the noble gases from the incoming liquid, and these gases are retained in the volume control tank vapor space. In addition, an overpressure of hydrogen cover gas is pr ovided for the tank to control the hydrogen concentration in the reactor coolant.
The volume control tank is located in a vault which is a PG&E Design Class I structure, so that in the event of a rupture or spill all liquids are retained in the vault. The volume of the tank vault is sufficient to contain the full contents of the tank without spillage from the vault. Any gases released from the volume control tank are collected by the auxiliary building ventilation system and discharged via the auxiliary building vent.
In the evaluation of the volume control tank rupture accident, the following fission
product accumulation and release assumptions are used for the design basis case:
(1) The reactor has been operating at full power with 1 percent defective fuel for an equilibrium core cycle.
(2) The volume control tank contains its maximum equilibrium inventory of radioactivity at the time of the ac cident. The failure of the tank releases the entire tank contents to the containment vault.
(3) All of the noble gases and 10E-4 of the iodines are released from the containment vault to the auxil iary building atmosphere. These effluents are exhausted from the auxiliary building at ground level over a 2-hour time period through the auxiliary buil ding filters, which have efficiencies of 90 percent for iodines and 0 percent for noble gases. A discussion of the
assumed effectiveness of the auxiliary build ing charcoal filters is given in Section 15.5.17. There is no decay in the auxiliary building. No liquids
escape from the vault during the accident.
The evaluation of the radiation ex posures resulting from the postulated accident for the design basis case is based on the maximum tank inventories given in Table 11.3-7.
The fission product accumulation and release assumptions for the expected case are
identical with those used for the design bas is case, with the exceptions that the tank inventories are based on operation with 0.2 percent defective fuel and the auxiliary
building filter efficiency is 99 percent for iodines. The evaluation of the resulting
radiation exposures for the expected case is based on the maximum tank inventories
given in Table 11.3-8.
The whole body radiation doses resulting from the rupture of a volume control tank were
calculated for the time period 0-2 hours using the semi-infinite cloud submersion model
as discussed in earlier sections, and the inhalation thyroid doses were calculated using the models described in Reference 4. Atmospheric dispersion factors used in the DCPP UNITS 1 &
2 FSAR UPDATE 15.5-80 Revision 22 May 2015 analysis are given in Tables 15.5-3 and 15.5-4, and the breathing rates used are given in Table 15.5-7. The resulting radiation ex posures are listed in Table 15.5-57. As shown in the table, the individual doses are well below the guideline doses of 10 CFR 100.11.
If credit for the Auxiliary Building fi lters is not taken, the dose contributions from noble gases are unchanged and the dose contributions from iodines are increased by a factor
of ten. The resulting thyroid doses are increased by a factor of ten from those in Table 15.5-57, and the resulting whole body doses are not significantly affected. These
results are still well below the guid eline doses of 10 CFR 100.11.
15.5.26.3 Conclusions The analysis demonstrates that the acceptance criteria are met as follows:
(1) The radiation dose to the whole body and to the thyroid of an individual located at any point on the boundary of the exc lusion area for the two hours immediately following the onset of the postulated fission product release are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-57.
(2) The radiation dose to the whole body and to the thyroid of an individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), are well below the dose limits of 10 CFR 100.11 as shown in Table 15.5-57.
15.5.27 REFERENCES
- 1. Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plant, N18.2, American Nuclear Society, 1972.
- 2. Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, US Atomic Energy Commission (AEC), Rev. 1, October 1972.
- 3. Regulatory Guide 4.2, Preparation of Environmental Reports for Nuclear Power Plants, Directorate of Regulatory Standards, AEC, March 1973.
- 4. W. K. Burnot, et al, EMERALD (REVISION I) - A Program for the Calculation of Activity Releases and Potential Doses, Pacific Gas and Electric Company, March 1974.
- 5. S. G. Gillespie and W. K. Brunot, EMERALD NORMAL - A Program for the Calculations of Activity Releases and Doses from Normal Operation of a Pressurized Water Plant, Program Description and User's Manual, Pacific Gas and Electric Company, March 1973.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-81 Revision 22 May 2015
- 6. Regulatory Guide Number 1.4, Assumptions Used for Evaluating the Potential Radiological Conse quences of a Loss-of-Coolant Accident for Pressurized Water Reactors, AEC, Rev. 1, June 1973.
- 7. D. H. Slade, ed., Meteorology and Atomic Energy 1968, AEC Report Number TID-24190, July 1968.
- 8. International Commission on Radiological Protection (ICRP) Publication 2, Report of Committee II, Permissible Dose for Internal Radiation, 1959.
- 9. R. L. Engel, et al, ISOSHLD - A Computer Code for General Purpose Isotope Shielding Analysis, BNWL-236, UC-34, Physics, Pacific Northwest Laboratory, Richland, WA, June 1966.
- 10. R. K. Hilliard, et al, "Removal of Iodine and Particles by Sprays in the Containment Systems Experiment," Nuclear Technology, April 1971.
- 11. G. L. Simmons, et al, ISOSHLD-II: Code Revision to Include Calculation of Dose Rate from Shielded Bremsstrahlung Sources, BNWL-236-SUP1, UC-34, Physics, Pacific Northwest Laboratory, Richland, WA, March 1967.
- 12. L. F. Parsly, Calculation of Iodine - Water Partition Coefficients, ORNL-TM-2412, Part IV, January 1970.
- 13. Westinghouse, Radiological Con sequences of a Fuel Handling Accident, December 1971.
- 14. F. J. Brutschy, et al, Behavior of Iodine in Reactor Water During Plant Shutdown and Startup, General Electric Co. Atomic Power Equipment Department Report, NEDO-10585, August 1972.
- 15. Deleted in Revision 16.
- 16. Proposed Addendum to ANS Standard N18.
2, Single Failure Criteria for Fluid Systems, American Nuclear Society, May 1974.
- 17. K. G. Murphy and K. M. Campe, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criteria 19," 13th AEC Air Cleaning Conference, August 1974.
- 18. M. L. Mooney and H. E. Cramer, Meteorological Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1970 (see also Appendix 2.3A in Reference 27 of Section 2.3 in this
FSAR Update).
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-82 Revision 22 May 2015
- 19. M. L. Mooney, First Supplement, Meteorological Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1971 (see also Appendix 2.3C in Reference 27 of Section 2.3 in this FSAR Update).
- 20. M. L. Mooney, Second Supplement, Me teorological Study of the Diablo Canyon Nuclear Power Plant Site, Meteorological Office, Gas Control Department, PG&E, 1972 (see also Appendix 2.3D in Reference 27 of Section 2.3 in this
FSAR Update).
- 21. International Commission on Radiological P rotection Publication 30, Limits for Intakes of Radionuclides by Workers, 1979.
- 22. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR 82, as amended.
- 23. Safety Guide 25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors, USNRC, March 1972.
- 24. Safety Guide 24, Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure, USNRC, March 1972.
- 25. Start, G. E., J.
H. Cate, C. R. Dickson, N. R. Ricks, G. H. Ackerman, and J. F. Sagendorf, "Rancho Seco B uilding Wake Effects on Atmospheric Diffusion, NOAA Technical Memorandum, ERL ARL-69, 1977.
- 26. Walker, D. H., R. N. Nassano, M. A. Capo, "Control Room Ventilation Intake Selection for the Floating Nuclear Power Plant," 14th ERDA Air Cleaning Conference, 1976.
- 27. Hatcher, R. N., R. N. Meroney, J. A. Peterka, K. Kothari, "Dispersion in the Wake of a Model Industrial Complex," NUREG-0373, 1978.
- 28. Meroney, R. N., and B. T. Yang, Wind Tunnel Study on Gaseous Mixing due to Various Stack Heights and Injection Rates Above an Isolated Structure, FDDL Report CER 71-72 RNM-BTY16, Colorado State University, 1971.
- 29. R. P. Hosker, Jr., "Dispersion in the Vicinity of Buildings," Preprints of Second Joint Conference on Applications in Air Pollution Meteorology and Second Conference on Industrial Meteorology, New Orleans, LA, March 24-28, 1980, pp.92-107, American Meteorological Society, Boston, Mass. Also in "Flow and
Diffusion Near Obstacles," Chapter 7 of Atmospheric Sciences and Power Production, D. Randerson, ed., USDOE.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-83 Revision 22 May 2015
- 30. D. J. Wilson, Contamination of Air Intakes from Roof Exhaust Vents, ASHRAE Trans. 82, Part 1, pp. 1024-1038, 1976.
- 31. R. J. B. Bouwmeester, K. M.
Kothari, R. N. Meroney, An Algorithm to Estimate Field Concentrations Under Nonsteady Meteorological Conditions from Wind Tunnel Experiments, NUREG/CR-1474, USNRC, September 1980.
- 32. R. Bhatia, J. Dodds, and J. Schulz, Building Wake /Qs for Post-LOCA Control Room Habitability, Bechtel Power Corporation, San Francisco, CA.
- 33. Report on the Methodology for the Resolution on the Steam Generator Tube Uncovery Issue, WCAP-13247, March 1992.
- 34. Deleted in Revision 18.
- 35. Regulatory Guide 1.4, Assumptions Used for Evaluating the Potential Radiological Consequ ences of a Loss of Coolant Accident for Pressurized Water Reactors, AEC, Revision 2, June 1974.
- 36. T. R. England and R. E. Schenter, ENDF-223, ENDF/B-IV Fission Product Files: Summary of Major Nuclide Data, October 1975.
- 37. Standard Review Plan, Section 15
.6.3, Radiological Consequences of Steam Generator Tube Failure (PWR), NUREG-0800, USNRC, July 1981.
- 38. Deleted in Revision 18.
- 39. Westinghouse Letter PGE-91-533, Safety Evaluation for Containment Spray Flow Rate Reduction, February 7, 1991.
- 40. Westinghouse Letter PGE-93-652 dated October 5, 1993, transmitting NSAL-93-016, Revision 1.
- 41. K. F. Eckerman et. al., Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988.
- 42. K. F. Eckerman and J. C. Ryman, External Exposure to Radi onuclides in Air, Water, and Soil, Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993.
- 43. Deleted in Revision 12.
DCPP UNITS 1 &
2 FSAR UPDATE 15.5-84 Revision 22 May 2015
- 44. Regulatory Guide 1.195, Methods and Assumptions for Evaluating Radiological Consequ ences of Design Basis Accidents at Light-Water Nuclear Power Reactors, 05/2003.
- 45. International Commission on Radiological Protection (ICRP) Publication 30, Limits For Intakes of Radionuclides by Workers, 07/1978.
- 46. Diablo Canyon Units 1 and 2 Replacement Steam Generator Program - NSSS Licensing Report, WCAP-16638 (Proprietary), September 2007.
- 47. LOCADOSE-NE319, A Computer Code System for Multi-Region Radioactive Transport and Dose Calculation, Release 6, Bechtel Corporation.
- 49. Diablo Canyon Units 1 and 2 T avg and T feed Ranges Program NSSS Engineering Report, WCAP-16985 (Proprietary), April 2009.
- 50. A. G. Croff, ORIGEN2 - A Revised and Updated Version of the Oak Ridge Isotope Generation and Depletion Code, ORNL-5621, Oak Ridge National Laboratory, July 1980.
- 51. NRC Letter, License Amendment No. 155/155, Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendment RE: Revision of Technical Specifications Section 3.9.4, Containment Penetrations (TAC Nos. MB3595 and MB3596), Accession No. ML021010606, October 21, 2002.
- 52. "RADTRAD: A Model for Radionuclide Transport, Removal and Dose," NUREG/CR-6604, SAND98-0272, April, 1998.
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 1 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 1 Uncontrolled control rod assembly withdrawal from a sub-critical condition (assuming the most unfavorable reactive conditions of the core and reactor coolant system), including control rod or temporary control device removal error during refueling.
15.2.1 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL FROM A SUBCRITICAL CONDITION 2 Uncontrolled control rod assembly withdrawal at the critical power (assuming the most unfavorable reactive conditions of the core and reactor coolant system) which yields the most severe results (hot at zero power, full power, etc). 15.2.2 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL AT POWER 3 Control rod misoperation or sequence of misoperations.
15.2.3 ROD CLUSTER CONTROL ASSEMBLY MISOPERATION 4 Chemical and volume control system malfunction.
15.2.4 UNCONTROLLED BORON DILUTION 5 Partial and total loss of reactor coolant flow force including trip of pumps and pump shaft seizures.
15.2.5 PARTIAL LOSS OF FORCED REACTOR COOLANT FLOW 15.3.4 COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 2 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 6 Start-up of an inactive reactor coolant loop or recirculating loop at incorrect temperature.
15.2.6 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP Precluded in Modes 1 and 2 due to Tech Spec 3.4.4 7 Loss of external electrical load and/or turbine stop valve closure, including, for BWRs closure of main steam isolation valve.
15.2.7 LOSS OF EXTERNAL ELECTRICAL LOAD AND/OR TURBINE TRIP 8 Loss of normal and/or emergency feedwater flow. 15.2.8 LOSS OF NORMAL FEEDWATER 9 Loss of all a-c power to the station auxiliaries (station blackout).
15.2.9 LOSS OF OFFSITE POWER TO THE STATION AUXILIARIES Station Blackout is beyond design basis. Refer to UFSAR Section 8.3.1.6 10 Heat removal greater than heat generation due to (1) feedwater system malfunctions, (2) a pressure regulator failure, or inadvertent opening of a relief valve or safety valve, and (3) a regulating instrument failure. 15.2.10 15.2.12 EXCESSIVE HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTIONS EXCESSIVE LOAD INCREASE INCIDENT Regarding (2) and (3):
There are no pressure regulators or regulating instruments in the Westinghouse pressurized water reactor (PWR) design whose failure could cause heat removal greater than heat generation.
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 3 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 11 Failure of the regulating instrumentation, causing for example, a power-coolant mismatch. Include reactor coolant flow controller failure resulting in increasing flow. N/A N/A N/A N/A N/A N/A Reactor coolant flow controller is not a feature of the Westinghouse PWR design. Treatment of the performance of the reactivity controller in a number of accident conditions is offered in this chapter. (Chapter 15)
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 4 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 12 Internal and external events such as major and minor fires, flood, storms or earthquakes.
N/A N/A N/A N/A N/A N/A Refer to the following Sections:
3.3 - Wind & Tornado Loadings 3.4 - Water level (flood) design 3.5 - Missile protection 3.7 - Seismic design 3.8 - Design of Class I structures 9.5.1 - Fire protection system DCPP UNITS 1 & 2 FSAR UPDATE Sheet 5 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 13 Loss of coolant accidents resulting from the spectrum of postulated piping breaks within the reactor coolant pressure boundary and relief and safety valve blowdowns.
15.2.13 ACCIDENTAL DEPRESSURIZATION OF THE REACTOR COOLANT SYSTEM 15.3.1 LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES THAT ACTUATE EMERGENCY CORE COOLING SYSTEM 15.4.1 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOCA) 14 Spectrum of postulated steam and feedwater system piping breaks inside and outside containment.
15.2.14 ACCIDENTAL DEPRESSURIZATION OF THE MAIN STEAM SYSTEM 15.3.2 MINOR SECONDARY SYSTEM PIPE BREAKS 15.4.2 MAJOR SECONDARY SYSTEM PIPE RUPTURE 15 Inadvertent loading and operation of a fuel assembly into an improper position.
15.3.3 INADVERTENT LOADING OF A FUEL ASSEMBLY INTO AN IMPROPER POSITION 16 Waste gas decay tank leakage or rupture.
15.4.7 RUPTURE OF A WASTE GAS DECAY TANK 17 Failure of air ejector lines (BWR).
N/A N/A N/A N/A N/A N/A This applies to BWR plants only 18 Steam generator tube rupture (PWR).
15.4.3 STEAM GENERATOR TUBE RUPTURE (SGTR) 19 Failure of charcoal or cryogenic system (BWR). N/A N/A N/A N/A N/A N/A This applies to BWR plants only DCPP UNITS 1 & 2 FSAR UPDATE Sheet 6 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 20 The spectrum of rod ejection accidents (PWR). 15.4.6 RUPTURE OF A CONTROL ROD DRIVE MECHANISM HOUSING (ROD CLUSTER CONTROL ASSEMBLY EJECTION) 21 The spectrum of rod drop accidents (BWR). N/A N/A N/A N/A N/A N/A This applies to BWR plants only 22 Break in instrument line or other lines from reactor coolant pressure boundary that penetrate containment.
N/A N/A N/A N/A N/A N/A No instrument lines from the RCS boundary in the DCPP design penetrate the containment. (For definition of the RCS boundary, refer to the 1972 issue of ANS N18.2, Nuclear Safety Criteria for the Design of Stationary PWR Plants.)
23 Fuel handling accident.
15.4.5 FUEL HANDLING ACCIDENT 24 Small spills or leaks of radioactive material outside containment.
N/A N/A N/A N/A N/A N/A The analysis of the consequences of such small spills and leaks is included within the cases evaluated in Chapter 11, and larger leaks and spills are analyzed in Section 15.5.
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 7 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 25 Fuel cladding failure (BWR, PWR) combined with steam generator leak (PWR).
N/A N/A N/A N/A N/A N/A The radiological consequences of this event are analyzed in Chapter 11, for the case of "Anticipated Operational Occurrences." 26 Control room uninhabitabi1ity.
N/A N/A N/A N/A N/A N/A Habitability of the control room following accident conditions is discussed in Chapter 6, and potential radiological exposures are reported in Section 15.5. In addition, Chapter 7 contains an analysis showing that the plant can be brought to, and maintained in, Mode 3 from outside the control room.
27 Failure or overpressurization of low pressure residual heat removal system.
N/A N/A N/A N/A N/A N/A Overpressurization of the residual heat removal system (RHRS) is considered extremely unlikely. PG&E reviewed possible RHRS overpressure scenarios and qualified the system for all credible high pressure transients in DCPP design change package N-049118.
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 8 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 28 Loss of condenser vacuum. N/A N/A N/A N/A N/A N/A This event is covered by the analyses of Section 15.2.7. Separate event analysis is not required.
29 Turbine trip with coincident failure of turbine bypass valves to open.
N/A N/A N/A N/A N/A N/A This event is covered by the analyses of Section 15.2.7. Separate event analysis is not required.
30 Loss of service water system. N/A N/A N/A N/A N/A N/A Malfunctions of auxiliary saltwater system and component cooling water system (CCWS) are discussed in Chapter 9, Sections 9.2.7 and 9.2.2 respectively.
31 Loss of one (redundant) d-c system. N/A N/A N/A N/A N/A N/A There are no significant safety-related consequences of this event. 32 Inadvertent operation of ECCS during power operation.
15.2.15 SPURIOUS OPERATION OF THE SAFETY INJECTION SYSTEM AT POWER 33 Turbine trip with failure of generator breaker to open.
N/A N/A N/A N/A N/A N/A The effects of turbine trip on the RCS are presented in Section 15.2.7.
Separate event analysis is not required.
34 Loss of instrument air system. N/A N/A N/A N/A N/A N/A Malfunctions of this system are discussed in Section 9.3.2.
DCPP UNITS 1 & 2 FSAR UPDATE Sheet 9 of 9 TABLE 15.0-1 REGULATORY GUIDE 1.70 REVISION 1, APPLICABILITY MATRIX Revision 22 May 2015 Event REG GUIDE 1.70, TABLE 15-1, REPRESENTATIVE TYPES OF EVENTS TO BE ANALYZED IN CHAPTER 15.0 OF THE SAR 15.2 Section UFSAR SECTION 15.2 CONDITION II:
FAULTS OF MODERATE FREQUENCY 15.3 Section UFSAR SECTION 15.3 CONDITION III:
INFREQUENT FAULTS 15.4 Section UFSAR SECTION 15.4 CONDITION IV:
LIMITING FAULTS Location of Analyses or Reason Why Not Applicable 35 Malfunction of turbine gland sealing system.
N/A N/A N/A N/A N/A N/A The radiological effects of this event are not significant for PWR plants.
Minor leakages are within the scope of the analysis cases presented in Chapter 11.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-1 Revision 22 May 2015 NUCLEAR STEAM SUPPLY SYSTEM POWER RATINGS Core Rated Thermal Power 3411 Thermal power generated by the reactor coolant
pumps minus heat losses to containment and
letdown system (b) 14 Nuclear steam supply system (NSSS) thermal power output (b) 3425 The engineered safety features design 3570 rating (maximum calculated turbine rating)(a)
(a) The units will not be operated at this rating because it exceeds the license ratings. (b) As noted on Table 15.1-4, some analyses assumed a full power NSSS thermal output of 3,423 MWt, based on the previous net reactor coolant pump heat of 12 MWt. An evaluation concluded that the effect of an additional 2 MWt for NSSS is negligible such that analyses based on 3,423 MWt remain valid.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-2 Sheet 1 of 1 Revision 23 December 2016 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Trip Limiting TripPoint Assumed Time Delay, Function In Analyses sec Power range high neutron flux, high setting 118% 0.5 Power range high neutron flux, low setting 35% 0.5 Power range high positive nuclear power rate 7% / 2 sec 0.65 Overtemperature T Variable, see Figure 15.1-1 7 (a) Overpower T Variable, see Figure 15.1-1 7 (a)
High pressurizer pressure 2460 psia 2
Low pressurizer pressure 1860 psia 2
High pressurizer water level 100% N/A (f) Low reactor coolant flow (from loop flow
detectors) 85% loop flow (b , d) 1 Undervoltage trip (b) N/A (b) Low-low steam generator level 8.2% of narrow range
level span 2 (c)
High steam generator level trip of the feedwater
pumps and closure of feedwater system valves
and turbine trips 100% of narrow
range level span (e) 2 _________________
(a) Total time delay consists of a maximum 5-second RTD lag time constant and a maximum 2-second electronics delay (b) Complete loss of flow analysis assumes that reactor trips, on low reactor coolant loop flow, not undervoltage.underfrequency. (c) When below 50% power, a variable trip time delay is utilized as discussed in Section 7.2.2.1.5.
(d) Value used in the analysis of the Locked Rotor event (Section 15.4.4) for RCS pressure and maximum cladding Temperature. Westinghouse letter PGE-96-582, Diablo Canyon Units 1 & 2 Evaluation of Revised Low Reactor Coolant Flow Reactor Trip Setpoint, June 27, 1996, concludes that a safety analysis setpoint of 85% loop flow
is acceptable for the Locked Rotor event (Section 15.4.4) and the Partial Loss of Flow event (Section 15.2.5),
for which 87% was assumed in the analysis. (e) The analysis assumed 100% narrow range level span for conservatism. The plant setpoint analytical limit is 98.8% narrow range level span for Model Delta 54 steam generators due to void effects. Although the turbine
trip is modeled for completeness it is not needed for DNBR analysis. (f) Westinghouse Letter PGE-02-72, Diablo Canyon Units 1 and 2 Evaluation of Reactor Trip Functions for Uncontrolled RCCA Bank Withdrawal at Power, December 13, 2002, documents that a specific response time is
not assumed since it is not a sensitive parameter for the generic evaluation results.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-4 Sheet 1 of 4 Revision 23 December 2016
SUMMARY
OF INITIAL CONDITIONS AND COMPUTER CODES USED Assumed Reactivity Coefficients Initial NSSS Thermal
Events Computer Codes Utilized Moderator Temp (a), pcm/°F (d) Moderator Density (a), k/gm/cc
Doppler (b) Power Output Assumed (c), MWt CONDITION II Uncontrolled RCCA bank withdrawal from a
subcritical condition
- TWINKLE, THINC, FACTRAN +5 - Least negative defect - 954
pcm 0 Uncontrolled RCCA bank withdrawal at power LOFTRAN
+5 0.43 Lower and Upper 3,423 RCCA misoperation THINC, ANC, LOFTRAN - - Lower 3,425 Uncontrolled boron dilution 0 and 3,423 Partial loss of forced reactor coolant flow LOFTRAN, THINC, FACTRAN +5 - Upper 3,423 Startup of an inactive reactor coolant loop LOFTRAN, FACTRAN, THINC - 0.43 Lower 2,396 Loss of external electrical load and/or turbine
trip-DNBR
- LOFTRAN, RETRAN-02 +5 0.43 Lower and Upper 3,423 Loss of external electrical load and/or turbine
trip - Overpressure RETRAN-02 +5 - Lower 3,425 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-4 Sheet 2 of 4 Revision 23 December 2016 Assumed Reactivity Coefficients Initial NSSS Thermal Events Computer Codes Utilized Moderator Temp (a), pcm/°F (d) Moderator Density (a), k/gm/cc
Doppler (b) Power Output Assumed (c), MWt CONDITION II (Cont'd)
Loss of normal feedwater RETRAN-02W 0 - Upper 3,425 Loss of offsite power to the plant auxiliaries RETRAN-02W 0 - Upper 3,425 Excessive heat removal due to feedwater
system malfunctions RETRAN-02W - 0.43 Lower 3,425 Excessive load increase LOFTRAN - 0 and 0.43 Lower and Upper 3,423 Accidental depressurization of the reactor
coolant system LOFTRAN +7 - Lower 3,425 Inadvertent operation of ECCS during power
operation - DNBR LOFTRAN +5 0.43 Lower and Upper 3,423 Inadvertent operation of ECCS during power RETRAN - - 3,425 Operation - Pressurizer Overfill CONDITION III Loss of reactor coolant from small ruptured
pipes or from cracks in large pipe which
actuate emergency core cooling NOTRUMP SBLOCTA - - - 3,479 Inadvertent loading of a fuel assembly into an
improper position PHOENIX-P, ANC - - - 3,483 Complete loss of forced reactor coolant flow LOFTRAN, THINC, FACTRAN 0 - Upper 3,425 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-4 Sheet 3 of 4 Revision 23 December 2016 Assumed Reactivity Coefficients Initial NSSS Thermal Events Computer Codes Utilized Moderator Temp (a), pcm/°F (d) Moderator Density (a), k/gm/cc
Doppler (b) Power Output Assumed (c), MWt CONDITION III (Cont'd)
Single RCCA withdrawal at full power ANC, THINC, PHOENIX-P - - - 3,425 (e)
CONDITION IV Major rupture of pipes containing reactor
coolant up to and including double-ended
rupture of the largest pipe in the reactor
coolant system (loss-of-coolant accident)
WCOBRA/TRAC HOTSPOT MONTECF Function of
moderator
density. See
Sec. 15.4.1 0 Function of fuel temp.
3,479 Major secondary system pipe rupture up to and
including double-ended rupture (rupture of a
steam pipe)
RETRAN-02W, ANC,THINC - Function of moderator
density. See
Figure 15.4.2-2.
See Figure
15.4.2-1 0.0 (Subcritical)
Major rupture of a main feedwater pipe RETRAN-02W 0.0 Lower 3,425
Rupture of a main steam line at power RETRAN-02W, ANC, THINC-IV 0.43 Lower 3,425 Waste gas decay tank rupture - - - - 3,577 Steam generator tube rupture RETRAN-02W - 0.0 Lower and Upper 3,425 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-4 Sheet 4 of 4 Revision 23 December 2016 Assumed Reactivity Coefficients Initial NSSS Thermal Events Computer Codes Utilized Moderator Temp (a), pcm/°F (d) Moderator Density (a), k/gm/cc
Doppler (b) Power Output Assumed (c), MWt CONDITION IV (Cont'd)
Single reactor coolant pump locked rotor LOFTRAN, THINC, FACTRAN
+5 (rods in DNB case) 0.0 (pressure case) Upper 3,425 Fuel handling accident 3,577 Rupture of a control rod mechanism housing (RCCA ejection)
TWINKLE,
FACTRAN PHOENIX-P
+5.2 BOL -23.EOL - Least negative defect. See
Table 15.4-11.
0 and 3,423
(a) Only one is used in analysis, i.e., either moderator temperature or moderator density coefficient. (b) Reference Figure 15.1-5.
(c) Two percent calorimetric error considered where applicable.
(d) Pcm means percent mille. See footnote Table 4.3-1.
(e) Analysis only models core thermal power of 3411 MWt
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 1 of 9 Revision 23 December 2016 TIME SEQUENCE OF EVENTS FOR CONDITION II EVENTS Accident Event Time, sec Uncontrolled RCCA Withdrawal from a Subcritical Condition Initiation of uncontrolled rod withdrawal 7.5 x 10
-4 k/sec reactivity insertion rate from 10
-9 of nominal power 0.0 Power range high neutron
flux low setpoint reached 9.6 Peak nuclear power occurs 9.8 Rods begin to fall into core 10.1 Peak heat flux occurs 11.9 Peak hot spot average
cladding temperature
occurs 12.3 Uncontrolled RCCA Withdrawal at Power
- 1. Case A Initiation of uncontrolled
RCCA withdrawal at
reactivity insertion rate of7.5 x 10
-4 k/sec 0.0 Power range high neutron flux high trip point reached 1.6 Rods begin to fall into core 2.1 Minimum DNBR occurs 3.0 2. Case B Initiation of uncontrolled
RCCA withdrawal at a
reactivity insertion rate of3.0 x 10
-5 k/sec 0.0 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 2 of 9 Revision 23 December 2016 Accident Event Time, sec Overtemperature T reactor trip signal initiated 31.8 Rods begin to fall into core 33.8 Minimum DNBR occurs 34.2 Uncontrolled Boron Dilution 1Dilution during refueling Dilution begins Shutdown margin lost 0.0 ~11884 2. Dilution during cold shutdown a. RCS filled Dilution begins 0.0 Shutdown margin lost
>900 b. RCS drained Dilution begins 0.0 Shutdown margin lost
>900 3. Dilution during hot shutdown a. One RCP operating Dilution begins 0.0 Shutdown margin lost
>900 b. RHR operating Dilution begins 0.0 Shutdown margin lost
>900 4. Dilution during hot standby Dilution begins 0.0 Shutdown margin lost
>900 5. Dilution during startup Dilution begins 0.0 Shutdown margin lost
~3696 6. Dilution during full power operation a. Automatic reactor control 1.6 % shutdown margin lost ~1180 b. Manual reactor control Dilution begins 0.0 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 3 of 9 Revision 23 December 2016 Accident Event Time, sec Reactor trip setpoint reached for high neutron
flux 40 Rods begin to fall into core 40.5 1.6 % shutdown is lost (if
dilution continues after trip) ~ 900 Historical Partial Loss of Forced Reactor Coolant Flow
- 1. All loops operating, two pumps coasting
down Coastdown begins Low-flow reactor trip (b) Rods begin to drop
Minimum DNBR occurs 0.0 1.43 2.43
3.9 Historical
Startup of an Inactive Reactor Coolant Loop Initiation of pump startup 0.0 Power reaches high
nuclear flux trip 3.2 Rods begin to drop 3.7 Minimum DNBR occurs 4
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 4 of 9 Revision 23 December 2016 Accident Event Time, sec Loss of External Electrical Load-DNBR
- 1. With pressurizer control (BOL) Loss of electrical load 0.0 High pressurizer pressure reactor trip setpoint
reached 11.9 Initiation of steam release
from steam generator
safety valves 12.0 Rods begin to drop 13.9 Peak pressurizer pressure
occurs 14.5 Minimum DNBR occurs 15 2. With pressurizer control (EOL) Loss of electrical load 0.0 Peak pressurizer pressure
occurs 9.0 Initiation of steam release
from steam generator
safety valves 12.5 Low-low steam generator
water level reactor trip 57 Rods begin to drop 59 Minimum DNBR occurs (a) 3. Without pressurizer control (BOL) Loss of electrical load 0.0 High pressurizer pressure
reactor trip point reached 6.1 Rods begin to drop 8.1 Minimum DNBR occurs (a)
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 5 of 9 Revision 23 December 2016 Accident Event Time, sec Peak pressurizer pressure occurs 9.5 Initiation of steam release
from steam generator
safety valves 12.0 4. Without pressurizer control (EOL) Loss of electrical load 0.0 High pressurizer pressure
reactor trip point reached 6 Rods begin to drop 8 Minimum DNBR occurs (a) Peak pressurizer pressure
occurs 8.5 Initiation of steam release
from steam generator
safety valves 12.5 Loss of External Electrical Load-Overpressure (Peak RCS Pressure)
- 1. With no pressurizer control (BOL) Reactor Trip 9.0 PSVs Open 9.1 Peak RCS Pressure 9.5 MSSVs Open 9.8 Peak Secondary Side
Pressure 16.0 Overpressure (Peak
Secondary Side
Pressure) 2. With pressurizer control (BOL) PORVs Open 3.6 MSSVs Open 9.1 Reactor Trip 15.1 PSVs Open 16.3 Peak RCS Pressure 16.5 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 6 of 9 Revision 23 December 2016 Accident Event Time, sec Peak Secondary Side pressure 20.0 W/Power W/O Power Loss of Normal Feedwater and Loss of Offsite Power to the Station Auxiliaries Main feedwater flow stops 0.0 0.0 Low-low steam generator water level reactor trip 52.7 54.2 Rods begin to drop 54.7 56.2 Reactor coolant pumps begin
to coast down - 58.2 Four SGs begin to receive
aux feed from both motor-
driven AFW pumps 112.7 114.2 Peak water level in
pressurizer occurs (post-trip) 1294 2030 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 7 of 9 Revision 23 December 2016 Accident Event Time, sec Excessive Feedwater at Full Load One main feedwater regulating valve fails full open 0.0 High-high steam generator
water level is reached 33.6 Turbine trip signal (from high-
high steam generator level, turbine stop valve fully closed
0.1second later 36.0 Reactor trip occurs from
turbine trip (rod motion
begins) 38.1 Minimum DNBR occurs 39.0 Initial pressurizer PORV
opens (all PORVs closed 1.3
seconds later) 39.7 Main feedwater isolation valves closed in all four loops (from high-high steam
generator level) 99.6 Excessive Load Increase
- 1. Manual reactor control (BOL
minimum moderator
feedback) 10% step load increase
Equilibrium conditions
reached (approximate times
only) 0.0 240 2. Manual reactor control (EOL
maximum moderator
feedback) 10% step load increase
Equilibrium conditions
reached (approximate times
only) 0.0 64 3. Automatic reactor control (BOL
minimum moderator
feedback) 10% step load increase
Equilibrium conditions
reached (approximate times
only) 0.0 150 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 8 of 9 Revision 23 December 2016 Accident Event Time, sec
- 4. Automatic reactor control (EOL maximum moderator
feedback) 10% step load increase
Equilibrium conditions
reached (approximate times
only) 0.0 150 Accidental Depressuri-zation of the Reactor Coolant System Inadvertent opening of one
pressurizer safety valve
Overtemperature T reactor trip setpoint reached 0.0 27.5 Rods begin to drop 29.5 Minimum DNBR occurs 29.8 Inadvertent Operation of ECCS During Power Operation - DNBR Charging pumps begin
injecting borated water 0.0 Low-pressure trip point
reached 23 Rods begin to drop 25 Inadvertent Operation of
ECCS During Power
Operation - Pressurizer Overfill Case 1 Reactor Trip/Safety
injection
Pressurizer fills 0
583 PSV opens 624 Last PSV relief 720 Case 2 Reactor Trip/Safety
Injection 0 Pressurizer fills 583 PORV opens 591 50 PORV cycles 1,560 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.2-1 Sheet 9 of 9 Revision 23 December 2016 Accident Event Time, sec Case 3 Reactor Trip/Safety Injection 0 PORV opens 77 Pressurizer fills 864 93 PORV cycles 1,560 (a) DNBR does not decrease below its initial value. (b) Analysis assumed low flow setpoint of 87 percent loop flow. An evaluation concludes that 85 percent loop flow is
acceptable. Refer to Table 15.1-2, footnote (d).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 21 September 2013 TABLE 15.3-1 TIME SEQUENCE OF EVENTS - SMALL BREAK LOCA Unit 1 2-inch 3-inch 4-inch 6-inch Transient Initiated, sec 0 0 0 0 Reactor Trip Signal, sec 43.58 18.32 10.55 5.9 Safety Injection Signal, sec 58 26.8 16.57 8.58 Safety Injection Begins (1), sec 85 53.8 43.57 35.58 Loop Seal Clearing Occurs (2), sec 1197 514 300 110 Top of Core Uncovered (3), sec 1796 941 635 N/A Accumulator Injection Begins, sec N/A 1984 885 385 Top of Core Recovered, sec 6500 3170 2545 N/A RWST Low Level, sec 1709 1689 1664 1640 Unit 2 2-inch 3-inch 4-inch 6-inch Transient Initiated, sec 0 0 0 0 Reactor Trip Signal, sec 44.72 18.78 10.82 6.11 Safety Injection Signal, sec 59.45 27.41 16.68 9 Safety Injection Begins (1), sec 86.45 54.41 43.68 36 Loop Seal Clearing Occurs (2), sec 1360 575 290 120 Top of Core Uncovered (3), sec 3200 722 770 N/A Accumulator Injection Begins, sec N/A 3050 985 400 Top of Core Recovered, sec N/A 3215 1630 N/A RWST Low Level, sec 1708 1690 1666 1641
(1) Safety Injection begins 27.0 seconds (SI delay time) after the safety injection signal is reached.
(2) Loop seal clearing is considered to occur when the broken loop seal vapor flow rate is sustained above 1 lbm/s.
(3) Top of core uncovery time is taken as the time when the core mixture level is sustained below the top of the core elevation.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 19 May 2010 TABLE 15.3-2 FUEL CLADDING RESULTS - SMALL BREAK LOCA Unit 1 2-inch 3-inch 4-inch PCT (°F) 907 1391 1241 PCT Time (s) 2173.3 1891.7 975.8 PCT Elevation (ft) 10.75 11.25 11.00 Burst Time (s)
(1) N/A N/A N/A Burst Elevation (ft)
(1) N/A N/A N/A Maximum Hot Rod Transient ZrO2 (%) 0.01 0.38 0.07 Maximum Hot Rod Transient ZrO2 Elev. (ft) 10.75 11.25 10.75 Hot Rod Average Transient ZrO2 (%) 0.01 0.06 0.01
Unit 2 2-inch 3-inch 4-inch PCT (°F) 814 1288 1004 PCT Time (s) 4838.3 1961.8 1079.2 PCT Elevation (ft) 11.00 11.25 10.75 Burst Time (s)
(1) N/A N/A N/A Burst Elevation (ft)
(1) N/A N/A N/A Maximum Hot Rod Transient ZrO2 (%) 0.01 0.18 0.01 Maximum Hot Rod Transient ZrO2 Elev. (ft) 11.00 11.25 10.75 Hot Rod Average Transient ZrO2 (%) 0 0.03 0.01 (1) Burst was not predicted to occur for any break size.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 15.3-3 TIME SEQUENCE OF EVENTS FOR CONDITION III EVENTS Accident Event Time sec (a)
Complete Loss of
Forced Reactor
Coolant Flow
All loops operating, all pumps coasting
down Coastdown begins
Rod motion begins
Minimum DNBR occurs 0.0 2.85 4.8
a) Event times are Unit 1; Unit 2 is 0.04 seconds later.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 21 September 2013 TABLE 15.4.1-1A UNIT 1 BEST ESTIMATE LARGE BREAK LOCA TIME SEQUENCE OF EVENTS FOR THE REFERENCE TRANSIENT Event Time (sec)
Start of Transient 0.0 Safety Injection Signal 6.0 Accumulator Injection Begins 11.0 End of Blowdown 29.0 Safety Injection Begins 33.0Bottom of Core Recovery 37.0 Accumulator Empty 50.0PCT Occurs 39.0 Hot Rod Quench
>300.0 End of Transient 500.0 DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 15.4.1-1B UNIT 2 BEST ESTIMATE LARGE BREAK SEQUENCE OF EVENTS FOR LIMITING PCT CASE Event Time (sec) Start of Transient 0.0 Safety Injection Signal 6.0 Accumulator Injection Begins 13.0 End of Blowdown 29.0 Safety Injection Begins 33.0 Bottom of Core Recovery 37.0 Accumulator Empty 48.0 PCT Occurs 110.0 Hot Rod Quench 285.0 End of Transient 500.0
DCPP UNITS 1 & 2 FSAR UPDATE Revision 21 September 2013 TABLE 15.4.1-2A UNIT 1 BEST ESTIMATE LARGE BREAK LOCA ANALYSIS RESULTS Component Blowdown Peak First Reflood Peak Second Reflood Peak PCTaverage <1485°F <1621°F <1486°F PCT 95% <1744°F <1900°F <1860°F Maximum Oxidation
<11% Total Oxidation
<0.89%
DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 15.4.1-2B UNIT 2 BEST ESTIMATE LARGE BREAK LOCA ANALYSIS RESULTS Result Criterion 95/95 PCT 1,872°F < 2,200°F 95/95 LMO 1.64% < 17% 95/95 CWO 0.17% < 1% PCT - Peak Cladding Temperature LMO - Local Maximum Oxidation CWO - Core Wide Oxidation
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-3A Sheet 1 of 4 Revision 21 September 2013 UNIT 1 KEY BEST ESTIMATE LARGE BREAK LOCA PARAMETERS AND REFERENCE TRANSIENT ASSUMPTIONS Parameter Reference Transient Uncertainty or Bias 1.0 Plant Physical Description
- a. Dimensions Nominal PCT MOD b. Flow resistance Nominal PCT MOD c. Pressurizer location Opposite broken loop Bounded d. Hot assembly location Under limiting location Bounded e. Hot assembly type 17x17 V5 w/ZIRLO clad Bounded f. SG tube plugging level High (15%)
Bounded (a) 2.0 Plant Initial Operating Conditions 2.1 Reactor Power
- a. Core average linear heat rate No minal - 100% of uprated power (3411 MWt) PCT PD b. Peak linear heat rate (PLHR) Derived from desired Technical Specifications (TS) limit and maximum baseload PCT PD c. Hot rod average linear heat rate (HRFLUX)
Derived from TS FH PCT PD DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-3A Sheet 2 of 4 Revision 21 September 2013 Parameter Reference Transi ent Uncertainty or Bias d. Hot assembly average heat rate HRFLUX/1.04 PCT PD e. Hot assembly peak heat rate PLHR/1.04 PCT PD f. Axial power distribution (PBOT, PMID) Figure 3-2-10 of Reference 60 PCT PD g. Low power region relative power (PLOW) 0.3 Bounded (a) h. Hot assembly burnup BOL Bounded i. Prior operating history Equilibrium decay heat Bounded j. Moderator Temperature Coefficient (MTC) TS Maximum (0) Bounded k. HFP boron 800 ppm Generic 2.2 Fluid Conditions
- a. Tavg Max. nominal Tavg = 577.3°F Nominal is bounded, uncertainty is in PCT IC b. Pressurizer pressure Nominal (2250.0 psia) PCT IC c. Loop flow 85000 gpm PCT MOD (b) d. T UH Best Estimate 0 e. Pressurizer level Nominal (1080 ft
- 3) 0 f. Accumulator temperature Nominal (102.5
°F) PCT IC g. Accumulator pressure Nominal (636.2 psia) PCT IC DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-3A Sheet 3 of 4 Revision 21 September 2013 Parameter Reference Transi ent Uncertainty or Bias h. Accumulator liquid volume Nominal (850 ft
- 3) PCT IC i. Accumulator line resistance Nominal PCT IC j. Accumulator boron Minimum Bounded 3.0 Accident Boundary Conditions
- a. Break location Cold leg Bounded b. Break type Guillotine PCT MOD c. Break size Nominal (cold leg area) PCT MOD d. Offsite power Off (RCS pumps tripped)
Bounded (a) e. Safety injection flow Minimum Bounded f. Safety injection temperature Nominal (68
°F) PCT IC g. Safety injection delay Max delay (27.0 sec, with loss of offsite power) Bounded h. Containment pressure Minimum based on WC/T M&E Bounded i. Single failure ECCS: Loss of 1 SI train Bounded j. Control rod drop time No control rods Bounded 4.0 Model Parameters
- a. Critical Flow Nominal (as coded) PCT MOD DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-3A Sheet 4 of 4 Revision 21 September 2013 Parameter Reference Transi ent Uncertainty or Bias b. Resistance uncertainties in broken loop Nominal (as coded) PCT MOD c. Initial stored energy/fuel rod behavior Nominal (as coded) PCT MOD d. Core heat transfer Nominal (as coded) PCT MOD e. Delivery and bypassing of ECC No minal (as coded)
Conservative f. Steam binding/entrainment Nomi nal (as coded)
Conservative g. Noncondensable gases/accumulator nitrogen Nominal (as coded) Conservative h. Condensation Nominal (as coded) PCT MOD (a) Confirmed by plant-specific analysis.
(b) Assumed to be result of loop resistance uncertainity.
Notes: 1. PCT MOD indicates this uncertainty is par t of code and global model uncertainty.
- 2. PCTPD indicates this uncertainty is part of power distribution uncertainty.
- 3. PCTIC indicates this uncertainty is part of initial condition uncertainty.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-3B Sheet 1 of 3 Revision 21 September 2013 UNIT 2 KEY BEST ESTIMATE LARGE BREAK LOCA PARAM ETERS AND INITIAL TRANSIENT ASSUMPTIONS Parameter Initial Transient Range/Uncertainty 1.0 Plant Physical Description
- a. Dimensions Nominal Sampled b. Flow resistance Nominal Sampled c. Pressurizer location Opposite broken loop Bounded d. Hot assembly location Under limiting location Bounded e. Hot assembly type 17x17 V5 + with ZIRLO TM cladding, Non-IFBA Bounded f. Steam generator tube plugging level High (15%)
Bounded(a) 2.0 Plant Initial Operating Conditions 2.1 Reactor Power
- a. Core average linear heat rate (AFLUX) Nominal - Based on 100% thermal power (3468 MWt)
Sampled b. Hot rod peak linear heat rate (PLHR)
Derived from desired Technical Specification limit F Q = 2.7 and maximum baseload F Q = 2.1 Sampled c. Hot rod average linear heat rate (HRFLUX)
Derived from Technical Specification FH = 1.7 Sampled d. Hot assembly average heat rate (HAFLUX)
HRFLUX/1.04 Sampled e. Hot assembly peak heat rate (HAPHR)
PLHR/1.04 Sampled f. Axial power distribution (PBOT, PMID)
Figure 15.4.1-15B Sampled g. Low power region relative power (PLOW) 0.3 Bounded(a) h. Cycle burnup
~2000 MWD/MTU Sampled i. Prior operating history Equilibrium decay heat Bounded DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-3B Sheet 2 of 3 Revision 21 September 2013 Parameter Initial Transient Range/Uncertainty 2.0 Plant Initial Operating Conditions (continued)
- j. Moderator temperature coefficient Technical Specification Maximum (0) Bounded k. HFP boron 800 ppm Generic 2.2 Fluid Conditions
- a. Tavg High Nominal Tavg = 577.6°F Bounded(a), Sampled b. Pressurizer pressure Nominal (2250.0 psia)
Sampled c. Loop flow 85,000 gpm Bounded d. Upper head fluid temperature T cold 0 e. Pressurizer level Nominal 0 f. Accumulator temperature Nominal (102.5°F)
Sampled g. Accumulator pressure Nominal (636.2 psia)
Sampled h. Accumulator liquid volume Nominal (850 ft
- 3) Sampled i. Accumulator line resistance Nominal Sampled j. Accumulator boron Minimum (2200 ppm)
Bounded 3.0 Accident Boundary Conditions
- a. Break location Cold leg Bounded b. Break type Guillotine (DECLG)
Sampled c. Break size Nominal (cold leg area)
Sampled d. Offsite power Loss of offsite power Bounded(a) e. Safety injection flow Minimum Bounded f. Safety injection temperature Nominal (68°F)
Sampled g. Safety injection delay Maximum delay (27.0 sec, with loss of offsite power)
Bounded DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-3B Sheet 3 of 3 Revision 21 September 2013 Parameter Initial Transient Range/Uncertainty 3.0 Accident Boundary Conditions (continued)
- h. Containment pressure Bounded - Lower (conservative) than pressure curve shown in Figure 15.4.1-14B.
Bounded i. Single failure ECCS: Loss of one safety injection train; Containment pressure: all trains
operational Bounded j. Control rod drop time No control rods Bounded 4.0 Model Parameters
- a. Critical flow Nominal (CD = 1.0)
Sampled b. Resistance uncertainties in broken loop Nominal (as coded)
Sampled c. Initial stored energy/fuel rod behavior Nominal (as coded)
Sampled d. Core heat transfer Nominal (as coded)
Sampled e. Delivery and bypassing of emergency core coolant Nominal (as coded)
Conservative f. Steam binding/entrainment Nominal (as coded)
Conservative g. Noncondensable gases/accumulator nitrogen Nominal (as coded)
Conservative
- h. Condensation Nominal (as coded)
Sampled (a) Per Confirmatory Study re sults (Section 15.4.1.1.2.5)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 21 September 2013 TABLE 15.4.1-4A UNIT 1 SAMPLE OF BEST ESTIMATE SENSITIVITY ANALYSIS RESULTS FOR ORIGINAL ANALYSIS (Reference 60)
Type of Study Parameter Varied Value PCT Results (°F) Blowdown Reflood 1 Reflood 2 Reference Transient See Table 15.4-3 1600 1852 1984 Confirmatory Cases Steam Generator Tube Plugging 0% 1569 1798 1878 Offsite Power Assumption Available 1500 1685 1781 Normalized Power in Outer Assemblies 0.8 1611 1805 1939 Vessel Average Temperature 565°F 1573 1843 1871 Initial Accumulator +50 ft 3 1601 1856 1823 Condition Volume 50 ft 3 1599 1863 2182 Global Models DECLG, CD 1.0 1600 1852 1984 SPLIT, CD 1.4 - 1596 1637 1.6 - 1784 1799 1.8 - 1790 1738 2.0 - 1765 1804
DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 15.4.1-4B UNIT 2 RESULTS FROM CONFIRMATORY STUDIES Transient Description PCT (°F) Reflood Initial Transient (High Tavg, High SGTP, Low PLOW, LOOP) 1595 SGTP Confirmatory Transient (High Tavg, Low SGTP, Low PLOW, LOOP) 1576 Tavg, Confirmatory Transient (Low Tavg, High SGTP, Low PLOW, LOOP) 1536 PLOW Confirmatory Transient (High Tavg , High SGTP, High PLOW, LOOP) 1657 LOOP Confirmatory Transient (High Tavg, High SGTP, Low PLOW, no-LOOP) 1425 Reference Transient (High Tavg , High SGTP, High PLOW, LOOP) 1657
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-5A Sheet 1 of 2 Revision 18 October 2008 UNIT 1 CONTAINMENT BACK PRESSURE ANALYSIS INPUT PARAMETERS USED FOR BEST ESTIMATE LOCA ANALYSIS Net Free Volume, cu ft 2,630,000 Initial Conditions Pressure, psia 14.7 Temperature, °F 85 RWST temperature, °F 35 Service water temperature, °F 45 Outside temperature, °F 33 Spray System Number of pumps operating 2 Runout flowrate per pump, gpm 3400 Actuation time, sec 40.8 Safeguards Fan Coolers Number of fan coolers operating 5
Fastest post-accident initiation of fan coolers, sec 0 Structural Heat Sinks Thickness, in.
Area, ft 2 42.0 concrete 65,749 12.0 concrete 24,054 24.0 concrete 14,313 12.0 concrete 48,183 12.0 concrete 15,725 108.0 concrete 20,493 30.0 concrete 33,867 1.68 steel 8,525 1.92 steel 4,015 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-5A Sheet 2 of 2 Revision 18 October 2008 Structural Heat Sinks (continued)
Thickness, in.
Area, ft 2 6.99 steel 1,771 0.5656 steel 43,396 0.088 steel 24,090 0.22 steel 10,597 0.088 steel 8,470 0.102 steel 23,438 0.071 steel 20,266 0.708 steel 26,050 0.127 steel 33,000 0.773 steel 11,004 0.375 steel 99,616 1.596 steel 1,530 1.098 steel 21,022 0.745 steel 6,755 0.96 steel 792 0.144 stainless steel 9,737 0.654 stainless steel 943 0.642 steel 1,373 3.0 steel 575 0.75 steel 17,542
DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 15.4.1-5B UNIT 2 CONTAINMENT BACK PRESSURE ANALYSIS INPUT PARAMETERS USED FOR BEST ESTIMATE LBLOCA ANALYSIS Net Free Volume 2,630,000 ft 3 Initial Conditions Pressure 14.7 psia Temperature 85.0°F RWST temperature 35.0°F Service water temperature 48.0°F Temperature outside containment 33.0°F Initial spray temperature 35.0°F Spray System Number of spray pumps operating 2 Post-accident spray system initiation delay 40.8 sec Maximum spray system flow from all pumps 6,800 gal/min.
Containment Fan Coolers Post-accident initiation fan coolers 0.0 sec (a) Number of fan coolers operating 5 (a) Bounds delay with and without LOOP
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-7A Sheet 1 of 3 Revision 22 May 2015 UNIT 1 PLANT OPERATING RANGE ALLOWED BY THE BEST-ESTIMATE LARGE BREAK LOCA ANALYSIS Parameter Operating Range 1.0 Plant Physical Description
- a. Dimensions No in-board assembly grid deformation assumed due to LOCA+DDE or LOCA + Hosgri b. Flow resistance N/A c. Pressurizer location N/A d. Hot assembly location Anywhere in core e. Hot assembly type Fresh 17X17 V5, ZIRLO, or Zircaloy cladding, 1.5X IFBA or non-IFBA f. SG tube plugging level 15% g. Fuel assembly type Vantage 5, ZIRLO, or Zircaloy cladding, 1.5X IFBA or non-IFBA 2.0 Plant Initial Operating Conditions 2.1 Reactor Power
- a. Core average linear heat rate Core power 102% of 3411 MWt
- b. Peak linear heat rate F Q 2.7 c. Hot rod average linear heat rate FH 1.7 d. Hot assembly average linear heat rate P HA 1.57 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-7A Sheet 2 of 3 Revision 22 May 2015 Parameter Operating Range e. Hot assembly peak linear heat rate F QHA 2.7/1.04
- f. Axial power distribution (PBOT, PMID) Figure 15.4.1-15A
- g. Low power region relative power (PLOW) 0.3 PLOW 0.8 h. Hot assembly burnup 75,000 MWD/MTU, lead rod
- i. Prior operating history All normal operating histories j. MTC 0 at HFP
- k. HFP boron Normal letdown 2.2 Fluid Conditions
- a. Tavg 560.0 T ave 582.3°F b. Pressurizer pressure 2190 P RCS 2310 psia
- c. Loop flow 85,000 gpm/loop
- d. T UH Current upper internals
- e. Pressurizer level Normal level, automatic control f. Accumulator temperature 85 accumulator temperature 120°F g. Accumulator pressure 579 P ACC 664 psig
- h. Accumulator volume 814 V acc 886 ft 3 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-7A Sheet 3 of 3 Revision 22 May 2015 Parameter Operating Range i. Accumulator fL/D Current line configuration
- j. Minimum accumulator boron 2200 ppm 3.0 Accident Boundary Conditions
- a. Break location N/A b. Break type N/A c. Break size N/A d. Offsite power Available or LOOP e. Safety injection flow Figure 15.4.1-13A f. Safety injection temperature 46 SI Temperature 90°F g. Safety injection delay 17 seconds (with offsite power) 27 seconds (with LOOP) h. Containment pressure Bounded - see Figure 15.4.1-14A i. Single failure Loss of one train j. Control rod drop time N/A DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-7B Sheet 1 of 2 Revision 22 May 2015 UNIT 2 PLANT OPERATING RANGE ALLOWED BY THE BEST-ESTIMATE LARGE BREAK LOCA ANALYSIS Parameter Operating Range 1.0 Plant Physical Description a) Dimensions No in-board assembly grid deformation during LOCA+DDE or LOCA +
Hosgri b) Flow resistance N/A c) Pressurizer location N/A d) Hot assembly location Anywhere in core interior (149 locations)(a) e) Hot assembly type Fresh 17x17 V5+ fuel with ZIRLO TM cladding f) Steam generator tube plugging level 15% g) Fuel assembly type 17x17 V5+ fuel with ZIRLO TM cladding, non-IFBA or IFBA 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Core average linear heat rate b) Peak linear heat rate F Q c) Hot rod average linear heat rate FH d) Hot assembly average linear heat rate HA P < 1.7/1.04 e) Hot assembly peak linear heat rate F QHA < 2.7/1.04 f) Axial power distribution (PBOT, PMID) See Figure 15.4.1-15B.
g) Low power region relative power (PLOW) h) Hot assembly burnup (a) i) Prior operating history All normal operating histories j) Moderator temperature coefficient k) HFP boron (minimum) 800 ppm (at BOL) 2.2 Fluid Conditions a) T avg avg DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4.1-7B Sheet 2 of 2 Revision 22 May 2015 Parameter Operating Range b) Pressurizer pressure RCS c) Loop flow d) T UH Converted upper internals, T COLD UH e) Pressurizer level Nominal level, automatic control f) Accumulator temperature ACC g) Accumulator pressure ACC h) Accumulator liquid volume 814 ft 3 ACC 3 i) Accumulator fL/D Current line configuration j) Minimum accumulator boron 3.0 Accident Boundary Conditions a) Break location N/A b) Break type N/A c) Break size N/A d) Offsite power Available or LOOP e) Safety injection flow See Figure 15.4.1-13B. f) Safety injection temperature g) Safety injection delay h) Containment pressure See Figure 15.4.1-14B and raw data in Table 15.4.1-5B. i) Single failure All trains operable (b) j) Control rod drop time N/A (a) 44 peripheral locations will not physically be lead power assembly. (b) Analysis considers loss of one train of pumped ECCS.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4-8 Sheet 1 of 5 Revision 23 December 2016 TIME SEQUENCE OF EVENTS FOR MAJOR SECONDARY SYSTEM PIPE RUPTURES Accident Event Time, sec Steam Line Rupture @ HZP
- 1. With Offsite Power Available Main steam line ruptures 0.0 Low steam line pressure setpoint reached 0.6 SIS flow begins(maximum flow assumed) 2.6 Steam line isolation occurs 8.6 Criticality attained 36.5 Borated water from the RWST reaches the
core ~40 Main feedwater isolation occurs 64.6 Accumulators inject 79.0 Peak core heat flux, minimum DNBR occurs 90.5
- 2. Without Offsite Power
Available Main steam line ruptures 0.0 Low steam line pressure setpoint reached 0.6 SIS flow begins (maximum flow assumed) 2.6 RCPs begin to coast down 3.0 Steam line isolation occurs 8.6 Criticality attained 44.4 Borated water from the RWST reaches the
core ~50
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4-8 Sheet 2 of 5 Revision 23 December 2016 Accident Event Time, sec Main feedwater isolation occurs 64.6 Peak core heat flux, minimum DNBR occurs 123.4 Accumulators inject 129.7 Rupture of Main Feedwater Pipe (Offsite Power Available)
Feedline rupture occurs 20 Low-low steam generator level reactor trip
setpoint reached in affected steam
generator 32 Rods begin to drop 34 Auxiliary feedwater is started 623 Pressurizer liquid water relief begins if
operator action is not assumed 2053 Total RCS heat generation (decay heat +
pump heat) decreases to auxiliary
feedwater heat removal capability 5900 Rupture of Main Feedwater
Pipe (Offsite Power
Unavailable)
Feedline rupture occurs 20 Low-low steam generator level reactor trip
setpoint reached in affected steam
generator 32 Rods begin to drop 34 Reactor coolant pump coastdown 36 Auxiliary feedwater is started 632 Peak pressurizer level after initial outsurge
reached 2091
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4-8 Sheet 3 of 5 Revision 23 December 2016 Accident Event Time, sec Total RCS heat generation decreases to auxiliary feedwater heat removal capability 2200 Steam Line Rupture at Power
(0.49 ft2)
Steam line ruptures
0.0 Peak core heat flux occurs 53.1 Rupture of a Main Feedwater Pipe for Pressurizer Filling Feedwater line rupture occurs 0.0 (Unblock Pressurizer PORV)
Low-low SG water level reactor trip setpoint (0% NRS) reached in faulted SG 15.9 Rods begin to drop 17.9 Turbine trip occurs 18.4 Steam line check valve closes in loop with faulted SG 18.5 Reactor coolant pumps begin to coast down (from loss of offsite power) 19.9 Low steam line pressure setpoint reached in loop with faulted SG 26.7 Safety injection actuation signal generated 28.7 Safety injection flow initation occurs 28.8 Steam line isolation occurs on low steam line pressure safety injection signal 34.7 Main steam safety valve relief begins 36.1 PSV steam relief begins 56.0 AFW flow initiation (390 gpm from a motor-driven AFW pump) occurs to intact SGs not connected to the faulted SG 75.9 Pressurizer reaches a water-solid condition 501.0 PSV water relief begins (maximum time for operator action to ensure a PORV is available) 518.8 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4-8 Sheet 4 of 5 Revision 23 December 2016 Accident Event Time, sec Operator action to isolate faulted SG to direct all available AFW flow to intact SGs 615.9 AFW flow initiation (195 gpm from the turbine-driven AFW pump and 195 gpm from the other motor-driven AFW pump) occurs to intact SG connected to the faulted SG 615.9 AFW flow addition (390 gpm from the turbine-driven AFW pump) occurs to intact SGs not connected to the faulted SG 615.9 Rupture of a Main Feedwater Pipe for Pressurizer Filling Feedline rupture occurs 0.0 (Isolate Charging Flow and Stop RCP Seal Injection Flow)
Pressurizer backup heater actuation on level deviation 13.5 Low-low SG water level reactor trip setpoint (0% NRS) reached in faulted SG 15.9 Rods begin to drop 17.9 Turbine trip occurs 18.4 Steam line check valve closes in loop with faulted SG 18.5 Reactor coolant pumps begin to coast down (from manual trip) 19.9 Pressurizer PORV steam relief begins 20.9 Low steam line pressure setpoint reached in loop with faulted SG 26.8 Safety injection actuation signal generated 28.8 Safety injection flow initation occurs 28.9 Steam line isolation occurs on low steam line pressure safety injection signal 34.8 Main steam safety valve relief begins 40.1 AFW flow initiation (390 gpm from a motor-driven AFW pump) occurs to intact SGs not connected to the faulted SG 75.9 Pressurizer reaches a water-solid condition 408.5 Operator action to isolate faulted SG to direct all available AFW flow to intact SGs 615.9 AFW flow initiation (195 gpm from the 615.9 DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.4-8 Sheet 5 of 5 Revision 23 December 2016 Accident Event Time, sec turbine-driven AFW pump and 195 gpm from the other motor-driven AFW pump) occurs to intact SG connected to the faulted SG AFW flow addition (390 gpm from the turbine-driven AFW pump) occurs to intact SGs not connected to the faulted SG 615.9 Operator action to isolate charging/SI flow 1516.0 Operator action to stop RCP seal injection flow 2715.9 Steam bubble forms again in pressurizer 6723.0 Maximum number of PORV cycles reached (capacity of backup nitrogen accumulators depleted) 7137.6 DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 15.4-10
SUMMARY
OF RESULTS FOR LOCKED ROTOR TRANSIENT 4 Loops Operating Initially 1 Locked Rotor Maximum RCS pressure, psia 2729 (1) Maximum clad average temperature, °F core hot spot 1963 Amount of Zr - H 2 O at core hot spot, % by weight 0.53%
(1) The locked rotor transient peak pressure of 2729 psia includes a conservative penalty of 41 psi, determined by Westinghouse in an evaluation to address Nuclear Safety Advisory Letter NSAL-09-2. Locked Rotor Analysis for Reactor Coolant System Overpressure. May 7, 2009.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 15 September 2003 TABLE 15.4-11 TYPICAL PARAMETERS USED IN THE VAN TAGE 5 RELOAD ANALYSIS OF THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT Time in Life Beginning Beginning End End Generic Vantage 5 Rel oad Analysis Values Power level, % 102 0.0 102 0.0 Ejected rod worth, % k 0.20 0.785 0.21 0.85 Delayed neutron fraction, % 0.55 0.55 0.44 0.44 Feedback reactivity weighting 1.30 2.071 1.30 3.55 Doppler - only power defect, pcm -955 -954 -829 -788 Trip reactivity, % k 4 2 4 2 F q before rod ejection 2.60 - 2.60 - F q after rod ejection 6.70 13 6.50 21.50 Number of operating pumps 4 2 4 2 Generic Vantage 5 Rel oad Analysis Results Maximum fuel pellet average temperature, °F 4154 3509 3812 3408 Maximum fuel center temperature, °F >4900 (a) 4025 >4800 (a) 3849 Maximum cladding average temperature, °F 2434 2660 2218 2632 Maximum fuel stored energy, cal/gm 183 149 165 144 Reload Analysis Ev aluation Values Power level, % 102 0.0 102 0.0
Ejected rod worth, % k 0.20 0.785 0.21 0.83 Delayed neutron fraction, % 0.55 0.55 0.44 0.44 Feedback reactivity weighting 1.30 2.071 1.30 3.55 Doppler - only power defect, pcm -995 -954 -829 -788 Trip reactivity, % k 4 2 4 2 F q before rod ejection 2.60 - 2.60 - F q after rod ejection 6.70 13 6.50 22.50 Number of operating pumps 4 2 4 2 (a) Less than 10% fuel pellet melt (at hot spot)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 19 May 2010 TABLE 15.4-12 OPERATOR ACTION TIMES FOR DESIGN BASIS SGTR ANALYSIS Action Time (min)
Identify and isolate ruptured SG 10 min or RETRAN-02W calculated time to reach 38% narrow range level in the ruptured SG, whichever is
longer Operator action time to initiate
cooldown 5 Cooldown Calculated by RETRAN-02W
Operator action time to initiate
depressurization 4
Depressurization Calculated by RETRAN-02W
Operator action time to initiate
SI termination 2
SI termination and pressure
equalization Calculated time for SI termination and
equalization of RCS and ruptured SG
pressures
DCPP UNITS 1 & 2 FSAR UPDATE Revision 20 November 2011 TABLE 15.4-13A TIMED SEQUENCE OF EVENTS - SGTR MTO ANALYSIS Event Time (sec)
SG Tube Rupture 100
Reactor Trip 274
SI Actuated 380
Turbine Driven AFW Pump Flow Isolated 700
Ruptured SG Steamline Isolation 700
Ruptured SG MDAFW Pump Flow Isolated 820
RCS Cooldown Initiated 1120
RCS Cooldown Terminated 1706
RCS Depressurization Initiated 1946
RCS Depressurization Terminated 2072
SI Terminated 2192
Break Flow Terminated 3475
DCPP UNITS 1 & 2 FSAR UPDATE Revision 20 November 2011 TABLE 15.4-13B TIMED SEQUENCE OF EVENTS - SGTR DOSE ANALYSIS Event Time (sec)
SG Tube Rupture 100
Reactor Trip 279
SI Actuated 315
Ruptured SG Isolated 953
Ruptured SG PORV Fails Open 953
Ruptured SG PORV Block Valve Closed 2753
RCS Cooldown Initiated 3053
RCS Cooldown Terminated 4424
RCS Depressurization Initiated 4664
RCS Depressurization Terminated 4839
SI Terminated 4959
Break Flow Terminated 5972
DCPP UNITS 1 & 2 FSAR UPDATE Revision 21 September 2013 TABLE 15.4-14 MASS RELEASE RESULTS - SGTR DOSE INPUT ANALYSIS 0 - 2 Hrs, lbm 2 - 8 Hrs, lbm Ruptured SG
- Condenser 294,500 0 - Atmosphere 140,200 27,000
- Feedwater 288,700 0 Intact SGs
- Condenser 878,100 0
- Atmosphere 367,100 922,600
- Feedwater 1,476,800 961,700
Break Flow 262,200 0
Flashed Break Flow 18,150 0 Note: The 0-2 hour releases to the condenser and feedwater flows include 100 seconds of steady state operation.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-1 REACTOR COOLANT FISSION AND CORROSION PRODUCT ACTIVITIES DURING STEADY STATE OPERATION AND PLANT SHUTDOWN OPERATION Operating PWR Plant Diablo Canyon - Design Basis Case (HISTORICAL)
Isotope Measured Activity Before Shutdown, mCi/gm Measured Peak Shutdown Activity, mCi/gm Calculated Activity Before Shutdown, mCi/gm Expected Peak Shutdown Activity, mCi/gm I -131 0.83 14.9 2.45 43.9 Xe-133 127.00 65.0 (a) 255.8 130.9 (a) Cs-134 1.29 1.7 0.198 0.26 Cs-137 1.67 2.14 0.31 0.39 Ce-144 0.00068 0.0058 0.00034 0.0029 Sr-89 0.0033 0.40 0.0026 0.32 Sr-90 0.00057 0.013 0.00013 0.003 Co-58 --- 0.95 0.026 1.04
(a) Activity reduced from steady state level by approximately 1 day of system degasification prior to plant shutdown.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-2 (HISTORICAL)
RESULTS OF STUDY OF EFFECTS OF PLUTONIUM ON ACCIDENT DOSES
Type of Accident Change in 30-day Thyroid Dose, %
Change in 30-day Whole Body Dose, %
Change in 2-hour Thyroid Dose, %
Change in 2-hour Whole Body Dose, %
Release from gas decay
tank 0 -4 0 -4 Fuel handling accident +6
-3 0 0 Loss of reactor primary
coolant - large break +6 -3 +5 -7 Steam generator tube
rupture accident +6 -2 +4 -2 Steam line rupture
accident +5 -2 +5 -2
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 15.5-3 DESIGN BASIS POST-ACCIDENT ATMOSPHERIC DILUTION FACTORS (SEC/M
- 3) Distance from Release Point, meters (a) Period, hrs 800 1200 2000 4000 7000 10,000 20,000 0-8 5.29x10
-4 3.40x10-4 1.87x10-5 7.78x10-5 3.59x10-5 2.20x10-5 8.85x10-6 8-24 2.15x10
-4 1.10x10-4 5.00x10-5 1.75x10-5 7.50x10-6 4.75x10-6 1.75x10-6 24-96 7.70x10
-5 3.90x10-5 1.75x10-5 5.70x10-6 2.50x10-6 1.54x10-6 5.50x10-7 96-720 1.75x10
-5 8.20x10-6 3.70x10-6 1.35x10-6 5.20x10-7 3.40x10-7 1.20x10-7
(a) Minimum exclusion area boundary radius is 0.5 miles (approximately 800 m). Radius of low population zone is 6 miles (approximately 10,000 m).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 23 December 2016 TABLE 15.5-4 EXPECTED POST-ACCIDENT ATMOSPHERIC DILUTION FACTORS (SEC/M
- 3) Distance from Release Point, meters (a) Period, hrs 800 1200 2000 4000 7000 10,000 20,000 0-8 5.29x10-5 3.40x10-5 1.87x10-5 7.78x10-5 3.59x10-6 2.20x10-6 8.85x10-7 8-24 2.15x10
-5 1.40x10-5 5.00x10-6 1.75x10-6 7.50x10-7 4.75x10-7 1.75x10-7 24-96 7.70x10
-6 3.90x10-6 1.75x10-6 5.70x10-7 2.50x10-7 1.54x10-7 5.50x10-8 96-720 1.75x10
-6 8.20x10-7 3.70x10-7 1.35x10-7 5.20x10-8 3.40x10-8 1.20x10-8 (a) Minimum exclusion area boundary radius is 0.5 miles (approximately 800 m). Radius of low population zone is 6 miles (approximately 10,000 m).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-5 ATMOSPHERIC DILUTION FACTORS /Q x 10 8 sec-m-3 Onshore Sector Sector Midpoint Downwind Distance, miles Midpoint
Directions 5 15 25 35 45 55 SSE 1.61 0.54 0.32 0.23 0.18 0.15 S 1.44 0.48 0.29 0.21 0.16 0.13 SSW 0.79 0.26 0.16 0.11 0.09 0.07 SW 0.54 0.18 0.11 0.08 0.06 0.05 WSW 0.65 0.22 0.13 0.09 0.07 0.06 W 1.08 0.36 0.22 0.15 0.12 0.10 WNW 1.19 0.40 0.24 0.17 0.13 0.11 NW 5.39 1.80 1.08 0.77 0.60 0.49 NNW 1.94 0.65 0.39 0.28 0.22 0.18
Atmospheric Dilution Factors /Q x 10 6 sec-m-3 Downwind Distance, meters Direction 800 1200 2000 4000 7000 10,000 20,000 SE 0.75 0.47 0.19 0.087 0.050 0.035 0.018
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-6 ASSUMED ONSITE ATMOSPHERIC DILUTION FACTORS (SEC/M
- 3) FOR THE CONTROL ROOM Base /Q (a) Modifying Factors Final Period, hrs.
Sec/m 3 f 1 f 2 f 3 f 4 f 5 f 6 Q (a)
A. For The Pressurization Case
0-8 1.084x10
-3 1 1 .2 1 .5 .65 7.05x10
-5 8-24 1.084x10
-3 .83 .92 .2 1 .5 .65 5.38x10-5 24-96 1.084x10
-3 .66 .84 .2 1 .5 .65 3.91x10
-5 96-720 1.084x10
-3 .48 .67 .2 1 .5 .65 2.27x10
-5 B. For The Infiltration Case
0-8 3.01x10
-3 1 1 .2 1 .5 .65 1.96x10
-4 8-24 3.01x10
-3 .83 .92 .2 1 .5 .65 1.49x10
-4 24-95 3.01x10
-3 .66 .84 .2 1 .5 .65 1.08x10
-4 96-720 3.01x10
-3 .48 .67 .2 1 .5 .65 6.29x10
-5 (a) The /Q calculated above do not account for credit for dual pressurization inlet and occupancy factors.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-7 BREATHING RATES (a) ASSUMED IN ANALYSIS Design Basis Case Expected Case Period Offsite Onsite Offsite Onsite 0-8 hrs 3.47x10
-4 3.47x10-4 2.32x10-4 3.47x10-4 8-24 hrs 1.75x10
-4 3.47x10-4 2.32x10-4 3.47x10-4 1-30 days 2.32x10
-4 3.47x10-4 2.32x10-4 3.47x10-4 (a) All breathing rates are expressed in m 3/sec. Values taken from Reference 8.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-8 POPULATION DISTRIBUTION Onshore Sector Sector Midpoint Downwind Distance, miles Total Midpoint Sector
Directions 5 15 25 35 45 55 Population SSE 1,014 4,727 2,700 5,433 1,567 697 16,138 S 1,000 4,666 2,000 4,234 466 466 12,832 SSW 1,367 20,334 7,000 4,933 1,167 1,100 35,901 SW 366 15,666 5,000 700 700 634 23,066 WSW 840 26,000 6,600 1,767 1,433 1,533 38,173 W 474 10,334 1,600 1,066 734 900 15,108 WNW 1,843 20,033 22,933 19,734 16,066 6,900 87,509 NW 0 9,700 21,334 18,666 15,334 6,000 71,034 NNW 0 0 21,333 22,267 19,133 6,500 69,233
Total Radial Population 6,904 111,460 90,500 78,800 56,600 24,730 368,944
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-9
SUMMARY
OF OFFSITE DOSES FROM LOSS OF ELECTRICAL LOAD Thyroid Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case 0.028 0.0065 Expected case 5.2 x 10-6 8.7 x 10-7 Whole Body Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 2.3 x 10-3 2.3 x 10-4 Expected case 7.2 x 10-7 6.9 x 10-8 Population Doses, man-rem Design basis case 0.15 Expected case 3.8 x 10-5
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-10 Sheet 1 of 1 Revision 22 May 2015
SUMMARY
OF OFFSITE DOSES FROM A SMALL LOSS-OF-COOLANT ACCIDENT NO FUEL DAMAGE Thyroid Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case 2.0 x 10-4 2.7 x 10-5 Expected case 9.0 x 10-7 1.2 x 10-7 Whole Body Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 1.8 x 10-4 5.4 x 10-5 Expected case 4.4 x 10-6 1.4 x 10-6 Population Doses, man-rem Design basis case 0.36 Expected case 0.013 DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-11
SUMMARY
OF OFFSITE DOSES FROM AN UNDERFREQUENCY ACCIDENT Thyroid Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case 0.021 0.0066 Expected case 4.0 x 10-6 1.2 x 10-6 Whole Body Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 0.0018 2.2 x 10-4 Expected case 5.3 x 10-7 6.6 x 10-8 Population Doses, man-rem Design basis case 0.15 Expected case 4.3 x 10-5
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-12
SUMMARY
OF OFFSITE DOSES FROM A SINGLE ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL Thyroid Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case 0.12 0.043 Expected case 9.5 x 10-5 3.4 x 10-5 Whole Body Doses, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 6.1 x 10-3 6.7 x 10-4 Expected case 6.5 x 10-6 6.9 x 10-7 Population Doses, man-rem Design basis case 0.42 Expected case 4.3 x 10-4
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-13 CALCULATED ACTIVITY RELEASES FROM LOCA - EXPECTED CASE (CURIES)
Nuclide 0-2 hr 2-8 hr 8-24 hr 24-96 hr 4-30 Days I-131 0.4105E-01 0.0 0.0 0.0 0.0 I-132 0.6675E-02 0.0 0.0 0.0 0.0 I-133 0.3110E-01 0.0 0.0 0.0 0.0 I-134 0.7392E-02 0.0 0.0 0.0 0.0 I-135 0.1629E-01 0.0 0.0 0.0 0.0 I-131ORG 0.1424E-01 0.3356E-01 0.4704E-01 0.1384E-01 0.1660E-03 I-132ORG 0.1780E-02 0.1556E-02 0.2210E-03 0.4311E-06 0.6136E-17 I-133ORG 0.1049E-01 0.2211E-01 0.2334E-01 0.3542E-02 0.5058E-05 I-134ORG 0.1314E-02 0.2862E-03 0.1670E-05 0.9077E-12 0.0 I-135ORG 0.5139E-02 0.8365E-02 0.4731E-02 0.1933E-03 0.1730E-08 I-131PAR 0.0 0.0 0.0 0.0 0.0 I-132PAR 0.0 0.0 0.0 0.0 0.0 I-133PAR 0.0 0.0 0.0 0.0 0.0 I-134PAR 0.0 0.0 0.0 0.0 0.0 I-135PAR 0.0 0.0 0.0 0.0 0.0 Kr-83M 0.3823E 00 0.3085E 00 0.3684E-01 0.4757E-04 0.1063E-15 Kr-65 0.5356E 01 0.1598E 02 0.4257E 02 0.9571E 02 0.8243E 03 Kr-85M 0.1750E 01 0.2889E 01 0.1689E 01 0.7387E-01 0.8775E-06 Kr-87 0.1285E 01 0.6227E 00 0.2430E-01 0.1922E-05 0.1515E-22 Kr-88 0.3503E 01 0.4192E 01 0.1180E 01 0.1097E-01 0.1648E-09 Xe-133 0.5617E 02 0.1648E 03 0.4139E 03 0.7359E 03 0.1469E 04 Xe-133M 0.9285E 00 0.2650E 01 0.6162E 01 0.8236E 01 0.5596E 01 Xe-135 0.6662E 01 0.1490E 02 0.1826E 02 0.3887E 01 0.1721E-01 Xe-135M 0.1289E 00 0.6268E-03 0.7107E-10 0.1070E-28 0.0 Xe-138 0.3957E 00 0.1035E-02 0.1840E-10 0.1979E-31 0.0 DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-14 CALCULATED ACTIVITY RELEASES FROM LOCA - DESIGN BASIS CASE (CURIES)
Nuclide 0-2 Hr 2-8 Hr 2-24 Hr 24-96 Hr 4-30 Days I-131 0.2703E 02 0.0 0.0 0.0 0.0 I-132 0.3985E 02 0.0 0.0 0.0 0.0 I-133 0.6207E 02 0.0 0.0 0.0 0.0 I-134 0.7063E 02 0.0 0.0 0.0 0.0 I-135 0.5712E 02 0.0 0.0 0.0 0.0 I-131ORG 0.7340E 02 0.2170E 03 0.5561E 03 0.1070E 04 0.3227E 04 I-132ORG 0.8325E 02 0.8763E 02 0.1862E 02 0.9240E-01 0.8557E-10 I-133ORG 0.1639E 03 0.4314E 03 0.8078E 03 0.5263E 03 0.5383E 02 I-134ORG 0.9847E 02 0.2469E 02 0.2045E 00 0.2811E-06 0.2665E-31 I-135ORG 0.1411E 03 0.2838E 03 0.2668E 03 0.3148E 02 0.1834E-01 I-131PAR 0.9175E 02 0.2713E 03 0.6951E 03 0.1338E 04 0.4033E 04 I-132PAR 0.1041E 03 0.1095E 03 0.2327E 02 0.1155E 00 0.1070E-09 I-P33PAR 0.2048E 03 0.5392E 03 0.1010E 04 0.6579E 03 0.6728E 02 I-134PAR 0.1231E 03 0.3086E 02 0.2557E 00 0.3514E-06 0.3331E-31 I-135PAR 0.1764E 03 0.3548E 03 0.3335E 03 0.3935E 02 0.2293E-01 Kr-83M 0.9280E 03 0.7487E 03 0.8940E 02 0.1154E 00 0.2578E-12 Kr-85 0.6379E 02 0.1913E 03 0.5097E 03 0.1145E 04 0.9827E 04 Kr-85M 0.2823E 04 0.4660E 04 0.2723E 04 0.1191E 03 0.1413E-02 Kr-87 0.3847E 04 0.1864E 04 0.7273E 02 0.5752E-02 0.4530E-19 Kr-88 0.7090E 04 0.8484E 04 0.2388E 04 0.2220E 02 0.3333E-06 Xe-133 0.1684E 05 0.4942E 05 0.1241E 06 0.2205E 06 0.4392E 06 Xe-133M 0.4250E 03 0.1212E 04 0.2819E 04 0.3766E 04 0.2556E 04 Xe-135 0.7402E 04 0.1655E 05 0.2028E 05 0.4316E 04 0.1910E 02 Xe-135M 0.8506E 03 0.4137E 01 0.4690E-06 0.7061E-25 0.0 Xe-138 0.2504E 04 0.6552E 01 0.1164E-06 0.1252E-27 0.0 DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-15 THYROID DOSE HOUR - CONTAINMENT LEAKAGE - EXPECTED CASE (REM)
Distance From Release Point, meters Nuclide 800 1200 2000 4000 7000 10000 20000 I-131 0.7456E-03 0.4792E-03 0.2636E-03 0.1097E-03 0.5060E-04 0.3101E-04 0.1247E-04 I-132 0.4383E-05 0.2817E-05 0.1549E-05 0.6445E-06 0.2974E-06 0.1823E-06 0.7332E-07 I-133 0.1527E-03 0.9814E-04 0.5398E-04 0.2246E-04 0.1036E-04 0.6350E-05 0.2554E-05 I-134 0.2268E-05 0.1458E-05 0.8017E-06 0.3336E-06 0.1539E-06 0.9432E-07 0.3794E-07 I-135 0.2479E-04 0.1593E-04 0.8763E-05 0.3646E-05 0.1682E-05 0.1031E-05 0.4147E-06 I-131ORG 0.2587E-03 0.1663E-03 0.9145E-04 0.3805E-04 0.1756E-04 0.1076E-04 0.4328E-05 I-132ORG 0.1169E-05 0.7514E-06 0.4132E-06 0.1719E-06 0.7933E-07 0.4862E-07 0.1956E-07 I-133ORG 0.5149E-04 0.3310E-04 0.1820E-04 0.7573E-05 0.3495E-05 0.2142E-05 0.8615E-06 I-134ORG 0.4031E-06 0.2591E-06 0.1425E-06 0.5929E-07 0.2736E-07 0.1677E-07 0.6744E-08 I-135ORG 0.7820E-05 0.5026E-05 0.2764E-05 0.1150E-05 0.5307E-06 0.3252E-06 0.1308E-06 I-131PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-132PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-133PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-134PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-135PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0
TOTAL 0.1249E-02 0.8030E-03 0.4416E-03 0.1837E-03 0.8479E-04 0.5196E-04 0.2090E-04
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-17 THYROID DOSE DAY - CONTAINMENT LEAKAGE - EXPECTED CASE (REM)
Distance From Release Point, meters Nuclide 800 1200 2000 4000 7000 10000 20000 I-131 0.7456E-03 0.4792E-03 0.2636E-03 0.1097E-03 0.5060E-04 0.3101E-04 0.1247E-04 I-132 0.4383E-05 0.2817E-05 0.1549E-05 0.6445E-06 0.2974E-06 0.1823E-06 0.7332E-07 I-133 0.1527E-03 0.9814E-04 0.5398E-04 0.2246E-04 0.1036E-04 0.6350E-05 0.2554E-05 I-134 0.2268E-05 0.1458E-05 0.8017E-06 0.3336E-06 0.1539E-06 0.9432E-07 0.3794E-07 I-135 0.2479E-04 0.1593E-04 0.8763E-05 0.3646E-05 0.1682E-05 0.1031E-05 0.4147E-06 I-131ORG 0.1252E-02 0.7543E-03 0.3960E-03 0.1587E-03 0.7223F-04 0.4452E-04 0.1762E-04 I-132ORG 0.2250E-05 0.1438E-05 0.7882E-06 0.3270E-06 0.1507E-06 0.9242E-07 0.3713E-07 I-133ORG 0.2091E-03 0.1280E-03 0.6797E-04 0.2751E-04 0.1257E-04 0.7734E-05 0.3074E-05 I-134ORG 0.4911E-06 0.3156E-06 0.1736E-06 0.7222E-07 0.3332E-07 0.2042E-07 0.8215E-08 I-135ORG 0.2352E-04 0.1473E-04 0.7955E-05 0.3264E-05 0.1498E-05 0.9202E-06 0.3679E-06 I-131PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-132PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-133PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-134PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-135PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0
TOTAL 0.2417E-02 0.1496E-02 0.8016E-02 0.3266E-03 0.1466E-03 0.9195E-04 0.3666E-04
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-19 WHOLE BODY DOSE HOUR - CONTAINMENT LEAKAGE - EXPECTED CASE (REM)
Distance From Release Point, meters Nuclide 800 1200 2000 4000 7000 10000 20000 I-131 0.3042E-06 0.1955E-06 0.1075E-06 0.4474E-07 0.2065E-07 0.1265E-07 0.5089E-08 I-132 0.2274E-06 0.1462E-06 0.8039E-07 0.3344E-07 0.1543E-07 0.9457E-08 0.3804E-08 I-133 0.4190E-06 0.2693E-06 0.1481E-06 0.6162E-07 0.2843E-07 0.1742E-07 0.7009E-08 I-134 0.1843E-06 0.1185E-06 0.6517E-07 0.2711E-07 0.1251E-07 0.7667E-08 0.3084E-08 I-135 0.4245E-06 0.2729E-06 0.1501E-06 0.6244E-07 0.2881E-07 0.1766E-07 0.7102E-08 I-131ORG 0.1055E-06 0.6784E-07 0.3731E-07 0.1552E-07 0.7163E-08 0.4389E-08 0.1766E-08 I-132ORG 0.6066E-07 0.3899E-07 0.2144E-07 0.8921E-08 0.4117E-08 0.2523E-08 0.1015E-08 I-133ORG 0.1413E-06 0.9082E-07 0.4995E-07 0.2078E-07 0.9589E-08 0.5876E-08 0.2364E-08 I-134ORG 0.3277E-07 0.2106E-07 0.1158E-07 0.4819E-08 0.2224E-08 0.1363E-08 0.5482E-09 I-135ORG 0.1339E-06 0.8608E-07 0.4734E-07 0.1970E-07 0.9089E-08 0.5570E-08 0.2241E-08 I-131PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-132PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-133PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-134PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-135PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Kr-83M 0.1915E-06 0.1231E-06 0.6771E-07 0.2817E-07 0.1300E-07 0.7965E-08 0.3204E-08 Kr-85 0.1469E-04 0.9439E-05 0.5192E-05 0.2160E-05 0.9967E-06 0.6108E-06 0.2457E-06 Kr-85M 0.9000E-05 0.5784E-05 0.3181E-05 0.1324E-05 0.6108E-06 0.3743E-06 0.1506E-06 Kr-87 0.4797E-04 0.3083E-04 0.1696E-04 0.7056E-05 0.3256E-05 0.1995E-05 0.8026E-06 Kr-88 0.1048E-03 0.6735E-04 0.3704E-04 0.1541E-04 0.7112E-05 0.4358E-05 0.1753E-05 Xe-133 0.1260E-03 0.8097E-04 0.4453E-04 0.1853E-04 0.8550E-05 0.5239E-05 0.2108E-05 Xe-133M 0.2658E-05 0.1708E-05 0.9396E-06 0.3909E-06 0.1804E-06 0.1105E-06 0.4447E-07 Xe-135 0.4764E-04 0.3062E-04 0.1684E-04 0.7006E-05 0.3233E-05 0.1981E-05 0.7969E-06 Xe-135M 0.8720E-06 0.5605E-06 0.3083E-06 0.1282E-06 0.5918E-07 0.3627E-07 0.1459E-07 Xe-138 0.9485E-05 0.6096E-05 0.3353E-05 0.1395E-05 0.6437E-06 0.3945E-06 0.1587E-06
TOTAL 0.3653E-03 0.2348E-03 0.1291E-03 0.5373E-04 0.2479E-04 0.1519E-04 0.6112E-05
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-21 WHOLE BODY DOSE DAY - CONTAINMENT LEAKAGE - EXPECTED CASE (REM)
Distance From Release Point, meters Nuclide 800 1200 2000 4000 7000 10000 20000 I-131 0.3042E-00 0.1955E-06 0.1075E-06 0.4474E-07 0.2065E-07 0.1265E-07 0.5089E-08 I-132 0.2274E-06 0.1462E-06 0.8039E-07 0.3344E-07 0.1543E-07 0.9457E-08 0.3804E-08 I-133 0.4190E-06 0.2693E-06 0.1481E-06 0.6162E-07 0.2843E-07 0.1742E-07 0.7009E-08 I-134 0.1843E-06 0.1185E-06 0.6517E-07 0.2711E-07 0.1251E-07 0.7667E-08 0.3084E-08 I-135 0.4245E-06 0.2729E-06 0.1501E-06 0.6244E-07 0.2881E-07 0.1766E-07 0.7102E-08 I-131ORG 0.5109E-06 0.3078E-06 0.1616E-06 0.6474E-07 0.2947E-07 0.1816E-07 0.7187E-08 I-132ORG 0.1167E-06 0.7463E-07 0.4090E-07 0.1697E-07 0.7822E-08 0.4796E-08 0.1927E-08 I-133ORG 0.5738E-06 0.3511E-06 0.1865E-06 0.7549E-07 0.3448E-07 0.2122E-07 0.8435E-08 I-134ORG 0.3992E-07 0.2566E-07 0.1411E-07 0.5870E-08 0.2709E-08 0.1660E-08 0.6677E-09 I-135ORG 0.4028E-06 0.2522E-06 0.1362E-06 0.5590E-07 0.2566E-07 0.1576E-07 0.6301E-08 I-131PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-132PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-133PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-134PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 I-135PAR 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Kr-83M 0.3536E-06 0.2263E-06 0.1241E-06 0.5151E-07 0.2375E-07 0.1456E-07 0.5851E-08 Kr-85 0.2189E-03 0.1162E-03 0.5620E-04 0.2106E-04 0.9086E-05 0.5697E-05 0.2150E-05 Kr-85M 0.2744E-04 0.1717E-04 0.9267E-05 0.3800E-05 0.1744E-05 0.1071E-05 0.4283E-06 Kr-87 0.7159E-04 0.4596E-04 0.2526E-04 0.1050E-04 0,4846E-05 0.2970E-05 0.1195E-05 Kr-88 0.2446E-03 0.1553E-03 0.8472F-04 0.3503E-04 0.1612E-04 0.9891E-05 0.3968E-05 Xe-133 0.1222E-02 0.6844E-03 0.3406E-03 0.1298E-03 0.5784E-04 0.3588E-04 0.1383E-04 Xe-133M 0.2137E-04 0.1224E-04 0.6179E-05 0.2385E-05 0.1072E-05 0.6632E-06 0.2578E-06 Xe-135 0.2113E-03 0.1283E-03 0.6775E-04 0.2729E-04 0.1244E-04 0.7664E-05 0.3040E-05 Xe-135M 0.8763E-06 0.5632E-06 0.3098E-06 0.1289E-06 0.5947E-07 0.3644E-07 0.1466E-07 Xe-138 0.9510E-05 0.6112E-05 0.3362E-05 0.1399E-05 0.6454E-06 0.3955E-06 0.1591E-06 TOTAL 0.2031E-02 0.1169E-02 0.5949E-03 0.2319E-03 0.1041E-03 0.6441E-04 0.2510E-04 DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-23
SUMMARY
OF EXPOSURE FROM CONTAINMENT LEAKAGE (a) Thyroid Doses, rem EAB - 2 Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case 95.9 17.7 Expected case 1.25 x 10-3 9.20 x 10-5 Whole Body Doses, rem EAB - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 5.61 (b) 0.57 (c) Expected case 3.65 x 10-4 6.44 x 10-5 Population Doses, man-rem Design basis case 932.1 Expected case 0.269 (a) These values correspond to the original analysis. See Table 15.5-75 for current analysis
(b) The EAB Whole Body dose of 5.61 rem is 3.69 rem gamma and 1.92 rem beta
(c) The LPZ Whole Body dose of 0.57 rem is 0.33 rem gamma and 0.24 rem beta
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-24 Sheet 1 of 5 ASSUMPTIONS USED TO CALCULATE OFFSITE EXPOSURES FROM POST-LOCA CIRCULATION LOOP LEAKAGE IN THE AUXILIARY BUILDING Revision 22 May 2015 Expected Expected DBA DBA Small Leakage Large Leakage Small Leakage Large Leakage A. ECCS, Containment Fan Cooler, Containment Spray System Operation 1. ECCS trains functioning 2 2 2 2 2. Containment fan coolers functioning 5 5 2 2 3. Containment spray system trains functioning 2 2 1 1 B. Activity Deposited in Containment Recirculation Sump Water 1. Iodine
& PU-239 fissions) 100% of gap
inventory per
Table 11.1-7; (I-
127, 129, rel.
fract. of 0.015; I-
131, 132, 133, 134, 135 rel.
fract. Table 11.1-
- 7) 100% of gap
inventory per Table
11.1-7; (I-127, 129, rel. fract. of 0.015;
I-131, 132, 133, 134, 135, rel. fract.
Table 11.1-7) 100% of gap
inventory per
Regulatory Guide
1.25; (I-127, 129, rel.
fract. of 0.30; I-131, 132, 133, 134, 135
rel. fract. of 0.10) 100% of gap
inventory per
Regulatory Guide
1.25; (I-127, 129, rel.
fract. of I-131, 132, 133, 134, 135, rel.
fract. of 0.10)
inventory 99.75% of gap iodine
inventory 99.75% of gap
iodine inventory (1) I-127 30.2g, 0 Ci 30.2g, 0 Ci 903g, 0 Ci 903g, 0 Ci (2) I-129 148.5g, 0 Ci 148.5g, 0 Ci 4,445g, 0 Ci 4,445g, 0 Ci (3) I-131, 132, 133, 134, 135 6.5g, 1.82x10 6Ci 6.5g, 1.82x10 6Ci 97g, 8.45x10 7 Ci 97g, 8.45x10 7 Ci
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-24 Sheet 2 of 5 ASSUMPTIONS USED TO CALCULATE OFFSITE EXPOSURES FROM POST-LOCA CIRCULATION LOOP LEAKAGE IN THE AUXILIARY BUILDING Revision 22 May 2015 Expected Expected DBA DBA Small Leakage Large Leakage Small Leakage Large Leakage
inventory 0.25% of gap iodine
inventory 0.25% of gap iodine
inventory (1) I-127 0.0 0.0 2g, 0 Ci 2g, 0 Ci (2) I-129 0.0 0.0 11g, 0 Ci 11g, 0 Ci (3) I-131, 132, 133, 134, 135 0.0 0.0 02g, 2.12x10 5 Ci 02g, 2.12x10 5 Ci
- c. Total iodine 185.2g, 1.82x10 6Ci 185.2g, 182x10 6Ci 5.458g, 8.47x10 7Ci 5.458g, 8.47x10 7 Ci
- 2. Noble Gases 0.0 0.0 0.0 0.0
Other fission products 0.0 0.0 0.0 0.0
C. Containment Recirculation Sump Decay and Cleanup 1. Radiological decay credit Yes Yes Yes Yes 2. Cleanup credit None None None None D. Volume of Water in Which Activity is Deposited (diluted) 1. Reactor coolant water, gal.
93,960 93,960 93,960 93,960 2. Accumulator water, gal.
25,040 25,040 25,040 25,040
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-24 Sheet 3 of 5 ASSUMPTIONS USED TO CALCULATE OFFSITE EXPOSURES FROM POST-LOCA CIRCULATION LOOP LEAKAGE IN THE AUXILIARY BUILDING Revision 22 May 2015 Expected Expected DBA DBA Small Leakage Large Leakage Small Leakage Large Leakage D. Volume of Water in Which Activity is Deposited (diluted) (Cont'd) 3. Refueling water storage tank, gal. (Table 6.3-1) 350,000 262,030 350,000 254,220
- 4. Total, gal.
469,000 381,030 469,000 373,220 E. Conditions of Loop Leakage Water
- 1. pH of leakage water (Figure 6.2-15) 8.8 8.4 8.5 7.85
- 2. Temperature of leakage water, °F 120 238 120 242 F. Loop Leakage Rate 1910 cc/hr.
50 gpm (Table 6.3-9) 1910 cc/hr 50 gpm (Table 6.3-9)
G. Duration of Loop Leakage
- 1. Time after LOCA leakage begins, hr (Table 6.3-5) 0.337 0.337 0.395 24
- 2. Time after LOCA leakage ends, hr 720 0.837 720 24.5
- 3. Total duration of loop leakage, hr 719.7 0.5 719.6 0.5
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-24 Sheet 4 of 5 ASSUMPTIONS USED TO CALCULATE OFFSITE EXPOSURES FROM POST-LOCA CIRCULATION LOOP LEAKAGE IN THE AUXILIARY BUILDING Revision 22 May 2015 Expected Expected DBA DBA Small Leakage Large Leakage Small Leakage Large Leakage H. Auxiliary Building Iodine Decontamination Factors
- 1. Elemental iodine decontamination factor a. M vapor, lbm - 2.78 x 102
-2 - 3.22 x 10-2 M liquid, lbm
- b. V vapor, ft 3/lbm - 1.60 x 10+3 - 1.60 x 10+3 V liquid, ft 3/lbm c. Partition coefficient, - 7.22 x 10+5 - 6.77 x 10+3 PC, (g/1) liquid (g/1) gas
- d. Partition factor, - 6.18 x 10-5 - 7.62 x 10-3 PF, (g) gas (g) liquid
- e. Decontamination factor, 1.0 1.62 x 10
+4 1.0 1.32 x 10+2 DF, (g) leak (g) gas
I. Auxiliary Building Decay, Plateout, and Filter Removal 1. Radiological decay credit None None None None 2. Plateout credit None None None None
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-24 Sheet 5 of 5 ASSUMPTIONS USED TO CALCULATE OFFSITE EXPOSURES FROM POST-LOCA CIRCULATION LOOP LEAKAGE IN THE AUXILIARY BUILDING Revision 22 May 2015 Expected Expected DBA DBA Small Leakage Large Leakage Small Leakage Large Leakage I. Auxiliary Building Decay, Plateout, and Filter Removal (Cont'd)
- 3. Auxiliary building None Yes None Yes filter credit
0.0 99.0 0.0 90.0 (2) Organic iodine, %
0.0 85.0 0.0 70.0 (3) Particulate iodine %
0.0 99.0 0.0 90.0 b. Noble gases 0.0 0.0 0.0 0.0 J. Atmospheric Dispersion
- 1. Down wind radiological None None None None decay credit
- 2. Atmospheric dilution Table 15.5-4 Table 15.5-4 Table 15.5-4 Table 15.5-4 factors K. Breathing Rates Table 15.5-7 Table 15.5-7 Table 15.5-7 Table 15.5-7
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-26 PERCENTAGE OCCURRENCE OF WIND DIRECTION AND CALM WINDS EXPRESSED AS PERCENTAGE OF TOTAL HOURLY OBSERVATIONS WITHIN EACH SEASON AT THE SITE (250-FOOT LEVEL)
Wind Direction
Season (a) Offshore (b) Onshore (c) Calm (d)
Annual 57% 38% 5%
Dry 55% 40% 5%
Wet 54% 42% 4%
Transitional 62% 34% 4%
(a) Dry Season - May through September Wet Season - November through March Transitional - April and October
(b) Offshore wind directions are defined as wind directions from northwest through east southeast measured clockwise.
(c) Onshore wind directions are defined as wind directions from southeast through west-northwest measured clockwise.
(d) Calm wind directions are defined as winds with speeds less than one 1 mph.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-27 Sheet 1 of 2 Revision 22 May 2015 DIABLO CANYON POWER PLANT SITE PROBABILITY OF PERSISTENCE OFFSHORE WIND DIRECTION SECTORS (250-FOOT LEVEL)
Conse- NNE NE ENE E ESE cutive Hours A (a) D (a) W (a) T (a) A D W T A D W T A D W T A D W T 1 0.461 0.652 0.432 0.390 0.426 0.655 0.373 0.477 0.505 0.769 0.401 0.732 0.678 0.906 0.577 0.840 0.400 0.430 0.370 0.434 2 0.245 0.107 0.327 0.178 0.252 0.241 0.237 0.318 0.190 0.231 0.180 0.195 0.158 0.038 0.189 0.160 0.152 0.180 0.154 0.110 3 0.129 0.054 0.039 0.248 0.136 0.103 0.151 0.102 0.133 - 0.176 0.073 0.069 0.057 0.090 - 0.082 0.075 0.068 0.124 4 0.073 0.071 0.049 0.118 0.041 - 0.047 0.045 0.064 - 0.090 - 0.066 - 0.100 - 0.086 0.120 0.046 0.138 5 0.058 0.045 0.077 0.030 0.041 - 0.044 0.057 0.064 - 0.090 - 0.000 - 0.000 - 0.058 0.100 0.028 0.069 6 0.010 0.000 0.000 0.036 0.050 - 0.071 - 0.019 - 0.027 - 0.000 - 0.000 - 0.060 0.060 0.034 0.124 7 0.012 0.000 0.022 - 0.000 - 0.000 - 0.000 - 0.000 - 0.000 - 0.000 - 0.050 0.035 0.080 -
8 0.013 0.071 - - 0.033 - 0.047 - 0.025 - 0.036 - 0.000 - 0.000 - 0.023 - 0.046 -
9 - - - - 0.000 - 0.000 - - - - - - - 0.045 - 0.013 - 0.026 -
10 - - - - 0.021 - 0.030 - - - - - - - - - 0.014 - 0.028 -
11 - - - - - - - - - - - - - - - - 0.000 - 0.000 -
12 - - - - - - - - - - - - - - - - 0.000 - 0.000 -
13 - - - - - - - - - - - - - - - - 0.037 - 0.074 -
14 - - - - - - - - - - - - - - - - 0.000 - 0.000 -
15 - - - - - - - - - - - - - - - - 0.000 - 0.000 -
16 - - - - - - - - - - - - - - - - 0.023 - 0.046 -
17 - - - - - - - - - - - - - - - - - - - -
18 - - - - - - - - - - - - - - - - - - - -
19 - - - - - - - - - - - - - - - - - - - -
20 - - - - - - - - - - - - - - - - - - - -
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-27 Sheet 2 of 2 Revision 22 May 2015 DIABLO CANYON POWER PLANT SITE PROBABILITY OF PERSISTENCE OFFSHORE WIND DIRECTION SECTORS (250-FOOT LEVEL)
Conse- NW NNW N Calm cutive Hours A D W T A D W T A D W T A D W T 1 0.105 0.086 0.157 0.105 0.364 0.453 0.318 0.348 0.504 0.701 0.461 0.442 0.364 0.329 0.393 0.426 2 0.091 0.073 0.142 0.090 0.194 0.194 0.201 0.178 0.200 0.124 0.217 0.231 0.237 0.214 0.239 0.313 3 0.078 0.064 0.137 0.056 0.125 0.115 0.135 0.118 0.088 0.051 0.078 0.144 0.152 0.166 0.151 0.101 4 0.067 0.058 0.111 0.049 0.085 0.089 0.103 0.042 0.104 0.090 0.122 0.077 0.103 0.081 0.164 0.054 5 0.058 0.040 0.086 0.085 0.067 0.051 0.071 0.078 0.018 0.000 0.022 0.048 0.055 0.074 0.016 0.068 6 0.062 0.048 0.098 0.068 0.036 0.036 0.046 0.016 0.049 0.034 0.052 0.058 0.018 0.022 0.000 0.041 7 0.050 0.045 0.046 0.068 0.034 0.028 0.045 0.018 0.016 - 0.030 - 0.014 0.026 0.000 -
8 0.047 0.039 0.046 0.072 0.039 0.032 0.021 0.084 0.009 - 0.017 - 0.000 0.000 0.000 -
9 0.045 0.044 0.029 0.059 0.016 - 0.023 0.024 0.000 - - - 0.027 0.050 0.000 -
10 0.038 0.044 0.008 0.049 0.006 - 0.000 0.026 0.012 - - - 0.030 0.037 0.031 -
11 0.046 0.060 0.009 0.054 0.007 - 0.000 0.029 - - - - - - - -
12 0.035 0.028 0.049 0.039 0.007 - 0.016 0.000 - - - - - - - -
13 0.038 0.054 0.011 0.011 0.000 - - 0.000 - - - - - - - -
14 0.038 0.043 0.011 0.045 0.000 - - 0.000 - - - - - - - -
15 0.019 0.027 0.000 0.012 0.009 - - 0.039 - - - - - - - -
16 0.020 0.025 0.000 0.026 0.010 - - - - - - - - - - -
17 0.022 0.031 0.000 0.014 - - - - - - - - - - - -
18 0.023 0.033 0.015 0.015 - - - - - - - - - - - -
19 0.030 0.034 0.016 0.046 - - - - - - - - - - - -
20 0.013 0.021 - 0.000 - - - - - - - - - - - -
21 0.003 0.005 - 0.000 - - - - - - - - - - - -
22 0.007 0.006 - 0.018 - - - - - - - - - - - -
23 0.004 0.006 - 0.000 - - - - - - - - - - - -
24 0.012 0.012 - 0.020 - - - - - - - - - - - -
25 0.012 0.019 - - - - - - - - - - - - - -
26 0.012 0.020 - - - - - - - - - - - - - -
27 0.004 0.007 - - - - - - - - - - - - - -
28 0.004 0.007 - - - - - - - - - - - - - -
29 0.000 0.000 - - - - - - - - - - - - - -
30 0.000 0.000 - - - - - - - - - - - - - -
31 0.005 0.008 - - - - - - - - - - - - - -
(a) A = Annual D = Dry season (May through September)
W = Wet season (November through March)
T = Transitional months (April and October)
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-28 Sheet 1 of 2 Revision 22 May 2015 ASSUMPTIONS USED TO CALCULATE ONSHORE CONTROLLED CONTAINMENT VENTING Expected Case DBA Case A. Activity Released to Containment Atmosphere
- a. Elemental 24.95% of gap iodine inventory 22.75% of core iodine inventory b. Organic 0.05% of gap iodine inventory 1.0% of core iodine inventory c. Particulate 0% of gap iodine inventory 1.25% of core iodine inventory
- 2. Noble gases 100% of gap inventory 100% of core inventory
- 3. Other fission products None None B. Decay, Cleanup, and Leakage in Containment Atmosphere
- 1. Radiological decay credit Yes Yes 2. Iodine spray cleanup
- a. Elemental 92.0 hr-1 31.0 hr-1 (a) b. Organic 0.58 hr
-1 0 hr-1 c. Particulate 0 0
- 3. Filter cleanup of containment atmosphere
- a. Iodines None None b. Noble gases None None 4. Containment leak rate 0.05%/per day 0.05%/per day DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-28 Sheet 2 of 2 Revision 22 May 2015 ASSUMPTIONS USED TO CALCULATE ONSHORE CONTROLLED CONTAINMENT VENTING Expected Case DBA Case C. Containment Atmosphere Volume 2.68 x 10 6 cubic feet 2.68 x 10 6 cubic feet D. Purge Schedule
- 1. Time after LOCA purging begins 1968 hours0.0228 days <br />0.547 hours <br />0.00325 weeks <br />7.48824e-4 months <br />, Chapter 6 672 hours0.00778 days <br />0.187 hours <br />0.00111 weeks <br />2.55696e-4 months <br />, Chapter 6
- 2. Time after LOCA purging ends 6792 hours0.0786 days <br />1.887 hours <br />0.0112 weeks <br />0.00258 months <br />, remainder of 1 yr.
8088 hours0.0936 days <br />2.247 hours <br />0.0134 weeks <br />0.00308 months <br />, remainder of 1 yr.
E. Purge Flowrate 10 cfm, Chapter 6 25 cfm, Chapter 6 F. Filter Efficiency
- 1. Iodines
- a. Elemental 99% 90% b. Organic 85% 70% c. Particulate 99% 90% 2. Noble gases None None G. Atmospheric Dispersion
- 1. Radiological decay credit None None 2. /Qs Table 15.5-30 Table 15.5-30 H. Breathing Rates Table 15.5-7 Table 15.5-7 (a) Although a subsequent safety evaluation showed that the Design Case coefficient of 31
-1 (for 2600 gpm spray header flow) should be reduced to approximately 29 hr
-1 (for 2466 gpm spray header flow), the potential offsite dose increase due to this change is extremely small and can be conside red insignificant (Reference 39).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-29 ONSHORE CONTROLLED CONTAINMENT VENTING EXPOSURES DBA Expected Thyroid exposure at site 2.21 9.83 x 10-25 boundary (800 meters), rem
Whole body exposure 0.0841 7.15 x 10
-3 at site boundary (800 meters), rem
DCPP UNITS 1 & 2 FSAR UPDATE Revision 11 November 1996 TABLE 15.5-30 ATMOSHPERIC DISPERSION FACTORS FOR ONSHORE CONTROLLED CONTAINMENT VENTING (STABILITY CATEGORY D)
Distance, km /Q, sec/m 3 0.8 1.437 x 10
-6 1.2 2.440 x 10
-6 2.0 1.968 x 10
-6 4.0 7.884 x 10
-7 7.0 3.135 x 10
-7 10.0 1.691 x 10
-7 20.0 6.099 x 10
-8 Meteorological Input Parameters:
Height of release = 70 meters Mixing depth = 350 meters Mean wind speed = 5.8 meters per second Sigma theta = 10 degrees Sigma phi = 3 degrees Vertical expansion rate beta, , = 0.9 Azimuth expansion rate alpha, , = 0.9 y = x and Z = x DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-31 Sheet 1 of 2 Revision 22 May 2015 CONTROL ROOM INFILTRATION ASSUMED FOR RADIOLOGICAL EXPOSURE CALCULATIONS Leakage Path Leakage Equation Leakage (cfm)
A. Windows No leakage, no windows. 0.0
B. Doors 3 C. Penetrations
- 1. Ducting (external seal)
No leakage: ducting penetrations caulked to full depth 0.0 and exterior surfaces sealed with FLAMEMASTIC 71A and control room will be positively pressurized.
- 2. Piping (external seal)
No leakage: concrete walls and floor poured with piping 0.0 in place and control room will be positively pressurized.
- 3. Conduits and trays
- a. External seal No leakage: space between exposed conductors and trays is 0.0 sealed with B&W KAOWOOL ceramic fiber 6 inches in depth, with two coats of FLAMEMASTIC 72A, and control room will be positively pressurized.
- b. Internal seal No leakage: conduits are sealed with THIXOTROPIC silicone 0.0 rubber compound, with a minimum depth of one diameter, and control room will be positively pressurized.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-31 Sheet 2 of 2 Revision 22 May 2015 CONTROL ROOM INFILTRATION ASSUMED FOR RADIOLOGICAL EXPOSURE CALCULATIONS Leakage Path Leakage Equation Leakage (cfm)
D. Dampers Q = A x q x p where:
Q = leakage, cfm A = damper area, square feet q = leakage per unit damper area per in. of water (a) p = pressure difference across damper, in. of water (b) 1. Mode damper #2 A = 6.00 ft 2 , q = 0.001 cfm/ft 2 - in. and p = 6.0 in. W.G.
<0.05 2. Mode damper #3 A = 1.84 ft 2 , q = 0.001 cfm/ft 2 - in. and p = 6.0 in. W.G.
<0.05 3. Mode damper #7 A = 6.00 ft 2 , q = 0.001 cfm/ft 2 - in. and p = 6.0 in. W.G.
<0.05 4. Mode damper #8 A = 1.78 ft 2 , q = 0.001 cfm/ft 2 - in. and p = 6.0 in. W.G.
<0.05 E. Total 3 (c) (a) From manufacturer's published data.
(b) Assume conservatively large value of 6 inches of water; dampers will never see a pressure differential this large.
(c) 10 cfm is conservatively assumed in the analysis.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-32 Sheet 1 of 3 ASSUMPTIONS USED TO CALCULATE POST-ACCIDENT CONTROL ROOM RADIOLOGICAL EXPOSURES Revision 22 May 2015 DBA Case A. Power Level 3580 MWt
B. Activity Released to Containment Atmosphere
- a. Elemental, % of core iodine inventory 22.75
- b. Organic, % of core iodine inventory 1.00
- c. Particulate, % of core iodine inventory 1.25
- 2. Noble gases, % of core inventory 100
- 3. Other fission product None
C. Decay, Cleanup, and Leakage in Containment Atmosphere
- 1. Radiological decay included Yes
- 2. Iodine spray cleanup
- a. Elemental 31 hr
-1 (a) b. Organic 0 hr
-1 c. Particulate 0 hr
-1 3. Decontamination factor (DF) cut-off for spray, elemental 100
- 4. Time post-LOCA spray starts 80 seconds
- 5. Filter cleanup of containment atmosphere None
- a. Iodines None
- b. Noble gases None
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-32 Sheet 2 of 3 ASSUMPTIONS USED TO CALCULATE POST-ACCIDENT CONTROL ROOM RADIOLOGICAL EXPOSURES Revision 22 May 2015 DBA Case 6. Containment leakrate
- a. First 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.1% per day
- b. Remainder of accident period 0.05% per day
D. Recirculation Loop Leakage
- 1. RHR leakage rate 50 gpm
- 2. Start of RHR leakage 24 hrs
- 3. Duration of RHR leakage 0.5 hr
- 4. Charcoal filter efficiency for release of RHR leakage
- a. Iodine filter efficiency
(1) Elemental, % 90.0
(2) Organic, % 70.0
(3) Particulate, % 90.0
- b. Noble gas filter efficiency 0.0
E. Meteorology (atmospheric dilution factors from the containment to the control room) Table 15.5-6
F. Control Room Ventilation Flowrates
- 1. Flowrate of contaminated air 10 cfm infiltrating into the control room
- 2. Flowrate of pressurization air into 2100 cfm the control room
- 3. Flowrate of recirculated control room air 2100 cfm through cleanup filters
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-32 Sheet 3 of 3 ASSUMPTIONS USED TO CALCULATE POST-ACCIDENT CONTROL ROOM RADIOLOGICAL EXPOSURES Revision 22 May 2015 DBA Case G. Decay and Cleanup in Control Room
- 1. Radiological decay included Yes
- 2. Filter cleanup of pressurization air Yes
- a. Iodines
(1) Elemental 95%
(2) Organic 95%
(3) Particulate 95%
- b. Noble gases 0%
H. Control Room Complex Volume (total for 170,000 ft 3 Units 1 and 2)
I. Control Room Occupancy Factors
- 1. 0-24 hours 1
- 2. 24-96 hours 0.6
- 3.96-720 hours 0.4
(a) Although a subsequent safety evaluation showed that the Design Case coefficient of 31 hr
-1 (for 2600 gpm spray header flow) should be reduced to approximately 29 hr
-1 (for 2466 gpm spray header flow), the potential offsite dose increase due to this change is extremely small and can be considered
insignificant (Reference 39).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-33 ESTIMATED POST-ACCIDENT EXPOSURE TO CONTROL ROOM PERSONNEL DBA Expected Case Accident Gamma Beta Thyroid Gamma Beta Thyroid Exposure, Exposure, Exposure, Exposure, Exposure, Exposure, Radiation Source rem rem rem rem rem rem 1. Radiation from airborne fission products postulated to enter the control room See Table 15.5-63 See Table 15.5-63 See Table 15.5-63 --- --- ---
- 2. Direct radiation to the control room from 0.032 0 0 6.8 x 10-5 0 0 fission products in the containment structure
- 3. Direct radiation to the control room from 0.022 0 0 1 x 10-5 0 0 fission products in the containment leakage plume 4. Radiation from airborne fission products 0.0066 0.0243 4.72 1.6 x 10-5 1.0 x 10-4 5 x 10-6 in the containment leakage plume to control room personnel during egress ingress
- 5. Direct radiation from fission products in 0.022 0 0 5.3 x 10-5 0 0 the containment structure to control room personnel during egress-ingress (53 5-minute trips)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-34 STEAM RELEASES FOLLOWING A MAJOR STEAM LINE BREAK Time Period
0-2 hr 2-8 hr Steam release from ruptured pipe, lbm 171,100 Steam release from relief valves, lbm 384,000 893,000 Note: All steam releases listed above are for RSGs. OSG MSLB steam releases, which are used in the MSLB dose analysis of record, are listed in item 11 of Section 15.5.18.2.1.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-40 (HISTORICAL)
LONG-TERM ACTIVITY RELEASE FRACTIONS FOR FUEL FAILURE ACCIDENTS Isotopes Release Fractions I-131 1.37 x 10
-9 I-132 2.51 x 10
-10 I-133 9.43 x 10
-10 I-134 1.02 x 10
-10 I-135 5.34 x 10
-5 Kr-83M 2.16 x 10
-5 Kr-85 0.98 Kr-85M 5.02 x 10
-5 Kr-87 1.51 x 10
-5 Kr-88 3.14 x 10
-5 Xe-133 1.68 x 10
-3 Xe-133M 5.52 x 10
-4 Xe-135 1.07 x 10
-4 Xe-135M 3.23 x 10
-7 Xe-138 2.89 x 10
-6
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-41 Sheet 1 of 2 ACTIVITY RELEASES FOLLOWING A LOCKED ROTOR ACCIDENT (CURIES)
Revision 22 May 2015 Design Basis Case Nuclide 0-2 hr 2-8 hr I-131 8.643E-1 4.1783E0 I-132 1.121E-1 2.505E-1 I-133 7.753E-1 3.4725E0 I-134 6.086E-2 4.067E-2 I-135 3.673E-1 1.4067E0 I-131ORG 0.0 0.0 I-132ORG 0.0 0.0 I-133ORG 0.0 0.0 I-134ORG 0.0 0.0 I-135ORG 0.0 0.0 I-131PAR 0.0 0.0 I-132PAR 0.0 0.0 I-133PAR 0.0 0.0 I-134PAR 0.0 0.0 I-135PAR 0.0 0.0 Kr-83M 9.975E-1 4.281E-1 Kr-85 1.2282E1 1.6610E1 Kr-85M 4.9366E0 4.1691E0 Kr-87 3.2949E0 8.647E-1 Kr-88 8.5004E0 5.4137E0 Xe-133 2.7392E2 3.6840E2 Xe-133M 3.8638E0 5.1188E0 Xe-135 2.0293E1 2.2393E1 Xe-135M 3.026E-1 0 Xe-138 9.120E-1 0
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-41 Sheet 2 of 2 ACTIVITY RELEASES FOLLOWING A LOCKED ROTOR ACCIDENT (CURIES)
Revision 22 May 2015 Expected Case Nuclide 0-2 hr 2-8 hr
I-131 1.109E-2 5.670E-2 I-132 1.223E-3 2.386E-3 I-133 8.495E-3 3.875E-2 I-134 6.708E-4 4.001E-4 I-135 3.984E-3 1.425E-2 I-131ORG 0.0 0.0 I-132ORG 0.0 0.0 I-133ORG 0.0 0.0 I-134ORG 0.0 0.0 I-135ORG 0.0 0.0 I-131PAR 0.0 0.0 I-132PAR 0.0 0.0 I-133PAR 0.0 0.0 I-134PAR 0.0 0.0 I-135PAR 0.0 0.0 Kr-83M 1.556E-2 3.770E-3 Kr-85 2.087E-1 2.379E-1 Kr-85M 7.267E-2 4.341E-2 Kr-87 5.210E-2 5.482E-3 Kr-88 1.401E-1 5.716E-2 Xe-133 2.8246E-0 3.1281E-0 Xe-133M 4.360E-2 4.702E-2 Xe-135 2.825E-1 2.366E-1 Xe-135M 5.116E-3 0 Xe-138 1.564E-2 0 DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-42
SUMMARY
OF OFFSITE DOSES FROM A LOCKED ROTOR ACCIDENT Thyroid Exposures, rem Site Boundary - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> LPZ - 30 days 10 CFR Part 100 300 300 Design basis case 0.30 0.076 Expected case 2.5 x 10-4 6.6 x 10-5 Whole Body Exposures, rem Site Boundary - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> LPZ - 30 days 10 CFR Part 100 25 25 Design basis case 1.3 x 10-2 1.1 x 10-3 Expected case 1.6 x 10-5 1.3 x 10-6 Population Doses, man-rem Design basis case 0.32 Expected case 2.8 x 10-4
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-44 Revision 22 May 2015 COMPOSITE SOURCE TERM FOR FUEL HANDLING ACCIDENT IN THE FUEL HANDLING BUILDING
Isotope Composite Source Term (Ci/assembly at shutdown)
Activity at 100
Hours After
Shutdown (Ci at 100 hrs)
Pool Activity (Ci at 100 hrs)
FHB Activity Based on DF200 for Iodines (Ci at 100 hrs) I-131 5.057E+05 3.625E+05 5.9813E+04 299.0625 I-132 7.283E+05 3.042E+05 5.0193E+04 250.965 I-133 1.032E+06 3.783E+04 6.2420E+03 31.21 I-134 1.165E+06 0 0 0 I-135 9.611E+05 2.689E+01 4.4369E+00 0.0222 Kr-83m 8.196E+04 9.554E-08 1.5764E-08 1.5764E-08 Kr-85m 1.901E+05 3.679E-02 0.0060704 0.0060704 Kr-85 6.353E+03 6.350E+03 3143.25 3143.25 Kr-87 3.828E+05 0 0 0 Kr-88 5.416E+05 1.350E-05 2.2275E-06 2.2275E-06 Kr-89 6.855E+05 0 0 0 Xe-131m 5.661E+03 5.469E+03 902.385 902.385 Xe-133m 3.187E+04 1.306E+04 2154.9 2154.9 Xe-133 9.993E+05 6.914E+05 114081 114081 Xe-135m 2.021E+05 4.264E+00 0.70356 0.70356 Xe-135 2.886E+05 1.327E+03 218.955 218.955 Xe-137 9.140E+05 0 0 0 Xe-138 9.477E+05 0 0 0 Where:
The activity/Assembly at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown = A 100. Pool activity at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> = (A 100)Pool = A 100 x 1.65 x release fraction
= A 100 x 1.65 x 0.1 for iodine and noble gases except Kr-85 and = A 100 x 1.65 x 0.3 for Kr-85 FHB activity at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> = (A 100)Pool / 200 for iodine
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-45 Sheet 1 of 2 ASSUMPTIONS FOR FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AREA Revision 22 May 2015 A. Pre-accident Operation
- 1. Core Power 3580 MWt B. Highest Power Fuel Assembly Characteristics
- 1. Radial peaking factor 1.65 C. Fuel Assembly Damage
- 1. Number of fuel rods per assembly 264
- 2. Number of fuel rods ruptured per assembly 264
- 3. Number of fuel assemblies damaged 1 D. Gap Activity Fractions 1. Iodine 0.10 a. Elemental 0.09975
- b. Organic 0.00025
- c. Particulate 0.0
- 2. Noble gases
- a. Other than Kr-85 0.10
- b. Kr-85 0.30
- 3. Other fission products None E. Gap Activity Release Fractions
- 1. Iodine 1
- 2. Noble gases 1
- 3. Other fission products None F. Fission Product Release Depth 23 feet G. Spent Fuel Pool Decontamination Factors
- 1. Iodine 200
- a. Elemental 500
- b. Organic 1
- c. Particulate None
- 2. Noble gases 1
- 3. Other fission products None
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-45 Sheet 2 of 2 ASSUMPTIONS FOR FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AREA Revision 22 May 2015 H. Decay and Cleanup in Fuel Handling Building 1. Radiological decay credit None
- 2. Radiological cleanup credit None I. Fuel Handling Building Volume 435,000 ft 3 J. Fuel Handling Building Filter Efficiencies Not credited K. Fuel Handling Building Exhaust Rate 40,000 cfm L. Atmospheric Dispersion 1. Radiological decay credit None 2. /Qs EAB (800m) 0 to 2 hr 9.9E-4 sec/m 3 LPZ (10 km) 0 to 8 hr 2.6E-5 sec/m 3 8 to 24 hr 4.5E-6 sec/m 3 24 to 96 hr 1.6E-6 sec/m 3 96 to 720 hr 3.3E-7 sec/m 3 M. Offsite Breathing Rates Table 15.5-7 N. Offsite Power (a) __________________
(a) Assumes the FHB ventilation operates continuously to maximize the FHB exhaust to the early stages of this event with or without offsite power.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 18 October 2008 TABLE 15.5-47
SUMMARY
OF DOSES FROM FUEL HANDLING ACCIDENT IN THE FUEL HANDLING AREA TEDE Exposures, rem Site Boundary 2 - Hours LPZ - 30 Days Regulatory Limit 6.3 6.3 Design basis case 4.265 0.112
Control Room
Regulatory Limit 5 Design basis case 0.689
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-48 Sheet 1 of 2 DESIGN INPUTS AND ASSUMPTIONS FOR FUEL HANDLING ACCIDENTS INSIDE CONTAINMENT Revision 22 May 2015 Parameter Value Containment:
Containment Free Volume (ft
- 3) 2.55E+06 Containment Volume above Fuel Pool (ft
- 3) 33600 Purge Line Flowrate to Environment (CFM) 13750 Depth of Water Above Damaged Fuel (ft) >23
Iodine Decontamination Factors:
Organic 1 Inorganic (Elemental) 500 Overall Effective 200
Exfiltration Rate (cfs) 2.55E+06 Duration of Release (sec) <2.0
Time of Accident after Shutdown (hr) 100 Number of Failed Rods 264
Gap Activity Released from Damaged Rods (%):
Kr-85 30 Noble Gases other than Kr-85 10 Iodines 10
Iodine Gap Inventory (%):
Iorganic 99.75 Organic 0.25
Values Assumed for Generation of Inventories:
Reactor Power (%RTP) 105 Reactor Power (MWt) 3580 Radial Peaking Factor 1.65
Dose Conversion Factors for Iodine Species (REM/Ci):
I-131 1.08E+06 I-132 6.44E+03 I-133 1.80E+05 I-134 1.07E+03 I-135 3.13E+04
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-48 Sheet 2 of 2 DESIGN INPUTS AND ASSUMPTIONS FOR FUEL HANDLING ACCIDENTS INSIDE CONTAINMENT Revision 22 May 2015 Parameter Value Control Room (CR) Input Data:
Control Room Volume (U1 +U2) (cubic feet) 170000 Flowrates (CFM)
Flowrate of contaminated air into CR 2110 Flowrate of recirc CR air thru filters 0 CR pressurization air filter:
Filter Depth 2 inches
Iodine Filter Efficiency:
Elemental 95%
Organic 95%
Particulate 95%
CR Occupancy Factors:
0 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1 24 -96 hours 0.6 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 0.4 Atmospheric Dispersion Factors (sec/m 3): Control Room Pressurization:
0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.05E-05 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 5.38E-05 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 3.91E-05 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 2.27E-05 Control Room Infiltration :
0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.96E-04 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.49E-04 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 1.08E-04 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 6.29E-05 Exclusion Area Boundary (EAB), 800 meters:
0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5.29E-04 Low Population Zone (LPZ), 10,000 meters:
0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.20E-05 8 -24 hours 4.75E-06 1 - 4 days 1.54E-06 4 - 30 days 3.40E-07 Control Room Breathing Rate (m 3/sec): 3.47E-04 Offsite Breathing Rates (m 3/sec): 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.47E-04 8 -24 hours 1.75E-04 1 - 30 days 2.32E-04
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-49 ACTIVITY RELEASES FROM FUEL HANDLING ACCIDENT INSIDE CONTAINMENT (CURIES)
Design Basis Case Nuclide 0-2 hr I-131 299.0625
I-132 250.965
I-133 31.21 I-134 0 I-135 0.0222
Kr-83m 1.5764E-08
Kr-85m 0.0060704
Kr-85 3143.25
Kr-87 0 Kr-88 2.2275E-06
Kr-89 0 Xe-131m 902.385
Xe-133m 2154.9
Xe-133 114081
Xe-135m 0.70356
Xe-135 218.955
Xe-137 0 Xe-138 0
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-50
SUMMARY
OF OFFSITE DOSES FROM FUEL HANDLING ACCIDENT INSIDE CONTAINMENT Thyroid Exposures, rem Control Room - 30 Days Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 30 (GDC 19) 300 300 Design-basis case 22.31 60.62 2.52 Whole Body Immersion Exposures, rem Control Room - 30 Days Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 5 (GDC 19) 25 25 Design-basis
case 7.57x10-3 0.43 0.018 Population Doses, man-rem Design basis case 8.53 Expected case 3 x 10
-3
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-51 Sheet 1 of 2 Revision 22 May 2015 ACTIVITY RELEASES FOLLOWING A ROD EJECTION ACCIDENT (CURIES)
Design Basis Case Nuclide 0-2 hr 2-8 hr 8-24 hr 24-96 hr 4-30 Days I-131 0.9765E-02 0.0 0.0 0.0 0.0 I-132 0.1578E-02 0.0 0.0 0.0 0.0 I-133 0.7394E-02 0.0 0.0 0.0 0.0 I-134 0.1729E-02 0.0 0.0 0.0 0.0 I-135 0.3867E-02 0.0 0.0 0.0 0.0 I-131ORG 0.0 0.0 0.0 0.0 0.0 I-132ORG 0.0 0.0 0.0 0.0 0.0 I-133ORG 0.0 0.0 0.0 0.0 0.0 I-134ORG 0.0 0.0 0.0 0.0 0.0 I-135ORG 0.0 0.0 0.0 0.0 0.0 I-131PAR 0.0 0.0 0.0 0.0 0.0 I-132PAR 0.0 0.0 0.0 0.0 0.0 I-133PAR 0.0 0.0 0.0 0.0 0.0 I-134PAR 0.0 0.0 0.0 0.0 0.0 I-135PAR 0.0 0.0 0.0 0.0 0.0 Kr-83M 0.7646E-01 0.61693-01 0.7366E-02 0.9508E-05 0.2124E-16 Kr-85 0.1066E 01 0.3193E 01 0.8511E 01 0.1912E 02 0.1641E 03 Kr-85M 0.3500E 00 0.5778E 00 0.3377E 00 0.1477E-01 0.1753E-06 Kr-87 0.2570E 00 0.1245E 00 0.4858E 02 0.3842E-06 0.3026E-23 Kr-88 0.7005E 00 0.8383E 00 0.2360E 00 0.2193E-02 0.3293E-10 Xe-133 0.1123E 02 0.3297E 02 0.8275E 02 0.1471E 03 0.2929E 03 Xe-133M 0.1857E 00 0.5299E 00 0.1232E 01 0.1646E 01 0.1117E 01 Xe-135 0.1332E 01 0.2979E 01 0.3650E 01 0.7768E 00 0.3437E-02 Xe-135M 0.2577E-01 0.1253E-03 0.1421E-10 0.2140E-29 0.0 Xe-138 0.7913E-01 0.2070E-03 0.3679E-11 0.3956E-32 0.0
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-51 Sheet 2 of 2 Revision 22 May 2015 ACTIVITY RELEASES FOLLOWING A ROD EJECTION ACCIDENT (CURIES)
Expected Case Nuclide 0-2 hr 2-8 hr 8-24 hr 24-96 hr 4-30 Days I-131 0.1645E-02 0.0 0.0 0.0 0.0 I-132 0.2675E-03 0.0 0.0 0.0 0.0 I-133 0.1247E-02 0.0 0.0 0.0 0.0 I-134 0.2963E-03 0.0 0.0 0.0 0.0 I-135 0.6529E-03 0.0 0.0 0.0 0.0 I-131ORG 0.0 0.0 0.0 0.0 0.0 I-132ORG 0.0 0.0 0.0 0.0 0.0 I-133ORG 0.0 0.0 0.0 0.0 0.0 I-134ORG 0.0 0.0 0.0 0.0 0.0 I-135ORG 0.0 0.0 0.0 0.0 0.0 I-131PAR 0.0 0.0 0.0 0.0 0.0 I-132PAR 0.0 0.0 0.0 0.0 0.0 I-133PAR 0.0 0.0 0.0 0.0 0.0 I-134PAR 0.0 0.0 0.0 0.0 0.0 I-135PAR 0.0 0.0 0.0 0.0 0.0 Kr-83M 0.3823E-01 0.3085E-01 0.3684E-02 0.4757E-05 0.1063E-16 Kr-85 0.5339E 00 0.1599E 01 0.4260E 01 0.9570E 01 0.8242E 02 Kr-85M 0.1750E 00 0.2889E 00 0.1689E 00 0.7387E-02 0.8775E-07 Kr-87 0.1285E 00 0.6227E-01 0.2430E-02 0.1922E-06 0.1515E-23 Kr-88 0.3503E 00 0.4192E 00 0.1180E 00 0.1097E-02 0.1648E-10 Xe-133 0.5617E 01 0.1648E 02 0.4139E 02 0.7359E 02 0.1469E 03 Xe-133M 0.9285E-01 0.2650E 00 0.6162E 00 0.8236E 00 0.5596E 00 Xe-135 0.6662E 00 0.1490E 01 0.1826E 01 0.3887E 00 0.1721E-02 Xe-135M 0.1289E-01 0.6268E-04 0.7107E-11 0.1070E-29 0.0 Xe-138 0.3957E-01 0.1035E-03 0.1840E-11 0.1979E-32 0.0
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-52
SUMMARY
OF OFFSITE DOSES FROM A ROD EJECTION ACCIDENT Thyroid Exposures, rem Site Boundary 2 - Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case 3.3 x 10-3 1.4 x 10-4 Expected case 3.7 x 10-5 1.6 x 10-6 Whole Body Exposures, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 7.3 x 10-4 1.3 x 10-4 Expected case 3.6 x 10-5 6.4 x 10-6 Population Doses, man-rem Design basis case 0.54 Expected case 0.027
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-53
SUMMARY
OF OFFSITE DOSES FROM A RUPTURE OF A GAS DECAY TANK Thyroid Exposures, rem Site Boundary 2 - Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case negligible negligible Expected case negligible negligible Whole Body Exposures, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 2.0 8.4 x 10-2 Expected case 4.4 x 10-2 1.8 x 10-3 Population Doses, man-rem Design basis case 55.1 Expected case 1.21 DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-56
SUMMARY
OF OFFSITE DOSES FROM RUPTURE OF A LIQUID HOLDUP TANK Thyroid Exposures, rem Site Boundary 2 - Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design basis case 1.41 0.432 Whole Body Exposures, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design basis case 0.152 6.70x10-3
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-57
SUMMARY
OF OFFSITE DOSES FROM RUPTURE OF A VOLUME CONTROL TANK Thyroid Exposures, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 300 300 Design Basis Case 3.31 x 10-5 1.38 x 10-6 Expected Case 4.43 x 10-8 1.84 x 10-9 Whole Body Exposures, rem Site Boundary - 2 Hours LPZ - 30 Days 10 CFR Part 100 25 25 Design Basis Case 0.465 0.0193 Expected Case 9.27 x 10-3 3.86 x 10-4 Population Doses, man-rem Design Basis Case 12.72 Expected Case 0.254
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-63 POST-LOCA DOSES WITH MARGIN RECIRCULATION LOOP LEAKAGE CONTROL ROOM OPERATOR DOSES (REM)
Gamma Beta Pathway Thyroid Whole Body Skin Notes Containment leakage 5.96 0.0394 0.480 RHR pump seal leakage 0.022 0.0 0.0 1 Expected recirculation loop leakage 0.85 0.00002 0.0014 Recirculation loop leakage:
1.85 gpm, with charcoal filtration, or 0.186 gpm, with no filtration 18.45 0.0006 0.0083 2 Plume radiation (egress-ingress) 4.72 0.0066 0.0243 3 Other direct radiation pathways 0.00 0.0760 0.00 3 TOTAL CONTROL ROOM OPERATOR DOSES 30.00 - 10 CFR PART 50 APPENDIX A, GDC 19 LIMITS 30 5 30 OFFSITE DOSES (REM)
SITE BOUNDARY Gamma Pathway Thyroid Whole Body Notes Containment leakage 107.06 3.24 RHR pump seal leakage 0.0 0.0 Expected recirculation loop leakage 8.22 0.03 Recirculation loop leakage:
1.88 gpm, with charcoal filtration, or 0.189 gpm, with no filtration 184.72 0.52 2 TOTAL SITE BOUNDARY DOSES 300.00 - - LPZ Gamma Pathway Thyroid Whole Body Notes Containment leakage 19.01 0.293 RHR pump seal leakage 0.09 0.0 Expected recirculation loop leakage 2.12 0.003 Recirculation loop leakage:
11.07 gpm, with charcoal filtration, or 1.11 gpm, with no filtration 278.78 0.44 2 TOTAL LPZ DOSES 300.00 - - 10 CFR PART 100 DOSE LIMITS 300 25 Notes:
- 1. RHR pump seal leakage of 50 gpm for 30 minutes, starting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the start of the LOCA, see Tables 15.5-24 and 15.5-33. 2. Additional recirculation loop leakage, existing at the start of the LOCA and continuing for 30 days.
- 3. Taken from Table 15.5-33.
DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-64 Sheet 1 of 2 PARAMETERS USED IN EVALUATING RADIOLOGICAL CONSEQUENCES FOR SGTR ANALYSIS Revision 22 May 2015 I. Source Data A. Core power level, MWt 3580 B. Total steam generator tube leakage, prior to accident, gpm 1.0 C. Reactor coolant activity:
- 1. Accident initiated spike The initial RC iodine activities based on 1 µCi/gram of D.E. I-131 are presented in Table 15.5-65. The iodine appearance rates based on an iodine
spiking factor of 335 assumed for the
accident initiated spike are presented in
Table 15.5-66
- 2. Pre-accident spike Primary coolant iodine activities based on 60 µCi/gram of D.E. I-131 are presented in Table 15.5-65
- 3. Noble gas activity The initial RC noble gas activities based
on 1% fuel defects are presented in
Table 15.5-67 D. Secondary system initial activity Dose equivalent of 0.1
µCi/gm of I-131, presented in Table 15.5-65 E. Reactor coolant mass, grams 2.27 x 10 8 F. Initial steam generator mass (each),
grams 4.07 x 10 7 G. Offsite power Lost at time of reactor trip H. Primary-to-secondary leakage duration for intact SG, hrs 8 I. Species of iodine 100 percent elemental DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.5-64 Sheet 2 of 2 PARAMETERS USED IN EVALUATING RADIOLOGICAL CONSEQUENCES FOR SGTR ANALYSIS Revision 22 May 2015 II. Activity Release Data A. Ruptured steam generator
- 1. Rupture flow See Figure 15.4.3-6b and Table 15.4-14
- 2. Flashed rupture flow See Figure 15.4.3-11 and Table 15.4-14
- 3. Iodine scrubbing efficiency Not Modeled
- 4. Total steam release, lbs See Figure 15.4.3-9 and Table 15.4-14
- 5. Iodine partition coefficient
- non-flashed
- flashed
100 1.0 B. Intact steam generators
- 1. Total primary-to-secondary leakage, gpm 1.0
- 2. Total steam release, lbs See Figure 15.4.3-10 and Table 15.4-14
- 3. Iodine partition coefficient 100
C. Condenser
- 1. Iodine partition coefficient 100
D. Atmospheric dispersion factors See Table 15.5-68
DCPP UNITS 1 & 2 FSAR UPDATE Revision 19 May 2010 TABLE 15.5-65 IODINE SPECIFIC ACTIVITIES IN THE PRIMARY AND SECONDARY COOLANT (a) - SGTR ANALYSIS Specific Activity (µCi/gm) Primary Coolant Secondary Coolant Nuclide 1 µCi/gm 60 µCi/gm 0.1 µCi/gm I-131 0.794 47.64 0.0794
I-132 0.204 12.24 0.0204
I-133 1.113 66.78 0.1113
I-134 0.139 8.34 0.0139
I-135 0.589 35.34 0.0589
(a) Based on 1, 60 and 0.1
µCi/gm of Dose Equivalent I-131 c onsistent with the DCPP Technical Specifications (Reference 22).
DCPP UNITS 1 & 2 FSAR UPDATE Revision 19 May 2010 TABLE 15.5-66 IODINE SPIKE APPEARANCE RATES (a) - SGTR ANALYSIS (CURIES/SECOND)
I-131 I-132 I-133 I-134 I-135 2.46 1.92 4.14 2.75 3.10 (a) The accident initiated spike appearance rate is 335 times the equilibrium appearance rate. The equilibrium appearance rate is calculated based on a total letdown flow of 143 gpm. This total is comprised of 120 gpm with perfect cleanup, a letdown flow uncertainty of 12 gpm, 10 gpm identified reactor coolant sy stem leakage, and 1 gpm unidentif ied leakage from the reactor coolant system.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 16 June 2005 TABLE 15.5-67 NOBLE GAS SPECIFIC ACTIVITIES IN THE REACTOR COOLANT (a) BASED ON 1% FUEL DEFECTS - SGTR ANALYSIS Nuclide Specific Activity (µCi/gm) Xe-131m 2.523
Xe-133m 3.911
Xe-133 256.3
Xe-135m 0.449
Xe-135 8.663
Xe-138 0.568
Kr-85m 2.141
Kr-85 6.209
Kr-87 1.232
Kr-88 3.907
(a) Based on a 2 year fuel cycle at a core power of 3580 MWt, a 75 gpm reactor coolant system letdown flow rate, and a 90% demineralizer iodine removal efficiency.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-68 ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES - SGTR ANALYSIS OFFSITE EXPOSURE Time Exclusion Area Boundary Low Population Breathing Rate (a) (hours) /Q (Sec/m 3) Zone /Q (Sec/m 3) (m 3/Sec) 0-2 5.29 x 10-4 2.2 x 10-5 3.47 x 10-4 2-8 -
2.2 x 10-5 3.47 x 10-4
CONTROL ROOM EXPOSURE Time Control Room Filtered Pressurization Control Room Unfiltered Pressurization Control Room Breathing Rate (a) (hours) /Q (Sec/m 3) Zone /Q (Sec/m 3) (m 3/Sec) 0-8 7.05 x 10-5 1.96 x 10-4 3.47 x 10-4 8-24 5.38 x 10
-5 1.49 x 10-4 3.47 x 10-4 24-96 3.91 x 10
-5 1.08 x 10-4 3.47 x 10-4
>96 2.27 x 10
-5 6.29 x 10-5 3.47 x 10-4 (a) Regulatory Guide 1.4, Revision 2, June 1974 (Note: Although revision 2 was referenced in the analysis, the breathing rates are the same as those in revision 1 which is the DCPP licensing basis)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-69 THYROID DOSE CONVERSION FACTORS (a) - SGTR ANALYSIS Nuclide I-131 1.07 x 10 6 (Rem/Curie)
I-132 6.29 x 10 3 (Rem/Curie)
I-133 1.81 x 10 5 (Rem/Curie)
I-134 1.07 x 10 3 (Rem/Curie)
I-135 3.14 x 10 4 (Rem/Curie)
(a) International Commission on Radiological Protection Publication 30, 1979.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 16 June 2005 TABLE 15.5-70 AVERAGE GAMMA AND BETA ENERGY FOR NOBLE GASES (a) - SGTR ANALYSIS (MeV/dis)
Nuclide E E I-131 0.38 0.19
I-132 2.2 0.52
I-133 0.6 0.42
I-134 2.6 0.69
I-135 1.4 0.43
Xe-131m 0.0029 0.16
Xe-133m 0.02 0.21
Xe-133 0.03 0.15
Xe-135m 0.43 0.099
Xe-135 0.25 0.32
Xe-138 1.2 0.66
Kr-85m 0.16 0.25
Kr-85 0.0023 0.25
Kr-87 0.79 1.3
Kr-88 2.2 0.25 (a) ENDF-223, October 1975 (Reference 36)
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-71 OFFSITE RADIATION DOSES FROM SGTR ACCIDENT Dose (Rem)
Calculated Allowable Guideline Value Value (Reference 37)
- 1. Accident Initiated Iodine Spike Exclusion Area Boundary (0-2 hr.)
Thyroid CDE 27 30.5 (a) Low Population Zone (0-8 hr.)
Thyroid CDE 1.5 30 2. Pre-Accident Iodine Spike
Exclusion Area Boundary (0-2 hr.)
Thyroid CDE 67 300
Low Population Zone (0-8 hr.)
Thyroid CDE 3.2 300
- 3. Accident Initiated Iodine Spike Whole-Body Gamma Dose
Exclusion Area Boundary (0-2 hr.)
Whole Body Gamma DDE 0.2 2.5 Low Population Zone (0-8 hr.)
Whole-Body Gamma DDE 0.02 2.5
- 4. Pre-Accident Initiated Iodine Spike Whole-Body Gamma Dose Exclusion Area Boundary (0-2 hr.)
Whole Body Gamma DDE 0.3 25 Low Population Zone (0-8 hr.)
Whole-Body Gamma DDE 0.02 25 (a) Note: A dose limit of 30.5 rem has been approved by the NRC based on OSGs. Refer to Section 15.5.20.2.1(5) for further discussion.
DCPP UNITS 1 & 2 FSAR UPDATE Revision 16 June 2005 TABLE 15.5-72 CONTROL ROOM PARAMETERS USED IN EVALUATING RADIOLOGICAL CONSEQUENCES FOR SGTR ANA;YSIS Control Room Isolation Signal Generated Time of Safety Injection Signal Delay in Control Room Isolation After Isolation Signal is Generated 35 seconds Control Room Volume 170,000 ft 3 Control Room Unfiltered In-Leakage 10 cfm Control Room Unfiltered Inflow Normal Mode 4200 cfm Emergency Mode 0 cfm Control Room Filtered Inflow Normal Mode 0 cfm Emergency Mode 2100 cfm Control Room Filtered Recirculation Normal Mode 0 cfm Emergency Mode 2100 cfm Control Room Filter Efficiency 95% ______________
DCPP UNITS 1 & 2 FSAR UPDATE Revision 19 May 2010 TABLE 15.5-74 CONTROL ROOM RADIATION DOSES FROM AIRBORNE ACTIVITY IN SGTR ACCIDENT Accident Initiated Pre-Accident GDC 19 Iodine Spike, rem Iodine Spike, rem Guideline, rem Thyroid CDE
(0-30 days) 0.2 0.8 30 Whole Body DDE
(0-30 days) .002 .002 5 Beta Skin SDE
(0-30 days) 0.09 0.09 30
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-75
SUMMARY
OF POST-LOCA DOSES FROM VARIOUS PATHWAYS (DF OF 100)
THYROID DOSES, rem EAB - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> LPZ - 30 days 10 CFR Part 100 300 300 Containment Leakage 107.06 19.01 RHR Pump Seal (50 gpm) 0 0.09
Pre-existing leak (1910 cc/hr) 8.22 2.12 Total 115.28 21.22 WHOLE BODY DOSES, rem EAB - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> LPZ - 30 days 10 CFR Part 100 25 25 Containment Leakage 3.24 0.293 RHR Pump Seal (50 gpm) 0.0 0.0
Pre-existing leak (1910 cc/hr) 0.03 0.003 Total 3.27 0.296
DCPP UNITS 1 & 2 FSAR UPDATE Revision 22 May 2015 TABLE 15.5-76 WHOLE BODY DOSE CONVERSION FACTORS (a)(b) DOSE EQUIVALENT XE-133 Nuclide Kr-85m 7.48E-15 (Sv m 3/Bq s) Kr-87 4.12E-14 (Sv m 3/Bq s) Kr-88 1.02E-13 (Sv m 3/Bq s) Xe-133m 1.37E-15 (Sv m 3/Bq s) Xe-133 1.56E-15 (Sv m 3/Bq s) Xe-135m 2.04E-14 (Sv m 3/Bq s) Xe-135 1.19E-14 (Sv m 3/Bq s) Xe-138 5.77E-14 (Sv m 3/Bq s) _______________________________________________________
(a) Table III.1 of Federal Guidance Report 12, EPA-402-R-93-081, 1993. (b) Note the AOR used conservative values with respect to the above.
30 35 40 45 50 55 60 65 70 75 80550560570580590600610620630640VESSEL AVERAGE TEMPERATURE (F)
VESSEL DELTA-T (F)SG SAFETY VALVES OPENOVERPOWER DELTA TCORE THERMAL LIMITS ATINDICATED PRESSURESOVERTEMPERATURE DELTA-T AT INDICATED PRESSURES 1860 PSIA 2250 PSIA2460 PSIA FIGURE 15.1-1 ILLUSTRATION OF OVERPOWER AND OVERTEMPERATURE T PROTECTION UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 14 November 2001 FIGURE 15.1-2 ROD POSITION VERSUS TIME ON REACTOR TRIP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.1-3 NORMALIZED RCCA REACTIVITY WORTH VERSUS PERCENT INSERTION UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.1-4 NORMALIZED RCCA BANK REACTIVITY WORTH VERSUS TIME AFTER TRIP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.1-5 DOPPLER POWER COEFFICIENT USED IN ACCIDENT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996
FIGURE15.1-6RESIDUAL DECAY HEAT(BEST ESTIMATE LBLOCA 1979 ANS DECAY HEAT)UNITS 1 AND2DIABLO CANYON SITEFSAR UPDATE(BESTESTIMATELBLOCA1979ANSDECAYHEAT)
FIGURE 15.1-7 1979 ANS DECAY HEAT CURVE(USED FOR NON-LOCA ANALYSES) UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.1-8 FUEL ROD CROSS SECTION UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.1-1 UNCONTROLLED ROD WITHDRAWAL FROM A SUBCRITICAL CONDITION NEUTRON FLUX VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996
AVERAGE CHANNEL FIGURE 15.2.1-3 UNCONTROLLED ROD WITHDRAWAL FROM A SUBCRITICAL CONDITION TEMPERATURE VERSUS TIME.
REACTIVITY INSERTION RATE 75 X 10-5 DELTA K/SEC UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 ROD WITHDRAWAL AT POWER Minimum Feedback, 75 pcm/sec Insertion Rate DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.2-1 Revision 11 November 1996 ROD WITHDRAWAL AT POWER Minimum Feedback, 75 pcm/sec Insertion Rate DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.2-2 Revision 11 November 1996 ROD WITHDRAWAL AT POWER Minimum Feedback, 3 pcm/sec Insertion Rate DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.2-3 Revision 11 November 1996 ROD WITHDRAWAL AT POWER Minimum Feedback, 3 pcm/sec Insertion Rate DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.2-4 Revision 11 November 1996 ROD WITHDRAWAL AT POWER Reactivity Insertion Rate vs. DNBR For 100% Power Cases DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.2-5 Revision 11 November 1996 ROD WITHDRAWAL AT POWER Reactivity Insertion Rate vs. DNBR For 60% Power Cases DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.2-6 Revision 11 November 1996 ROD WITHDRAWAL AT POWER Reactivity Insertion Rate vs. DNBR For 10% Power Cases DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.2-7 Revision 11 November 1996 FIGURE 15.2.3-1 TRANSIENT RESPONSE TO DROPPED ROD CLUSTER CONTROL ASSEMBLY UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.3-2 TRANSIENT RESPONSE TO DROPPED ROD CLUSTER CONTROL ASSEMBLY UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.4-1 VARIATION IN REACTIVITY INSERTION RATE WITH INITIAL BORON CONCENTRATION FOR A DILUTION RATE OF 262 GPM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.5-1 ALL LOOPS OPERATING TWO LOOPS COASTING DOWN CORE FLOW VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 HISTORICAL Revision 23 December 2016 FIGURE 15.2.5-2 ALL LOOPS OPERATING TWO LOOPS COASTING DOWN FAILED LOOP FLOW VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 HISTORICAL Revision 23 December 2016 FIGURE 15.2.5-3 ALL LOOPS OPERATING TWO LOOPS COASTING DOWN HEAT FLUX VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 HISTORICAL Revision 23 December 2016 FIGURE 15.2.5-4 ALL LOOPS OPERATING TWO LOOPS COASTING DOWN NUCLEAR POWER VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 HISTORICAL Revision 23 December 2016 FIGURE 15.2.5-5 ALL LOOPS OPERATING TWO LOOPS COASTING DOWN DNBR VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 HISTORICAL Revision 23 December 2016 FIGURE 15.2.6-1 NUCLEAR POWER TRANSIENT DURING STARTUP OF AN INACTIVE LOOP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.6-2 AVERAGE AND HOT CHANNEL HEAT FLUX TRANSIENTS DURING STARTUP OF AN INACTIVE LOOP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.6-3 CORE FLOW DURING STARTUP OF AN INACTIVE LOOP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.6-4 PRESSURIZER PRESSURE TRANSIENT AND CORE AVERAGE TEMPERATURE TRANSIENT DURING STARTUP OF AN INACTIVE LOOP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.2.6-5 DNBR TRANSIENT DURING STARTUP OF AN INACTIVE LOOP UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 LOSS OF LOAD With Pressurizer Spray and Power Operated Relief Valve For DNB Concern at Beginning of Life DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.7-1 Revision 11 November 1996 LOSS OF LOAD With Pressurizer Spray and Power Operated Relief Valve For DNB Concern at Beginning of Life DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.7-2 Revision 11 November 1996 FIGURE 15.2.7-3 LOSS OF LOAD WITH PRESSURIZER SPRAY AND POWER OPERATED RELIEF VALVE FOR DNB CONCERN AT END OF LIFE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996
FIGURE 15.2.7-4 LOSS OF LOAD WITH PRESSURIZER SPRAY AND POWER OPERATED RELIEF VALVE FOR DNB CONCERN AT END OF LIFE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 LOSS OF LOAD Without Pressurizer Spray and Power Operated Relief Valves For Overpressure Concern at Beginning of Life DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.7-9 Revision 11 November 1996
OUT CONCERN
LOSS OF LOAD With Pressurizer Spray and Power Operated Relief Valves For Overpressure Concern at Beginning of Life DIABLO CANYON UNITS 1 & 2 FIGURE 15.2.7-12 Revision 11 November 1996 FIGURE 15.2.8-1 LOSS OF NORMAL FEEDWATER - RCS TEMPERATURES AND STEAM GENERATOR MASS TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010 FIGURE 15.2.8-2 LOSS OF NORMAL FEEDWATER PRESSURIZER WATER VOLUME AND PRESSURIZER PRESSURE TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010 FIGURE 15.2.8-3 LOSS OF NORMAL FEEDWATER NUCLEAR POWER AND STEAM GENERATOR PRESSURE TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
Revision 22 May 2015
FIGURE 15.2.10-1 FEEDWATER VALVE MALFUNCTION FULL POWER, MANUAL ROD CONTROL NUCLEAR POWER AND CORE HEAT FLUX TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010 Revision 23 December 2016 FIGURE 15.2.10-2 FEEDWATER VALVE MALFUNCTION FULL POWER, MANUAL ROD CONTROL PRESSURIZER PRESSURE AND FAULTED LOOP DELTA-T TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010 Revision 23 December 2016 FIGURE 15.2.10-3 FEEDWATER VALVE MALFUNCTION FULL POWER, MANUAL ROD CONTROL CORE AVERAGE TEMPERATURE AND DNBR TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010 Revision 23 December 2016 2
2 2
0.0 0.2 0.4 0.6 0.8 1.0 1.20102030405060TIME (SEC)NUCLEAR POWER (FRACTION OF NOMINAL) 1.0 2.0 3.0 4.0 5.0 6.00.010.020.030.040.050.060.0TIME (SEC)DNBR FIGURE 15.2.1-1 NUCLEAR POWER AND DNBR TRANSIENTS FOR ACCIDENTAL DEPRESSURIZATION OF THE REACTOR COOLANT SYSTEM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 14 November 2001 1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 24000102030405060TIME SECPRESSURIZER PRESSURE (PSIA) 520.0 540.0 560.0 580.0 600.0 620.00.010.020.030.040.050.060.0TIME (SEC)CORE AVERAGE TEMPERATURE (*F)
FIGURE 15.2.1-2 PRESSURIZER PRESSURE AND CORE AVERAGE TEMPERATURE TRANSIENTS FOR ACCIDENTAL DEPRESSURIZATION OF THE REACTOR COOLANT SYSTEM UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 14 November 2001 FIGURE 15.2.15-1 SPURIOUS ACTUATION OF SAFETY INJECTION SYSTEM AT POWER DNBR ANALYSIS - PRESSURIZER WATER VOLUME AND PRESSURIZER PRESSURE VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 16 June 2005 FIGURE 15.2.15-2 SPURIOUS ACTUATION OF SAFETY INJECTION SYSTEM AT POWER DNBR ANALYSIS - NUCLEAR POWER, STEAM FLOW, AND CORE WATER TEMPERATURE VERSUS TIME UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 16 June 2005
2100 2200 2300 2400 2500020040060080010001200140016001800Time (sec)Pressurizer Pressure (p sia)Case 3 w/o Sprays Case 2 Case 1 - w/o Sprays and PORVS FIGURE 15.2.15-3 SSI PRESSURIZER OVERFILL ANALYSIS TYPICAL PRESSURIZER PRESSURE RESPONSE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
1000 1200 1400 1600 1800020040060080010001200140016001800Time (sec)Pressurizer Li q uid Volume (ft3)Case 3 - w/o SpraysCase 2Case 1 - w/o Sprays and PORVs FIGURE 15.2.15-4 SSI PRESSURIZER OVERFILL ANALYSIS TYPICAL PRESSURIZER LIQUID VOLUME RESPONSE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
550 555 560 565 570020040060080010001200140016001800Time (sec)RCS Tav g (F)Case 3 w/o SpraysCase 2Case 1 - w/o Sprays and PORVS FIGURE 15.2.15-5 SSI PRESSURIZER OVERFILL ANALYSIS TYPICAL RCS AVERAGE TEMPERATURE RESPONSE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.3-1 SAFETY INJECTION FLOW RATE FOR SMALL BREAK LOCA UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 13 April 2000 DCPP Unit 1
FIGURE 15.3-2 (Sheet 1 of 2)
RCS DEPRESSURIZATION 4-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013 DCPP Unit 2
FIGURE 15.3-2 (Sheet 2 of 2)
RCS DEPRESSURIZATION 4-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013 DCPP Unit 1
FIGURE 15.3-3 (Sheet 1 of 2) CORE MIXTURE ELEVATION 4-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013 DCPP Unit 2
FIGURE 15.3-3 (Sheet 2 of 2) CORE MIXTURE ELEVATION 4-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013 DCPP Unit 1
FIGURE 15.3-4 (Sheet 1 of 2) CLADDING TEMPERATURE TRANSIENT4-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 2
FIGURE 15.3-4 (Sheet 2 of 2) CLADDING TEMPERATURE TRANSIENT4-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013
10 0 10-1 10 4 10 3 10 2 10 1 10 0 10-1 10-2TIME AFTER S H U TD O WN (S E CO ND S)TOTAL RESIDUAL HEAT (WITH 4% SHUTDOWN MARGIN)WATTS/WATT AT POWER FIGURE 15.3-8 LOCA CORE POWER TRANSIENT UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 13 April 2000 DCPP Unit 1
FIGURE 15.3-9 (Sheet 1 of 2)
RCS DEPRESSURIZATION 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013
DCPP Unit 2
FIGURE 15.3-9 (Sheet 2 of 2)
RCS DEPRESSURIZATION 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 1
FIGURE 15.3-11 (Sheet 1 of 2) CORE MIXTURE ELEVATION 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013 DCPP Unit 2 FIGURE 15.3-11 (Sheet 2 of 2) CORE MIXTURE ELEVATION 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013
DCPP Unit 1
FIGURE 15.3-13 (Sheet 1 of 2) CLAD TEMPERATURE TRANSIENT 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013
DCPP Unit 2
FIGURE 15.3-13 (Sheet 2 of 2) CLAD TEMPERATURE TRANSIENT 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 4026-385 RPNMLKJHGFEDCBA-8.9-7.4 1-5.6-9.1-8.5 2-8.2-6.8-4.10.2 3-7.9-8.2-7.7 4-8.4-6.0-3.8-1.8 5-8.5-8.4-7.4-5.5-0.3 6-7.7-5.0-1.2-1.0 7-7.7-7.3-5.9-3.21.53.23.43.6 8-6.92.75.96.09-3.40.710.6 10-5.3-1.85.917.111.4111.312.324.6 120.10.77.723.6 132.54.711.117.6 142.16.515CASE A THE NUMBERS REPRESENT THE PERCENT DEVIATION FROM ASSEMBLY AVERAGE POWER FIGURE 15.3-15 INTERCHANGE BETWEEN REGION 1 AND REGION 3 ASSEMBLY UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 4026-386 RPNMLKJHGFEDCBA0.31.513.20.83.26.021.20.01.610.330.02.96.54-2.2-1.02.26.96.65-1.70.58.86-3.25.216.75.47-3.5-3.4-2.6-0.711.411.35.84.48-3.6-2.0-2.32.29-3.8-3.8-3.6-2.90.510-3.9-4.3-4.6-1.511-2.8-3.1-4.512-4.8-4.4-2.61.413-0.4-4.8-4.814-4.8-4.515CASE B-1 THE NUMBERS REPRESENT THE PERCENT DEVIATION FROM ASSEMBLY AVERAGE POWER.
FIGURE 15.3-16 INTERCHANGE BETWEEN REGION 1 AND REGION 2ASSEMBLY, BURNABLE POISON RODS BEING RETAINED BY THE REGION 2 ASSEMBLY UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 4026-387
THE NUMBERS REPRESENT THE PERCENT DEVIATION FROM ASSEMBLY AVERAGE POWER.
FIGURE 15.3-17 INTERCHANGE BETWEEN REGION 1 AND REGION 2ASSEMBLY, BURNABLE POISON RODS BEING TRANSFERRED TO THE REGION 1 ASSEMBLY UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 R PN ML K JHGFEDC BA 1.0 1.1 1 5.1 1.01.0 21.11.11.94.9 31.71.7 1.4 41.11.81.10.7 50.00.2 1.83.94.0 6 0.05.2 2.2-0.3 7-0.7-0.60.35.1 1.5-0.3-0.6-0.7 8-1.0-1.1-0.8-0.9 9-1.4-3.1-1.3 10-0.9-1.7-1.7-0.911-2.5-2.9-1.1 12 0.7 -1.9-2.92.5 13 2.3 -2.8-2.4-0.8 14-2.1-2.8 15CASE B-2 4026-388
THE NUMBERS REPRESENT THE PERCENT DEVIATION FROM ASSEMBLY AVERAGE POWER.
FIGURE 15.3-18 ENRICHMENT ERROR: A REGION 2 ASSEMBLYLOADED INTO THE CORE CENTRAL POSITIONUNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 RPNMLKJHGFEDCBA-2.2-2.112.0-2.0-2.12-1.5-1.6-1.02.0 3-0.9-1.0-0.4 4-0.41.2-0.5-1.4 5-2.1-1.62.35.7-2.0 6-3.29.74.4-1.7 7-2.3-1.61.813.65.6-0.4-1.6-2.1 8-2.29.71.1-2.290.34.5-0.910-1.9-0.41.8-0.5-1.911-0.9-0.6-1.1120.4-1.4-1.52.0 132.0-2.1-2.0-0.9 14-1.9-2.215CASE C 4026-389
THE NUMBERS REPRESENT THE PERCENT DEVIATION FROM ASSEMBLY AVERAGE POWER.
FIGURE 15.3-19 LOADING A REGION 2 ASSEMBLY INTO A REGION 1 POSITION NEAR CORE PERIPHERYUNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 RPNMLKJHGFEDCBA-11.0-14.010.4-9.2-12.02-12.0-14.0-15.0-13.0 33.21.2-11.0 4-1.5-12.0-15.0-16.0 59.87.1-1.6-8.0-16.0 69.2-2.3-12.0-14.0 720.017.810.80.8-10.0-14.0-15.0-16.0 827.2-5.5-11.0-15.0920.75.8-12.0 1042.023.61.9-8.6-13.01114.0-1.7-8.9 1238.620.42.8-7.0 1335.97.0-3.3-6.31415.32.9 15CASE D DCPP Unit 1
FIGURE 15.3-33 (Sheet 1 of 2) TOP CORE NODE VAPOR TEMPERATURE3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 2
FIGURE 15.3-33 (Sheet 2 of 2) TOP CORE NODE VAPOR TEMPERATURE3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 1
FIGURE 15.3-34 (Sheet 1 of 2) ROD FILM COEFFICIENT 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLOCANYON SITE FSAR UPDATERevision 21 September 2013 DCPP Unit 2
FIGURE 15.3-34 (Sheet 2 of 2) ROD FILM COEFFICIENT 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013
DCPP Unit 1
FIGURE 15.3-35 (Sheet 1 of 2) HOT SPOT FLUID TEMPERATURE3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 2
FIGURE 15.3-35 (Sheet 2 of 2) HOT SPOT FLUID TEMPERATURE3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 1 FIGURE 15.3-36 (Sheet 1 of 2) BREAK MASS FLOW 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 2 FIGURE 15.3-36 (Sheet 2 of 2) BREAK MASS FLOW 3-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 1
FIGURE 15.3-37 (Sheet 1 of 2)
RCS DEPRESSURIZATION 2-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 2
FIGURE 15.3-37 (Sheet 2 of 2)
RCS DEPRESSURIZATION 2-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 1
FIGURE 15.3-38 (Sheet 1 of 2) CORE MIXTURE ELEVATION 2-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 2
FIGURE 15.3-38 (Sheet 2 of 2) CORE MIXTURE ELEVATION 2-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 1
FIGURE 15.3-39 (Sheet 1 of 2) CLADDING TEMPERATURE TRANSIENT2-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 2
FIGURE 15.3-39 (Sheet 2 of 2) CLADDING TEMPERATURE TRANSIENT2-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 DCPP Unit 1 DCPP Unit 2 FIGURE 15.3-40 RCS DEPRESSURIZATION 6-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013 DCPP Unit 1 DCPP Unit 2 FIGURE 15.3-41 CORE MIXTURE ELEVATION 6-INCH COLD LEG BREAK UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013
Note: The results for Unit1 are nearly identical to those for Unit 2; therefore, the figures for Unit 1 are representative of the results for Unit 2.Revision 11 November 1996Core Coolant Mass Flow Rate (Fraction of Nominal) Time (s) 1 0.9 0.8 0.7 0.6 0.5 0.4 COMPLETE LOSS OF FORCEDREACTORCOOLANTFLOWRevision23December2016
Note: The results for Unit1 are nearly identical to those for Unit 2; therefore, thefigures for Unit 1 are representative of the results for Unit 2.1.0 0.9 0.8 0.7 0.6 0.5 0.4 0.3 1.1 0 2 4 6 8 10 Hot Channel Heat Flux (Fraction of Nominal) Time (s) Revision23December2016 COMPLETE LOSS OF FORCEDREACTORCOOLANTFLOW
Note: The results for Unit1 are nearly identical to those for Unit 2; therefore, thefigures for Unit 1 are representative of the results for Unit 2.Time (s) Nuclear Power (Fraction of Nominal)
COMPLETE LOSS OF FORCEDREACTORCOOLANTFLOWRevision23December2016
Note: The results for Unit1 are nearly identical to those for Unit 2; therefore, thefigures for Unit 1 are representative of the results for Unit 2.
Time (s) DNBR (Thimble Cell)
COMPLETE LOSS OF FORCEDREACTORCOOLANTFLOWRevision23December2016
FIGURE 15.4.1-1A REFERENCE TRANSIENT PCT AND PCT LOCATION UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-1B LIMITING PCT CASE AND PCT LOCATION UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-2A REFERENCE TRANSIENT VESSEL SIDE BREAK FLOW UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-2B LIMITING PCT CASE VESSEL SIDE BREAK FLOW UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-3A REFERENCE TRANSIENT LOOP SIDE BREAK FLOW UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-3B LIMITING PCT CASE LOOP SIDE BREAK FLOW UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-4A REFERENCE TRANSIENT BROKEN AND INTACT LOOP PUMP VOID FRACTION UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-4B LIMITING PCT CASE BROKEN AND INTACT LOOP PUMP VOID FRACTION UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-5A REFERENCE TRANSIENT HOT ASSEMBLY/TOP OF CORE VAPOR FLOW UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-5B LIMITING PCT CASE HOT ASSEMBLY/TOP OF CORE VAPOR FLOW UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-6A REFERENCE TRANSIENT PRESSURIZER PRESSURE UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-6B LIMITING PCT CASE PRESSURIZER PRESSURE UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-7A REFERENCE TRANSIENT LOWER PLENUM COLLAPSED LIQUID LEVEL UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-7B LIMITING PCT CASE LOWER PLENUM COLLAPSED LIQUID LEVEL UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-8A REFERENCE TRANSIENT VESSEL WATER MASS UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-8B LIMITING PCT CASE VESSEL FLUID MASS UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-9A REFERENCE TRANSIENT LOOP 1 ACCUMULATOR FLOW UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-9B LIMITING PCT CASE LOOP 1 ACCUMULATOR FLOW UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-10A REFERENCE TRANSIENT LOOP 1 SAFETY INJECTION FLOW UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-10B LIMITING PCT CASE LOOP 1 SAFETY INJECTION FLOW UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-11A REFERENCE TRANSIENT CORE AVERAGE CHANNEL COLLAPSED LIQUID LEVEL UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-11B LIMITING PCT CASE CORE AVERAGE CHANNEL COLLAPSED LIQUID LEVEL UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-12A REFERENCE TRANSIENT LOOP 1 DOWNCOMER COLLAPSED LIQUID LEVEL UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-12B LIMITING PCT CASE LOOP 1 DOWNCOMER COLLAPSED LIQUID LEVEL UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-13A TOTAL ECCS FLOW (3 LINES INJECTING)
UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
05001000150020002500 3000 3500 40000200400600800100012001400160018002000Pressure (psia)Total SI Flow (gpm)
FIGURE 15.4.1-13B TOTAL ECCS FLOW (3 LINES INJECTING)
UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008 FIGURE 15.4.1-14A REFERENCE TRANSIENT PRESSURE TRANSIENT UNIT 1 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-14B LOWER BOUND COCO CONTAINMENT PRESSURE TRANSIENT UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
FIGURE 15.4.1-15A AXIAL POWER DISTRIBUTION LIMITS UNIT 1 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013 0.46, 0.20.46, 0.37250.28, 0.450.28, 0.30.150.20.250.30.350.40.450.50.250.30.350.40.450.5 PMID PBOT
FIGURE 15.4.1-15B AXIAL POWER DISTRIBUTION LIMITS UNIT 2 DIABLO CANYON SITE FSAR UPDATE Revision 18 October 2008
-1500-1000-500 00.000.050.100.150.200.250.30Power (fraction of nominal)Integral of Power Coefficient (pcm
)
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.4.2-1 RUPTURE OF A MAIN STEAM LINE VARIATION OF REACTIVITY WITH POWER AT CONSTANT CORE AVERAGE TEMPERATURE Revision 19 May 2010 Zero Power 1050 psia EOL Rodded Core One RCCA Stuck Full Out
FIGURE 15.4.2-2 RUPTURE OF A MAIN STEAM LINE VARIATION OF K EFF WITH CORE AVERAGE TEMPERATURE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
FIGURE 15.4.2-3 RUPTURE OF A MAIN STEAM LINESAFETY INJECTION CURVE UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 21 September 2013
FIGURE 15.4.2-4 RUPTURE OF A MAIN STEAM LINE WITH OFFSITE POWER AVAILABLE CORE HEAT FLUX AND STEAM FLOW TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
FIGURE 15.4.2-5 RUPTURE OF A MAIN STEAM LINE WITH OFFSITE POWER AVAILABLE LOOP AVERAGE TEMPERATURE AND REACTOR COOLANT PRESSURE TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-6 RUPTURE OF A MAIN STEAM LINE WITH OFFSITE POWER AVAILABLE REACTIVITY AND CORE BORON TRANSIENTS FSAR UPDATE Revision 19 May 2010 UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-7 RUPTURE OF A MAIN STEAM LINE WITHOUT OFFSITE POWER AVAILABLE CORE HEAT FLUX AND STEAM FLOW TRANSIENTS FSAR UPDATE Revision 19 May 2010 UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-8 RUPTURE OF A MAIN STEAM LINE WITHOUT OFFSITE POWER AVAILABLE LOOP AVERAGE TEMPERATURE AND REACTOR COOLANT PRESSURE TRANSIENTS FSAR UPDATE Revision 19 May 2010 UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-9 RUPTURE OF A MAIN STEAM LINE WITHOUT OFFSITE POWER AVAILABLE REACTIVITY AND CORE BORON TRANSIENTS FSAR UPDATE Revision 19 May 2010 FIGURE 15.4.2-10 MAIN FEEDLINE RUPTURE WITH OFFSITE POWER AVAILABLE NUCLEAR POWER AND CORE HEAT FLUX TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
FIGURE 15.4.2-11 MAIN FEEDLINE RUPTURE WITH OFFSITE POWER AVAILABLE PRESSURIZER PRESSURE AND WATER VOLUME TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
FIGURE 15.4.2-12 MAIN FEEDLINE RUPTURE WITH OFFSITE POWER AVAILABLE REACTOR COOLANT TEMPERATURE TRANSIENTS FOR THE FAULTED AND INTACT LOOPS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.4.2-13 MAIN FEEDLINE RUPTURE WITH OFFSITE POWER AVAILABLE STEAM GENERATOR PRESSURE AND TOTAL MASS TRANSIENTS Revision 19 May 2010 FIGURE 15.4.2-14 MAIN FEEDLINE RUPTURE WITHOUT OFFSITE POWER AVAILABLE NUCLEAR POWER AND CORE HEAT FLUX TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
FIGURE 15.4.2-15 MAIN FEEDLINE RUPTURE WITHOUT OFFSITE POWER AVAILABLE PRESSURIZER PRESSURE AND WATER VOLUME TRANSIENTS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
FIGURE 15.4.2-16 MAIN FEEDLINE RUPTURE WITHOUT OFFSITE POWER AVAILABLE REACTOR COOLANT TEMPERATURE TRANSIENTS FOR THE FAULTED AND INTACT LOOPS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 19 May 2010
UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.4.2-17 MAIN FEEDLINE RUPTURE WITHOUT OFFSITE POWER AVAILABLE STEAM GENERATOR PRESSURE AND TOTAL MASS TRANSIENTS Revision 19 May 2010 UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.4.2-18 MAIN STEAM LINE RUPTURE AT FULL POWER, 0.49 ft 2 BREAK NUCLEAR POWER AND CORE HEAT FLUX TRANSIENTS Revision 19 Ma y 2010 UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.4.2-19 MAIN STEAM LINE RUPTURE AT FULL POWER, 0.49 ft 2 BREAK PRESSURIZER PRESSURE AND WATER VOLUME TRANSIENTS Revision 19 Ma y 2010 UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.4.2-20 MAIN STEAM LINE RUPTURE AT FULL POWER, 0.49 ft 2 BREAK REACTOR VESSEL INLET TEMPERATURE AND LOOP AVERAGE TEMPERATURE TRANSIENTS Revision 19 Ma y 2010 UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE FIGURE 15.4.2-21 MAIN STEAM LINE RUPTURE AT FULL POWER, 0.49 ft 2 BREAK TOTAL STEAM FLOW AND STEAM PRESSURE TRANSIENTS Revision 19 Ma y 2010 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-22 MAIN FEEDLINE RUPTURE FOR PRESSURIZER FILLING (UNBLOCK PRESSURIZER PORV) PRESSURIZER PRESSURE AND WATER VOLUME TRANSIENTS Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-23 MAIN FEEDLINE RUPTURE FOR PRESSURIZER FILLING (UNBLOCK PRESSURIZER PORV) PSV RELIEF FLOW RATE AND ENTHALPY TRANSIENTS Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-24 MAIN FEEDLINE RUPTURE FOR PRESSURIZER FILLING (ISOLATE CHARGING AND STOP RCP SEAL INJECTION FLOW) PRESSURIZER PRESSURE AND WATER VOLUME TRANSIENTS Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-25 MAIN FEEDLINE RUPTURE FOR PRESSURIZER FILLING (ISOLATE CHARGING AND STOP RCP SEAL INJECTION FLOW) PORV AND PSV RELIEF FLOW RATE TRANSIENTS Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-26 MAIN FEEDLINE RUPTURE FOR PRESSURIZER FILLING (ISOLATE CHARGING AND STOP RCP SEAL INJECTION FLOW) PORV/PSV RELIEF FLOW ENTHALPY AND TOTAL NUMBER OF PORV CYCLES TRANSIENTS Revision23December2016
FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITE FIGURE 15.4.2-27 MAIN FEEDLINE RUPTURE FOR PRESSURIZER FILLING (ISOLATE CHARGING AND STOP RCP SEAL INJECTION FLOW) COLD LEG INJECTION FLOW RATE TRANSIENT Revision23December2016 FIGURE 15.4.3-1A PRESSURIZER LEVEL SGTR MTO ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 FIGURE 15.4.3-1B PRESSURIZER LEVEL SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-2A PRESSURIZER PRESSURE SGTR MTO ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 FIGURE 15.4.3-2B PRESSURIZER PRESSURE SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-3A SECONDARY PRESSURE SGTR MTO ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 FIGURE 15.4.3-3B SECONDARY PRESSURE SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-4A INTACT LOOP HOT AND COLD LEG RCS TEMPERATURES SGTR MTO ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 FIGURE 15.4.3-4B INTACT LOOP HOT AND COLD LEG RCS TEMPERATURES SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-5B RUPTURED LOOP HOT AND COLDLEG RCS TEMPERATURES SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-6A PRIMARY TO SECONDARY BREAK FLOW RATE SGTR MTO ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 FIGURE 15.4.3-6B PRIMARY TO SECONDARY BREAK FLOW RATE SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-7A RUPTURED STEAM GENERATOR WATER VOLUME SGTR MTO ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 FIGURE 15.4.3-7B RUPTURED STEAM GENERATOR WATER VOLUME SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-8A RUPTURED STEAM GENERATOR WATER MASS SGTR MTO ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATERevision 20 November 2011 FIGURE 15.4.3-8B RUPTURED STEAM GENERATOR WATER MASS SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-9 RUPTURED SG MASS RELEASE RATETO THE ATMOSPHERE SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FIGURE 15.4.3-10 INTACT SGs MASS RELEASE RATETO THE ATMOSPHERE SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013
FIGURE 15.4.3-11 TOTAL FLASHED BREAK FLOW SGTR DOSE INPUT ANALYSIS UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 21 September 2013 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITEFIGURE 15.4.4-1 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR MAXIMUM RCS PRESSURE VS. TIME Comment [ F1]: CN-TA-04-116, page 160Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITEFIGURE 15.4.4-2 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR CLAD AVERAGE TEMPERATURE VS. TIME Comment [ F2]: CN-TA-04-116, page 162Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITEFIGURE 15.4.4-3 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR CORE FLOW VS. TIMEComment [ F3]: CN-TA-04-116, page 164Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITEFIGURE 15.4.4-4 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR HOT CHANNEL HEAT FLUX VS. TIME Comment [ F4]: CN-TA-04-116, page 166Revision23December2016 FSAR UPDATE UNITS 1 AND 2 DIABLO CANYON SITEFIGURE 15.4.4-5 SINGLE REACTOR COOLANT PUMP LOCKED ROTOR NUCLEAR POWER VS. TIMEComment [ F5]: CN-TA-04-116, page 168Revision23December2016 FIGURE 15.4.6-1 NUCLEAR POWER TRANSIENT, BOL, HZP, ROD EJECTION ACCIDENT UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.4.6-2 HOT SPOT FUEL AND CLAD TEMPERATURES VERSUS TIME, BOL, HZP, ROD EJECTION ACCIDENT UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.4.6-3 NUCLEAR POWER TRANSIENT, EOL, HFP, ROD EJECTION ACCIDENT UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 FIGURE 15.4.6-4 HOT SPOT FUEL AND CLAD TEMPERATURES VERSUS TIME, EOL, HZP, ROD EJECTION ACCIDENT UNITS 1 AND 2 DIABLO CANYON SITE FSAR UPDATE Revision 11 November 1996 Ratio of Short-term Release Concentration to Continuous Release Concentration vs. Release Duration DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-1 Revision 11 November 1996 Thyroid Dose at 800 Meters Verses Weight of Steam Dumped to Atmosphere (Design Basis Case Assumptions)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-2 Revision 11 November 1996 Thyroid Dose at 10,000 Meters Verses Weight of Steam Dumped to Atmosphere (Design Basis Case Assumptions)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-3 Revision 11 November 1996 Thyroid Dose at 10,000 Meters Verses Weight of Steam Dumped to Atmosphere (Expected Case Assumptions)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-4 Revision 11 November 1996 Thyroid Dose at 800 Meters Verses Weight of Steam Dumped to Atmosphere (Expected Case Assumptions)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-5 Revision 11 November 1996 Thyroid Exposures for 15 Percent Nonremovable Iodine (Normalized to Exposures for Zero Spray Removal Constant)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-6 Revision 11 November 1996 Historical Revision 22 May 2015 DBA 2-Hour 800-Meter Thyroid Exposures Verses Spray Removal Constant and Percent Nonremovable Iodine (Normalized to Exposures with Zero Spray Removal Constant and Zero Percent Nonremovable Iodine)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-7 Revision 11 November 1996 Historical Revision 22 May 2015 DBA 30-Hour 800-Meter Thyroid Exposures Verses Spray Removal Constant and Percent Nonremovable Iodine (Normalized to Exposures with Zero Spray Removal Constant and Zero Percent Nonremovable Iodine)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-8 Revision 11 November 1996 Historical Revision 22 May 2015 Containment Recirculation Sump Activity Pathway to the Atmosphere for Small Leak Case DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-9 Revision 11 November 1996 Containment Recirculation Sump Activity Pathway to the Atmosphere for Large Leak Case DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-10 Revision 11 November 1996 Equilibrium Elemental Iodine Partition and Decontamination Factors for the Expected Case - Large Circulation Loop Leakage in the Auxiliary Building DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-11 Revision 11 November 1996 Equilibrium Elemental Iodine Partition and Decontamination Factors for the DBA Case - Large Circulation Loop Leakage in the Auxiliary Building DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-12 Revision 11 November 1996 Potential Radiation Exposures as a Result of Accidents Involving Failure of Fuel Cladding (Design Basis Case Assumptions)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-14 Revision 11 November 1996 Potential Radiation Exposures as a Result of Accidents Involving Failure of Fuel Cladding (Expected Case Assumptions)
DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-15 Revision 11 November 1996 Incremental Long-term Doses From Accidents Involving Failure of Fuel Cladding DIABLO CANYON UNITS 1 & 2 FIGURE 15.5-16 Revision 11 November 1996 Historical Revision 22 May 2015