ML19150A497

8 to Updated Final Safety Analysis Report, Chapter 15, Table of Contents

ML19150A497
Person / Time
Site:	Ginna
Issue date:	05/10/2019
From:	Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
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Page 1 of 24 Revision 28 5/2019 15 ACCIDENT ANALYSES 1

15.0 GENERAL 2

15.0.1 INITIAL CONDITIONS 2

15.0.1.1 Assumed Values of Initial Conditions 2

15.0.2 POWER DISTRIBUTION 3

15.0.3 REACTIVITY COEFFICIENTS ASSUMED IN THE ACCIDENT ANALYSES 4

15.0.4 ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS 4

15.0.5 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN THE ACCIDENT ANALYSES 4

15.0.6 INSTRUMENTATION DRIFT AND CALORIMETRIC ERRORS -

POWER RANGE NEUTRON FLUX 5

15.0.7 COMPUTER CODES 5

15.0.7.1 FACTRAN 6

15.0.7.2 RETRAN 6

15.0.7.3 TWINKLE 6

15.0.7.4 VIPRE 7

15.0.7.5 ADVANCED NODAL CODE (ANC) 7 15.0.8 CLASSIFICATION OF PLANT CONDITIONS 7

15.0.8.1 Condition I - Normal Operation 8

15.0.8.2 Condition II - Faults of Moderate Frequency 8

15.0.8.3 Condition III - Infrequent Faults 8

15.0.8.4 Condition IV - Limiting Faults 8

15.0.9 UFSAR Re-write 9

15.0.9.1 General Layout 9

15.0.9.2 Interpretation of Operator Action Times 9

15.0 REFERENCES

FOR SECTION 15.0 10 Table 15.0-1 NSSS PCWG Parameters for Ginna Station Uprate Program 11 Table 15.0-1 NSSS PCWG Parameters for Ginna Station Uprate Program 12 Table 15.0-2 Non-LOCA Analysis Limits and Analysis Results 13 Table 15.0-3 Non-LOCA Plant Initial Condition Assumptions 16 Table 15.0-4 Pressurizer and Main Steam System (MSS) Pressure Relief Assumptions 17 Table 15.0-5 Core Kinetics Parameters and Reactivity Feedback Coefficients 21 Table 15.0-6 Summary of RPS and ESFAS Functions Actuated 22

Page 2 of 24 Revision 28 5/2019 Table 15.0-7 Overtemperature and Overpower T Setpoints 25 Table 15.0-8 DETERMINATION OF MAXIMUM OVERPOWER TRIP POINT

- POWER RANGE NEUTRON FLUX CHANNEL - BASED ON NOMINAL SETPOINT CONSIDERING INHERENT INSTRU-MENT ERRORS 26 Table 15.0-9 Summary of Initial Conditions and Computer Codes Used 27 15.1 INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 30 15.1.1 DECREASE IN FEEDWATER TEMPERATURE 30 15.1.1.1 Description of Event 30 15.1.1.2 Frequency of Event 31 15.1.1.3 Event Analysis 31 15.1.1.3.1 Protective Features 31 15.1.1.3.2 Single Failures Assumed 31 15.1.1.3.3 Operator Actions Assumed 31 15.1.1.3.4 Chronological Description of Event 31 15.1.1.3.5 Impact on Fission Product Barriers 31 15.1.1.4 Reactor Core and Plant System Evaluation 32 15.1.1.4.1 Input Parameters and Initial Conditions 32 15.1.1.4.2 Methodology 32 15.1.1.4.3 Acceptance Criteria 32 15.1.1.4.4 Results 32 15.1.1.5 Radiological Consequences 32 15.1.1.6 Conclusions 32 15.1.2 INCREASE IN FEEDWATER FLOW 33 15.1.2.1 Increase in Feedwater Flow at Full Power 33 15.1.2.1.1 Description of Event 33 15.1.2.1.2 Frequency of Event 33 15.1.2.1.3 Event Analysis 33 15.1.2.1.3.1 Protective Features 34 15.1.2.1.3.2 Single Failures Assumed 34 15.1.2.1.3.3 Operator Actions Assumed 35 15.1.2.1.3.4 Chronological Description of Event 35 15.1.2.1.3.5 Impact on Fission Product Barriers 35 15.1.2.1.4 Reactor Core and Plant System Evaluation 35 15.1.2.1.4.1 Input Parameters and Initial Conditions 35 15.1.2.1.4.2 Method of Analysis 36

Page 3 of 24 Revision 28 5/2019 15.1.2.1.4.3 Acceptance Criteria 36 15.1.2.1.4.4 Results 36 15.1.2.1.5 Radiological Consequences 37 15.1.2.1.6 Conclusion 37 15.1.2.2 Increase in Feedwater Flow at Zero Power 37 15.1.2.2.1 Description of Event 37 15.1.2.2.2 Frequency of Event 37 15.1.2.2.3 Event Analysis 37 15.1.2.2.3.1 Protective Features 37 15.1.2.2.3.2 Single Failures Assumed 38 15.1.2.2.3.3 Operator Actions Assumed 38 15.1.2.2.3.4 Chronological Description of Event 38 15.1.2.2.3.5 Impact on Fission Product Barriers 38 15.1.2.2.4 Reactor Core and Plant System Evaluation 38 15.1.2.2.4.1 Input Parameters and Initial Conditions 38 15.1.2.2.4.2 Methodology 39 15.1.2.2.4.3 Acceptance Criteria 39 15.1.2.2.5 Radiological Consequences 39 15.1.2.2.6 Conclusion 39 15.1.3 EXCESSIVE LOAD INCREASE INCIDENT 39 15.1.3.1 Description of Event 39 15.1.3.2 Frequency of Event 40 15.1.3.3 Event Analysis 40 15.1.3.3.1 Protective Features 40 15.1.3.3.2 Single Failures Assumed 41 15.1.3.3.3 Operator Actions Assumed 41 15.1.3.3.4 Chronological Description of Event 41 15.1.3.3.5 Impact on Fission Product Barriers 41 15.1.3.4 Reactor Core and Plant System Evaluation 41 15.1.3.4.1 Input Parameters and Initial Conditions 41 15.1.3.4.2 Methodology 42 15.1.3.4.3 Acceptance Criteria 42 15.1.3.5 Radiological Consequences 42 15.1.3.6 Conclusions 42 15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR RELIEF/

43 SAFETY VALVE 15.1.5 SPECTRUM OF STEAM SYSTEM PIPING FAILURES INSIDE AND 43 OUTSIDE OF CONTAINMENT

Page 4 of 24 Revision 28 5/2019 15.1.5.1 Description of Event 43 15.1.5.2 Frequency of Event 43 15.1.5.3 Event Analysis 44 15.1.5.3.1 Protective Features 44 15.1.5.3.2 Single Failures Assumed 45 15.1.5.3.3 Operator Actions Assumed 45 15.1.5.3.4 Chronological Description of Event 45 15.1.5.3.5 Impact on Fission Product Barriers 45 15.1.5.4 Reactor Core and Plant System Evaluation 46 15.1.5.4.1 Input Parameters and Initial Conditions 46 15.1.5.4.2 Methodology 47 15.1.5.4.3 Acceptance Criteria 48 15.1.5.4.4 Results 48 15.1.5.5 Radiological Consequences 49 15.1.5.6 Conclusions 50 15.1.5.7 Supplemental Evaluations 50 15.1.5.7.1 SEV-1073 50 15.1.5.7.2 HZP 6 Inch Steamline Break 50 15.1.5.7.3 High Steam Flow Setpoint Increase Evaluation 50 15.1.5.7.4 Steamline Rupture a Full Power 51 15.1.5.8 Potential for Containment Overpressurization 51 15.1.6 COMBINED STEAM GENERATOR ATMOSPHERIC RELIEF VALVE (ARV) AND MAIN FEEDWATER REGULATING VALVE (MFRV)

FAILURES 51 15.1.6.1 Description of Event 51 15.1.6.2 Frequency of Event 52 15.1.6.3 Event Analysis 52 15.1.6.3.1 Protective Features 53 15.1.6.3.2 Single Failures Assumed 53 15.1.6.3.3 Operator Actions Assumed 53 15.1.6.3.4 Chronological Description of Event 54 15.1.6.3.5 Impact on Fission Product Barriers 54 15.1.6.4 Reactor Core and Plant System Evaluation 54 15.1.6.4.1 Input Parameters and Initial Conditions 54 15.1.6.4.2 Methodology 55 15.1.6.4.3 Acceptance Criteria 56 15.1.6.4.4 Results 56 15.1.6.5 Radiological Consequences 56

Page 5 of 24 Revision 28 5/2019 15.1.6.6 Conclusions 57

15.1 REFERENCES

FOR SECTION 15.1 58 Table 15.1-1 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC-TION TRANSIENTS HOT FULL POWER - SINGLE LOOP - WITH ROD CONTROL 59 Table 15.1-2 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC-TION TRANSIENTS HOT FULL POWER - SINGLE LOOP -

WITHOUT ROD CONTROL 60 Table 15.1-3 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC-TION TRANSIENTS HOT FULL POWER - MULTI LOOP -

WITH ROD CONTROL 61 Table 15.1-4 TIME SEQUENCE OF EVENTS FOR FEEDWATER MALFUNC-TION TRANSIENTS HOT FULL POWER - MULTI LOOP -

WITHOUT ROD CONTROL 62 Table 15.1-5 Table DELETED 63 Table 15.1-6 TIME SEQUENCE OF EVENTS FOR STEAM LINE RUPTURE 64 Table 15.1-7

SUMMARY

OF MAIN FEEDWATER REGULATING VALVES (MFRV)/STEAM GENERATOR ATMOSPHERIC RELIEF VALVE (ARV) COMBINATION FAILURE CASES EVALUATED 66 Table 15.1-8 MSLB DOSE ANALYSIS ASSUMPTIONS 67 Table 15.1-9 RESULTS FOR MAIN STEAM LINE BREAK, REM TEDE 69 Table 15.1-10 TIME SEQUENCE OF EVENTS FOR THE COMBINED FAILURE 70 OF TWO MFRV's AND TWO ARV's AT HOT FULL POWER 15.2 DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 71 15.2.1 STEAM PRESSURE REGULATOR MALFUNCTION OR FAILURE THAT RESULTS IN DECREASING STEAM FLOW 71 15.2.2 LOSS OF EXTERNAL ELECTRICAL LOAD 71 15.2.2.1 Description of Event 71 15.2.2.2 Frequency of Event 71 15.2.2.3 Event Analysis 71 15.2.2.3.1 Protective Features 72 15.2.2.3.2 Single Failures Assumed 72 15.2.2.3.3 Operator Actions Assumed 72 15.2.2.3.4 Chronological Description of Event 73 15.2.2.3.5 Impact on Fission Product Barriers 73 15.2.2.4 Reactor Core and Plant System Evaluation 73 15.2.2.4.1 Input Parameters and Initial Conditions 73 15.2.2.4.2 Method of Analysis 74 15.2.2.4.3 Acceptance Criteria 75 15.2.2.4.4 Results 75 15.2.2.5 Radiological Consequences 76

Page 6 of 24 Revision 28 5/2019 15.2.2.6 Conclusions 76 15.2.2.7 Supplemental Evaluations 76 15.2.3 TURBINE TRIP 76 15.2.4 LOSS OF CONDENSER VACUUM 76 15.2.5 LOSS OF ALL ALTERNATING CURRENT POWER TO THE STATION AUXILIARIES 77 15.2.5.1 Description of the event 77 15.2.5.2 Frequency of Event 77 15.2.5.3 Event Analysis 78 15.2.5.3.1 Protective Features 78 15.2.5.3.2 Single Failures Assumed 79 15.2.5.3.3 Operator Actions Assumed 79 15.2.5.3.4 Chronological Description of Event 79 15.2.5.3.5 Impact on Fission Product Barriers 79 15.2.5.4 Reactor Core and Plant System Evaluation 80 15.2.5.4.1 Input Parameters and Initial Conditions 80 15.2.5.4.2 Method of Analysis 81 15.2.5.4.3 Acceptance Criteria 81 15.2.5.4.4 Results 82 15.2.5.5 Radiological Consequences 82 15.2.5.6 Conclusions 82 15.2.5.7 Supplemental Evaluations 83 15.2.6 LOSS OF NORMAL FEEDWATER FLOW 83 15.2.6.1 Description of Event 83 15.2.6.2 Frequency of Event 84 15.2.6.3 Event Analysis 84 15.2.6.3.1 Protective Features 84 15.2.6.3.2 Single Failures Assumed 85 15.2.6.3.3 O

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Operator Actions Assumed 85 15.2.6.3.4 Chronological Description of Event 85 15.2.6.3.5 Impact on Fission Product Barriers 85 15.2.6.4 Reactor Core and Plant System Evaluation 85 15.2.6.4.1 Input Parameters and Initial Conditions 85 15.2.6.4.2 Method of Analysis 87 15.2.6.4.3 Acceptance Criteria 87 15.2.6.4.4 Results 87 15.2.6.5 Radiological Consequences 88 15.2.6.6 Conclusions 88

Page 7 of 24 Revision 28 5/2019 15.2.6.7 Supplemental Evaluations 88 15.2.7 FEEDWATER SYSTEM PIPE BREAKS 89 15.2.7.1 Description of Event 89 15.2.7.2 Frequency of Event 89 15.2.7.3 Event Analysis 89 15.2.7.3.1 Protective Features 89 15.2.7.3.2 Single Failures Assumed 90 15.2.7.3.3 Operator Actions Assumed 91 15.2.7.3.4 Chronological Description of Event 91 15.2.7.3.5 Impact on Fission Product Barriers 91 15.2.7.4 Reactor Core and Plant System Evaluation 91 15.2.7.4.1 Input Parameters and Initial Conditions 91 15.2.7.4.2 Method of Analysis 93 15.2.7.4.3 Acceptance Criteria 93 15.2.7.4.4 Results 94 15.2.7.5 Radiological Consequences 95 15.2.7.6 Conclusions 95

15.2 REFERENCES

FOR SECTION 15.2 96 Table 15.2-1 TIME SEQUENCE OF EVENTS FOR LOSS OF EXTERNAL ELEC-TRICAL LOAD 97 Table 15.2-2 TIME SEQUENCE OF EVENTS FOR LOSS OF OFFSITE ALTER-NATING CURRENT POWER TO THE STATION AUXILIARIES 99 Table 15.2-3 Table DELETED 100 Table 15.2-4 TIME SEQUENCE OF EVENTS FOR LOSS OF NORMAL FEEDWATER FLOW 101 Table 15.2-5 TIME SEQUENCE OF EVENTS FOR THE FEEDWATER LINE PIPE BREAK (0.3 FT2 BREAK AREA) 102 15.3 DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE 103 15.3.1 FLOW COASTDOWN ACCIDENTS 103 15.3.1.1 Description of Event 103 15.3.1.2 Frequency of Event 103 15.3.1.3 Event Analysis 103 15.3.1.3.1 Protective Features 104 15.3.1.3.2 Single Failures Assumed 105 15.3.1.3.3 Operator Actions Assumed 105 15.3.1.3.4 Chronological Description of Event 105 15.3.1.3.5 Impact on Fission Product Barriers 105 15.3.1.4 Reactor Core and Plant System Evaluation 105 15.3.1.4.1 Input Parameters and Initial Conditions 105

Page 8 of 24 Revision 28 5/2019 15.3.1.4.2 Method of Analysis 106 15.3.1.4.3 Acceptance Criteria 106 15.3.1.4.4 Results 106 15.3.1.5 Radiological Consequences 107 15.3.1.6 Conclusions 107 15.3.2 LOCKED ROTOR ACCIDENT 108 15.3.2.1 Description of Event 108 15.3.2.2 Frequency of Event 108 15.3.2.3 Event Analysis 108 15.3.2.3.1 Protective Features 108 15.3.2.3.2 Single Failures Assumed 109 15.3.2.3.3 Operator Actions Assumed 109 15.3.2.3.4 Chronological Description of Event 109 15.3.2.3.5 Impact on Fission Product Barriers 109 15.3.2.4 Reactor Core and Plant System Evaluation 110 15.3.2.4.1 Input Parameters and Initial Conditions 110 15.3.2.4.2 Method of Analysis 110 15.3.2.4.3 Acceptance Criteria 111 15.3.2.4.4 Results 112 15.3.2.5 Radiological Consequences 112 15.3.2.6 Conclusions 112

15.3 REFERENCES

FOR SECTION 15.3 113 Table 15.3-1 TIME SEQUENCE OF EVENTS FOR LOSS OF REACTOR COOL-ANT FLOW 114 Table 15.3-2

SUMMARY

OF LIMITING RESULTS FOR LOCKED ROTOR ACCIDENT 115 Table 15.3-3 TIME SEQUENCE OF EVENTS FOR LOCKED ROTOR INCIDENT 116 Table 15.3-4 LR Dose Analysis Assumptions 117 Table 15.13-5 RESULTS FOR LOCKED ROTOR 118 15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES 119 15.4.1 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITH-DRAWAL FROM A SUBCRITICAL CONDITION 119 15.4.1.1 Description of Event 119 15.4.1.2 Frequency of Event 119 15.4.1.3 Event Analysis 119 15.4.1.3.1 Protective Features 119 15.4.1.3.2 Single Failures Assumed 120 15.4.1.3.3 Operator Actions Assumed 120 15.4.1.3.4 Chronological Description of Event 120 15.4.1.3.5 Impact on Fission Product Barriers 120

Page 9 of 24 Revision 28 5/2019 15.4.1.4 Reactor Core and Plant System Evaluation 120 15.4.1.4.1 Input Parameters and Initial Conditions 120 15.4.1.4.2 Methodology 122 15.4.1.4.3 Acceptance Criteria 122 15.4.1.4.4 Results 122 15.4.1.5 Radiological Evaluation 123 15.4.1.6 Conclusions 123 15.4.2 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITH-DRAWAL AT POWER 123 15.4.2.1 Description of Event 123 15.4.2.2 Frequency of Event 123 15.4.2.3 Event Analysis 123 15.4.2.3.1 Protective Features 124 15.4.2.3.2 Single Failures Assumed 124 15.4.2.3.3 Operator Actions Assumed 124 15.4.2.3.4 Chronological Description of Event 124 15.4.2.3.5 Impact on Fission Product Barriers 124 15.4.2.4 Reactor Core and Plant System Evaluation 125 15.4.2.4.1 Input Parameters and Initial Conditions 125 15.4.2.4.2 Methodology 126 15.4.2.4.3 Acceptance Criteria 126 15.4.2.4.4 Results 127 15.4.2.5 Radiological Evaluation 128 15.4.2.6 Conclusions 128 15.4.3 STARTUP OF AN INACTIVE REACTOR COOLANT LOOP 129 15.4.3.1 Description of Event 129 15.4.3.2 Frequency of Event 129 15.4.3.3 Event Analysis 129 15.4.3.3.1 Protective Features 129 15.4.3.3.2 Single Failures Assumed 129 15.4.3.3.3 Operator Actions Assumed 129 15.4.3.3.4 Chronological Description of Event 130 15.4.3.3.5 Impact on Fission Product Barriers 130 15.4.3.4 Reactor Core and Plant System Evaluation 130 15.4.3.4.1 Input Parameters and Initial Conditions 130 15.4.3.4.2 Methodology 130 15.4.3.4.3 Acceptance Criteria 131

Page 10 of 24 Revision 28 5/2019 15.4.3.4.4 Results 131 15.4.3.4.5 Effect of 18 Month Fuel Cycle Changes 132 15.4.3.5 Radiological Evaluation 132 15.4.3.6 Conclusions 132 15.4.4 CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION 132 15.4.4.1 Description of Event 132 15.4.4.2 Frequency of Event 133 15.4.4.3 Event Analysis 133 15.4.4.3.1 Protective Features and Single Failures Assumed 133 15.4.4.3.1.1 Reactor in Mode 1 or Mode 2 133 15.4.4.3.1.2 Reactor in MODES 3 to 6 134 15.4.4.3.1.3 Indication and Alarms 134 15.4.4.3.2 Operator Actions Assumed 134 15.4.4.3.3 Chronological Description of Event 135 15.4.4.3.4 Impact on Fission Product Barriers 135 15.4.4.4 Reactor Core and Plant System Evaluation 135 15.4.4.4.1 Methodology 135 15.4.4.4.2 Acceptance Criteria 135 15.4.4.4.3 Dilution During Refueling (MODE 6) 136 15.4.4.4.3.1 Input Parameters and Initial Conditions 136 15.4.4.4.3.2 Results 137 15.4.4.4.4 Dilution During Cold Shutdown (MODE 5) 137 15.4.4.4.5 Dilution at Startup (MODE 2) 137 15.4.4.4.5.1 Input Parameters and Initial Conditions 137 15.4.4.4.5.2 Results 138 15.4.4.4.6 Dilution at Power (MODE 1) 138 15.4.4.4.6.1 Input Parameters and Initial Conditions 138 15.4.4.4.6.2 Results 139 15.4.4.4.7 Dilution from a Single Failure While in Residual Heat Removal Mode -

Inadvertent Draining of the Spray Additive Tank.

15.4.4.4.8 Dilution from a Single Failure While in Residual Heat Removal Mode (MODE 5) -Boron Dilution from the Reactor Coolant Drain Tank.

139 139 15.4.4.4.8.1 Input Parameters and Initial Conditions 139 15.4.4.4.8.2 Results 140 15.4.4.4.9 Dilution from a Single Failure While in Residual Heat Removal Mode (MODE 5) -Boron Dilution Due to Resin Changing in the Purification System.

140

Page 11 of 24 Revision 28 5/2019 15.4.4.4.9.1 Input Parameters and Initial Conditions 140 15.4.4.4.9.2 Results 141 15.4.4.4.10 Dilution from a Single Failure While in Residual Heat Removal Mode (MODE 6) -Boron Dilution from Reactor Coolant Drain Tank After Refueling.

141 15.4.4.4.10.1 Input Parameters and Initial Conditions 141 15.4.4.4.10.2 Results 141 15.4.4.5 Radiological Evaluation 142 15.4.4.6 Conclusions 142 15.4.5 RUPTURE OF A CONTROL ROD DRIVE MECHANISM HOUSING

- ROD CLUSTER CONTROL ASSEMBLY EJECTION 142 15.4.5.1 Description of Event 142 15.4.5.1.1 Nuclear Design 143 15.4.5.1.2 Effects on Adjacent Housings 143 15.4.5.2 Frequency of Event 143 15.4.5.3 Event Analysis 143 15.4.5.3.1 Protective Features 143 15.4.5.3.2 Single Failures Assumed 144 15.4.5.3.3 Operator Actions Assumed 144 15.4.5.3.4 Chronological Description of Event 144 15.4.5.3.5 Impact on Fission Product Barriers 144 15.4.5.4 Reactor Core and Plant System Evaluation 145 15.4.5.4.1 Input Parameters and Initial Conditions 145 15.4.5.4.2 Methodology 145 15.4.5.4.2.1 Average Core Analysis 146 15.4.5.4.2.2 Ejected Rod Worths and Hot Channel Factors 146 15.4.5.4.2.3 Hot Spot Analysis 146 15.4.5.4.2.4 Reactivity Feedback Weighting Factors 147 15.4.5.4.2.5 System Overpressure Analysis 147 15.4.5.4.3 Acceptance Criteria 148 15.4.5.4.4 Results 148 15.4.5.4.4.1 Beginning of Life, Full Power - Case (1) 149 15.4.5.4.4.2 Beginning of Life, Zero Power - Case (2) 149 15.4.5.4.4.3 End of Life, Full Power - Case (3) 149 15.4.5.4.4.4 End of Life, Zero Power - Case (4) 149 15.4.5.4.4.5 Pressure Surge 149 15.4.5.4.4.6 Lattice Deformations 150

Page 12 of 24 Revision 28 5/2019 15.4.5.5 Radiological Evaluation 150 15.4.5.6 Conclusions 150 15.4.6 ROD CLUSTER CONTROL ASSEMBLY DROP 150 15.4.6.1 Description of Event 150 15.4.6.2 Frequency of Event 151 15.4.6.3 Event Analysis 151 15.4.6.3.1 Protective Features 151 15.4.6.3.2 Single Failures Assumed 152 15.4.6.3.3 Operator Actions Assumed 152 15.4.6.3.4 Chronological Description of Event 152 15.4.6.3.5 Impact on Fission Product Barriers 152 15.4.6.4 Reactor Core and Plant System Evaluation 152 15.4.6.4.1 Input Parameters and Initial Conditions 152 15.4.6.4.2 Methodology 153 15.4.6.4.2.1 One or More Dropped Rod Cluster Control Assemblies From the Same Group 153 15.4.6.4.2.2 Dropped Rod Cluster Control Assembly Bank 153 15.4.6.4.2.3 Statically Misaligned Rod Cluster Control Assembly 153 15.4.6.4.3 Acceptance Criteria 153 15.4.6.4.4 Results 154 15.4.6.4.4.1 One or More Dropped Rod Cluster Control Assemblies 154 15.4.6.4.4.2 Dropped Rod Cluster Control Assembly Bank 154 15.4.6.4.4.3 Statically Misaligned Rod Cluster Control Assembly 154 15.4.6.5 Radiological Evaluation 155 15.4.6.6 Conclusions 155

15.4 REFERENCES

FOR SECTION 15.4 156 Table 15.4-1 TIME SEQUENCE OF EVENTS FOR UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL FROM A SUB-CRITICAL CONDITION 158 Table 15.4-2 TIME SEQUENCE OF EVENTS FOR UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL AT POWER 159 Table 15.4-3 PARAMETERS USED IN THE ANALYSIS OF THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT 160 Table 15.4-4 TIME SEQUENCE OF EVENTS FOR ROD CLUSTER CONTROL ASSEMBLY EJECTION 161 Table 15.4-5 REA CONTAINMENT ASSUMPTIONS 162 Table 15.4-6 RESULTS FOR REA DOSE, REM TEDE 164 15.5 INCREASE IN REACTOR COOLANT INVENTORY 165

15.5 REFERENCES

FOR SECTION 15.5 166 15.6 DECREASE IN REACTOR COOLANT INVENTORY 167

Page 13 of 24 Revision 28 5/2019 15.6.1 INADVERTENT OPENING OF A PRESSURIZER SAFETY VALVE OR PRESSURIZER POWER OPERATED RELIEF VALVE (PORV) 167 15.6.1.1 Description of Event 167 15.6.1.2 Frequency of Event 167 15.6.1.3 Event Analysis 167 15.6.1.3.1 Protective Features 167 15.6.1.3.2 Single Failures Assumed 167 15.6.1.3.3 Operator Actions Assumed 167 15.6.1.3.4 Chronological Description of Event 167 15.6.1.3.5 Impact on Fission Product Barriers 167 15.6.1.4 Reactor Core and Plant System Evaluation 168 15.6.1.4.1 Input Parameters and Initial Conditions 168 15.6.1.4.2 Methodology 168 15.6.1.4.3 Acceptance Criteria 168 15.6.1.4.4 Results 169 15.6.1.5 Radiological Consequences 169 15.6.1.6 Conclusions 169 15.6.2 RADIOLOGICAL CONSEQUENCES OF SMALL LINES CARRYING PRIMARY COOLANT OUTSIDE CONTAINMENT 169 15.6.3 Steam Generator Tube Rupture 170 15.6.3.1 Description of Event 170 15.6.3.2 Frequency of Event 170 15.6.3.3 Event Analysis 171 15.6.3.3.1 Protective Features 171 15.6.3.3.2 Single Failures Assumed 172 15.6.3.3.2.1 Single Failure - Margin to Overfill 172 15.6.3.3.2.2 Single Failure - Mass Release 173 15.6.3.3.3 Operator Actions Assumed 173 15.6.3.3.3.1 Operator Actions to Terminate Tube Rupture Flow 173 15.6.3.3.3.2 Operator Actions Due to Single Failures 175 15.6.3.3.3.3 Operator Actions for Cooldown to MODE 5 (Cold Shutdown) 175 15.6.3.3.4 Chronological Description of Event 176 15.6.3.3.5 Impact on Fission Product Barriers 176 15.6.3.4 Reactor Core and Plant System Evaluation 177 15.6.3.4.1 Input Parameters and Initial Conditions 177 15.6.3.4.2 Methodology 178 15.6.3.4.3 Acceptance Criteria 179

Page 14 of 24 Revision 28 5/2019 15.6.3.4.4 Results 179 15.6.3.4.4.1 SGTR Margin to Overfill Transient Analysis 179 15.6.3.4.4.2 SGTR Mass Release Transient Analysis 181 15.6.3.5 Radiological Consequences 182 15.6.3.6 Conclusions 183 15.6.4 PRIMARY SYSTEM PIPE RUPTURES 183 15.6.4.1 Loss of Reactor Coolant from Small Ruptured Pipes or From Cracks in Large Pipes Which Actuates Emergency Core Cooling System (ECCS) 183 15.6.4.1.1 Description of Event 183 15.6.4.1.2 Frequency of Event 184 15.6.4.1.3 Event Analysis 184 15.6.4.1.3.1 Protective Features 184 15.6.4.1.3.2 Single Failures Assumed 185 15.6.4.1.3.3 Operator Actions Assumed 185 15.6.4.1.3.4 Chronological Description of Event 185 15.6.4.1.3.5 Impact on Fission Product Barriers 186 15.6.4.1.4 Reactor Core and Plant System Evaluation 186 15.6.4.1.4.1 Input Parameters and Initial Conditions 186 15.6.4.1.4.2 Methodology 187 15.6.4.1.4.3 Acceptance Criteria 188 15.6.4.1.4.4 Results 188 15.6.4.1.4.5 Effect of Emergency Core Cooling System (ECCS) Evaluation Model Modi-fications 189

Page 15 of 24 Revision 28 5/2019 15.6.4.1.5 Radiological Evaluation 189 15.6.4.1.6 Conclusions 189 15.6.4.2 Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident) 189 15.6.4.2.1 Description of Event 189 15.6.4.2.2 Frequency of Event 191 15.6.4.2.3 Event Analysis 191 15.6.4.2.3.1 Protective Features 191 15.6.4.2.3.2 Single Failures Assumed 192 15.6.4.2.3.3 Operator Actions Assumed 192 15.6.4.2.3.4 Chronological Description of Event 192 15.6.4.2.3.5 Impact on Fission Product Barriers 194 15.6.4.2.4 Reactor Core and Plant System Evaluation 194 15.6.4.2.4.1 Input Parameters and Initial Conditions 194 15.6.4.2.4.2 Methodology 197 15.6.4.2.4.3 Acceptance Criteria 202 15.6.4.2.4.4 Results 202 15.6.4.2.5 Radiological Evaluation 203 15.6.4.2.6 Conclusions 204

15.6 REFERENCES

FOR SECTION 15.6 205 Table 15.6-1 COMPARISON OF NOMINAL AND PLANT PARAMETERS USED IN STEAM GENERATOR TUBE RUPTURE (SGTR) ANALYSIS 209 Table 15.6-2 OPERATOR ACTION TIMES 210 Table 15.6-3 SEQUENCE OF EVENTS - MARGIN TO OVERFILL ANALYSIS 211 Table 15.6-4 OPERATOR ACTION TIMES FOR DESIGN BASIS STEAM GEN-ERATOR TUBE RUPTURE ANALYSIS 212 Table 15.6-5 SEQUENCE OF EVENTS - OFFSITE RADIATION DOSE ANALY-SIS 213 Table 15.6-6 SGTR DOSE ANALYSIS ASSUMPTIONS 214 Table 15.6-7 STEAM RELEASES AND RUPTURE FLOW 216 Table 15.6-8 RESULTS FOR SGTR, REM TEDE 217 Table 15.6-9 TIME SEQUENCE OF EVENTS - ACCIDENTAL DEPRESSURIZA-TION OF THE RCS 218 Table 15.6-10 TOTAL SMALL BREAK LOSS-OF-COOLANT ACCIDENT SAFETY INJECTION AND SPILL FLOW 219 Table 15.6-11 SMALL BREAK LOSS-OF-COOLANT ACCIDENT KEY ASSUMP-TIONS 220 Table 15.6-12 SMALL BREAK LOSS-OF-COOLANT ACCIDENT MAIN STEAM SAFETY VALVE (MSSV) ASSUMPTIONS 222

Page 16 of 24 Revision 28 5/2019 Table 15.6-13 SMALL BREAK LOSS-OF-COOLANT ACCIDENT TIME SEQUENCE OF EVENTS 223 Table 15.6-14 SMALL BREAK LOSS-OF-COOLANT ACCIDENT FUEL 224 CLADDING RESULTS Table 15.6-15 LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS TIME SEQUENCE OF EVENTS FOR DECLG BREAK 225 Table 15.6-16 Key LBLOCA Parameters and Initial Transient Assumptions for R.

E. Ginna Analysis 226 Table 15.6-17 LARGE BREAK LOCA ANALYSIS SAFETY INJECTION FLOW VERSUS PRESSURE 229 Table 15.6-18a PARAMETERS FOR CONTAINMENT PRESSURE - DRY CONTAINMENT DATA 231 Table 15.6-18b STRUCTURAL HEAT SINK DATA 232 Table 15.6-19 PLANT OPERATING RANGE ALLOWED BY THE BEST-ESTIMATE LARGE BREAK LOCA ANALYSIS (R. E. GINNA) 234 Table 15.6-20 LIMITING LARGE BREAK PCT AND OXIDATION RESULTS FOR R. E. GINNA 236 Table 15.6-21 ASSUMPTIONS FOR ANALYSIS OF RADIOLOGICAL CONSE-QUENCES OF THE LOSS-OF-COOLANT ACCIDENT 237 Table 15.6-21A LBLOCA DOSE

SUMMARY

, REM TEDE 239 Table 15.6-22 Total Core Activity (Curies) at End of 525-day Fuel Cycle - including Decay 240 Table 15.6-23 Core Inventory Fraction Released into Containment 243 Table 15.6-24 TABLE DELETED 245 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 246 15.7.1 RADIOACTIVE GAS WASTE SYSTEM FAILURE 246 15.7.1.1 Gas Decay Tank Rupture 246 15.7.1.1.1 Description of Event 246 15.7.1.1.2 Frequency of Event 246 15.7.1.1.3 Event Analysis 246 15.7.1.1.3.1 Single Failures Assumed 247 15.7.1.1.3.2 Operator Actions Assumed 247 15.7.1.1.3.3 Chronological Description of Event 247 15.7.1.1.3.4 Impact on Fission Product Barriers 247 15.7.1.1.4 Reactor Core and Plant System Evaluation 247 15.7.1.1.4.1 Input Parameters and Initial Conditions 247 15.7.1.1.4.2 Methodology 248 15.7.1.1.4.3 Acceptance Criteria 248 15.7.1.1.4.4 Results 248

Page 17 of 24 Revision 28 5/2019 15.7.1.1.5 Radiological Evaluation 248 15.7.1.1.6 Conclusions 248 15.7.1.2 Volume Control Tank Rupture 249 15.7.1.2.1 Description of Event 249 15.7.1.2.2 Frequency of Event 249 15.7.1.2.3 Event Analysis 249 15.7.1.2.3.1 Single Failures Assumed 249 15.7.1.2.3.2 Operator Actions Assumed 249 15.7.1.2.3.3 Chronological Description of Event 249 15.7.1.2.3.4 Impact on Fission Product Barriers 250 15.7.1.2.4 Reactor Core and Plant System Evaluation 250 15.7.1.2.4.1 Input Parameters and Initial Conditions 250 15.7.1.2.4.2 Methodology 250 15.7.1.2.4.3 Acceptance Criteria 251 15.7.1.2.4.4 Results 251 15.7.1.2.5 Radiological Evaluation 251 15.7.1.2.6 Conclusions 251 15.7.2 RADIOACTIVE LIQUID WASTE SYSTEM FAILURE 251 15.7.2.1 Description of Event 251 15.7.2.2 Frequency of Event 252 15.7.2.3 Event Analysis 252 15.7.2.3.1 Single Failures Assumed 253 15.7.2.3.2 Operator Actions Assumed 253 15.7.2.3.3 Chronological Description of Event 253 15.7.2.3.4 Impact on Fission Product Barriers 253 15.7.2.4 Reactor Core and Plant System Evaluation 253 15.7.2.4.1 Input Parameters and Initial Conditions 253 15.7.2.4.2 Methodology 254 15.7.2.4.3 Acceptance Criteria 254 15.7.2.4.4 Results 254 15.7.2.4.4.1 Accidental Release of Liquid Waste Assessment 254 15.7.2.4.4.2 Spent Resin Storage Tank Assessment 255 15.7.2.4.5 Effects of 18-month Fuel Cycle 256 15.7.2.5 Radiological Evaluation 256 15.7.2.6 Conclusions 256 15.7.3 FUEL HANDLING ACCIDENTS 256 15.7.3.1 Description of Event 256

Page 18 of 24 Revision 28 5/2019 15.7.3.1.1 MODE 6 (Refueling) Preparations 256 15.7.3.1.2 Fuel Handling Equipment Safety Features 257 15.7.3.1.3 Fuel Handling Operations Precautions 258 15.7.3.1.4 Consequence of Dropped Fuel Assembly 258 15.7.3.2 Frequency of Event 259 15.7.3.3 Event Analysis 259 15.7.3.3.1 Protective Features 260 15.7.3.3.2 Single Failures Assumed 260 15.7.3.3.3 Operator Actions Assumed 260 15.7.3.3.4 Chronological Description of Event 260 15.7.3.3.5 Impact on Fission Product Barriers 260 15.7.3.4 Reactor Core and Plant System Evaluation 261 15.7.3.4.1 Input Parameters and Initial Conditions 261 15.7.3.4.2 Methodology 261 15.7.3.4.3 Acceptance Criteria 261 15.7.3.4.4 Results 261 15.7.3.5 Radiological Evaluation 261 15.7.3.6 Conclusions 261

15.7 REFERENCES

FOR SECTION 15.7 262 Table 15.7-1 FISSION PRODUCT INVENTORY AND ACTIVITY RELEASED FROM POOL 265 Table 15.7-2 FHA DOSE ANALYSIS ASSUMPTIONS 266 Table 15.7-3 FHA DOSE. REM TEDE 267 Table 15.7-4 Table DELETED 268 Table 15.7-5 Table DELETED 269 Table 15.7-6 Table DELETED 270 15.8 ANTICIPATED TRANSIENTS WITHOUT SCRAM 271 15.8.1 ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS) 271 15.8.2 Frequency of event 271 15.8.3 Event Analysis 271 15.8.3.1 Single Failures Assumed 271 15.8.3.2 Operator Actions Assumed 271 15.8.3.3 Chronological Description of Event 272 15.8.3.4 Impact on Fission Product Barriers 272 15.8.4 Reactor Core and Plant System Evaluation 273 15.8.4.1 Input Parameters and Initial Conditions 273 15.8.4.2 Methodology 274 15.8.4.3 Acceptance Criteria 274

Page 19 of 24 Revision 28 5/2019 15.8.4.4 Results 274 15.8.5 Radiological Evaluation 275 15.8.6 Conclusions 275

15.8 REFERENCES

FOR SECTION 15.8 276 FIGURES Figure 15.0-1 Core Limits and Overpower-Overtemperature Delta T Setpoints (Tref =

576.0F)

Figure 15.0-2 Reactivity Coefficients Used in Non-LOCA Safety Analysis Figure 15.0-3 Reactivity Insertion Scram Curves Figure 15.1-1 Feedwater Flow Increase at Full Power, Nuclear Power and Loop Average Temperature Versus Time Figure 15.1-2 Feedwater Flow Increase at Full Power, Pressurizer Pressure and Steam Gen-erator Pressure Versus Time Figure 15.1-3 Feedwater Flow Increase at Full Power, Steam Generator Mass Versus Time Figure 15.1-4 Steam Line Rupture, Multiplication Factor Versus Core Average Temperature (Calculated at 1050 psia)

Figure 15.1-5 Steam Line Rupture, Integrated Doppler Defect Versus Fraction of Power Figure 15.1-6 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-7 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Pressur-izer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-8 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Loop TAVG and Cold Leg Loop Temperature Versus Time Figure 15.1-9 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-10 Steam Line Rupture, 1.4ft2 Break with Power, Two Loops in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-11 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-12 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Pressurizer Water Volume and Pressurizer Pressure Versus Time Figure 15.1-13 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Loop TAVG and Cold Leg Loop Temperatures Versus Time Figure 15.1-14 Steam Line Rupture, 1.4ft2 Break Without Power, Two Loops in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-15 Steam Line Rupture, 1.4ft2 Break without Power, Two Loops in Service, Core Averaged Boron and Reactivity Versus Time Figure 15.1-16 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Core Heat Flux and Nuclear Power Versus Time Figure 15.1-17 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Pressur-izer Water Volume and Pressurizer Pressure Versus Time

Page 20 of 24 Revision 28 5/2019 Figure 15.1-18 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Loop TAVG and Cold Leg Loop Temperatures Versus Time Figure 15.1-19 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Faulted Loop Steam Flow and Total Feedwater Flow Versus Time Figure 15.1-20 Steam Line Rupture, 1.4ft2 Break with Power, One Loop in Service, Core Figure 15.1-21 Averaged Boron and Reactivity Versus Time Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Nuclear Power and Core Heat Flux Versus Time Figure 15.1-22 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Loop Average Temperature and Pressurizer Pressure Versus Time Figure 15.1-23 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, DNBR Versus Time Figure 15.1-24 Combined Atmospheric Relief Valve and Main Feedwater Regulating Valve Failure, Steam Generator Level and Steam Generator Mass Versus Time Figure 15.2-1 Loss of Load, with Automatic Pressure Control, Nuclear Power and DNBR Versus Time Figure 15.2-2 Loss of Load, with Automatic Pressure Control, RCS Average Temperature and Pressurizer Water Volume Versus Time Figure 15.2-3 Loss of Load, with Automatic Pressure Control, Steam Generator Pressure and Pressurizer Pressure Versus Time Figure 15.2-4 Loss of Load, Without Pressure Control, Nuclear Power Versus Time Figure 15.2-5 Loss of Load, Without Pressure Control, RCS Average Temperature and Pres-surizer Water Volume Versus Time Figure 15.2-6 Loss of Load, Without Pressure Control, Steam Generator Pressure and Reac-tor Coolant System Pressures Versus Time Figure 15.2-7 Loss of Load, Peak MSS Pressure Case, Nuclear Power Versus Time Figure 15.2-8 Loss of Load, Peak MSS Pressure Case, RCS Average Temperature and Pres-surizer Water Volume Versus Time Figure 15.2-9 Loss of Load, Peak MSS Pressure Case, Steam Generator Pressure and Pres-surizer Pressure Versus Time Figure 15.2-10 Figure Deleted Figure 15.2-11 Figure Deleted Figure 15.2-12 Figure Deleted Figure 15.2-13 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Nuclear Power and Pressurizer Pressure Versus Time Figure 15.2-14 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Pressur-izer Water Volume and Pressurizer Steam Relief Rate Versus Time Figure 15.2-15 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Reactor Coolant Flow and Core Inlet/Outlet Temperatures Versus Time Figure 15.2-16 Loss of Offsite Alternating Current Power to the Station Auxiliaries, Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-17 Loss of Normal Feedwater With Power, Nuclear Power and Pressurizer Pres-sure Versus Time

Page 21 of 24 Revision 28 5/2019 Figure 15.2-18 Loss of Normal Feedwater With Power, Pressurizer Water Volume and Pres-surizer Steam Relief Rate Versus Time Figure 15.2-19 Loss of Normal Feedwater With Power, Reactor Coolant Flow and Core Inlet/

Outlet Temperatures Versus Time Figure 15.2-20 Loss of Normal Feedwater With Power, Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-21 Feedline Break With Offsite Power; Nuclear Power and Pressurizer Pressure Versus Time Figure 15.2-22 Feedline Break With Offsite Power; Pressurizer Water Volume and Pressur-izer Steam Relief Rate Versus Time Figure 15.2-23 Feedline Break With Offsite Power; Cold Leg, Hot Leg and Saturation Tem-peratures Versus Time Figure 15.2-24 Feedline Break With Offsite Power; Steam Generator Mass and Steam Gener-ator Pressure Versus Time Figure 15.2-25 Feedline Break With Offsite Power; Feedwater Mass Flow Rates Versus Time Figure 15.2-26 Feedline Break Without Offsite Power; Nuclear Power and Pressurizer Pres-sure Versus Time Figure 15.2-27 Feedline Break Without Offsite Power; Pressurizer Water Volume and Pres-surizer Steam Relief Rate Versus Time Figure 15.2-28 Feedline Break Without Offsite Power; Cold Leg, Hot Leg and Saturation Temperatures Versus Time Figure 15.2-29 Feedline Break Without Offsite Power; Steam Generator Mass and Steam Generator Pressure Versus Time Figure 15.2-30 Feedline Break Without Offsite Power; Feedwater Mass Flow Rates Versus Time Figure 15.3-1 Full Loss of Flow (Undervoltage), Nuclear Power and RCS Flow Versus Time Figure 15.3-1a Full Loss of Flow (Underfrequency), Nuclear Power and RCS Flow Versus Time Figure 15.3-2 Full Loss of Flow (Undervoltage), Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-2a Full Loss of Flow (Underfrequency), Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-3 Full Loss of Flow (Undervoltage), RCS Pressures and DNBR Versus Time Figure 15.3-3a Full Loss of Flow (Underfrequency), DNBR and Reactor Coolant System Pressures Versus Time Figure 15.3-4 Partial Loss of Flow, Nuclear Power and RCS Flow Versus Time Figure 15.3-5 Partial Loss of Flow, RCS Pressures and RCS Loop Flows Versus Time Figure 15.3-6 Partial Loss of Flow, Core Average and Hot Channel Heat Flux Versus Time Figure 15.3-7 Partial Loss of Flow, DNBR Versus Time Figure 15.3-8 Locked Rotor, RCS Pressures and RCS Loop Flows Versus Time Figure 15.3-9 Locked Rotor, Nuclear Power and RCS Flow Versus Time Figure 15.3-10 Locked Rotor, Core Average Heat Flux and Cladding Inside Temperature Ver-sus Time

Page 22 of 24 Revision 28 5/2019 Figure 15.4-1 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From Subcritical Conditions, Heat Flux and Nuclear Power Versus Time (422V+Fuel)

Figure 15.4-2 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From Subcritical Conditions, Clad Inside and Fuel Average Temperature Versus Time(422V+Fuel)

Figure 15.4-3 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Nuclear Power and Heat Flux Versus Time Figure 15.4-4 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Pressurizer Pressure and Pressurizer Water Volume Versus Time Figure 15.4-5 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum Feedback, 100 pcm/sec, Tavg and DNBR Versus Time Figure 15.4-6 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, Nuclear Power and Heat Flux Versus Time Figure 15.4-7 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, Pressurizer Water Volume and Pressurizer Pressure Versus Time Figure 15.4-8 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Maximum Feedback, 5 pcm/sec, TAVG and DNBR Versus Time Figure 15.4-9 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Insertion Rate Figure 15.4-10 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from 60%

Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Inser-tion Rate Figure 15.4-11 Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from 10%

Power, Minimum and Maximum Feedback, DNBR Versus Reactivity Inser-tion Rate Figure 15.4-12 Startup of an Inactive Coolant Loop, Nuclear Power Versus Time Figure 15.4-13 Startup of an Inactive Coolant Loop, TAVG Versus Time Figure 15.4-14 Startup of an Inactive Coolant Loop, Core Inlet Temperature Versus Time Figure 15.4-15 Startup of an Inactive Coolant Loop, Pressurizer Pressure Versus Time Figure 15.4-16 Rod Cluster Control Assembly Ejection Beginning-of-Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-16a Rod Cluster Control Assembly Ejection, Beginning of Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-16b Rod Cluster Control Assembly Ejection, Beginning of Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17 Rod Cluster Control Assembly Ejection Beginning-of-Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17a Rod Cluster Control Assembly Ejection, End of Life, Full Power, Fuel and Clad Temperature and Nuclear Power Versus Time Figure 15.4-17b Rod Cluster Control Assembly Ejection, End of Life, Zero Power, Fuel and Clad Temperature and Nuclear Power Versus Time

Page 23 of 24 Revision 28 5/2019 Figure 15.4-18 Rod Cluster Control Assembly Drop Heat Flux and Nuclear Power Versus Time Figure 15.4-19 Rod Cluster Control Assembly Drop Pressurizer Pressure and Core Average Temperature Versus Time Figure 15.4-20 Uncontrolled Rod Cluster Control Assembly Bank Withdrawl from 8% Power (RCS Pressure Case), Minimum Feedback, 55 pcm/sec, Nuclear Power and Heat Flux Versus Time Figure 15.4-21 Uncontrolled Rod Cluster Control Assembly Bank Withdrawl from 8% Power (RCS Pressure Case), Minimum Feedback, 55 pcm/sec, Pressurizer Pressure and Tavg Versus Time Figure 15.6-1 Steam Generator Tube Rupture (Overfill), Maximum Safety Injection Flow Versus Pressure Figure 15.6-1a RCS Depressurization, Nuclear Power Versus Time Figure 15.6-1b RCS Pressurization, Pressurizer Pressure Versus Time Figure 15.6-1c RCS Depressurization, Indicated Loop Average Temperature Versus Time Figure 15.6-1d RCS Depressurization, DNBR Versus Time Figure 15.6-2 SGTR (Overfill), Pressurizer Level and Pressurizer Pressure Versus Time Figure 15.6-3 SGTR (Overfill), Secondary Pressure and Steam Generator Liquid Mass Ver-sus Time Figure 15.6-4 SGTR (Overfill), Hot and Cold Leg Temperatures for Intact and Ruptured Steam Generators Versus Time Figure 15.6-5 SGTR (Overfill), Total Primary to Secondary Leakage and Total Integrated Primary to Secondary Leakage Versus Time Figure 15.6-6 SGTR (Overfill), Steam Generator Relief Flow and Integrated Steam Genera-tor Relief Flow Versus Time Figure 15.6-7 SGTR (Overfill), Steam Generator Water Volume Versus Time Figure 15.6-8 SGTR (Dose), Pressurizer Level and Pressurizer Pressure Versus Time Figure 15.6-9 SGTR (Dose), Secondary Pressure and Steam Generator Liquid Mass Versus Time Figure 15.6-10 SGTR (Dose), Hot and Cold Leg Temperatures for Intact and Ruptured Steam Generators Versus Time Figure 15.6-11 SGTR (Dose), Total Primary to Secondary Leakage and Total Integrated Primary to Secondary Leakage Versus Time Figure 15.6-12 SGTR (Dose), Steam Generator Relief Flow and Integrated Steam Generator Relief Flow Versus Time Figure 15.6-13 SGTR (Dose), Steam Generator Water Volume Versus Time Figure 15.6-14 SGTR (Dose), Tube Rupture Flow Flashing Fraction and Integrated Flashed Break Versus Time Figure 15.6-15 Small Break LOCA Inch Break, Pressurizer Pressure Versus Time Figure 15.6-16 Small Break LOCA Inch Break, Core Mixture Level Versus Time Figure 15.6-17 Small Break LOCA Inch High Break, Peak Cladding Temperature at PCT Elevation Versus Time Figure 15.6-18 Small Break LOCA Inch High Break, Core Exit Vapor Flow Versus Time Figure 15.6-19 Small Break LOCA Inch Break, Hot Rod Heat Transfer Coefficient at PCT Elevation Versus Time

Page 24 of 24 Revision 28 5/2019 Figure 15.6-20 Small Break LOCA Inch Break, Fluid Temperature at PCT Elevation Versus Time Figure 15.6-21 Small Break LOCA - Axial Power Distribution, Heat Rate Versus Core Elevation Figure 15.6-22 Small Break LOCA - 1.5-Inch Break, Pressurizer Pressure Versus Time Figure 15.6-23 Small Break LOCA Inch High Break, Pressurizer Pressure Versus Time Figure 15.6-24 Small Break LOCA - 1.5-Inch Break, Core Mixture Level Versus Time Figure 15.6-25 Small Break LOCA Inch Break, Core Mixture Level Versus Time Figure 15.6-26 Small Break LOCA - 1.5-Inch Break, Peal Cladding Temperature at PCT Elevation Versus Time Figure 15.6-27 Small Break LOCA Inch Break, Peak Cladding Temperature at PCT Elevation Versus Time Figure 15.6-28 Figure Deleted Figure 15.6-29 Figure Deleted Figure 15.6-30 Figure Deleted Figure 15.6-31 R.E. Ginna Vessel Model Noding Diagram1 Figure 15.6-32 R.E. Ginna Loop Model Noding Diagram Figure 15.6-33 R.E. Ginna Initial Transient Axial Power Distributions Figure 15.6-34 Containment Pressure Used for the R.E. Ginna Best-Estimate Large Break LOCA Initial Transient Figure 15.6-35 Peak Clad Temperature of the 5 rods for the Initial Transient Figure 15.6-36 Split Break Flow for the Initial Transient Figure 15.6-37 Total Flow at the Bottom of the Core for the Initial Transient Figure 15.6-38 Accumulator Injection Flow for the Initial Transient Figure 15.6-39 High Head Safety Injection Flow for the Initial Transient Figure 15.6-40 Low Head Safety Injection Flow for the Initial Transient Figure 15.6-41 Average Collapsed Liquid Level in the Downcomer for the Initial Transient Figure 15.6-42 Lower Plenum Collapsed Liquid Level for the Initial Transient Figure 15.6-43 Core Collapsed Liquid Levels for the Initial Transient Figure 15.6-44 Vessel Liquid Mass for the Initial Transient Figure 15.6-45 Pressurizer Pressure for the Initial Transient Figure 15.6-46 Hot Rod Peak Clad Temperature and Elevation for the Initial Transient Figure 15.6-47 R.E. Ginna PBOT/PMID Analysis and Operating Limits Figure 15.6-48 Lower Bound Containment Pressure for R.E. Ginna Analysis