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NUREG-2187, Vol. 2, Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models-Byron Unit 1. Appendices D to G.
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NUREG-2187 Volume 2 Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk ModelsByron Unit 1 Appendices D to G Office of Nuclear Regulatory Research

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NUREG-2187 Volume 2 Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk ModelsByron Unit 1 Appendices D to G Manuscript Completed: May 2015 Date Published: January 2016 Prepared by:

J. Corson,1 D. Helton,1 M. Tobin,1 A. Bone1 M. Khatib-Rahbar,2 A. Krall2 L. Kozak3 R. Buell4 1Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 2Energy Research Inc.

P.O. Box 2034 Rockville, MD 20847-2034 3Region III U.S. Nuclear Regulatory Commission 2443 Warrenville Road, Suite 210 Lisle, IL 60532-4352 4Idaho National Laboratory P.O. Box 1625 Idaho Falls, ID 83415

ABSTRACT This report extends the work documented in NUREG-1953, Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models-Surry and Peach Bottom to the Byron Station, Unit 1. Its purpose is to produce an additional set of best-estimate thermal-hydraulic calculations that can be used to confirm or enhance specific success criteria (SC) for system performance and operator timing found in the agencys probabilistic risk assessment (PRA) tools. Along with enhancing the technical basis for the Agencys independent standardized plant analysis risk (SPAR) models, these calculations are expected to be a useful reference to model end-users for specific regulatory applications (e.g.,

the Significance Determination Process). The U.S. Nuclear Regulatory Commission selected Unit 1 of the Byron Station for this study because it is generally representative of a group of four-loop Westinghouse plants with large, dry containment designs.

This report first describes major assumptions used in this study, including the basis for using a core damage (CD) surrogate of 2,200 degrees Fahrenheit (1,204 degrees Celsius) peak cladding temperature (PCT). The justification for this PCT is documented in NUREG/CR-7177, Compendium Of Analyses To Investigate Select Level 1 Probabilistic Risk Assessment End-State Definition And Success Criteria Modeling Issues. The major plant characteristics for Byron Unit 1 are then described, in addition to the MELCOR model used to represent the plant.

Finally, the report presents the results of MELCOR calculations for selected initiators and compares these results to SPAR SC, the licensees PRA sequence timing and SC, or other generic studies.

The study results provide additional timing information for several PRA sequences, confirm many of the existing SPAR model modeling assumptions, and provide a technical basis for a few specific SPAR modeling changes. Potential SPAR model changes supported by this study include:

Small-Break Loss-of-Coolant Accident (SLOCA) Sequence Timing for Alignment of Sump RecirculationFor sequences where operator cooldown is credited as an alternative to high-pressure recirculation (HPR), the SPAR success criteria related to containment cooling could be enhanced by requiring one containment fan cooler to prevent containment spray actuation. Avoiding spray actuation extends the time available prior to refueling water storage tank depletion and allows the operators to successfully depressurize the plant using the post-LOCA procedures for cases when HPR is not available.

SLOCA Success Criteria for Steam Generator (SG) Depressurization and Condensate FeedAction to depressurize the SGs early and align condensate feed is a candidate for inclusion in the SPAR model. This would provide an additional success path for a loss of auxiliary feedwater event. If this is done, hotwell refill or alignment of alternate feedwater later in the scenario would also need to be modeled. Early depressurization to achieve condensate feed was not found to require primary-side depressurization actions (e.g., opening a power-operated relief valve (PORV)).

SLOCA Success Criteria for Primary Side Bleed and Feed (B&F)These calculations have demonstrated a potential conservatism that can be removed from the applicable SPAR models. It is proposed that the SC for SLOCA B&F be changed from (one safety iii

injection (SI) or centrifugal charging pump (CCP) and two PORVs) to (one SI pump and two PORVs) or (one CCP and one PORV). In other words, for SLOCAs the requirement for availability of a second PORV can be removed when a CCP is available.

Loss of DC Bus-111 - Unavailable Diesel-Driven Auxiliary Feedwater, and Subsequent Primary Side B&FThese calculations are generally representative of non-loss-of-coolant accident (non-LOCA) B&F situations and have demonstrated a potential improvement that can be implemented in the Byron SPAR model. It is proposed that the SC for non-LOCA B&F be changed from (one SI or CCP and two PORVs) to (one CCP and one PORV). In other words, the same one CCP and one PORV enhancement as above is credited, but credit is eliminated for cases with no CCP available. This initiator was chosen because it was qualitatively felt to be more restrictive than those scenarios categorized as general transients in the PRA, and thus the conclusions are believed to be applicable to those initiators as well. Note that the applicability of the loss of DC bus SC may vary, (e.g., due to the unique reactor coolant pump trip situation that this initiator creates) and should be evaluated on a case-by-case basis before implementation for other plant models.

SGTR - Spontaneous Steam Generator Tube Rupture with No Operator ActionFor sequences with successful high-pressure injection (HPI) and auxiliary feedwater, but with steam generator isolation having failed, an additional success path or additional recovery credit may be justifiable pending additional consideration of closely-related accident sequence and human reliability modeling assumptions.

Medium-Break Loss-of-Coolant Accident (MLOCA) - Injection SC For breaks in the lower half of the MLOCA range, it was found that an early operator-induced depressurization based on the Functional Restoration Procedure (FRP) for inadequate core cooling would be needed to avoid core damage if HPI fails. The time available to implement these actions following the FRP entry criterion being met could be short. The accident sequence modeling and human reliability analysis associated with secondary-side cooldown for these situations (MLOCA with HPI failed) should be reviewed.

iv

FOREWORD The U.S. Nuclear Regulatory Commissions (NRCs) standardized plant analysis risk (SPAR) models are used to support a number of risk-informed initiatives. The fidelity and realism of these models is ensured through a number of processes, including cross-comparison with industry models, review and use by a wide range of technical experts, and confirmatory analysis. The following reportprepared by staff in the Office of Nuclear Regulatory Research in consultation with staff from the Office of Nuclear Reactor Regulation, experts from Energy Research Incorporated and Idaho National Laboratory, and the agencys senior reactor analystsrepresents a major confirmatory analysis activity.

Probabilistic risk assessment (PRA) models for nuclear power plants rely on underlying modeling assumptions known as success criteria (SC) and sequence timing assumptions.

These criteria and assumptions determine what combination of system and component availabilities will lead to postulated core damage (CD), as well as the timeframes during which components must operate or operators must take particular actions. This report investigates certain thermal-hydraulic aspects of a particular SPAR model (which is generally representative of other models within the same class of plant design), with the goal of further strengthening the technical basis for decisionmaking that relies on the SPAR models. This report augments the existing collection of contemporary Level 1 PRA SC analyses, and as such, supports (1) maintaining and enhancing the SPAR models that the NRC develops, (2) supporting the NRCs risk analysts when addressing specific issues in the accident sequence precursor program and the significance determination process, and (3) informing other ongoing and planned initiatives. This analysis employs the MELCOR computer code and uses a plant model developed for this project.

The analyses summarized in this report provide the basis for confirming or changing SC in the SPAR model for the Byron Station Unit 1. Further evaluation of these results will be performed to extend the results to similar plants. In addition, future work is planned to perform similar analysis for other design classes, and past work has already considered other design classes (see NUREG-1953, Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis Risk Models - Surry and Peach Bottom). In addition, work has been recently completed to scope other aspects of this topical area, including the degree of variation typical in common PRA sequences and the quantification of conservatisms associated with CD surrogates (see NUREG/CR-7177, Compendium of Analyses to Investigate Select Level 1 Probabilistic Risk Assessment End-State Definition and Success Criteria Modeling Issues). Where applicable, insights from that work are referenced in this report. The confirmation of SC and other aspects of PRA modeling using the agencys state-of-the-art tools (e.g., the MELCOR computer code) is expected to receive continued focus as the agency continues to develop and improve its risk tools.

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CONTENTS ABSTRACT ...............................................................................................................................iii FOREWORD ............................................................................................................................. v CONTENTS ..............................................................................................................................vii LIST OF FIGURES ....................................................................................................................ix LIST OF TABLES ......................................................................................................................xi ABBREVIATIONS AND ACRONYMS .....................................................................................xiii

1. INTRODUCTION AND BACKGROUND ............................................................................. 1

2. MAJOR ASSUMPTIONS .................................................................................................... 3 2.1 Selection of a Core Damage Surrogate........................................................................ 5

3. RELATIONSHIP TO THE AMERICAN SOCIETY OF MECHANICAL ENGINEERS/AMERICAN NUCLEAR SOCIETY PROBABILISTIC RISK ASSESSMENT STANDARD .............................................................................................. 9

4. MAJOR PLANT AND MELCOR MODEL CHARACTERISTICS........................................13 4.1 Byron Station Unit 1 ....................................................................................................13 4.2 Byron MELCOR Model ...............................................................................................14 4.3 MELCOR Validation ....................................................................................................15

5. MELCOR RESULTS..........................................................................................................17 5.1 Small-Break Loss-of-Coolant Accident-Sequence Timing for Alignment of Sump Recirculation.....................................................................................................18 5.2 Small-Break Loss-of-Coolant Accident-Success Criteria for Steam Generator Depressurization and Condensate Feed .....................................................................29 5.3 Small-Break Loss-of-Coolant Accident-Success Criteria for Primary Side Bleed and Feed ..........................................................................................................35 5.4 Loss of DC Bus 111, Unavailable DD-AFW, and Subsequent Primary Side Bleed and Feed ..........................................................................................................42 5.5 Spontaneous Steam Generator Tube Rupture with No Operator Action......................50 5.6 Medium-Break LOCA Injection Success Criteria .........................................................59 5.7 Medium-Break LOCA Cooldown Timing for Low-Pressure Recirculation ....................67 5.8 Loss of Shutdown Cooling ..........................................................................................76 5.8.1 Changes to the MELCOR Input Deck for Loss of Shutdown Cooling Calculations .........................................................................................................77 5.8.2 Mode 4 Calculations ............................................................................................77 5.8.3 Mode 5 Calculations ............................................................................................82

6. APPLICATION OF MELCOR RESULTS TO THE SPAR MODELS ..................................87

7. CONCLUSIONS ................................................................................................................93

8. REFERENCES ..................................................................................................................95 APPENDIX A DETAILED INFORMATION ON BASE MELCOR MODEL A.1 Byron MELCOR Input Model Description..................................................................... A-1 A.2 Input Deck Revisions and MELCOR Code Versions .................................................... A-6 A.3 Additional Notes on MELCOR ..................................................................................... A-7 A.4 References .................................................................................................................. A-7 vii

APPENDIX B DETAILED SMALL-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS RESULTS B.1 Small-Break Loss-of-Coolant Accident - Sequence Timing for Alignment of Sump Recirculation .................................................................................................... B-1 B.2 Small-Break Loss-of-Coolant Accident - Success Criteria for Steam Generator Depressurization and Condensate Feed.................................................................... B-85 B.3 Small-Break Loss-of-Coolant Accident - Success Criteria for Primary Side Bleed and Feed ....................................................................................................... B-133 APPENDIX C DETAILED LOSS OF DC BUS 111 ANALYSIS RESULTS C.1 Loss of DC Bus 111 and Unavailable DD-AFW, Leading to Primary Side Bleed and Feed ........................................................................................................... C-1 APPENDIX D DETAILED STEAM GENERATOR TUBE RUPTURE ANALYSIS RESULTS D.1 Spontaneous SG Tube Rupture with No Operator Action ............................................ D-1 APPENDIX E DETAILED MEDIUM-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS RESULTS E.1 Medium-Break Loss-of-Coolant Accident Injection Success Criteria ............................ E-1 E.2 Medium-Break Loss-of-Coolant Accident Cooldown Timing for Low-Pressure Recirculation.............................................................................................................. E-91 APPENDIX F DETAILED LOSS OF SHUTDOWN COOLING RESULTS F.1 Mode 4 Calculations .................................................................................................... F-1 F.2 Mode 5 Calculations .................................................................................................. F-47 APPENDIX G EVENT TREE MODELS FOR STUDIED INITIATORS G.1 Byron SPAR Model Event Trees..................................................................................G-1 viii

LIST OF FIGURES Main Report Figure 1 Example of variation in core damage timing from (NRC, 2014b) ............................ 6 Figure 2 Schematic of the Byron MELCOR RCS model ..................................................... 15 Figure 3 Time of RWST depletion as a function of RWST volume ...................................... 27 Figure 4 Peak containment pressure as a function of containment volume and the number of available fan coolers for Case 11 ................................................... 28 Appendix G Figure G-1 Small-break loss-of-coolant accident (SLOCA) event tree .................................. G-2 Figure G-2 Loss of 125V vital DC bus 111 event tree ........................................................... G-3 Figure G-3 Steam generator tube rupture (SGTR) event tree ............................................... G-4 Figure G-4 Medium-break loss-of-coolant accident (MLOCA) event tree .............................. G-5 ix

LIST OF TABLES Main Report Table 1 Summary of Accident Scenarios Examined ............................................................ 2 Table 2 Major Assumptions ................................................................................................. 4 Table 3 Comparison of this Project to the ASME/ANS PRA Standard ............................... 10 Table 4 Major Plant Characteristics for Byron Unit 1 ......................................................... 13 Table 5 SLOCA-Sump Recirculation Boundary Conditions............................................... 19 Table 6 SLOCA-Sump Recirculation Results.................................................................... 20 Table 7 SLOCA-Sump Recirculation Key Event Timings .................................................. 21 Table 8 SLOCA-Sump Recirculation Margins ................................................................... 22 Table 9 SLOCA-Sump Recirculation Cooldown Rates ..................................................... 22 Table 10 SLOCA-Sump Recirculation Sensitivity Studies ................................................... 23 Table 11 SLOCA-Condensate Feed Boundary Conditions ................................................. 30 Table 12 SLOCA-Condensate Feed Results ...................................................................... 30 Table 13 SLOCA-Condensate Feed Key Event Timings .................................................... 31 Table 14 SLOCA-Condensate Feed Margins ..................................................................... 31 Table 15 SLOCA-Condensate Feed Cooldown Rates ........................................................ 32 Table 16 SLOCA-Condensate Feed Sensitivity Studies ..................................................... 33 Table 17 SLOCA-Bleed and Feed Boundary Conditions .................................................... 36 Table 18 SLOCA-Bleed and Feed Results ......................................................................... 37 Table 19 SLOCA-Bleed and Feed Key Event Timings........................................................ 37 Table 20 SLOCA-Bleed and Feed Margins ........................................................................ 38 Table 21 SLOCA-Bleed and Feed Sensitivity Studies ........................................................ 39 Table 22 Loss of DC Bus 111 Boundary Conditions ............................................................ 43 Table 23 Loss of DC Bus 111 Results ................................................................................. 43 Table 24 Loss of DC Bus 111 Key Event Timings ............................................................... 44 Table 25 Loss of DC Bus 111 Margins ................................................................................ 44 Table 26 Loss of DC Bus 111 Sensitivity Studies ................................................................ 46 Table 27 SGTR Boundary Conditions ................................................................................. 52 Table 28 SGTR Results ...................................................................................................... 53 Table 29 SGTR Key Event Timings ..................................................................................... 54 Table 30 SGTR Margins...................................................................................................... 55 Table 31 SGTR Sensitivity Studies ..................................................................................... 56 Table 32 MLOCA Injection Success Criteria Boundary Conditions ...................................... 60 Table 33 MLOCA Injection Success Criteria Results ........................................................... 60 Table 34 MLOCA Injection Success Criteria Key Event Timings ......................................... 61 Table 35 MLOCA Injection Success Criteria Margins .......................................................... 62 Table 36 MLOCA Injection Success Criteria Sensitivity Studies .......................................... 64 Table 37 MLOCA Cooldown Timing Boundary Conditions .................................................. 68 Table 38 MLOCA Cooldown Timing Results ....................................................................... 68 Table 39 MLOCA Cooldown Timing Key Event Timings ...................................................... 70 Table 40 MLOCA Cooldown Timing Margins....................................................................... 71 Table 41 MLOCA Cooldown Timing Cooldown Rates ......................................................... 71 Table 42 MLOCA Cooldown Timing Sensitivity Studies ...................................................... 72 Table 43 Loss of Shutdown Cooling (Mode 4) Boundary Conditions ................................... 78 Table 44 Loss of Shutdown Cooling (Mode 4) Results ........................................................ 78 Table 45 Loss of Shutdown Cooling (Mode 4) Key Event Timings ...................................... 79 xi

Table 46 Loss of Shutdown Cooling (Mode 5) Boundary Conditions ................................... 82 Table 47 Loss of Shutdown Cooling (Mode 5) Results ........................................................ 83 Table 48 Loss of Shutdown Cooling (Mode 5) Key Event Timings ...................................... 84 Table 49 Mapping of MELCOR Analyses to the Byron SPAR (8.27) Model ......................... 88 Table 50 Potential Success Criteria Updates Based on Byron Unit 1 Results ..................... 89 Appendix A Table A-1 Reactor Trip Signals ............................................................................................ A-1 Table A-2 Charging Pump Performance ............................................................................. A-2 Table A-3 SI Pump Performance ........................................................................................ A-2 Table A-4 RHR Pump Performance ..................................................................................... A-3 Table A-5 Reactor Coolant Pump Motive and Control Power Configuration ......................... A-5 Table A-6 Opening and Closing Pressures for Pressurizer PORVs and SRVs..................... A-5 Table A-7 Input Models Used for Documented Calculations ................................................ A-6 xii

ABBREVIATIONS AND ACRONYMS

°C degree(s) Celsius

°C/hr degree(s) Celsius per hour

°F degree(s) Fahrenheit

°F/hr degree(s) Fahrenheit per hour T temperature difference ACC accumulator ADAMS Agencywide Documents Access and Management System AFW auxiliary feedwater ANS American Nuclear Society ASME American Society of Mechanical Engineers ASP accident sequence precursor B&F bleed and feed BAF bottom of active fuel BEP Byron Emergency Procedure BWR boiling-water reactor CCP centrifugal charging pump CCW component cooling water CD core damage CDF core damage frequency CET core exit temperature CFR Code of Federal Regulations cm centimeter(s)

CNMT containment COR MELCOR core package CS containment spray CST condensate storage tank CVH control volume hydrodynamics (MELCOR package)

CVTR Carolinas Virginia Tube Reactor DC direct current DD-AFW diesel-driven auxiliary feedwater ECA emergency contingency action ECCS emergency core cooling system EOP emergency operating procedure EPRI Electric Power Research Institute ESF Engineered Safety Features FCL fan cooler FRP functaionl restoration procedure FSAR Final Safety Analysis Report ft foot/feet ft3 cubic foot/feet FW feedwater gal gallon(s) gpm gallon(s) per minute HEM homogeneous equilibrium model HEP human error probability HFM homogeneous frozen model HPI high-pressure [ECCS] injection xiii

HPR high-pressure [ECCS] recirculation hr hour(s)

HS heat structure in. inch(es) iPWR integral pressurized-water reactor K Kelvin kg kilogram(s) kg/s kilogram(s) per second kPa kilopascal(s) lb/s pound(s) per second LBLOCA large-break loss-of-coolant accident lbm/hr pound(s) mass per hour LOCA loss-of-coolant accident LoDCB-111 loss of DC bus 111 LOFT loss-of-fluid test LPI low pressure [ECCS] injection LPR low pressure [ECCS] recirculation LTOP low temperature overpressure protection m meter(s) m3 cubic meter(s) m3/min cubic meter(s) per minute m3/s cubic meter(s) per second MAAP4 Modular Accident Analysis Program version 4 MD-AFW motor-driven auxiliary feedwater MELCOR Not an acronym MFW main feedwater min minute(s)

MLOCA medium-break loss-of-coolant accident MPa megapascal(s)

MPa abs megapascal(s) absolute MSIV main steam isolation valve MUR measurement uncertainty recapture MW megawatt(s)

MWt megawatt(s) thermal NPSH net positive suction head NR narrow range [water level]

NRC U.S. Nuclear Regulatory Commission PCT peak cladding temperature PORV power- (or pilot-) operated relief valve PRA probabilistic risk assessment PRT pressurizer relief tank PSA Probabilistic Safety Assessment psi pound(s) per square inch psia pound(s) per square inch absolute psid pound(s) per square inch differential psig pound(s) per square inch gage PWR pressurized-water reactor PZR pressurizer RCFC reactor containment fan cooler RCP reactor coolant pump RCS reactor coolant system xiv

recirc recirculation RHR residual heat removal RHR HX residual heat removal heat exchanger RPS reactor protection system RPV reactor pressure vessel RWST refueling water storage tank s second(s)

SC success criterion/criteria SDP significance determination process scfm standard cubic foot/feet per minute SG steam generator SG-x steam generator in loop x SGTR steam generator tube rupture SI safety injection SLOCA small-break loss-of-coolant accident SOARCA State-of-the-Art Reactor Consequence Analyses SPAR standardized plant analysis risk SRV safety relief valve TAF top of active fuel Tavg loop average temperature TBV turbine bypass valve TCL cladding temperature TRACE TRAC/RELAP5 Advanced Computational Engine VCT volume control tank WR wide range [water level]

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APPENDIX D DETAILED STEAM GENERATOR TUBE RUPTURE ANALYSIS RESULTS

D.1 Spontaneous SG Tube Rupture with No Operator Action Case 1: 0.5 Tube, Min ECCS, No Steam Dumps D-1

D-2 D-3 D-4 D-5 D-6 Case 2: 2 Tubes, Max ECCS, No Steam Dumps D-7

D-8 D-9 D-10 D-11 D-12 Case 3: 0.5 Tube, Min ECCS, No Steam Dumps D-13

D-14 D-15 D-16 D-17 D-18 Case 4: 2 Tubes, Max ECCS, No Steam Dumps D-19

D-20 D-21 D-22 D-23 D-24 D.1.4.1 Case 4a: 2 Tubes, Max ECCS, SG PORV Sticks Open D-25

D-26 D-27 D-28 D-29 D-30 Case 5: 0.5 Tube, Min ECCS, Steam Dumps Available D-31

D-32 D-33 D-34 D-35 D-36 Case 6: 2 Tubes, Max ECCS, Steam Dumps Available D-37

D-38 D-39 D-40 D-41 D-42 Case 7: 0.5 Tube, Min ECCS, Automatic Scram, No Steam Dumps D-43

D-44 D-45 D-46 D-47 D-48 Case 8: 0.5 Tube, Max ECCS, Automatic Scram, No Steam Dumps D-49

D-50 D-51 D-52 D-53 D-54 APPENDIX E DETAILED MEDIUM-BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS RESULTS

E.1 Medium-Break Loss-of-Coolant Accident Injection Success Criteria Case 1: 2-in. Break with 1/2 SI, 1/2 RHR, 2/2 CS Pumps E-1

E-2 E-3 E-4 E-5 E-6 E-7 Case 2: 3.33-in. Break with 1/2 SI, 1/2 RHR, 2/2 CS Pumps E-8

E-9 E-10 E-11 E-12 E-13 E-14 Case 3: 4.67-in. Break with 1/2 SI, 1/2 RHR, 2/2 CS Pumps E-15

E-16 E-17 E-18 E-19 E-20 E-21 E.1.3.1 Case 3a: 4.67-in. Break with 1/2 SI, 1/2 RHR, 2/2 CS Pumps, No RHRHX E-22

E-23 E-24 E-25 E-26 E-27 E-28 Case 4: 6-in. Break with 1/2 SI, 1/2 RHR, 2/2 CS Pumps E-29

E-30 E-31 E-32 E-33 E-34 E-35 E.1.4.1 Case 4a: 6-in. Break with 1/2 SI Pumps, 1/2 RHR Pump, 2/2 CS Pumps, CS Recirc E-36

E-37 E-38 E-39 E-40 E-41 E-42 E.1.4.2 Case 4b: 6-in. Break with 1/2 SI Pumps, 1/2 RHR Pump, 0/2 CS Pumps E-43

E-44 E-45 E-46 E-47 E-48 Case 5: 2-in. Break with 2 Accumulators, 1 RHR, 2/2 CS Pumps E-49

E-50 E-51 E-52 E-53 E-54 E-55 Case 6: 3.33-in. Break with 2 Accumulators, 1 RHR, 2/2 CS Pumps E-56

E-57 E-58 E-59 E-60 E-61 E-62 Case 7: 4.67-in. Break with 2 Accumulators, 1 RHR, 2/2 CS Pumps E-63

E-64 E-65 E-66 E-67 E-68 E-69 E.1.7.1 Case 7a: 4.67-in. Break with 2 Accumulators, 1 RHR, 2/2 CS Pumps, No RHRHX E-70

E-71 E-72 E-73 E-74 E-75 E-76 Case 8: 6-in. Break with 2 Accumulators, 1 RHR, 2/2 CS Pumps E-77

E-78 E-79 E-80 E-81 E-82 E-83 E.1.8.1 Case 8a: 6-in. Break with 2 Accumulators (1 in Broken Loop), 1 RHR, 2/2 CS Pumps E-84

E-85 E-86 E-87 E-88 E-89 E-90 E.2 Medium-Break Loss-of-Coolant Accident Cooldown Timing for Low-Pressure Recirculation Case 1: 2-in. Break, 100 °F/hr Cooldown at 20 min, 2/2 CS Pumps, 0/4 RCFC E-91

E-92 E-93 E-94 E-95 E-96 E-97 Case 2: 6-in. Break, 100 °F/hr Cooldown at 20 min, 2/2 CS Pumps, 0/4 RCFC E-98

E-99 E-100 E-101 E-102 E-103 E-104 E.2.2.1 Case 2a: 6-in. Break, 100 °F/hr Cooldown at 20 min, 2/2 CS Pumps, 0/4 RCFC, CS Recirc E-105

E-106 E-107 E-108 E-109 E-110 E-111 Case 3: 2-in. Break, 100 °F/hr Cooldown at 20 min, 0/2 CS Pumps, 0/4 RCFC E-112

E-113 E-114 E-115 E-116 E-117 E-118 Case 4: 6-in. Break, 100 °F/hr Cooldown at 20 min, 0/2 CS Pumps, 0/4 RCFC E-119

E-120 E-121 E-122 E-123 E-124 E-125 Case 5: 2-in. Break, 100 °F/hr Cooldown at 20 min, 0/2 CS Pumps, 4/4 RCFC E-126

E-127 E-128 E-129 E-130 E-131 E-132 Case 6: 6-in. Break, 100 °F/hr Cooldown at 20 min, 0/2 CS Pumps, 4/4 RCFC E-133

E-134 E-135 E-136 E-137 E-138 E-139 Case 7: 2-in. Break, 100 °F/hr Cooldown at 40 min, 2/2 CS Pumps, 0/4 RCFC E-140

E-141 E-142 E-143 E-144 E-145 E-146 Case 8: 6-in. Break, 100 °F/hr Cooldown at 40 min, 2/2 CS Pumps, 0/4 RCFC E-147

E-148 E-149 E-150 E-151 E-152 E-153 E.2.8.1 Case 8a: 6-in. Break, 100 °F/hr Cooldown at 40 min, 2/2 CS Pumps, 0/4 RCFC, CS Recirc E-154

E-155 E-156 E-157 E-158 E-159 E-160 E.2.8.2 Case 8b: 6-in. Break, 100 °F/hr Cooldown at 40 min, 2/2 CS Pumps, 0/4 RCFC, No RHRHX E-161

E-162 E-163 E-164 E-165 E-166 E-167 Case 9: 2-in. Break, 100 °F/hr Cooldown at 40 min, 0/2 CS Pumps, 0/4 RCFC E-168

E-169 E-170 E-171 E-172 E-173 E-174 Case 10: 6-in. Break, 100 °F/hr Cooldown at 40 min, 0/2 CS Pumps, 0/4 RCFC E-175

E-176 E-177 E-178 E-179 E-180 E-181

APPENDIX F DETAILED LOSS OF SHUTDOWN COOLING RESULTS

F.1 Mode 4 Calculations Notes The following list identifies the major changes that were made to the MELCOR input deck in order to perform Mode 4 shutdown calculations.

Logic has been added to model the shutdown cooling function of the residual heat removal (RHR) system. This logic is set up such that RHR flow rate is adjusted in order to maintain a constant coolant temperature, up to the maximum flow rate of the system.

The logic also includes provisions to achieve a target cooldown rate; however, this feature is not used in any of the shutdown calculations performed for this report.

Pressurizer level control logic has been modified to control water level at the no-load setpoint (25 percent level) during the steady-state portion of the calculation.

Similarly, pressurizer heater logic has been modified to achieve the desired pressure during the steady-state portion of the Mode 4 calculations.

Logic that makes it possible to turn off emergency core cooling system (ECCS) flow to prevent overfilling the pressurizer has been modified in order to simulate recovery actions in which operators inject using a charging pump when reactor pressure vessel (RPV) level is low. This feature is exercised in Mode 4 Cases 2 and 5.

The decay heat curves have been shifted in order to simulate the desired times after trip.

For example, the decay heat curve is shifted by 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for Mode 4 Cases 1-5. Note that during the steady-state portion of the calculation, the decay heat is assumed to be constant and to equal the decay power at 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The same is true for all other times since subcriticality that are analyzed in Section 5.8.2 of the report.

Initial temperature and pressure of reactor coolant system (RCS) control volumes have been set to 275 degrees Fahrenheit (F) (408.15 Kelvin (K)) and 350 pounds per square inch absolute (psia) (2.413 megapascals (MPa)).

Secondary-side temperatures (including feedwater temperature) have been set to 275 degrees F (408.15 K).

Logic for the steam dump valves has been modified to maintain secondary-side pressure at 45 psia (0.313 MPa), which is the saturation pressure at 275 degrees F (408.15 K).

Steam generator water level logic has been modified so that steady-state water level is controlled at 18 percent narrow range (NR) or 27 percent wide range (WR) level, depending on the case being analyzed.

Cold volumes have been used in place of hot volumes for RCS control volumes. This decreases the RCS volume by approximately 1 percent.

F-1

Case 1: SG at 18% NR Level, 12 hr after Shutdown, No Recovery Actions F-2

F-3 F-4 F-5 Case 2: SG at 18% NR Level, 12 hr after Shutdown, Start CCP on Low RPV Level F-6

F-7 F-8 F-9 F-10 Case 3: SG at 18% NR Level, 12 hr after Shutdown, Recover RHR at 2 hr F-11

F-12 F-13 F-14 Case 4: SG at 18% NR Level, 12 hr after Shutdown, Initiate AFW at 3 hr F-15

F-16 F-17 F-18 F-19 Case 5: SG at 18% NR Level, 12 hr after Shutdown, Initiate Bleed & Feed at 5 hr F-20

F-21 F-22 F-23 F-24 Case 6: SG at 27% WR Level, 12 hr after Shutdown, Recover RHR at 2 hr F-25

F-26 F-27 F-28 Case 7: SG at 27% WR Level, 12 hr after Shutdown, Initiate AFW at 3 hr F-29

F-30 F-31 F-32 F-33 Case 8: SG at 18% NR Level, 6 hr after Shutdown, No Recovery Actions F-34

F-35 F-36 F-37 Case 9: SG at 18% NR Level, 6 hr after Shutdown, Recover RHR at 2 hr F-38

F-39 F-40 F-41 Case 10: SG at 18% NR Level, 6 hr after Shutdown, Initiate AFW at 3 hr F-42

F-43 F-44 F-45 F-46 F.2 Mode 5 Calculations Notes The following list identifies some of the changes that were made to the MELCOR input deck in order to perform Mode 5 shutdown calculations.

Logic has been added to model the shutdown cooling function of the RHR system. This logic is set up such that RHR flow rate is adjusted in order to maintain a constant coolant temperature, up to the maximum flow rate of the system. The logic also includes provisions to achieve a target cooldown rate; however, this feature is not used in any of the shutdown calculations performed for this report.

Pressurizer level control logic has been modified to control water level during the steady-state portion of the calculation. For the Mode 5 calculations, level control is based on RPV level because the level is assumed to be at the vessel flange, which is below the bottom of the pressurizer.

Pressurizer heaters have been disabled because the pressurizer is empty.

Logic that makes it possible to turn off ECCS flow to prevent overfilling the pressurizer has been modified in order to simulate recovery actions in which operators inject using a charging pump when RPV level is low. This feature is exercised in Mode 5 Cases 2, 5, and 8.

The decay heat curves have been shifted in order to simulate the desired times after trip.

For example, the decay heat curve is shifted by 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months for Mode 5 Cases 1-3. Note that during the steady-state portion of the calculation, the decay heat is assumed to be constant and to equal the decay power at 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months . The same is true for all other times since subcriticality that are analyzed in Section 5.8.3 of the report.

Initial temperature and pressure of RCS control volumes have been set to 170 degrees F (349.8 K) and atmospheric pressure.

Flow paths have been added to model the antisiphon hole in the line leading from the pressurizer power-operated relief valves to the pressurizer relief tank (PRT). It is necessary to include this flow path because, otherwise, the RCS will draw a vacuum when RHR is operating.

The flow path representing the PRT rupture disk is held open throughout the Mode 5 calculations. It is expected that the PRT would be vented to containment during this operating stage; however, the characteristics of this vent path are unknown. In the absence of better information, the PRT rupture disk flow path is used as the vent path for this model.

F-47

Flow paths from CV 310 and 311 to 320 and 321 have been deleted, or the valves in the flow paths have been closed, to simulate loop stop valve closure. The same is true for flow paths between CV 346 and 348 in the cold leg and between analogous control volumes in the other loops.

Cold volumes have been used in place of hot volumes for RCS control volumes. This decreases the RCS volume by approximately 1 percent.

F-48

Case 1: 40 hr after Shutdown, No Recovery Actions F-49

F-50 F-51 Case 2: 40 hr after Shutdown, Start CCP on Low RPV Level F-52

F-53 F-54 Case 3: 40 hr after Shutdown, Recover RHR at 23 Minutes F-55

F-56 F-57 Case 4: 30 hr after Shutdown, No Recovery Actions F-58

F-59 F-60 Case 5: 30 hr after Shutdown, Start CCP on Low RPV Level F-61

F-62 F-63 Case 6: 30 hr after Shutdown, Recover RHR at 23 Minutes F-64

F-65 F-66 Case 7: 60 hr after Shutdown, No Recovery Actions F-67

F-68 F-69 Case 8: 60 hr after Shutdown, Start CCP on Low RPV Level F-70

F-71 F-72 Case 9: 60 hr after Shutdown, Recover RHR at 27 Minutes F-73

F-74 F-75

APPENDIX G EVENT TREE MODELS FOR STUDIED INITIATORS

Byron SPAR Model Event Trees This section provides the relevant event trees from the Byron (v8.27) Standardized Plant Analysis Risk model dated April 2014. These event trees show the sequences described in the main report.

G-1

SMALL LOCA CONDITIONAL LOOP REACTOR TRIP FEEDWATER HIGH PRESSURE FEED AND BLEED SECONDARY RCS COOLDOWN LOW PRESSURE RESIDUAL HEAT LOW PRESSURE HIGH PRESSURE # End State GIVEN A LOCA INJECTION COOLING INJECTION REMOVAL RECIRC RECIRC (Phase - CD)

RECOVERED IE-SLOCA COND-LP-SL RPS FW HPI FTF-SYS-NLOSP FAB FTF-SYS-NLOSP SSCR SSC LPI FTF-SYS-NLOSP RHR LPR FTF-SYS-NLOSP HPR FTF-SYS-NLOSP 1 OK 2 OK 3 CD 4 OK 5 CD 6 OK 7 CD 8 OK 9 CD 10 CD 11 CD 12 OK 13 OK 14 CD 15 OK 16 CD 17 OK 18 CD G-2 19 OK 20 CD 21 CD 22 CD 23 @LOCA-LP Figure G-1 Small-break loss-of-coolant accident (SLOCA) event tree

LOSS OF DC BUS 111 REACTOR TRIP AUXILIARY PORVs ARE CLOSED LOSS OF SEAL HIGH PRESSURE FEED AND BLEED SECONDARY RCS COOLDOWN RESIDUAL HEAT HIGH PRESSURE # End State FEEDWATER COOLING INJECTION COOLING REMOVAL RECIRC (Phase - CD)

RECOVERED IE-LDCA RPS AFW FTF-SYS-NLOSP PORV LOSC HPI FTF-SYS-NLOSP FAB FTF-SYS-NLOSP SSCR SSC RHR HPR FTF-SYS-NLOSP 1 OK 2 LOSC 3 OK 4 OK 5 CD 6 OK 7 CD 8 CD 9 OK 10 OK 11 CD 12 CD 13 ATWS Figure G-2 Loss of 125V vital DC bus 111 event tree G-3

SG TUBE RUPTURE REACTOR TRIP FEEDWATER HIGH PRESSURE FAULTED STEAM RCS COOLDOWN TERMINATE OR FEED AND BLEED RWST REFILL HIGH PRESSURE RESIDUAL HEAT DECAY HEAT # End State INJECTION GENERATOR CONTROL SAFETY RECIRC REMOVAL REMOVAL/ (Phase - CD)

ISOLATION INJECTION RECOVERY (ECA-IE-SGTR RPS FW HPI FTF-SYS-NLOSP SGI SSC CSI FAB FTF-SYS-NLOSP RFL HPR FTF-SYS-NLOSP RHR ECA 3.1/3.2) 1 OK 2 OK CST-REFILL 3 CD 4 OK 5 OK 6 CD 7 OK 8 OK RFL1 9 CD 10 OK 11 OK RFL1 12 CD 13 OK RHR-LPI 14 CD SSC1 15 CD 16 CD 17 OK 18 CD G-4 19 CD 20 CD 21 CD 22 CD Figure G-3 Steam generator tube rupture (SGTR) event tree

MEDIUM LOCA CONDITIONAL LOOP REACTOR TRIP HIGH PRESSURE ACCUMULATORS AUXILIARY RCS COOLDOWN LOW PRESSURE HIGH PRESSURE LOW PRESSURE # End State GIVEN A LOCA INJECTION FEEDWATER INJECTION RECIRC RECIRC (Phase - CD)

IE-MLOCA COND-LP-SL RPS HPI FTF-SYS-NLOSP ACC AFW FTF-SYS-NLOSP SSC LPI FTF-SYS-NLOSP HPR FTF-SYS-NLOSP LPR FTF-SYS-NLOSP 1 OK 2 CD 3 OK 4 CD 5 OK 6 CD 7 OK 8 CD 9 CD 10 CD 11 CD 12 CD 13 CD 14 @LOCA-LP G-5 Figure G-4 Medium-break loss-of-coolant accident (MLOCA) event tree