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| number = ML16022A062
| number = ML16022A062
| issue date = 01/31/2016
| issue date = 01/31/2016
| title = 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.
| title = 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
| author name = Bone A, Buell R, Corson J, Helton D, Khatib-Rahbar M, Kozak L, Krall A, Tobin M
| author name = Bone A, Buell R, Corson J, Helton D, Khatib-Rahbar M, Kozak L, Krall A, Tobin M
| author affiliation = Energy Research, Inc, Idaho National Lab, NRC/RES, NRC/RGN-III
| author affiliation = Energy Research, Inc, Idaho National Lab, NRC/RES, NRC/RGN-III
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{{#Wiki_filter: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
{{#Wiki_filter:}}
 
AVAILABILITY OF REFERENCE MATERIALS IN NRC PUBLICATIONS NRC Reference Material                                      Non-NRC Reference Material As of November 1999, you may electronically access          Documents available from public and special technical NUREG-series publications and other NRC records at          libraries include all open literature items, such as books, NRCs Library at www.nrc.gov/reading-rm.html. Publicly      journal articles, transactions, Federal Register notices, released records include, to name a few, NUREG-series        Federal and State legislation, and congressional reports.
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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.
v
 
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]
xv
 
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 hours 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 hours. 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 hours 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 hours. 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
 
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]
xv
 
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 hours 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 hours. 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 hours 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 hours. 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
 
NUREG-2187, Vol. 2 Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria January 2016 in the Standardized Plant Analysis Risk ModelsByron Unit 1}}

Latest revision as of 05:27, 10 January 2025

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
ML16022A062
Person / Time
Site: Byron Constellation icon.png
Issue date: 01/31/2016
From: Alysia Bone, Buell R, James Corson, Don Helton, Khatib-Rahbar M, Laura Kozak, Alfred Krall, Margaret Tobin
Energy Research, Idaho National Lab, Office of Nuclear Regulatory Research, NRC/RGN-III
To:
Blount B
References
NUREG-2187 V02
Download: ML16022A062 (349)
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